Water-cooled Solid Breeder Research Articles

Development of water-cooled cylindrical blanket in JA DEMO

The concept of the tritium breeding blanket for Japan’s DEMOnstration fusion reactor (JA DEMO) has been developed with pressure tightness against in-box loss-of-coolant accidents based on a water-cooled solid breeder concept. The cooling conditions are designed on the pressurized-water reactor water conditions which are the coolant temperature of 290 °C–325 °C and the operating pressure of 15.5 MPa, respectively. The point of the blanket design is to reduce the amount of structural material in casing as well as to ensure its pressure tightness. This is because a decrease in the amount of structural material improves tritium breeding ratio (TBR). A cylindrical structure, a thin wall casing structure which ensures pressure tightness and could increase TBR, is feasible. However, a relatively larger useless space is expected between modules when the cylindrical blanket modules are arranged in a vacuum vessel, which could decrease TBR. Therefore, the cylindrical blanket modules are to be in a close-packed arrangement to reduce useless space. The Be12Ti block (which shows a minor swelling compared to Be) is selected to achieve the target TBR as the net Be density is equivalent to the case of the Be pebble. The use of Be12Ti blocks can reduce or remove the cooling piping inside the module as the Be12Ti block has a higher thermal conductivity than pebbles. As a result of the neutronics, finite element method, and computational fluid dynamics analyses, it was found that the target TBR value can be achieved in cylindrical structure blanket that ensure pressure tightness.

Investigation and Analysis on Tritium Accumulation Phenomenon Based on a Water-Cooled Blanket for CFETR at 1.5 GW

China Fusion Engineering Test Reactor (CFETR) is a major scientific project independently designed and developed by China, which will demonstrate the engineering feasibility of continuous large-scale fusion energy for safe and stable power generation. One of the most critical issues for fusion reactors is to achieve tritium self-sufficiency with a tritium breeding ratio (TBR) of no less than 1.0, as it relates to the ability of the fusion reactor to fulfill the fuel cycle and continue stable operation. Moreover, the release and inventory of tritium is also a significant factor affecting the operation of CFETR. The temperature distribution inside the breeder blanket is precisely related to the tritium release. Therefore, it is particularly important for CFETR to optimize the temperature field inside the breeding blanket. It is significant that to ensure that the temperature of the mixed pebble bed zone reaches the limit of 900 °C as much as possible, which could maximize the tritium release efficiency. Simultaneously, it must make sure that other structural materials meet the allowable value of temperature. Tritium accumulation close to the cooling pipe is investigation, which will be adverse to tritium release. Hence, in order to solve this difficulty, the Be12 Ti tube was sleeved on the cooling pipe to increase the temperature of the mixed pebble bed zone near the cooling pipe. In this work, a water-cooled solid breeder blanket was proposed and the nuclear heat was calculated based on the 1.5 GW fusion power. The nuclear heat was used as the preprocessing data for thermal calculation carried out by the Monte Carlo N-Particle Transport Code (MCNP). With the given conditions, the thermal hydraulic was simulated with the commercial CFD code, the ANSYS CFX. In this way, a breeding blanket structure with a uniform temperature field of mixed bed zone close to 900 °C can be obtained. In addition, the selection of Be12 Ti tubes with the thickness of 0.525 cm can ensure a high TBR and increase the temperature around the cooling tubes to improve the tritium release efficiency. This work provides an important reference for the subsequent engineering design of CFETR breeding blanket.

Reaction kinetics of lithium extraction and a recycling process for lithium meta-titanate pebbles

ABSTRACT Fusion reactors require large amounts of lithium meta-titanate (Li2TiO3) pebbles as tritium breeding materials for Japanese water-cooled solid breeder blanket. The development of a recycling process for tritium-breeding materials is essential for the early realization of fusion reactors. This study was proposed to understand the kinetics of Li extraction from Li2TiO3 by inducing chemical reactions and applying microwave heating and establish a highly efficient recycling process for spent Li2TiO3 pebbles. The results of Li extraction experiments performed under various conditions clarified that this extraction reaction could be applied via a surface chemical reaction with kct = 1 − (1 − f)1/3. The apparent activation energy associated with this extraction reaction was evaluated to be 52.78 kJ/mol, which is in agreement with that of the chemical reaction-controlled mechanism. Additionally, the reaction order for the dissolution of Li2TiO3 in the HCl solution was first-order. The highly efficient recycling process of Li2TiO3 pebbles was suggested.

Neutronic and Thermo-Hydraulic Analyses of Water-Cooled Blanket Based on Pressurized/Supercritical Water Conditions for CFETR

China Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between International Thermonuclear Experimental Reactor (ITER) and future fusion power plant. As one of the candidates, a water-cooled solid breeder blanket based on pressurized water and supercritical water conditions were proposed. In the concept, multiplying layers separated by three breeding layers were designed and optimized for higher tritium breeding ratio (TBR) and uniform heat distribution. This blanket adopts the Li2TiO3 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized blanket on both water conditions were performed by numerical simulation, to discuss thermo-hydraulic performance of the blanket using pressurized water/supercritical water as its coolant. At first, the neutronic analysis was performed and based on the typical outboard equatorial blanket. Then, thermal and fluid dynamic analysis of the 3D model was carried out by CFX with fluid–solid coupling approach. It was found that the blanket can be effectively cooled on both water conditions, certified the feasibility of the blanket design with pressurized/supercritical water conditions. It indicated that supercritical water blanket had smaller safety margin than pressurized water blanket, but supercritical water blanket would lead to higher outlet temperature, thermal conductivity, and heat exchange efficiency also. In addition, the beryllium fraction was considered as one of the dominant factor, which leading a higher TBR in our simulations.

Development of water-cooled blanket concept with pressure tightness against in-box LOCA for JA DEMO

A conceptual design of breeding blanket module with pressure tightness against in-box LOCA has been carried out, based on a pressurized water-cooled solid breeder blanket. The cooling water for DEMO is operated at the PWR water conditions of 15.5 MPa and 290 ºC-325 ºC. In this design, the breeding area of the module is divided into 0.1-m-squared cells with rib structure and has simple interior for mass production using a mixed pebbles bed of Li2TiO3pebbles and Be12Ti ones. As a result, a rib with the thickness of 0.015m is needed to withstand the design pressure of 17.2 MPa by a stress analysis. The cooling system for the blanket module is designed by fluid dynamics analysis based on the PWR water conditions, and the outlet coolant temperature and the pressure drop are 321 ºC and 0.32 MPa, respectively. It was found that the self-sufficient production of tritium is likely to be satisfied with the blanket radial width thickness of 0.70 m or more and with an idea to improve TBR that the coolant is changed from existing light water to heavy water.

Reactivity and thermal stability of ternary Be-Zr-V beryllides

A water-cooled solid breeder blanket of a fusion reactor has been adopted in Japan, which focuses on safety concerns; safety is one of the most critical issues. In particular, as a multiplier, Be pebbles are well known to possibly generate hydrogen and react in an exothermic manner due to reaction with H2O during the loss of coolant accident. In contrast to these Be pebbles, beryllium intermetallic compounds (beryllides) demonstrate promise due to their more stable chemical reactivity at high temperatures. Currently, several studies have reported the development of advanced neutron multipliers in Japan, and the EU is part of the demonstration R&D activities at the International Fusion Energy Research Center project, which forms a part of the Broader Approach program. Fabrication methods for beryllide pebbles have been successfully developed by the combination of plasma sintering synthesis and a rotating electrode granulation method.By employing these methods, ternary beryllide pebbles with mixtures of Be13Zr and Be12V were first synthesized with ratios of 1.0:1.0, 1.0:0.8, 1.0:0.6, 1.0:0.4, 1.0:0.2, and 1.0:0.1. The small-sized Be13Zr phase as the precipitate and the Be12V phase were formed without the formation of the Be phase as these compositions did not exhibit the peritectic reaction. In addition, ternary beryllide pebbles exhibited a lower reactivity to water vapor as well as a higher thermal stability than Be pebble. In this study, not only the synthesis but also characterizations at high temperatures were discussed.

Tritium transport analysis for CFETR WCSB blanket

Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

Numerical analysis of heat transfer in the first wall of CFETR WCSB blanket

China Fusion Engineering Test Reactor (CFETR), the first fusion reactor experiment project planned in China, is now being investigated in detail. Recently, a conceptual structural design of the Water-Cooled-Solid-Breeder (WCSB) blanket was proposed as one of the breeding blanket candidates for CFETR. In this research, based on the present design of the CFETR WCSB blanket, the heat transfer performance in the first wall (FW) under the pressurized water cooling condition was analyzed. The 3D computational fluid dynamics (CFD) results show that the maximal temperature of the FW will not exceed the limited temperature under normal or even higher heat flux condition. In addition, the effect of buoyancy on heat transfer is negligible under both conditions. The influence of roughness becomes increasingly important when the roughness height lies in the fully turbulent regime. The maximal temperature increases approximately linearly as the thickness of the front wall increases. It is also found that the heat flux and the local heat transfer coefficient are extremely non-uniform in the circumferential direction. Two 1D CFD methods are also evaluated by 3D CFD data, with the conclusion that both 1D results have some differences with the 3D data. The improved 1D method is more accurate than the former one. However, we ascertain that 1D methods should be used with caution for the water-cooling FW design.

Design study of blanket structure based on a water-cooled solid breeder for DEMO

Blanket concept with a simplified interior for mass production has been developed using a mixed bed of Li2TiO3 and Be12Ti pebbles, coolant conditions of 15.5MPa and 290–325°C and cooling pipes without any partitions. Considering the continuity with the ITER test blanket module option of Japan and the engineering feasibility in its fabrication, our design study focused on a water-cooled solid breeding blanket using the mixed pebbles bed. Herein, we propose blanket segmentation corresponding to the shape and dimension of the blanket and routing of the coolant flow. Moreover, we estimate the overall tritium breeding ratio (TBR) with a torus configuration, based on the segmentation using three-dimensional (3D) Monte Carlo N-particle calculations. As a result, the overall TBR is 1.15. Our 3D neutronics analysis for TBR ensures that the blanket concept can achieve a self-sufficient supply of tritium.

Preliminary Neutronics and Activation Analysis of the Water Cooled Solid Breeder Blanket for CFETR

China Fusion Engineering Test Reactor is an “ITER-like” superconducting tokomak reactor. It’s necessary to assess the neutronics parameters, such as tritium breeding ratio, neutron wall loading and nuclear heating of the Water Cooled Solid Breeder (WCSB) blanket for its initial design. In this paper, the neutronics calculations were performed by using MCNP/4C code with the Fusion Evaluated Nuclear Data Library FENDL-2.0. The results show that the neutronics parameters can meet the engineering design requirement. Different source models affect the magnitude and distribution of the neutron wall loading and the nuclear heating were also discussed. The activation calculations were carried out by using FISPACT-2007 with the EAF-2007 library. Activity, afterheat and dose rate of the blanket form the beginning of shutdown to 106 years were presented. The results show that at shutdown the total activity of the equatorial outboard blanket module is 7.27 × 1016 Bq, and the total afterheat is 11.52 kW. These results will provide useful indications for further optimization of the WCSB blanket.

Flow Distribution Analysis of Water-Cooled Solid Breeder Test Blanket Module

Blanket system is one of the most important systems in a fusion reactor, which plays an important role in heat removing, radiation shielding and tritium breeding. Water-cooled ceramic breeder (WCCB) blanket module (BM) is one of tritium breeding blanket module concepts for Chinese Fusion Engineering Testing Reactor. According to the preliminary design of WCCB BM, there are complicated cascade and parallel cooling channels, in which maybe exists the nonuniform distribution of flow rate, and resulting adverse effect on heat transfer and safety. In this paper, the whole model of one BM is built by thermal hydraulic analytical code named Relap 5 and the flow distribution issue of water-cooled solid breeder (WCSB) test blanket module (TBM) is analyzed. The systematic analysis results show that the flow rate differences of most parts of the WCSB TBM are less than 4 % in a steady state. Start-up, operational transient and loss of flow accident are also studied, and flow instability in these transient cases is found and needs for further analysis. Three dimensional local model of First Wall is also built by CFX, to investigate flow characteristics at partial WCSB TBM, which shows that flow distribution calculated by CFX is consistent with the results from the thermal hydraulic analytical code. Both of the results of the steady state and transient analysis show that the thermal hydraulic analytical code is appropriate in analyzing the flow distribution and transient issue from the systematic view.

Reactivity of plasma-sintered beryllium–titanium intermetallic compounds with water vapor

Beryllium–titanium intermetallic compounds (beryllides) are candidates for advanced neutron multipliers in water-cooled solid breeder demonstration (DEMO) reactor. In contrast to beryllium, which is highly reactive with water vapor above 873K, the beryllide is considerably more stable with much lower hydrogen-gas-generation rates. To investigate the effect of phase compositions on the reactivity with water vapor, the beryllides were plasma-sintered for different times to produce different phase compositions. They were examined by thermogravimetric analysis and the H2-generation rate was measured at 1273 and 1473K for 24h under Ar gas containing 1% H2O. The resulting weight gain of the beryllides under 1% H2O was in good agreement with previous results showing that with an increase in the area fraction of Be, the oxidation resistance under Ar gas containing 1% H2O decreases. The result of the H2-generation rate at 1273K demonstrates that the larger the fraction of Be, the higher the H2-generation rate. It is clear that the Be phase in the plasma-sintered beryllides attributes to an increase in H2-gas generation. We report here on the effect of phase compositions on reactivity with water vapor of the plasma-sintered beryllides.

Packing experiment of breeder pebbles into water cooled solid breeder test blanket module for ITER

This paper describes an experiment in which Li 2TiO 3 tritium breeder pebbles were packed into a full-scale breeder pebble container (BPC) mockup. The BPC mockup, part of the Japanese water cooled solid breeder (WCSB) – Test Blanket Module (TBM) project, has been successfully developed using a reduced activation ferritic/martensitic steel, F82H. The BPC mockup was fabricated using a fiber laser welding technique, selected for its high focusing capability. Membrane panels, tube plates and side walls were joined by fiber laser welding to form the BPC. Its dimensions are 74 mm × 112 mm × 990 mm. It was confirmed to be gastight under pressurized helium up to 0.5 MPa. Packing tests were performed with real breeder material Li 2TiO 3 pebbles of 1 mm diameter. X-ray tomography revealed that dense packing was uniformly achieved throughout the BPC mockup.

Progress of Design and R&D of Water Cooled Solid Breeder Test Blanket Module

Development of Water Cooled Solid Breeder (WCSB) TBM, the primary candidate of ITER Test Blanket Module (TBM), has been performed in Japan, according to the TBM milestones, which are necessary for acceptance of the TBM in ITER for testing from the first day of plasma operation. The TBM milestones consist of milestones on safety assessment, module qualification and design integration in ITER. For the safety milestones, essential source terms were evaluated, and failure modes and effect analysis (FMEA) was performed. Based on the results of FMEA, safety assessment was performed. For the milestones of the design integration, detailed structural design of the TBM and the interface structure with the ITER test port were performed. Based on the design, performance analysis such as thermo-mechanical analysis in over-pressurization was performed. For the milestones of the qualification of fabrication technology, essential fabrication technology was developed and near full size first wall of the TBM was successfully fabricated and demonstration of the integrity in heat flux equivalent to ITER. The development of the WCSB TBM is showing steady progress toward the installation in ITER.

Compact DEMO, SlimCS: design progress and issues

The design progress in a compact low aspect ratio (low A) DEMO reactor, ‘SlimCS’, and its design issues are reported. The design study focused mainly on the torus configuration including the blanket, divertor, materials and maintenance scheme. For continuity with the Japanese ITER-TBM, the blanket is based on a water-cooled solid breeder blanket. For vertical stability of the elongated plasma and high beta access, the blanket is segmented into replaceable and permanent blankets and a sector-wide conducting shell is arranged inbetween these blankets. A numerical calculation indicates that fuel self-sufficiency can be satisfied when the blanket interior is ideally fabricated. An allowable heat load to the divertor plate should be 8 MW m−2 or lower, which can be a critical constraint for determining a handling power of DEMO.

Achievements in the development of the Water Cooled Solid Breeder Test Blanket Module of Japan to the milestones for installation in ITER

As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, a water cooled solid breeder (WCSB) TBM is being developed. This paper shows the recent achievements towards the milestones of ITER TBMs prior to the installation, which consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, targeting the detailed design final report in 2012, structure designs of the WCSB TBM and the interfacing components (common frame and backside shielding) that are placed in a test port of ITER and the layout of the cooling system are presented. As for the module qualification, a real-scale first wall mock-up fabricated by using the hot isostatic pressing method by structural material of reduced activation martensitic ferritic steel, F82H, and flow and irradiation test of the mock-up are presented. As for safety milestones, the contents of the preliminary safety report in 2008 consisting of source term identification, failure mode and effect analysis (FMEA) and identification of postulated initiating events (PIEs) and safety analyses are presented.

R&D of a Li2TiO3 pebble bed for a test blanket module in JAEA

At JAEA, a test blanket module (TBM) with a water-cooled solid breeder is being developed. This paper presents recent achievements of research activities for the TBM, particularly addressing the pebble bed of the tritium breeder materials and tritium behaviour. For the breeder material, the chemical stability of Li2TiO3 was improved using Li2O additives. To analyse the pebble bed behaviour, thermomechanical properties of the Li2TiO3 pebble bed were assessed experimentally. To verify the pebble bed's nuclear properties, the activation foil method was proposed and a preliminary experiment was conducted. To reduce the tritium permeation, the chemical densified coating method was developed and the coating was attached to F82H steel. For tritium behaviour, the tritium recovery system was modified in consideration of the design change of the TBM.

Development of water-cooled solid breeder test blanket module in JAEA

One of the most important missions of ITER is to provide a test bed for breeding blanket modules, which are called as test blanket module (TBM). JAEA has been extensively developing a water-cooled solid breeder test blanket module (WCSB TBM) for ITER. JAEA developed fabrication technology of F82H rectangular cooling tubes, and has successfully fabricated the near-full scale first wall mock-up of WCSB TBM by hot isostatic press (HIP) technique, which is fully made of F82H. The mock-up has been high-heat flux tested in the DATS facility in JAEA, which is an ion beam test facility. The inlet temperature of the cooling water is about 280 °C with 15 MPa, which is almost the same as the WCSB TBM design conditions. The mock-up has endured a heat load of 0.5 MW/m 2, 30 s for 80 thermal cycles. Neither hot spots nor thermal degradation have been observed.

Torus configuration and materials selection on a fusion DEMO reactor, SlimCS

SlimCS is the conceptual design of a compact fusion DEMO plant assuming technologies foreseeable in 2020–2030s. For continuity of blanket technology from the Japanese ITER-TBM, the prime option of blanket is water-cooled solid breeder with Li 2TiO 3 (or Li 4SiO 4) and Be. A reduced-activation ferritic–martensitic steel (RAFM) and subcritical water are chosen as the structural material and coolant, respectively. The reactor has somewhat complex torus configuration with a sector-wide conducting plate slipped in between the replaceable (front) and permanent (back) blanket. In order to allow flexibility of maintenance in such a complex configuration, sector transport hot cell maintenance scheme is adopted. This paper describes characteristics of SlimCS with a focus on materials selection.

Heatup event analyses of the Water Cooled Solid Breeder Test Blanket Module

This paper describes the safety evaluation on the heatup events of the Water Cooled Solid Breeder (WCSB) Test Blanket Module (TBM) and proposes the basic method to ensure safety, incorporating the chemical reaction between beryllium and water. One-dimensional analyses are carried out for the three representative event sequences, which were derived by the Failure Modes and Effect Analysis. With respect to the event sequence of loss of cooling of the TBM during plasma operation, once temperature of the TBM totally increases, then a plasma disruption takes place when the temperature of the first wall armor becomes too high. In this event, the requirement to the TBM design becomes apparent that no cooling pipe rupture can be guaranteed under the predicted highest temperature condition to prevent the activation of the chemical reaction under the high temperature condition. With respect to the event sequence of ingress of coolant into the TBM during plasma operation, heat load due to beryllium–water chemical reaction is added to the normal heat load of the TBM. The requirement for the cooling system of the TBM becomes apparent that the cooling system of the TBM should be designed to continue its operation against partial loss of coolant due to the ingress of coolant. With respect to the loss of off-site power after the ingress of coolant, it is confirmed that the event converges.