Water Cooled Ceramic Breeder Blanket Research Articles

Blanket test facility for mockup testing on water cooled ceramic breeder blanket system

Intensive research and development activities have been conducted for designing a water-cooled ceramic breeder (WCCB) blanket system. In this system 325 °C/15.5 MPa of pressurized water removes heat injected and accumulated in blanket containers made of a reduced activation ferritic/martensitic steel, F82H. The container has lithium ceramics and beryllium pebbles as tritium breeder and neutron multiplier, respectively. In order to demonstrate performance of the system, especially for high-temperature pressurized water coolant, experimental apparatus have been installed in Rokkasho Fusion Institute, National Institutes for Quantum Science and Technology, Japan. These apparatuses enable to conduct the following experiment on full-scale mockups: 1) high heat flux (HHF) test to demonstrate heat removal of surface heat flux by electron beam gun simulating plasma, 2) flow assisted corrosion, 3) coolant ingress into dummy pebble beds, and 4) beryllium-water reaction between the coolant and the pebbles. In this paper, overviews, specifications, early results of these apparatus are reported.

Overview of progress on water cooled ceramic breeder blanket in Japan

National institutes for Quantum Science and Technology (QST) are the implementing body of ITER project in Japan and takes the lead in fusion DEMO reactor development in Japan. The primary candidate of the breeder blanket for JA DEMO is Water Cooled Ceramic Breeder (WCCB) blanket. ITER-Test Blanket Module (TBM) program is an opportunity for the demonstration of DEMO breeding blanket concept under actual fusion environment. QST has led the development of the WCCB TBM Concept in the program since the signing of the TBM Arrangement in 2014. The preliminary design of test blanket system (TBS) is on-going. Recently, the status of the preliminary design of TBS was assessed between QST and the ITER Organization (IO). So, the status of the latest TBS design and the DEMO relevancy is presented.

Electromagnetic load evaluating and optimizing approach of the blanket system during VDEs considering halo current

In most scenarios, the vertical displacement event (VDE) represents the most extreme electromagnetic (EM) event within the tokamak device. The significant EM loads experienced during this event have the potential to compromise the structural stability of in-vessel components. This study investigates the EM loads on the water-cooled ceramic breeder blanket system of China Fusion Engineering Test Reactor (CFETR) using finite element analysis methods in two characteristic events: hot-VDE and cold-VDE. The study discusses the EM load effects resulting from changes in magnetic flux and induced electromotive force, respectively, with a specific focus on halo currents. The results reveal that, with similar current quech time, the difference in EM load on the blanket system during the VDEs primarily depends on the halo currents. When the electrical connection of the back supporting structure (BSS) is open, the halo current path within the blanket system and vacuum vessel (VV) changes, and a substantial portion of the halo current in the blanket system is conducted to the VV via the BSS. Consequently, a portion of the EM load on the blankets and BSS is transferred to the VV due to the transfer of halo current. Inspired by this, the conceptual use of ‘shunts’ is proposed to provide a dedicated circuit for shunting halo currents away from critical device components, such as the VV and blankets. This approach allows for the sharing of EM loads caused by halo currents and reduces the threat posed by halo currents to the structural integrity of these essential components.

Parametric study of density wave instability in parallel channels of a water-cooled blanket in a fusion reactor

In fusion reactors, many blanket concepts are designed with water as a coolant to transfer high-density heat from the fusion reaction out of the reactor core. The coolant temperature and pressure are maintained as the validated use in water-cooled fission reactors. However, the flow channel in a water-cooled blanket is independent of each other, and there is no flow mixing between coolant channels. Therefore, flow instability may occur in the independent parallel channels in a water-cooled blanket due to its unique structure and heat distribution, especially under the high heat flux caused by plasma rupture. In this study, the parametric analysis of density wave instability is performed using a thermal-hydraulic code developed for independent parallel channels based on the homogeneous model for the two-phase flow. The parallel-channel system in a water-cooled ceramic breeder (WCCB) blanket of the China Fusion Engineering Experimental Reactor (CFETR) is established for its first wall structure. A small disturbance is introduced into the system to determine if it is stable under different conditions. It is found that the channel number has no obvious influence on the prediction of the flow instability boundary. Therefore, the two-channel system is adopted to investigate the influence of different parameters, such as the pressure, resistance, flow rate, and inclination, on the flow instability boundary of the parallel-channel system in the CFETR WCCB blanket. The results show that flow instability occurs more easily in this study compared to the traditional instability analysis, especially under high-pressure conditions. In general, conditions of high pressure, large flow rate, and no inclination can stabilize the system, while the influence of resistance is quite different under different conditions of resistance and pressure. The research work indicates that more attention should be paid to the joint influence of different parameters for the water-cooled blanket during its design and operation.

Preliminary neutronics/thermal-hydraulics design and analyses of the helium-cooled mixed bed breeder blanket for CFETR under 1.5 GW fusion power

Chinese Fusion Engineering Test Reactor (CFETR) is a tokamak device designed with different power levels to validate fusion engineering technologies in China. Helium-cooled ceramic breeder (HCCB) blanket and Water-cooled ceramic breeder (WCCB) blanket are the main candidates. Compared with WCCB, HCCB generally has a higher tritium breeding ratio (TBR) because helium has smaller neutron absorption cross sections than water. However, the TBR of the HCCB decreases with the increase of the fusion power. In this paper, the preliminary neutronics and thermal-hydraulics design of the Helium-Cooled Mixed bed Breeder (HCMB) blanket is proposed. The blanket structure with a fusion power of 1.5 GW is designed based on the current HCCB design of the CFETR. The HCMB adopts helium as the coolant and the mixed Li4SiO4/Be12Ti as the tritium breeder. The neutronics and thermal-hydraulics performances of HCMB blanket have been analyzed in the paper. The results show that the HCMB blanket design has a TBR up to 1.25 with the thermal-hydraulics requirements satisfied.

A new method to reduce the toroidal field ripple effect due to the ferromagnetic materials in updated CFETR

The toroidal field ripple (TFR) is one of the critical factors affecting the long-pulse high-performance discharges for the tokamak device. The TFR is not only determined by the toroidal field coils (TFCs) but is also affected by the ferromagnetic materials in the plasma facing components such as diagnostic system and blanket system. In the phase Ⅱ design of the China Fusion Engineering Test Reactor (CFETR), the inner ferromagnetic constituent and the poloidal segmentation of water-cooled ceramic breeder (WCCB) blanket are modified and the diagnostic system also has major change compared with the phase Ⅰ design, therefore it is necessary to investigate and reduce the TFR effect of the updated blanket system and diagnostic system for the phase Ⅱ designed CFETR. Firstly, the effect of the blanket system on the TFR is evaluated systematically, considering the diagnostic shielding modules (DSMs) of the diagnostic system at the equatorial ports (EPs). Moreover, a new TFR reducing method by adjusting the distribution of ferromagnetic materials filling ratios in the toroidal direction is proposed. According to the results, the local TFR nearby the plasma separatrix line already exceeds the design specification limit with the blanket system, and this phenomenon will be worse due to installing the DSMs and the EP. After adjusting the percentage of the ferromagnetic materials in the outboard-side (OBs) segment, the plasma TFR is reduced obviously and within the 0.5% limit.

Microstructure and Mechanical Characteristics of the First Wall Welded Joints

Vacuum electron beam welding (EBW) is selected as the welding of the water-cooled ceramic breeder (WCCB) blanket. In the process of welding the first wall (FW) of WCCB blanket, it is necessary to add a backing plate to the arc section. At this stage, the thickness of reduced activation ferrite/martensite (RAFM) steels is up to 31 mm. In this article, thick plate RAFM steels after hot isostatic pressing (HIP) are welded by EBW, and then different heat treatments are performed. The results showed that, in the welding condition, hardness of welded joints was changed slightly. The highest hardness of the joint is 440 HV while the lowest is 290 HV, and a softening zone was observed in the heat-affected zone (HAZ). Moreover, after different heat treatments, the welding seam changed into the tempered martensite. In addition, the hardness of base metal and welded joints is all reduced, and the performance of welded joints in quenched and tempered conditions is better than that in normalized and tempered conditions.

What Would Happen to a Fusion Power Plant in the Case of Plasma Disruption? A Dynamic Model Simulation

It is a widespread view in the fusion community that steady-state, water-cooled fusion power plants can utilize the power generation systems of conventional pressurized water reactor (PWR) fission plants as is. However, what would happen to a fusion power plant in the case of plasma disruption? The authors constructed a dynamic simulation model of a water-cooled ceramic breeder blanket fusion power plant model on Modelica language [300.0-MW(electric) electrical output/1138-MW(thermal) fusion output] and evaluated the applicability of a PWR power generation system. Simulation results suggest that while the PWR system would function as intended during steady-state operation, the conventional system may not be able to cope with a sudden loss of energy influx in the event of plasma disruption without modification: The PWR system’s steam generator experienced a water overflow in less than 150 s from the plasma disruption.

Optimization of a WCCB blanket module with detached first wall for CFETR

State-of-the-art blanket designs in fusion reactor engineering tend to use a first wall to protect the plasma facing components from high surface heat flux. This first wall is prone to get damaged, and thus should better be easily replaceable for convenient maintenance. The wall should also have a tungsten cladding of reasonable thickness to balance the blanket's tritium breeding and plasma shielding capabilities. In this regard, optimization of a CFETR water-cooled ceramic breeder (WCCB) blanket module is proposed. Compared to the present design, the optimized module features a detached first wall bolted to the back part of the blanket, with thickened tungsten armor covered on the surface. The new design not only allows convenient replacement and maintenance by remote handling, but is expected to provide improved protection against plasma surface loads. This paper presents the detailed design of the proposed blanket module, as well as feasibility analysis results from thermo-mechanical and thermal-hydraulics aspects.

Expression of Concern: Effect of Heat Treatment on the Properties of Electron Beam Welded Joint of First Wall

The structural material of the water-cooled ceramic breeder (WCCB) blanket is reduced activated ferrite/martensite (RAFM) steels. The first wall (FW) of one of its components is realized by hot isostatic pressure (HIP) diffusion welding and vacuum electron beam welding (EBW). In order to study the performance difference of electron beam welded joint of RAFM steels before and after heat treatment, the EBW test of RAFM steels was carried out, and the welded joint was quenched and tempered treatment. The properties of welded joints before and after quenched and tempered treatment were studied. The results show that the microstructure of the weld is composed of <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\delta $ </tex-math></inline-formula> ferrite and coarse lath martensite. After quenching and tempering treatment, the microstructure of the joint changes into tempered martensite with no obvious martensite characteristics. Moreover, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\delta $ </tex-math></inline-formula> ferrite disappears, and carbide precipitates out at the original austenite grain boundaries. Here, the type of precipitated carbide is M <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">23</sub> C <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">6</sub> as characterized by the electron backscatter diffraction analysis.

Progress on neutronic analysis for CFETR

The neutron-induced irradiation field is a key problem in a fusion reactor related to nuclear responses, shielding design, nuclear safety and thermo-hydraulic analysis. To support the system design of the China Fusion Engineering Test Reactor (CFETR), a comprehensive analysis of the irradiation field has been conducted in support of many newly developed advanced tools. This paper first summarizes recent progress on the related effort to develop neutronics code including the geometry conversion tool cosVMPT, Monte Carlo variance reduction technology and ‘on-the-fly’ global variance reduction. Such developed tools have been fully validated and applied to the CFETR nuclear analysis. The neutron irradiation has been evaluated on the CFETR water-cooled ceramic breeder blanket, divertor, vacuum vessel (VV), superconductive coils and four kinds of heating systems: electron cyclotron resonance heating, ion cyclotron resonance heating, low hybrid wave and neutral beam injection. The nuclear responses of tritium breeding ratio, heating, irradiation damage and the hydrogen/helium production rate of material have been analyzed. In the case of neutron damage and overheating, deposition on the superconductive coils and VV, the interface and the shielding design among heating systems, and the blanket and other systems, has been initialized. The results show the shielding design can meet the requirement of coil and VV after several iterated neutronics calculations.

Design of the Water-Cooled Ceramic Breeder blanket for CFETR

The Water-Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for the Chinese Fusion Engineering Testing Reactor (CFETR). In the current version, the WCCB blanket was designed to fulfill the requirements of operating in multiple fusion power modes (200 MW, 500 MW, 1 GW and 1.5 GW) of CFETR while satisfying the tritium self-sufficiency. Structural designs were carried out for typical inboard and outboard modules. Adequate neutron shielding capability and acceptable thermo-mechanical performance were achieved according to the analyses. The global Tritium Breeding Ratio (TBR) was 1.165 with full blanket coverage. It reduced to 1.052 considering the ports occupied by auxiliary heating systems and diagnostic systems. Eventually, the TBR increased to 1.123 by applying the approach of divertor blanket. Besides, two independent cooling systems were applied to adapt to different fusion powers. One cooling system provided coolant for the first wall, the other for breeder zones. The corresponding Primary Heat Transfer System (PHTS) was designed and main operation parameters were obtained under the fusion power of 200 MW∼1.5 GW.

An Innovative Investigation on Fluid-to-Fluid Modeling of Post-Dryout Heat Transfer in Thermal Energy Systems

Post-dryout heat transfer is an important thermal hydraulics phenomenon happening in the loss-of-coolant accident of water-cooled nuclear reactors and the off-design conditions of water-cooled ceramic breeder blanket in fusion reactors. It is necessary to research the flow and heat transfer mechanisms of post-dryout. To economically and comprehensively study on the post-dryout heat transfer characteristics, this work conducted fluid-to-fluid modeling research on water and R-134a data banks by using Buckingham dimensional analysis methodology. In addition, a new interpolation method was developed and verified to make the scaling between the two data banks possible. Finally, a scaling method of post-dryout heat transfer was proposed and assessed by using experimental data. The results showed that the obtained scaling method is reliable in certain conditions.

Tritium transport code development based on COMSOL and MATLAB and its application to the WCCB blanket

Water cooled ceramic breeder (WCCB) blanket was proposed as one of the candidate blanket schemes of the Chinese Fusion Engineering Test Reactor (CFETR). In this paper, a code simulating tritium transport in WCCB blanket and its external loops has been developed, where zero and two dimensional models are integrated to get more reliable results. The purge gas flow in the breeder zone, heat transfer in the blanket and tritium isotopes mass transport are considered. The isotope exchange in the purge gas and coolant loop are also taken into account in the model. The two dimensional model of the WCCB blanket and zero dimensional model of external loops are integrated to calculate tritium related quantities. The amount of tritium inventory in different components, tritium losses, tritium extracted by the tritium extraction system (TES) and purified by the coolant purification system (CPS) are obtained. The sensitivity analysis about key parameters has also been conducted.

Study on electron beam welding process of the first wall of water-cooled ceramic breeder (WCCB) blanket

Water- cooled ceramic breeder (WCCB) blanket is one of the candidate blanket for China Fusion Engineering Test Reactor (CFETR) independently developed by China Institute of plasma physics. The first wall (FW) is a component of WCCB, which directly faces the plasma. FW contains flow channels and has complex shape structure, therefore, it is difficult to complete the full-scale manufacturing by one kind of welding process. Hence, in the development process, the components of FW are first connected by hot isostatic pressing (HIP), and then connected by electron beam welding (EBW), so as to form the full-scaled FW connection. In this paper, the EBW process of FW was studied. The results showed that when the electron beam current was 50 mA and the focus was on the surface, the welded joint with 20 mm penetration and good shape can be formed. After the one time welding and the twice welding (overlapping first welding) of the weld, the tensile test fracture occurred in the base metal. The impact energy of the one time welding was 13 MPa, 18 MPa, and 21 MPa, respectively, and the impact energy of the twice welding was 19 MPa, 20 MPa, and 51 MPa, respectively. The microstructure of the joint was composed of δ-ferrite and lath martensite. When the weld penetration was increased to 50 mm, the lower the focus is, the smaller the porosity at the bottom of the weld is. Similarly, when the lower focus was used, the porosity at the bottom of the weld disappeared basically; The EBW test on the scaled FW with W/Cu sheets showed that when the we took the test pieces with the W-Cu sheet spacing is 4 mm, the welded joint with good surface shape and no internal defects can be formed without scanning deflection electron beam welding, and ultrasonic testing showed that EBW does not affect on the HIP interface. This provides a favorable basis for the future test of equal ratio FW test piece and the welding of actual FW.

Investigation on the Tritium Self-Sufficiency and Neutron Shielding Based on a Water-Cooled Blanket for CFETR at 1.5 GW

China Fusion Engineering Test Reactor (CFETR) is a new test tokamak device in China, which is to bridge the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve tritium self-sufficiency with a tritium breeding ratio (TBR) of no less than 1.05. In addition, the neutron shielding evaluation is also a significant issue, which could provide a significant reference for the water-cooled ceramic breeder (WCCB) blanket modification and improvements. In this work, a 3-D simplified neutronics model containing the WCCB blanket is developed and the neutronics performance is calculated based on the 1.5 GW fusion power. The nuclear analyses are carried out by the Monte Carlo N-particle (MCNP) transport code, including the TBR, neutron wall loading (NWL), fast neutron flux on inboard side, and nuclear heating deposition on main in-vessel components. It is found that the net TBR reaches 1.23, which meets the condition of tritium self-sufficiency. Moreover, it is found that the peak NWL of inboard and outboard occurs in the equatorial plane. The inboard peak NWL is 1.89 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> and the outboard peak NWL is 1.97 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> . In addition, the fast neutron flux and nuclear heating deposition at the inboard midplane in toroidal field coil (TFC) are obtained. According to the radiation limits of the CFETR TFC referring to ITER, the results show the nuclear responses of the TFC meet the limit criteria.

Conceptual design of a blanket module integrated with divertor for CFETR

Achieving tritium self-sufficiency (tritium breeding ratio (TBR)≥1.1) is one of the most important missions of the Chinese Fusion Engineering Test Reactor (CFETR). The neutronics analysis shows that the global TBR of the CFETR water-cooled ceramic breeder (WCCB) blanket is 1.147 when the plasma is surrounded by blankets (without any windows), which can meet the tritium self-sufficiency requirement. In the engineering design phase, however, a number of blanket modules are replaced by auxiliary systems. It would lead to decline of TBR and failure of tritium self-sufficiency. Therefore, a new blanket module concept extended to the divertor region and integrated with the divertor is proposed to increase TBR. The blanket module is located at the back of the divertor target plates. This paper presents the design of the proposed blanket module. The feasibility of the blanket design is evaluated considering neutronics, thermal and mechanical performances.

Optimization of Electron Beam Welding Joint of Water-Cooled Ceramic Breeder Blanket.

The water-cooled ceramic breeder (WCCB) blanket is a component of the China Fusion Engineering Test Reactor (CFETR). The Reduced Activation Ferrite/Martensite (RAFM) steels are the preferred structural materials for WCCB blanket. As a kind of RAFM steels, China low activation martensitic (CLAM) steel was welded by electron beam welding (EBW), and then quenched-tempered treatment was carried out. The results show that at the welding state, the welding seam was composed of large martensite and δ ferrite and the organization of the heat-affected zone (HAZ) was changed slightly with the different heat input. Moreover, the hardness of welded joints was higher than that of base material (BM), but the impact toughness was very low. After quenched-tempered treatment, the δ ferrite in the weld was eliminated, the residual stress of the test plate decreased as a whole, and the mechanical properties were improved significantly.

Annihilation kinetics of irradiation defects in promising tritium breeding pebbles

• γ-Ray irradiation mainly introduces E′-center and O-related (O - -center, O 2 - center et.al) defects in Li 2 TiO 3 , Li 4 SiO 4 and Li 2 TiO 3 -Li 4 SiO 4 pebbles. • Isochronal and isothermal annealing experiment were carried out to illustrate the annihilation kinetics of different breeding pebbles. • Li 2 TiO 3 pebbles have a better irradiation stability than Li 4 SiO 4 pebbles. Lithium-based tritium breeding materials will be adopted to produce tritium in the D-T fusion reactor. Li 2 TiO 3 and Li 4 SiO 4 pebbles have been proposed as candidates in water-cooled ceramic breeder blanket (WCCB) and helium-cooled ceramic breeder blanket (HCCB) respectively. Biphasic ceramics of core–shell Li 2 TiO 3 -Li 4 SiO 4 have been considered as advanced breeding materials due to combining the superiority of Li 2 TiO 3 and Li 4 SiO 4 . The defects will be introduced into tritium breeders in the operation of fusion reactor. They have important effects on tritium release. It is necessary to carry out the experiment on irradiation defects evolution. The annihilation kinetics of defects induced by γ-ray irradiation were investigated. Fluka and Flair were used to calculate the DPA (displacement per atom) of these irradiated pebbles. EPR (Electron Paramagnetic Resonance) characterization and EasySpin simulation were adopted to analyze the evolution and annihilation kinetics processes of irradiation defects. The concentration of defects decreased as the annealing temperature increased. There were still a certain amount of defects in Li 4 SiO 4 and Li 2 TiO 3 -Li 4 SiO 4 when the defects in Li 2 TiO 3 disappeared by annealing. Li 2 TiO 3 pebbles have a better irradiation stability than that of Li 4 SiO 4 pebbles. According to the results of EasySpin simulation, the defect concentration of E’-center and O-related was obtained respectively. The kinetics parameters for the defects of E’-center and O-related center in Li 2 TiO 3 and Li 4 SiO 4 were acquired. The evolution behavior of Ti 3+ in core–shell Li 2 TiO 3 -Li 4 SiO 4 contributes to the recovery of defects. The correlation between annihilation of irradiation defects and tritium release was discussed. The anti-irradiation damage stability of three kinds of ceramic breeders were evaluated. This work carries out the research on the defect kinetics of the new core–shell structure tritium breeder. Meanwhile, a comprehensive comparison of the annihilation properties of three promising breeders has been made. Li 2 TiO 3 pebbles present excellent tritium release performance due to higher annihilation rate constant. Compared the performances of these breeder pebbles, Li 2 TiO 3 plays a positive role in irradiation tolerance and tritium release performance, and Li 4 SiO 4 has higher lithium density which benefit for tritium production.

Experimental study on tritium breeding in water-cooled ceramic breeder blanket mock-up under D–T neutron irradiation conditions

The accurate assessment of tritium breeding parameters within fusion blankets is crucial for future magnet-confined fusion machines to realize fuel self-sufficiency. Such an assessment can be conducted using the simulation approach with nuclear data of high-fidelity and, most importantly, validated against experimental data. In this paper, we report neutronics experiment studies carried out on a mock-up of the water-cooled ceramic breeder blanket of the China fusion engineering test reactor (CFETR), irradiated at a deuterium–tritium (D–T) neutron source. The mock-up’s nuclear responses to 14 MeV neutrons, including tritium production rates (TPR) and neutron-induced reaction rates, are simulated and validated by experimental results. Redundant measurement techniques are used, including Li2TiO3 pellets for offline TPR measurements and the developed lithium glass detector with a significantly reduced size for online TPR measurements. The validation of the TPR value is complemented by the experimental evaluation of the neutron-induced reaction rates for Au and Zr foils. All experimental results are analyzed using the MCNP-4C code and FENDL-3.0 nuclear data library. The source term for the Monte Carlo simulation is built using a newly-developed method based upon the modeling of the depth profiling of tritium in the tritiated target. The experiments are in good agreement with the simulations; ratios of the calculation to experimental results (C/E) on TPR are found to be 0.97–1.08. The influence of the first wall tungsten armor on the mock-up nuclear responses is also studied.