Water Cooled Ceramic Breeder Research Articles

Preliminary tritium transport analysis for the water cooled ceramic breeder blanket system based on a system level code

This paper presents a dynamic analysis tool for evaluating the tritium inventory, tritium extraction and tritium losses for the Water Cooled Ceramic Breeder (WCCB) blanket system of Chinese Fusion Engineering Test Reactor (CFETR). We have developed a program for the tritium transport analysis at the system level, which aims to estimate the tritium related quantities for the WCCB blanket system. The tritium inventory in the material defects is likewise taken into account. The sensitivity analysis is conducted on key parameters is carried out. The key output quantities of the WCCB blanket system include the tritium inventory in structural material, purge gas and coolant loops, tritium extracted by the TES and CPS, and tritium losses.

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.

Investigation of hydrogen generation from the reaction of beryllium pebbles with water vapor for WCCB-TBM

Be pebble, as a neutron multiplier material, is being explored for the development of a Water-Cooled Ceramic Breeder Test Blanket Module (WCCB-TBM) in Japan for the International Thermonuclear Experiment Reactor (ITER) project. The Be pebbles were fabricated by the rotating electrode process (REP). However, as the safety of the test blanket module (TBM) could be compromised by hydrogen generation via Be–H2O reaction owing to the leakage of pressurized water from the cooling pipes inside TBM, the safety assessment of TBM must be conducted using the Be–H2O reactivity data for the Be pebbles whose fabrication method is the same as that for TBM (i.e., REP). Thus, in this study, a Be–H2O reactivity test was conducted using Be pebbles that were fabricated by REP to characterize the generated hydrogen and evaluate the hydrogen generation (HG) rates toward WCCB-TBM safety assessment. First, the Be–H2O reactivity data were obtained at ≥ 300 °C (the coolant temperature for the TBM). The experimental results and oxidation rate of the Be pebbles revealed that only the Be surface may have been oxidized at 300 °C–600 °C, although its inside was oxidized at 700 °C–1000 °C. Further, the experimental data–derived HG rates at 300 °C–1000 °C increased with the temperature and the tendency changed in the three temperature regions. The HG rates did not contradict those reported by previous studies. Additionally, these regions corresponded to each expected Be–H2O reaction behavior at 300 °C–1000 °C. At a Be temperature of < 600 °C, the generated hydrogen from TBM will not exceed 2.5 kg of a limitation by the water leakage on the estimation based on the HG rates and the qualitative effects of the safety functions.

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.

Main Nuclear Responses of the DEMO Tokamak with Different In-Vessel Component Configurations

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.

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.

Neutronic Activity for Development of the Promising Alternative Water-Cooled DEMO Concepts

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&amp;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.

Evaluation on Nonirradiated Thermo-mechanical Response of WCCB TBM Following RCC-MRx

Since the submodules as the most important structural component in Water-Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) contributes to tritium breeding and extracting, structural integrity assessment is necessary to keep normal operation and function of WCCB TBM-set. Recently, the nonirradiated material properties of F82H which is considered as the structural material of submodule has been summarized following the requirement of RCC-MRx, combined with EU directive, French order, and pressure vessel construction code (RCC-MRx), structural integrity assessment is investigated under the ITER relevant thermo-mechanical load in this study. Meanwhile, cooling water conditions of WCCB TBM-set with its operation scenario, relationship, and detailed values (temperature and pressure) are constituted. Based on the determined structural integrity assessment procedure from RCC-MRx and numerical analysis results, structural integrity of current design of submodule in WCCB TBM-set could be validated. According to the current achievements and established structural integrity assessment procedure, further evaluation on structural integrity combined with electromagnetic load shall be investigated.

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.

Research on hot isostatic pressing process of CFETR WCCB components

Chinese Fusion Engineering Test Reactor (CFETR) is a superconducting magnet tokamak reactor of equivalent size and complementary functionality to that of ITER. The water-cooled ceramic breeder (WCCB) breeder blanket designed by the Institute of Plasma Physics, Chinese Academy of Sciences is one of the candidate breeder blanket concepts for CFETR. As an important part of the fusion reactor, the WCCB directly determines the safe operation of the fusion reactor. In the WCCB manufacturing technology, the hot isostatic pressing (HIP) process is applied to the manufacture of WCCB due to its unique technical characteristics. This paper introduces the process research and inspection during the development of WCCB components, and the results will be the guidance for the manufacture of WCCB Components.

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.

Research of flow instabilities in parallel channels with opposite flow directions for WCCB blankets

In the current water-cooled ceramic breeder (WCCB) blankets, a special cooling scheme, the flow direction of adjacent channels is opposite, has been designed. In this paper, flow instabilities in parallel channels with opposite flow directions (OFD) for WCCB blankets were studied using RELAP5/MOD3.4. Flow instabilities were investigated under the operating condition matrix composed of different pressure, inlet subcooling, and mass flux. Flow excursion and density-wave oscillations (DWOs) were identified. Their boundaries were analyzed with the variation of operating conditions and synthesized using empirical correlations respectively. The cooling scheme of identified flow directions (IFD) was designed and flow instabilities were calculated as well. By comparing the OFD and IFD schemes, it is concluded that the IFD scheme is more stable than the OFD scheme for flow excursion; for out-phase DWO, the relative stability depends on the operating conditions. This research results can provide reference for the design of WCCB blankets.