Ceramic Breeder Research Articles

Study on the EM Loads Distribution on Updated HCCB Blanket Inboard and Outboard Blanket Vertical Segments

As part of the ongoing China Fusion Engineering Test Reactor (CFETR) studies, this work focused on the study of the electromagnetic (EM) loads distribution on inboard (IB) and outboard (OB) blanket vertical segments. In this paper, a major central disruption scenario with a 36-ms linear current quench was performed in the ANSYS finite element software. EM forces and moments, both on the single module and the complete blanket segment, were calculated for the updated helium-cooled ceramic breeder (HCCB) blanket. Furthermore, the effect of EM loads with respect to poloidal segmentations on the back plate (BP) was studied. Results show that the EM forces and moments in the IB blanket barely changed, but in the OB blanket, great changes occurred. With poloidal segmentations increasing, EM forces loads on the BP will be reduced. This paper may provide important data to support subsequent optimization of the BP of CFETR and could also be a technical reference for other fusion organizations.

Evaluation on the method of reducing CFETR TF ripple generated by WCCB blanket

Reduced Activation Ferritic/Martensitic (RAFM) steel is selected as structural material for the water cooled ceramic breeder (WCCB) blanket for China Fusion Engineering Test Reactor (CFETR). Due to the magnetization of RAFM steel, the toroidal field (TF) ripple is generated by the WCCB blanket. The TF ripple induces significant losses of high-energy particles, and then causes issues of heat load on plasma-facing components (PFCs) and energy balance of plasma. Hence, TF ripple generated by WCCB blanket is calculated, and several effective optimization methods are explored in this paper. The simulation results indicate that TF ripple in plasma target separatrix is greater than design limits (δTF<0.5%) in present design. Four methods are employed to study and evaluate the effect of decreasing TF ripple including installing ferromagnetic inserts, using correction coils, changing the related parameter of diagnostic system and moving the WCCB blanket, respectively. Based on the analysis results, it is found that these four methods can reduce TF ripple to some extend through different optimizing methods. The location of five TF ripple peaks of a 22.5° sector are at 0°, ±3.07° and ±7.67° respectively. The TF ripple can only reduce from 0.25% to 0.21% at ±3.07° plane using optimization method of installing ferromagnetic inserts. The rest of the location has almost little impact. The TF ripple can only reduce from 0.31% to 0.20% at 0° plane using correction coils, but the peak value will increase to 0.33% at ±3.07° plane. Moving diagnostic system 40 mm outward in radial direction will effectively decrease the TF ripple from 3.54% to 0.37% at -3.07° plane. Moving WCCB blankets 20 mm outward in radial direction makes the maximum TF ripple from 0.56% to 0.09%. Discussions are made about the effect of the optimization methods for reducing TF ripple.

Release kinetics of tritium generation in neutron irradiated biphasic Li2TiO3–Li4SiO4 ceramic breeder

Biphasic Li2TiO3–Li4SiO4 tritium breeders with different phase ratios were prepared by the solution combustion synthesis method. The as-prepared biphasic Li2TiO3–Li4SiO4 and single-phase materials were irradiated by the thermal neutron with the flux of 2.75 × 1013 n cm−2 s−1 at Kyoto University Research Reactor (KUR). Tritium is generated based on the following reaction: 6Li+1n→3T+4He+4.8 MeV. The results of tritium release experiment show that tritium release starts at the lower temperature range for the biphasic materials. The tritium release temperature and the shape of tritium-TDS (Thermal desorption spectroscopy) spectra are dependent on the phase ratio of Li2TiO3 to Li4SiO4. In the case of Li2TiO3/Li4SiO4 = 2.0, tritium release starts at 400 K, and the tritium migration is mainly controlled by the diffusion process. With the increase of Li4SiO4 content in the biphasic materials, tritium release is influenced by the de-trapping process.

Mechanical compression behaviors of single phase Be and binary Be12Ti pebbles

Beryllium pebbles are planned to be used as a neutron multiplier in the Helium-cooled ceramic breeder (HCCB) tritium breeding blanket module (TBM) of ITER, which is also the primary option of the Chinese TBM program. Advanced neutron multipliers such as Be12Ti beryllide with high stability at elevated temperature are desired for demonstration power plant (DEMO) reactors. Prototypic pebbles 1 mm in diameter of single phase Be and 0.7 mm in diameter of binary Be12Ti beryllide were successfully fabricated by a combination method involving hot isostatic pressing and rotating electrode process. Surface microstructure of Be and Be12Ti pebbles were analyzed by using AFM and SEM. Compression properties of Be and Be12Ti pebbles were evaluated by mechanical compression. Beryllium pebbles displayed high ductility (50% deformation without fracture), which is better than that of beryllide pebbles. One method based on contact mechanics was introduced to evaluate the range of effective modulus and stiffness of Be and Be12Ti from compression load curves in the elastic regime.

Considering for the blanket structure scheme of HCCB DEMO

ABSTRACT For the solid blanket concept of helium cooled ceramic breeder (HCCB) demonstration fusion power plant (DEMO), a feasible blanket structure with configuration 2×X is proposed as considering relatively low temperature limit of neutron multiplier beryllium pebbles. Based on that, preliminary design for the typical blanket module of HCCB DEMO has been carried out and verified by thermal-hydraulic analysis and structural analysis. Furthermore, the specific relationship of maximum temperature depended on the surface heating of blanket key part first wall (FW) is also analyzed.

Progress on design and related R&D activities for the water-cooled breeder blanket for CFETR

The water-cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for Chinese fusion engineering testing reactor (CFETR) and is being developed at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). This paper reviews design and evolution of the WCCB blanket for CFETR, and presents a new WCCB blanket design according to the latest CFETR core parameters (major and minor radii are R = 7.2 m and a = 2.2 m, respectively) and missions. This new design is expected to satisfy multiple CFETR operation modes of 0.2, 0.5, 1.0, and 1.5 GW fusion power and achieve tritium self-sufficiency. The feasibility of the updated blanket design is evaluated from the aspects of neutronics and thermo-hydraulics. Furthermore, the research and development (R&D) activities supporting to the WCCB blanket for CFETR are reported, including the design code, the water loop experiments, the pebble bed modeling and experiments, and the components fabrication technology.

Steady states and LOFA analyses of the updated WCCB blanket for multiple fusion power modes of CFETR

The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates of Chinese Fusion Engineering Test Reactor (CFETR). The core parameters of CFETR have been updated to major radius R = 7.2 m, minor radius a = 2.2 in 2018. Accordingly, the WCCB blanket design was also updated to satisfy multiple operation modes of CFETR with the fusion power of 200 MW, 500 MW, 1.0 GW and 1.5 GW. The WCCB blanket is characterized by 3 independent cooling systems (CSs), namely CS1 for the cooling of the First Wall (FW), CS2&3 for Breeding Zones (BZs). At 200 MW of low fusion power, CS3 is idle. At 500 MW to 1.5 GW of high fusion power, all 3 CSs are put into use to ensure adequate cooling. So it is indispensable to evaluate the feasibility of the updated WCCB blanket from the thermal hydraulic and safety point of view. A system analysis code RELAP5, was employed to model the blanket. The steady states were analyzed under 200 MW, 500 MW, 1.0 GW and 1.5 GW of fusion power. Then, the Loss of Flow Accident (LOFA) was simulated with different assumptions to verify the robustness of 3 cooling systems design under different initial postulations. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature, can basically meet the design criterion, which preliminarily verifies the feasibility of the WCCB blanket from the thermal hydraulic safety point of view. Comprehensive system design and analysis is under way.

Design and analysis of mock-up of CFETR water cooled ceramic blanket for neutronics experiment

The water cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for China Fusion Engineering Test Reactor (CFETR). The neutronics experiment of the WCCB blanket is to validate the neutron transport codes and nuclear data libraries. A scaled down mock-up of the WCCB blanket should be designed considering constraints of economic costs and experimental conditions before the neutronics experiment. In this work, the mock-up was designed keeping neutronics and geometric characteristics of the original equatorial outboard blanket module of CFETR. The radial arrangement of the mock-up was designed firstly, including tungsten armor, first wall, breeding zone 1, beryllium layer 1, cooling plate 1, breeding zone 2, beryllium layer 2, cooling plate 2. Then the toroidal and poloidal dimensions of mock-up were optimized. Five mock-ups with the same radial arrangement and different toroidal and poloidal dimensions were selected as initial mock-ups. The nuclear response of these five mock-ups, including neutron intensity, neutron spectra and tritium production rate were analyzed. After the optimization, the size of the experimental mock-up was determined as 20.00 cm (Toroidal) × 20.00 cm (Poloidal) × 17.27 cm (Radial). The analyzed results showed that the neutronics performance of the mock-up were similar with the real blanket module and it can be employed to carry out neutronics experiments.

Assessment of the activation induced by neutron irradiation in K-DEMO and thermal response under the decay heat

The radioactivation of in-vessel components due to the fusion neutrons is an unavoidable trade-off in a tokamak reactor. We investigate the radioactivity level of nuclides and decay heat of conceptual water-cooled ceramic breeder blanket and divertor modules in K-DEMO, and demonstrate that a cooling scheme related to the decay time is important to prevent thermal failure of those components due to decay heat during their maintenance. For K-DEMO with a fusion power of 2.2 GW, the activation levels of blankets and divertors are evaluated regarding a regulatory low level limit in Korea. Total decay heat from radioactivated blankets and divertors reaches 63.4 MW after the full power operation of two years. In an outboard module, immediately after the plasma shutdown, local maximum temperature reaches 1300°C with the temperature difference up to 840 °C inside the module. Conservative natural convection was assumed to validate its integrity to remain within the allowable temperature range of the materials used, when provided that the cool-down time is secured at least a couple of days, the overall temperature of the module is reduced to about 200°C in 10 days. It is worth consideration that a forced convective scheme such an internal passage cooling can be tailored until a critical time corresponding to configurations of each module.

Numerical investigation of mechanical and thermal characteristics of binary-sized pebble beds using discrete element method

The characteristic identification of the functional materials is important in predicting the performance of a breeding blanket, which is one of the main components of the fusion power plant. The functional materials of the solid-type ceramic breeding blanket are mainly used in the form of pebble beds. Various experimental methods such as laser flash, hot wire, and hot disk method have been utilized to assess the thermal properties of the pebble bed. In recent years, numerical techniques have been also introduced to evaluate mechanical characteristics as well as thermal characteristics of pebble beds. In this study, the mechanical and thermal behaviors of the binary-sized pebbles in a cuboidal container are simulated considering the contact interaction of the pebbles by means of discrete element method. Through a series of numerical simulations of pouring the pebbles into the vibrating container, we investigate the effect of the size difference of the binary-sized pebbles on the packing structures of the pebble beds. The connectivity of binary-sized pebbles is also discussed in terms of the average contact force of the pebbles. Furthermore, heat conduction between the pebbles is considered based on the packing configurations obtained from the mechanical analysis, and the thermal properties of the pebble beds are qualitatively evaluated in terms of the effective thermal conductivity of the pebble beds.

A DEM-CFD study of the effects of size distributions and packing fractions of pebbles on purge gas flow through pebble beds

In a solid-type ceramic breeding blanket, the functional materials, such as the breeder, multiplier, and reflector, are equipped in the form of pebble beds, while helium purge gas flows through pebble beds. The flow characteristics of purge gas need to be investigated in order to design and evaluate the performance of the blanket system, which depends on the configuration of pebble beds. In the current study, various pebble bed models were constructed according to the discrete uniform distribution and normal distribution of pebble sizes, and the packing fractions of the pebble beds were controlled using the discrete element method. Computational fluid dynamics simulations for the laminar flow of helium purge gas through pebble beds were then conducted. In particular, the effects of size distributions and packing fractions of pebble beds on the flow resistance were investigated in terms of pressure drop. Through a series of numerical analyses, it was found that the pressure drop increases not only, obviously, in proportion to the packing fraction of pebbles, but also in inverse proportion to the difference in pebble size. Since the pressure drop is related to the surface areas of pebbles, a smaller difference in the pebble size leads to a higher pressure drop due to a larger surface effect at the same volume of pebbles. The numerical results were also validated by comparing them to the semi-empirical formulae derived for the prediction of pressure drop in packed beds, such as the Ergun equation and the Kozeny-Carman equation. The current study will provide a basis for resolving the practical issues on blanket system design by extending it further to the purge gas flow and conjugate heat transfer that involve fragmentation and resettlement of pebbles.

Tritium breeding capability of water cooled ceramic breeder blanket with different container designs

Water cooled ceramic breeder (WCCB) blankets have been regarded as a primary concept in Japan. We investigated the tritium breeding capability of a WCCB blanket with a cylindrical structure, an advantageous shape for withstanding high coolant pressure, in this study. The breeding area in the blanket was modeled as a homogeneous mixture or heterogeneous structures. Nuclear analyses with Monte Carlo method were conducted for evaluating the tritium breeding ratio (TBR) of the blanket. The neutron balance of the blanket was analyzed to elucidate the mechanism related to the increase in TBR when an additional thin breeding layer (envelope) was applied to the blanket. The neutron multiplication in (n, 2n) reaction increased concomitantly with increase of the beryllium volume ratio. However, numerous neutrons were not used efficiently for tritium production, but were captured by the container, which was made of the reduced activation ferritic martensitic steel, F82H. Capture by the container can be reduced by introducing an envelope. The effect of the envelope was considerable when modeling the internal structure such as U-shaped pipes in the breeding area. The envelope was also applied to a blanket with different container designs. An increase in TBR was shown irrespective of blanket design. The envelope effect was remarkable when it was difficult to achieve the internal configuration, which was equivalent to the homogeneous mixture in the breeding area.

Effect of hydrogen on corrosion properties of reduced activation ferritic/martensitic steel, F82H

Reduced activation ferritic/martensitic (RAFM) steel, e.g., F82H, is the leading candidate structural material for fusion blanket. Of many blanket concepts, the water-cooled ceramic breeder blanket is an attractive concept because of its compactness and its compatibility with the technologies which using high-temperature water as coolant in conventional light water reactor. For tritium breeding, it is necessary to manage the corrosion of F82H by controlling not chemicals but hydrogen and oxygen in the cooling water. Understanding the effects of oxygen and hydrogen on corrosion property was therefore required for the blanket design. This work aims to investigate the corrosion behavior of F82H in high-temperature water with hydrogen added. Based on the results of the corrosion test, it is speculated that hydrogen addition expected no considerable change of the flow-accelerated corrosion (FAC). In contrast, it was found that the number of cycles to fracture was decreased in the high-temperature water in comparison to the fatigue test at room temperature in air.

Study of lithium germanate additions to advanced ceramic breeder pebbles

1. Abstract Worldwide, lithium rich ceramic pebbles are being developed to be used as tritium breeders inside future fusion reactors. Upon irradiation, the pebbles will produce tritium which will be used alongside deuterium to fuel the fusion reaction. It is essential that these pebbles are of sufficient quality to withstand the various forces experienced within the wall of the reactor. Currently, a melt-based process is used at the Karlsruhe Institute of Technology for the production of pebbles composed of lithium orthosilicate with a secondary strengthening phase of lithium metatitanate. Recent studies have focused on incorporating lithium orthogermanate, which forms a solid solution with lithium orthosilicate. Various compositions were used to perform fundamental studies to determine phase transitions as well as melting points for processing. The melt-based KALOS process was then used to produce pebbles with selected compositions. Thereafter the pebbles underwent a series of characterisation tests including microstructure and surface analysis using REM as well as elemental mappings using EDX. XRD was used for determining the crystal structures present in various compositions and the unit cell volumes. The measurements showed that the unit cell volume of each structure linearly depends on the composition. Additionally, pebbles were held at reactor relevant temperatures to determine any changes to the mechanical strength. It was found that pebbles with 30 mol% lithium metatitanate and a lithium orthosilicate/orthogermanate ratio in the range of 1.5–4 showed an increased strength as well as good thermal stability and may have the potential to be used as tritium breeders.

Updated design of water-cooled breeder blanket for CFETR

The water-cooled ceramic breeder (WCCB) blanket is one candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), and it is being developed at the Institute of Plasma Physics of the Chinese Academy of Sciences. This paper presents an updated WCCB blanket concept design that is expected to cover the CFETR operation modes of 200 MW, 500 MW, 1 GW, and 1.5 GW fusion power and achieve tritium self-sufficiency with the latest core parameters (major radius R = 7.2 m; minor radius a = 2.2 m) and mission. The feasibility of the new blanket design is evaluated considering neutronics and thermal-hydraulics aspects.

Experimental investigation of thermal diffusivity and heat capacity of ceramic breeder beds

In the solid Breeder Blanket (BB) concepts both tritium release and heat recovery depend on the thermal performances of the breeding zone. Within the R&D activities of the Helium Cooled Pebble Bed (HCPB) breeding blanket, the knowledge of the thermal diffusivity of the breeder beds is of fundamental importance to model the transient heat transfer during the power pulses of the fusion machine. The aim of the present study is to investigate the thermal diffusivity of the breeder beds at BB relevant conditions; to this end the line heat source probe method was employed together with differential scanning calorimetry. An experimental facility, based on the line heat source probe method, was devised for the investigation of the thermal diffusivity of granular beds at breeder blanket relevant temperatures, mechanical state, purge gas type and pressure. Besides the experimental approach, literature values were used to estimate the specific heat capacity of the breeder materials starting from the specific heat capacity of the constituent compounds based on the mole fraction. In addition to the thermal diffusivity and heat capacity, a preliminary insight on the phase transitions of the reference and advanced ceramic breeder beds is given.

Neutronic assessment of HCCR breeding blanket for DEMO

Helium-cooled ceramic reflector (HCCR) blanket concept was proposed as a DEMO breeding blanket in Korea. It adopted lithium ceramic breeder, graphite reflector to reduce the beryllium multiplier, reduced activation ferritic-martensitic steel as a structural material, and high pressure/temperature helium as a coolant. As a pre-conceptual design study, neutronic assessment of the HCCR blanket was carried out by inserting the newly developed model into the K-DEMO three-dimensional neutronic calculation model. As results, the peak neutron wall load of 4.06 MW/m2 was obtained at the outboard midplane blanket first wall for the fusion power of 3 GW. The overall tritium breeding ratio (TBR) was estimated as 1.10. And the poloidal mean global TBR distribution at the blankets were calculated.

Study on ripple effect of ferromagnetic WCCB blanket modules in CFETR

China Fusion Engineering Test Reactor (CFETR), the next fusion device in China, is proposed to bridge the gaps between ITER and the demonstration reactor (DEMO). The Water Cooled Ceramic Breeder (WCCB) blanket, one candidate option of breeding blanket conception for CFETR, impacts the toroidal field (TF) ripple in CFETR, as the reduced activation ferritic/martensitic (RAFM) steel was used as its structure material. The purpose of this paper is to investigate and evaluate the ripple effect of WCCB blanket on TF ripple in CFETR. An analysis is performed using three-dimensional finite element model with 16 TF coils and WCCB blanket modules. ANSYS code is employed for this study. The results indicated that the TF ripple was less than 0.5% at all points in target separatrix when it was estimated with TF coils only. We also calculated TF ripple in case of WCCB blanket modules and various port configuration schemes. It is found that TF ripple cannot meet the design requirements by the presence of WCCB blanket modules, even in difference port configuration. Discussions are made on the implication of the WCCB blanket modules and ports on TF ripple. Potential methods for reducing the TF ripple are studied in this paper.

An enhanced, near-term HCPB design as driver blanket for the EU DEMO

The Helium Cooled Pebble Bed (HCPB) breeding blanket is a candidate as driver blanket for the EU DEMO. The reference design of the HCPB is based on a cooling plate “sandwich” arrangement built in Multi-Module Segments. This architecture significantly improved the tritium breeding performance (TBR = 1.15) and the plant circulating power (≈130 MW) compared to the former ITER-like “beer-box”-like design (TBR<1.10, plant circulating power>200 MW). However, several issues remain with this design, in which (1) the still large power required per He circulator (beyond the state-of-the-art for these components) and (2) the large tritium inventory foreseen in Be have been identified as the most critical. After a basic research for a concept that mitigates these key issues, a fission-like configuration based on a hexagonal matrix of radial fuel-breeder pin assemblies has been found to solve many of the outstanding issues present in the reference design. This design has been developed for the latest EU-DEMO 2017 baseline. It features Be12Ti neutron multiplier, which largely reduces the tritium retention, swelling and reactivity issues compared to pure Be. The substitution of the relatively complex CPs by simpler fuel-breeder pins improves the heat transfer capability of the breeder zone (BZ), allowing 20°C higher blanket outlet temperature and drastically reducing an order of magnitude the pressure drop in the BZ. The paper reports the successful set of nuclear, thermo-hydraulic and thermo-mechanical performances, together with some notes about the manufacturing and assembly strategy, the improved RAMI features and emphasizes the reduction of the plant circulating power to a remarkably low figure (˜90 MW), enabling the use state-of-the-art He turbomachinery. Due to some inherent issues connected to the use of Be as neutron multiplier which cannot be simply solved by design, a parallel long research has identified molten lead as a cost-effective, low risk substitute for Be. Taking as basis the HCPB fuel-breeder pin concept, an alternative helium cooled Molten Lead Ceramic Breeder has been derived and some key performance figures are reported. The paper concludes with a discussion on the results and their implications on future development steps.

Progress on the helium cooled Molten Lead Ceramic Breeder concept, as a near-term alternative blanket for EU DEMO

Within the EUROfusion framework, a Molten Lead Ceramic Breeder (MLCB) breeding blanket is being developed at KIT as an alternative, cost-effective and low technology-risk solid breeder blanket for EU DEMO. The use of molten lead as neutron multiplier has been identified very promising, solving key potential issues coupled with the use of Be in solid breeder blankets. This MLCB concept has been derived from the enhanced Helium Cooled Pebble Bed, which is based on fission-like “fuel-breeder pin” configuration. Compared to former cooling-plate configuration, this concept greatly reduce the pressure drop, enabling the use of existing helium-circulator. After initial design studies with this concept for the previous EU DEMO baseline (BL2015), the status of the design activities on MLCB for the latest EU DEMO baseline (BL2017) is presented in this paper.First, neutronics analyses have been performed to assess and optimize the tritium breeding performance. Then, structural integrity under in-box LOCA has been assessed and the design has been iterated to fulfill the design criteria. Thereafter, thermo-hydraulic analyses have been conducted to evaluate the temperatures and pressure drops. After several design iterations to satisfy the temperature and pressure drop requirements, a structural assessment under normal condition has been conducted respecting the RCC-MRx rules. The results show that the current MLCB design fulfills the basic nuclear and thermo-mechanic-hydraulic requirements, setting the path for a consolidated design of this concept. Due to the absence of water-reactive elements in MLCB concept, water can be also used as coolant, which is under investigation.