Water-cooled Solid Breeder Blanket Research Articles

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.

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.

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.

Key achievements in elementary R&D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

This paper presents the significant progress made in the research and development (R&D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li2TiO3 pebble bed under neutron-pulsed operation conditions are summarized.With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 °C followed by normalizing it at 930 °C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R&D on the breeder material, Li2TiO3, the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li2TiO3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li2TiO3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation.

Design study of fusion DEMO plant at JAERI

Three options of fusion DEMO plant are proposed characterized by functions of the center solenoid (CS). The prime option uses a downsized CS, which does not provide sufficient V-s for plasma current ramp-up but supplies enough coil current for plasma shaping. This option produces a fusion output of 3 GW with a major radius of 5.5 m, aspect ratio of 2.6, normalized beta of 4.3 and maximum field of 16.4 T. The estimated reactor weight is lighter than that of other conventional tokamak reactors, suggesting an economic advantage. The plant uses rather conservative technologies such as Nb 3Al superconductor, water-cooled solid breeder blanket, low activation ferritic steel as the structural material and tungsten monoblock divertor plate. The design philosophy and key issues related to the constituent technologies of the plant are described in the present paper.

A European proposal for an ITER water-cooled solid breeder blanket

Abstract The water-cooled solid breeder blanket concept proposed here aims to replace the shielding blanket for the enhanced performance phase of the international thermonuclear experimental reactor (ITER). The nominal performances are as follows: an average neutron wall load of 1 MW m−2 which corresponds to a fusion power of about 1.5 GW, and an average neutron fluence of 1 MWy m−2. The proposed blanket concept has been designed to accept a power increase of about 30% and power transients up to 3–5 GW for a short time. This blanket concept is based on a breeder inside tube (BIT)-type blanket with poloidal breeding elements made of 316 L-type stainless steel and filled with lithium metazirconate and beryllium pebbles. Inlet and outlet water temperatures of 160 and 200°C have been considered with a medium-pressure cooling system during plasma burn. The diameters of the breeding elements are compatible with the space available in test fission reactor core channels, making in-pile testing, required for blanket development and qualification, easier. A conservative approach using qualified materials, a blanket concept easily testable in fission reactors and on-going mock-up testing, which can be qualified using blanket test modules during the basic performance phase of ITER, will allow the blanket reliability required for the enhanced performance phase to be achieved.

A European proposal for an ITER water-cooled solid breeder blanket

A European proposal for an ITER water-cooled solid breeder blanket

A Parametric Study of Water-Cooled Solid Breeder Blanket Designs

An assessment of the role of the design and operating parameters on the performance of solid breeder blankets based on lithium ceramics (Li/sub 2/O and ..gamma..-LiAlO/sub 2/) was carried out. The results indicate that not only poor thermophysical properties but also uncertainties associated with the property data base are the design-limiting factors. In addition, the operating conditions such as the upper and the lower temperature limits, the choice of breeder materials either in the form of sintered pellets or in sphere-pac form, the interfacial contact resistance between the coolant channels and the solid breeder, and the diffusion characteristics of tritium and chemical interactions between tritium and the solid breeder play a prominent role in selection of blanket concepts. Designs to account for the expected degradation of the thermophysical properties due to thermal sintering and nuclear irradiation lead to high coolant and structural material fractions, and thus may result in a lower tritium breeding ratio. The results of the parametric studies show that water-cooled solid breeder blanket designs require a firmer data base for the operating temperature limits, the thermophysical properties, the gap conductances, and the tritium retention and release characteristics of solid breeders.