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In order to verify the performance and safety of the solid-type breeding blanket concept, the thermal conductivity of the pebble bed is required as an essential parameter for design and analysis of the blanket. The hot wire, hot disk, guarded hot plate, and laser flash method are considered as measurement technique for the thermal conductivity of pebble bed. This study aims at the provision of the thermal conductivity of Li2TiO3 pebble bed, which is one of the candidates for tritium breeder material in the solid-type breeding blanket. The laser flash method was adopted due to its advantages compared with other methods, such as high accuracy and repeatability, easy sample preparation, absolute measurement technique, and so on. In this method, the thermal diffusivity is directly measured by the laser flash system, and then the thermal conductivity is obtained by multiplying density and specific heat to the measured thermal diffusivity. The thermal conductivity of the Li2TiO3 pebble bed with packing factor of 57% increased with increasing temperature up to 500 °C and was saturated to 1.3 W/m/K at more than 500 °C.

Korea has own programs toward DEMO and fusion reactors that will require further improved DEMO blanket and energy utilization systems. Primary goals of DEMO blanket development in Korea are to develop and verify the integrated blanket design tools; to develop blanket materials, cooling and tritium fuel cycle technologies; and to develop and evaluate fabrication and joining technologies. The concept of helium-cooled ceramic reflector (HCCR) blanket is adopted to be tested in ITER as Test Blanket Module (TBM). Currently, the design and R&D activities are mainly performed through the ITER TBM program in Korea. It is expected to demonstrate the major objectives of the breeding blanket: extraction of heat from burning plasma, tritium self-sufficiency and its integrity within the whole systems considering safety features. Although TBM program will provide very essential data and experience, there will be a considerable technical gap from DEMO blanket. This paper presents the technical gap in the main technical categories of the breeding blanket based on the experience of the development of HCCR TBM and breeding blanket. It is found that still about 60–70% of the DEMO blanket technology can be achieved through the HCCR TBM depending on the technology maturity level, and some ways to DEMO blanket from HCCR TBM are proposed.

The design of diagnostics system for the Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) has now progressed from generic approaches to the conceptual design integration phases. It is proposed to monitor the temperature, strain, magnetic field strength and neutron flux in selected positions of the corresponding Test Blanket Module (TBM). The temperature can be measured by thermocouple sensors to be integrated for the first wall and breeding zones of TBM. The strain will be monitored by appropriate gauges on the first wall of TBM. Several magnetic pick-up coils could be installed to measure magnetic field strength inside the TBM structure. For the neutron flux monitoring, pairs of Neutron Activation System (NAS) probes and Micro-Fission Chambers (MFCs) would be installed in the TBM structure. In this paper, a brief description on the proposed diagnostics system as well as the key components is presented, and the integration challenges of HCCR TBS diagnostics system are outlined.

Development of the Helium Cooled Ceramic Reflector (HCCR) breeding blanket is underway according to the development strategy of core technology for K-DEMO under the fusion energy development roadmap in Korea. The main goals are to validate integrated design tools and to develop core technologies in the field of materials, manufacturing and joining technologies, helium cooling and tritium technologies, system integration, and safety technologies. A unique graphite reflector concept was adopted to save cost by reducing the amount of beryllium neutron multiplier. The main functions and design concepts of the HCCR breeding blanket are to be verified in ITER through the HCCR Test Blanket Module (TBM) program. Not only breeder modules are to be tested in ITER but also the technologies of cooling, coolant purification, tritium extraction, etc. are to be tested or proved in ITER before employed in DEMO. This paper introduces the current updates of the design and R&D activities of the HCCR TBM systems in the preliminary design stage.

Since the stress due to the thermal load on breeding blanket structure is one of the main design driver, the thermal conductivity is necessary input data for thermal-structural and thermo-hydraulic analyses performed in order to understand the heat transfer phenomena and estimate the thermal stress. Since for the functional materials of solid type breeding blanket a pebble-bed form is mainly adopted instead of a bulk form such as a block or a disk, it should be needed to measure the thermal conductivity of pebble-bed. In this study, the effective thermal conductivity of pebble-bed is measured by laser flash method, which is one of the various thermal conductivity measurement methods, because this method has several advantages such as a wide thermal conductivity range of the measurement and a small amount of pebbles. A sample holder considering the heat transfer mechanism from the laser source to pebble-bed has been specially designed in order to apply the laser flash technique to the pebble-bed sample and it has been validated by the experiments. This paper introduces preliminary results of the effective thermal conductivity on the pebble-bed using this sample holder.

Using a conceptual design of the Korean helium cooled ceramic reflector (HCCR) test blanket module (TBM) including the TBM-shield for testing in ITER, a structural analysis with electro-magnetic (EM) loads is performed. From a previous analysis of the material magnetization due to the use of reduced activation ferritic-martensitic (RAFM) steel as the TBM structure material and EM analysis considering the major disruption of MD-I, MD-II, and MD-IV, the forces are obtained and used for the current structural analysis. The results indicate that the maximum stress occurs at the He purge line at the upper and lower region of the breeding zone (BZ) box including the graphite reflector region, which meets the design requirement. In addition, displacements are lower than the designed gaps from the TBM port plug (PP) frame. The results are provided to the load combination analysis.

Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) will be installed at the #18 equatorial port of the Vaccum Vessel in order to test the feasibility of the breeding blanket performance for forthcoming fusion power plant in the ITER TBM Program. Since ITER tokamak contains Vaccum Vessel and set of electromagnetic coils, the TBM as well as other components is greatly influenced by magnetic field generated by these coils. By the electromagnetic (EM) fast transient events such as major disruption (MD), vertical displacement event (VDE) or magnet fast discharge (MFD) occurred in tokamak system, the eddy current can be induced eventually in the conducting components. As a result, the magnetic field and induced eddy current produce extremely huge EM load (force and moment) on the TBM. Therefore, EM load calculation is one of the most important analyses for optimized design of TBM. In this study, a 20-degree sector model for tokamak system including central solenoid (CS) coil, poloidal field (PF) coil, toroidal field (TF) coil, vaccum vessel, shield blankets and TBM set (TBM, TBM key, TBM shield, TBM frame) is prepared for analysis by ANSYS-EMAG tool. Concerning the installation location of the TBM, a major disruption scenario is particularly applied for fast transient analysis. The final goal of this study is to evaluate the EM load on HCCR TBM during plasma major disruption.

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements.

Korea plans to test a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER. The HCCR TBM adopts a four sub-module concept considering the fabricability and the transfer of irradiated TBM for post irradiation examination. Each sub-module has seven-layer breeding zone, including three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebble bed packed tritium breeder layers, and a reflector layer packed with graphite pebbles. Based on this configuration, neutronic and electromagnetic calculations were performed and their results were applied for the conceptual design of HCCR TBM that considers manufacturing feasibility. Also, a design and safety analysis of HCCR Test Blanket System (TBS) was performed using integrated design tools modifying nuclear system codes for helium coolant and tritium behavior evaluation. The Advanced Reduced Activation Alloy (ARAA) is being developed as a structural material. A total of 73 candidate ARAA alloys were designed and their out-of-pile performance was evaluated. The graphite pebbles as the neutron reflector were fabricated by using mechanical machining and grounding method with the surface coated with SiC. The hydrogen permeation characteristics of structural materials were evaluated using the Hydrogen PERmeation (HYPER) facility. The recent design and R&D progress on these areas are addressed in this paper.

We evaluate electromagnetic (EM) loads on the main systems of the ITER machine using a single finite element model. The 20° sector of the full ITER machine includes the main in-vessel components as well as the vacuum vessel. Narrow slits of the in-vessel components are effectively modeled by using the element splitting method without significant increase of computation memory and time as well as without sacrificing the accuracy. Furthermore, the halo current is taken into account at the same time together with the plasma current. To apply both currents concurrently, dedicated conversion codes are utilized to transfer the plasma simulation results by DINA to the electromagnetic analysis by ANSYS-EMAG used here. The electromagnetic loads on the ITER machine are calculated for various disruption scenarios. Investigation on the analysis results is made to find the worst plasma disruption case and the design-driving load component for each system as well as to compare load contribution from eddy and halo currents. The effect of the narrow slits on load reduction is also examined.

Recently Korea has decided to test helium cooled ceramic reflector (HCCR) concept for ITER test blanket module (TBM) which will be inserted in TBM frame in dedicated ITER equatorial port. As the previous design was performed based on the TBM frame with the horizontal opening and the openings of all the TBM frames have been changed to vertical orientation, design change to vertical configuration is inevitable. This design change has significant impact on the TBM performance and design parameters. Therefore, new design iteration is required. In this paper, the updated design of the Korean HCCR TBM to vertical configuration is presented. While maintaining the main feature of employing graphite reflector, the new vertical HCCR TBM with four sub-modules was proposed considering manufacturability and post irradiation examination aspect. Various performance analyses have been performed to optimize the design. Current design and performance results including nuclear, thermo-hydraulic and thermo-mechanical analyses results are introduced.

Korea plans to install and test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in the ITER, because the HCCR blanket concept is one of options of the DEMO blanket. Currently, many design and R&D activities have been performed to develop the Korean HCCR TBM. An integrated design tool for a fusion breeder blanket has been developed based on nuclear technologies including a safety analysis for obtaining a license for testing in the ITER. A half-scale sub-module mockup of the first wall with the manifold was fabricated, and the manufacturability and thermo-hydraulic performances were evaluated. High heat load and helium cooling test facilities have been constructed. Next, the recent status of TBM material development in Korea was introduced including Reduced Activation Ferritic Martensitic (RAFM) steel, lithium ceramic pebbles and silicon carbide (SiC) coated graphite pebbles. Several fabrication methods of RAFM steel, lithium ceramic pebbles, and silicon carbide coating on graphite pebbles were investigated. Recent design and R&D progress on these areas are introduced here.

Several R&Ds are being performed for Korean helium cooled solid breeder (HCSB) test blanket module (TBM) in the field of hydrogen isotopes permeation characteristics measurement in the helium purge line, joining technologies of structural materials, breeder pebble materials development, and the measurement of pebble bed characteristics. Electron beam welding for reduced activated ferritic–martensitic (RAFM) steel is evaluated to find optimal welding conditions. Also, a hydrogen permeation measurement apparatus is newly installed for the evaluation of the permeation barrier characteristics of stainless steel and RAFM steels. Two fabrication methods of lithium orthosilicate pebbles are investigated using slurry droplet methods. As methods of silicon carbide coating on the graphite pebble, microwave coating and chemical vapor deposition coating are evaluated. Two apparatuses are established to assess the thermo-mechanical properties of graphite and breeder pebble beds. The current status of R&D activities on these areas is introduced and the main progresses are addressed in this paper.

Electromagnetic (EM) loads due to eddy current and halo current during plasma disruptions are evaluated for the ITER diagnostic upper port plug. To reduce strong EM loads acting on the port plug fixed to the vacuum vessel like a cantilever beam, three design options have been considered: removal of the diagnostic first wall, slitting of the diagnostic shield module and recess of the port plug. The main focus of the present study is to examine the efficacy of these options in terms of EM loads on the upper port plug. It is found that making slits is more effective than removing the first wall. It is also shown that the upper port plug needs to be recessed to reduce the EM load induced by halo current.

R&D activities currently being undertaken for Korean Helium Cooled Solid Breeder (HCSB) Test Blanket Module (TBM) include joining technologies of structural materials, breeder pebble material development, and measurement of pebble bed characteristics. The joining performance of ferritic–martensitic steel is evaluated. TIG welding for ferritic–martensitic steel is evaluated to find welding conditions with two different current modes. The fabrication methods of lithium silicate powder and lithium titanate pebble are under development using special process. The Li2TiO3 pebbles showing an average diameter of 1.8 mm was fabricated by water-based sol–gel method. The Li4SiO4 powder with the particle size of about 200 nm was synthesized by PVA polymer solution method. The method of silicon carbide coating on the graphite pebble is investigated, and its possibility is verified based on the coating results on the graphite plate. Sputtered SiC coating layer was crystallized by high temperature heat treatment. The measurement method and database of the thermal properties of the pebble and pebble bed are important for the design of TBM as well as breeding blanket. The initial characteristics of the pebbles are assessed measuring the effective thermal diffusivity of graphite pebble by laser flash method.

Thermo-hydraulic design of test blanket module (TBM) is challenging in that each component of TBM should be properly controlled within narrow temperature margins in the extreme conditions such as high heat flux from plasma and volumetric heat generation due to neutron irradiation. This paper presents thermo-hydraulic analysis results on the Korean helium cooled solid breeder (HCSB) TBM with the updated design. In the previous study, computation was performed on the TBM model with assumption that the coolant is uniformly distributed at back manifolds, without including the back manifolds in the computational model. In the present study, the updated configuration of the back manifolds is introduced. Flow distribution in each manifold is investigated as well as flow characteristics. Then, thermo-hydraulic analysis on the TBM model is carried out with flow distributions calculated from the flow analysis results on the back manifolds. It is found that the current manifolds are reasonably designed so that the TBM components are effectively cooled against associated thermal loads.

Korea is developing a Helium Cooled Solid Breeder Test Blanket Module for ITER. The primary candidate structural material is a Reduced Activation Ferritic/Martensitic steel. The complex TBM structure requires developing joining technologies for successful fabrication. The characteristics of Tungsten Inert Gas welding and Hot Isostatic Pressing joining of RAFM steel were investigated. Metallurgical examinations showed that the steel grain size was increased after HIP joining and recovered by post joining heat treatment. Both TIG welding and HIP joining are found to be acceptable for ITER TBM based on mechanical tests and microstructure examination.

An analysis is carried out on the three-dimensional modeling and computation of the magnetic field in ITER. The commercial finite element code ANSYS-EM is employed for this study. In particular, an emphasis is put on the analysis of the characteristics of non-axisymmetric magnetic fields produced by ferromagnetic materials, including ferromagnetic inserts (FIs) and helium cooled solid breeder test blanket modules (TBMs). It is found that the ITER design requirement for toroidal field ripple is violated by the presence of TBMs, even in the presence of FIs. Calculations of TBM-produced error fields also show that TBM produces a significant error field at q = 2 surface exceeding the ITER design requirement. Discussions are made of the potential implication of the TBM-produced non-axisymmetric fields on plasma performance and the design of a TBM emulation system.

A volume integral equation method (VIEM) is used to calculate the elastostatic field in an isotropic elastic half-plane containing circular inclusions subject to remote loading parallel to the traction-free boundary. The material of the inclusions may be either isotropic or anisotropic and they are assumed to be distributed in square or hexagonal array. A detailed analysis of the stress field at the interface between the matrix and one of the inclusions is carried out for different distances between the inclusion and the surface of the half-plane. The results of the calculations are compared with available results. The VIEM is shown to be very accurate and effective for investigating the local stresses in the presence of multiple inclusions. The method can be applied to multiple inclusions of arbitrary geometry and elastic properties embedded in extended isotropic elastic media.

R&D activities currently being undertaken for HCSB TBM include joining technologies of structural material, breeder and reflector pebble material development, the effect of TBM ferritic–martensitic steel on the ripple of toroidal magnetic field, and ceramic coating on graphite pebble. The HIP joining performance of FM steel is evaluated. Lithium ceramic breeder and graphite reflector pebble fabrication methods are under development using special fabrication process, and the initial characteristics of the pebbles are assessed. Silicon carbide coating on graphite pebble is also investigated and its preliminary results are mentioned. Finally, an accurate evaluation of the effect of TBM and ferromagnetic inserts on magnetic field are implemented. The current results of these R&D issues are addressed in this paper.

Helium Cooled Solid Breeder (HCSB) blanket is one of two blanket concepts being considered as Korean Test Blanket Module (TBM) for ITER with the aim for testing and verifying the capability of the breeding blanket concepts. R&D activities being carrying out for HCSB TBM include the development of materials, fabrication technologies, and TBM associated system design. A small sample of ferritic/martensitic (FM) steel was fabricated. It was found that the tensile strength was close to previous value. A fabrication technique of sphere pebbles of lithium titanate breeder and graphite reflector was developed. A FM/FM TIG welding was performed and the results showed that tensile strength of the welded zone was decreased about 10 %. A small punch test method of mechanical property evaluation was introduced to verify the suitability of small specimen for irradiation test by examining the relationship of the conventional uniaxial tensile test, and the tensile strengths were compared. As TBM design has complicated square channel configuration, short square channel was fabricated successfully. Finally, the components and specifications of the TBM associated systems are described in this paper.

One of the major ITER goals is test blanket module (TBM) program which is for the demonstration of the breeding capability that would lead to tritium self-sufficiency in a reactor and the extraction of high-grade heat suitable for electricity generation under the ITER fusion environment. While the engineering design of Korean helium cooled solid breeder (HCSB) TBM and its ancillary systems has been performed, a safety assessment on different possible accident scenarios should be carried out for the purpose of licensing. In this paper, accident analyses for several loss of coolant accident (LOCA) cases were performed in order to assess safety aspects of the TBM design using RELAP5/MOD3.2. Since the TBM forms a loop with helium cooling system (HCS) which is one of ancillary systems required for removing heat deposited in the TBM by neutron wall loading and surface heat flux from plasma, it is necessary to model the complete loop for accident analysis. In this study, the helium passage including the TBM and HCS was nodalized for each accident scenario. The TBM and HCS components were modeled as the associated heat structures provided by RELAP5 to include heat transfer across solid boundaries. Based on computational results it was found that current design of the TBM is robust from the safety point of view.

A volume integral equation method (VIEM) is used to calculate the plane elastostatic field in an isotropic elastic half-plane containing multiple isotropic or anisotropic inclusions subject to remote loading. A detailed analysis of stress field at the interface between the matrix and the central inclusion in the first column of square packing is carried out for different values of the distance between the center of the central inclusion in the first column of square packing of inclusions and the traction-free surface boundary in an isotropic elastic half-plane containing multiple isotropic or anisotropic inclusions. The method is shown to be very accurate and effective for investigating the local stresses in an isotropic elastic half-plane containing multiple isotropic or anisotropic inclusions.

Since the Test Blanket Module (TBM) will be operated at the wide range of elevated temperatures under cyclic thermal loading condition, thermal fatigue is one of the key life-limiting factors. The primary candidate for the structural material of TBM is the Reduced Activation Ferritic/Martensitic (RAFM) steel. A TIG welding of RAFM steel is one of key technologies for the fabrication of the TBM. Especially, the integrity of TIG welded RAFM steel under long-term thermal load condition should be guaranteed considering the TBM operation in ITER. In the previous study, a TIG welding condition for the RAFM steel was derived and main results were published. In this study, Low Cycle Fatigue (LCF) test was performed on the base metal and TIG weldment. To confirm the reliability of the TIG welded RAFM steel, the effects of Heat Affected Zone (HAZ) on LCF properties at room temperature and 550 °C were investigated.