Helium Cooled Ceramic Reflector Test Blanket Module Research Articles

Status of HCCR TBM program for DEMO blanket

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.

Investigation of technical gaps between DEMO blanket and HCCR TBM

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.

Components Design and Analysis for the HCCR TBM Preliminary Design

After Conceptual Design (CD) Approval, Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) design is being updated for the preparation of the preliminary design phase, in which the manufacturability is the main concern based on the TBM-set model of CD phase. The major design parameters at the preliminary design phase were introduced. According to these parameters, the design of the main components such as first wall (FW), side wall (SW), and the breeding zone (BZ) was updated through the thermal-hydraulic analysis with ANSYS CFX v14.5. Also, the detailed design was performed on the back manifold (BM), TBM-shield, and the connecting supports. The internal channels for the He coolant and purge gas are designed in BM, in which the characteristics of the He channel were analyzed to meet the requirements of the temperature. The design concept of the connecting supports was described. The design of the TBM-shield is updated in order to simplify production and assembly.

Numerical Investigation on Cooling Performance of Breeding Zone in HCCR TBM

The helium cooled ceramic reflector (HCCR) test blanket module (TBM) has been designed to be installed in ITER and to verify the tritium production and the heat extraction in Korea. Lithium, beryllium, and graphite are used as a breeder, a neutron multiplier, and a reflector, respectively, which called as breeding zone (BZ) including cooling plate. The BZ was operated with the highest temperature in the TBM due to the nucler heating not only in breeding material but also structure. The margin to the allowable temperature for the breeder is very small in the current conceptual design of HCCR TBM. In the present study, feasible methods were investigated to lower the maximum temperature of the BZ. The thermal resistance and the effect of each factor were studied with a conventional CFD code, ANSYS-CFX v14.5. It is found that the thermal resistance related to the pebble beds layer was main factor to determine the breeder temperature, and the installation of the cooling fins could reduce the heat transfer resistance and lower the maximum temperature of breeder about 80°C.

Accident analysis on LOCA in HCCR-TBS towards CCWS-1

Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be operated at elevated temperature with high pressure helium coolant during normal operation in ITER. One of the main ancillary systems of HCCR-TBS is Helium Cooling System (HCS) which play an important role to extract heat from HCCR Test Blanket Module (TBM) by the helium coolant to keep the operational temperature and the extracted heat is finally transferred to ITER CCWS-1 (Component Cooling Water System) by a Printed Circuit Heat Exchanger (PCHE) in the HCS. In such circumstances if Loss Of Coolant Accident (LOCA) occurs in the PCHE, the high pressure helium coolant in the primary side goes into the lower pressure water in the secondary side thus pressurizing CCWS-1. In addition, since the helium coolant contains tritium due to permeation from the TBM, tritium migrates into CCWS-1, a non-nuclear system. In this paper, accident analysis for LOCA in the heat exchanger is presented. For the analysis, GAMMA-FR code which has been developed for fusion applications was used. Main components in the HCS and CCWS-1 were modelled as volume and junctions. The accident analysis was performed for the reference case with ten channels rupture and sensitivity study was also performed by changing the crack size. The results show that pressure and tritium requirement of CCWS-1 can be met in spite of LOCA in the heat exchanger of the HCCR-TBS HCS.

Mechanical properties of ARAA steel after electron beam welding

Korea has designed a helium cooled ceramic reflector (HCCR) test blanket module (TBM) that includes a TBM shield, called a TBM set, that will be tested in ITER. The HCCR TBM is composed of four sub-modules and a back manifold. In addition, each sub-module is composed of a first wall (FW), a breeding box with a seven-layer breeding zone (BZ), and side walls with a cooling path. Korean RAFM steel was developed as a structural material for the HCCR TBM, and advanced reduced activation alloy (ARAA) was selected as the primary candidate from various program alloys. Welding technologies for fabrication of the HCCR TBM were developed using ARAA. Tensile, impact, bend tests were performed after post-weld heat treatment, and hardness, microstructure characteristics were determined before and after post-weld heat treatment to evaluate the welded specimen under the chosen welding conditions.

Evaluation of ARAA steel E-beam welding characteristics for the fabrication of KO HCCR TBM

Korea has designed a helium cooled ceramic reflector (HCCR) test blanket module (TBM), including a TBM shield, called a TBM set, that will be tested in ITER. Korean RAFM steel was developed as a structural material for the HCCR TBM, and advanced reduced activation alloy (ARAA) was selected as the primary candidate from various program alloys. Fabrication technologies for the HCCR TBM were developed using ARAA to provide the method and procedure for fabricating the TBM for testing in ITER based on RCC-MRx, which was selected as the design and fabrication code and standard for the HCCR TBM. To establish and optimize welding procedures for electron beam welding of an ARAA material, variations in welding current and speed were investigated. A series of performance tests was performed before and after post-weld heat treatment to evaluate the welded specimen under the determined welding conditions.

Development of Zr-containing advanced reduced-activation alloy (ARAA) as structural material for fusion reactors

Korea has developed an advanced reduced-activation alloy (ARAA) as a structural material for helium-cooled ceramic reflector test blanket module (HCCR-TBM) applications. The present paper describes the history of alloy development and the properties of ARAA, which has been produced at a 5t scale using vacuum induction melting and electro-slag re-melting methods. ARAA is a 9Cr–1.2W based ferritic–martensitic steel with 0.01wt.% Zr. The mechanical properties, thermal properties and physical and magnetic properties of ARAA show similar temperature dependencies to those observed for Eurofer 97. However, ARAA exhibits a much longer creep–rupture time than conventional RAFM steel, which suggests a positive effect on Zr addition. The enhanced creep strength of ARAA by the addition of Zr is attributed to the reduced temperature-dependence of the yield strength.

Tritium Transport Evolutions in HCCR TBM Under ITER Inductive Operations

First-of-a-kind numerical simulation was performed to evaluate time dependent tritium transport properties for Korea’s HCCR (Helium-Cooled Ceramic Reflector) TBM (Test Blanket Module) design under ITER inductive operating conditions. The estimation of tritium inventories in various components of the HCCR submodule and its permeation amount into the helium coolant was obtained through three computational models involving: 1) a 3D FW standalone model where diffusion and permeation into FW He coolant through tritium ion implantation was studied, 2) a 2D Poloidal-Radial (P-R) mid-plane model where the effect of increased tritium concentration in the purge gas stream was accounted for, and 3) a 2D Toroidal-Radial (T-R) mid-plane model to study tritium concentration accumulation in the He coolant. The analysis shows that tritium inventory in the breeder reaches an equilibrium value in about 10 cycles, and is about 0.373 mg per submodule. Tritium inventory in the ferritic steel structure reaches its equilibrium value in less than 10 cycles, and has about 0.0012 mg per submodule at the end of the plasma burn. The amount of the tritium permeated into helium coolant is about 1.8% of the amount of tritium produced per cycle.

Strength Evaluation of HAZ in Electron Beam Welded ARAA by Small Punch Test for HCCR TBM in ITER

The Korean helium cooled ceramic reflector (HCCR) test blanket module (TBM) has been developed for ITER, and Korean reduced activation ferritic martensitic (RAFM) steel, called advanced reduced activation alloy (ARAA), has also been developed for a structural material of the HCCR TBM. One case of limited optimized electron beam (EB) welding conditions was selected based on previous work, and the weldability of an EB weld was evaluated for TBM fabrication. The micro-hardness was measured from the base to the weld region, and the microstructures were also observed. A small punch (SP) test considering the HAZ was carried out at room and high (550°C) temperatures. The empirical mechanical properties of HAZ in the EB weld were evaluated, and the fracture behavior was investigated after the SP test. The SP results show that the estimated yield and tensile strength of the HAZ were higher than the base metal at both temperatures. A rupture occurred in the base metal region, and an elongated ductile fracture was observed on the fractured surface at both temperatures.

Electromagnetic analysis on Korean Helium Cooled Ceramic Reflector (HCCR) TBM during plasma major disruption

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.

Strength Evaluation of a Single Layer Passed TIG Weld Joint in Advanced Reduced Activation Alloy by Small Punch Test for HCCR TBM in ITER

Korea has developed a helium-cooled ceramic reflector (HCCR) test blanket module (TBM) for ITER, and Korean reduced activation ferritic martensitic (RAFM) steel, which is named ARAA (advanced reduced activation alloy), has also been developed for a structural material of the KO HCCR TBM. To evaluate the welding fabrication technology in the TBM, one case of TIG welding conditions was selected based on the previous work by Yoon et al. (2013), and a single pass with one side weld procedure through a thickness in TIG weld was carried out using ARAA, Batch 2 (F206). The microstructure was observed in the base, heat affected zone (HAZ), and weld region, and the micro-hardness was measured from the base to the weld region. In addition, a small punch (SP) test considering the base metal and HAZ region was carried out at room and high (550°C) temperatures. The empirical mechanical properties of the HAZ were estimated based on the correlation between the tensile and SP test in the base metal, and the fracture morphology was observed after the SP test.

Investigation into the in-box LOCA consequence and structural integrity of the KO HCCR TBM in ITER

Korea has developed a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) and its auxiliary system in ITER. In parallel with its design, safety analysis has performed including accident analysis with the selected reference accidents. Among them, the effect of in-box LOCA to the structural integrity of the TBM was investigated. From the transient analysis of the GAMMA-FR on the in-box LOCA, it is found that the pressure of the internal TBM can be increased up to 8MPa with the same pressure of the operating coolant through the Tritium Extraction System (TES) and He purge lines in the TBM. Structural analysis with ANSYS code for TBM was performed with this condition and it is confirmed that the TBM can endure and it does not affect the ITER machine by the failure.

Progress of Functional Components Design and Analysis of a Korean HCCR TBM in ITER

Korea has developed a helium cooled ceramic reflector (HCCR) test blanket module (TBM) for testing in a ITER, which consists of functional components to distribute the He coolant to each region such as the first wall (FW), breeding zone (BZ), side wall (SW), and back manifold (BM). In this paper, the detailed design of each component is introduced as follows: 1) FW considering cooling under a structural material temperature limit (550 °C); 2) BZ layer for obtaining tritium breeding ratio and cooling with a breeder, reflector, and multiplier pebbles; 3) SW considering the flow distribution to BZ and internal pressure; 4) BM for uniform flow to FW cooling channels; and 5) He purge line in BZ considering a purge gas distribution in BZ. From the performance analysis of each functional component using the CFD code, ANSYS-CFX with the results of nuclear heating from a neutronic analysis, the results show that the design requirements of KO HCCR TBM were satisfied.

Design and R&D progress of Korean HCCR TBM

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.

Structural components design and analysis of side wall in a Korean HCCR TBM for ITER

Korea has developed a helium cooled ceramic reflector (HCCR) test blanket module (TBM) consisting of four sub-modules in an ITER. From the draft design of the side wall (SW) according to a thermal-hydraulic analysis, a mechanical analysis was performed considering a design channel pressure of 10MPa. The SW comprised of sixteen grids with the seventeen partitions for the manifold function satisfied 1.5Sm of the allowable stress (Sm) according to RCC-MR code at the maximum stress region in the SW. In addition, an elastic analysis of the draft design of the back manifold (BM) was carried out, which supported the four sub-modules in the HCCR TBM and has the main inlet/outlet of the He cooling pipe, the measurement pipes, and He purge gas lines from the port cell. The results show that the maximum stress was higher than 1.5Sm, and the BM design has been modified to satisfy the BM function and requirements.

Development of fabrication procedure for Korean HCCR TBM

Korea has developed and plans to test a helium cooled ceramic reflector (HCCR) test blanket module (TBM) in the ITER. The HCCR TBM is composed of four sub-modules and a back manipulator (BM). Each sub-module is composed of a first wall (FW), breeding box, and side walls (SW). The fabrication procedure was developed to confirm the fabrication method for the HCCR TBM. The test specimens of the ARAA were prepared to test the weldability for tungsten inert gas (TIG) welding and electron beam (EB) welding. To establish and optimize the welding procedure in an EB weld from ARRA material, the variation in the bead width and penetration depth according to the welding current and welding speed were investigated. To verify the weldability and fabrication procedure for a complex structure such as the breeding zone, a small box with a cooling channel is being fabricated using the ARAA steel under development.

Surface coating of graphite pebbles for Korean HCCR TBM

The new concept of the recently modified Helium-Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) is to adopt a graphite reflector in the form of a pebble bed. A protective SiC coating is applied to the graphite pebbles to prohibit their reaction with steam or air as well as dust generation during TBM operation. In this research, the chemical vapor reaction (CVR) method was applied to fabricate SiC-coated graphite pebbles in a silica source. Relatively dense CVR-SiC coating was successfully formed on the graphite pebbles through the reduction of the graphite phase with SiO gas that was simply created from the silica source at 1850°C (2h). The microstructural features, XRD patterns, pore-size distribution and oxidation behavior of the SiC-coated graphite pebbles were investigated. To develop the practical process, which will be applied for mass production hereafter, a novel alternative method was applied to form the layer of SiC coating on the graphite pebbles over the silica source.

CVR-SiC coating of graphite pebbles for fusion blanket application

Abstract The Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM), recently planned for installation and testing in the nuclear fusion energy project of International Thermonuclear Experimental Reactor (ITER), uses graphite in the form of pebbles as a neutron reflector, opening the possibility of avoiding the use of a beryllium neutron multiplier. A silicon carbide (SiC) coating should be formed on the graphite pebbles to prohibit the reaction of graphite with steam and/or air and to reinforce the surface strength. In this work, a dense β-SiC coating was formed over the surface of the graphite pebbles by a chemical vapor reaction (CVR) process of the solid–solid (SS) and vapor–solid (VS) reaction of the SiO2(s)–C(s)–SiO(v)–CO(v) system. The microstructural features, XRD patterns, pore size distribution, porosity and oxidation behavior of the SiC-coated graphite pebbles were investigated. A SiC coating with a thickness of approximately 30 μm remarkably improved the oxidation resistance and the density of the graphite pebbles. In an isothermal oxidation test conducted at 700 °C, the SiC-coated graphite pebbles showed strong oxidation resistance and a weight loss of approximately 2 wt% over the course of 2 h. This paper describes the major results obtained from our experiments on the CVR-SiC coating of nuclear graphite pebbles, recently conducted in the Korean HCCR TBM team of the ITER project.

Overview of Helium Cooled Ceramic Reflector Test Blanket Module development in Korea

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.