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An advanced reduced activation alloy (ARAA) has been developed as a structural material for breeding blankets in fusion reactors. A breeding blanket has a relatively complex structure, so various welding methods, such as electron beam welding and similar and dissimilar tungsten inert gas (TIG) welding have been developed for the manufacture of breeding blankets. In this article, mechanical properties of joints of dissimilar TIG welding are introduced using ARAA and SS316L. Charpy impact, tensile, and bending tests were carried out for the as-postweld heat-treated (PWHT) samples manufactured with defined welding conditions, and hardness tests were conducted for the as-welded and the PWHT samples. The average hardness values of the dissimilar welded samples were 338 HV in the heat-affected zone (HAZ) and 199 HV in the weld metal (WM). The PWHT decreased hardness to 223 HV in the HAZ but did not affect the hardness in the WM. The yield strength (YS) and ultimate tensile strength (UTS) of the dissimilar TIG welds were 335 and 617 MPa, respectively.

In accordance with the national fusion energy program of Korea, Korea Atomic Energy Research Institute (KAERI) has established a road map for the development of a volumetric fusion neutron source (V-FNS) and begun development of a compact fusion neutron source (C-FNS) for fusion, fission, and industrial applications. Regarding the C-FNS development, KAERI has developed a radio frequency (RF) ion source inclusive of an RF driver and generator, and has designed, fabricated, and tested a target at the “Korea heat load test facility with an electron beam (KoHLT-EB)” prior to assembly with an ion source for investigating the integrity of the target under a high heat flux condition. Test conditions such as heat flux, coolant flow rate, and pressure were prepared using a preliminary analysis and later compared against test results. From the comparison, the design of the target was optimized and its thermal integrity confirmed. Thus far, KAERI has been successful in its development of a C-FNS and has begun preparations for on-site C-FNS and V-FNS designs.

KoHLT-EB (Korea Heat Load Test facility with Electron Beam gun) is operated for the performance qualification testing of the plasma facing components (PFC), ITER first wall (FW), ITER TBM (Test Blanket Module) FW, and DEMO PFCs in Korea. A thermal fatigue test shall be performed on the fabricated mockups to validate manufacturing technology, thermo-hydraulic performance and design validation using high heat load testing. The test parameters are defined along with numerically simulated conditions. For the high heat flux testing, various mockups were fabricated by each bonding technique, such as HIP (Hot Isostatic Pressing), vacuum plasma spray (VPS) coating technique. Each fabricated mockup was installed inside the vacuum chamber of KoHLT-EB, and thermo-hydraulic performance tests and thermal fatigue tests were performed to qualify the mockups specification and bonding techniques.

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.

A helium cooled liquid lithium (or lead lithium) concept has been developed to design a liquid breeder blanket in South Korea. Ferritic–martensitic steel (FMS) was selected as a structural material for fusion reactors, and a commercial-scale advanced reduced-activation alloy (ARAA) has been developed. An experimental loop for a liquid breeder (ELLI), which we designed and fabricated ourselves, was constructed to validate the electromagnetic (EM) pump as well as to test the effects of the magnetohydrodynamics of a liquid metal flow, and to investigate the compatibility of PbLi with the structural materials of FMS. Thus far, performance tests on each component, such as the heaters and control systems used for heating the loop, were conducted, and characteristic tests using a magnet and an EM pump were carried out. Corrosion tests using ELLI were conducted using grade 91 FMS steel and ARAA for a 250-h period (150+100 h). In this paper, a commercial FMS with grade 91 and developed ARAA were used for corrosion test specimens to compare the corrosion characteristics in the flowing PbLi loop. A corrosion test was conducted for investigating the compatibility of PbLi using structural materials. The effects of the oxide coatings on the prevention of the corrosion progress were also investigated.

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.

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.

The Korea heat load test facility using electron beam (KoHLT-EB) was constructed and commissioned to perform high heat flux testing on plasma facing components (PFCs), such as ITER blanket first wall, the tungsten divertor, and heat transfer experiments on other cooling devices. Also it will be used for the thermo-hydraulic performance test of ITER helium cooled ceramic reflector test blanket module in Korea. Qualification tests will be performed to evaluate the use of the high heat flux test facility on PFC and fusion reactor materials. For the thermal fatigue test of PFCs, two types of tungsten mockups were fabricated. One is a coated tungsten on the ferritic–martensitic steel (FMS). The other is a hot isostatic pressing bonded tungsten with FMS. Also, the integrity and cooling performance should be tested under the same or similar operation conditions of the ITER PFCs. After the completion of the preliminary mockup test and the facility qualification, the high heat flux test facility, KoHLT-EB, will provide the performance test for the various PFCs in fusion reactor materials and PFCs.

A helium cooling system (HCS) circulates helium coolant to remove heat from a tritium breeding blanket in a fusion reactor. A small amount of tritium can permeate into the HCS from the blanket, and it should be removed as per safety requirements in the system. A coolant purification system (CPS) is connected to the HCS, and one percent of the helium flow is bypassed into the CPS from the HCS for purification of the coolant. An ambient molecular sieve bed (AMSB) will play a role in capturing the tritiated water, Q2O. In order to develop and confirm the AMSB design and its function, a test facility is constructed especially for the experimental validation of the adsorption characteristics under a very low concentration of vapor. An experiment will be conducted under a helium flow. However, before the main helium test, air tests are performed at room temperature and 0.33 MPa. The facility can be divided into two parts, one part containing a device generating a small amount of moisture and the other containing the test section including the AMSB. The vapor generator can continuously supply vapor of 1 to hundreds of ppm to the dry gas. Various AMSBs can be installed in the test section. The adsorption performance and saturation characteristics according to the diameter, length, and flow rate are compared. The experiment was conducted on two columns of different lengths. Specifically, when the performance validation test with the 64 mm AMSB was conducted over a period of more than 10 days, it showed very stable adsorption performance. The length of the AMSB estimated in the experiment was several times shorter than that obtained through a correlation equation. If these results are reflected in the design, the design can be more compact or the operation cycle can be increased. These experiments can contribute to the development of AMSB designs.

Korea has developed a helium cooled ceramic reflector test blanket module (TBM) for fusion reactors. An advanced reduced activation alloy has been developed as a structural material and various joining methods such as E-beam welding, tungsten inert gas (TIG) welding, laser welding, and hot isostatic pressing have been applied to establish its fabrication method and procedure. Based on the current design of the TBM and breeding blanket, TIG welding was selected as the main joining method, and the U-shaped first wall (FW) was adopted. For welding of the breeding blanket, a special joining technique without a permanent backing strip with simultaneous full penetration welding in accordance with the design rules of class N2RX of RCC-MRx is required for the pressure vessel. For the U-shaped FW, distortion of the drilled cooling hole during the bending should be limited. Three small mock-ups were fabricated in this study to investigate these issues. The joints and bending parts were visually observed after cutting, and the hydrostatic pressure test was performed for regions with low thickness. The results show that the proposed fabrication methods are applicable in the fabrication procedure development for the breeding blanket.

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 is performed according to the selected Load Combinations (LCs), which is described in the TBM Port Plug (TBM PP) System Load Specifications. Single load analyses are performed such as dead weight (DW), operating pressure (PresO), operating temperature (THO), electromagnetic (EM; MD-I, MD-II, and MD-IV), seismic (SL-1, SL-2, and SMHV), and in-TBM LOCA-IV. In addition, their results are superposed for the selected LCs. Through an analysis with 14 single load cases and 10 LC cases, it is confirmed that all load combination results meet the design criteria from the stress breakdown analysis according to the RCC-MRx.

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.

Joining technology for dissimilar materials was developed for the fabrication of an ITER blanket first-wall, which consisted of Be, CuCrZr, and stainless steel (SS). The Be/CuCrZr/SS joint was fabricated using a hot isostatic pressing (HIP) method. Beryllium armor was joined to the CuCrZr/SS block at 580 °C under 100 MPa. The optimal interlayer coatings of Cr/Cu and Ti/Cr/Cu were developed using an ion-beam assisted physical vapor deposition. Beryllium is also a candidate armor material for the TBM first-wall. Successful joining of Be to ferritic-martensitic steel (FMS) was accomplished using an HIP method by introducing the thick diffusion barrier. A thick diffusion barrier of a Cu foil(10 μm) limited the excessive diffusion and prevented the formation of brittle phases at the Be/FMS interface. Be and FMS were bonded at 650–850 °C; however, a temperature of lower than 750 °C was recommended to avoid material degradation of FMS. In addition, the joining of W to FMS has been developed. Tungsten is another armor material applicable to more severe plasma conditions. The large difference in the thermal expansion between W and FMS was resolved by introducing the Ti interlayer and Mo separator. Moreover, the double-staged HIP (the first stage at 900 °C and 100 MPa and the second stage at 750 °C and 70 MPa) was applied to suppress the edge delamination of W/FMS joints during thermal history.

Using the conceptual design of the Korean helium cooled ceramic reflector (HCCR) test blanket module (TBM) including the TBM-shield for testing in ITER, a seismic analysis is performed. According to the ITER TBM port plug (TBM PP) system load specifications, seismic events are selected as SL-1 (seismic level-1), SL-2 (seismic level-2), and SMHV (seismes maximaux historiquement vraisemblables, Maximum Histroically Probable Earthquakes). In a modal analysis a total of 50 modes are obtained. Then, a spectra response analysis for each seismic event is carried out using ANSYS based on the modal analysis results. For each event, the obtained Tresca stress is evaluated to confirm the design integrity, by comparing the resulting stress to the design criteria. The Tresca strain and displacement are also estimated for the HCCR TBM-set. From the analysis, it was concluded that the maximum stresses by the seismic events meet the design criteria, and the displacements are lower than the designed gap from the TBM PP frame. The results are provided to a load combination analysis.

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.

The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated, and an integrity and thermo-hydraulic performance test should be completed under the same or similar operation conditions of ITER. The test plan for a thermo-hydraulic analysis was developed by using a high heat flux test facility, called the Korean heat load test facility by using electron beam (KoHLT-EB). This facility is utilized for a qualification test of the plasma facing component (PFC) for the ITER first wall and DEMO divertor, and for the thermo-hydraulic experiments. In this work, KoHLT-EB will be used for the plan of the performance qualification test of the ITER HCCR-TBS mockups. This qualification tests should be performed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization (IO), which describe the specifications and qualifications of the heat flux test facility and test procedure for ITER PFC.

With a conceptual design of Korean helium cooled ceramic reflector test blanket module (TBM), including TBM-shield for testing in ITER, thermo-mechanical analyses are performed with the temperature data through thermal-hydraulic analyses and the inner pressure for the electromagnetic (EM)-TBM and INTegral (INT)-TBM, including a TBM-shield, which are the same models according to the ITER operation conditions. The Tresca stress, strain, and deformation are investigated, and the results confirm that they satisfy the RCC-MRx design requirements. In addition, the deformation results give a small value, which is not in contact with the TBM frame.

With a conceptual design of Korean helium-cooled ceramic reflector test blanket module (TBM), including TBM shield for testing in ITER, thermal-hydraulic analyses are performed with a conventional CFD code, ANSYS-CFX v14.5, for the electromagnetic module and the integral module, including the TBM shield. With the same model and meshes, according to the ITER operation conditions of the H/He and D-T phases, the temperature distribution, flow rate, and pressure drop are investigated to meet the design requirements, and the temperature data are directly provided to a mechanical analysis.

In the design of a fusion reactor, a helium cooling system (HCS) is used to extract heat from the blanket. To reduce the permeated tritium to HCS, coolant purification system (CPS) has been developed and experimental loop has been constructed. In the present paper, the main components of the CPS loop such as oxide bed and absorption bed were introduced focusing with design, fabrication and their performance calculation. Since the molecular size of tritium is too small to be captured, it is oxidized to Q2O using a copper oxide bed. Then, the adsorption bed using molecular sieve captures Q2O through physical adsorption. For the design of CPS, correlation of previous studies and general properties of molecular sieve were used. Since the concentration of Q2 in this system is estimated as 0.4 Pa, which is very low compared to the helium pressure of 8 MPa, the used correlations for design and saturation characteristics of the absorption bed under low partial pressure of Q2O should be validated. In the present study, component test was prepared to verify the designed absorption bed, which can simulate the helium flow condition with low concentration of Q2. The test was performed under various conditions such as temperature, pressure, concentration of Q2 and velocity. The results will be used for CPS loop design.

A helium-cooled ceramic reflector (HCCR) blanket was studied as a candidate breeder blanket for the Korean DEMO reactor. A detailed three-dimensional neutronics model based on a K-DEMO model and a simplified thermal-hydraulic model were established for the HCCR-DEMO blanket conceptual design analysis. The insertion of a graphite reflector resulted in more than 33 % reduction of the required total beryllium multiplier volume, while its influence on the tritium breeding ratio (TBR) was evaluated to be less than 2 %. The final overall TBR of the HCCR-DEMO blanket was estimated as being more than 1.15 even after operation of 20 effective full power years (EFPYs) without replacement.

A hypervapotron mockup was fabricated and used in the Neutral Beam Injection (NBI) heating system at KAERI. During the test, especially with the JAEA ion source, which has a focused beam and a long...

The characteristics of thick tungsten layers coated onto ferritic–martensitic (FM) steel blocks using a vacuum plasma spray system were investigated. Because of the large differences in thermal properties between tungsten and FM steel, it is difficult to coat a thick tungsten layer onto FM steel without delamination or detachment of the coated layer from the substrate. To avoid detachment of the tungsten layer from the FM steel surface, the optimization of the coating process was studied in this work. The tungsten layer having a 3.7 mm thickness was coated using 43 kW DC torch power at 16 kPa of chamber pressure with a mixture of argon, hydrogen, and helium. The adhesion between the coating layer and the FM steel was enhanced by inserting pre-heating process before coating process, which increased the penetration of molten tungsten droplets into the FM substrate. The hardness of the coating layer was about 380–400 HV, which was 80 to 95% of the hardness levels for commercial bulk tungsten monoblocks, and the porosity, purity, and morphology properties of the tungsten coating layer were also discussed. To evaluate the homogeneity of the layer bonding between the FM steel and the tungsten coating layer, an ultrasonic test was also performed.

For the application to plasma facing component (PFC) in a fusion reactor, joining methods between tungsten (W) and ferritic-martensitic steel (FMS) have been developed and three W/FMS mockups were fabricated by HIP (hot isostatic pressing) joining method. Because the high heat flux test should be performed over the thermal lifetime of the mockup to confirm the integrity of joining technology, test conditions are found by performing a thermal-hydraulic and thermo-mechanical analysis with the conventional codes such as ANSYS-CFX and ANSYS-mechanical, respectively. From the analysis, the heating and the cooling conditions are determined, for 1.0 MW/m2 heat flux, to be 30 s heating and 30 s cooling with given test facility cooling system. And the test cycle number for thermal-fatigue testing is determined to be 2500 cycles because the estimated thermal-lifetime of the mockup is about 2324 cycles from the results of elastic-plastic analysis. The high heat flux test with KoHLT-EB will be performed with these test conditions in the near future.

To develop tritium breeding technology for a Fusion Reactor, Korea has participated in the Test Blanket Module (TBM) program in ITER. The He Cooled Ceramic Reflector (HCCR) TBM consists of functional components such as First Wall (FW), Breeding Zone (BZ), Side Wall (SW), and Back Manifold (BM) and it was designed based on the separate analyses for each component in 2012. Based on the each component analysis model, the integrated model is prepared and thermal-hydraulic analysis for the HCCR TBM is performed in the present study. The coolant flow distribution from BM and SW to FW and BZ, and resulted structure temperatures are obtained with the integrated model. It is found that the non-uniform flow rate occurs at FW and BZ and it causes excess of the design limit (550 °C) at some region. Based on this integrated model, we will perform the design optimization for obtaining uniform flow distribution for satisfying the design requirements.

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 act...

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 10 MPa. The SW comprised of sixteen grids with the seventeen partitions for the manifold function satisfied 1.5 S m of the allowable stress ( S m ) 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 1.5 S m , and the BM design has been modified to satisfy the BM function and requirements.

Korea has designed a Helium-Cooled Ceramic Reflector (HCCR)-based Test Blanket System (TBS) for International Thermonuclear Experimental Reactor (ITER). Among seven selected reference accidents in Korean TBS, in-box loss of coolant accident (LOCA) is one of them. This is initiated by a double-ended break of the coolant pipe in the Breeding Zone (BZ), pressurizing the BZ box structure, causing pressurization of the Tritium Extraction System (TES) and purging of pipelines. When the accident is detected, the Plant Safety System (PSS) isolates the Helium Cooling System (HCS) and TES, and requests plasma shutdown to Fusion Power Shutdown System (FPSS). To prevent aggravating failure of the system, the safety function is automatically activated when the accident is detected, the device being the isolation valve of HCS and TES. One important observation of this accident is that instant isolation is not a good measure to take. In terms of the possibility of aggravating failure, system isolation is an important safety procedure but isolated TES volume is exposed to high pressure and temperature conditions in the early move of the accident transient. The result of system safety analysis shows that delayed isolation keeps the system safe for a while. In this article, given the preliminary accident analysis results for the current HCCR TBS, case studies were performed regarding the delayed isolation timing effect. For this transient simulation, Korean nuclear fusion reactor safety analysis code (GAMMA-FR) was used.

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 8 MPa 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.

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.

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.

The Korean Helium Cooled Ceramic Reflector Test Blanket Module (TBM) consisted of four submodules for International Thermonuclear Experimental Reactor program. The thermal-structural analysis considered a design pressure (10 MPa) and a thermal stress in the changed first wall design, which has a 15 mm \(\times 11\) mm rectangular U-shape, was carried out according to the RCC-MR code. The contact stress between the TBM submodules was also evaluated to confirm the structural integrity. The results showed that the design requirements were satisfied.

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 $^{\circ}{\rm C}$ ); 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.

The ITER first wall (FW) includes beryllium armor tiles joined to a CuCrZr heat sink. The FWs are one of the critical components in an ITER machine with a surface heat flux of 4.7 MW/ ${\rm m}^{2}$ or above. The small-scale mockup shall be a part of the qualification tests and used to validate the performance of the dominant manufacturing technologies before the production of larger scale components, and this mockup shall be equipped with a hypervapotron heat sink and manufacturing processes developed for a semiprototype design. The small-scale mockup includes 48 beryllium armor tiles (12 mm $\times\,$ 12 mm) capable of withstanding the specified heat flux values. The tile thickness shall be 6 mm to minimize the beryllium surface temperature and evaporation under high thermal loads. The detailed fabrication process of semiprototype small-scale mockups was developed for a qualification test in Korea. For the CuCrZr and stainless steel, the canned materials are processed into an hot isostatic pressing (HIP) device. In the case of beryllium-to-CuCrZr joining, the HIP was conducted at 580 $^{\circ}{\rm C}$ and 100 MPa. For nondestructive tests of the fabricated semiprototypes, visual and dimension inspections were performed whenever needed during the fabrication process, and ultrasonic tests were performed using ultrasonic probes. Destructive tests for the qualification semiprototype were performed on a small-scale mockup, which was fabricated together with semiprototypes. The Korea heat load test facility using an electron beam system was constructed with an electron gun (maximum electric power of 800 kW) for a high heat flux application with a 300-kW high-voltage power supply and maximum accelerating voltage of 60 kV. This facility was operated to evaluate the performance test of plasma facing components. A cyclic heat flux test will be performed to evaluate the ITER qualification program.

Korea has designed a helium cooled ceramic reflector (HCCR) based test blanket system (TBS) for an ITER. An in-vessel loss of coolant accident is one eight selected reference accidents in the Korean TBS. This accident is initiated by a single or multiple rupture of the test blanket module first wall cooling channels, causing a plasma disruption, and pressurization of the vacuum vessel (VV). In this type of accident, the governing parameters are various, for example, the operating pressure, gas temperature, TBS volume, VV volume, and mass flow rate. Thus, a scoping study is an essential strategy when attempting to determine the proper design specification for a Korean TBS. In this paper, given the preliminary accident analysis results for the current HCCR TBS, a parametric study was performed. For this transient simulation, the Korean nuclear fusion reactor safety analysis code (GAMMA-FR) was used.

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.

Korea has developed a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) testing in ITER, which was considered one of the fusion DEMO-relevant blankets in Korea. The design and performance analysis of the TBM body have been carried out considering the uniqueness of the KO TBM and design requirements by the IO and KO design concept: (1) KO TBM has 4 sub-modules considering a post irradiation test (PIE) and its delivery. (2) A first wall (FW) design was changed into a 15 × 11 rectangular shape and its performance was confirmed by thermal-hydraulic and thermo-mechanical analyses using commercial ANSYS code. The results showed that the revised design model satisfied 1.5 S m and 3 S m of the allowable stress ( S m ) in the RCC-MR code at the maximum stress region of the components for mechanical and thermo-mechanical analyses, respectively. (3) Considering the tritium breeding and cooling, a breeding zone (BZ) design was investigated. Three Li and Be layers, and one graphite layer, were proposed by the iteration, and the appropriate temperature distribution was obtained. The design for other components such as a side wall (SW) and back manifold (BM) is on-going considering 9 MPa of channel pressure and its functions of flow distribution as a manifold.

A scaled-down Helium Supplying System (HeSS), which is linked with an electron beam heat load facility, has been constructed in Korea. HeSS is designed to supply a high temperature and high pressure helium gas flow into the first wall mock-ups of the HCCR TBM. The electron beam facility (KoHLT-EB) is connected with HeSS to apply a high heat load (up to 5 MW/m2 at 300×200 mm2) to the first wall mock-up target. A heat load test, with a constant heat flux of 0.3 to 0.5 MW/m2, with the first wall mock-up is scheduled under inlet conditions of 8 MPa, 300 °C, and a 0.5 kg/s helium flow rate, which is based on the operating condition of HCCR in 2013.

Korean high heat flux test facility for the plasma facing components of nuclear fusion machines will be constructed to evaluate the performance of each component. This facility for plasma facing ma...

Since the decision of blanket redesign by 2007 ITER design review, the blanket system is being developed in the framework of Blanket Integrated Product Team (BIPT) composed mainly of ITER Organizat...

In the ITER tokamak, the toroidal magnetic field (TF) ripple is estimated with TF coils only, with the installation of ferromagnetic inserts (FIs), and with test blanket modules (TBMs) by using a 2-D code for easy and fast calculation. We assessed the effects of the thickness of the FIs on the TF ripple in order to optimize the FI. And we analyzed how the TBMs distort the TF, and calculated the TF ripple for various amounts of a ferromagnetic material and the positions of the TBMs. Even in the case of moving the TBMs outward up to 60-cm, and reducing the ferromagnetic material to 52%, the TF ripple is not decreased below 0.38%. So we had to adopt ripple correction coils. With a 52% reduced amount of the ferromagnetic material in a TBM, we could reduce the TF ripple to 0.28% at a coil current of 100 kA turn per each coil. And with an outward recess of the TBM up to 60 cm, we could reduce the TF ripple to 0.23% at a coil current of 250 kA turn per each coil. As a combined approach, if we reduce the amount of a ferromagnetic material in a TBM to 30%, and recess the TBM to 15 cm, we can efficiently obtain the TF ripple of 0.25% at a coil current of 150 kA turn per each coil.

Korea Domestic Agency (KO-DA) was responsible for the procurement of the ITER blanket modules 1, 2, and 6 in the original procurement allocation. According to the procurement reallocation of the blanket system, Korea will procure the blanket shield block in place of the blanket first wall. Nevertheless, several R&D activities in Korea have been performed including optimization of the hot isostatic pressing (HIP) bonding process between Be/CuCrZr and CuCrZr/SS, the nondestructive test method, fabrication feasibility study, high heat flux tests, and design analysis. Especially, KO-DA participated in the qualification program for the mock-up manufacture and high heat flux tests. Several mock-ups were fabricated and tested during the qualification program. The details of the mock-up manufacture and test results are described in this paper. Also, two heat flux facilities were installed based on the graphite heating, and a new electron beam heat flux facility will be built in the near future for the enh...

A Korean high heat flux test facility for the semi-prototype (SP) qualification of an ITER first wall (FW) will be constructed to evaluate the fabrication technologies required for the ITER FW, and the acceptance of these developed technologies will be established for the ITER FW manufacturing procedure. Korea participated in this qualification program, and is responsible for suitable arrangements for the heat flux test of our fabricated SPs. Qualification testing can be started provided that adequate quality and control measures are implemented and validated by the ITER Organization (IO). The controlling measures required for all heat flux tests shall be concrete and demonstrate the satisfaction of the IO test programs. Each country shall provide a test plan covering the quality and controlling measures in the high heat flux test facility to be implemented throughout the test program. Korean high heat flux testing for these ITER plasma facing materials will be performed by using a 60 kV electron beam and a power supply system of 300 kW, where the allowable target dimension is 70 cm × 50 cm in a vacuum chamber. In addition, this facility needs a cooling system for a high-temperature target and decontamination system for beryllium filtration.

Korea has developed a liquid breeder blanket for the test blanket module (TBM) program in ITER with a helium-cooled molten lithium concept. Since ferritic martensitic steel is used as the structura...

A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5–1.0 MW/m 2 . Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and circulator, a 9 MPa He loop was constructed, and it supplies high temperature (500 °C) and pressure (8 MPa) He to the high heat flux test facility. For an electromagnetic (EM) pump development for circulating the liquid breeder, magnetohydrodynamic (MHD) experiment, and flow corrosion test, a PbLi breeder loop was constructed. From the performance test, the EM pump and magnet show their capability, and flow and static corrosion tests including oxide coating for corrosion protection were performed. For tritium extraction from the liquid breeder, a gas–liquid contact method was adopted and a tritium extraction chamber was constructed. For measurement of the tritium amount in the liquid breeder, permeation sensors have been developed.