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Neutron generators (NG) are being increasingly used in various industrial and research areas such as neutron activation analysis, neutron radiography, and neutron capture therapy. In such applications of neutron generators, compactness is one of the most important issues. Since a neutron source is generated by deuterium-deuterium (D-D) or deuterium-tritium (D-T) fusion reaction, a relatively thick shield for both fast neutrons and related photons is usually required. In this study, optimization of shielding designs for D-D and D-T neutron generators were investigated by adopting appropriate moderator and shielding materials. Based on the optimized moderator and shield thicknesses, the final dimensions of neutron generators were derived for various source strengths of D-D and D-T neutron conditions. Considering conventional condition of a container, we concluded that a 1010 n/sec D-D source and a 108 n/sec D-T source could be a portable NG.

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.

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.

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.

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.

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...

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 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 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.

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.

The design scheme and system codes for fusion application have been developed for the ITER Test Blanket Module (TBM) program in Korea in parallel with the breeding blanket development, which were based on the developed system codes in Gen. IV reactor development projects such as MARS (Multi-dimensional Analysis of Reactor Safety) and GAMMA (GAs Multi-component Mixture Analysis). Considering the unique and common features with both the Fusion and Gen. IV reactors, four approaches have been carried out: (1) modifying the heat transfer model and suggesting a 3D analysis for considering the one-sided heating with extreme temperature differences, (2) implementing a tritium permeation model for a simulation of its behavior and amount simulation in a fusion coolant system, (3) developing a physical properties generation model for PbLi and Li considering the liquid metal breeders in these codes, and (4) implementing the magnetohydrodynamics (MHD) model by Miyazaki et.al. To integrate these separate codes ...

s The nuclear desalination based on the high temperature gas-cooled reactor (HTGR) with gas turbomachinery and multi-effect distillation (MED) is attracting attention because the coupling system can utilize the waste heat of the nuclear power system for the MED desalination system. In previous work, KAIST proposed the new HTGR + MED coupling scheme, evaluated desalination performance, and performed cost analysis for the system. In this paper, in order to confirm the safety and the performance of the coupling system, we performed the transient analysis with GAMMA-T (GAs Multidimensional Multicomponent mixture Analysis–Turbomachinery) code for the KAIST HTGR + MED systems. The experimental benchmarks of GAMMA-T code were set up before the transient analysis for several accident scenarios. The GAMMA-T code was well validated against steady state and transient scenarios of the He–Water test loop such as changes in water mass flow rate and water inlet temperatures. Then, for transient analysis, the GTHTR300 was chosen as a reference plant. The GTHTR300 + MED systems were made, based on the KAIST HTGR + MED coupling scheme. Transient analysis was performed for three kinds of accidents scenarios: (1) loss of heat rejection through MED plant, (2) loss of heat rejection through heat sink, and (3) overcooling due to abnormal cold temperature of seawater. In all kinds of accident scenarios, maximum peak fuel temperatures were well below than the fuel failure criterion, 1600 °C and the GTHTR300 + MED system could be operated continuously and safely. Specially, in the loss of heat rejection through MED plants, which is related with the failure of the MED plant, the performance reduction was negligible because the reactor thermal power and the generator power decreased by only 1% and 2%, respectively. As the KAIST coupling scheme uses the double stage precooler, it turns out to have lower potential of total loss of heat rejection accidents than the single purpose nuclear power system using a single stage precooler.

The Korea nuclear industry has been developing the thermal-hydraulic system analysis Safety and Performance Analysis CodE (SPACE) and the GAs Multicomponent Mixture Analysis (GAMMA) code fo...

An air ingress accident is a major safety issue pertaining to high-temperature gas-cooled reactors (HTGRs). When a helium depressurization accident occurs, air enters the core through broken pipes. The natural circulation (NC) phase of this accident leads to massive air ingress, causing extensive oxidation of the core graphite structure. The oxidation results in three main safety issues: an increase in the core temperature, degradation of the core graphite structure, and explosive CO accumulation in the upper plenum. As a result, the onset time of the natural circulation is one of the key parameters related to reactor safety. Using GAMMA (GAs Multi-component Mixture transient Analysis) we performed a one-dimensional (D) analysis of GT-MHR in the same manner used by the conventional safety tools. A 2D analysis was also done to determine the 2D effects, including the degree of stratified-flow induced-air ingress in the horizontal ducts. The 1D and 2D simulations showed that NC occurs at around 360 h and over 500 h, respectively ( Jin et al., 2011 ). In an effort to determine the effect of the onset time of NC on safety, we performed a sensitivity study of the NC onset time when it is intentionally controlled. Due to the limitation of the 1D simulation to represent multi-dimensional stratified flows in hot and cold pipes, the 1D and 2D results are quite different. In the 1D calculation, the bottom reflector temperature is very sensitive to the onset time of NC, whereas it is insensitive in the 2D simulation. According to the 2D analysis, however, a strong-back flow from the air in the cavity is observed in the lower part of the reactor core and at the entrance of the broken pipes. As a result, the bottom reflector temperature gradually decreases in the 2D analysis irrespective of the onset time of NC. Based on these calculation results, the possibility of CO detonation and damage to the graphite structure are examined. Our conclusions are that there are large safety margins of the maximum fuel temperature, CO detonation, and damage to the graphite structure and even larger margins considering the more realistic 2D simulation.

An air-ingress accident is a major safety issue pertaining to high-temperature gas-cooled reactors. To see the effect of a stratified flow, which is a multi-dimensional phenomenon that occurs in large broken pipes, we perform 1-D and 2-D air-ingress simulations in the guillotine break of the main coaxial pipe of a 600 MWth GT-MHR with the GAs multicomponent mixture transient analysis (GAMMA) code. We used a 2-D fluid volume to build the coaxial inlet pipe, the lower plenum of the reactor core, and the cavity and simplified the other components as 1-D fluid blocks. After the guillotine break of the main coaxial pipe, the air in the reactor cavity flows into the reactor core in four phases: the blow-down phase, the stratified flow phase, the molecular diffusion phase, and the natural convection phase. In the early stage of a broken pipe, the lower plenum region of the reactor is filled with air within 30 s by a density-driven airflow. In a 1-D simulation, the process of filling the lower plenum with air ingressed from a cavity caused by the diffusion process takes 30 min. However, after 30 s, the flow velocity of air ingressed into the broken pipe decreases and the diffusion phase eventually begins. The natural circulation in this scenario starts after more than 360 h for the 1-D simulation but fails to commence after more than 500 h for the 2-D simulation. The belated natural circulation in the 2-D simulation is mainly attributed to the slower diffusion process in the core region. In turn, the slower diffusion occurs because the temperature of the air in the lower plenum is lower in the 2-D simulation than in the 1-D simulation. The maximum core temperature in the 2-D simulation was by 60 °C lower than that in the 1-D simulation.

Conventional I–S cycle flowsheets suffer from low thermal efficiency and highly corrosive streams. To alleviate these problems, KAIST has proposed the optimal operating condition for the Bunsen reaction and devised a new flowsheet that produces highly enriched HI through spontaneous L–L phase separation and simple flash processes under low pressure. A series of experiments were performed at KAIST to validate the new flowsheet and extend its feasibility. The experimental procedure, measurement method with a rich iodine condition, and results of experiments are discussed in this paper. When the molar ratio of I 2 /H 2 SO 4 in the feed increased from 2 to 4, the molar ratio of HI/(HI + H 2 O) in the HIx phase improved from 0.157 to 0.22, which is high enough to generate highly enriched HI gas through flashing. An inverse Bunsen reaction and a sulfur formation were observed when the temperature was increased from 313 K to 343 K and the molar ratio of I 2 /H 2 SO 4 was decreased from 4 to 1. 10–50 wt% of HI in the feed turned into I 2 when an inverse Bunsen reaction and a sulfur formation occurred. The experimental data utilized in the previous parametric study of KAIST has been validated.

Based on the Bunsen reaction process whose operating conditions are optimized to yield an over-azeotropic HI liquid solution, we devised a new flowsheet of iodine–sulfur thermo-chemical cycle. A highly enriched hydrogen-iodide gas can be generated through a series of processes of liquid–liquid separation of product mixture from Bunsen reaction and flash of over-azeotropic HI solution. Operating temperature and pressure for HI enrichment need not to be increased as high as those for existing flowsheets; as a result, the operating conditions become less corrosive. Chance of pipe clogging due to iodine solidification is low because there is no process where iodine is concentrated that high. Enrichment of HI through spontaneous liquid–liquid separation and simple flash processes avoiding complicated separate process is considered to be an additional benefit. Analysis of overall and component material balances showed that excess amount of feed to each process to get a desired output depends on the efficiency of flash and decomposition processes. Compared to previous ones, the proposed flowsheet requires more recirculation flows throughout the whole cycle mainly because only a portion of HI content exceeding the azeotrope is allowed to evaporate in the flash without employing a separate HI enrichment process. Thermal efficiency of the proposed flowsheet was evaluated, together with a series of parametric analyses for the sensitivity to key operating parameters and component performances. It was observed that the thermal efficiency can be raised above 60% at optimal condition.

Korea has designed a Helium Cooled Ceramic Reflector Test Blanket Module (HCCR TBM) including TBM-shield, which is called TBM-set to be tested in ITER. Thermal-hydraulic analyses are performed with conventional CFD code, ANSYS-CFX v14.5 for Electro-Magnetic Module (EM-TBM) and INTegral Module (INT-TBM) including TBM-shield, respectively. With the same model and meshes according to the ITER operation conditions of H/He and D-T phases, respectively, temperature distribution, flow rate and pressure drop are investigated to meet the design requirements and the temperature data are directly provided to mechanical analysis.

Korea plans to test a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER. Ferritic-martensitic steel (FMS) was considered a structural material, and a commercial-scale Advanced Reduced Activation Alloy (ARAA) was developed as an HCCR TBM structural material. A liquid-type TBM, i.e., a Helium Cooled Molten Lithium (HCML) TBM, is a candidate TBM for KO DEMO. In this concept, helium (He) and liquid lithium (Li) were used as a coolant and breeder, respectively. However, according to our strategy for developing a liquid breeder TBM and its more relevant DEMO concept, liquid breeders not only consider liquid lithium but also lead-lithium (PbLi). An Experimental Loop for a Liquid breeder (ELLI) was constructed for the purpose of validating the electromagnetic (EM) pump design, which we designed and fabricated ourselves, as well as for testing the effects of magneto-hydro-dynamics (MHD), and investigating the compatibility of PbLi using structural materials such as ferritic martensitic steel. Performance tests on each component such as the heaters and control systems for heating the loop were conducted, and characteristic tests using a magnet and EM pump were carried out. Corrosion tests using ELLI were conducted using grade 91 FMS steel and ARAA for a 250-hour period. In this study, a commercial FMS with grade 91 and developed ARAA steel was 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.

Korea has designed a Helium Cooled Ceramic Reflector Test Blanket Module (HCCR TBM) including TBM-shield, which is called TBM-set to be tested in ITER. In this study, the thermo-mechanical analyses are performed with the temperature data by the thermal-hydraulic analyses and inner pressure for Integral Module (INT-TBM) and EM Module (EM-TBM) including TBM-shield, respectively, which are the same model according to the ITER operation conditions. Tresca stress, strain, and deformation are investigated, and the results are confirmed that they give a satisfaction of RCC-MRx design requirement. In addition the deformation result gives a small value, which is not contacted to the TBM frame.

A new 2 MW NB (Neutral Beam) ion source for supplying 3.5 MW NB heating for the KSTAR campaign was developed in 2012 and its grid was made from OFHC (Oxygen Free High Conductivity) copper with rectangular cooling channels. However, the plastic deformation such as a bulging in the plasma grid of the ion source was found during the overhaul period after the 2012 campaign. A thermal-hydraulic and a thermo-mechanical analysis using the conventional code, ANSYS, were carried out and the thermal fatigue life assessment was evaluated. It was found that the thermal fatigue life of the OFHC copper grid was about 335 cycles in case of 0.165 MW/m(2) heat flux and it gave too short fatigue life to be used as a KSTAR NB ion source grid. To overcome the limited fatigue life of the current design, the following methods were proposed in the present study: (1) changing the OHFC copper to CuCrZr, copper-alloy or (2) adopting a new design with a pure Mo metal grid and CuCrZr tubes. It is confirmed that the proposed methods meet the requirements by performing the same assessment.

The KO HCCR TBM (Korean Helium Cooled Ceramic Reflector Test Blanket Module) was consisted of 4 sub-modules. A thermal-structural analysis of the changed FW design, which has a 15 × 11 rectangular U-shape, was carried out according to the RCC-MR code considered the design pressure (10 MPa) and thermal stress results based on the thermal-hydraulic analysis. The contact stress between the TBM sub-modules was also evaluated to confirm the structural integrity. The results showed that the design requirements were satisfied.

The ITER first wall (FW) includes beryllium armour tiles joined to a CuCrZr heat sink with stainless steel cooling tubes. The FW panels are one of the critical components in an ITER machine with a surface heat flux of 4.7 MW/m2 or above. The small-scale mockup shall be part of the qualification program and used to validate the performance of the key manufacturing technologies before the production of larger scale components, and this mockup shall utilize the hypervapotron heat sink and manufacturing processes developed for a semi-prototype design. The small-scale mockup includes 48 beryllium armour tiles (12 mm × 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 semi-prototype small-scale mockups was developed for a qualification test in Korea. For the CuCrZr and stainless steel, the canned materials are processed into an HIP (Hot Isostatic Pressing) device. In the case of beryllium to CuCrZr joining, the HIP was conducted at 580 °C and 100 MPa. For non-destructive tests of the fabricated semi-prototypes, visual and dimension inspections were performed whenever needed during the fabrication process, and ultrasonic tests were performed with ultrasonic probes. Destructive tests for the qualification semi-prototype were performed on a small mockup, which was fabricated together with semi-prototypes. The Korea heat load test facility, KoHLT-EB (Electron Beam) was constructed with an electron gun (Max. 800 kW) for a high heat flux application with a 300 kW high voltage power supply, and maximum accelerating voltage of 60 kV. This test facility was operated to evaluate the performance of these small-scale mockups. A cyclic heat flux test will be performed to evaluate the ITER qualification program.

Korea has developed a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) for testing in ITER and it consists of the functional components to distribute the He coolant to each region such as First Wall (FW), Breeding Zone (BZ), Side Wall (SW), and Back Manifold (BM). In the current study, the detailed design of each components were introduced as follows; (1) FW considering cooling under structural material temperature limit (550 °C); (2) BZ layer for obtaining tritium breeding ratio and cooling with 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 purge gas distribution in BZ. From the analysis with the CFD code, ANSYS-CFX with the results of nuclear heating by neutronic analysis, the performances of each functional components were investigated and it is confirmed that their performances were satisfied the design requirements of KO HCCR TBM.