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This paper describes the manufacturing progress of ITER Vacuum Vessel Thermal Shield (VVTS) sectors, which are on-going since the start of material buffing in October 2014. Fabrication of VVTS proceeds according to the following main processes: 1) plate cutting, 2) bending and forming, 3) welding, 4) flange final machining, 5) pre-assembly of 40 ° sector, 6) silver coating and 7) final acceptance test. All VVTS shell segments are to be assembled by the flange joints, which are welded to the shells. Two kinds of inspection methods are presented for the cooling pipe welding: endoscope and leak test. A specific endoscope is developed for a long welded cooling pipe. Vacuum leak test is performed in the test vacuum chamber with the helium pressurization of cooling pipe by 3 MPa, which is the same pressure differential with the operating condition of the VVTS. One of the dimensional inspection, 3D laser scanning is also described to see the effect of shape correction for the flange to shell welding. Silver coating jig design has been carried out, focusing on the electrode position. Structural analysis results are shown for the design of pre-assembly jig of 40 ° sector. Finally near term schedule of the manufacturing is summarized.

Summary This overview discusses the status of fusion energy development in the Republic of Korea. Korea studies fusion energy through research and development in 2 ways: one by performing the Korea Superconducting Tokamak Advanced Research (KSTAR) for a domestic approach and the other by participating in the international collaboration project ITER. Korea had remarkable progresses not only in KSTAR but also in the ITER project in 2016. In KSTAR, a world record of more than 70 seconds in high-performance plasma (H-mode) operation in the superconducting tokamak has been achieved. For the ITER project, toroidal field conductors have been successfully accomplished and the other major equipment such as vacuum vessel sectors, ports, and assembly tooling seem to be reaching an end. This paper deals with current research and development results on both fields of KSTAR study and ITER project progress.

Korea is operating 24 nuclear power plants and a highly advanced neutron application reactor HANARO (High-flux Advanced Neutron Application Reactor). In addition, Korea is designing a tritium storage and delivery system (SDS) for ITER. We have been developing detritiation and tritium storage technologies since the operation of Wolsong CANDU (Canada Deuterium-Uranium) station in 1983. The Wolsong Tritium Removal System (TRF) was designed to remove tritium generated in heavy water of the moderator and heat transport. Catalysts transfer tritium from the tritiated heavy water to gaseous tritiated deuterium. The hydrogen isotopes, including tritium, are transported to a cryogenic distillation system where the tritium is removed for safe storage. Conventional high-pressure storage tanks can be dangerous for the storage of radioactive tritium gas. We have been studying various kinds of metal hydride, such as titanium, zirconium cobalt, and depleted uranium. Titanium was proven to store tritium safely and...

Control of dehydriding (desorption of hydrogen or hydrogen isotope) rate from a hydride bed in fusion fuel cycle is one important design point to estimate a real supplying amount of hydrogen from the hydride bed. In a real system tens of batch-type hydride beds are to be utilized for supplying a certain amount of hydrogen isotope at the same time. A study on efflux time from a hydraulic water tank was applied as a fundamental similarity test of the gas fueling system. As a result, liquid efflux from a tank shows a similar behavior with the desorption pattern of the depleted uranium hydride bed system. As much important as keeping a hydride buffer vessel pressure in a hydride bed system, similar tendency was studied in the tank efflux system; i.e., to keep the secondary vessel height there needs a certain amount of liquid flow from the upper tank and the tank height difference. From one tank with connected another tank flow with understanding of tank efflux model a complicated multi-tanks behavior could be understood by simulating its complex efflux characteristics, and it is likely to be applied to the multi-hydride beds system.

Major progress in tritium research in the Republic of Korea began when Korea became responsible for ITER tritium Storage and Delivery System (SDS) procurement package which is part of the ITER Fuel Cycle. To deliver the tritium SDS package, a variety of research institutes, universities and industry have respectively taken roles and responsibilities in developing technologies that have led to significant progress. This paper presents the current work and status of tritium related technological research and development (R&D) in Korea and introduces future R&D plans in the area of fuel cycle systems for fusion power generation.

A tritium plant for nuclear fusion power plants consists of an SDS (Storage and Delivery System), an ISS (Hydrogen Isotope Separation System), a TEP (Tokamak Exhaust Processing system), and an ANS (tritium plant Analytical System). Korea has been developing an SDS. The main purpose of this tritium storage and delivery system is to store and supply the D-T gas needed for DT plasma operation and to provide the necessary infrastructure for short- and long-term storage of large amounts of tritium. We have been developing tritium storage beds for the SDS.The primary role of the metal hydride beds in the SDS is to store and supply D-T fuel during DT plasma operation. ZrCo and depleted uranium (DU) have been extensively studied. Compared to the use of ZrCo, which is disproportionate at temperatures of higher than 350°C, DU hydride can be heated up to very high temperatures sufficient to pump hydrogen isotopes without using gas compressors. Our experimental apparatus used to test the experimental DU bed consists ...

Consecutive absorption/desorption cycles of the ZrCo–H2 system were studied to simulate the real International Thermonuclear Experimental Reactor (ITER) hydrogen getter system. A ZrCo getter was used in this paper instead of the depleted uranium (DU) getter material, which has been recently considered as the hydrogen getter in ITER. In a cyclic pressure-composition isotherm (PCI) measurement, the high-pressure Sievert apparatus seems impractical to describe the equilibrium state of the ZrCo–H2 system in detail, especially for the desorption stage. This high-pressure PCI apparatus, however, shows cause and effect well, from the previous getter state to the following state in presenting hydriding/dehydriding performance. In case of the ZrCo–H2 system or in case of the DU–H2 system having multiple getter bed battery, a similar affection by previous getter status might be related and a similar aspect could be shown to consider further ITER design; for example, a need for control logic from PCI measurements using a high-pressure Sievert apparatus.

Based on the preliminary design of the ITER blanket shield block (SB) #8, the full scale prototype (FSP) has been manufactured and tested in accordance with pre-qualification program, and related R&D was performed to resolve the technical issues of fabrication. The objective of the SB pre-qualification program is to demonstrate the acceptable manufacturing quality by successfully passing the formal test program. 316L(N)-IG stainless steel forging blocks with 1.80L × 1.12W × 0.43t (m) were developed by using an electric arc furnace, and as a result, the material properties were satisfied with technical specification. In the course of applying conventional fabrication techniques such as cutting, milling, drilling and welding of the forged stainless steel block for the manufacturing of the SB #8 FSP, several technical problems have been addressed. And also, the hydraulic connector of cross-forged material re-melted by electro slag or vacuum arc requires the application of advanced joining techniques such as automatic bore TIG and friction welding. Many technical issues – drilling, welding, slitting, non-destructive test and so on – have been raised during manufacturing. Associated R&D including the computational simulation and coupon testing has been done in collaboration with relevant industries in order to resolve these engineering issues. This paper provides technical key issues and their possible resolutions addressed during the manufacture and formal test of the SB #8 FSP, and related R&D.

Korea has been operating a CANDU nuclear power plant since 1983. Tritium generated in the heavy water of the plant is removed by the Wolsong TRF (Tritium Removal Facility) and measurement campaigns of tritium near the power plant have shown the efficiency of the TRF system. The HANARO reactor uses heavy water as both reflector and moderator. In HANARO the tritiated water removal system consists of compressors, condensers, and adsorption beds. A tritium behavior analysis code (TRIBAC) for a Very High Temperature Gas-Cooled Reactor (VHTR) is under development at KAERI. The TRIBAC computer software has been equipped with models for tritium production, purification, and leakage, as well as chemisorption and tritium behavior, in the hydrogen production system. Korea takes part into the ITER program and is responsible for the supply of an SDS (Tritium Storage and Delivery System). Within this program Korea has launched an experimental program to study the physico-chemical properties of metal and their hydrides in which hydrogen isotope gases can be stored and removed safely.

Korea has been developing nuclear fusion fuel storage and delivery system (SDS) technologies including a basic scientific study on hydrogen storage. To develop nuclear fusion technology, it will be...

The main objective of ITER tritium Storage and Delivery System (SDS) is contracted to develop an optimal metal hydride bed that can be reveal the unprecedented fueling performance for the Tokamak. One function of the hydride bed is to keep safety requirements in terms of confinement of tritium. The hydride material for storing the deuterium and tritium fuelling gases is being made narrow with depleted uranium (DU) by its good performance. DU also has its own uncertainties, however, in applying it to realize the getter bed system having an all-round capability, especially in aspect of safety. This paper deals with from bed design target to the design variables in terms of comparison of risk-based multi-criteria using HAZOP (risk matrix) analysis. In analysis of the risks, important variables that denotes safety-effective, or cost-effective, or maintainability-effective, or manufacturability-effective are sometimes mutually interrelated with each other. As a conclusion the authors could recommend the way to concentrate and minimize the bed design variables with most meaningful risk-containing components that can be applied to increase the performance of hydride bed. It needs, however, that further study of comparison of risk analyses should be proceeded to complete the hydride bed design.

The storage and delivery system bed provides rapid and safe functions for storage and delivery of tritium gas in a fusion power plant. In order to investigate the hydrogen isotope effect, we prepared a small-sized (1/10) ZrCo bed. Based on hydrogen-deuterium gas mixtures with various composition 0, 25, 50, 75 and 100 volume %, we observed the hydriding and dehydriding dynamic behaviors. In the case of hydriding performances at room temperature, 100% hydrogen gas provided the fastest delivery rate among the mixtures. In the case of the dehydriding performance at temperatures below 350 °C, however, there was no significant difference among the mixtures. Vacuum driving can provide a stronger effect for this apparatus than various hydrogen compositions.

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.

With the development of fusion technology, it will be necessary to store large amounts of tritium during the nuclear fusion fuel cycle. Stable metal tritides are viewed as potential candidates for the high-density storage of tritium. Metal tritide formers offer a safe and convenient method for tritium storage. For the storage, supply, and recovery of hydrogen isotopes, zirconium cobalt (ZrCo) and depleted uranium (DU) have been extensively proposed. Thus, we have designed and fabricated two identical small-scale getter beds for a comparison of ZrCo with DU on the hydriding/dehydriding properties. After the powderization of the metals, the hydriding/dehydriding performance at different stoichiometries of ZrCo and DU was measured. We provide preliminary experimental results of our ZrCo and DU beds.

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.

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...

Korea Domestic Agency (KODA) is developing a nuclear fusion fuel storage and delivery system (SDS) as one of the Korean procurement packages. Korea Atomic Energy Research Institute (KAERI) is operating the following basic scientific research laboratories for an SDS and tritium supply study: a metal hydride bed preparation laboratory, hydrogen isotope recovery and delivery performance test rig, in-bed calorimetry performance test rig, and tritium shipping container integrity test facility. Furthermore, the development of a test blanket module (TBM) is required to test and validate the design concept of tritium breeding blankets relevant to fusion power plants. KAERI is also operating the following laboratories for the TBM research, such as a tritium extraction performance test rig, High-flux Advanced Neutron Application Reactor (HANARO), and Experimental Loop for Liquid Breeder (ELLI).

국제핵융합실험로(ITER)가 2019년까지 7개국의 공동개발사업으로 건설될 예정이다. ITER의 핵융합연료주기는 핵융합진공용기, 삼중수소 플랜트, 연료공급부로 구성되어 있다. 이중에서 삼중수소 플랜트는 핵융합연료주기를 위한 중 수소와 삼중수소의 저장, 공급, 분리, 제거, 회수 등의 기능을 제공한다. 삼중수소 플랜트는 외부에서 중수소와 삼중수소를 공급받아 저장 공급하는 SDS, 토카막배출처리의 TEP, 수소동위원소 분리의 ISS, 삼중수소수 및 대기 처리의 WDS ADS, 정성 정량분석의 ANS 등으로 구성된다. 이 논문에서는 삼중수소 플랜트를 구성하는 주요 공정에 대한 기능 및 설계요건을 기술하였다. 한국은 SDS 개발에 참여하고 있으며 월성원전 삼중수소 제거설비(WTRF) 건설 및 운전경험을 통해 WDS 대한 기술을 일부 확보하였다. 향후 ISS 및 TEP에 대한 기술확보를 위한 여러 분야에서의 참여 확대를 기대하고 있다. 【International Thermonuclear Experimental Reactor (ITER) will be constructed in 2019 according to the JIA (Joint Implementation Agreement) of 7 countries. The ITER fusion fuel cycle consists of fusion vacuum vessel, tritium plant and fuelling system. The tritium plant provides the functions of storage, delivery, separation, removal and recovery of the deuterium and tritium used as fusion fuels for the ITER. The tritium plant systems supply deuterium and tritium from external sources and treat all tritiated fluids from ITER operation through Storage and Delivery System (SDS), Tokamak Exhaust Processing (TEP), Isotope Separation System (ISS), Water Detritiation System & Atmosphere Detritiation System (WDS & ADS) and Analysis System (ANS). In this paper, the functions and design requirements of the major systems in the tritium plant and the status of R&D are described. Korean party is developing the SDS for ITER tritium plant and partially attaining the WDS technology through the construction and operation experience of the Wolsong Tritium Removal Facility (WTRF). Now it is expected that researchers in other fields such as chemical engineering take part in the development of upcoming technologies for ISS and TEP.】

Since the recommendation of blanket redesign by 2007 ITER design review, the blanket system has been developed in the framework of Blanket Integrated Product Team (BIPT) composed mainly of ITER organization (IO) and procuring parties. After the completion of blanket Conceptual Design Review (CDR), the design teams were organized to efficiently implement the preliminary and detailed design in a timely manner. The blanket shield block is a bulk structure to provide the nuclear shielding; therefore it should be designed to accommodate the nuclear heating in parallel with the mitigation of electromagnetic loading by slit optimization. This paper briefly describes the cooling optimization for the preliminary design of ITER blanket shield block 01, 02, 08 and 16. Three-dimensional hydraulic and thermo-mechanical analyses for the preliminary design modules of hybrid structure with poloidal and radial cooling concepts are performed under the inductive operation as a representative loading condition. The pressure drop, heat transfer and coolant uniformity in cooling passages are investigated in detail. In addition, the manufacturability of a blanket shield block is also considered as the important design constraint in the cooling optimization. Our stress evaluation follows the relevant codes and standards outlined in the design protocol provided by the IO. This paper presents the analysis results, identifies issues on the preliminary configuration and makes suggestions on the design improvements.

A nuclear fusion fuel cycle plant is composed of various subsystems such as a hydrogen isotope storage and delivery system, a tokamak exhaust processing system, and a hydrogen isotope separation system. Korea shares in the construction of the International Thermonuclear Experimental Reactor fuel cycle plant with the EU, Japan and US, and is responsible for the development and supply of the storage and delivery system. We thus present details on the hydrogen isotope process safety. The main safety analysis procedure is to use a hazard and operability study. Nine segments were studied how the plant might deviate from its design purpose. We present a detailed description of the process, examine every part of it to determine how deviations from the design intent can occur and decide whether these deviations can give rise to hazards. We determine possible causes and note protective systems, evaluate the consequences of the deviation, and recommend actions to achieve our safety goal.

The ITER fuel cycle is designed for DT operation in equimolar ratio. It involves not only a group of fuelling system and torus cryo-pumping system of the exhaust gases through the divertor from the torus in tokamak plant, but also from the exhaust gas processing of the fusion effluent gas mixture connected to the hydrogen isotope separation in cryogenic distillation to the final safe storage & delivery of the hydrogen isotopes in tritium plant. Tritium plant system supplies deuterium and tritium from external sources and treats all tritiated fluids in ITER operation. Every operation and affairs is focused on the tritium inventory accountancy and the confinement. This paper describes the major fuel cycle processes and interfaces in the tritium plant in aspects of upcoming technologies for future hydrogen and/or hydrogen isotope utilization.

Hydrogen isotopes accountancy and storage are important functions in a nuclear fusion fuel cycle. The hydrogen isotopes are safely stored in metal hydride beds. The tritium inventory of the bed is determined from the decay heat of tritium. The decay heat is measured by circulating helium through the metal hydride bed and measuring the resultant temperature increase of the helium flow. We are reporting our preliminary experimental results on the hydrogen isotopes accountancy and storage performance in a metal hydride bed.

The compressibility effect on the hydride reaction of ZrCo, intermetallic compound, was studied using a visual cell reactor. In this study the compressibility effect on the hydride reaction of the ZrCo getter material was investigated by measuring the hydriding rate and monitoring the behavior through a visual cell. As a result, the ZrCo hydride reaction is found that the compressibility to the hydride reaction state has an effect on the initial hydride reaction rate and the final saturation state. For the hydride reaction of the ZrCo under pressurized reaction the log t 1/2 vs. log( p / p 0 ) is well represented by a straight line as the overall irreversible reaction having n th order reaction, where n is nearly 0.1. Even though the hydride reaction rate of the ZrCo is strongly hindered by the system compressibility, the inclination of hydride formation and the volume expansion could be propagated in a limited container space.

To verify an applicability of ZrCo hydride to the D-T gas storage and delivery system (SDS) in nuclear fusion fuel cycles, a new design of the ZrCo bed using rectangular tray configuration was developed by the Korea Atomic Energy Research Institute (KAERI). The objectives of this new design are (i) achievement of rapid hydrogen delivery performance on a ZrCo hydride bed, (ii) simplification of the bed internal configuration for manufacturing cost saving and (iii) improvement of operation reliability in 20 years operation of the nuclear fusion fuel cycles. The ZrCo bed composed of three trays was fabricated. ZrCo powder of 414 g per tray packed in thin wire mesh bags was placed on the surface of each tray. A filter plate of a large surface area, which enables to minimize pressure drop through the filter during rapid delivery of hydrogen from the hydride bed, was welded on the top of each tray. This paper presents details of the tray configuration and test results of hydrogen recovery and delivery performance of the bed.

Since the recommendation of blanket redesign by 2007 ITER design review, the blanket system has been developed in the framework of blanket integrated product team composed mainly of ITER organizati...

The Republic of Korea is participating in the ITER project as a full member and procuring components for the ITER tokamak and its facility. This paper presents the current stage of preparation of the ITER components to be delivered by Korea including an overview of Korean activities related to the ITER project, as of October 2009.

Korea is supposed to develop the ITER tritium storage and delivery system (SDS), which is one of the main components of the ITER tritium plant. For successful procurement, there are several ongoing R&D activities in the detailed design phase. Investigation of design parameters of the storage and delivery beds has been performed. Small and large-scale mock-ups of ZrCo beds are used to test the capability of desirable rapid delivery and recovery performance and to establish the pertinent procedure of in-bed calorimetry. An experimental apparatus is prepared to develop the integration and verification technologies for the unit processes of the tritium SDS. The performance test of a tritium-compatible metal bellows pump is examined, and the results show a reasonable agreement with the catalog data of the pump. A tritium storage and delivery bed simulator has been developed to simulate various bed operation scenarios under normal and abnormal conditions. A prototype of the SDS simulator is fabricated, and the bed operation scenario generation program to be applied to this simulator is developed. The design requirement of the tritium loading station (TLS) calorimeter is prepared based on a benchmarking mock-up calorimeter, namely, Korea Electric Power Research Institute Tritium Laboratory (KEPTL) calorimeter. Documents for the procurement of the TLS calorimeter will be developed through the experience on the KEPTL calorimeter operation.

Thermophysical properties of the complex metal hydride system such as zirconium cobalt hydride, an intermetallic hydride compound, in a massive state were estimated by introducing a crystal lattice structure in a stepwise formation and applying a mixing rule for each property. Experimental data in rarity in metal hydride system was used to calculate and to correlate the consistency of the mixed thermal and physical properties of the complex atomic structure in a unit cell. As a result, the volume expansion of the ZrCoH x was greatly influenced by the hydrogen content and increased to a maximum range of 36% at ZrCoH 3 system, but no meaning in the thermal expansion in engineering concept. In consideration of the heat capacity the temperature effect due to the hydrogen―an interstitial heat quantity― in the metal complex formation was mainly attributed, but not much for the hydrogen content (H/ZrCo ratio). In the temperature range between 200K and 600K the heat capacity of hydrogen atom was taken into account to reveal a sharp discrepancy in its non-hydriding property, especially in the lower temperature range. Atomic hydrogen was expected to behave from a gas to a solid property in heat capacity in the temperature ranges from 600K to 200K.

The ITER blanket shield block will be fabricated by the conventional process based on drilling, milling and welding of forged stainless steel blocks. For the adaptation of conventional method, one of the most important tasks to be verified is to develop the electron beam welding (EB W) technologies for the block to block joint. In addition, after drilling and milling for cooling passages, plugging techniques satisfying the requirements for the welding section should be developed. Three joining technologies, the electron beam welding for thick blocks, joining of lids and plugs, and attachment of flow drivers, were investigated and the results are summarized in this paper. Electron beam welding parameters for 70 and/or 110 mm thickness blocks satisfying the requirements in accordance with ASME code were established. It was also found that slower welding speed could suppress the formation of voids and porosities in weld metal. From the welding test results of mock-ups, optimal concepts on the weld design for front header lids or plugs were suggested.

One of the main research and development issues concerning the test blanket module (TBM) is the development of joining technologies for fabrication of the first wall. The objectives of the present study are to investigate the effects of thermal history corresponding to the TBM fabrication process on reduced activation ferritic martensitic (RAFM) steel microstructure, and to establish the appropriate hot isostatic pressure (HIP) conditions for the fabrication of RAFM/RAFM steels and beryllium (Be)/RAFM steels joints without degradation of the mechanical properties of the RAFM steel or delamination of the joined interface. In this study, RAFM and RAFM steels were joined by HIPing at 1050 °C under 100 MPa for 2 h. During the HIP process, the thermally altered microstructure and mechanical properties were recovered to the as-received state by subsequent normalizing at 950 °C for 2 h and tempering at 750 °C for 2 h. Be and RAFM steels were also bonded successfully by the application of Ti/Cu interlayers and HIPing at 850 °C under 100 MPa for 2 h.

Korea has been developing the metal hydride beds for the ITER storage and delivery system (SDS). There are two major concerns for the bed: one is the rapid delivery/recovery rate according to the tokamak plasma operation scenarios, and the other is the accurate measurement of tritium inventory using in-bed calorimetry within limited time. One of the bed concepts considering now consists of the thin and double layers of ZrCo which are confined by the cylinder-shaped SUS filter and the primary vessel wall. The primary vessel wall is wound by the heating coils, and has the helium channels for the in-bed calorimetry. In this study, the three-dimensional thermo-hydraulic analysis is performed for the bed to investigate the in-bed calorimetry performance. Based on the transient simulation results with an assumption of complete removal of the decay heat by He-loop, it is found that the bed under pre-cooling condition satisfies the accountancy of ±1 g-T accuracy in 24 h and ±3 g-T accuracy in 8 h.

It is known that the test of the capability of tritium self-sufficiency and the extraction of high-grade heat using tritium breeding blanket concepts is one of the major ITER missions. This requires the development of test blanket modules based on a corresponding DEMO blanket design, despite the differences in operating conditions between DEMO and ITER. Two blanket concepts such as a Helium-Cooled Solid Breeder (HCSB) blanket and a Helium-Cooled Molten Lithium (HCML) blanket are being considered as Korean Test Blanket Module (TBM) for ITER with the aim for verifying the capability of the breeding blanket design and manufacturability. In this paper, the current status of the design and analysis for the HCSB TBM is described. The key features of the design include the use of graphite neutron reflectors to reduce the amount of beryllium multiplier. Major analyses have been executed in order to optimize the design. The results of nuclear and thermo-hydraulic analysis show that the design parameters and requirements are satisfied. Also, the safety analysis results show that the design of the TBM system is sufficient to endure several accident events.

Recent activities and progress on the design and fabrication of the ITER blanket shield block in Korea are described in this paper. Hydraulic analyses, using a flow driver model for determining the gap between the radial cooling passages and flow drivers inside the shield block, were performed. The thermo-hydraulic analysis of half of a shield block was also conducted to investigate the uniformity of the flow stream in cooling passages and to evaluate the temperature distribution in the structure. The maximum temperature is below the allowable value, although hot spots occurred in the corner edge in the shield block. A manufacturing feasibility study for the development of the blanket shield block was performed in cooperation with KO industries. It was found that specific techniques would be required for the successful fabrication of an ITER blanket shield block, specifically electron-beam welding at a thickness up to 110 mm. The development of joining and drilling technologies for the thick shield block and lid joints is in progress. In addition, a full scale mock-up fabrication and the development of NDT techniques are planned in the near future.

ITER tokamak assembly is mainly composed of lower cryostat activities, sector sub-assembly, sector assembly, in-vessel activities and ex-vessel activities. The main tools for sector sub-assembly procedures consists of upending tool, sector lifting tool, vacuum vessel support and bracing tool and sector sub-assembly tool. Conceptual design of assembly tools for sector sub-assembly procedures is described herein. The basic structure for upending tool has been developed under the assumption that upending is performed with crane which will be installed in Tokamak building. Sector lifting tool is designed to adjust the position of a sector to minimize the difference between the center of the tokamak building crane and the center of gravity of the sector. Sector sub-assembly tool is composed of special frame for the fine adjustment of position control with 6 degrees of freedom. The design of VV support and bracing tool for four kinds of VV 40° sectors has been developed. Also, structural analysis for upending tool, sector sub-assembly tool has been studied using ANSYS for the situation of an applied load with the same dead weight multiplied by 3/4. The results of structural analyses for these tools were below the allowable values.

The ITER fuel cycle plant is composed of various subsystems such as a long term tritium storage system (LTS), a fuel storage and delivery system (SDS), a tokamak exhaust processing system, a hydrogen isotope separation system, and a tritium plant analytical system. Korea shares in the construction of the ITER fuel cycle plant with the EU (Japan)) and US, and is responsible for the development and supply of the SDS and LTS. The authors thus present details on the development status of the tritium transport container, the long term tritium storage beds, the short-term delivery system T{sub 2}, DT, and the D{sub 2} storage beds, the calorimetry system, and the associated He-3 recovery loop, the over pressure protection systems, and the gas analysis manifold connected to the tritium plant's analytical systems. (authors)

By means of a linear sweep voltammetry, a cyclic voltammetry and a chronopotentiometry, the electrolytic reduction of uranium oxide has been studied to establish the reduction mechanisms, which are based on a simultaneous uranium oxide reduction and a Li2O electrowinning, and the formation and electrolysis of lithium uranate. From the voltammograms, the reduction potentials of the uranium oxide and Li2O were obtained. From the chronopotentiometries based on the results of the voltammograms, the uranium oxide was reduced to uranium metal through the reduction mechanisms showing a more than 99% conversion. For a verification of the reduction mechanisms feasibility, basic data on the electrolytic reduction of the uranium oxide was obtained from the experiments and the characteristics of the closed recycle of Li2O were discussed.

An advanced spent fuel management process using a molten LiCl salt for the purpose of reducing spent oxide fuel to a metallic form generates a waste salt containing alkali, alkaline-earth, and some rare-earth fission products. A periodic removal of the high heat-generating Cs and Sr should be accomplished to reuse the salt since a recycling of the LiCl waste salt to a process stream is required to decrease the total amount of waste to be disposed of. In this study, zeolite 4A was proven to have desirable properties for the removal of the Cs and Sr elements from an LiCl molten salt phase, and the ion-exchange characteristics of zeolite in the molten salt were investigated. The adsorption of the Cs and Sr elements in an LiCl molten salt reaches nearly a constant value after 2–4 h of contact with the zeolite. The salt-occluded zeolite was produced in an LiCl molten salt, and then its ionexchange and salt occlusion properties were studied experimentally. The result indicates that zeolite 4A occluded between 10 and 11.5 salt molecules, and the salt-occluded zeolite was found to be a very effective molecular sieve for sorbing the Cs and Sr in the LiCl waste salt.

Consecutive absorption/desorption cycles of the ZrCo-H 2 system were studied to simulate the real ITER hydrogen getter system. ZrCo getter was used in this study instead of the depleted uranium (DU) getter material which was recently considered as the hydrogen getter in ITER. In a cyclic PCI measurement the high-pressure Sievert apparatus seems impractical to describe the equilibrium state of the ZrCo-H 2 system in detail, especially for the desorption stage. This high-pressure Pressure-Composition Isotherm (PCI) apparatus, however, shows a cause-and-effect well, from the previous getter state to the following state in presenting hydriding/dehydriding performance. In case of the ZrCo-H 2 system or in case of the DU-H 2 system, having multiple getter bed battery, a similar affection by the previous getter status might be related and a similar aspect could be shown to should consider further in ITER design, for example a need for control logic, from PCI measurements using a high-pressure Sievert apparatus.

Korea has been developing several concepts of the metal hydride beds for the ITER storage & delivery system (SDS). A thin double-layered annulus metal hydride bed is under development. It consists of the thin and double layers of ZrCo which are confined by the cylinder-shaped SUS filter and the primary vessel wall. The helium channels for the in-bed calorimetry are drilled at the primary vessel wall. Three-dimensional numerical simulation of the in-bed calorimetry of the metal hydride bed is performed. The temperature and flow rates are the major parameters of the helium flow condition of the in-bed calorimetry. The optimal helium flow condition determines the performance of the in-bed calorimetry which is evaluated by the time to steady state of heat exchange and the accuracy achieved. The various helium flow conditions are investigated and the pressure drop of helium flow is monitored. Based on the simulation results, the operation scenario of in-bed calorimetry of metal hydride bed is obtained. The results will be compared with the experimental test results on in-bed calorimetry performance of the 1:1 full scale mock-up bed.