Seungyon Cho

ITER storage and delivery system R&D in Korea

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.

Disproportionation Characteristics of a Zirconium-Cobalt Hydride Bed Under ITER Operating Conditions

Abstract Quantitative assessment of a disproportionation in the ZrCo-hydrogen system under ITER-relevant operating conditions was performed by means of experimental tests and a theoretical calculation. In the static temperature experiments with equilibrium hydrogen pressures, a 10% disproportionation of ZrCoHx (x = 2.0 and 2.5) was observed in 5.5 h at 415° (~78 kPa), 9 h at 400° (~72 kPa), 172 h at 380° (~51 kPa), and 1626 h at 350° (~28 kPa). An experimental formula [log τ = 17 268/T (K) - 25.814, where τ is the reaction time (day) of 10% disproportionation] was derived from these experiments. Experiments with a temperature cycling of up to 125 cycles (from room temperature to 350 to 360°) proved that no enhancement of a disproportionation occurs in the ZrCoHx (1.7 < x ≤ 2.0). Typical operation conditions of the ZrCo hydride bed for the D-T gas storage delivery system were proposed based on the ITER FDR 2000 plasma operation scenarios. The disproportionation rate estimated conservatively by the theoretical model indicates that a disproportionation in the ITER basic performance phase can be reduced by <4% even when there is a direct supply from the fuel storage and delivery system beds for all the D-T pulses and by <0.1% when the supply is from the hydrogen isotope separation system.

R&D Activities on the Tritium Storage and Delivery System in Korea

Abstract R&D activities on the tritium storage and delivery system include the development of getter beds to increase tritium recovery and delivery performance, the investigation of tritium reaction characteristics with ZrCo metal-hydride, in-bed calorimetry as tritium measurement techniques, and the development of process design technologies for the storage and delivery system such as pump performance test and bed simulator. The current status of the R&D activities on these subjects is addressed in this paper.

INITIAL TEST RESULTS OF A FAST HEAT TRANSFER RESPONSE ZrCo HYDRIDE BED

Abstract We are developing an innovative ZrCo hydride bed design, which is characterized by a large cylindrical filter, very thin cylindrical metal hydride powder packed layer, and large relative heating area per unit weight of ZrCo powder for ITER fuel cycle application. To validate this design concept, two ZrCo bed models each loaded with 127 g of ZrCo were tested by using H2 gas. In the first model, ZrCo powder was packed into the 3 mm gap between the filter cylinder and the vessel, and mold heater elements were attached to the outer surface of the vessel. The second model consisted of a layer of ZrCo powder packing (7 mm thickness), coiled cable heaters attached independently to the outer surface of the primary vessel and the inner surface of the filter cylinder. This paper presents detailed design features of the ZrCo bed models, and test results of the beds performances, i.e., temperature transient of the ZrCo packed bed during fast heating, hydriding rate up to 90-99% recovery, and 90-98% delivery fraction.

Fabrication of a 1/6-Scale Mock-Up for the Korea TBM First Wall in ITER

Abstract 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 structural material for the TBM first wall (FW), various joining methods have been developed with hot isostatic pressing in order to develop a TBM FW fabrication method. In this study, three small mock-ups were fabricated in order to develop and verify the manufacturing method of the TBM FW through the pressure and helium leak tests. They were successfully fabricated. After fabrication and checking the performance of the mock-ups, a 1/6-scale mock-up was fabricated with a 260-mm height, 444-mm width, and 435-mm depth, in which width and depth were preserved and the number of channels was reduced from 60 to 10. The mock-up has a U-type shape and ten channels with a size of 20-mm height and 10-mm width for cooling. A manifold for flow testing and high heat flux testing of the 1/6-scale mock-up was designed and fabricated to distribute fluid uniformly to the mock-up.

Modeling of tritium transport in ceramic breeder single crystal

Abstract Simple existing models were found not to adequately reproduce recent tritium release data for LiAlO 2 single crystal, indicating the need for a more comprehensive model. To help understand and interpret the data, a new model, MISTRAL-SC, was developed, incorporating bulk diffusion as well as the four major surface processes, and allowing for the variation of surface activation energies with coverage and for the presence of H 2 in the purge. The model is described in this paper and an analysis of the single crystal data presented. Based on the analysis a bulk diffusion coefficient for tritium diffusing in LiAlO 2 is estimated and compared to previous estimated values from past experiments. Discrepancies are discussed and recommendations are proposed for values of the diffusion coefficient to be used in LiAlO 2 experiment and blanket analyses, and for future work.

Current R&D Activities on Korean Helium Cooled Solid Breeder Test Blanket Module

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

EXPERIMENTAL STUDY ON THE DELIVERY RATE AND RECOVERY RATE OF ZrCo HYDRIDE FOR ITER APPLICATION

Abstract To investigate the key design aspects of the storage and delivery system (SDS) bed in ITER, rates of a hydriding, dehydriding and isotope effects on the H/D composition during a rapid delivery were experimentally investigated by using small tube-type reactors with different packing heights. Hydrogen recovery times for a shorter packing-height bed (20~40mm) decreased exponentially with an increasing initial hydrogen pressure, but increased by approximately two orders of a magnitude in a longer packing-height bed (145mm). Dehydriding rate increases exponentially with an increase in the relative heating area per unit weight of ZrCo powder and decreases in the packing-height of ZrCo hydride. Continuous isotopic compositional change inevitably occurs during the entire delivery time due to the known isotope effect in the metal-hydrogen systems. To overcome the isotope effect during a delivery from the SDS beds, an alternative operation method was suggested for the fuel supply from the SDS.

Modeling of tritium release from beryllium in fusion blanket applications

Only limited data are available on tritium release from irradiated Be. Few models have been suggested for tritium release from beryllium. Simple diffusion/desorption models have been proposed, but lack the capability of accounting for important phenomena such as burst release through interconnected porosity and the effect of the Be0 impurity layer. This paper describes the application of a new model for tritium release from Be, BETTY, to the analysis of recent experimental data. The model includes diffusion through Be and BeO, surface desorption and diffusion through interconnected porosity (as-fabricated or irradiation-induced). Tritium diffusion and desorption coefficients for Be and Be0 are estimated from the analysis and compared to previously estimated values.

Design analysis of electromagnetic forces on the KSTAR vacuum vessel interfaces

The KSTAR vacuum vessel consists of double-wall configuration with major ports, reinforcing ribs and supporting structures. The vacuum vessel structure supports all in-vessel structures. The electromagnetic loads on the plasma facing components induced during plasma disruptions would be transferred to the vessel through the supporting structure and produce a large localized stress on the interface between the vessel and the supporting structure. Due to the element limitation of load calculation code, a simplified supporting structure model was defined to obtain the electromagnetic loads on the plasma facing components. A new transferring method of these loads to the interface area was developed. In this paper the electromagnetic loads on the in-vessel components were calculated and stress analyses were performed to help design the interface area based on the equivalent forces estimated using the developed EM load transferring method.