Hyung Gon Jin

Manufacturing and Examination for ITER Blanket First Wall Small-Scale Mockups With KoHLT-EB in Korea

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/m2 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 × 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°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.

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

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

Scoping Study on In-Vessel LOCA of a Korean TBS in ITER

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.

Thermal-Hydraulic Analysis for Conceptual Design of Korean HCCR TBM Set

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.

Assessment and modification of an ion source grid design in KSTAR neutral beam systema)

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.

Functional components design and analysis of a Korean HCCR TBM in ITER

Korea has developed a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) for testing in 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.

Korean Development of Advanced Thermal-Hydraulic Codes for Water Reactors and HTGRs: Space and Gamma

Abstract 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 for safety analysis of pressurized water reactors (PWRs) and high-temperature gas-cooled reactors (HTGRs), respectively. SPACE will replace outdated vendor-supplied codes and will be used for the safety analysis of operating PWRs and for the design of an advanced PWR. SPACE consists of up-to-date physical models of two-phase flow dealing with multidimensional two-fluid, three-field flow. GAMMA consists of multidimensional governing equations consisting of the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of n species. GAMMA is based on a porous media model so that thermofluid and chemical reaction behaviors in a multicomponent mixture system and heat transfer within solid components, free and forced convection between a solid and a fluid, and radiative heat transfer between solid surfaces can be dealt with. GAMMA has a two-dimensional helium turbine model based on the throughflow calculation and a coupled neutronics-thermal-hydraulic model. Extensive code assessment has been performed for the verification and validation of SPACE and GAMMA.

Thermo-Mechanical Analysis for the Conceptual Design of Korean HCCR TBM-Set

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.

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

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

Status and Plan of the System Codes Development for ITER Test Blanket Module and Fusion Breeding Blanket in Korea

Abstract 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 into single ones, called MARS-FR (Fusion Reactor) and GAMMA-FR, their environments were carefully handled during their development procedure.