• Nuclear Engineering and Technology
• /
• v.40 no.1
• /
• pp.87-92
• /
• 2008

Tokamak reactor system analysis code was developed at KAERI (Korea Atomic Energy Research Institute) and is used here for the conceptual development of a DEMO reactor. In the system analysis code, prospects of the development of plasma physics and the relevant technology are included in a simple mathematical model, i.e., the overall plant power balance equation and the plasma power balance equation. This system analysis code provides satisfactory results for developing the concept of a DEMO reactor and for identifying the necessary R&D areas, both in the physics and technology areas for the realization of the concept. With this system analysis code, the performance of a DEMO reactor with a limited extension of the plasma physics and technology adopted in the ITER design. The main requirements for the DEMO reactor were selected as: 1) demonstrate tritium self-sufficiency, 2) generate net electricity, and 3) achieve a steady-state operation. It was shown that to access an operational region for higher performance, the main restrictions are presented by the divertor heat load and the steady-state operation requirements.

• Korean Journal of Materials Research
• /
• v.30 no.12
• /
• pp.655-659
• /
• 2020

Hydrogen isotopes (i.e. deuterium and tritium) are supplied to the tokamak in the International Thermonuclear Experimental Reactor (ITER) fuel cycle. One important part of the ITER fuel cycle is the recycling of unused fuel back to the tokamak, as almost 99 % of fuel is unburned during fusion reaction. For this, cryogenic distillation has been used in the isotope separation system (ISS) of ITER, but this technique tends to be energy-intensive and to have low selectivity (typically below 1.5 at 24 K). Recently, efficient isotope separation by porous materials has been reported in the so-called quantum sieving process. Hence, in this study, hydrogen isotope adsorption behavior is studied using chemically stable ZIF-11. At low temperature (40 K ~ 70 K), the adsorption increases and the sorption hysteresis becomes stronger as the temperature increases to 70K. Molar ratio of deuterium to hydrogen based on the isotherms shows the highest (max. 14) ratio at 50 K, confirming the possibility of use as a potential isotope separation material.

• Nuclear Engineering and Technology
• /
• v.55 no.1
• /
• pp.100-108
• /
• 2023

In this study, a neural network model inspired by a one-dimensional convolution U-net is developed to automatically accelerate edge localized mode (ELM) detection from big diagnostic data of fusion devices and increase the detection accuracy regardless of the hyperparameter setting. This model recognizes the input signal patterns and overcomes the problems of existing detection algorithms, such as the prominence algorithm and those of differential methods with high sensitivity for the threshold and signal intensity. To train the model, 10 sets of discharge radiation data from the KSTAR are used and sliced into 11091 inputs of length 12 ms, of which 20% are used for validation. According to the receiver operating characteristic curves, our model shows a positive prediction rate and a true prediction rate of approximately 90% each, which is comparable to the best detection performance afforded by other algorithms using their optimized hyperparameters. The accurate and automatic ELM-burst detection methodology used in our model can be beneficial for determining plasma properties, such as the ELM frequency from big data measured in multiple experiments using machines from the KSTAR device and ITER. Additionally, it is applicable to feature detection in the time-series data of other engineering fields.

• Journal of the Korean Vacuum Society
• /
• v.16 no.5
• /
• pp.388-396
• /
• 2007

KSTAR (Korea Superconducting Tokamak Advanced Research) current lead system (CLS) has a role to interconnect magnet power supply (MPS) in room temperature (300 K) and superconducting (SC) bus-line, electrically. For the first plasma experiments, it should be assembled 4 current leads (CL) on toroidal field (TF) current lead box (CLB) and 14 leads on poloidal field (PF) CLB. Two current leads, with the design currents 17.5 kA, and SC bus-lines are connected in parallel to supply 35 kA DC currents on TF magnet. Whereas, it could supply $20\;{\sim}\;26\;kA$ to each pairs of PF magnets during more than 350 s. At the cold terminals of the leads, there are joined SC bus-lines and it was constructed helium coolant control system, aside from main tokamak system, to protect heat flux through current leads and enhanced Joule heat due to supplied currents. Throughout the establishment processes, it was tested the high vacuum pumping, helium leak of the helium lines and hardwares mounted between the helium lines, flow controls for CL, and liquid nitrogen cool-down of possible parts (current leads, CL helium lines, and thermal shield helium lines for CLB), for the accomplishment of the required performances.

• Proceedings of the KIEE Conference
• /
• 1988.11a
• /
• pp.206-209
• /
• 1988

Fixed boundary MHD static equilibrium for the axisymmetric toroidal plasma is analyzed numerically. The Grad-Shafranov equation is solved using FFM. The toroidal current tenn is expressed by plasma pressure p($\psi$) and toroidal field function g($\psi$). The numerical results are compared to the Solovev analytic equilibrium for the verification of the analysis. For SNUT-79 tokamak device in Seoul National University, flux surfaces and toroidal current profiles according to the variation of p and g profiles are observed. Also the safety factor q and average beta are obtained.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
• /
• 2006.05a
• /
• pp.438-443
• /
• 2006

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
• /
• 2000.02a
• /
• pp.119-120
• /
• 2000

The thermal shield for the TF coils and PF coils has been located between the coils and vacuum vessel. The thermal shielding cryopanel is cooled under 80K by a fored flow of helium gas using cooing pipes on the cryopanel. Design of the KSTAR thermal shield of vacuum vessel is described.

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
• /
• 2000.02a
• /
• pp.121-122
• /
• 2000

KSTAR coils use super-conducting magnet systems. These coils operate around 4.5K and therefore require a thermal shield to reduce the heat load from outer cryostat. 80K thermal shields must be cooled by a forced flow of He gas at 20 bars without the pressure drop of 0.45 bar. Designed thermal shield shows that the pressure drop is lower than that of 0.5 bar.

• Proceedings of the KIPE Conference
• /
• 2010.07a
• /
• pp.279-280
• /
• 2010

본 논문의 Quench System은 KSTAR의 토카막의 초전도 코일인 플로이달 코일과 토로이달 코일의 보호장치로서 코일에서 Arc 발생 시 코일과 전원을 차단, 분리하고 코일전류를 Dump 저항으로 급속하게 방전시켜, 코일을 보호하는 System 이다.

• Proceedings of the KIPE Conference
• /
• 2017.07a
• /
• pp.425-426
• /
• 2017

본 논문은 미래 녹색에너지원으로 주목받는 핵융합에너지 개발 장치인 한국형 초전도 핵융합 연구로(KSTAR : Korea Superconducting Tokamak Advanced Research)의 가열장치인 중성입자빔(NBI : Neutral Beam Injector) 장치를 소개하고, 2016년 KSTAR 실험에서 빔 에너지 90/70/80 keV, 빔 출력 3.81 MW, 70.6초의 최대 장 펄스와 빔 에너지 95/90/90 keV, 빔 출력 5.13 MW, 10.6 초로 입사 운전한 NBI 장치 실험 결과를 정리한다.