• 한국수소및신에너지학회논문집
• /
• 제24권2호
• /
• pp.121-127
• /
• 2013

Tritium was attracted with high energy source in neutron fusion energy systems. A number of research was performed in tritium storage materials. The Korea was raised storage and delivery systems (SDS) of international thermonuclear experimental reactor (ITER) research. However, bottles of SDS would be important because of stability. The bottles have a welding zone, this zone will be vulnerable to hydrogen embrittlement. This zone have a high thermodynamic energy and heat deterioration. Therefore bottles were studied about hydrogen embrittlement to retain stability. The heat treatment of hydrogen was carried under pressure-composition-temperature (PCT) apparatus because of checking at real time. And then, mechanical properties were evaluated by tensile test and hardness test. In results of this study, hydrogen atmosphere condition is very important by tensile test and kinetics test. The samples were evaluated, that is more weak hydrogen pressure, increasing temperature and time. This results could be useful in SDS bottle designs.

• 한국에너지공학회:학술대회논문집
• /
• 한국에너지공학회 2000년도 추계 학술발표회 논문집
• /
• pp.115-120
• /
• 2000

The analysis of flux distribution under stead-state in large power reactors with assymetry reactivity insertions requires the use of three-dimensional diffusion calculations. For the purpose, consistently formulated modern nodal methods based on higher order interface techniques have become popular tools for flux distributions in large commercial nuclear reactors. Among the earlier developments, the nodal Green's function method obtains its nodal interface equation from the transverse-integrated integral diffusion equation using a finite-medium Green's function. In this method, the outgoing current from a node surface is formulated as a response of the incoming currents and the spatially integrated neutron source within the same node. The well-known nodal expansion method is also based on an interface partial current formulation. Nodal methods high-level interface variables, i.e., interface net current and flux, may be more computationally efficient than the nodal Green's function method because they have one fewer unknown per interface. The Analytic Nodal Method(ANM), which can be classified as an interface net current technique and, was faster in solving some standard benchmark problems than the other two methods.(omitted)

• 한국방사선학회논문지
• /
• 제9권5호
• /
• pp.295-300
• /
• 2015

현재 우리나라는 포항방사광가속기와 국립암센터의 양성자 치료용 가속기와 경주 양성자 가속기가 운영되고 있고 중이온 가속기, 4세대 방사광가속기 등 대형가속기 시설이 건설 중에 있다. 이들 시설에서 고에너지로 가속된 입사입자는 표적물질과 상호작용 후 2차 중성자를 발생시키고, 이 중성자는 가속기 구조물 및 주변 콘크리트, 토양, 지하수 등을 방사화 시킨다. 따라서 이러한 가속기 시설의 안전적 측면을 고려할 때 방사화를 일으키는 중성자의 차폐가 중요하다. 본 연구는 차폐해석에 사용되는 몬테카를로 코드 중 MCNPX를 이용하여 $^{12}C$ beam빔과 표적물질(Cu)과의 상호작용 후 생성되는 중성자를 계산하고, 그 중성자의 철 차폐체와 콘크리트 차폐체의 두께별 투과 후 스펙트럼을 MCNPX의 JENDL/HE 07과 la150을 이용해 비교하여 계산하였다. 빔의 방향과 차폐체의 종류 및 두께에 따라 그 결과를 실험값과 비교하여 검증함으로써 핵자료의 특성을 확인하였으며 향후 대형가속기시설의 선량평가용 기반기술로 활용하고자 하였다.

• Nuclear Engineering and Technology
• /
• 제52권10호
• /
• pp.2410-2414
• /
• 2020

Heavy ions have a high potential for destroying deep tumors that carry the highest dose at the peak of Bragg. The peak caused by a single-energy carbon beam is too narrow, which requires special measures for improvement. Here, carbon-12 (¹²C) ion with different energies has been used as a source for calculating the dose distribution in the water phantom, soft tissue and bone by the code of Monte Carlobased FLUKA code. By increasing the energy of the initial beam, the amount of absorbed dose at Bragg peak in all three targets decreased, but the trend for this reduction was less severe in bone. While the maximum absorbed dose per bone-mass unit in energy of 200 MeV/u was about 30% less than the maximum absorbed dose per unit mass of water or soft tissue, it was merely 2.4% less than soft tissue in 400 MeV/u. The simulation result showed a good agreement with experimental data at GSI Darmstadt facility of biophysics group by 0.15 cm average accuracy in Bragg peak positioning. From 200 to 400 MeV/u incident energy, the Bragg peak location increased about 18 cm in soft tissue. Correspondingly, the bone and soft tissue revealed a reduction dose ratio by 2.9 and 1.9. Induced neutrons did not contribute more than 1.8% to the total energy deposited in the water phantom. Also during ¹²C ion bombardment, secondary fragments showed 76% and 24% of primary 200 and 400 MeV/u, respectively, were present at the Bragg-peak position. The combined treatment of carbon ions with neutron or electron beams may be more effective in local dose delivery and also treating malignant tumors.

• Nuclear Engineering and Technology
• /
• 제37권4호
• /
• pp.363-374
• /
• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• Nuclear Engineering and Technology
• /
• 제56권7호
• /
• pp.2756-2766
• /
• 2024

Small rod-controlled pressurized water reactors (PWR) are the ideal energy source for vessel propulsion, benefiting from their high reactivity control efficiency. Since the control rods (CRs) increase the complexity of reactivity control, this paper seeks to study the performance of CRs in small rod-controlled PWRs to extend the lifetime and reduce power offset due to CRs. This study investigates CR grouping, move-in/out strategies, and axially non-uniform design effects on core neutron physics metrics. These metrics include axial offset (AO), core lifetime (CL), fuel utilization (FU), and radial power peaking factor (R-PPF). To simulate the movement of the CRs, a "Critical-CR-burnup" function was developed in OpenMC. In CR designs, the CRs are grouped into three banks to study the simultaneous and prioritized move-in/out strategies. The results show CL extension from 590 effective full power days (EFPDs) to 638-698 EFPDs. A lower-worth prioritized strategy minimizes AO and the extremum values decrease from -0.69 and + 0.81 to -0.28 and + 0.51. Although an axially non-uniform CR design can improve AO at the beginning of cycle (BOC), considering the overall CR worth change is crucial, as a significant decrease can adversely impact axial power distribution during the middle of cycle (MOC).