• Nuclear Engineering and Technology
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• 제51권4호
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• pp.949-953
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• 2019

The burnup equation of nuclides is one of the most equations in nuclear reactor physics, which is generally coupled with transport calculations. The burnup equation describes the variation of the nuclides with time. Because of its very stiffness and the need for large time step, this equation is solved by special methods, for example transmutation trajectory analysis (TTA) or the matrix exponential methods where the matrix exponential is approximated by CRAM. However, TTA or CRAM functions well when the flux is constant. In this work, a new method is proposed when the flux changes. It's an improved method compared to TTA or CRAM. Furtherly, this new method is based on TTA or CRAM, and it is more accurate than them. The accuracy and efficiency of this method are investigated. Several cases are used and the results show the accuracy and efficiency of this method are great.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2001년도 ICCAS
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• pp.133.6-133
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• 2001

Reactor vessel is one of the most important equipment of Nuclear Power Plant (NPP) with regard to the nuclear safety. Thus reactor vessel must be examined periodically by certified experts. Currently, ultra-sonic(UT) non-destructive inspection is executed on reactor vessel. Two different techniques are used in this inspection. One is using the movable manipulator fixed with the support-guide placed on the vessel, and the other is using mobile robot moving in the vessel. Movable manipulator machine is very heavy, hard to handle, and very expensive. Mobile robot equipment is small and convenient but has a weak point on positional precision. To solve these problems we developed a reactor inspection system based on laser-driven mobile robot. This paper describes the main concept and structure of integrated inspection units and the feature of implemented units.

• 한국신재생에너지학회:학술대회논문집
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• 한국신재생에너지학회 2008년도 춘계학술대회 논문집
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• pp.210-213
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• 2008

The purpose of this study is to investigate the performances of organic removal and methane recovery in the full scale two-phase anaerobic system. The full scale two-phase anaerobic system was consists of an acidogenic ABR (Anaerobic Baffled Reactor) and a methanognic UASB (Upflow Anaerobic Sludge Blanket) reactor. The volume of acidogenic and methanogenic reactors is designed to 28.3 $m^3$ and 75.3 $m^3$. The two-phase anaerobic system represented 60-82% of COD removal efficiency when the influent COD concentration was in the range of 7,150 to 16,270 mg/L after screening (average concentration is 10,280 mg/L). After steady-state, the effluent COD concentration in the methanogenic reactor showed 2,740 $\pm$ 330 mg/L by representing average COD removal efficiency was 71.4 $\pm$ 8.1% when the operating temperature was in the range of 19-32$^{\circ}C$. The effluent SCOD concentration was in the range of 2,000-3,000 mg/L at the steady state while the volatile fatty concentration was not detected in the effluent. Meanwhile, the COD removal efficiency in the acidogenic reactor showed less than 5%. The acidogenic reactor played key roles to reduce a shock-loading when periodic shock loading was applied and to acidify influent organics. Due to the high concentration of alkalinity and high pH in the effluent of the methanogenic reactor, over 80% of methane in the biogas was produced consistently. More than 70 % of methane was recovered from theoretical methane production of TCOD removed in this research. The produced gas can be directly used as a heat source to increase the reactor temperature.

• 대한기계학회논문집 C: 기술과 교육
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• 제3권4호
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• pp.299-305
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• 2015

미래에너지의 해법으로 원자력에너지를 이용한 물분해 수소생산시스템의 핵심기술을 개발하였다. 안전성을 보장할 수 있는 제4세대 원자로인 초고온가스로의 고열을 이용하여 황요오드 열화학적인 방법으로 물을 분해하여 수소를 생산하는 기술이다. 원자력수소생산 핵심기술은 초고온에서의 열을 공급하는 것을 모사하는 초고온 실험기술, 초고온가스로의 안전성을 모사하는 연구, 초고온가스로의 노심과 안전성을 해석할 수 있는 도구의 개발, 초고온가스로에 사용하는 연료제조기술, 물을 분해하여 열화학적인 방법으로 수소를 생산하는 기술로 구성된다. 원자력수소생산에 필요한 핵심기술을 개발하고 실험실 규모로 입증하였으며, 대규모 실용화를 위해서 선결되어할 미완성 기술을 제시하였다. 본 기술은 제4세대 원자로개발 국제공동연구로 수행한 기술로서 향후 미래의 원자로 기술이다.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.1950-1958
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• 2024

For the pressurized water reactor two-step calculation, the traditional assembly homogenization and two-group neutron diffusion calculation have been widely used. When it comes to the core pin-by-pin simulation, many models and techniques are different and unsettled. In this paper, the homogenization methods based on the pin discontinuity factors and super homogenization factors are used to get the pin-cell homogenized parameters. The heterogeneous leakage model is applied to modify the infinite flux spectrum of the single assembly with reflective boundary condition and to determine the diffusion coefficients for the SP3 solver which is used in the core simulation. To reduce the environment effect of the single-assembly reflective boundary condition, the online method for the SPH factors updating is applied in this paper, and the functionalization of SPH factors based on the least-squares method will be pre-made alone with the table of the group constants. The fitting function will be used to update the thermal-group SPH factors with a whole-core pin-by-pin homogeneous solution online. The three-dimensional Watts Bar Nuclear Unit 1 (WBN1) problem was utilized to test the performance of pin-by-pin calculation. And numerical results have demonstrated that PWR pin-by-pin core calculation has more accurate results compared with the traditional assembly-homogenization scheme.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.90-95
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• 1997

In this work, linear dynamics of a circulating fluid-fueled subcritical reactor system with temperature feedback and external neutron source was modeled and examined. In a circulating fluid-fuel system, the stable region is slightly moved by a circulation fluid effect. The effects of subcriticality and temperature feedback coefficient on the reactor stability were tested by calculating frequency response of neutron density originated from reactivity perturbation or external source oscillation of system. The amplitude transfer function has a different shape near subcritical region due to the exponential term in the transfer function. The results of the study show that at a slightly subcritical region, low frequency oscillation in accelerator current or reactivity can be amplified depending on the temperature feedback. However, as the subcriticality increases, the oscillation becomes negligible regardless of the magnitude of the temperature feedback coefficient.

• Nuclear Medicine and Molecular Imaging
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• 제41권1호
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• pp.1-8
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• 2007

This review discusses the production of alpha-particle-emitting radionuclides in radioimmunotherapy. Radioimmunotherapy labeled with alpha-particle is expected to be very useful for the treatment of monocellular cancer (e.g. leukemia) and micrometastasis at an early stage, residual tumor remained in tissues after chemotherapy and tumor resection, due to the high linear energy transfer (LET) and the short path length in biological tissue of alpha particle. Despite of the expected effectiveness of alpha-particle in radioimmunotherapy, its clinical research has not been activated by the several reasons, shortage of a suitable a-particle development and a reliable radionuclide production and supply system, appropriate antibody and chelator development. Among them, the establishment of radionuclide development and supply system is a key factor to make an alpha-immunotherapy more popular in clinical trial. Alpha-emitter can be produced by several methods, natural radionuclides, reactor irradiation, cyclotron irradiation, generator system and elution. Due to the sharply increasing demand of $^{213}Bi$, which is a most promising radionuclide in radioimmunotherapy and now has been produced with reactor, the cyclotron production system should be developed urgently to meet the demand.

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1323-1332
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• 2009

The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.813-820
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• 2009

A rapid pneumatic transfer system (PTS) for an instrumental neutron activation analysis (INAA) is developed as an automatic irradiation facility involving the measurement of a short half-life nuclide and a delayed neutron counting system. Three new PTS designs with improved functions were constructed at the HANARO research reactor in 2006. The new system is composed of a manual system and an automatic system for both an INAA and a delayed neutron activation analysis (DNAA). The design and basic conception of a modified PTS are described, and the functions of system operation and control, radiation protection and emissions of radioactive gas are improved. In addition, a form of capsule transportation of these systems is tested. The experimental results pertaining to the irradiation characteristics with variation of the neutron flux and the temperature of the irradiation position with the irradiation time are presented, as is an analysis of the reference material for analytical quality control and uncertainty assessments.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.476-483
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• 2017

The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of $0.22^{\circ}C/W$, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.