• 한국수소및신에너지학회논문집
• /
• 제5권2호
• /
• pp.81-90
• /
• 1994

The hydrogen fuel tanks having hydrogen storing capacity of about 300g and 1200g are manufactured using $MmNi_{4.7}Al_{0.25}V_{0.05}Fe_{0.001}$ alloy. They are composed of several unit reactor made of Cu-tube(outer diameter = 50.1mm, thickness = 2mm). In order to increase the heat and mass transfer property of the hydride bed, Al-plates are inserted perpendicular to axial direction at intervals of 5mm and three arteries of diameter 8mm are installed symmetrically in each unit reactor. Hydrogen absorption is proceeded about 80% within 30 minute and is completed within 60 minute at the conditions of charging hydrogen pressure of 25atm and temperature of $22^{\circ}C$. On desorbing hydrogen at a constant rate of 30 slm at $20^{\circ}C$, discharging hydrogen pressure is sustained at 3~5atm for 120 minutes. The discharging pressure is increased upto 5~8atm as the increase of the reactor temperature to $30^{\circ}C$. From the experimental results and the brief discussions about the hydrogen absorption and disorption behaviors of the hydrogen storage tank, it is suggested that the behaviors of hydrogen charging and discharging could be controlled by adjusting the operating parameters and the reactor design parameters.

• 한국압력기기공학회 논문집
• /
• 제8권2호
• /
• pp.14-20
• /
• 2012

Full load rejection capability of nuclear power plant depends primarily on steam dump capacity (SDCAP) and steam generator level control capability. Recently, Ulchin Units 1&2 have performed stretched power uprate (SPU) and replacement steam generator (RSG) projects, which increase the power by 4.5 percent. They change major design or operating parameters and especially reduces steam dump capacity at full power due to increase of the steam flow. The reduction of SDC after SPU results in degradation of heat removal capability in full load rejection transients. Therefore, we should perform evaluation to determine whether reactor trips occur in large load rejection transients. Uchin Units 1&2 have experienced full load rejection (FLR) three times from 2004 to 2010. Operating data from the plant occurrence of FLR at Ulchin Units 1&2 showed that steam generator (SG) level transients were limiting in point of reactor trip. However the plant had never reached reactor trip in the FLR and successfully continued in house load operation. The parameters and setpoints for the SG will be changed if the SG is replaced. Therefore, we evaluated the appropriateness of steam dump, main feedwater and steam generator water level control system preventing the plant from reactor trip in case of FLR by the parameter sensitivity study whether SG water level operated smoothly after SPU and RSG projects.

• Nuclear Engineering and Technology
• /
• 제14권3호
• /
• pp.122-137
• /
• 1982

고리 1호기 일부 반응도사고에 대한 민감도 분석을 수행하였다. 본 민감도 분석에 고려한 반응도 사고는 비교적 진행속도가 빠른 사고로서 미임계나 저출력 시동조건에서 발생한 제어뱅크 인출사고와 제어봉 일출사고가 이에 속한다. 본 분석작업은 다음과 같이 세단계로 수행하는 바 원자로 평균출력의 변동 계산과 최고점에서의 열전달계산 그리고 DNBR계산 단계로 나눌 수 있다. 계산에 사용된 전산코드들은 본 분석을 위하여 개발하거나 기존 전산코드들을 수정ㆍ보완하여 제작하였으며 신뢰도도 평가하였다. 원자로 설계 및 운전변수들이 사고발생시 원자로의 거동에 미치는 영향을 조사하기 위하여 민감도 분석을 수행하였다. 본 민감도 분석 결과에 근거하여 고리 1호기 반응도사고 분석에 사용된 최종안전분석보고서의 가정과 초기조건이 타당한가를 조사하였고, 또한 계산 결과도 보수적이고 신뢰할 수 있는지 판별하였다. 고리 1호기 반응도사고 분석에 사용된 가정 및 초기조건을 재검토하고 민감도를 분석한 결과 최종안전분석보고서의 해석결과는 보수적이고 신뢰 할 수 있는 것으로 평가되었다.

• Journal of Microbiology and Biotechnology
• /
• 제8권1호
• /
• pp.81-88
• /
• 1998

The determination of appropriate parameters and parameter conditions is very important for the optimization of production of target materials. Taguchi's method has been used widely as the basis for development trials and optimization during industrial process design. Reaction variables which influence product yield are easily determined and their effects are revealed by just a few reactions, negating the need for extensive experimental investigation. There are usually some factors that are responsible for variations in process characteristics, so called noise factors. Controlling noise factors is very costly and difficult or impossible. Taguchi's experimental design method was examined to determine the control factor's level that is less sensitive to the changes in environmental conditions and other noise factors without control of noise factors. In this study, optimization of lipase-catalyzed production of lysophosphatidic acid (LPA) which has various physiological functions was performed by Taguchi's method. We obtained LPA yields ($66.5\%$) with low variance (5.32) at 400 RPM, molar ratio of 40 : 3 (mol) (fatty acid: G-3-P), 48 h, and $50^{\circ}C$. Thus, bioactive LPA with a desired fatty acid moiety could be produced with high yields and low variance despite various environmental noise factors.

• 반도체디스플레이기술학회지
• /
• 제18권3호
• /
• pp.115-121
• /
• 2019

Response surface methodology (RSM) is used to investigate the complex coupling effects of different operating parameters on silicon growth rate in planetary CVD reactor. Based on the computational fluid dynamics (CFD) model, an accurate RSM model is obtained to predict the growth rate with different parameters, including temperature, pressure, rotation speed of the wafer, and the mole fraction of dichlorosilane (DCS). Analysis of variance is used to estimate the contributions of process parameters and their interactions. Among the four operating parameters that have been studied, the influences of susceptor temperature and the operating pressure were the most significant factors that affect silicon growth rate, followed by the mole fraction of DCS. The influence of wafer rotation is the least. The validation tests show that the results of silicon deposition rate obtained from the regression model are in good agreement with those from CFD model and the maximum deviations is 2.15%.

• 방사성폐기물학회지
• /
• 제18권spc호
• /
• pp.75-87
• /
• 2020

The Korea Atomic Energy Research Institute (KAERI) has developed geological repository systems for the disposal of high-level wastes and spent nuclear fuels (SNFs) in South Korea. The purpose of the most recently developed system, the improved KAERI Reference Disposal System Plus (KRS⁺), is to dispose of all SNFs in Korea with improved disposal area efficiency. In this paper, a system-level safety assessment model for the KRS⁺ is presented with long-term assessment results. A system-level model is used to evaluate the overall performance of the disposal system rather than simulating a single component. Because a repository site in Korea has yet to be selected, a conceptual model is used to describe the proposed disposal system. Some uncertain parameters are incorporated into the model for the future site selection process. These parameters include options for a fractured pathway in a geosphere, parameters for radionuclide migration, and repository design dimensions. Two types of SNF, PULS7 from a pressurized water reactor and Canada Deuterium Uranium from a heavy water reactor, were selected as a reference inventory considering the future cumulative stock of SNFs in Korea. The highest peak radiological dose to a representative public was estimated to be 8.19×10^-4 mSv·yr^-1, primarily from ¹²⁹I. The proposed KRS⁺ design is expected to have a high safety margin that is on the order of two times lower than the dose limit criterion of 0.1 mSv·yr^-1.

• Nuclear Engineering and Technology
• /
• 제53권10호
• /
• pp.3413-3420
• /
• 2021

Applying the available neutron flux for medical and industrial purposes is the most important application of research reactors. The neutron radiography system is used for non-destructive testing (NDT) of materials so that it is one of the main applications of nuclear research reactors. One of these research reactors is the 5 MW pool-type light water research reactor of Tehran (TRR). This work aims to investigate on materials and location of the beam tube (BT) of the TRR radiography system to improve the index parameters of BT. Our results showed that a through-type BT with 20 cm thick carbon neutron filter, 1.2 cm and 9.4 cm of the diameter of inlet (D₁) and output (D₂) BT, respectively gives thermal neutron flux almost 25.7, 5.6 and 1.1 times greater than the former design of the TRR (with D₁ = 1.8 cm and D₁ = 9.4 cm), previous design of the TRR with D₁ = 3 cm and D₁ = 9.4 cm, and another design with D₁ = 5 cm and D₁ = 9.4 cm, respectively. Therefore, the design proposed in this paper could be a better alternative to the current BT of the TRR.

• 한국연소학회:학술대회논문집
• /
• 한국연소학회 2005년도 제31회 KOSCO SYMPOSIUM 논문집
• /
• pp.58-63
• /
• 2005

As fuel cells approach commercialization, hydrogen production becomes a critical step in the overall energy conversion pathway. Reforming is a process that produces a hydrogen-rich gas from hydrocarbon fuels. Hydrogen production via autothermal reforming (ATR) is particularly attractive for applications that demand a quick start-up and response time in a compact size. However, further research is required to optimize the performance of autothermal reformers and accurate models of reactor performance must be developed and validated. The design includes the requirement of accommodating a wide range of experimental set ups. Factors considered in the design of the reformer are capability to use multiple fuels, ability to vary stoichiometry, precise temperature and pressure control, implementation of enhancement methods, capability to implement variable catalyst positions and catalyst arrangement, ability to monitor and change reactant mixing, and proper implementation of data acquisition. A model of the system was first developed in order to calculate flowrates, heating, space velocity, and other important parameters needed to select the hardware that comprises the reformer. Predicted performance will be compared to actual data once the reformer construction is completed. This comparison will quantify the accuracy of the model and should point to areas where further model development is required. The end result will be a research tool that allows engineers to optimize hydrogen production via autothermal reformation.

• Nuclear Engineering and Technology
• /
• 제52권8호
• /
• pp.1847-1861
• /
• 2020

A design study was performed to improve the limit aseismic performance (LSP) of a primary heat transport system (PHTS) pump. This pump is part of the primary equipment of a prototype generation IV sodium-cooled fast reactor (PGSFR). The LSP is the maximum allowable seismic load that still ensures structural integrity. To calculate the LSP of the PHTS pump, a structural analysis model of the pump was developed and its dynamic characteristics were obtained by modal analysis. The floor response spectrum (FRS) initiated from a safety shutdown earthquake (SSE), 0.3 g, was applied to the support points of the PHTS pump, and then the seismic induced stresses were calculated. The structural integrity was evaluated according to the ASME code, and the LSP of the PHTS pump was calculated from the evaluation results. Based on the results of the modal analysis and LSP of the PHTS pump, design parameters affecting the LSP were selected. Then, ways to improve the LSP were proposed from sensitivity analysis of the selected design variables.

• Journal of Radiation Protection and Research
• /
• 제32권4호
• /
• pp.190-193
• /
• 2007

The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II Research Reactor Facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit ($10{\mu}Sv/hr$). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete.