• 한국정보처리학회논문지
• /
• 제4권6호
• /
• pp.1601-1614
• /
• 1997

본 논문에서는 ATM-LAN 환경의 MOD 서버들로부터 클라이언트들에게 프레임 기반의 CM(Continuous Media) 데이터 스트림을 전송하기 위한 기술들을 제안한다. 여기서 프레임이란 CM의 디스플레이 단위를 의미한다. 세부적으로는 수송 계층과 IP 계층의 개입없이 애플리케이션과 AAL이 직접적으로 접속함으로써 ATM-Specific한 native ATM 서비스를 적용한다. 또한 네비게이션 서버를 통한 투명한 브라이징 메카니즘, 서버와 클라이언트들간의 PVC를 통한 세션 설정 과정, 애플리케이션과 AAL과 QoS 협상 및 예약 과정, MIU 사이즈를 초과하는 프레임들에 대한 쪼갬/붙임 알고리즘 등을 제안하고 실험 결과를 제시한다.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
• /
• pp.423-428
• /
• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• Bulletin of the Korean Chemical Society
• /
• 제26권4호
• /
• pp.569-574
• /
• 2005

A coupled column liquid chromatography system was applied for the separation of the burnup monitors in spent nuclear fuel sample solutions. A reversed phase column was studied for the adsorption behavior of uranyl ions using alpha-hydroxyisobutyric acid as an eluent and used for the separation of plutonium and uranium. A cation exchange column prepared by coating 1-eicosylsulfate onto the reversed phase column was used for the separation of the lanthanides. In addition, retention of Np was checked with the reversed phase column and cation exchange column, respectively, according to the oxidation states to observe the interference effect for the separation of burnup monitors. This chromatography system showed a great reduction in separation time compared to a conventional anion exchange method. A good agreement from the burnup data was obtained between for this method and a conventional anion exchange method to within 1% of a difference for the spent nuclear fuel samples of about 40 GWD/MTU.

• Nuclear Engineering and Technology
• /
• 제46권3호
• /
• pp.423-430
• /
• 2014

An in-pile performance test of HANA claddings was conducted at up to 67 GWD/MTU in the Halden research reactor in Norway over a 6.5 year period. Four types of HANA claddings (HANA-3, HANA-4, HANA-5, and HANA-6) and a reference Zircaloy-4 cladding were used for the in-pile test. The evaluation parameters of the HANA claddings were the corrosion behavior, dimensional changes, hydrogen uptake, and tensile strength after the claddings were tested under the simulated operation conditions of a Korean commercial reactor. The oxide thickness ranged from 15 to 37 mm at a high flux region in the test rods, and all HANA claddings showed corrosion resistance superior to the Zircaloy-4 cladding. The creep-down rate of all HANA claddings was lower than that of the Zircaloy-4 cladding. In addition, the hydrogen content of the HANA claddings ranged from 54 to 96 wppm at the high heat flux region of the test rods, whereas the hydrogen content of the Zircaloy-4 cladding was 119 wppm. The tensile strength of the HANA and Zircaloy-4 claddings was similarly increased when compared to the un-irradiated claddings owing to the radiation-induced hardening.

• 제어로봇시스템학회논문지
• /
• 제5권6호
• /
• pp.753-763
• /
• 1999

The existing SCADA(Supervisory Control and Data Acquisition) system software is usually developed to suitable for the specific hardware platforms. However, as per rapid improvement of computer performance and development of network technology, it is required to support scalability and inter-operability in existing different SCADA systems. In order to meet such requirements, in this paper, we propose a new type of SCADA testbed using Java for electric distribution applications. The system consists of three modules; development support tools, client and server modules. The basic architecture of the proposed SCADA system is similar to existing one, however, we improve the function of MTU and MMI interface to facilitate LAN and WAN environment. Also, the proposed system can deals with alarm and history data by using heterogeneous DBMS. Since the system is built in Java environment, the development cost is cheap and it can support sacalability and portability. Our experience can be utilized to develop next generation of small and medium size of SCADA system.

• Journal of Advanced Marine Engineering and Technology
• /
• 제24권4호
• /
• pp.446-453
• /
• 2000

The closed cycle diesel system is operated in closed circuit system where there is non air breathing with working fluid consisted of the combination of oxygen, argon and recycled exhaust gas for obtaining underwater or underground power sources. this study has been carried out to analysis the performance of closed cycle system by means of investigation on the combustion characteristics of diesel engine MTU8V183TE52 operating in open, semi-closed, and closed cycle modes. The combustion in closed mode starts a little bit earlier than in open cycle mode. The oxygen concentration and fuel consumption at 240kW closed cycle running are 21∼24% by volume and 77∼79kg/h, respectively. The maximum cylinder pressure and ignition delay time are investigated 110bar and 8.9degree. Also, The combustion simulation program has been studied to predict whether or not combustion. The results from numerical prediction for the basic, cylinder averaged quantities such as the cylinder pressure and the heat release showed excellent with the experimental data.

• Nuclear Engineering and Technology
• /
• 제50권5호
• /
• pp.709-716
• /
• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• 한국멀티미디어학회논문지
• /
• 제24권11호
• /
• pp.1526-1537
• /
• 2021

Due to the difference between the underwater communication environment and the terrestrial communication environment, the radio communication mainly used on the ground cannot be used in underwater. For this reason, in the underwater communication environment, various communication media such as acoustic waves, infrared rays, light and so on has been studied, but there exist several difficulties in operating them individually due to their physical limitations. The concept for overcoming these difficulties is the very underwater multi-media communication, a method to select a communication medium best suitable for the current underwater environment among underwater communication multimedia whenever there occurs underwater communication failure. In this paper, we present an underwater multi-media communication network based on star topology and a fragmentation and reassembly technique to solve the problems caused by the different MTU (Maximum Transmission Unit) sizes among different underwater communication media. We also present the estimations and analysis on processing times in each of fragmentation and reassembly and the total data amount for transmitting fragments in our proposed underwater multi-media communication network.

• Nuclear Engineering and Technology
• /
• 제39권5호
• /
• pp.673-682
• /
• 2007

The contents of transuranic elements in high-burnup spent fuel samples were determined. The activity amounts of $^{238}Pu,\;^{239}Pu,\;^{240}Pu,\;^{241}Am,\;^{244}Cm\;and\;^{242}Cm$ were measured by alpha spectrometry using $^{242}Pu\;and\;^{243}Am$ as tracers, respectively. A spike addition method for $^{237}Np$ was established by an alpha and gamma spectrometry using $^{239}Np$ as a spike after the optimum conditions for the measurements of $^{237}Np\;and\;^{239}Np$, respectively, were obtained. A separation system using anion exchange chromatography and diethylhexylphosphoric acid extraction chromatography was applied for the separation of these elements. This method was applied to high-burnup spent nuclear fuel samples $(40{\sim}60GWD/MTU)$. The contents of the transuranic elements were compared with those by ORIGEN-2 code. Measurements and the calculations of the contents of the plutonium isotopes $^{238}Pu,\;^{239}Pu\;and\;^{240}Pu$ agreed to within 10% on average. The contents of $^{237}Np$ agreed to within approximately 5% except for one instance of a calculation, while those of $^{241}Am,\;^{244}Cm\;and\;^{242}Cm$ showed higher values by approximately 19%, 35% and 14% on average, respectively, compared to the calculations according to the burnup.

• Nuclear Engineering and Technology
• /
• 제34권3호
• /
• pp.187-201
• /
• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.