• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 추계학술발표회논문집(1)
• /
• pp.191-196
• /
• 1996

The potential value of the Application Specific Integrated Circuits(ASIC's) in safety systems of Nuclear Power Plants(NPP's) is being increasingly recognized because they are essentially hardwired circuitry on a chip, the reliability of the system can be proved more easily than that of software based systems which is difficult in point of software V&V(Verification and Validation). There are two types of ASIC, one is a full customized type, the other is a half customized type. PLD(Programmable Logic Device) used in this paper is a half customized ASIC which is a device consisting of blocks of logic connected with programmable interconnections that are customized in the package by end users. This paper describes the RPS(Reactor Protection System) composed of ASICs which provides emergency shutdown of the reactor to protect the core and the pressure boundary of RCS(Reactor Coolant System) in NPP's. The RPS is largely composed of five logic blocks, each of them was implemented in one PLD, as the followings. A). Bistable Logic B). Matrix Logic C).Initiation Logic D). MMI(Man Machine Interface) Logic E). Test Logic.

• Nuclear Engineering and Technology
• /
• 제37권1호
• /
• pp.79-90
• /
• 2005

In this paper are reviewed the current status of nuclear fuel management and reactor operational aid tools. In addition, we indicate deficiencies in current capabilities and what future research is judged warranted. For the nuclear fuel management review the focus is on light water reactors and the utilization of stochastic optimization methods applied to the lattice, fuel bundle, core loading pattern, and for BWRs the control rod pattern/core flow design decision making problems. Significant progress in addressing separately each of these design problems on a single cycle basis is noted; however, the outstanding challenge of addressing the integrated design problem over multiple cycles under conditions of uncertainty remains to be addressed. For the reactor operational aid tools review the focus is on core simulators, used to both process core instrumentation signals and as an operator aid to predict future core behaviors under various operational strategies. After briefly reviewing the current status of capabilities, a more in depth review of adaptive core simulation capabilities, where core simulator input data are adjusted within their known uncertainties to improved agreement between prediction and measurement, is presented. This is done in support of the belief that further development of adaptive core simulation capabilities is required to further significantly advance the utility of core simulators in support of reactor operational aid tools.

• Nuclear Engineering and Technology
• /
• 제53권8호
• /
• pp.2477-2487
• /
• 2021

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

• Nuclear Engineering and Technology
• /
• 제53권12호
• /
• pp.3966-3978
• /
• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.11-18
• /
• 2022

In this research, curium oxide (CmO₂) is studied as fuel for VVER-1200 reactor to get an attention to its energy value and possibilities. For this purpose, CmO₂ is used in fuel rods or integrated burnable absorber (IBA) rods with and without UO₂ and then compared with the conventional fuel assembly of VVER-1200 reactor. It is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library. From these studies, it is found that CmO₂ is competent like UO₂ as a fuel due to higher fission cross-section of ²⁴³Cm and ²⁴⁵Cm isotopes and neutron capture cross-section of ²⁴⁴Cm and ²⁴⁶Cm isotopes. As a result, when some or all of the UO₂ of fuel rods or IBA rods are replaced by CmO₂, we get a similar k-inf like the reference even with lower enrichment UO₂ fuels. These studies show that the use of CmO₂ as IBA rods is more effective than the fuel rods considering the initially loaded amount, power peaking factor (PPF), fuel temperature and void coefficient, and the quality of spent fuel. From a detailed study, 3% CmO₂ with inert material ZrO₂ in IBA rods are recommended for the VVER-1200 reactor assembly from the once through concept.

• Nuclear Engineering and Technology
• /
• 제54권10호
• /
• pp.3919-3927
• /
• 2022

This study presents an optimal design of the coolant system of a non-refueling full-life small reactor by analyzing the space-integrated geometrical and electromagnetic variables of an extra vessel electromagnetic pump (EVEMP) for the circulation of a lead-bismuth eutectic (LBE) coolant. The EVEMP is an ideal alternative to the thermal-hydraulic system of non-refueling full-life micro reactors as it possesses no internal structures, such as impellors or sealing structures, for the transportation of LBE. Typically, the LBE passes through the annular flow channel of a reactor, is cooled by the heat exchanger, and then circulates back to the EVEMP flow channel. This thermal-hydraulic flow method is similar to natural circulation, which enhances thermal efficiency, while providing a golden time for cooling cores in the event of an emergency. When the forced circulation technology of the EVEMP was applied, the non-refueling full-life micro reactor achieve an output power of 60 MWt, which is higher than that achievable via the natural circulation method (30 MWt). Accordingly, an optimized EVEMP for Micro URANUS with a flow rate of 4196 kg/s and developed pressure of 73 kPa under a working temperature of 250 ℃ was designed.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.877-878
• /
• 2005

• Nuclear Engineering and Technology
• /
• 제41권4호
• /
• pp.447-454
• /
• 2009

This paper is a technical review of the research and development results of the Korea Nuclear Instrumentation and Control System (KNICS) project and Nu-Tech 2012 program. In these projects man-machine interface system architecture, two digital platforms, and several control and protection systems were developed. One platform is a Programmable Logic Controller (PLC) for a digital safety system and another platform is a Distributed Control System (DCS) for a non-safety control system. With the safety-grade platform PLC, a reactor protection system, an engineered safety feature-component control system, and reactor core protection system were developed. A power control system was developed based on the DCS. A logic alarm cause tracking system was developed as a man-machine interface for APR1400. Also, Integrated Performance Validation Facility (IPVF) was developed for the evaluation of the function and performance of developed I&C systems. The safety-grade platform PLC and the digital safety system obtained approval for the topical report from the Korean regulatory body in February of 2009. A utility and vendor company will determine the suitability of the KNICS and Nu- Tech 2012 products to apply them to the planned nuclear power plants.

• 한국수소및신에너지학회논문집
• /
• 제26권5호
• /
• pp.409-415
• /
• 2015

This study was performed to optimize the integrated anaerobic digestion (AD) and microbial electrolysis cells (MECs) for the enhanced hydrogen production. The optimum operational conditions of integrated AD and MECs were obtained using response surface methodology. The optimum substrate concentration and operational pH were 10 g/L and 6.8, respectively. In the confirm test, 1.43 mol $H_2/mol$ hexose was achieved, which was 2.5 times higher than only AD. After 40 to 60 hour at seeding, the volatile fatty acids (VFAs) in reactor of AD were not changed. However the VFAs of reactor of AD-MECs were reduced by 61.3% (acetate: 76.4%, butyrate: 50.0%, lactate: 55.0%).

• KSBB Journal
• /
• 제14권5호
• /
• pp.543-549
• /
• 1999

본 연구에서는 미생물의 성장시 수반하는 대사열을 측정할 수 있는 직접 열량계의 기능을 가질 수 있도록 발효조를 개량하였다. 발효조의 온도제어신호의 길이를 측정하고 이 신호를 컴퓨터에서 전달하여 계산하고 발효조에 공급된 누적 열량을 측정할 수 있도록 발효조를 일부 개조하고 온라인 측정 시스템을 구성하였다. 균체 없이 발효조를 운전하면서 여러 조건에서 누적 열량을 측정하여 통기량의 따라 30.9kJ/vvm의 열손실과 공기중으로 전도 및 대류, 복사에 의한 0.5 kJ/L/hr/$^{\circ}C$의 열손실이 발생함을 측정할 수 있었다. 그리고 소형의 가열체를 반응액에 발생함을 측정할 수 있었다. 그리고 소형의 가열체를 반응액에 투입하여 열량측정의 정확도를 확인하였으며 누적열량은 $\pm$0.2%의 오차 범위 내에서 측정되었다. 본 시스템에소 효모와 대장균을 이용하여 대사열량을 성공적으로 측정할 수 있음을 보였다.