• Journal of Energy Engineering
• /
• v.8 no.2
• /
• pp.213-223
• /
• 1999

In recently, the safety system software of the nuclear power plant has been verified and validated according to ANSI/IEEE-ANS-7-4.3.2-1982 to improve the reliability. This standard requires that safety-related software should be tested in the static and dynamic environments. In case of Inadequate Core Cooling Monitoring System (ICCMS), the static test procedure and related techniques are developed but the dynamic test procedure and related techniques are not developed. Therefore, this paper discusses the undeveloped techniques, and suggests the dynamic test procedure and the program for generation of test input data. The performance of the program was identified using accident analysis report of Ulchin 3&4 Final Safety Analysis Report (FSAR).

• Journal of the Korean Society for Aeronautical & Space Sciences
• /
• v.35 no.9
• /
• pp.805-815
• /
• 2007

In order to build a conceptual design program for a liquid rocket engine system, performance based sub-programs for each core component of the engine system were made. Parts included were the combustion chamber, supersonic nozzle, centrifugal pump, and impulsive turbine. Simple mathematical models based on classical thermodynamic and inviscid theories were adopted with proper tuning by empirical data. In Part I, aiming to validate each sub-program, we examined the results of each program qualitatively, and parametrically investigated the sensitivity due to the change in design parameters.

• Nuclear Engineering and Technology
• /
• v.44 no.8
• /
• pp.855-870
• /
• 2012

The paper deals with three main topics: a) the definition of System Thermal-Hydraulics (SYS TH), b) a historical outline for SYS TH and, c) the description of elements for reflection when planning research projects or improvement activities, this last topic being the main reason for the paper. Distinctions between basic thermal-hydraulics and computational Fluid-Dynamics (CFD) on the one side and SYS TH on the other side are considered under the first topic; stakeholders in the technology are identified. The proposal of Interim Acceptance Criteria for Emergency Core Cooling Systems in 1971 by US NRC (AEC at the time) is recognized as the starting date or the triggering event for SYS TH (second topic). The complex codes and the main experimental programs (list provided in the paper) constitute the pillars for SYS TH. Caution or warning statements are introduced in advance when discussing the third topic: a single person (or a researcher) has little to no possibility, or capability, of streamlining the forthcoming investments or to propose a roadmap for future activities. Nevertheless, the ambitious attempt to foresee developments in this area has been pursued without constraints connected with the availability of funds and with industrial benefits or interests. Demonstrating the acceptability of current SYS TH limitations and training in the application of those codes are mentioned as the main challenges for forthcoming research activities.

• Journal of Korea Society of Digital Industry and Information Management
• /
• v.11 no.1
• /
• pp.27-35
• /
• 2015

Increasing the number of GPU (Graphic Processor Unit) cores, the studies on High Performance Computing Platform using GPU have actively been made in recent. This trend has led to the development of GPGPU (General Purpose GPU) and CUDA (Compute Unified Device Architecture) Framework. In this paper, we explain the many benefits of the GPU based system, and propose the ICIDF(Identify Compute-Intensive Data set and Function) methodology to apply GPU technology to legacy military information system for performance improvement. To demonstrate the efficiency of this methodology, we applied this method to AES CPU based program obtained from the Internet web site. Simply changing the data structure made improved the performance of AES program. As a result, the performance of AES based GPU program is improved gradually up to 10 times. Depending on the developer's ability, additional performance improvement can be expected. The problem to be solved is heat issue, but this problem has been much improved by the development of the cooling technology.

• Nuclear Engineering and Technology
• /
• v.53 no.8
• /
• pp.2488-2498
• /
• 2021

The wall film-wise condensation plays an important role in the heat transfer processes of heat exchangers, refrigerators, and air conditioner. In the field of nuclear engineering, steam condensation is often utilized in safety systems to remove the core decay heat under both transient and accident conditions. In particular, passive containment cooling system (PCCS), are designed to ensure containment safety under severe accident conditions. A computational fluid dynamics (CFD) scale analysis has been conducted to calculate the heat transfer rate of the PCCS. However, despite the increase in computing power, there are challenges in the long-term transient simulation of containment using CFD scale codes. In this study, a heat structure coupling between the CFD and system analysis codes was performed to efficiently analyze PCCS. In addition, the component unstructured program for interfacial dynamics (CUPID) was improved to analyze the condensation behavior of ternary gas mixtures. Thereafter, the condensation heat transfer on the primary side was calculated using the improved CUPID and CFD code, whereas that on the secondary side was simulated using MARS. Both the coupled codes were validated against the CONAN facility database. Finally, conjugate heat transfer simulations with wall condensation in the presence of non-condensable gases were appropriately performed.

• Journal of Radiation Protection and Research
• /
• v.41 no.3
• /
• pp.237-244
• /
• 2016

Background: Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). Materials and Methods: This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. Results and Discussion: The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ${\pm}6%$. Conclusion: It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.11a
• /
• pp.104-104
• /
• 2018

Kori Unit 1 is the first permanent shutdown nuclear power plant in Korea and it is on June 18th, 2017. Spent fuel assemblies began to be discharged from the reactor core to the spent fuel pool(SFP) within one week after shutdown of Kori unit 1 and the campaign was completed on June 27th, 2017. The total number of spent nuclear fuel assemblies in SFP of Kori Unit-1 is 485 and their discharging date is different respectively. So, decay heat was evaluated considering the actual enrichment, operation history and cooling time of the spent fuel assemblies stored in SFP of the Kori Unit-1. The code used in the evaluation is the ORIGEN-based CAREPOOL system developed by KHNP. Decay heat calculation of PWR fuel is based on ANSI/ANS 5.1-2005, "Decay heat power in light water reactors" and ISO-10645, "Nuclear energy - Light water reactors - Calculation of the decay heat power in nuclear fuels. Also, we considered the contribution of fission products, actinide nuclides, neutron capture and radioactive material in decay heat calculation. CAREPOOL system calculates the individual and total decay heat of all of the spent fuel assemblies in SFP of Kori Unit-1. As a result, the total decay heat generated in SFP on June 28th, 2017 when the spent fuel assemblies were discharged from the reactor core, is estimated to be about 4,185.8 kw and to be about 609.5 kw on September 1st, 2018. It was also estimated that 119.6 kw is generated in 2050 when it is 32 years after the permanent shutdown. Figure 1 shows the trend of total decay heat in SFP of Kori Unit-1.

• The Journal of the Korea institute of electronic communication sciences
• /
• v.14 no.3
• /
• pp.553-558
• /
• 2019

Global auto parts companies are making efforts to develop EWP(: Electric Water Pump) which is one of the core parts of environment friendly car. In eco-friendly automobiles, an independent cooling system is used rather than a cooling system that is linked to an internal combustion engine. Therefore, the research and development of the water pump operating separately from the engine and the related production system are being actively carried out. In order to overcome the shortcoming of EWP of PPS material suitable for injection system, G company which is a global parts company that researches and develops EWP around SUS and is in the process of developing robot-based production equipment for mass production. In this paper, a heat shock simulation system is designed and implemented that works with the robot-based production system to test the performance of the produced EWP. By using this system, it is possible to test the EWP in an virtual environment similar to the actual environment, thereby reducing the defect rate of the product. At the same time, all the data produced during the entire process for testing can be stored, which can be utilized in the future development of CPS(: Cyber Physical System) of EWP system based on big data.

• Journal of international Conference on Electrical Machines and Systems
• /
• v.2 no.4
• /
• pp.460-465
• /
• 2013

Electrical equipments of nuclear power plant are divided into class 1E and non-class 1E. Electrical equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, are classified as class 1E. batteries of nuclear power plant are divided into four channels, which are physically and electrically separate and independent. The battery bank of class 1E DC power system of the nuclear power plant use lead-acid batteries in present. The lead acid battery, which has a high energy density, is the most popular form of energy storage. Kt factor of lead-acid battery is used to determine battery size and it is one of calculatiing coefficient for capacity. this paper analyzes Kt factor of lead-acid battery for the DC power system of nuclear power plant. In addition, correlation between Kt parameter and peukert's exponent of lead-acid battery for nuclear plant are discussed. The analytical results contribute to optimize of determining size Lead-acid battery bank.

• Proceedings of the KSME Conference
• /
• 2004.11a
• /
• pp.383-387
• /
• 2004

The Fuel Test Loop (FTL) consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). The test condition in IPS such as pressure, temperature and quality of the main cooling water, can be controlled by the OPS. The FTL has been developed to be able to irradiate three pins to the core irradiation hole (IR1 hole) by considering for its utility and user's irradiation requirement. The IPS vessel assembly (IVA) consists of IPS head, outer pressure vessel, inner pressure vessel, inner assembly and test fuel carrier. The IVA is approximately 5.6 m long and fits within a 74 mm in diameter envelope over the full height of the chimney. Above the top of the chimney, the head of the IPS is enlarged to allow the closure flanges and pipe work connections. IVA was designed to test the CANDU and PWR nuclear fuel pin together. Specially, wished to minimize interference by nuclear fuel change in design and synthesize these items and shape design for IVA.