• Journal of Radiation Protection and Research
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• v.22 no.3
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• pp.183-193
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• 1997

The amount and kinds of radionuclide contained in waste volume should be known to prepare for occupational exposure management, perform safety assessment and finally to license a repository. Although the volume of filters and resins are small, activities of them comprise most of the radioactivity that made during power generation. This study aims at developing a method of estimating the radionuclide accumulation at the filters and resins of coolant systems. In this study, accumulated amount of radionuclides is estimated by a computer program which makes use of instantaneous decontamination factor, DF, instead of average DF. A FORTRAN program was developed for the estimation. Data from in-plant source-term measurements at Rancho-Seco nuclear power plant in the United States are employed for verification of the estimating method. And experimental data are employed, too. The instantaneous-DF-method showed smaller error than the average-DF-method. Accumulated amount of radionuclides can be calculated with only the DF and the radionuclide concentration, which are measured periodically according to the operating guide. However, especially, when the operating condition of nuclear power plant changes rapidly, the measuring term of DF and radionuclide should be shortened to ensure the accurate estimation.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.74-83
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• 1995

The accurate determination of the fast neutron flux/fluence onto the pressure vessel is an essential part of the reactor pressure vessel surveillance program. It has been reported recently that the iron cross section data in ENDF/B versions III through V might underestimate the flux/fluence of fast neutrons in steel structures such as reactor pressure vessel. In this study, for the comparison of iron data of ENDF/B-IV and VI we produced two 47-group cross section sets, CXFe-IV and CXFe-Ⅵ, which are based on Yonggwang nuclear unit-3/4 model and the iron data of ENDF/B-IV and VI, respectively. A comparison was made of the results obtained from DOT4.3 calculation using CXFe-IV and CXFe-VI. From the results, it was found that the fast flux(E 〉 1.0 MeV), which is important for the pressure vessel embrittlement analysis, increases by about 7.6％ at the inner wall and 20％ at the outer wall of the vessel, if the iron data are used from ENDF/B-VI instead of ENDF/B-IV.

• Journal of the Society of Disaster Information
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• v.5 no.1
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• pp.78-100
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• 2009

This study is to protect peoples' life, minimize the property damage by coping with threats quickly and take more preventive measures in advance against nuclear bomb, CBR, and potential explosive. For this, CBRNE(Chemical, Biological, Radiological, Nuclear, Explosive) program research was used. Thanks to advance in technology, terrorist groups and even individuals make or keep nuclear and CBR weapons. And also it's likely that disaster and threats from a toxic gas, acute pathogens, accidents in the nuclear power plants and a high explosive could be happened a lot. Recently more organized terrorist groups maintain random attacks for unspecified individuals and also it's highly likely that a large-scale terrorist attack by WMD and CBRNEwill be done. To take strict measures against CBRNE attacks by terrorists is on the rise as an urgent national task. Moreover biological weapons are relatively easy and inexpensive to obtain or produce and cause mass casualties with a small amount. For this reason, more than 25 countries have already possessed them. In the 21 st century, the international safety environment marks the age of complicated threats : transnational threats such as comprehensive security and terror, organized crime, drug smuggling, illegal trade of weapons of mass destruction, and environmental disruption along with traditional security threats. These cause military threats, terror threats, and CBRNE threats in our daily life to grow. Therefore it needs to come up with measures in such areas as research development, policy, training program. Major industrial nations on CBRNE like USA, Canada, Switzerland, and Israel have implemented various educational programs. These researches could be utilized as basic materials for drawing up plans for civil defense, emergency services and worldwide countermeasures against CBRNE.

• Journal of Radiation Industry
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• v.5 no.2
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• pp.185-189
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• 2011

As nuclear industry has been developed, a various types of radiological contamination has occurred. After 9.11 terror in U.S.A., it has been concerned that terrorists' active area has been enlarged to use nuclear or radioactive substance. Recently, the most powerful earth-quake stroke, which triggered a massive tsunami in Japan and then Fukushima nuclear power plant reactor has suffered from a serious accident in history. The Fukushima reactor accident has occurred an anxiety of radiation leaks and about 170,000 people have been evacuated from the accidental area near the nuclear power plant. For these reasons, a social chaos can be occurred if radiological contamination occurs to the supply system for the drinking water. As such, the establishment of the radiation monitoring system for the city main water system is compelling for the national security. In this study, a feasibility test of radiation monitoring system which consists of unified hybrid-type radiation detectors was experimented for multi detection system by using gamma-ray imaging. The hybrid-type radiation sensors were fabricated with CsI(Tl) scintillators and photodiodes. A preamplifier and amplifier was also fabricated and assembled with the sensor in the shielding case. For the preliminary test of detection of radiological contamination in the river, multi CsI(Tl)-PIN photodiode radiation detectors and $^{137}Cs$ gamma-ray source were used. The DAQ was done by Linux based ROOT program and NI DAQ system with Labview program. The simulated contamination was assumed to be occurred at Gapcheon river in Daejeon city. Multi CsI(Tl)-PIN photodiode radiation detectors were positioned at the Gapcheon river side. Assuming that the radiological contaminations flows in the river the $^{137}Cs$ gamma-ray source has been moved and then, the contamination region was reconstructed.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2881-2892
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• 2024

Annular flow refers to a special type of two-phase flow pattern in which liquid flows as a thin film at the periphery of a pipe, tube, or conduit, and gas with relatively high velocity flows at the center of the flow section. This gas also includes dispersed liquid droplets. The liquid film flow rate continuously changes inside the tube due to two processes-entrainment and deposition. To determine the liquid holdup, pressure drop, the onset of dryout, and heat transfer characteristics in annular flow, it is important to have proper knowledge of flow characteristics. Especially a better understanding of entrainment fraction is important for the heat transfer and safe operation of two-phase flow systems operating in an annular two-phase flow regime. Therefore, the objective of this work is to develop a computational model for the simulation of the annular two-phase flow regime and assess the various existing models for the entrainment rate. In this work, Computational Fluid Dynamics (CFD) in ANSYS FLUENT has been applied to determine annular flow characteristics such as liquid film thickness, film velocity, entrainment rate, deposition rate, and entrainment fraction for various gas-liquid flow conditions in a vertical upward tube. The gas core with droplets was simulated using the Discrete Phase Model (DPM) which is based on the Eulerian-Lagrangian approach. The Eulerian Wall Film (EWF) model was utilized to simulate liquid film on the tube wall. Three different models of Entrainment rate were implemented and assessed through user-defined functions (UDF) in ANSYS. Finally, entrainment for fully developed flow was determined and compared with the experimental data available in the literature. From the simulations, it was obtained that the Bertodano correlation performed best in predicting entrainment fraction and the results were within the ±30 % limit when compared to experimental data.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2875-2880
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• 2024

The development of nuclear power plants is in three phases. The first phase is a consideration before the decision on the NPP construction program is approved, the second phase is the preparatory work for making contracts and preparing for the construction of NPP after the NPP construction policy is approved, and the third phase is contracting, licensing and building the first NPP. As a volcanically active country, Indonesia contains over 130 active volcanoes that are part of the Pacific Ring of Fire. The volcanic aspect is one of the safety factors considered while deciding the location of an NPP. Research on the potential of natural external risks to the determination of nuclear power plants in Indonesia, including the volcanic aspect, has been conducted based on the safety reference or safety guide of the IAEA and the Nuclear Energy Regulatory Body (BAPETEN) Regulation. Due to technological advancements, safety needs have evolved so the existing Indonesia National Standard (SNI) must be updated to comply with BAPETEN regulations. The substance in SNI 18-2034-1990 relating to volcanic features seems less relevant in actual conditions, given that more complete and exact criteria for determining a site guarantee the safety and health of residents and surrounding the environment site. The study intends to conduct a gap analysis of volcanic issues in SNI and volcanic regulations. The method used is identification requirements for volcanic aspects in SNI 18-2034-1990 about Determining Site of Nuclear Reactor Guidance with BAPETEN Chairman Regulation (BCR) number 4 of 2018 about Nuclear Installation Site Evaluation Safety Provisions and BCR number 5 of 2015 about Evaluation of Nuclear Installation Sites for Volcanic Aspects, and analysis uses a qualitative method of inductive techniques. The outcome of this research applies to suggesting a revision of SNI number 18-2034-1990, especially the volcanic aspect.

• The Korean Journal of Nuclear Medicine Technology
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• v.15 no.2
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• pp.88-93
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• 2011

Purpose: As to analytical method of data, the AutoQUANT software in which it is used quantitative rating of the myocardial perfusion SPECT are reported that there is a difference. Therefore the measured value error of the mutual program is expected to be generated even if the quantitative analysis is made data of the same patient. The purpose of this study is to offer the comparative analysis of myocardial extract method in Quantitative Perfusion SPECT. Materials and methods: We analyzed the 51 patients who were examined by Tc-99m MIBI gated myocardial SPECT in nuclear medicine department of Pusan National University Hospital from June to December 2010(34 men, 17 women, mean age $66.5{\pm}9.9$). We acquired the extracted image in myocardial extract protocol. QPS program that uses the AutoQUANT software measured TID(Transient Ischemic Dilation), ESD(Extent of Stress Defect), SSS(Summed Stress Score). Then analyzed the results. Results: The correlation of appyling myocardial extract is TID(r=0.98), ESD(r=0.99), SSS(r=0.99). In the 95% confidence limit, there was no satistically significant difference(TID p=0.78, ESD p=0.31, SSS p=0.19). After blinding test with a physician for making a qualitative analysis, there was no difference. Conclusion: Quantitative indices in QPS program showed good correlation and the results showed no statistically signigicant difference. The variance between method was small. therefore, the functional parameters by each method can be used interchangeably. Also, we expect patient's satisfaction.

• Journal of Radiation Protection and Research
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• v.38 no.2
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• pp.124-131
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• 2013

Utilities have tried to ensure that radiological hazards to the environment and residents are kept as low as reasonably achievable by monitoring and controlling planned releases. However, since groundwater contamination was reported to occur due to unplanned releases mostly in the United States nuclear power plants, the interest of the stakeholders has increased to a point where it is now one of the most important issues in the United States nuclear power industry. This paper aims to help to implement an effective on-site groundwater monitoring program at domestic nuclear power plants by briefing the experiences of the United States nuclear power plants on groundwater contaminations and groundwater monitoring, and responses of the United States nuclear industry and regulator body for them.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.1-6
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• 2014

PID(Proportional, Integral, Derivative) controller is one of the most used process controllers in nuclear power plants. The optimized parameter setting of process controller contributes to the stable operation and efficiency in the operating nuclear power plants. PID parameter setting is tuned when new process control system is established or process control system is changed. It is a burdensome work for I&C(Instrument and Control) engineers to tune the PID controller because it requires a lot of experience and knowledge. When the plant is in operation, inadequate PID parameter setting can be the cause of the unstable process of the plant. Therefore the results of PID parameter setting should be compared, simulated, verified and finally optimized. The practical PID tuning methods used in process controller are tuning operation calculation(Ziegler-Nicholes, Minimum TIAE, Lambda, IMC), exclusive tuning program based on computer and Matlab application. This paper introduces the various tuning methods and suggests an optimized PID tuning process in the operating nuclear power plants.