• Nuclear Engineering and Technology
• /
• v.56 no.6
• /
• pp.2011-2018
• /
• 2024

Since RF breakdown is one of the primary limitations to improving the performances of RF accelerators, extensive efforts have been dedicated to locating the breakdowns. However, most existing methods rely on specialized techniques, resulting in high financial burdens. Although the method based on transient response of transmission line (TL) is suitable for facilities with sporadic recoverable breakdowns, practical operations are susceptible to notable errors. This study revisits the fundamental theories of lossless TL and investigates the wave process to understand the characteristics of the reversed pulse induced by the breakdowns. By utilizing steadystate response of the TL and employing phasor method, we derive analytical formulas to determine the exact location of breakdowns within the faulty cell for constant-gradient TW accelerator. Furthermore, the derived formulas demonstrate their independence from RF phase, thereby distinguishing them from traditional phasebased methods. Additionally, experimental validations are conducted at the HUST injector, and the results confirm the consistency of the analysis. Thus, the proposed method represents a promising improvement over the TL-based approaches and serves as a valuable complement to current techniques. Importantly, this method demonstrates particular advantages for constructed TW accelerators seeking to achieve a balance among high performance, low costs, and compact layouts.

• Nuclear Engineering and Technology
• /
• v.55 no.6
• /
• pp.2335-2347
• /
• 2023

As medical facilities are usually built at urban areas, special concrete aggregates and evaluation methods are needed to optimize the design of concrete walls by balancing density, thickness, material composition, cost, and other factors. Carbon treatment rooms require a high radiation shielding requirement, as the neutron yield from carbon therapy is much higher than the neutron yield of protons. In this case study, the maximum carbon energy is 430 MeV/u and the maximum current is 0.27 nA from a hybrid particle therapy system. Hospital or facility construction should consider this requirement to design a special heavy concrete. In this work, magnetite is adopted as the major aggregate. Density is determined mainly by the major aggregate content of magnetite, and a heavy concrete test block was constructed for structural tests. The compressive strength is 35.7 MPa. The density ranges from 3.65 g/cm³ to 4.14 g/cm³, and the iron mass content ranges from 53.78% to 60.38% from the 12 cored sample measurements. It was found that there is a linear relationship between density and iron content, and mixing impurities should be the major reason leading to the nonuniform element and density distribution. The effect of this nonuniformity on radiation shielding properties for a carbon treatment room is investigated by three groups of Monte Carlo simulations. Higher density dominates to reduce shielding thickness. However, a higher content of high-Z elements will weaken the shielding strength, especially at a lower dose rate threshold and vice versa. The weakened side effect of a high iron content on the shielding property is obvious at 2.5 µSv=h. Therefore, we should not blindly pursue high Z content in engineering. If the thickness is constrained to 2 m, then the density can be reduced to 3.3 g/cm³, which will save cost by reducing the magnetite composition with 50.44% iron content. If a higher density of 3.9 g/cm³ with 57.65% iron content is selected for construction, then the thickness of the wall can be reduced to 174.2 cm, which will save space for equipment installation.

• Economic and Environmental Geology
• /
• v.54 no.2
• /
• pp.285-297
• /
• 2021

Globally, nuclear-decommissioning facilities have been increased in number, and thereby hundreds of thousands of wastes, such as concrete, soil, and metal, have been generated. For this reason, there have been numerous efforts and researches on the development of technology for volume reduction and recycling of solid radioactive wastes, and this study reviewed and examined thoroughly such previous studies. The waste concrete powder is rehydrated by other processes such as grinding and sintering, and the processes rendered aluminate (C3A), C4AF, C3S, and ��-C2S, which are the significant compounds controlling the hydration reaction of concrete and the compressive strength of the solidified matrix. The review of the previous studies confirmed that waste concretes could be used as recycling cement, but there remain problems with the decreasing strength of solidified matrix due to mingling with aggregates. There have been further efforts to improve the performance of recycling concrete via mixing with reactive agents using industrial by-products, such as blast furnace slag and fly ash. As a result, the compressive strength of the solidified matrix was proved to be enhanced. On the contrary, there have been few kinds of researches on manufacturing recycled concretes using soil wastes. Illite and zeolite in soil waste show the high adsorption capacity on radioactive nuclides, and they can be recycled as solidification agents. If the soil wastes are recycled as much as possible, the volume of wastes generated from the decommissioning of nuclear power plants (NPPs) is not only significantly reduced, but collateral benefits also are received because radioactive wastes are safely disposed of by solidification agents made from such soil wastes. Thus, it is required to study the production of non-sintered cement using clay minerals in soil wastes. This paper reviewed related domestic and foreign researches to consider the sustainable recycling of concrete waste from NPPs as recycling cement and utilizing clay minerals in soil waste to produce unsintered cement.

• Journal of Energy Engineering
• /
• v.24 no.3
• /
• pp.67-73
• /
• 2015

Almost base-loaded power plants such as flaming coal and nuclear energy require large-scale transmission facilities (LTFs) in order to send electricity to remote consumption areas. As well known, LTFs incur various social costs. However, a decentralized generation source such as integrated energy business (IEB)-based combined heat and power (CHP) plant is located in nearby electricity-consuming area, and thus it does not demand LTFs, providing the benefits from avoiding the damages caused by them. This study attempts to measure the benefits of avoiding the damages from the LTFs by the use of the contingent valuation (CV) method. To this end, a national survey of randomly chosen 1,000 households was implemented and the public's willingness to pay (WTP) for substituting consumption of electricity generated from flaming coal-fired power plant, currently a dominant generation source in Korea, with that produced from IEB-based CHP plant. The results show that the WTP for the substitution is estimated to be about 41.4 won per kWh. Considering that this value amounts to 33% of the average price of residential electricity in 2014, the external benefit of the IEB-based CHP as a decentralized generation appears to be large.

• Journal of the Korean Society of Radiology
• /
• v.14 no.5
• /
• pp.503-510
• /
• 2020

Concrete is one of the most widely used materials as the shielding structures of a nuclear facilities. It is also the most generated radioactive waste in quantity while dismantling facilities. Since the concrete captures neutrons and generates various radionuclides, radiation measurement and analysis of the sample was fulfilled prior to dismantle facilities. An HPGe detector is used in general for the radiation measurement, and effective correction factors such as geometrical correction factor, self-absorption correction, and absolute detector efficiency have to be applied to the measured data to decide exact radioactivity of the sample. Correction factors are obtained by measuring data using a standard source with the same geometry and chemical states as the sample under the same measurement conditions. However, it is very difficult to prepare standard concrete sources because concrete is limited in pretreatment due to various constituent materials and high density. In addition, the concrete sample obtained by core drill is a volumetric source, which requires geometric correction for sample diameter and self absorption correction for sample density. Therefore in recent years, many researchers are working on the calculation of effective correction factors using Monte carlo simulation instead of measuring them using a standard source. In this study we calculated, using Geant4, one of the Monte carlo codes, the correction factors for the various diameter and density of the concrete core sample at the gamma ray energy emitted from the nuclides ¹⁵²Eu and ⁶⁰Co, which are the most generated in radioactive concrete.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.4
• /
• pp.385-410
• /
• 2016

The disposal facility in Gyeongju is planning to dispose of 800,000 packages of low- and intermediate- level radioactive waste. This facility will be developed as a complex disposal facility that has various types of disposal facilities and accompanying management. In this study, based on the comprehensive development plan of the disposal facility, a preliminary post-closure safety assessment is performed to predict the phase development of the total capacity for the 800,000 packages to be disposed of at the site. The results for each scenario meet the performance target of the disposal facility. The assessment revealed that there is a significant impact of the inventory of intermediate-level radionuclide waste on the safety evaluation. Due to this finding, we introduce a disposal limit value for intermediate-level radioactive waste. With stepwise development of safety case, this development plan will increase the safety of disposal facilities by reducing uncertainties within the future development of the underground silo disposal facilities.

• Journal of the Korean Society of Radiology
• /
• v.16 no.6
• /
• pp.741-749
• /
• 2022

Radiography-Testing that verify the quality of welding structures without destruction are overwhelmingly used in industries, but many safety precautions are required as radiation is used. The workers for Radiography-Testing perform the inspection by moving the Iridium-192 radiation source embedded in the transport container of the gamma-ray irradiator within or outside the facility. The general facility is completely blocked about radiation from the outside with thick concrete, but if it is difficult for worker to handle object of inspection, facilities ceiling can be opened. A general facility may be constructed using a theoretical dose evaluation method because all exterior facilities are blocked, but if the ceiling is open, it is not appropriate to evaluate radiation safety with a simple theoretical calculation method due to the skyshine effect. Therefore, in this study, the radiation safety of the facility was evaluated in the actual field through an ion chamber survey-meter and an accumulated dose-meter called as OSLD, and the actual evaluation environment was modeled and evaluated using the Monte Carlo simulation code as FLUKA. According to the direction of the irradiation, the radiation dose at the facility boundary was difficult to meet the standards set by the regulatory authority, and radiation safety could be secured through additional methods. In addition, it was confirmed that the simulation results using the Iridium-192 source were valid evaluation with the actual measured results.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.4
• /
• pp.375-387
• /
• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Journal of the Korean housing association
• /
• v.17 no.2
• /
• pp.151-158
• /
• 2006

Medical technology being developed, the increase of the aged population brings about many changes in financial standard, consciousness and lifestyle. And the increase of a nuclear family and a professional woman makes their family not to be able to support them anymore. Because aged people also don't want to rely on their family, aged people households are growing gradually. These causes make a house and a living environment of aged people to new social problems and these became elements to determine the living quality. In case of advanced nations a house considered of the physical character of aged people has been planned and they can live with various services in their houses and community that they are living without moving: But a hou! se for aged people is initial stage in Korea. And most facilities are poor and just for protection and accommodation. Although concerning a house for the aged according to revised the aged welfare law, companies are waiting to build because of problems of cognition and regulation. Therefore the plan considered of the character of the aged is being needed for the independent and comfortable living of the aged.

• Proceedings of the Korean Institute of Interior Design Conference
• /
• 2006.05a
• /
• pp.254-258
• /
• 2006

Decrease of children caused by economic growth, scientific and technological advancement, long life, spread of nuclear families, increase of women's entry in public affairs, etc. in modern society has given rise to aging societies as another social problem. This has resulted in the advent of problems of the aged and necessity of geriatric hospitals specializing in providing medical services for the old. Even though the geriatric hospitals currently operated are mainly used by aged persons, however, their color environment has been decided, for the most pin, not in view of characteristics of the old but in view of supporting families' criteria for selection of facilities. This study intends to help geriatric hospital designers recognize importance of color environment considering characteristics of the old and select the relevant colors In designing geriatric hospitals in future to elevate remedial value and prevent accidents in space use. To this end, this study suggests problems of color environment found in surveys of the existing geriatric hospitals currently operated throughout the nation and further make some proposals for improvement.