• Progress in Superconductivity and Cryogenics
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• v.25 no.3
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• pp.13-17
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• 2023

Magnetic separation technology for small paramagnetic particles has been desired for the volume reduction of contaminated soil from the Fukushima nuclear power plant accident and for the separation of scale and crud from nuclear power plants. However, the magnetic separation for paramagnetic particles requires a superconducting high gradient magnetic separation system applied, hence expanding the bore diameter of the magnets is necessary for mass processing and the initial and running costs would be enormous. The use of high magnetic fields makes safe onsite operation difficult, and there is an industrial need to increase the magnetic separation efficiency for paramagnetic particles in as low a magnetic field as possible. Therefore, we have been developing a magnetic separation system combined with a selection tube, which can separate small paramagnetic particles in a low magnetic field. In the previous technique we developed, a certain range of particle size was classified, and the classified particles were captured by magnetic separation. In this new approach, the fluid control method has been improved in order to the selectively classify particles of various diameters by using a multi-stage selection tube. The soil classification using a multi-stage selection tube was studied by calculation and experiment, and good results were obtained. In this paper, we report the effectiveness of the multi-stage selection tube was examined.

• Journal of Radiation Protection and Research
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• v.45 no.3
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• pp.101-107
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• 2020

Background: An important lesson learned from the Fukushima accident is that the transition to the mid- and long-term phases from the emergency-response phase requires less than a year, which is not very long. It is necessary to know how much radioactive material has been deposited in an urban area to establish mid- and long-term countermeasures after a radioactive accident. Therefore, an urban deposition model that can indicate the site-specific characteristics must be developed. Materials and Methods: In this study, the generalized urban deposition velocity and the subsequent variation in radionuclide contamination were estimated based on the characteristics of the Korean urban environment. Furthermore, the application of the obtained generalized deposition velocity in a hypothetical scenario was investigated. Results and Discussion: The generalized deposition velocities of ¹³⁷Cs, ¹⁰⁶Ru, and ¹³¹I for each residence type were obtained using three-dimensional (3D) modeling. For all residence types, the deposition velocities of ¹³¹I are greater than those of ¹⁰⁶Ru and ¹³⁷Cs. In addition, we calculated the generalized deposition velocities for each residential types. Iodine was the most deposited nuclide during initial deposition. However, the concentration of iodine in urban environment drastically decreases owing to its relatively shorter half-life than ¹⁰⁶Ru and ¹³⁷Cs. Furthermore, the amount of radioactive material deposited in nonresidential areas, especially in parks and schools, is more than that deposited in residential areas. Conclusion: In this study, the generalized urban deposition velocities and the subsequent deposition changes were estimated for the Korean urban environment. The 3D modeling was performed for each type of urban residential area, and the average deposition velocity was obtained and applied to a hypothetical accident. Based on the estimated deposition velocities, the decision-making systems can be improved for responding to radioactive contamination in urban areas. Furthermore, this study can be useful to predict the radiological dose in case of large-scale urban contamination and can support decision-making for long-term measurement after nuclear accident.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.423-434
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• 2016

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.240-252
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• 2015

In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of in-vessel retention for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of in-vessel retention against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.

• Composites Research
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• v.29 no.1
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• pp.40-44
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• 2016

Nuclear fuel cladding used in a nuclear power plant must possess superior oxidation resistance in the coolant atmosphere of high temperature/high pressure. However, as was the case for the critical LOCA (loss-of-coolant accident) accident that took place in the Fukushima disaster, there is a risk of hydrogen explosion when the nuclear fuel cladding and steam reacts dramatically to cause a rapid high-temperature oxidation accompanied by generation of a huge amount of hydrogen. Hence, an active search is ongoing for an alternative material to be used for manufacturing of nuclear fuel cladding. Studies are currently aimed at improving the safety of this cladding. In particular, ceramic-based nuclear fuel cladding, such as SiC, is receiving much attention due to the excellent radiation resistance, high strength, chemical durability against oxidation and corrosion, and excellent thermal conduction of ceramics. In the present study, cladding with $SiC_f/SiC$ protective films was fabricated using a process that forms a matrix phase by polymer impregnation of polycarbosilane (PCS) after filament-winding the SiC fiber onto an existing Zry-4 cladding tube. It is analyzed the oxidation and microstructure of the metal cladding with $SiC_f/SiC$ composite protective films using a drop tube furnace for thermal shock test.

• Journal of the Korean Society of Radiology
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• v.10 no.1
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• pp.53-58
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• 2016

March 11, 2011 Fukushima nuclear power plant accident has occurred, the last four years, his is the current state of anxiety remains on the consumer. Therefore, 2015 the current progress continues to have continued research for the purpose of resolving the insecurity of fishery in the country. We measured the radiation of the origin of fishery products, by 2015, research was carried out through a number of references to get additional data or studies that influence this to some degree. Create a consumer and one for the seller for the survey to find out the thoughts of both consumers and sellers on fisheries was to create the statistics by direct survey in 2014 year survey meter using PM1405 equipment dynamics of several species of aquatic origin in accordance with the (Taiwan, China, Russia), mackerel, pomfret, hairtail, saury, shrimp, squid measurements of radiation were investigated. The measurement in 2014 was $0.043{\sim}0.073{\mu}Sv/h$. The seller influenced the survey (90%) on fishery products sales, consumer safety, without this fishery (90%), radioactive contamination very high (28%) and the polarization was not as nearly 72%. The study of the stability of the radiation determined by the radiation levels of the aquatic marine products in a comparison the radiation levels result was determined in ICRP below the standard value (1 mSv/y).

• Journal of Environmental Policy
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• v.13 no.4
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• pp.79-105
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• 2014

Natech risk is a type of complex disasters that natural hazards trigger technological disaster or industrial accidents. Research on Natech risk has been started from the mid-1990s in European countries and the Unites States, and drawn much more attention after the Fukushima nuclear accident caused by the 2011 East Japan earthquake. While early studies on Natech risk have focused on the causal natural hazards and possibility to occur, and the resulting spill of hazardous materials from the perspective of science and engineering, the recent research interests lie on effective Natech risk management. Especially, emphasizing the difference of Natech risk management from traditional disaster management, issues of uncertainty management, integration between natural disaster and technological disaster, and responsibility, has been drawn attention. In Korea, Natech risk has not been introduced as a research topic. Although some regulatory improvements have been made in nuclear safety and chemical Substance management after the Fukushima disaster, the potential impact of natural hazards in these areas has not been considered yet. It is necessary to raise the issues of Natech risk management in research and policy areas through active discussion and interdisciplinary approaches.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.605-612
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• 2013

Removal of decay heat from an operating reactor during a prolonged station blackout condition is a big concern for reactor designers, especially after the recent Fukushima accident. In the case of a prolonged station blackout condition, heat removal is possible only by passive means since no pumps or active systems are available. Keeping this in mind, the AHWR has been designed with many passive safety features. One of them is a passive means of removing decay heat with the help of Isolation Condensers (ICs) which are submerged in a big water pool called the Gravity Driven Water Pool (GDWP). The ICs have many tubes in which the steam, generated by the reactor core due to the decay heat, flows and condenses by rejecting the heat into the water pool. After condensation, the condensate falls back into the steam drum of the reactor. The GDWP tank holds a large amount of water, about 8000 $m^3$, which is located at a higher elevation than the steam drum of the reactor in order to promote natural circulation. Due to the recent Fukushima type accidents, it has been a concern to understand and evaluate the capability of the ICs to remove decay heat for a prolonged period without escalating fuel sheath temperature. In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS) tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.

• Journal of Radiation Protection and Research
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• v.46 no.1
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• pp.14-23
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• 2021

Background: For radiological protection and control, the International Commission on Radiological Protection (ICRP) provides the nominal risk coefficients related to radiation exposure, which can be extrapolated using the excess relative risk and excess absolute risk obtained from the Life Span Study of atomic bomb survivors in Hiroshima and Nagasaki with the dose and dose-rate effectiveness factor (DDREF). Materials and Methods: Since it is impossible to directly estimate the radiation risk at doses less than approximately 100 mSv only from epidemiological knowledge and data, support from radiation biology is absolutely imperative, and thus, several national and international bodies have advocated the importance of bridging knowledge between biology and epidemiology. Because of the accident at the Tokyo Electric Power Company (TEPCO)'s Fukushima Daiichi Nuclear Power Station in 2011, the exposure of the public to radiation has become a major concern and it was considered that the estimation of radiation risk should be more realistic to cope with the prevailing radiation exposure situation. Results and Discussion: To discuss the issues from wide aspects related to radiological protection, and to realize bridging knowledge between biology and epidemiology, we have established a research group to develop low-dose and low-dose-rate radiation risk estimation methodology, with the permission of the Japan Health Physics Society. Conclusion: The aim of the research group was to clarify the current situation and issues related to the risk estimation of low-dose and low-dose-rate radiation exposure from the viewpoints of different research fields, such as epidemiology, biology, modeling, and dosimetry, to identify a future strategy and roadmap to elucidate a more realistic estimation of risk against low-dose and low-dose-rate radiation exposure.

• Journal of Energy Engineering
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• v.25 no.1
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• pp.43-47
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• 2016

KHNP CRI has developed APR+ nuclear power plant since 2007, which is GEN III+ model with 1500 MWe capacity. To develop safer nuclear power plant than APR1400, we investigated advanced design features of ALWR being constructed in Korea and being developed/constructed in foreign countries. We applied the advanced design features and lessons learned from Fukushima accident to develop APR+ standard design suitable for both domestic construction and overseas construction business. One economic assessments have performed during safety design improvement phase(2013.1 ~ 2015.12) of APR+. The result of the economic analysis for APR+ safety inhancement design showed that APR+ N-th plant is about 39.2% more economical than coal-fired 1,000MW power plant. Also APR+ plant is more cost advantage over foreign advanced nation ALWRs.