• Journal of the Korea Concrete Institute
• /
• v.28 no.6
• /
• pp.673-683
• /
• 2016

In recent years, frequent terror or military attack by explosion, impact, fire accidents have occurred. Particularly, World Trade Center collapse and US Department of Defense Pentagon attack on Sept. 11 of 2001. Also, nuclear power plant incident on Mar. 11 of 2011. These attacks and incidents were raised public concerns and anxiety of potential terrorist attacks on major infrastructures and structures. Therefore, the extreme loading researches were performed of prestressed concrete (PSC) member, which widely used for nuclear containment vessel and gas tank. In this paper, to evaluate the blast resistance capacity and its protective performance of bi-directional unbonded prestressed concrete member, blast tests were carried out on $1,400{\times}1,000{\times}300mm$ for reinforced concrete (RC), prestressed concrete without rebar (PSC), prestressed concrete with rebar (PSRC) specimens. The applied blast load was generated by the detonation of 55 lbs ANFO explosive charge at 1.0 m standoff distance. The data acquisitions not only included blast waves of incident pressure, reflected pressure, and impulse, but also included displacement, acceleration, and strains at steel, concrete, PS tendon. The results can be used as basic research references for related research areas, which include protective design and blast simulation under blast loading.

• Journal of Radiation Protection and Research
• /
• v.30 no.1
• /
• pp.17-25
• /
• 2005

Since the safety of nuclear power plant has been becoming a big social issue the exposure dose of radiation for workers has been one of the important factors concerning the safety problem. The existing calculation methods of radiation dose used in the planning of radiation work assume that dose rate does not depend on the location within a work space thus the variation of exposure dose by different work path is not considered. In this study, a modified numerical method was presented to estimate the exposure dose during radiation work in radwaste storage considering the effects of the distance between a worker and sources. And a new numerical algorithm was suggested to search the optimal work path minimizing the exposure dose in pre-defined work space with given radiation sources. Finally, a virtual work simulation program was developed to visualize the exposure dose of radiation doting radiation works in radwaste storage and provide the capability of simulation for work planning. As a numerical example, a test radiation work was simulated under given space and two radiation sources, and the suggested optimal work path was compared with three predefined work paths. The optimal work path obtained in the study could reduce the exposure dose for the given test work. Based on the results, tile developed numerical method and simulation program could be useful tools in the planning of radiation work.

• Journal of Radiation Protection and Research
• /
• v.42 no.1
• /
• pp.42-47
• /
• 2017

Background: One of the most urgent issues following the accident at the Fukushima Daiichi nuclear power plant (FDNPP) was the remediation of the land, in particular, for residential area contaminated by the radioactive materials discharged. In this study, the effect of decontamination on reduction of ambient dose equivalent outdoors and indoors was evaluated. The latter is essential for residents as most individuals spend a large portion of their time indoors. Materials and Methods: From December 2012 to November 2014, thirty-seven Japanese single-family detached wooden houses were investigated before and after decontamination in evacuation zones. Outdoor and indoor dose measurements (n = 84 and 114, respectively) were collected based on in situ measurements using the NaI (Tl) scintillation surveymeter. Results and Discussion: The outdoor ambient dose equivalents [$H^*(10)_{out}$] ranged from 0.61 to $3.71{\mu}Sv\;h^{-1}$ and from 0.23 to $1.32{\mu}Sv\;h^{-1}$ before and after decontamination, respectively. The indoor ambient dose equivalents [$H^*(10)_{in}$] ranged from 0.29 to $2.53{\mu}Sv\;h^{-1}$ and from 0.16 to $1.22{\mu}Sv\;h^{-1}$ before and after decontamination, respectively. The values of reduction efficiency (RE), defined as the ratio by which the radiation dose has been reduced via decontamination, were evaluated as $0.47{\pm}0.13$, $0.51{\pm}0.13$, and $0.58{\pm}0.08$ ($average{\pm}{\sigma}$) when $H^*(10)_{out}$ < $1.0{\mu}Sv\;h^{-1}$, $1.0{\mu}Sv\;h^{-1}$ < $H^*(10)_{out}$ < $2.0{\mu}Sv\;h^{-1}$, and $2.0{\mu}Sv\;h^{-1}$ < $H^*(10)_{out}$, respectively, indicating the values of RE increased as $H^*(10)_{out}$ increased. It was found that the values of RE were $0.53{\pm}0.12$ outdoors and $0.41{\pm}0.09$ indoors, respectively, indicating RE was larger outdoors than indoors. Conclusion: Indoor dose is essential as most individuals spend a large portion of their time indoors. The difference between outdoors and indoors should be considered carefully in order to estimate residents' exposure dose before their returning home.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.3
• /
• pp.201-209
• /
• 2016

The object of this study was to evaluate the separation distance from a temporary storage facility satisfying the dose criteria. The calculation of ambient dose rates took into account cover soil thickness, facility size, and facility type by using MCNPX code. Shielding effects of cover soil were 68.9%, 96.9% and 99.7% at 10 cm, 30 cm and 50 cm respectively. The on-ground type of storage facility had the highest ambient dose rate, followed by the semi-ground type and the underground type. The ambient dose rate did not vary with facility size (except $5{\times}5{\times}2m\;size$) due to the self-shielding of decontamination waste in temporary storage. The separation distances without cover soil for a $50{\times}50{\times}2m\;size$ facility were evaluated as 14 m (minimum radioactivity concentration), 33 m (most probably radioactivity concentration), and 57 m (maximum radioactivity concentration) for on-ground storage type, 9 m, 24 m, and 45 m for semi-underground storage type, and 6 m, 16 m, and 31 m for underground storage type.

• Journal of Radiation Protection and Research
• /
• v.47 no.4
• /
• pp.204-213
• /
• 2022

Background: The accident at the Fukushima Daiichi nuclear power plant increased the level of anxiety related to the radioactive contamination of various foods sourced in Japan. Particularly, after the accident, the detection of artificial radionuclides in locally produced foods raised food safety concerns. In this study, the radioactivity concentrations and annual ingestions of ⁴⁰K and ¹³⁷Cs in food products commonly and frequently consumed by the general public were investigated, and the annual effective dose of each was evaluated. Materials and Methods: The 2016-2018 data from the Radiation Safety Management Report released by the Korea Nuclear Safety Technology Center was referenced for the evaluation of the amounts of ⁴⁰K and ¹³⁷Cs contained in food. Using the food-ingestion survey mentioned above as a reference, we selected 62 foods to include in our radioactivity concentration and dose assessment. We also developed a questionnaire and evaluated the responses from the subjects who answered the questionnaire. Results and Discussion: The radioactivity concentration of ¹³⁷Cs was found to be close to or below the level of minimum detectable activity. Additionally, the annual ingestion of 62 foods was 294.77 kg/yr, the effective doses from ⁴⁰K and ¹³⁷Cs were 136.4 and 0.163 μSv/yr, respectively. Conclusion: Thus, the findings confirmed that the effective dose from ⁴⁰K and ¹³⁷Cs in food tends to be lower than the effective dose limit of 1 mSv/yr suggested by the International Commission on Radiological Protection (ICRP) Publication 60. The questionnaire developed in this study is expected to be useful for estimating the annual effective dose status of Korean adults who consume foods containing ⁴⁰K and ¹³⁷Cs.

• Journal of Nutrition and Health
• /
• v.56 no.5
• /
• pp.455-468
• /
• 2023

Purpose: Abeliophyllum distichum (A.distichum) is a plant native to Korea. In this study, we investigated the mechanism of antioxidant and anti-inflammatory effects of the leaf extract of A.distichum. Methods: The antioxidant capacity of the A.distichum leaf extract was determined based on the total polyphenol content, 2,2-diphenyl-1-picrylhydrazyl (DPPH) assay, 2,2'-azino-bis (3-ethylbenzothiazoline-6-sulfonic acid) (ABTS) assay, and the ferric reducing antioxidant power (FRAP) assay. The anti-inflammatory effects of the A.distichum leaf extract were evaluated by measuring the production of nitric oxide (NO) and the expression levels of proinflammatory cytokines including tumor necrosis factor (TNF)-α, interleukin (IL)-1β, and IL-6 using the enzyme-linked immunosorbent assay (ELISA) and reverse transcription quantitative real-time PCR (RT-qPCR). In addition, the expression of heme oxygenase-1 (HO-1), nuclear transcription factor-erythroid 2 related factor (Nrf2), inducible nitric oxide synthase (iNOS), and cyclooxygenase 2 (COX-2), as well as the activation of nuclear factorkappa B (NF-ĸB) were examined using the western blot analysis. Results: The total polyphenol content of the A.distichum leaf extract was 329.89 ± 30.17 gallic acid equivalents mg/g and the DPPH and ABTS scavenging activities were 55% and 70%, respectively. Additionally, the FRAP value of the extract was 743.68 ± 116.59 mg/mL. After 12-hour treatment with the A.distichum leaf extract, there was a tendency for the Nrf2 expression to increase, and the expression of HO-1 was significantly elevated in the RAW264.7 cells. The A.distichum leaf extract treatment resulted in decreased levels of NO, TNF-α, IL-6, and IL-1β, as well as reduced expression of iNOS and COX-2, along with inhibition of NF-κB activation in lipopolysaccharide-stimulated RAW264.7 cells. Conclusion: These results suggest that the A.distichum leaf extract exerts antioxidative and anti-inflammatory effects by upregulating the expression of HO-1 and downregulating NF-κB activation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.3
• /
• pp.199-212
• /
• 2007

If a type IP-2 transport package were to be subjected to a free drop test and a penetration test under the normal conditions of transport, it should prevent a loss or dispersal of the radioactive contents and a more than 20% increase in the maximum radiation level at any external surface of the package. In this paper, we suggested the analytic method to evaluate the structural safety of a type IP-2 transport package using a thick steel plate for a structure part and a bolt for tying a bolt. Using an analysis a loss or dispersal of the radioactive contents and a loss of shielding integrity were confirmed for two kinds of type IP-2 transport packages to transport radioactive waste drums from a waste facility to a temporary storage site in a nuclear power plant. Under the free drop condition the maximum average stress at the bolts and the maximum opening displacement of a lid were compared with the tensile stress of a bolt and the steps in a lid, which were made to avoid a streaming radiation in the shielding path, to evaluate a loss or dispersal of radioactive waste contents. Also a loss of shielding integrity was evaluated using the maximum decrease in a shielding thickness. To verify the impact dynamic analysis for free drop test condition and evaluate experimentally the safety of two kinds of type IP-2 transport packages, free drop tests were conducted with various drop directions. For the tests we examined the failure of bolts and the deformation of flange to evaluate a loss or dispersal of radioactive material and measured the shielding thickness using a ultrasonic thickness gauge to assess a loss of shielding integrity. The strains and accelerations acquired from tests were compared with those by analyses to verify the impact dynamic analysis. The analytic results were larger than the those of test so that the analysis showed the conservative results. Finally, we evaluated the safety of the type IP-2 transport package under the stacking test condition using a finite element analysis. Under the stacking test condition, the maximum Tresca stress of the shielding material was 1/3 of the yielding stress. Two kinds of a type IP-2 transport package were safe for the free drop test condition and the stacking test condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.4
• /
• pp.319-327
• /
• 2010

A nuclear plant ESF ACS simulator was designed, built, and verified to perform experiment related to ESF ACS of nuclear power plants. The dimension of 3D CAD model was based on drawings of the main control room(MCR) of Yonggwang units 5 and 6. The CFD analysis was performed based on the measurement of the actual flow rate of ESF ACS. The air flowing in ACS was assumed to have $30^{\circ}C$ and uniform flow. The flow rate across the HEPA filter was estimated to be 1.83 m/s based on the MCR ACS flow rate of 12,986 CFM and HEPA filter area of 9 filters having effective area of $610{\times}610mm^2$ each. When MCR ACS was modeled, air flow blocking filter frames were considered for better simulation of the real ACS. In CFD analysis, the air flow rate in the lower part of the active carbon adsorber was simulated separately at higher than 7 m/s to reflect the measured value of 8 m/s. Through the CFD analyses of the ACSes of fuel building emergency ventilation system, emergency core cooling system equipment room ventilation cleanup system, it was confirmed that all three EFS ACSes can be simulated by controlling the flow rate of the simulator. After the CFD analysis, the simulator was built in nuclear grade and its reliability was verified through air flow distribution tests before it was used in main tests. The verification result showed that distribution of the internal flow was uniform except near the filter frames when medium filter was installed. The simulator was used in the tests to confirm the revised contents in Reg. Guide 1.52 (Rev. 3).

• Fire Science and Engineering
• /
• v.26 no.6
• /
• pp.15-23
• /
• 2012

Indirect Test Method is often used instead of direct test method in test method for extinguishing performance of $CO_2$ extinguishing facility because of high cost, environment problems and difficulties of procedure. But in the danger facilities for a unit of nation, such as a petrochemical plant, a nuclear power plant, or etc. is better to verify the performance of the extinguishment through direct discharge test. In $CO_2$ extinguishing system for total flooding system installed in dangerous facilities in Korea, each protected area in surface fire and deep-seated fire had selected and verified of extinguishing performance of $CO_2$ extinguishing facilities. To get recognized as extinguishing performance, discharged $CO_2$ concentration to protected area should be equivalence with design concentration standards (NFSC and NFPA). The Design Concentration means that $CO_2$ extinguishing agent is considerate of concentration for percentage of allowance (20 %) from extinguishing concentration which available to control of flame. As test result, surface fire and deep seated fire in protected area is obtained $CO_2$ design concentration and maintained design concentration more than 20 minutes as deep-seated fire. Through this study, we introduced direct discharging test method and decision method. And furthermore, especially in the dangerous facilities as a unit of Nation, we suggested necessity about reliability of extinguishing facilities to use direct test method.

• Journal of the Korea institute for structural maintenance and inspection
• /
• v.21 no.2
• /
• pp.87-94
• /
• 2017

In the nuclear power plant, the steel or polymer liner plates are adopted to prohibit the inner concrete surface from contacting with gas or liquid materials. If there is an accident, the plate may be damaged, and, in this case, concrete shall have the final responsibility to safety requirements. In this paper, an experimental research was carried out to investigate the effects of construction joint and wet and loading conditions on the permeability of concrete. The test results showed that, under a construction joint in the wet condition, leakage of gas pressure has been started from $1kg/cm^2$. However, when there are no construction joints, it is initiated from $2kg/cm^2$. In addition, under the air dried and unloading condition, regardless of with or without the presence of the construction joint, since the gas passage that exist in concrete is constant, leakage has a constant tendency to increase. Finally, under the loading condition, as described in Reference 1, since leakage is inversely proportional to the thickness of the wall, and, considering the wall thickness of the actual plant, it is found that there will not be no problem in the sealing of the gas.