• 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.37 no.3
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• pp.116-122
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• 2012

A short-term atmospheric dispersion factor (${\chi}/Q$) is an essential element for radiological dose assessment following a hypothetical accidental releases of light-water nuclear power plants. The U. S. NRC developed PAVAN program to comply with the U. S. NRC's Regulatory Guide 1.145. Meteorological data is an essential element for atmospheric dispersion, and PAVAN uses a joint frequency distribution data, which represents the occurrence probability of wind speed and wind direction for atmospheric stability. Using the meteorological data measured at Kori and Wolsung sites for the last 5 years (from 2006 to 2010), a variety of joint frequency distribution data were prepared to evaluate ${\chi}/Q$ values with different wind speed classifications (U. S. NRC's recommendation and even distribution of occurrence probability) and periods of meteorological data to be analyzed (1 year, 2 year, 3 year, 4 year, 5 year). As a result, it was found that the influence of the wind speed classification on ${\chi}/Q$ values is little, while the influence of the periods of meteorological data to be analyzed is relatively significant, representing more than 1.5 times in the ratio of maximum to minimum values.

• Journal of Radiation Protection and Research
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• v.42 no.1
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• pp.42-47
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• 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 Energy Engineering
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• v.25 no.2
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• pp.1-20
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• 2016

A condensation heat transfer model is very important for the safety analysis of nuclear power plants. Especially, condensation under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NCGs in the vapor largely reduces the condensation rate. In this study, the condensation heat transfer model of the severe accident analysis code MELCOR 1.8.6 has been assessed using a set of condensation experiments performed under the thermal-hydraulic conditions similar to those inside a containment during design-basis accidents or severe accidents. Experiment conditions are categorized into 4 types according to the shape of the condensation surface: vertical flat plates, outer surface of vertical pipes, inner surface of vertical pipes, the inner surface of horizontal pipes. The results of the calculations show that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on inner surface of vertical pipes.

• Nuclear Engineering and Technology
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• v.27 no.2
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• pp.203-215
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• 1995

A continued operation of RCPs during a certain small break LOCA may increase unnecessary inventory loss from the RCS causing a severe core uncovery which might lead to a fuel failure. After TMI-2 accident, the CEOG developed RCP trip strategy called “Trip-Two/Leave-Two” (T2/L2) in response to NRC requests and incorporated it in the generic EPG for CE plants. The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis determines the RCP trip setpoint and demonstrates the safe operational aspects of RCP trip strategy during a small break LOCA for YGN 3&4. The trip setpoint of the first too RCPs for YGN 3&4 is calculated to be 1775 psia in pressurizer pressure based on the limiting small break LOCA with 0.15 ft$^2$ break size in the hot leg. The analysis results show that YGN 3&4 can maintain the core coolability even if the operator fails to trip the second too RCPs or trips at worst time. Also, the YGN 3&4 RCP trip strategy demonstrates that both the 10 CFR 50.46 requirements on PCT and the ANSI standards 58.8 requirements on operator action time can be satisfied with enough margin. Therefore, it is concluded that the T2/L2 RCP trip strategy with a trip setpoint of 1775 psia for YGN 3&4 can provide improved operator guidance for the RCP operation during accidents.

• Nuclear Engineering and Technology
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• v.19 no.2
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• pp.85-98
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• 1987

To analyze the aerosol dynamics in severe accidents of LMFBR, a new computer code entitled MCAD (Multicomponent Aerosol Dynamics) has been developed. The code can treat two component aerosol system using relative collision probability of each particles as sequences of accident scenarios. Coagulation and removal mechanisms incorporating Brownian diffusion and gravitational sedimentation are included in this model. In order to see the effect of particle geometry, the code makes use of the concept of density correction factor and shape factors. The code is verified using the experimental result of NSPP-300 series and compared to other code. At present, it fits the result of experiment well and agrees to the existing code. The input variables included are very uncertain. Hence, it requires uncertainty and sensitivity analysis as a supplement to code development. In this analysis, 14 variables are selected to analyze. The input variables are compounded by experimental design method and Latin hypercube sampling. Their results are applied to Response surface method to see the degree of regression. The stepwise regression method gives an insight to which variables are significant as time elapse and their reasonable ranges. Using Monte Carlo Method to the regression model of LHS, the confidence level of the results of MCAD and their variables is improved.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4348-4358
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• 2022

The thermal-hydraulics phenomena in a containment during an accident will necessarily include radiative heat transfer (i) within the gas mixture due to the high radiative absorption and emission of steam and (ii) between the gas mixture and the surrounding structures. The analysis of some previous PANDA experiments (PSI, Switzerland) demonstrated the importance of the proper modelling of radiation for the benefit of numerical simulations. These results together with dedicated scoping calculations conducted for the present experiments indicated that the radiative heat transfer is considerable, even for a very low amount of steam (≈2%). The H2P2 series conducted in the large-scale PANDA facility at the Paul-Scherrer-Institut (PSI) in the framework of the OECD/NEA HYMERES-2 project is intended to enhance the understanding of thermal radiation phenomena and to provide a benchmark for corresponding numerical simulations. Thus, the test matrix was tailored around the two opposite extremes: either gas compositions with small steam content such that radiative heat transfer phenomena can be neglected. Or gas mixtures containing larger amounts of steam, so that radiative heat transfer is expected to play a dominant role. The H2P2 series consists of 5 experiments designed to isolate the radiation phenomena from convective and diffusive effects as much as possible. One vessel with a diameter of 4 m and a height of 8 m was preconditioned with different mixtures of air / steam at room and elevated temperatures. This was followed by the build-up of a stable helium stratification at constant pressure in the upper part of the vessel. After that, helium was injected from the top into the vessel which leads to an increase of the vessel pressure and a corresponding elevation-dependent and transient rise of the gas temperature. It is shown that even the addition of small amounts of steam in the initial gas atmosphere considerably impacts the radiative heat transport throughout all phases of the experiments and markedly influences i) the monitored gas peak temperature, ii) the temperature history during the compression and iii) the following relaxation phase after the compression was stopped. These PANDA experiments are the first of its kind conducted in a large scale thermal-hydraulic facility.

• Journal of the Korean Society of Radiology
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• v.11 no.5
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• pp.321-328
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• 2017

The purpose of the study is to provide basic data for the management of individual exposure and the monitoring of natural radiation dose using D-Shuttle dosimeter (Chiyoda Technol Corporation, Tokyo, Japan). The dose was calculated using D-Shuttle dosimeter. The dose was 1.346 mSv when exposed for 400 days, the annual dose per year was 1.228 mSv/year and the average dose per hour was $0.014{\mu}Sv/hr$. Domestic individual external dose (1.295 mSv/year = Korea average natural individual external dose) and domestic additional dose per year is -0.0663 mSv/year. D-Shuttle is a personal dosimeter for radiation monitoring. It can be used as a very useful dosimeter for ALARA because of its excellent detection capability of radiation, real-time radiation exposure management, alarm function of radiation work, and efficient and easy to use personal radiation dose management.. Radiation monitoring equipment for radiation workers and local residents can be used for radiation monitoring in hospitals, industry, medical sites, nuclear accident areas and hazardous areas in non-destructive areas.

• Environmental Analysis Health and Toxicology
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• v.31
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• pp.1.1-1.13
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• 2016

Objectives To investigate the accuracy and scientific validity of the current very low risk factor for hereditary diseases in humans following exposures to ionizing radiation adopted by the United Nations Scientific Committee on the Effects of Atomic Radiation and the International Commission on Radiological Protection. The value is based on experiments on mice due to reportedly absent effects in the Japanese atomic bomb (A-bomb) survivors. Methods To review the published evidence for heritable effects after ionising radiation exposures particularly, but not restricted to, populations exposed to contamination from the Chernobyl accident and from atmospheric nuclear test fallout. To make a compilation of findings about early deaths, congenital malformations, Down's syndrome, cancer and other genetic effects observed in humans after the exposure of the parents. To also examine more closely the evidence from the Japanese A-bomb epidemiology and discuss its scientific validity. Results Nearly all types of hereditary defects were found at doses as low as one to 10 mSv. We discuss the clash between the current risk model and these observations on the basis of biological mechanism and assumptions about linear relationships between dose and effect in neonatal and foetal epidemiology. The evidence supports a dose response relationship which is non-linear and is either biphasic or supralinear (hogs-back) and largely either saturates or falls above 10 mSv. Conclusions We conclude that the current risk model for heritable effects of radiation is unsafe. The dose response relationship is non-linear with the greatest effects at the lowest doses. Using Chernobyl data we derive an excess relative risk for all malformations of 1.0 per 10 mSv cumulative dose. The safety of the Japanese A-bomb epidemiology is argued to be both scientifically and philosophically questionable owing to errors in the choice of control groups, omission of internal exposure effects and assumptions about linear dose response.

• Journal of the Korea Institute of Information and Communication Engineering
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• v.19 no.6
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• pp.1405-1410
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• 2015

In this study we designed the radiation-hardened sensor signal processing modules that can be commonly used for a variety of sensors during normal operation and even in high-radiation environments caused by an accident. First development module was designed to receive the change of the R and C value from the sensors and to process the signal as a PWM modulation scheme. This module was assessed to have ± 10% error to the Full-Scale in the radiation test in the range of 12 kGy TID. The main cause of the error was analyzed as the annealing of the common circuit in the switching element and the consequent increase in the duty ratio of the pulse width modulation circuit according to the radiation dose increasement. The redesigned module for higher radiation resistivity with Stub transistor circuit was found to have less than 5% error to the Full-scale from the radiation test results for 20.7 kGy TID range.