• Journal of The Korean Association For Science Education
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• v.32 no.4
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• pp.775-788
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• 2012

This study was done at a scientific camp sponsored by Nara Women's University Secondary School, Japan. In this school, $10^{th}$ grade students from 4 East Asian countries: Korea, Japan, Taiwan, and Singapore, participated. We made a research on students' perceptions about nuclear energy. Sample populations include 77 students in total, with 12 Korean, 46 Japanese, 9 Taiwanese and 10 Singaporean students. Overall perceptions comparison about nuclear energy shows average values from the order of highest Korea, Taiwan, Singapore, and to lowest, Japan. We implemented a T-test to identify perception differences about nuclear energy, with one group that include 3 countries (Korea, Taiwan and Singapore) and another group that includes all the Japanese students. T-test results of perceptions about nuclear energy shows students from the 3 countries of Korea, Taiwan and Singapore having higher average than Japanese students. (p<.05). Korean average scores regarding overall perceptions about nuclear energy show as the highest in all 4 East Asian countries and also highest in all subcategories. On the contrary in Japan, they have lower and negative perceptions of nuclear energy. In spite of these facts, perceptions of Japanese students about nuclear energy seem lowest and negative mainly because of the recent Fukushima nuclear power plant disaster, caused by the tsunami and its subsequent damages and fears of radiation leaks, etc. This shows that negative information about future disasters and its resulting damages like the Chernobyl nuclear accident could influence more on people's risk perception than general information like nuclear energy-related technologies or the news that the plant is operating normally, etc. Even if the possibility of this kind of accident is very low, just one accident could bring abnormal risks to technology itself. This strong signal makes negative image and strengthens its perceptions to the people. This could bring a stigma about nuclear energy. This study shows that Government's policy about the highest priority for nuclear energy safety is most important. As long as such perception and decision are fixed, we found that it might not be easy to get changed again because they were already fortified and maintained.

• Journal of Mechanical Science and Technology
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• v.18 no.6
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• pp.990-1001
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• 2004

The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a Pressurized Water Reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.241-253
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• 2001

Under severe accidents, the pressure and temperature response has an important role for the integrity of a nuclear power plant containment. The history of the pressure and temperature is characterized by the amount and state of steam/air mixture in a containment. Recently, the heat transfer rate to the structure surface is supposed to be increased by the wavy interface formed on condensate film. However, in the calculation by using CONTAIN code, the condensation heat transfer on a containment wall is calculated by assuming the smooth interface and has a tendency to be underestimated for safety. In order to obtain the best- estimate heat transfer calculation, we investigated the condensation heat transfer model in CONTAIN 1.2 code and adopted the new forced convection correlation which is considering wavy interface. By using the film tracking model in CONTAIN 1.2 code, the condensate film is treated to consider the effect of wavy interface. And also, it was carried out to investigate the effect of the different cell modelings - 5-cell and 10-cell modeling - for KNGR(Korean Next Generation Reactor) containment phenomena during a severe accident. The effect of wavy interface on condensate film appears to cause the decrease of peak temperature and pressure response . In order to obtain more adequate results, the proper cell modeling was required to consider the proper flow of steam/air mixture.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.59-65
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• 2015

Background: After the Fukushima accident, the importance of hazard analysis for extreme external events was raised. Methods: To analyze typhoon-induced hazards, which are one of the significant disasters of East Asian countries, a statistical analysis using the extreme value theory, which is a method for estimating the annual exceedance frequency of a rare event, was conducted for an estimation of the occurrence intervals or hazard levels. For the four meteorological variables, maximum wind speed, instantaneous wind speed, hourly precipitation, and daily precipitation, the parameters of the predictive extreme value theory models were estimated. Results: The 100-year return levels for each variable were predicted using the developed models and compared with previously reported values. It was also found that there exist significant long-term climate changes of wind speed and precipitation. Conclusion: A fragility analysis should be conducted to ensure the safety levels of a nuclear power plant for high levels of wind speed and precipitation, which exceed the results of a previous analysis.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.534-545
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• 2015

In the very high temperature reactor (VHTR), which is a next generation nuclear reactor system, ceramics are used as a fuel coating material and graphite is used as a core structural material. Even if a depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change only slowly. This is because the thermal capacity of the core is so high. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel. The objectives of this study are to investigate the heat transfer characteristics of natural convection of a one-side heated vertical channel with inserted porous materials of high porosity and also to develop the passive cooling system for the VHTR. An experiment was carried out using a one-side heated vertical rectangular channel. To obtain the heat transfer and fluid flow characteristics of the vertical channel with inserted porous material, we have also carried out a numerical analysis using a commercial Computational Fluid Dynamics (CFD) code. This paper describes the thermal performances of the one-side heated vertical rectangular channel with an inserted copper wire of high porosity.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.692-699
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• 2017

The determination of the shape and rising velocity of gas bubbles in a liquid pool is of great importance in analyzing the radioactive aerosol emissions from nuclear power plant accidents in terms of the fission product release rate and the pool scrubbing efficiency of radioactive aerosols. This article suggests a simple parameterization for the gas bubble rising velocity as a function of the volume-equivalent bubble diameter; this parameterization does not require prior knowledge of bubble shape. This is more convenient than previously suggested parameterizations because it is given as a single explicit formula. It is also shown that a bubble shape diagram, which is very similar to the Grace's diagram, can be easily generated using the parameterization suggested in this article. Furthermore, the boundaries among the three bubble shape regimes in the $E_o-R_e$ plane and the condition for the bypass of the spheroidal regime can be delineated directly from the parameterization formula. Therefore, the parameterization suggested in this article appears to be useful not only in easily determining the bubble rising velocity (e.g., in postulated severe accident analysis codes) but also in understanding the trend of bubble shape change due to bubble growth.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.738-745
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• 2019

Coprecipitation using hydrous ferric oxide (HFO) has been effectively used for the removal of radionuclides from radioactive wastewater. This work studied the dynamic behavior of HFO floc formation during the neutralization of acidic ferric iron in the presence of several radionuclides by using a photometric dispersion analyzer (PDA). Then the coagulation-flocculation system using HFO-anionic poly acrylamide (PAM) composite floc system was evaluated and compared in seawater and distilled water to find the effective condition to remove the target nuclides (Co-60, Mn-54, Sb-125, and Ru-106) present in wastewater generated in the severe accident of nuclear power plant like Fukushima Daiichi case. A ferric iron dosage of 10 ppm for the formation of HFO was suitable in terms of fast formation of HFO flocs without induction time, and maximum total removal yield of radioactivity from the wastewater. The settling time of HFO flocs was reduced by changing them to HFO-PAM composite floc. The optimal dosage of anionic PAM for HFO-anionic PAM floc system was approximately 1-10 ppm. The total removal yield of Mn-54, Co-60, Sb-125, Ru-106 radionuclides by the HFO-anionic PAM coagulation-flocculation system was higher in distilled water than in seawater and was more than 99%.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1146-1156
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• 2021

Single-unit probabilistic safety assessment (SUPSA) has complex Boolean logic equations for accident sequences. Multi-unit probabilistic safety assessment (MUPSA) model is developed by revising and combining SUPSA models in order to reflect plant state combinations (PSCs). These PSCs represent combinations of core damage and non-core damage states of nuclear power plants (NPPs). Since all these Boolean logic equations have complemented gates (not gates), it is not easy to generate exact Boolean solutions. Delete-term approximation method (DTAM) has been widely applied for generating approximate minimal cut sets (MCSs) from the complex Boolean logic equations with complemented gates. By applying DTAM, approximate conditional core damage probability (CCDP) has been calculated in SUPSA and MUPSA. It was found that CCDP calculated by DTAM was overestimated when complemented gates have non-rare events. Especially, the CCDP overestimation drastically increases if seismic SUPSA or MUPSA has complemented gates with many non-rare events. The objective of this study is to suggest a new quantification method named probability subtraction method (PSM) that replaces DTAM. The PSM calculates accurate CCDP even when SUPSA or MUPSA has complemented gates with many non-rare events. In this paper, the PSM is explained, and the accuracy of the PSM is validated by its applications to a few MUPSAs.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.917-925
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• 2022

Steam generator (SG) tubes in a nuclear power plant can undergo rapid changes in pressure and temperature during an accident; thus, an accurate model to predict short-term creep damage is essential. The theta (𝜃) projection method has been widely used for modeling creep-strain behavior under constant stress. However, many creep test data are obtained under constant load, so creep rupture behavior under a constant load cannot be accurately simulated due to the different stress conditions. This paper proposes a novel methodology to obtain the creep curve under constant stress using a modified 𝜃 projection method that considers the increase in true stress during creep deformation in a constant-load creep test. The methodology is validated using finite element analysis, and the limitations of the methodology are also discussed. The paper also proposes a creep-strain model for alloy 690 as an SG material and a novel creep hardening rule we call the damage-fraction hardening rule. The creep hardening rule is applied to evaluate the creep rupture behavior of SG tubes. The results of this study show its great potential to evaluate the rupture behavior of an SG tube governed by creep deformation.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1483-1490
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• 2021

Many studies are undertaken into nuclear power plants (NPPs) in preparation for accidents exceeding design standards. In this paper, we analyze the applicability of various wireless communication technologies as accident countermeasures in different NPP environments. In particular, a commercial wireless communication module (WCM) is investigated by measuring leakage current and packet error rate (PER), which vary depending on the intensity of incident radiation on the module, by testing at a Co-60 gamma-ray irradiation facility. The experimental results show that the WCMs continued to operate after total doses of 940 and 1097 Gy, with PERs of 3.6% and 0.8%, when exposed to irradiation dose rates of 185 and 486 Gy/h, respectively. In short, the lower irradiation dose rate decreased the performance of WCMs more than the higher dose rate. In experiments comparing the two communication protocols of request/response and one-way, the WCMs survived up to 997 and 1177 Gy, with PERs of 2% and 0%, respectively. Since the request/response protocol uses both the transmitter and the receiver, while the one-way protocol uses only the transmitter, then the electronic system on the side of the receiver is more vulnerable to radiation effects. From our experiments, the tested module is expected to be used for design-based accidents (DBAs) of "Category A" type, and has confirmed the possibility of using wireless communication systems in NPPs.