• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.543-558
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• 2023

Currently, Japan is undertaking a nationwide project to measure and map radioactive contamination around Fukushima, as part of the efforts to restore normalcy following the nuclear accident. The Japan Atomic Energy Agency (JAEA) manages the Fukushima Environmental Safety Center, located approximately 20 km north of the Fukushima Daiichi nuclear power plant in Minamisōma City, Fukushima Prefecture. In collaboration with the JAEA, this study involved conducting comparison experiments and analyses with radiation detectors in high radiation environments, a challenging task in Korean environments. Environmental radiation surveys were conducted using three types of detectors: CZT, NaI(Tl), and LaBr₃(Ce), across two contaminated areas. Dose rate values were converted using dose rate conversion factors for each detector type, and dose rate maps were subsequently created and compared. The detectors yielded similar results, demonstrating their feasibility and reliability in high radiation environments. The findings of this study are expected to be a crucial reference for enhancing the verification and supplementation of procedures and methods in future radiation measurements and mobile surveys in high-radiation environments, using these three types of radiation instruments.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.196-205
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• 2016

Background: The emergency planning zone (EPZ) of the city of Busan is divided into the precautionary actions zone (PAZ) and the urgent protective action planning zone; which have a 5-km radius and a 20-km to 21-km radius from the nuclear power plant site, respectively. In this study, we assumed that a severe accident occurred at Shin-Kori nuclear unit 3 and evaluated the dispersion speed of radiological material at each distance at various wind speeds, and estimated the effective dose equivalent and the evacuation time of PAZ residents with the goal of supporting off-site emergency action planning for the nuclear site. Materials and Methods: The total effective dose equivalent, which shows the effect of released radioactive materials on the residents, was evaluated using the RASCAL 4.2 program. In addition, a survey of 1,036 residents was performed using a standardized questionnaire, and the resident evacuation time according to road and distance was analyzed using the VISSIM 6.0 program. Results and Discussion: According to the results obtained using the VISSIM and RASCAL programs, it would take approximately 80 to 252.2 minutes for permanent residents to move out of the PAZ boundary, 40 to 197.2 minutes for students, 60 to 232.2 minutes for the infirm, such as elderly people and those in a nursing home or hospital, and 30 to 182.2 minutes for those temporarily within the area. Consequently, in the event of any delay in the evacuation, it is estimated that the residents would be exposed to up to $10mSv{\cdot}h^{-1}$ of radiation at the Exclusion Area Boundaries (EAB) boundary and $4-6mSv{\cdot}h^{-1}$ at the PAZ boundary. Conclusion: It was shown that the evacuation time for the residents is adequate in light of the time lapse from the initial moment of a severe accident to the radiation release. However, in order to minimize the evacuation time, it is necessary to maintain a system of close collaboration to avoid traffic congestion and spontaneous evacuation attempts.

• Journal of the Korean Society of Radiology
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• v.14 no.7
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• pp.857-862
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• 2020

In March 2011, a tsunami off Japan caused radioactive material that had seeped into the sea from the Fukushima nuclear accident to flow to the Pacific Ocean, causing pollution to sea life. For a comparative evaluation with the area surrounding the site of a nuclear power plant by the release of radioactive materials, an area 20 to 30 km away from the emergency protection plan area was selected as a comparative point considering weather conditions, population distribution, etc. In addition, the government intends to analyze the minimum detection radiation received by residents around the nuclear power plant and evaluate the effective dose. Analysis of tritium radiation from water samples showed that most of the samples were not detected and that 0.0014 % to 0.777 % of the annual legal standard of 1 mSv for the general public had little effect on the human body. Therefore, the measurement and analysis of water samples around the nuclear power plant site is expected to help relieve anxiety, such as exposure to the general public and neighboring residents due to radiation release.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.97-106
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• 1986

System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAPS/MODl/NSC), based upon the sequence of events for the KNU1 (Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximates the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV settings etc. are recognized to be important in the transient analyses on a bestestimate basis.

• Journal of the Ergonomics Society of Korea
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• v.32 no.1
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• pp.125-137
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• 2013

Objective: The aim of this study was to comprehend major concepts and flows that penetrate international guides or standards for developing a quantitative possibility measure of human errors that can be committed or omitted in nuclear power plants. Background: For a few past decades, lots of researchers have studied the effect of stress and/or fatigue which can result in human errors. Thus, this study was carried out on the assumption that much of them were summarized as an international guidelines or manuals, if any, for human error prevention. Method: A literal survey was conducted with materials and documentation published by international organizations related with safety and standardization, such as ISO, OSHA, NIOSH, NASA, and so on with special reference to human error prevention through management of work stress and fatigue as major Performance Shaping Factors. Results: International guides or management manuals on stress or fatigue management for human error prevention hardly were found, and most researches seemed to concentrate on one of them individually. Conclusion: There was few vestige of research that studied both concurrently. However, it was verified that not a few researches have been tried to develop quantitative measures to estimate probability or job characteristics for human error prevention and/or performance downgrading. Application: The results of this study would help to develop a causal model of human errors due to work stress and fatigue that can result in unexpected accidents in nuclear power plant.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.24 no.11
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• pp.1447-1456
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• 2000

Simulated experiments of high pressure melt ejection(HPME) are performed to measure the released fraction of corium simulant from the French type PWR cavity. The experiments are carried out on a 1/20th linear scaled model of the Ulchin 1&2 cavity. Water or woods metal and nitrogen is used as simulant of molten corium and steam, respectively. Experimental parameters are water mass, annulus area and breach size. It is shown that only breach size effects is very important while the mass and the annulus area do not affect the released fraction. It is found that the liquid film transport is much more dominant mechanism than the entrainment droplet transport, especially in linear scale down simulated HPME experiment.

• Journal of Radiation Industry
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• v.5 no.4
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• pp.337-345
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• 2011

In order to provide an effective preparedness for a nuclear or radiological emergency happening in the domestic or neighborhood countries and to solve the vague fear of the people for the ingestion of radioactive livestock products, the establishment of national guideline level for radionuclides in feed is urgently necessary. This is because it is important to secure the safety and to manage the crisis in the agricultural, fishery and food sector by performing the effective safety control during and after nuclear incident. This study was performed to investigate the report cases of international organizations and foreign countries to set up a domestic control standard for managing radioactive substances that may be contaminated in animal feeds due to the nuclear power plant incident. In addition, an attempt was made to provide a useful reference that can help prepare a domestic control standard, using a coefficient that can consider the transfer into livestock through the intake of radioactive contaminated animal feeds. The standard radioisotopes investigated were confined to radioactive cesium ($^{137+134}Cs$) and iodine ($^{131}I$). Guideline level for the radionuclides was calculated by using the transfer coefficient factor and the maximum daily intake of animal feed provided by IAEA. For example, the maximum daily intake of animal feed was set as $25kg\;d^{-1}$ for dairy cows, $10kg\;d^{-1}$ for beef cattle, $3.0kg\;d^{-1}$ for pigs and $0.15kg\;d^{-1}$ for chickens. The result values for radioactive cesium were calculated as $8,696Bq\;kg^{-1}$, $4,545Bq\;kg^{-1}$, $1,667Bq\;kg^{-1}$ and $2,469Bq\;kg^{-1}$, respectively. The results for radioactive iodine showed the ranges between $741Bq\;kg^{-1}$ and $76,628Bq\;kg^{-1}$. These data can be utilized as a scientific reference for the preparation of a crisis management manual for the emergency control due to nuclear power plant accident in Korea and neighboring country. These results will contribute to establish the safe feed management system at national level as manual for responding the radioactive exposure of agricultural products and animal feeds, which are currently not established.