• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.12 no.5
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• pp.380-388
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• 2002

This paper describes the optimum design for low-pressure steam turbine rotor of 1,000 MW nuclear power plant by using a combined genetic algorithm, which uses both a genetic algorithm and a local concentrate search algorithm (e.g. simplex method). This algorithm is not only faster than the standard genetic algorithm but also supplies a more accurate solution. In addition, this algorithm can find the global and local optimum solutions. The objective is to minimize the resonance response (Q factor) and total weight of the shaft, and to separate the critical speeds as far from the operating speed as possible. These factors play very important roles in designing a rotor-bearing system under the dynamic behavior constraint. In the present work, the shaft diameter, the bearing length, and clearance are used as the design variables. The results show that the proposed algorithm can improve the Q factor and reduce the weight of the shaft and the 1st critical speed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.38-43
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• 2011

For the efficient and safe operation of nuclear power plant, evaluating quantitatively aging phenomenon of major components is necessary. Especially, typical aging parameters such as stresses and cumulative usage factors should be determined accurately to manage the lifetime of the plant facility. The 3-D finite element(FE) model is generated to calculate the aging parameters. Mechanical and thermal transfer functions called Green's functions are developed for the FE model with standard step input. The stress results estimated from transfer functions are verified by comparing with 3-D FE analyses results. Lastly, we suggest an effective fatigue evaluation methodology by using the transfer functions. The usefulness of the proposed fatigue evaluation methodology can be maximized by combining it with an on-line monitoring system.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.278-283
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• 2005

Cement and paraffin waste form were prepared with a acid extraction method for the analysis of radionuclides generated from nuclear power plants. The acid extraction method was carried out with $HNO_3-HCl$ acid. At first, we compared the method with the microwave acid digestion method using SRM. The solutions of decomposed SRM were then analyzed by AAS and ICP-AES. The acid extraction method had shown good results as microwave acid digestion method. This method provided recovery values greater than $80\%$ for metallic elements.

• Journal of the Korean Society of Safety
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• v.36 no.2
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.124-129
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• 2003

In case that the mixing weight ratio of waste form between boric acid and paraffin was 3.3/l, which had been adopted in the concentrate waste drying system (CWDS) of domestic nuclear power plants. Using several specimens with different diameters and heights, 50/100mm specimens. compressive strength were measured. The experiment result showed that the small diameter specimens of compressive strength are increased more than large diameter specimens. (d=50＞75＞100mm) The average compressive strength of specimens showed that the range from 22.43 $\kg/textrm{cm}^2$ to 38.57$\kg/textrm{cm}^2$ (NRC standard$\geq$4.1 $\kg/textrm{cm}^2$). NRC standard is recommended that the compressive strength test specimens be right circular cylinders, 2 to 3 inches in diameter, with a height-to-diameter(H/D) ratio of approximately two. and compressive strength were increased more than large loading rate. As test result, this conditions are a good agreement, and estimated.

• Weed & Turfgrass Science
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• v.4 no.1
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• pp.18-25
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• 2015

A universal DNA barcoding for agricultural noxious weeds is a powerful technique for species identification without morphological knowledge, by using short sections of DNA from a specific region of the genome. Two standard barcode markers, chloroplast rbcL and matK, and a supplementary nuclear ribosomal Internal Transcribed Spacer (ITS) region were used to examine the effectiveness of the markers for Pooideae barcoding using 163 individuals of 29 taxa across 16 genera of Korean Pooideae. The rbcL and ITS revealed a good level of amplification and sequencing success while matK did not. Barcode gaps were 78.6% for rbcL, 96.2% for matK, and 91.7% for ITS, respectively. Resolving powers were 89.3% for rbcL, 92.3% for matK, and 79.1% for ITS. The matK obtained the best both barcode gap and resolving power. However, it should be considered not to employ matK for Pooideae barcode because of low rate of PCR amplification and sequencing success. As a single DNA marker, rbcL and ITS were reasonable for Pooideae barcode. Barcode gap and resolving power were increased when ITS was incorporated into the rbcL. The barcode sequences were deposited to the National Center for Biotechnology Information (NCBI) database for public use.

• The KSFM Journal of Fluid Machinery
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• v.17 no.1
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• pp.5-11
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• 2014

The purpose of this paper is to investigate an analytical approach for opening performance evaluation of the nuclear pressure safety valve based on standard codes such as ASME or KEPIC. It is well-known that safety valve is considered as one of pressure relief valves for protecting a boiler or pressure vessel from exceeding the maximum allowable working pressure. When pressure in a container reaches its set pressure, the safety valve commences discharging the internal fluid by a sudden opening called as popping. Safety valve is usually evaluated by set pressure, full open, blow-down, leakage and flow capacity. The test procedure and technical requirement for performance evaluation is described in international code of ASME code such as BPVC. The opening characteristics of steam safety valve can be analyzed by computational fluid dynamics (CFD) and steam shaft dynamics. First, the flow analysis along opening process is simulated by running the CFD models of the ten types of opening steps from 0 to 100%. As a analysis result, the various CFD outputs of flow pattern, pressure, forces on the disc and mass flow at each simulation step is demonstrated. The lift force is calculated by using the forces applied on disc from static pressure and secondary flow. And, the effect of huddle chamber or control chamber is studied by dynamic analysis based on CFD simulation results such as lift force. As a result, dynamics analysis shows opening features according to the sizes of control chamber.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.959-964
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• 2022

Lead is an important material, both for fusion or fission reactors. The cross sections of natural lead should be validated because lead is a main component of lithium-lead modules suggested for fusion power plants and it directly affects the crucial variable, tritium breeding ratio. The presented study discusses a validation of the lead transport libraries by dint of the activation of carefully selected activation samples. The high emission standard ²⁵²Cf neutron source was used as a neutron source for the presented validation experiment. In the irradiation setup, the samples were placed behind 5 and 10 cm of the lead material. Samples were measured using a gamma spectrometry to infer the reaction rate and compared with MCNP6 calculations using ENDF/B-VIII.0 lead cross sections. The experiment used validated IRDFF-II dosimetric reactions to validate lead cross sections, namely ¹⁹⁷Au(n, 2n)¹⁹⁶Au, ⁵⁸Ni(n,p)⁵⁸Co, ⁹³Nb(n, 2n)^92mNb, ¹¹⁵In(n,n')^115mIn, ¹¹⁵In(n,γ)^116mIn, ¹⁹⁷Au(n,γ)¹⁹⁸Au and ⁶³Cu(n,γ)⁶⁴Cu reactions. The threshold reactions agree reasonably with calculations; however, the experimental data suggests a higher thermal neutron flux behind lead bricks. The paper also suggests ²⁵²Cf isotropic source as a valuable tool for validation of some cross-sections important for fusion applications, i.e. reactions on structural materials, e.g. Cu, Pb, etc.

• Journal of Preventive Medicine and Public Health
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• v.57 no.1
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• pp.55-64
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• 2024

Objectives: This study investigated the prevalence and characteristics of comorbid conditions in patients exposed to ionizing radiation and those who were involved in the Soviet-Afghan war. Methods: This study analyzed the frequency and spectrum of morbidity and comorbidity in patients over a long-term period (30-35 years) following exposure to ionizing radiation at the Semipalatinsk nuclear test site or the Chornobyl nuclear power plant, and among participants of the Soviet-Afghan war. A cohort study, both prospective and retrospective, was conducted on 675 patients who underwent comprehensive examinations. Results: Numerical data were analyzed using the Statistica 6 program. The results are presented as the mean±standard deviation, median, and interquartile range (25-75th percentiles). The statistical significance of between-group differences was assessed using the Student t-test and Pearson chi-square test. A p-value of less than 0.05 was considered statistically significant. We found a high prevalence of cardiovascular diseases, including hypertension (55.0%) and cardiac ischemia (32.9%); these rates exceeded the average for this age group in the general population. Conclusions: The cumulative impact of causal occupational, environmental, and ultra-high stress factors in the combat zone in participants of the Soviet-Afghan war, along with common conventional factors, contributed to the formation of a specific comorbidity structure. This necessitates a rational approach to identifying early predictors of cardiovascular events and central nervous system disorders, as well as pathognomonic clinical symptoms in this patient cohort. It also underscores the importance of selecting suitable methods and strategies for implementing treatment and prevention measures.

• Journal of Radiation Industry
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• v.17 no.4
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• pp.509-518
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• 2023

The Kori Unit 1 nuclear power plant, which is planned to be dismantled after permanent shutdown, is expected to generate a large amount of various types of radioactive waste during the dismantling process. For the disposal of Very-low-level waste, which is expected to account for the largest amount of generation, the Korea Radioactive waste Agency (KORAD) is in the process of detailed design to build a 3-phase landfill disposal facility in Gyeongju. In addition, a large container is being developed to efficiently dispose of metal and concrete waste, which are mainly generated as Very low-level waste of decommissioning. In this study, based on the design characteristics of the 3-phase landfill disposal facility and the large container under development, radiation exposure dose evaluation was performed considering the normal and accident scenarios of radiation workers during operation. The direct exposure dose evaluation of workers during normal operation was performed using the MCNP computer program, and the internal and external exposure dose evaluation due to damage to the decommissioning waste package during a drop accident was performed based on the evaluation method of ICRP. For the assumed scenario, the exposure dose of worker was calculated to determine whether the exposure dose standards in the domestic nuclear safety act were satisfied. As a result of the evaluation, it was confirmed that the result was quite low, and the result that satisfied the standard limit was confirmed, and the radiational disposal suitability for the 3-phase landfill disposal facility of the large container for dismantled radioactive waste, which is currently under development, was confirmed.