• Journal of the Korean Society for Nondestructive Testing
• /
• v.35 no.4
• /
• pp.268-274
• /
• 2015

Weld overlay was first used in power plants in the US in the early 1980s as an interim method of repairing the welds of flawed piping joints. Weld overlaid piping joints in nuclear power plants must be examined periodically using ultrasonic examination technology. Portable phased array ultrasonic technology has recently become available. Currently, the application of preemptive weld overlays as a mitigation technique and/as a method to improve the examination surface condition for more complex configurations is becoming more common. These complex geometries may require several focused conventional transducers for adequate inspection of the overlay, the original weld, and the base material. Alternatively, Phased array ultrasonic probes can be used to generate several inspection angles simultaneously at various focal depths to provide better and faster coverage than that possible by conventional methods. Thus, this technology can increase the speed of examinations, save costs, and reduce radiation exposure. In this paper, we explain the general sequence of the inspection of weld overlay and the results of signal analysis for some PAUT (phased array ultrasonic testing) signals detected in on-site inspections.

• Fire Science and Engineering
• /
• v.31 no.2
• /
• pp.52-60
• /
• 2017

Numerical simulations were performed to evaluate the habitability of an operator for a cabinet fire in the main control room of a nuclear power plant presented in NUREG-1934. To this end, a Fire Dynamics Simulator (FDS), as a representative fire model, was used. As the criteria for determining the habitability of operator, toxic products, such as CO, were also considered, as well as radiative heat flux, upper layer temperature, smoke layer height, and optical density of smoke. As a result, the probabilities of exceeding the criteria for habitability were evaluated through the sensitivity analysis of the major input parameters and the uncertainty analysis of fire model for various fire scenarios, based on V&V (Verification and Validation). Sensitivity analyses of the maximum heat release rate, CO and soot yields, showed that the habitable time and the limit criterion, which determined the habitability, could be changed. The present methodology will be a realistic alternative to enhancing the reliability for a habitability evaluation in the main control room using uncertain information of cabinet fires.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.2 no.3
• /
• pp.165-174
• /
• 2004

In order to simultaneously vitrify the ion exchange resin(IER) and combustible dry active waste(DAW) generated from Korean nuclear power plants, a vitrification pilot test was conducted using an induction cold crucible melter(CCM) . The energy necessary for startup of the glass using a Ti-ring was evaluated as about 290 kWh. The power supplied from a high frequency generator to melt the glass properly was ranged from 160 to 190 kW without any interruption. When the mixture of the IER and DAW was fed into the CCM, the concentration of CO was lowered up to 1/40 compared to feeding the IER solely. It may be caused by the DAW which can produce about 1.8 times higher heat compared to the IER. When the swelling phenomenon occurred in the glass melt, the concentration of $NO_2$, oxidizing gas, was higher than NO, reducing gas. Total feed amounts of the IER and DAW were 368 and 751 kg, respectively. And then, about 74 of volume reduction factor was achieved.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.35 no.10
• /
• pp.1337-1345
• /
• 2011

In pressurized water reactors (PWRs), the reactor pressure vessel (RPV) upper head contains numerous control rod drive mechanism (CRDM) nozzles. In the last 10 years, the incidences of cracking in alloy 600 CRDM nozzles and their associated welds has increased significantly. Several axial and circumferential cracks have been found in CRDM nozzles in European PWRs and U.S. nuclear power plants. These cracks are caused by primary water stress corrosion cracking (PWSCC) and have been shown to be driven by welding residual stresses and operational stresses in the weld region. Therefore, detailed finite-element (FE) simulations for the Korea Nuclear Reactor Pressure Vessel have been conducted in order to predict the magnitudes of the weld residual stresses in the tube materials. In particular, the weld residual stress results are compared in terms for nozzle location, geometry factor$r_o$/t, geometry of fillet, and adjacent nozzle.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.205-210
• /
• 2005

The proposed regulation for low and intermediate level radioactive waste disposal facility, scheduled to be revised, recommends that the waste generator should verify the radioactive waste conforms to the disposal requirements before disposing of it. According to the regulation, the radionuclide concentration of the radioactive waste, and its physical and chemical characteristics and safety must be confirmed prior to the disposal of low and intermediate level radioactive wastes, and the waste generator is required to deliver this information to the disposal facility operator. In addition, the disposal facility operator must assess the safety of the disposal site to establish the SWAC (Site Specific Waste Acceptance Criteria) in consideration of the characteristics of the site, whereas the waste generator must comply with the criteria in managing, disposing of and delivering low and intermediate level radioactive wastes. To abide by the afore-mentioned regulation and criteria, the waste generator must verify that the radioactive wastes to be disposed of are suitable for disposal before they are transported to the disposal facility, and to this end a radioactive waste certification program must be developed. This study conducted an in-depth analysis of the radioactive waste certification programs enforced in countries advanced in atomic energy to develop a draft of a certification program applicable to local power plants, and the program is currently applied as pilot to Uljin Power Plants No. 1 & 2 to prove its applicability. This study is going to analyze the results of the pilot application with a view to developing a radioactive waste certification program suitable to local conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.4
• /
• pp.311-320
• /
• 2006

Removal characteristics of $H^{14}CO_3$ ion from IRN-150 mixed resin contaminated with $^{14}C$ radionuclide and a gasification behavior of $^{14}C$ radionuclide to $^{14}CO_2$ were investigated. The stripping solutions used for the removal of $^{14}C$ from spent resin were $NaNO_3,\;Na_3PO_4,\;NH_4H_2PO_4,\;H_3PO_4$. The influence of stripping solution concentration on the desorption characteristics of inactive $HCO_3$ ion into stripping solution from IRN-150 mixed resin and the gasification of this ion to $CO_2$ was analyzed. The gasification behavior to $CO_2$ by using NaOH, $HNO_3$, HCl was also compared to that of phosphate solution. Real spent resin stored in Wolsung nuclear power plant was used to evaluate the gasification characteristics of $^{14}C$ radionuclide to $^{14}CO_2$. Gamma radionuclides such as $^{137}Cs,\;^{60}Co$ in residual striping solutions after desorption experiment were analyzed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.2
• /
• pp.179-188
• /
• 2016

Spent fuels from nuclear power plants, as well as high-level radioactive waste from the recycling of spent fuels, should be safely isolated from human environment for an extremely long time. Recently, meaningful studies on the development of deep borehole radioactive waste disposal system in 3-5 km depth have been carried out in USA and some countries in Europe, due to great advance in deep borehole drilling technology. In this paper, domestic deep geoenvironmental characteristics are preliminarily investigated to analyze the applicability of deep borehole disposal technology in Korea. To do this, state-of-the art technologies in USA and some countries in Europe are reviewed, and geological and geothermal data from the deep boreholes for geothermal usage are analyzed. Based on the results on the crystalline rock depth, the geothermal gradient and the spent fuel types generated in Korea, a preliminary deep borehole concept including disposal canister and sealing system, is suggested.

• Journal of KIISE:Computing Practices and Letters
• /
• v.5 no.4
• /
• pp.457-466
• /
• 1999

원자력 발전소 계측 제어 시스템, 의료 관련 시스템, 항공 관련 시스템 등 실생활과 밀접한 시스템에 소프트웨어의 사용이 점차 증가하고 있다. 이러한 시스템에서 소프트웨어의 오류는 예기치 않는 사고를 유발하여 인명, 재산상의 심각한 타격을 줄 수 있다. 그러므로 고신뢰도 소프트웨어의 개발 시에는 반드시 시스템의 안전성을 보장해 주어야 한다. 역방향 안전성 분석 방법은 시스템의 안전성을 분석하는 한가지 방법으로서 시스템의 위험 상태를 정의하고 그 위험의 원인들을 추적, 분석함으로써 안전성에 대한 효율적인 분석을 수행할 수 있는 장점을 갖는다. 이 논문에서는 소프트웨어 개발 초기 단계에서 안전성을 분석할 수 있는 방법으로 Colored Petri Nets(CPN)에 기반을 둔 역방향 안전성 분석 방법을 제시한다. 또한 CPN 역방향 안전성 분석 도구인 SAC(Safety Analyzer for CPN)의 설계 및 구현에 대해 언급한다. SAC은 기존의 상용 CPN 모델링 도구인 Design/CPN과 연계하여 사용될 수 있으므로 CPN으로 모델링된 시스템의 안전성을 분석할 수 있다는 장점이 있다. 이 논문에서는 예제로 자동 교통 제어 시스템의 일부를 CPN으로 모델링하고 SAC을 이용한 분석 과정을 기술한다.Abstract In safety-critical systems such as nuclear power plants, medical machines, and avionic systems which are closely related with our livings, the usage of software in the controlling part is growing rapidly. Since software errors in safety-critical systems may cause serious accidents leading to financial or human damages, system safety should be ensured during and after development of a system. A backward safety analysis technique defines system hazards and tries to trace their causes by analyzing system states backward. In this paper, we provide a backward safety analysis technique based on Colored Petri Nets(CPN), which is applicable to the early software development phase. Also Safety Analyzer for CPN(SAC), the supporting tool, is designed and implemented. Since SAC is compatible with Design/CPN, a commercial tool for supporting CPN, it can be applicable to analyze safety in practical problems. As an example, we model a part of the traffic light control system using CPN and analyze safety properties of the model using the SAC tool.

• Journal of the Korea institute for structural maintenance and inspection
• /
• v.22 no.5
• /
• pp.82-90
• /
• 2018

The nuclear structures are composed of large diameter bars over No.36. If the hooked bars are used for anchorage of large diameter bars, too long length of the tail extension of the hook plus bend create congestion and make an element difficult to construct. To address those problems, headed bars were developed. Provisions of ACI 318-08 specify the development length of headed bars and ignore the effect of transverse reinforcement based on the background researches. However, if headed bars are used at the cut-off or lap splice, longitudinal reinforcements, which are deformed in flexural members, induce tensile stress in cover concrete and increase the tensile force in the transverse reinforcement. The object of this research is to evaluate the effects of transverse reinforcement on the anchorage capacity of headed bar so anchorage test with variable of transverse rebar spacing was conducted. Specimens, which can consider the behavior at the cut-off, were tested. Test results show that failure of specimen without transverse reinforcement was sudden and brittle with concrete cover lifted and developed stress of headed bars was less than half of yield strength of headed bars. On the other hand, in the specimen with transverse reinforcement, transverse rebar directly resist the load of free-end so capacity of specimens highly increased.

• Applied Chemistry for Engineering
• /
• v.7 no.6
• /
• pp.1164-1173
• /
• 1996

Treating methods and characteristics of waste from a nuclear fuel powder conversion plant were studied. To recovery or treat a trace uranium in liquid waste, the ammonium uranyl carbonate(AUC) filtrate must be heated for $CO_2$ expelling, essentially. Uranium content of final treated waste solution from fuel powder processes for a heavy water reactor(HWR) could be lowered to 1 ppm by the lime treatment after the ammonium di-uranate(ADU) precipitation by simple heating. Otherwise, in case of the waste from fuel powder processes for a pressurized light water reactor(PWR), it is result in 0.8 ppm as a form of uranium peroxide such as $UO_4{\cdot}2NH_4F$ compounds. Optimum condition was found at $101^{\circ}C$ by the simple heating method in case of HWR powder process waste. And in case of PWR powder process waste, optimum condition could be obtained by precipitating with adding hydrogen peroxide and adjusting at pH 9.5 with ammonia gas at $60^{\circ}C$ after heating the waste In order to expelling $CO_2$. As the characteristics of recovered uranium compounds, median particle size of ADU was increased with pH increasing in case of HWP waste. Also, in case of uranium proxide compound recovered from PWR waste, the property of $U_3O_8$ power obtained after thermal treatment in air atmosphere was similar to that of the powder prepared from AUC conversion plant.