• Environmental Engineering Research
• /
• 제21권4호
• /
• pp.384-392
• /
• 2016

The aim of this study was to investigate the removal efficiency of $17{\alpha}$-methyltestosterone (MT) from aqueous solution by Salvinia cucullata Roxb. ex Bory in an active plant-based reactor with a specific focus on linear regression analysis for the sorption phenomena of MT onto the plant roots. A high performance liquid chromatographic method using UV detection (245 nm) was used to analyse the samples. The batch experiments of the active plant reactor (APR) were established to investigate the ability of Salvinia cucullata to remove MT from the liquid phase. The results revealed that 40% and 60% removal of MT from the liquid phase was observed at 5 min. and at 4 h, respectively. Salvinia cucullata can effectively remove MT from the aqueous solution in APRs. Kinetic studies revealed that the sorption phenomena of MT by Salvinia is best described using a linearized pseudo - second order model. Based on the kinetic parameters, it is likely that during the first 4 h of the contact (t = 0 to t = 4 h) sorption is the major driving mechanism of the disappearance of MT from aqueous solutions. However, at higher MT concentrations, diffusivity of MT has a significant effect on the migration of MT from the bulk stream to the root surface. The isotherm analysis revealed that the sorption kinetics favourably followed pseudo second-order. The results of isotherm analysis have indicated that the sorption of MT onto the root surfaces of Salvinia cucullata was favourable and almost irreversible.

• Nuclear Engineering and Technology
• /
• 제48권4호
• /
• pp.1047-1057
• /
• 2016

Design engineering process for field programmable gate array (FPGA)-based reactor trip functions are developed in this work. The process discussed in this work is based on the systems engineering approach. The overall design process is effectively implemented by combining with design and implementation processes. It transforms its overall development process from traditional V-model to Y-model. This approach gives the benefit of concurrent engineering of design work with software implementation. As a result, it reduces development time and effort. The design engineering process consisted of five activities, which are performed and discussed: needs/systems analysis; requirement analysis; functional analysis; design synthesis; and design verification and validation. Those activities are used to develop FPGA-based reactor bistable trip functions that trigger reactor trip when the process input value exceeds the setpoint. To implement design synthesis effectively, a model-based design technique is implied. The finite-state machine with data path structural modeling technique together with very high speed integrated circuit hardware description language and the Aldec Active-HDL tool are used to design, model, and verify the reactor bistable trip functions for nuclear power plants.

• Nuclear Engineering and Technology
• /
• 제56권2호
• /
• pp.575-600
• /
• 2024

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%-8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ~4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ~6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

• 한국안전학회지
• /
• 제28권3호
• /
• pp.123-128
• /
• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.

• Nuclear Engineering and Technology
• /
• 제49권7호
• /
• pp.1431-1441
• /
• 2017

The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP) design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower), which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
• /
• pp.862-868
• /
• 1998

Korean Next Generation Reactor(KNGR) is a standardized evolutionary Advanced Light Water Reactor design under development Korea Power Company(KEPCO). It incorporates design enhncements such as active and passive advanced design features(ADFs) to increase the plant safety. A Preliminary level 1 Probabilistic Safety Assessment(PSA) has been performed for KNGR to examine the effect of these safety features. The preliminary PSA result shows that it meets the KNGR safety goal on core damage frequency(CDF). The result of this safety assessment shows that the four-train safety systems, and the ADFs such as Passive Secondary Cooling System (PSCS) contributes greatly to the reduction the CDF. Furthermore, several design changes are made or proposed for detailed review based on the PSA insights.

• 환경위생공학
• /
• 제10권1호
• /
• pp.98-105
• /
• 1995

In the treatment of wastewater or sewage plant sludge with high solid concentration, high rate digestion process in which heating and mixing occur at a time is mainly used, and in the case of wastewater containing solid matter below 1000mg/ℓ the recently developed AF or UASB is developed Recently and commonly utilized. But these processes have weakpoints such as clogging of packing media and need of long period of trial run after microorganism granulation. In this point of view, there are active researches on the ASBR( anaerobic sequence batch reaction ) that is capable of treating of organic matter with reactor that has no packing materials and controlling the inflow time, reaction time sedimentation time and outflow time by time control without loss of microorganisms. The objectives of this study are to evaluate the efficiency of ASBR process according to the reaction time, change of treated water quality and gas output rate in the treatment of wheat plant wastewater.

• Nuclear Engineering and Technology
• /
• 제47권3호
• /
• pp.293-305
• /
• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

• 한국안전학회지
• /
• 제36권3호
• /
• pp.24-30
• /
• 2021

Heat is generated during the synthesis and mixing process of chemical compounds due to a change in activation energy during the reaction. A runaway reaction occurs when sufficient heat is not removed during the heat control process within a reactor, rapidly increasing the temperature, reaction speed, and rate of heat generation inside the reactor. A risk assessment was executed using an RC-1 (Reaction Calorimeter) during Friedel-Crafts acylation. Friedel-Crafts acylation runs the risk of rapid heat generation during Active Pharmaceutical Ingredient (API) manufacturing; it was used to confirm the risk of a runaway reaction at each synthesis stage and during the mixing process. This study used experimental data to develop a safety efficiency improvement plan to control the risks of runaway and other exothermic reactions, which was implemented at the production site of a chemical plant.

• Nuclear Engineering and Technology
• /
• 제54권8호
• /
• pp.2915-2923
• /
• 2022

There is an increasing interest in passive safety systems to minimize the need for operator intervention or external power sources in nuclear power plants. Because a passive system has a weak driving force, there is greater uncertainty in the performance compared with an active system. In previous studies, several methods have been suggested to evaluate passive system reliability, and many of them estimated the failure probability using thermal-hydraulic analyses and the Monte Carlo method. However, if the functional failure of a passive system is rare, it is difficult to estimate the failure probability using conventional methods owing to their high computational time. In this paper, a procedure for the application of the Chernoff bound to the evaluation of passive system reliability is proposed. A feasibility study of the procedure was conducted on a passive decay heat removal system of a micro modular reactor in its conceptual design phase, and it was demonstrated that the passive system reliability can be evaluated without performing a large number of thermal-hydraulic analyses or Monte Carlo simulations when the system has a small failure probability. Accordingly, the advantages and constraints of applying the Chernoff bound for passive system reliability evaluation are discussed in this paper.