• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.14 no.5
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• pp.1-11
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• 2000

A control system using a programmable logic controller (PLC) has been developed for a dual tower dryer (DTD) in a canadian deuterium natural uranium reactor (CADND) type nuclear power plant. This system will replace the existing DTD control system which was implemented with mechanical timers and relays. The new control system makes it possible for an operator to perform more precise time and dew point control for the DTD, thanks to the high efficiency and flexibility of the PLC. The operational cost for the control system is much reduced compared to the existing system.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.253-259
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• 2015

The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ~265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.227-239
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• 2015

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

• Nuclear Engineering and Technology
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• v.30 no.6
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• pp.531-540
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• 1998

The DYLAM-3 code which overcomes the limitation of event tree/fault tree was applied to LOOP (Loss of Off-site Power) in the mid-loop operation employing HEPs (Human Error Probabilities) supplied by the ASEP (Accident Sequence Evaluation Program) and the SEPLOT (Systematic Evaluation Procedure for Low power/shutdown Operation Task) procedure in this study. Thus the time history of core uncovery frequency during the mid-loop operation was obtained. The sensitivity calculations in the operator's actions to prevent core uncovery under LOOP in the mid-loop operation were carried out. The analysis using the time dependent HEP was performed on the primary feed & bleed which has the most significant effect on core uncovery frequency. As the result, the increment of frequency is shown after 200 minutes duration of simulation conditions. This signifies the possibility of increment in risk after 200 minutes. The primary feed & bleed showed the greatest impact on core uncovery frequency and the recovery of the SCS (Shutdown Cooling System) showed the least impact. Therefore the efforts should be taken on the primary feed & bleed to reduce the core uncovery frequency in the mid-loop operation. And the capability of DYLAM-3 in applying to the time dependent concerns could be demonstrated.

• Nuclear Engineering and Technology
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• v.17 no.1
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• pp.1-7
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• 1985

Non-destructive gamma-ray spectrometry was carried out on the spent PWR fuel assemblies at the spent fuel pool of reactor-site. Attention was focused on the determination of burnup distribution and cooling time. For the measurement of burnup distribution, the concentration ratio of $^{134}$ Cs$^{137}$ Cs was used and the results showed these ratios varied with the positions of assemblies in the core during their irradiation. For the measurement of cooling time, $^{144}$ Ce$^{137}$ Cs was used and the results were agreed considerably well with the operator declared cooling time.

• Journal of the Korean Society of Systems Engineering
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• v.15 no.1
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• pp.51-59
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• 2019

After Fukushima Dai-ichi NPP accident, the need for implementation of diverse and flexible coping strategies (FLEX) became evident. However, to ensure the effectiveness of the safety strategy, it is essential to quantify the uncertainties associated with the station blackout (SBO) scenario as well as the operator actions. In this paper, a systems engineering approach for uncertainty analysis (UA) of a SBO scenario in advanced pressurized water reactor is performed. MARS-KS is used as a best estimate thermal-hydraulic code and is loosely-coupled with Dakota software which is employed to develop the uncertainty quantification framework. Furthermore, the systems engineering approach is adopted to identify the requirements, functions and physical architecture, and to develop the verification and validation plan. For the preliminary analysis, 13 uncertainty parameters are propagated through the model to evaluate the stability and convergence of the framework. The developed framework will ultimately be used to quantify the aleatory and epistemic uncertainties associated with an extended SBO accident scenario and assess the coping capability of APR1400 and the effectiveness of the implemented FLEX strategies.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2828-2839
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• 2022

This study developed a safety assessment tool for geological disposal systems called APro, a systemically integrated modeling system based on modularizing and coupling the processes which need to be considered in a geological disposal system. Thermal, hydraulic, chemical, canister failure, radionuclide release and transport processes were considered in the current version of APro. Each of the unit processes in APro consists of a single Default Module, and several Alternative Modules which can increase the flexibility of the model. As an initial stage of developing the modularization concept and modeling interface, the Default Modules of each unit process were described, with one Alternative Module of chemical process. The computation part of APro is mainly a MATLAB workspace controlling COMSOL and PHREEQC, which are coupled by an operator splitting scheme. The APro model domain is a stylized geological disposal system employing the Swedish disposal concept (KBS-3 type), but the repository layout can be freely adjusted. In order to show the applicability of APro to the total system performance assessment of geological disposal system, some sample simulations were conducted. From the results, it was confirmed that coupling of the thermal and hydraulic processes and coupling of the canister failure and the radionuclide release processes were well reflected in APro. In addition, the technical connectivity between COMSOL and PHREEQC was also confirmed.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1431-1440
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• 2024

The study aimed to develop a laser-based distance meter (LDM) to improve water surface identification for clinical MeV electron beam dosimetry, as inaccurate water surface determination can lead to imprecise positioning of ionization chambers (ICs). The LDM consisted of a laser ranging sensor, a signal processing microcontroller, and a tablet PC for data acquisition. I₅₀ (the water depth at which ionization current drops to 50 % of its maximum) measurements of electron beams were performed using six different types of ICs and compared to other water surface identification methods. The LDM demonstrated reproducible I₅₀ measurements with a level of 0.01 cm for all six ICs. The uncertainty of water depth was evaluated at 0.008 cm with the LDM. The LDM also exposed discrepancies between I₅₀ measurements using different ICs, which was partially reduced by applying an optimum shift of IC's point of measurement (POM) or effective point of measurement (EPOM). However, residual discrepancies due to the energy dependency of the cylindrical chamber's EPOM caused remained. The LDM offers straightforward and efficient means for precision water surface identification, minimizing reliance on individual operator skills.

• The Korean Journal of Nuclear Medicine Technology
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• v.13 no.3
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• pp.147-151
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• 2009

Purpose: 2-[$^{18}F$]Fluoro-2-deoxy-D-glucose ([$^{18}F$]FDG) particularly plays as a important role in Positron Emission Tomography (PET) imaging in nuclear medicine. Domestic [$^{18}F$]FDG auto synthesizers are installed in Seoul National University Bundang Hospital (SNUBH) at June 2008, these modules were known that it's synthetic yields were guaranteed in average $45{\pm}5%$ so far. To improve yields and convenience of domestic [$^{18}F$]FDG auto synthesizer, numerous trials in reaction time, base concentration, pressure and temperature were performed to increase [$^{18}F$]FDG yields. Materials and Methods: Several synthetic factors (temperature, time and pressure) and shortcoming were corrected based on many evaporation test. Syringe dispensing of tetra-butylammonium bicarbonate (TBAB) was replaced with micro pipette to prepare tetrabutyl ammonium fluoride salt ([$^{18}F$]TBAF). Troublesome refill of liquid nitrogen every 2 hours which was used to protect vacuum system was changed to charcoal cartridge, base guard filter. To monitor the volume of delivered $[^{18}O]OH_2$ from cyclotron by surveillance camera, we set up the volumetric vial on the cover of the module. In addition to, the recovery vial was added in [$^{18}F$]FDG production system to recover [$^{18}F$]FDG loss due to the leak of valve ($V_{13,14}$) in [$^{18}F$]FDG purification process. Results: When we used micro pipette for adding TBAB ($30\;{\mu}L$ in 12% $H_2O$ in acetonitrile), this quantitative dispensation has enabled to improve $5.5{\pm}1.7%$ residual fluorine-18 activity in fluorine separation cartridge compared to syringe adding. Besides, the synthetic yields of [$^{18}F$]FDG has increased $58{\pm}2.6%$ (n=19), $58{\pm}2.9%$ (n=14), $60%{\pm}2.5%$ (n=17) for 3 months. The life cycle of charcoal cartridge and base vacuum was 3 months prior to filling liquid nitrogen every 2 hours and additional side separator can prevent pump corrosion by organic solvent. After setting of volumetric indicator vial, the operator can easily monitor the total volume of irradiated $[^{18}O]OH_2$ from cyclotron. The recovery vial can be used for the stabilizer when an irregular [$^{18}F$]FDG loss was generated by the leak of valves ($V_{13,14}$). Conclusions: We has optimized appropriate synthetic conditions (temperature, time, pressure) in domestic [$^{18}F$]FDG auto synthesizer. In addition to, the remodeling with several accessories improve yields of domestic [$^{18}F$]FDG auto synthesizer with reliable reproducibility.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.1 s.173
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• pp.196-205
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• 2000

Computed Tomography (CT), Nuclear Magnetic Resonance Imaging (MR1) and some ultrasound techniques make it possible to obtain cross sections of human body or mechanical parts. In CAD system, a series of sectional surfaces can also be obtained from solid models of 3D objects. In this paper we introduce a tetrahedral meshing algorithm from these series of general sections using basic operators. In this scheme. general sections of three-dimensional object are triangulated first and side surfaces between two sections are triangulated by the use of tiling process. Finally tetrahedral meshing process is performed on each layer of 3D objects, which is composed of two general sections and one side surface.