• KIEE International Transaction on Electrical Machinery and Energy Conversion Systems
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• v.5B no.3
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• pp.289-295
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• 2005

Instrumentation and Control (I&C) of the Neutral Beam Test Stand (NB- TS) Facility at the Korea Atomic Energy Research Institute (KAERI) for the Korea Superconducting Tokamak Advanced Research (KSTAR) project has been underway since the start of the project to answer the diverse requests arising from the various facets of the development and construction phases of the project. Optical signal transmission constitutes a significant portion of I&C works and has been performed for the entirety of the project. During the NB- TS construction and related experiments, significant achievements to a more accurate as well as more refined optical signal transmissions have been made. Examples of those I&C works that utilized the optical signal transmission are the Langmuir probe signal transmission, gradient grid current signal transmission, gas flow control and signal transmission, ion source temperature measurement, beam line component temperature monitoring, and coolant flow signal transmission, etc. These optical signal transition provisions are now performing part of the indispensable functions for the proper operation of the NB- TS facility. Attained experience and expertise are expected to be well applied to the upcoming main neutral beam injection (NBI) system construction for the KSTAR project.

• Journal of the Korean Society of Safety
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• v.30 no.5
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• pp.131-137
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• 2015

For the evaluation of safety margin of a nuclear power plant using a conservative methodology, the influence of applied assumptions such as initial conditions and boundary conditions needs to be assessed deliberately. Usually, a combination of the most conservative initial conditions is determined, and the safety margin for the transient is evaluated through the analysis for this conservative conditions. In existing conservative methodologies, a most-conservative condition is searched through the analyses for the maximum, minimum, and nominal values of the major parameters. In the present study, we investigates a new approach which can be applied to choose a most-conservative initial condition effectively when a best-estimate computer code and a conservative evaluation methodology are utilized for the evaluation of safety margin of transients. By constituting the band of various initial conditions using the random sampling of input parameters, the sensitivity study for various parameters are performed systematically. A method of sampling the value of control or operation parameters for a certain range is adopted by use of MOSAIQUE program, which enables to minimize the efforts for achieving the steady-state for various different conditions. A representative control parameter is identified, which governs the reactor coolant flow rate, pressurizer pressure, pressurizer level, and steam generator level, respectively. It is shown that an appropriate distribution of input parameter is obtained by adjusting the range and distribution of the control parameter.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.191-196
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• 1996

The potential value of the Application Specific Integrated Circuits(ASIC's) in safety systems of Nuclear Power Plants(NPP's) is being increasingly recognized because they are essentially hardwired circuitry on a chip, the reliability of the system can be proved more easily than that of software based systems which is difficult in point of software V&V(Verification and Validation). There are two types of ASIC, one is a full customized type, the other is a half customized type. PLD(Programmable Logic Device) used in this paper is a half customized ASIC which is a device consisting of blocks of logic connected with programmable interconnections that are customized in the package by end users. This paper describes the RPS(Reactor Protection System) composed of ASICs which provides emergency shutdown of the reactor to protect the core and the pressure boundary of RCS(Reactor Coolant System) in NPP's. The RPS is largely composed of five logic blocks, each of them was implemented in one PLD, as the followings. A). Bistable Logic B). Matrix Logic C).Initiation Logic D). MMI(Man Machine Interface) Logic E). Test Logic.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3653-3664
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• 2021

In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO₂ benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.209-217
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• 1986

The inlet coolant temperature reduction technique has been used to extend the load follow operation further in the end-of-cycle-life(EOL). In order to simulate the technique and calculate the nuclear characteristics of a PWR core according to the load follow operation, the three dimensional computing system has been established. The analysis was performed in both MINB and SPINR modes of typical 12-3-6-3 locad follow operation for the EOL of KNU-1 plant. Moreover, the capability of return-to-power has been also tested for those two modes with the system analysis by the RETRAN-02 code. The results show that it has no difficulty to extend the load follow operation further in the EOL by applying the inlet coolant temprature reduction, and also the spinning reserve capacity(SRC) increases by 13% in MINB mode and 14% in SPINR mode Bore that used by control rods only, for 14$^{\circ}$ F drop in the inlet temperature.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.401-422
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• 2005

We review here research and development progress achieved in US Plasma Chamber technology roughly over the last decade. In particular, we focus on two major programs carried out in the US: the APEX project (1998-2003) and the US ITER TBM activities (2003-present). The APEX project grew out of the US fusion program emphasis in the late 1990s on more fundamental science and innovation. APEX was commissioned to investigate novel technology concepts for achieving high power density and high temperature reactor coolants. In particular, the idea of liquid walls and the related research is described here, with some detailed examples of liquid metal and molten salt magnetohydrodynamic and free surface effects on flow control and heat transfer. The ongoing US ITER Test Blanket Module (TBM) program is also described, where the current first wall/blanket concepts being considered are the dual coolant lead lithium concept and the solid breeder helium cooled concepts, both using ferritic steel structures. The research described for these concepts includes both thermofluid MHD issues for the liquid metal coolant in the DCLL, and thermomechanical issues for ceramic breeder packed pebble beds in the solid breeder concept. Finally, future directions for ongoing research in these areas are described.

• Proceedings of the Korean Society of Machine Tool Engineers Conference
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• 2005.05a
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• pp.216-227
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• 2005

The head penetrations for control rod drive mechanism and instrumentation systems are installed at the reactor pressure vessel head of PWRs. Primary coolant water and the operating conditions of PWR plants can cause cracking of these nickel-based alloy through a process called primary water stress corrosion cracking (PWSCC). Inspection of the head penetrations to ensure the integrity of the head penetrations has been interested since reactor coolant leakages were found at U. S. reactors in 2000 and 2001. The complex geometry of the head penetrations and the very low echo amplitude from the fine, multiple flaws due to the nature of the see made it difficult to detect and size the flaws using conventional pulse-echo UT methods. Time-of-flight-diffraction technique, which utilizes the time difference between the flaw tips while pulse-echo does the flaw response amplitude from the flaw, has been selected for this inspection for it's best performance of the detection and sizing of the head penetration see flaws. This study defines the limits of the detectable and accurately sizable minimum flaw size which can be detected by the General TOFD and the Delta TOFD techniques for circumferentially and axially oriented flaws respectively. These results assures the reliability of the inspection techniques to detect and accurately size for various kind of flaws, and will also be utilized for the future development and qualifications of the TOFD techniques to enhance the detecting sensitivity and sizing accuracy of the flaws of the reactor head penetrations in nuclear power plants.

• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.199-212
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• 2008

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3275-3285
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• 2021

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

• Journal of the Korea Convergence Society
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• v.13 no.3
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• pp.191-196
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• 2022

Unlike an international combustion engine car, a battery-powered electric vehicle requires an additional heat source for its heating system. A high-voltage coolant heater has the advantages of high efficiency and a wide operating temperature range. In its development, the geometry design of the coolant flow path is essential. This paper presents the thermal flow simulations of a 7kW high-voltage heater with symmetric serpentine flow channels arranged parallelly. The heater performance was evaluated from the simulation results in terms of the pressure and temperature differences and the flow uniformity. The proposed design showed a greater flow resistance and similar heat exchanging capability than the existing parallel serpentine design. It has the advantage of a relatively wide low-temperature surface area, where the control circuit board susceptible to high temperatures can be located.