• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.502-516
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• 2002

Wolsong NPP are CANDU6 type reactors and there are 4 CANDU6 type reactors in commercial operation. CANDU type reactors require on-power refuelling by two remote controlled F/Ms (Fuelling Machines). Most of channels, fuel bundles are float by channel coolant flow and move toward downstream, however in about 30% of channels the coolant flow are not sufficient enough to carry fuel bundles to downstream. For those channels a special device, FARE (Flow Assist Ram Extension) device, is used to create additional force to push fuel bundles. It has been showing that during fuelling operation of some channels the channel coolant flow rate is reduced below specified limit (80% of normal), and consequently trip alarm signal turns on. This phenomenon occurs on selected channels that are instrumented for the channel flow and required to use the FARE device for refuelling. Hence it is believed that the FARE device causes the problem. It is also suspected that other channels that do not use the FARE device for refuelling might also go into channel flow low state. The analysis revealed that the channel How low occurs as the FARE device is introduced into the core and disappears as the FARE device is removed from the core. This paper presented the FARE device behavior, detailed fuelling operation sequence with the FARE device and effect on channel flow low phenomena. The FARE device components design changes are also suggested, such as increasing the number or now holes in the tube and flow slots in the ring orifice.

• Journal of the Korean Society of Fisheries and Ocean Technology
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• v.14 no.2
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• pp.89-95
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• 1978

A SCR static breaker was studied on the Resistive and inductive load, then on the overload break circuit using operational Amplifier. In this paper, the principal circuit required for forced commutation was voltage commutation by the introduction of a parallel Capacitor. The results obtained are follows; 1. In thecondition that the tima constant of R-C circuit is larger than the turn off time of SCR, the breaker has low transient phenomena and no recovery vol age. 2. By using OP Amplifier on the load circuit, overcurrent trip point will be able to adjust to the wide range of over current. 3. In the over current qrcuit, the power loss was reduced remarkably.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.515-520
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• 1996

Since the conventional Westinghouse DNB (Departure from Nucleate Boiling) protection logic is implemented on analog circuits, the logic must be very simple. However, if the DNB protection logic is implemented in a digital processor, a little bit of complexity can be allowed to increase the thermal (or operation) margin. The Westinghouse OTΔT DNB protection logic heavily restricts the operation region by applying the same logic for a full range of pressure in order to maintain its simplicity. In this work, the different DNB protection logic is used for several regions of pressure. The proposed method is applied to Yonggwang 1&2 nuclear power plants and it is calculated that the improved OTΔT can have 5.07% percent more thermal margin than the conventional OTΔT trip logic.

• Journal of the Korean Society of Systems Engineering
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• v.17 no.2
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• pp.49-60
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• 2021

The Core Protection Calculator System (CPCS) is vital for plant safety as it ensures the required Specified Acceptance Fuel Design Limit (SAFDL) are not exceeded. The CPCS generates trip signals when Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD) exceeds their predetermined setpoints. These setpoints are established based on the operating margin from the analysis that produces the SAFDL values. The goal of this research is to create a simplified CPCS that optimizes operating margin for I-SMRs. Because the I-SMR is compact in design, instrumentation placement is a challenge, as it is with Ex-core detectors and RCP instrumentation. The proposed CPCS addresses the issue of power flux measurement with In-Core Instrumentation (ICI), while flow measurement is handled with differential pressure transmitters between Steam Generators (SG). Simplification of CPCS is based on a Look-Up-Table (LUT) for determining the CEA groups' position. However, simplification brings approximations that result in a loss of operational margin, which necessitates compensation. Appropriate compensation is performed based on the result of analysis. FPGAs (Field Programmable Gate Arrays) are presented as a way to compensate for the inadequacies of current systems by providing faster execution speeds and a lower Common Cause Failure rate (CCF).

• Journal of the Korea Academia-Industrial cooperation Society
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• v.12 no.5
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• pp.2279-2284
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• 2011

In most of radial distribution systems, the overcurrent protection coordination is adopted for the protection of apparatus and the improvement of electrical power system reliability. The protection coordination structure in distribution substation is composed of several circuit breakers(CB) with distribution lines originating from one substation bus under one transformer, which trip for their fault current. But sufficient analysis is necessary for the capacity of CB's in distribution systems with several distribution generations(DG). In this paper, a protection coordination method not to exceed the traditional capacity of CB's was proposed and certified through simulation by the PSCAD-EMTDC S/W.

• Journal of the Korean Society of Safety
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• v.29 no.6
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• pp.166-171
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• 2014

Nuclear power plants are equipped with the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS) to improve safety on the normal operation. In the event of the design basis accident (DBA), a various of post accident monitor(PAM)systems support to provide important details (e.g. Containment pressure, temperature and pressure of reactor cooling system and core exit temperature) to determine action of main control room (MCR). Operator should be immediately activated for the accident mitigation with the information. Especially, core exit temperature is a critical parameter because the operating mode converts from normal mode to emergency mode when the temperature of core exit reaches $649^{\circ}C$. In this study, uncertainty which was caused by exterior environment, characteristic of thermocouple/connector and accuracy of calibrator/indicator was evaluated in accordance with ANSI-ISA 67.04. The square root of the sum of square (SRSS) methodology for combining uncertainty terms that are random and independent was used in the synthesis. Every uncertainty that may exist in the hardware which is used to measure the core exit temperature was conservatively applied and the associative relation between the elements of uncertainty was considered simultaneously. As a result of uncertainty evaluation, the channel statistical allowance (CSA) of single channel of core exit temperature was +1.042%Span. The range of uncertainty, -0.35%Span ($-4.05^{\circ}C$) ~ +2.08%Span($24.25^{\circ}C$), was obtained as the operating criteria of core exit temperature.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.463-478
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• 2019

The most important non-functional requirements for dependability of any Embedded Real-Time Safety Systems are safety, availability and reliability requirements. System architecture plays the primary role in achieving these requirements. Compliance with these non-functional requirements should be ensured early in the development cycle with appropriate considerations during architectural design. In this paper, we present an application of system architecture modeling for quantitative assessment of system dependability. We use probabilistic model checker (PRISM), for dependability analysis of the DTMC model derived from system architecture model. In general, the model checking techniques do not scale well for analyzing large systems, because of prohibitively large state space. It limits the use of model checking techniques in analyzing the systems of practical interest. We propose abstraction based compositional analysis methodology to circumvent this limitation. The effectiveness of the proposed methodology has been demonstrated using the case study involving the dependability analysis of safety system of a large Pressurized Water Reactor (PWR).

• 한국전기화학회:학술대회논문집
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• 2000.12a
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• pp.51-62
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• 2000

Large plastic Li-ion (PLI) cells (25 to 28-Ah) were fabricated for an EV application. The 28-Ah cells showed high specific energy (160 Wh/kg), high specific power (526 W/g), excellent round-trip energy efficiency $(92\%)$, and low self-discharge rate ($6\%$ in 30 days). A 25-Ah cell of an earlier design showed good cycle life of up to 750 cycles at $100\%$ DOD to $80\%$ of its initial capacity, while cycle life test of a 28-Ah cell of a later design is in progress. Preliminary safety tests were also carried out using 6-Ah cells of a similar electrode design giving very encouraging results for development of a safe hish-energy density PLI battery for EV application.

• 유공압시스템학회:학술대회논문집
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• 2010.06a
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• pp.116-125
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• 2010

In this study, a novel hydraulic energy-regenerative system was presented from its proposal through its modeling to its control. The system was based on a closed-loop hydrostatic transmission and used a hydraulic accumulator as the energy storage system in a novel configuration to recover the kinetic energy without any reversion of the fluid flow. The displacement variation in the secondary unit was reduced, which widened the uses of several types of hydraulic pump/motors for the secondary unit. The proposed system was modeled based on its physical attributes. Simulation and experiments were performed to evaluate the validity of the employed mathematical model and the energy recovery potential of the system. The experimental results indicated that the round trip recovery efficiency varied from 22% to 59% for the test bench.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.116-124
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• 2019

$M^2LFR-1000$ is a medium-power modular lead-cooled fast reactor, developed by University of Science and Technology of China (USTC), aiming at achieving a reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing the optimum safety and economics. In order to evaluate the safety performance of $M^2LFR-1000$ reactor core, three typical transients are selected from initiating events, which are unprotected transient overpower (UTOP), unprotected loss of offsite power (ULOHS+ULOF) and increase of feedwater flowrate with primary pumps trip (IFW+PLOF). These three transients presented and discussed in this paper are performed with the code Analysis of THermal-hydraulics of LEaks and Transients (ATHLET), which is developed by Gesellschaft $f{\ddot{u}}r$ Anlagen-und Reaktorsicherheit gGmbH (GRS). The results indicate that the $M^2LFR$ is safe enough under these three transients due to the good inherent safety features of the reactor, without human intervention, the reactor will reach a new steady state under UTOP condition.