• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.544-549
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• 1995

An analysis for Steam Line Break (SLB) events which result in a return-to-power conditions after reactor trip was performed for a postulated Yonggwang Nuclear Power Plant Unit 3 cycle 8. Analysis methodology for post-trip return-to-power SLB is quite different from that of a no return-to-power SLB and is more complicated. Therefore, it is necessary to develop an methodology to analyze the response of the NSSS parameter and the fuel performance for the post-trip return-to-power SLB events. In this analysis, the cases with and without offsite power were simulated by crediting 3-D reactivity feedback effect due to local heatup around stuck CEA and compared with the cases without 3-D reactivity feedback with respect to fuel performance, departure from nucleate boiling ratio (DNBR) and linear heat generation rate (LHGR).

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1845-1853
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• 2024

The setpoint of the reactor trip system shall be set to consider the measurement uncertainty of the instrument channel and provide a reasonable and sufficient margin between the analytical limit and the trip setpoint. A comparative analysis was conducted to find out an appropriate uncertainty combination method through an example problem. The four methods were evaluated; 1) ISA-67.04.01 method, 2) the GUM95 method, 3) the modified GUM method developed by Fotowicz, and 4) the modified IEC61888 method proposed by authors for the pressure instrument channel presented in ISA-RP67.04.02 example. The appropriateness of each method was validated by comparing it with the result of Monte Carlo simulation. As a result of the evaluation, all methods are appropriate when all measurement uncertainty elements are normally distributed as expected. But ISA-67.04 method and GUM95 method overestimated the channel uncertainty if there is a dominant input element with rectangular distribution among the uncertainty input elements. Modified GUM95 methods developed by Fotowicz and modified IEC61888 method by authors are able to produce almost the same level of channel uncertainty as the Monte Carlo method, even when there is a dominant rectangular distribution among the uncertainty components, without computer-assisted simulations.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.58-67
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• 1999

The purpose is to produce a model of nuclear reactor trip logic caused by the steam generator water level of Wolsong 2/3/4 unit through an activity chart and a statechart and to produce C language automatically using Statechart-based Formalism and Stalemate MAGNUM toolset suggested by David Harel Formalism. It was worth attempting auto-generation of C language though we manually made Software Requirement Specification(SRS) for safety-critical software using statechart-based formalism. Most of the phases of the software life-cycle except the software requirement specification of an analysis phase were generated automatically by Computer Aided Software Engineering (CASE) tools. It was verified that automatically produced C language has high productivity, portability, and quality through the simulation.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.437-442
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• 1997

To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are peformed. The methodology is developed by evaluating "decrease in secondary heat removal" events such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling, The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. pressure.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.84.1-84
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• 2002

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.316-327
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• 2000

The Emergency Core Cooling System in current nuclear power plants typically has a considerable number of complex functions and largely cumbersome operator interfaces. Functions for initiation, switch-over between various phases of operation, interlocks, monitoring, and alarming are usually performed by relays and analog comparator logic which are difficult to maintain and test. To improve problems of an analog based ECC (Emergency Core Cooling) System, the trip computer for ECCS based on Dynamic Safety System (DSS) is implemented. The DSS is a computer based reactor protection system that has fail-safe nature and performs a dynamic self-testing. The most important feature of the DSS is the introduction of test signal that send the system into a tripped state. The test signals are interleaved with the plant signals to produce an output which switches between a tripped and health state. The dynamic operation is a key feature of the failsafe design of the system. In this work, a possible implementation of the DSS using PLC is presented for a CANDU Reactor. ECC System of the CANDU Reactor is selected as the reference system.

• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.307-318
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• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3595-3603
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• 2022

Nuclear safety-class DCS is used for nuclear reactor protection function, which is one of the key facilities to ensure nuclear power plant safety, the maintenance for DCS to keep system in a high reliability is significant. In this paper, Nuclear safety-class DCS system developed by the Nuclear Power Institute of China is investigated, the model of reliability estimation considering nuclear power plant emergency trip control process is carried out using Markov transfer process. According to the System-Subgroup-Module hierarchical iteration calculation, the evolution curve of failure probability is established, and the preventive maintenance optimization strategy is constructed combining reliability numerical calculation and periodic overhaul interval of nuclear power plant, which could provide a quantitative basis for the maintenance decision of DCS system.

• Proceedings of the KIEE Conference
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• 1998.07b
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• pp.807-809
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• 1998

Plant Protection Systems(PPS) are those systems that initiate safety actions to mitigate the consequences of design basis events by sending signals to Reactor Trip Switch Gear System(RTSS) and Engineered Safety Features-Component Control Systems(ESF-CCS). This paper illustrates distinctive features and improved design concepts of Korea Next Generation Reactor(KNGR) based on the experience obtained through prototyping of PPS.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.697-708
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• 2012

The majority of instrumentation and control (I&C) systems in today's nuclear power plants (NPPs) are based on analog technology. Thus, most existing I&C systems now face obsolescence problems. Existing NPPs have difficulty in repairing and replacing devices and boards during maintenance because manufacturers no longer produce the analog devices and boards used in the implemented I&C systems. Therefore, existing NPPs are replacing the obsolete analog I&C systems with advanced digital systems. New NPPs are also adopting digital I&C systems because the economic efficiencies and usability of the systems are higher than the analog I&C systems. Digital I&C systems are based on two technologies: a microprocessor based system in which software programs manage the required functions and a programmable logic device (PLD) based system in which programmable logic devices, such as field programmable gate arrays, manage the required functions. PLD based systems provide higher levels of performance compared with microprocessor based systems because PLD systems can process the data in parallel while microprocessor based systems process the data sequentially. In this research, a bistable trip logic in a reactor protection system (RPS) was developed using very high speed integrated circuits hardware description language (VHDL), which is a hardware description language used in electronic design to describe the behavior of the digital system. Functional verifications were also performed in order to verify that the bistable trip logic was designed correctly and satisfied the required specifications. For the functional verification, a random testing technique was adopted to generate test inputs for the bistable trip logic.