• Nuclear Engineering and Technology
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• 제48권4호
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• pp.1047-1057
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• 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
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• 제55권8호
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• pp.2952-2965
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• 2023

Advanced reactors, such as Small Modular Reactors or existing Nuclear Power Plants, often use Field Programmable Gate Array (FPGA) based controllers in new Instrumentation and Control (I&C) system architectures or as an alternative to existing analog-based I&C systems. Compared to CPU-based Programmable Logic Controllers (PLCs), FPGAs offer better overall performance. However, programming functions on FPGAs can be challenging due to the requirement for a hardware description language that does not explicitly support the operation of real numbers. This study aims to implement the Reactor Trip (RT) functions of the existing analog-based Reactor Protection System (RPS) using FPGAs. The RT equations for Overtemperature delta Temperature and Overpower delta Temperature involve dynamic compensators expressed with the Laplace transform variable, 's', which is not directly supported by FPGAs. To address this issue, the trip equations with the Laplace variable in the continuous-time domain are transformed to the discrete-time domain using the Z-transform. Additionally, a new operation based on a relative value for the equation range is introduced for the handling of real numbers in the RT functions. The proposed approach can be utilized for upgrading the existing analog-based RPS as well as digitalizing control systems in advanced reactor systems.

• Nuclear Engineering and Technology
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• 제40권7호
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• pp.615-624
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• 2008

In this paper, the contribution of task types and error types involved in the human-related unplanned reactor trip events that have occurred between 1986 and 2006 in Korean nuclear power plants are analysed in order to establish a strategy for reducing the human-related unplanned reactor trips. Classification systems for the task types, error modes, and cognitive functions are developed or adopted from the currently available taxonomies, and the relevant information is extracted from the event reports or judged on the basis of an event description. According to the analyses from this study, the contributions of the task types are as follows: corrective maintenance (25.7%), planned maintenance (22.8%), planned operation (19.8%), periodic preventive maintenance (14.9%), response to a transient (9.9%), and design/manufacturing/installation (6.9%). According to the analysis of the error modes, error modes such as control failure (22.2%), wrong object (18.5%), omission (14.8%), wrong action (11.1 %), and inadequate (8.3%) take up about 75% of the total unplanned trip events. The analysis of the cognitive functions involved in the events indicated that the planning function had the highest contribution (46.7%) to the human actions leading to unplanned reactor trips. This analysis concludes that in order to significantly reduce human-induced or human-related unplanned reactor trips, an aide system (in support of maintenance personnel) for evaluating possible (negative) impacts of planned actions or erroneous actions as well as an appropriate human error prediction technique, should be developed.

• Nuclear Engineering and Technology
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• 제32권4호
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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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• 제44권6호
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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.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 심포지엄 논문집 정보 및 제어부문
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• pp.84-86
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• 2005

The digital systems such as PLC or DCS have been applied to non-safety systems of nuclear power plants because of many difficulties in using analog systems. Nowadays, digital systems have been applied to safety systems of the plants such as reactor protection system. One of the main advantages of digital systems is applicability of automatic testing methods to the systems. The protection system requires high-reliability and high-availability because it shall minimize the propagation of abnormal or accident conditions of nuclear power plants. The calculation of reliability and availability of systems depends on the maintenance period of the system. In general, the maintenance period of the protection system is one-month in case of the manual test. However, the cycle of test can be shortened in several hours by using automatic periodic testing. The reliability and availability of the system is better when test period is shortened because the reliability and availability is inverse proportion to the test period. In this research, we developed the automatic periodic testing method for KNICS Reactor Protection System, which can test the system automatically without an operator or a tester. The automatic testing contained all functions of reaction protection systems from analog-to-digital conversion function of the bistable Processor to the coincident trip function of the coincident processor. By applying the automatic periodic testing to reaction system, the maintenance cost can be cut down and the reliability can be increased.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.55-62
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• 2014

A method for estimating software reliability for nuclear safety software is proposed in this paper. This method is based on the software reliability growth model (SRGM), where the behavior of software failure is assumed to follow a non-homogeneous Poisson process. Two types of modeling schemes based on a particular underlying method are proposed in order to more precisely estimate and predict the number of software defects based on very rare software failure data. The Bayesian statistical inference is employed to estimate the model parameters by incorporating software test cases as a covariate into the model. It was identified that these models are capable of reasonably estimating the remaining number of software defects which directly affects the reactor trip functions. The software reliability might be estimated from these modeling equations, and one approach of obtaining software reliability value is proposed in this paper.

• 전자공학회논문지
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• 제49권9호
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• pp.279-286
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• 2012

최근 범용소프트제어기 설계는 원자력발전소의 첨단주제어실에 적용되고 있다. 범용소프트제어기는 고집적 주제어실에서 비안전 기기뿐만이 아니라 안전기기를 제어할 수 있는 소프트웨어 기반의 수동제어 수단이다. 따라서 범용소프트제어기는 신형 주제어실의 단일 워크스테이션 구현을 위한 필수적인 설계특성을 갖고 있다. 전통적인 주제어실은 컴퓨터 기반으로 하는 통합 운전원 인터페이스 체계로 대체되고 있다. 범용소프트제어기의 오작동신호 발생 가능성을 줄이기 위해 어떠한 기기의 조작을 위해서는 2단계의 구분된 운전원 조작을 요구하는 설계를 고려하였다. 범용소프트제어기 오작동 가능성은 매우 낮기 때문에 범용소프트제어기 그자체로 발전소의 트립 가능성을 증가시키지는 않는다. 범용소프트제어기는 원자력발전소의 계측제어분야/인간연계 분야의 혁신을 대표한다. 범용소프트제어기는 인간연계를 기반으로 하는 단일 표시장치에 다양한 디비젼의 제어와 표시기를 통합하고 있다. 범용소프트제어기의 고장으로부터 안전기능 수행의 영향을 막기 위해 안전기기 및 기능에는 공학적 안전설비 신호가 적용된다. 또한 안전등급 수동스위치는 범용소프트제어기의 신호보다 우선한다. 그러므로 범용소프트제어기의 오작동 신호는 안전관련 스위치로부터의 제어신호에 의해 차단되어질 수 있다.