• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.619-620
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• 2004

• Proceedings of the KIEE Conference
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• 2005.05a
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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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• v.41 no.4
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• pp.447-454
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• 2009

This paper is a technical review of the research and development results of the Korea Nuclear Instrumentation and Control System (KNICS) project and Nu-Tech 2012 program. In these projects man-machine interface system architecture, two digital platforms, and several control and protection systems were developed. One platform is a Programmable Logic Controller (PLC) for a digital safety system and another platform is a Distributed Control System (DCS) for a non-safety control system. With the safety-grade platform PLC, a reactor protection system, an engineered safety feature-component control system, and reactor core protection system were developed. A power control system was developed based on the DCS. A logic alarm cause tracking system was developed as a man-machine interface for APR1400. Also, Integrated Performance Validation Facility (IPVF) was developed for the evaluation of the function and performance of developed I&C systems. The safety-grade platform PLC and the digital safety system obtained approval for the topical report from the Korean regulatory body in February of 2009. A utility and vendor company will determine the suitability of the KNICS and Nu- Tech 2012 products to apply them to the planned nuclear power plants.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.591-592
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.86-86
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• 2003

• Journal of KIISE:Software and Applications
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• v.31 no.6
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• pp.750-759
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• 2004

The software of the nuclear power plant digital control system is a safety-critical system where many techniques must be applied to it in order to preserve safety in the whole system. Formal specifications especially allow the system to be clearly and completely specified in the early requirements specification phase therefore making it a trusted method for increasing safety. In this paper, we discuss the NuSCR, which is a qualified formal specification method for specifying nuclear power plant digital control system software requirements. To investigate the application of NuSCR, we introduce the experience of using NuSCR in formally specifying the plant protection system's software requirements, which is presently being developed at KNICS. Case study that shows that the formal specification approach NuSCR is very much qualified and specialized for the nuclear domain is also shown.

• International Journal of Safety
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• v.3 no.1
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• pp.38-46
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• 2004

This paper introduces the software life-cycle V&V (verification and validation) tasks for the KNICS (Korea nuclear instrumentation and control system) project. The objectives of the V&V tasks are mainly to develop a programmable logic controller (PLC) for safety critical instrumentation and control (I&C) systems, and then to apply the PLC to developing the prototype of an engineered safety features-component control system (ESF-CCS) in nuclear power plants. As preparative works for the software V&V, various kinds of software plans and V&V task procedures have been developed according to the software life-cycle management. A number of software V&V tools have been adopted or developed to efficiently support the V&V tasks. The V&V techniques employed in this work include a checklist-based review and inspection, a requirement traceability analysis, formal verification, and life-cycle based software testing.

• Proceedings of the KIEE Conference
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• 2008.07a
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• pp.1787-1788
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• 2008

본 논문은 한국원전계측제어시스템(KNICS) 개발 사업의 일환으로 개발되는 안전등급 PLC(Programmable Logic Controller) 소프트웨어 개발 도구의 컴파일러에 대한 시험을 다룬다. 개발된 컴파일러에 대하여 외부 시스템과 연계하지 않고 내부에서 각 컴포넌트 별로 시험하는 컴포넌트 시험과 컴파일러 외부 시스템과 연계하여 시험하는 통합시험을 수행하였다. 시험 과정은 먼저 시험 계획 단계에서 시험 항목을 선정하고 각 항목 별로 사례와 시험 절차를 작성하며 이에 따라 시험을 수행한 결과를 분석하여 컴파일러의 수정 및 보완에 반영하는 것이다.

• Proceedings of the Korean Information Science Society Conference
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• 2006.06c
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• pp.190-192
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• 2006

원자력 발전소의 제어 소프트웨어는 안전성이 중요시 되는 시스템이다. KNICS 컨소시엄의 APR-1400 RPS 개발 프로젝트에서는 시스템의 안전성과 품질을 높이기 위하여 요구사항을 NuSCR 정형명세로 기술하였다. 명세에 대한 분석을 위하여 SMV를 이용한 자동화된 정형검증 기법이 사용되는데, 본 논문에서는 테이블 형태의 명세인 SDT까지 그 범위를 확장하는 방법을 제안한다. 제안하는 방법의 효율성을 입증하기 위하여 실제 프로젝트에서 개발중인 시스템의 일부를 예제로 사용하였다.

• Proceedings of the Korean Information Science Society Conference
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• 2005.07b
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• pp.427-429
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• 2005

원자력 발전소의 제어 시스템은 safety-critical 소프트웨어로서 안정성이 중요시되는 시스템이다. 최근 기존의 시스템이 PLC 기반의 디지털 제어장치로 대체되면서 이에 사용되는 소프트웨어의 안정성과 품질을 보장하기 위한 정형검증 기법이 요구되고 있다. 특히 PLC 프로그램의 설계에 사용되는 FBD의 모델체킹을 통한 정형검증에 대한 연구는 미비한 수준이다. VIS 검증기는 위의 요구에 부합하는 도구로서 이를 사용하면 여러 종류의 정형 검증이 가능하다. 본 논문에서는 VIS를 이용한 FBD의 검증을 위해서 FBD를 Verilog로 변환 하는 기법을 제안한다. 제안하는 방법의 효율성을 검증하기 위해서 현재 KNICS 사업단에서 개발중인 APR-1400용 원자로 보호 시스템의 운전정지회로를 예로 사용하였다.