• Proceedings of the Korea Information Processing Society Conference
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• 2008.05a
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• pp.257-260
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• 2008

디지털 컴퓨터와 정보처리기술의 급속한 발전과 함께 산업계 전반적으로 아날로그 기술은 쇠퇴하고 디지털 기술로 전환되고 있다. 심지어 안전-필수 기능을 담당하는 원자력발전소의 계측제어시스템에서도 제한적으로 디지털 기술을 채택하여 사용하기 시작했다. 지금까지 소프트웨어의 신뢰도의 정량화에 대한 연구는 많이 이루어져 왔으나 소프트웨어가 가지는 특수성 때문에 연구결과에 대해 전문가들의 동의를 얻지 못하고 있는 상태이다. 원자력발전소에서는 확률적 안전성 평가(PSA)를 수행할 때 소프트웨어의 실패에 기인한 위험은 무시하고 있다. 하지만, 소프트웨어를 기반으로 한 디지털 시스템의 사용이 점점 늘어남에 따라 소프트웨어 신뢰도에 대한 정량화가 점점 더 요구되고 있다. 본 연구에서는 소프트웨어의 실패모드를 정의하고 해당 실패모드에 의해 사고가 발생할 확률을 베이지안 통계이론을 이용하여 정량화하였다.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.39 no.11
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• pp.905-911
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• 2015

In this study, to examine the effect of a hole size change(smaller hole diameter) in the outer region of the lower-support-structure-bottom plate(LSSBP) on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD software, ANSYS CFX R.15. The predicted results were compared with those of the original LSSBP. Through these comparisons, it was concluded that a more uniform distribution of the mass flow rate at the core-inlet plane could be obtained by reducing the hole size in the outer region of the LSSBP. Therefore, from the nuclear regulatory perspective, design change of the hole pattern in the outer region of the LSSBP may be desirable in terms of improving both the mechanical integrity of the fuel assembly and the core thermal margin.

• Journal of KIISE:Software and Applications
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• v.27 no.2
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• pp.120-133
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• 2000

Major obstruction of using formal methods for hybrid real-time systems in industry is the difficulty that engineers have in understanding and applying the quantitative methods in an abstract requirements phase. While formal methods technology in safety-critical systems can help increase confidence of software, difficulty and complexity in using them can cause another hazard. In order to overcome this obstruction, we propose a framework for qualitative requirements engineering of the hybrid real-time systems. It consists of a qualitative method for requirements specification, called QFM (Qualitative Formal Method), and a safety analysis method for the requirements based on a causality information, called CRSA (Causal Requirements Safety Analysis). QFM emphasizes the idea of a causal and qualitative reasoning in formal methods to reduce the cognitive burden of designers when specifying and validating the software requirements of hybrid safety systems. CRSA can evaluate the logical contribution of the software elements to the physical hazard of systems by utilizing the causality information that is kept during specification by QFM. Using the Shutdown System 2 of Wolsong nuclear power plants as a realistic example, we demonstrate the effectiveness of our approach.

• 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

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.285-290
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• 1998

본 논문은 David Harel이 제안한 Statechart based Formalism과 Statemate MAGNUM toolset을 이용하여 월성 원전 2/3/4호기 증기발생기 수위로 인한 원자로 정지를 activity chart 및 Statechart로 모델링하고 K&R C 코드를 자동으로 생산하였다. 이는 종전의 몇몇 소프트웨어 전문가에 의해서 개발될 수 밖에 없었던 원전 필수만전(Safety-critical) 소프트웨어를 정형화된 Computer Aided Software Engineering 도구를 활용하여 소프트웨어 생명주기중 요구사양명세 및 설계까지만 수행하고 그 이하는 모두 자동으로 생산하는 소프트웨어 공학의 핵심기술을 연구한 것이다. 자동으로 생산된 K&R C 코드는 품질이 우수하고 생산성이 높으며 이식성이 뛰어남을 확인할 수 있었다.

• Communications of the Korean Institute of Information Scientists and Engineers
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• v.25 no.2 s.213
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• pp.35-42
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• 2007

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4367-4381
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• 2023

Nuclear power plants are increasingly being equipped with digital I&C systems. Although some probabilistic safety assessment (PSA) models for the digital I&C of nuclear power plants have been constructed, there is currently no specific internationally agreed guidance for their modeling. This paper presents an initiative by the OECD Nuclear Energy Agency called "Digital I&C PSA - Comparative application of DIGital I&C Modelling Approaches for PSA (DIGMAP)", which aimed to advance the field towards practical and defendable modeling principles. The task, carried out in 2017-2021, used a simplified description of a plant focusing on the digital I&C systems important to safety, for which the participating organizations independently developed their own PSA models. Through comparison of the PSA models, sensitivity analyses as well as observations throughout the whole activity, both qualitative and quantitative lessons were learned. These include insights on failure behavior of digital I&C systems, experience from models with different levels of abstraction, benefits from benchmarking as well as major contributors to the core damage frequency and those with minor effect. The study also highlighted the challenges with modeling of large common cause component groups and the difficulties associated with estimation of key software and common cause failure parameters.

• Journal of Convergence Society for SMB
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• v.2 no.1
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• pp.11-18
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• 2012

As scale of software has been expanded and complicated, it is difficult to detect hazards which induce functional failure of software. Functional failure of safety-critical system (nuclear power plant, air traffic control systems, railway operating system) could result in a disaster (personal injury, environmental pollution). Therefore, it is necessary to conduct a safety analysis for preventing functional failure and increasing safety of the software. However, there are some reasons (time and effort problem, low knowledge of various safety analysis techniques, selecting conventional technique in company, organization) which disturb selecting an apposite one. This paper presents some traditional safety analysis techniques, recently presented techniques and combined models. We expect that it helps stakeholders to choice adequate one for target system.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.27 no.4
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• pp.220-227
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• 2015

Complex thermal hydraulic characteristics exist inside the reactor because the reactor internals consist of fuel assembly, internal structures and so on. In this study, to examine the effect of Reynolds-Averaged Navier-Stokes (RANS)-based two-equation turbulence models in the analysis of flow distribution inside a 1/5 scaled-down APR+, simulation was performed using the commercial computational fluid dynamics software, ANSYS CFX R.13 and the predicted results were compared with the measured data. It was concluded that reactor internal flow pattern was locally different depending on the turbulence models. In addition, the prediction accuracy of k-${\varepsilon}$ model was superior to that of other two-equation turbulence models and this model predicted the relatively uniform distribution of core inlet flow rate.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.119-130
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• 2023

Among the various elements of probabilistic safety assessment (PSA), human failure events (HFEs) and their dependencies are major contributors to the quantification of risk of a nuclear power plant. Currently, the dependency among HFEs is reflected using a post-processing method in PSA, wherein several drawbacks, such as limited propagation of minimal cutsets through the fault tree and improper truncation of minimal cutsets exist. In this paper, we propose a method to model the HFE dependency directly in a fault tree using the if-then-else logic. The proposed method proved to be equivalent to the conventional post-processing method while addressing the drawbacks of the latter. We also developed a software tool to facilitate the implementation of the proposed method considering the need for modeling the dependency between multiple HFEs. We applied the proposed method to a specific case to demonstrate the drawbacks of the conventional post-processing method and the advantages of the proposed method. When applied appropriately under specific conditions, the direct fault-tree modeling of HFE dependency enhances the accuracy of the risk quantification and facilitates the analysis of minimal cutsets.