• 한국정밀공학회지
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• 제8권4호
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• pp.64-75
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• 1991

The design of structures of modern industry has developed to satisfy stringent service, realiability and safety. Up to now, geometric profile which means surface foughness and dimension accuracy is mainly treated in manufacturing process of work surface. But it is inevitable to evaluate changes of surface geometry as well as the nature of alterations in surface layers because surface of workpiece changes as a result of phase transformation, chemical changes, plastic deformation and stress changes. This paper is to present principal data for safety design by verifying the effect of grinding conditions and method in grinding layers and to explain the method of measuring surface integrity. In this paper, structural steel(SM20C) is used as a workpiece. Of integrity, surface roughness in view of surface texture is analyzed by frequency domain and residual stress, structures and defect of ground layers in view of surface metallurgy are investigated.

• 한국철도학회:학술대회논문집
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• 한국철도학회 2010년도 춘계학술대회 논문집
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• pp.1389-1394
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• 2010

This paper describes the evaluation of structural integrity and dynamic characteristic of inertial load test equipments for performance test of railway vehicle propulsion control system. The propulsion control system of railway vehicle has to be confirmed of safety and reliability prior to it's application. Therefore, inertial load test equipments were designed through theoretical equation for performance test of propulsion control system. The structural analysis of inertial load test equipments was conducted using Ansys v11.0 and it's dynamic characteristic was evaluated the designed using Adams. The results showed that the structural integrity of inertial load test equipment was satisfied with a safety factor of 10.2. Also, the structural stability was proved by maximum dynamic displacement of 0.82mm.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.952-964
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• 2016

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium-water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium-water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

• 한국철도학회:학술대회논문집
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• 한국철도학회 2003년도 춘계학술대회 논문집
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• pp.511-516
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• 2003

It is very important to ensure system safety during the process of developing a system. Railway system is also devoting a great portion for the safety. Nowadays many countries leading railway industry have their own system assessment principles according to the situation of their train control systems. In this paper, several principles to derive Safety Integrity Level are represented in the railway signalling system. The characteristics of those principles are also considered respectively.

• Journal of Mechanical Science and Technology
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• 제14권2호
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• pp.215-225
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• 2000

The structural integrity of the fuel test loop (FTL) in a Korean experimental reactor is evaluated when the FTL, submerged in a water environment, is subjected to a postulated pipe rupture. The analyses are performed under static and dynamic conditions, imposing the thrust force history at each postulated pipe rupture section. Through analysis the following results are found: l) A double ended guillotine can not be expected based on the toughness of the material, 2) the structural integrity of the chimney surrounding the FTL would not impede the structural integrity by the pipe whip. All analyses are performed by finite element methods.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.185-191
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• 2020

Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events. The vessel failure probabilities from the probabilistic fracture mechanics analyses are combined with the transient frequencies to generate the through-wall cracking frequencies, which are compared to the acceptance criterion. Several sensitivity analyses are performed, focusing on the orientations and sizes of cracks, the copper content, and a flaw distribution model. The results show that the integrity of the reactor vessel is expected to be maintained for long-term operation beyond the design lifetime from the PTS perspective using the design data of the advanced reactor. Moreover, a fluence level exceeding 9×10¹⁹ n/㎠ is found to be acceptable, generating a sufficient margin beyond the design lifetime.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.359-374
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• 2024

A high-fidelity numerical analysis methodology was proposed for evaluating the fuel rod cladding integrity of a Prototype Gen IV Sodium Fast Reactor (PGSFR) during normal operation and Design basis events (DBEs). The MARS-LMR code, system transient safety analysis code, was applied to analyze the DBEs. The results of the MARS-LMR code were used as boundary condition for a 3D computational fluid dynamics (CFD) analysis. The peak temperatures considering HCFs satisfied the cladding temperature limit. The temperature and pressure distributions were calculated by ANSYS CFX code, and applied to structural analysis. Structural analysis was performed using ANSYS Mechanical code. The seismic reactivity insertion SSE accident among DBEs had the highest peak cladding temperature and the maximum stress, as the value of 87 MPa. The fuel cladding had over 40 % safety margin, and the strain was below the strain limit. Deformation behavior was elucidated for providing relative coordinate data on each active fuel rod center. Bending deformation resulted in a flower shape, and bowing bundle did not interact with the duct of fuel assemblies. Fuel rod maximum expansion was generated with highest stress. Therefore, it was concluded that the fuel rod cladding of the PGSFR has sufficient structural safety margin during DBEs.

• 한국화재소방학회논문지
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• 제20권3호
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• pp.1-8
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• 2006

본 논문의 목적은 Ethyl Benzene 플랜트의 공정에서 과압 현상이 Column 상부의 반응폭주 및 화재 폭발의 원인이 되기 때문에 안전장치시스템의 신뢰도가 압력방출밸브가 요구하는 안전건전성수준으로 설계되어 있는지를 정량적으로 분석한 것이다. 압력방출밸브의 요구시 실패확률은 일반신뢰도 자료 조사결과를 근거로 하여 안전장치시스템에 대한 안전건전성수준의 목표등급을 SIL3으로 설정하였고, 이에 대한 PFD를 1.00E-3에서 1.00E-4로 결정하였다. 신뢰도 모델의 구축 및 결함수 분석기법을 이용하여 SIS의 요구시 실패확률에 대한 정량화를 수행한 결과 SIS에 대한 PFD는 Benzene Prefractionator Column, Benzene Column, EB Column에 대해 각각 8.97E-04, 5.37E-04, 5.37E-04로 계산되었다. 따라서 SIS의 신뢰도가 SIL3 등급에 요구되는 안전건전성수준으로 설계되어 있다고 판단되며 컨트롤밸브에 대한 6개월 주기의 Partial Stroke Test가 수행될 경우 각 Column의 SIS는 약 $22{\sim}27%$의 신뢰도 향상이 기대된다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• 대한조선학회논문집
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• 제49권6호
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• pp.447-460
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• 2012

In this study, the shutdown system of the fuel gas supply system is designed based on the Safety Integrity Level of IEC 61508 and IEC 61511. First of all, the individual risk($10^{-4}$/year) and the risk matrix which are the risk acceptance criteria are set up for the qualitative risk assessment such as the HAZOP study. The natural gas leakage at the gas supply pipe is identified as the highest risk among the hazards identified through the HAZOP study and as a safety instrumented function the shutdown function for leakage was defined. SIL 2 and PFD($2.5{\cdot}10^{-3}$) for the shutdown function are determined by the layer of protection analysis(LOPA). The shutdown system(SIS) carrying out the shutdown function(SIF) is verified and designed according to qualitative and quantitative requirements of IEC 61508 and IEC 61511. As a result of SIL verification and SIS conceptual design, the shutdown system is composed of two gas detectors voted 1oo2, one programmable logic solver, and two shutdown valve voted 1oo2.