• Nuclear Engineering and Technology
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• 제41권1호
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• pp.91-98
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• 2009

The "3+3 Process" for safety critical software for nuclear power plants' I&C (Instrumentation and Control system) has been developed in this work. The main idea of the "3+3 Process" is both to simplify the software development and safety analysis in three steps to fulfill the requirements of a software safety plan [1]. The "3-Step" software development process consists of formal modeling and simulation, automated code generation and coverage analysis between the model and the generated source codes. The "3-Step" safety analysis consists of HAZOP (hazard and operability analysis), FTA (fault tree analysis), and DV (design validation). Put together, these steps are called the "3+3 Process". This scheme of development and safety analysis minimizes the V&V work while increasing the safety and reliability of the software product. For assessment of this process, validation has been done through prototyping of the SDS (safety shut-down system) #1 for PHWR (Pressurized Heavy Water Reactor).

• 한국정보통신학회:학술대회논문집
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• 한국정보통신학회 2012년도 추계학술대회
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• pp.495-497
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• 2012

의료기기 소프트웨어 테스트 위험관리 적용 방안에 대해 연구하였다. 먼저 설문조사를 통해 의료기기 제조업체의 소프트웨어 밸리데이션 및 위험관리 현황을 분석하고 두 번째로 블랙박스 테스트와 화이트 박스 테스트를 비교하여 차이점을 분석하였다. 세 번째로 소프트웨어 분석 도구를 활용한 코드 분석 후 잠재적인 위해요인을 도출하고 이를 정량화 하였으며, 마지막으로 도출된 위해요인을 FMEA 기법을 이용하여 의료기기 위험관리 프로세스에 적용할 수 있도록 프레임워크를 구축하였다. 이를 통해 의료기기 품질관리(GMP) 업무 중 소프트웨어 밸리데이션 및 위험관리 프로세스를 구축하기 어려운 제조업체를 위한 업무 지원에 활용하고자 한다.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.191-195
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• 2016

Background: The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Materials and Methods: Validation calculations with recent nuclear data evaluations ENDF/BV-II. 0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. Results and Discussion: It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and ${chi}^2$ values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the $k_{eff}$ values. Conclusion: It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.

• 한국전산유체공학회지
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• 제7권2호
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• pp.25-32
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• 2002

The velocity and temperature profiles in the cabin of the automobile affect greatly the comfortableness of passengers. In this paper, the three dimensional flow and temperature analysis in the cabin of real automobile have been peformed. The three dimensional Navier-Stokes equation solver was validated by comparing with the other numerical data of a 1/5 scale model. The temperature field of cavity was also analyzed for the validation of energy equation solver. After the code validation, the numerical analysis of real field of flow and temperature of an automobile was peformed and the present result provides the insight of flow and temperature field of the inside of cabin.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 학술대회
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• pp.79-83
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• 2008

An approach to CFD code validation is developed that gives proper consideration to experimental and simulation uncertainties. The comparison errors include the difference between the data, simulation values and represents the combination of all errors. The uncertainties of modeling and numerical analysis in the CFD prediction were estimated by a Coleman's theory. In this paper, the numerical solutions are calculated by A-type standard uncertainty and Richardson extrapolation Method.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년 추계학술대회논문집
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• pp.79-83
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• 2008

An approach to CFD code validation is developed that gives proper consideration to experimental and simulation uncertainties. The comparison errors include the difference between the data, simulation values and represents the combination of all errors. The uncertainties of modeling and numerical analysis in the CFD prediction were estimated by a Coleman's theory. In this paper, the numerical solutions are calculated by A-type standard uncertainty and Richardson extrapolation Method.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.34-45
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• 2001

We have needs to develop a performance analysis system that can be used in domestic nuclear power plants to determine performance status of turbine cycle. We developed new NOPAS system to aid performance analysis of turbine cycle . Procedures of performance calculation are improved using several adaptations from standard calculation algorithms based on ASME (American Society of Mechanical Engineers) PTC (Performance Test Code). Robustness in the performance analysis is increased by verification & validation scheme for measured input data. The system also provides useful aids for performance analysis such as graphic heat balance of turbine cycle and components, turbine expansion lines, automatic generation of analysis reports.

• 대한기계학회논문집B
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• 제29권3호
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• pp.409-416
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• 2005

We need to develop a on-line thermal performance analysis system for nuclear power plant to determine performance status and heat rate of turbine cycle. We have developed PERUPS(PERformance Upgrade System) to aid the effective performance analysis of turbine cycle. Procedures of performance calculation are improved using several adaptations from standard calculation algorithms based on PTC(Performance Test Code). Robustness in the on-line performance analysis is increased by verification & validation scheme for measured input data. The system also provides useful web interfaces for performance analysis such as graphic heat balance of turbine cycle and components, turbine expansion lines, automatic generation of analysis report. The system was successfully applied for YongGwang nuclear plant unit #3,4.

• Nuclear Engineering and Technology
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• 제37권2호
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• pp.167-176
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• 2005

We developed the PERUPS system to aid the on-line performance analysis for the turbine cycle of the YongGwang 3 and 4 nuclear power plants. Procedure of measurement validation is included in the performance calculation to obtain heat balance. Precision of on-line performance calculation is increased via practical modifications of standard calculation algorithms based on the PTC (Performance Test Code). The proposed system also provides useful Web-based aids for performance analysis, including performance data management, a graphic viewer for heat balance and turbine expansion lines, and synthesized reports of performance.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2003년도 추계 학술발표회논문집
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• pp.49-57
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• 2003

Lately, a guideline for seismic design of electric substation facilities has been newly prepared and issued. The guideline adopts a new simplified dynamic analysis technique called "pseudo-resonance method" from the design code of Japan. The method can effectively consider resonance effect of structural system during earthquake motion. However, reliability and practical applicability of the method have not been fully guaranteed under the different seismic and engineering situations in Korea. This paper presents a validation study of the pseudo-resonance method for practical seismic analysis. For this purpose, various parametric studies have been performed using recorded earthquake motions and idealized substation equipment models. The results showed that the pseudo-resonance method can be effectively used for the practical seismic design of substation facilities in Korea.