• Nuclear Engineering and Technology
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• 제54권1호
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• pp.138-151
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• 2022

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident. Most severe accident analysis codes determine the cladding rupture based on simple parameter models. In this paper, a FRTMB module was developed using the thermal-mechanical model to analyze the fuel mechanical behavior. The purpose is to judge the cladding rupture with the severe accident analysis code. The FRTMB module was integrated into the self-developed severe accident analysis code ISAA to simulate the PHEBUS FPT3 experiment. The predicted rupture time and temperature of the cladding were basically consistent with the measured values, which verified the correctness and effectiveness of the FRTMB module. The results showed that the rising of gas pressure in the fuel rod and high temperature led to cladding ballooning. Consequently, the cladding hoop strain exceeded the strain limit, and the cladding burst. The developed FRTMB module can be applied not only to rod-type fuel, but also to plate-type fuel and other types of reactor fuel rods. Moreover, the FRTMB module can improve the channel blockage model of ISAA code and make contributions to analyzing the effect of clad ballooning on transient and subsequent parts of core degradation.

• 한국시뮬레이션학회논문지
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• 제19권2호
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• pp.9-16
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• 2010

본 사례 연구는 가상환경에서 자동차 의장 라인의 PLC 코드 검증 시스템을 구현하였다. 그간 인력과 PLC 검증 시간을 줄이고자 표준화된 PLC 에디터를 개발하려고 노력해왔으나, 실제 현장 환경에서 PLC 에디터를 적용하는데 어려움이 있었다. 본 연구는 현장의 PLC 모듈과 통신하며, 특별한 프로토콜을 이용하여 연결 환경을 구축하는 가상장비를 개발했다. PLC 모듈과 가상장비가 OPC 프로토콜에 의해 실시간으로 통제되므로 실제 장비를 설치하지 않고 PLC 코드를 검증할 수 있게 되었다. 실험적 모델을 자동차 공장의 최종 조립라인에 적용하고, DELMIA Automation을 이용하여 검증하였다. 결론적으로 PLC 코드 검증 과정과 가상장비 작동에 이 시스템이 유용하게 이용되었다. 이 시스템은 PLC 코드 개발 시간을 절약하며 조립라인의 생산성, 무결성을 높이는데 기여했다.

• 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.

• 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.

• 대한조선학회논문집
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• 제49권1호
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• pp.6-13
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• 2012

As ocean resources in shallow water areas are being exhausted, deep sea development is becoming common these days. Therefore floating type offshore structures are more competitive than fixed type structures, and FPSO is the most popular one these days. FPSO's are generally operated in a specific region and positioned to meet mostly head or bow waves in order to reduce roll motions. However this makes these vessels more vulnerable to green water around the bow region, and therefore the bow shape must be properly designed to mitigate green water damage. In the present study, experimental results for three different FPSO bow shapes in regular head waves were analyzed and compared to each other. Also CFD computations were carried out as a sample validation case for the database built for CFD code validation.

• 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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• 제26권3호
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• pp.215-224
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• 2001

In order to comply with revised national regulationson radiological protection and to implement recent nuclear data and dose conversion factors, KOPEC developed a new point kernel gamma and beta ray shielding analysis computer program. This new code, named VisualShield, adopted mass attenuation coefficient and buildup factors from recent ANSI/ANS standards and flux-to-dose conversion factors from the International Commission on Radiological Protection (ICRP) Publication 74 for estimation of effective/equivalent dose recommended in ICRP 60. VisualShieid utilizes graphical user interfaces and 3-D visualization of the geometric configuration for preparing input data sets and analyzing results, which leads users to error free processing with visual effects. Code validation and data analysis were performed by comparing the results of various calculations to the data outputs of previous programs such as MCNP 4B, ISOSHLD-II, QAD-CGGP, etc.

• 한국분무공학회지
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• 제2권1호
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• pp.36-45
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• 1997

A spray-wall impingement model for fuel sprays is proposed and implemented as a module into the KIVA-POSTECH code. The model is based on the single droplet experiments. The droplet behaviors after impingement are determined from experimental correlations. Different behaviors of impinged droplets depend on the wall temperature and the critical temperature of the fuel. Fuel film formation is taken into account so that the model can be applicable to any wall temperature and injection conditions. Computational results on a normal and on inclined wall are in good agreement for the spray shape and penetration. More validation against experiments and development of the heat transfer model are needed for further improvement.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.542-547
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• 1997

In the SSC(Super System Code), the reactivity feedback models of the Doppler effect and fuel axial expansion were modified to evaluate the safety performance of the metal-fueled core. The core radial expansion model was developed and implemented into the code as well. The transient analyses have been performed by the modified SSC for UTOP, ULOHS, ULOF/LOHS, and UTOP/LOF/LOHS events for one of the core design options being considered. Analysis results shows that the reactivity feedbacks can provide an inherent shutdown capability in response to key anticipated events without scram. Development of other reactivity feedback models and validation of these models against experimental data would make the SSC suitable for the assessment of the metal-fueled core safety performance.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.237-244
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• 2016

Background: Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). Materials and Methods: This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. Results and Discussion: The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ${\pm}6%$. Conclusion: It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.