• 한국안전학회지
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• 제27권5호
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• pp.219-223
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• 2012

For an off-site consequence analysis at nuclear power plant, MELCOR Accident Consequence Code System(MACCS) II code is widely used as a software tool. In this study, the algorithm of web-based off-site consequence analysis program(OSCAP) using the MACCS II code was developed for an Integrated Leak Rate Test (ILRT) interval extension and Level 3 probabilistic safety assessment(PSA), and verification and validation(V&V) of the program was performed. The main input data for the MACCS II code are meteorological, population distribution and source term information. However, it requires lots of time and efforts to generate the main input data for an off-site consequence analysis using the MACCS II code. For example, the meteorological data are collected from each nuclear power site in real time, but the formats of the raw data collected are different from each site. To reduce the efforts and time for risk assessments, the web-based OSCAP has an automatic processing module which converts the format of the raw data collected from each site to the input data format of the MACCS II code. The program also provides an automatic function of converting the latest population data from Statistics Korea, the National Statistical Office, to the population distribution input data format of the MACCS II code. For the source term data, the program includes the release fraction of each source term category resulting from modular accident analysis program(MAAP) code analysis and the core inventory data from ORIGEN. These analysis results of each plant in Korea are stored in a database module of the web-based OSCAP, so the user can select the defaulted source term data of each plant without handling source term input data.

• 에너지공학
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• 제24권2호
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• pp.91-102
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• 2015

관류식 직관형 증기발생기의 열수력 설계와 성능분석을 위한 해석코드를 개발하였다. 개발한 물리적 모델과 수치 해석 코드를 검증하기 위해 설계 제원이 공개되어 사용되고 있는 관류식 직관형 증기발생기를 개발된 코드를 이용해 해석하고 설계 자료와 비교하였다. 또한 동일한 증기발생기를 최적 열수력 안전해석코드인 MARS를 이용하여 해석한 뒤 비교분석하였다. 열전달면적, 압력 및 온도분포 등의 계산 결과는 설계 자료 및 MARS 코드의 계산 결과와 대부분 일치하게 나타났다. 최종적으로 개발된 코드가 직관형 증기발생기의 열적 설계 최적화 및 민감도 분석을 목적으로 폭넓게 사용될 수 있음을 확인하였다.

• 한국소프트웨어감정평가학회 논문지
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• 제10권1호
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• pp.19-26
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• 2014

본 논문에서는 안드로이드 애플리케이션(앱)을 역공학 공격으로부터 방어하는 기법을 제안한다. 이 기법에서 서버는 안드로이드 패키지 파일인 APK 내에 있는 원본 실행코드(DEX)를 암호화하고, 실행 시 이를 복호화 할 수 있는 스텁(stub) 코드를 APK에 삽입하여 배포한다. 스텁 코드는 자신에 대한 공격을 탐지하기 위해 무결성 검증 코드를 포함한다. 사용자가 해당 APK를 설치·실행할 때, 스텁 코드는 자체의 무결성을 검증한 후, 암호화된 원본 실행코드를 복호화하고, 이를 동적 로딩(dynamic loading)하여 실행한다. 앱의 원본 실행코드는 암호화되어 배포되므로 지적재산권을 효과적으로 보호할 수 있다. 또한, 스텁 코드에 대해 무결성을 검증하므로, 제안 기법의 우회 가능성을 차단한다. 우리는 15개의 안드로이드 앱에 제안 기법을 적용하여 그 유효성을 평가하였다. 실험 결과, 13개의 앱이 정상적으로 동작함을 확인하였다.

• Nuclear Engineering and Technology
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• 제46권5호
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• pp.641-654
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• 2014

A new computer code, named CORONA (Core Reliable Optimization and thermo-fluid Network Analysis), was developed for the core thermo-fluid analysis of a prismatic gas cooled reactor. The CORONA code is targeted for whole-core thermo-fluid analysis of a prismatic gas cooled reactor, with fast computation and reasonable accuracy. In order to achieve this target, the development of CORONA focused on (1) an efficient numerical method, (2) efficient grid generation, and (3) parallel computation. The key idea for the efficient numerical method of CORONA is to solve a three-dimensional solid heat conduction equation combined with one-dimensional fluid flow network equations. The typical difficulties in generating computational grids for a whole core analysis were overcome by using a basic unit cell concept. A fast calculation was finally achieved by a block-wise parallel computation method. The objective of the present paper is to summarize the motivation and strategy, numerical approaches, verification and validation, parallel computation, and perspective of the CORONA code.

• Nuclear Engineering and Technology
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• 제41권7호
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• pp.953-968
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• 2009

It is important to accurately predict the temperature and density distributions in large stratified enclosures both for design optimization and accident analysis. Current reactor system analysis codes only provide lumped-volume based models that can give very approximate results. Previous scaling analysis has shown that stratified mixing processes in large stably stratified enclosures can be described using one-dimensional differential equations, with the vertical transport by jets modeled using integral techniques. This allows very large reductions in computational effort compared to three-dimensional CFD simulation. The BMIX++ (Berkeley mechanistic MIXing code in C++) code was developed to implement such ideas. This paper summarizes major models for the BMIX++ code, presents the two-plume mixing experiment simulation as one validation example, and describes the codes' application to the liquid salt buffer pool system in the AHTR (Advanced High Temperature Reactor) design. Three design options have been simulated and they exhibit significantly different stratification patterns. One of design options shows the mildest thermal stratification and is identified as the best design option. This application shows that the BMIX++ code has capability to provide the reactor designers with insights to understand complex mixing behavior with mechanistic methods. Similar analysis is possible for liquid-metal cooled reactors.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.43-54
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• 2016

In the present paper, development of the three-dimensional (3D) computational code based on Galerkin finite element method (GFEM) for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA)-3D and Water-Water Energetic Reactor (VVER)-1000 reactor cores. In addition, validation of the calculations against the $P_1$ approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.

• 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.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4460-4473
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• 2022

The behaviour of the fission gas plays an important role in the fuel rod performance. In a previous work, we presented a physics-based model describing intra- and inter-granular behaviour of radioactive fission gas. The model was implemented in SCIANTIX, a mesoscale module for fission gas behaviour, and assessed against the CONTACT 1 irradiation experiment. In this work, we present the multi-scale coupling between the TRANSURANUS fuel performance code and SCIANTIX, used as mechanistic module for stable and radioactive fission gas behaviour. We exploit the coupled code version to reproduce two integral irradiation experiments involving standard fuel rod segments in steady-state operation (CONTACT 1) and during successive power transients (HATAC C2). The simulation results demonstrate the predictive capabilities of the code coupling and contribute to the integral validation of the models implemented in SCIANTIX.

• Structural Engineering and Mechanics
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• 제89권4호
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• pp.363-374
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• 2024

Significant changes have been made to estimate the punching shear capacity for edge column-slab joints in the latest editions of most current codes. The revised equations account for axial forces as well as moments conveyed to columns from slabs, which have a substantial impact on the punching resistance of such joints. Many key design parameters, such as reinforcement-ratio, concrete strength, size-effect, and critical-section perimeter, were treated differently or even ignored in various code provisions. Consequently, wide ranges of predicted punching shear strength were detected by applying different code formulas. Therefore, it is essential to assess the various current Codes' design-equations. Because of the similarity in estimated outcomes, only the ACI, EC, and SNiP are used in this study to cover a wide range of estimation ranges from highly conservative to unconservative. This paper is devoted to analyzing the techniques in these code provisions, comparing the estimated punching resistance with available experimental data, and finally developing efficient models predicting the punching capacity of edge column-slab connections. 63 samples from past investigations were chosen for validation. To appropriately predict the punching shear, newly updated equations for ACI and SNiP are provided based on nonlinear regression analysis. The proposed equations'results match the experimental data quite well.

• 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.