• 인터넷정보학회논문지
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• 제17권5호
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• pp.33-42
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• 2016

Pin은 프로그램 동적 분석 도구를 생성할 수 있는 프레임워크로, 리눅스와 윈도우에서 사용자 영역의 프로그램 분석을 수행할 수 있게 한다. 역공학 방지 프로그램이나 악성코드는 프로그램 분석을 방해하는 안티디버깅이 적용되어 있기 때문에 Pin을 사용한 분석이 어렵다. 본 논문에서는 Pin을 이용해서 프로그램에 적용된 안티디버깅을 우회하여 동적 분석을 진행할 수 있는 Pin 플러그인 프로그램을 설계한 내용과 구현한 내용을 제안한다. Pin 탐지 안티디버깅을 우회할 수 있는 각각의 Pin 코드를 작성하고, Pin 코드를 하나로 합쳐 여러 안티디버깅을 우회할 수 있는 Pin 도구를 구현한다. 구현된 Pin 도구는 안티디버깅을 지원하는 프로텍터로 생성한 파일로 안티디버깅 우회 실험을 진행한다. 본 기법은 추후에 발견되는 안티디버깅 우회 코드 작성의 참고자료가 될 것이고 발견된 안티디버깅에 맞춰 수정 후 추가 적용이 가능할 것으로 예상된다.

• 한국정보통신학회논문지
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• 제15권12호
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• pp.2670-2678
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• 2011

IPTV 기술의 발전으로 인하여 사용자는 지리적 위치에 상관없이 멀티미디어 데이터의 서비스를 손쉽게 제공받고 있다. 그러나 가입 가구내 IPTV 게임을 불법적으로 서비스 받으려는 사용자 또한 증가하고 있다. 이 논문에서는 IPTV 서비스를 제공받는 청소년이 청소년이용불가 게임에 쉽게 접근할 수 없도록 PIN코드를 이용한 IPTV 가구내 사용자 개별 인증 프로토콜을 제안한다. 제안 프로토콜은 청소년이 청소년이용불가 게임에 불법적으로 접근하는 것을 예방하기 위해서 PIN 코드를 이용한 일회용 패스워드를 생성하여 IPTV 가구내 주어진 권한 정보를 쌍으로 묶어 인증서버와 셋톱박스에 저장하여 청소년이 강제적으로 게임에 접근하려는 시도를 예방한다.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3133-3150
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• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.797-806
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• 2009

The Multichannel Analyzer for Transient and steady-state in Rod Array - Liquid Metal Reactor for Flow Blockage analysis (MATRA-LMR-FB) code for the analysis of a subchannel blockage has been developed and evaluated through several experiments. The current version of the code is improved here by the implementation of a distributed resistance model which accurately considers the effect of flow resistance on wire spacers, by the addition of a turbulent mixing model, and by the application of a hybrid scheme for low flow regions. Validation calculations for the MATRA-LMR-FB code were performed for Oak Ridge National Laboratory (ORNL) 19-pin tests with wire spacers and Karlsruhe 169-pin tests with grid spacers. The analysis of the ORNL 19-pin tests conducted using the code reveals that the code has sufficient predictive accuracy, within a range of 5 $^{\circ}C$, for the experimental data with a blockage. As for the results of the analyses, the standard deviation for the Karlsruhe 169-pin tests, 0.316, was larger than the standard deviation for the ORNL 19-pin tests, 0.047.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3543-3562
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• 2021

The methods and performance of a 3D pin-by-pin neutronics code based on the 2D/1D decoupling method are presented. The code was newly developed as an effort to achieve enhanced accuracy and high calculation performance that are sufficient for the use in practical nuclear design analyses. From the 3D diffusion-based finite difference method (FDM) formulation, decoupled planar formulations are established by treating pre-determined axial leakage as a source term. The decoupled axial problems are formulated with the radial leakage source term. To accelerate the pin-by-pin calculation, the two-level coarse mesh finite difference (CMFD) formulation, which consists of the multigroup node-wise CMFD and the two-group assembly-wise CMFD is implemented. To enhance the accuracy, both the discontinuity factor method and the super-homogenization (SPH) factor method are examined for pin-wise cross-section homogenization. The parallelization is achieved with the OpenMP package. The accuracy and performance of the pin-by-pin calculations are assessed with the VERA and APR1400 benchmark problems. It is demonstrated that pin-by-pin 2D/1D alternating calculations within the two-level 3D CMFD framework yield accurate solutions in about 30 s for the typical commercial core problems, on a parallel platform employing 32 threads.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• 한국소성가공학회:학술대회논문집
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• 한국소성가공학회 2009년도 춘계학술대회 논문집
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• pp.458-461
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• 2009

Thermo-mechanical simulation of the Friction Stir Spot Welding (FSSW) processes was performed for the AA5083-H18 sheets, utilizing commercial Finite Element Method (FEM) and Finite Volume Method (FVM) which are based on Lagrangian and Eulerian formulations, respectively. The Lagrangian explicit dynamic FEM code, PAM-CRASH, and the Eulerian Computational Fluid Dynamics (CFD) FVM code, STAR-CD, were utilized to understand the effect of pin geometry on weld strength and material flow under the unsteady state condition. Using FVM code, material flow pattern near the tool boundary was analyzed to explain the weld strength difference between the weld by cylindrical pin and the weld by triangular pin, while the frictional energy concept using the FEM code had limitation to explain the weld strength difference.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3388-3400
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• 2023

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1339-1345
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• 2017

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.21-27
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• 1996

KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER validation analyses, which are in progress aiming to submit the Uncertainty Topical Report to KINS in the first half of 1996, include global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations. The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification results are in details presented in the separate papers.