• Journal of Radiation Protection and Research
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• 제41권2호
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• pp.123-131
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• 2016

Background: With the increase in the number of particle accelerator facilities under either operation or construction, the accurate calculation using Monte Carlo codes become more important in the shielding design and radiation safety evaluation of accelerator facilities. Materials and Methods: The calculations with different physics models were applied in both of cases: using only physics model and using the mix and match method of MCNPX code. The issued conditions were the interactions of 600 MeV proton and $290MeV{\cdot}n^{-1}$ oxygen with a carbon target. Both of cross-section libraries, JENDL High Energy File 2007 (JENDL/HE-2007) and LA150, were tested in this calculation. In the case of oxygen ion interactions, the calculation results using LAQGSM physics model and JENDL/HE-2007 library were compared with D. Satoh's experimental data. Other Monte Carlo calculations using PHITS and FLUKA codes were also carried out for further benchmarking study. Results and Discussion: It was clearly found that the physics models, especially intra-nuclear cascade model, gave a great effect to determine proton-induced secondary neutron spectrum in MCNPX code. The variety of physics models related to heavy ion interactions did not make big difference on the secondary particle productions. Conclusion: The variations of secondary neutron spectra and particle transports depending on various physics models in MCNPX code were studied and the result of this study can be used for the shielding design and radiation safety evaluation.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.280-291
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• 2018

In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010-2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II) was thus launched early in 2014, which has been organized in two complementary activities: (1) comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2) assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1715-1720
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• 2019

Jozef Stefan Institute (JSI), TRIGA Mark II reactor employs the homogeneous mixture of uranium and zirconium hydride fuel type. Since its upgrade, a series of fresh fuel steady state experimental benchmarks have been conducted. The benchmark results have provided data for testing computational neutronics codes which are important for reactor design and safety analysis. In this work, we investigated the JSI TRIGA Mark II reactor neutronics characteristics: the effective multiplication factor and two safety parameters, namely the control rod worth and the fuel temperature reactivity coefficient using SuperMC. The modeling and real-time cross section generation methods of SuperMC were evaluated in the investigation. The calculation analysis indicated the following: the effective multiplication factor was influenced by the different cross section data libraries; the control rod worth evaluation was better with Monte Carlo codes; the experimental fuel temperature reactivity coefficient was smaller than calculated results due to change in water temperature. All the results were in good agreement with the experimental values. Hence, SuperMC could be used for the designing and benchmarking of other TRIGA Mark II reactors.

• 정보보호학회논문지
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• 제32권2호
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• pp.171-179
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• 2022

자체 수정 코드(Self-Modifying-Code)란 실행 시간 동안 스스로 실행 코드를 변경하는 코드를 말한다. 이런 기법은 특히 악성코드가 정적 분석을 우회하는 데 악용된다. 따라서 이러한 악성코드를 효과적으로 검출하려면 자체 수정 코드를 파악하는 것이 중요하다. 그동안 동적 분석 방법으로 자체 수정 코드를 분석해왔으나 이는 시간과 비용이 많이 든다. 만약 정적 분석으로 자체 수정 코드를 검출할 수 있다면 악성코드 분석에 큰 도움이 될 것이다. 본 논문에서는 LLVM IR로 변환한 바이너리 실행 프로그램을 대상으로 자체 수정 코드를 탐지하는 정적 분석 방법을 제안하고, 자체 수정 코드 벤치마크를 만들어 이 방법을 적용했다. 본 논문의 실험 결과 벤치마크 프로그램을 컴파일로 변환한 최적화된 형태의 LLVM IR 프로그램에 대해서는 설계한 정적 분석 방법이 효과적이었다. 하지만 바이너리를 리프팅 변환한 비정형화된 LLVM IR 프로그램에 대해서는 자체 수정 코드를 검출하기 어려운 한계가 있었다. 이를 극복하기 위해 바이너리를 리프팅 하는 효과적인 방법이 필요하다.

• Nuclear Engineering and Technology
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• 제42권6호
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• pp.656-661
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• 2010

Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. By utilizing the HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER$^2$) facility, a thermal hydraulic benchmarking study has been conducted for the prediction of pressure loss in lead-alloy cooled advanced nuclear energy systems (LACANES). The loop has several complex components that cannot be readily characterized with available pressure loss correlations. Among these components is the core, composed of a vessel, a barrel, heaters separated by complex spacers, and the plenum. Due to the complex shape of the core, its pressure loss is comparable to that of the rest of the loop. Detailed CFD simulations employing different CFD codes are used to determine the pressure loss, and it is found that the spacers contribute to nearly 90 percent of the total pressure loss. In the system codes, spacers are usually accounted for; however, due to the lack of correlations for the exact spacer geometry, the accuracy of models relies strongly on assumptions used for modeling spacers. CFD can be used to determine an appropriate correlation. However, application of CFD also requires careful choice of turbulence models and numerical meshes, which are selected based on extensive experience with liquid metal flow simulations for the KALLA lab. In this paper consistent results of CFX and Star-CD are obtained and compared to measured data. Measured data of the pressure loss of the core are obtained with a differential pressure transducer located between the core inlet and outlet at a flow rate of 13.57kg/s.

• 한국정보처리학회:학술대회논문집
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• 한국정보처리학회 2001년도 추계학술발표논문집 (상)
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• pp.313-316
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• 2001

At the end of 1999, the TeraCluster project in the KISTI Supercomputing Center was initiated to explore the possibility of PC clusters as a scientific computing platform to replace the Cray T3E system in KISTI by the year 2002. In order to understand whether an application is scalable to tera-flops sized cluster system, running test is inevitable. Extensive performance tests using well-known benchmarking codes with real applications' characteristics in them were carried out with different combinations of CPUs, system boards, network devices. The lessen learned shows the relationships between system performances and varied applications' different needs resulting in promises of How-Tos in building large scale cluster system. The 64/16 node clusters with Alpha EV6(466MHz), Pentium III(667 MHz) i inter-node network of Fast Ethernet. SCI[1] and Myrinet[2] were evaluated. More detailed specifications of the Linux clusters are described in Table 1.

• 한국소프트웨어감정평가학회 논문지
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• 제17권1호
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• pp.51-58
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• 2021

최근 주목받고 있는 마이크로서비스는 독립적으로 개발될 뿐만 아니라 독립적으로 실행 및 배포가 가능하다는 장점 때문에, 클라우드 컴퓨팅 환경에서 보다 유연한 확장과 효율적인 협력을 보장할 수 있다. 이러한 영향으로 최근 마이크로서비스지향 애플리케이션 환경으로의 전환이 급격히 증가하고 있다. 마이크로서비스의 도입을 위해서는 무엇보다 모노리식 아키텍처로 구축된 단일 애플리케이션의 구성요소를 마이크로서비스 단위로 식별하는 문제가 선결되어야 한다. 본 논문에서는 레거시 시스템으로부터 마이크로서비스 식별의 문제를 알고리즘 기반으로 해결하기 이한 접근방법을 제안한다. 코드의 메타정보를 이용하여 그래프를 생성하고 클러스터링 알고리즘을 적용하여 마이크로서비스 후보를 추출한다. 추출된 마이크로서비스 후보에 대해 메트릭을 이용하여, 모듈화 품질을 평가한다. 또한 제안된 식별 방법의 효과를 검증하기 위해 벤치마크를 위해 많이 사용되는 공개 소프트웨어의 코드를 이용하여 후보 서비스를 도출하고, 메트릭을 이용하여 모듈화 수준을 평가한다. 결과적으로 좀더 작은 단위의 마이크로서비스로 식별해 내면서 모듈품질을 향상시키는 결과를 확인할 수 있다.