• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1789-1803
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• 2022

RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate the feasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal code verification results for the MET-1000 and CAR-3600 benchmark cores. Three different multi-group cross-section calculation schemes are employed to improve the agreement between the nodal and reference solutions. The reference solution is obtained by the MCS code using continuous-energy nuclear data. Additionally, the MCS/RAST-F nodal solution is verified with results based on cross-section generated by collision probability code TULIP. A good agreement between MCS/RAST-F and reference solution is observed with less than 120 pcm discrepancy in k_eff and less than 1.2% root-mean-square error in power distribution. This study confirms the two-step approach MCS/RAST-F as a reliable tool for the three-dimensional simulation of reactor cores with fast spectrum.

• 정보처리학회논문지A
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• 제14A권3호
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• pp.167-172
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• 2007

가상기계(Virtual Machine)는 소프트웨어로 제작되어 논리적인 시스템 구성을 갖는 컴퓨터이기 때문에 그 수행 속도와 필요 저장 공간 측면에서 성능이 떨어질 수밖에 없다. 이러한 환경에서의 가상기계 코드 최적화는 실행 성능을 향상시킬 수 있기에 중요하다. 특별히 임베디드 장치(Embedded Device)에서 작동하는 가상기계 환경에서의 최적화는 기존의 최적화에 비해 수행 비용 대비 효과에서 높은 효율을 요구한다. 이에 따라 프로파일링을 통하여 성능에 크게 영향을 주는 함수 또는 기본 블록(Basic Block)을 찾아 최적화하는 것이 효과적이다. 본 논문에서는 프로파일링을 이용한 가상기계 코드 최적화기를 설계하고 구현하였다. 먼저, 가상기계 코드 최적화를 위해 코드를 실행하여 얻을 수 있는 동적 정보인 프로파일링 데이터(Profiling Data)를 정의하였고, 프로파일링 정보를 이용한 가상기계 코드 최적기를 구현하였다. 또한, 구현과 실험에 있어서 가상기계 코드는 EVM(Embedded Virtual Machine)의 중간 언어인 SIL(Standard Intermediate Language)를 사용하였고, 구현된 최적화기에 대한 실험을 통해 최적화기의 효과를 확인하였다.

• Journal of Mechanical Science and Technology
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• 제20권12호
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• pp.2218-2229
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• 2006

A conservative finite-volume numerical method for unstructured grids with the cell-centered method has been developed for computing flow and heat transfer by combining the attractive features of the existing pressure-based procedures with the advances made in unstructured grid techniques. This method uses an integral form of governing equations for arbitrary convex polyhedra. Care is taken in the discretization and solution procedure to avoid formulations that are cell-shape-specific. A collocated variable arrangement formulation is developed, i.e. all dependent variables such as pressure and velocity are stored at cell centers. For both convective and diffusive fluxes the forms superior to both accuracy and stability are particularly adopted and formulated through a systematic study on the existing approximation ones. Gradients required for the evaluation of diffusion fluxes and for second-order-accurate convective operators are computed by using a linear reconstruction based on the divergence theorem. Momentum interpolation is used to prevent the pressure checkerboarding and a segregated solution strategy is adopted to minimize the storage requirements with the pressure-velocity coupling by the SIMPLE algorithm. An algebraic solver using iterative preconditioned conjugate gradient method is used for the solution of linearized equations. The flow analysis code (PowerCFD) developed by the present method is evaluated for its application to several 2-D structured-mesh benchmark problems using a variety of unstructured quadrilateral and triangular meshes. The present flow analysis code by using unstructured grids with the cell-centered method clearly demonstrate the same accuracy and robustness as that for a typical structured mesh.

• JSTS:Journal of Semiconductor Technology and Science
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• 제16권5호
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• pp.582-594
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• 2016

Various test data compression techniques have been developed to reduce the test costs of system-on-a-chips. In this paper, a scan chain reordering algorithm for code-based test data compression techniques is proposed. Scan cells within an acceptable relocation distance are ranked to reduce the number of conflicts in all test patterns and rearranged by a positioning algorithm to minimize the routing overhead. The proposed method is demonstrated on ISCAS '89 benchmark circuits with their physical layout by using a 180 nm CMOS process library. Significant improvements are observed in compression ratio and test power consumption with minor routing overhead.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1388-1395
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• 2017

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal-hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal-hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.356-368
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• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2274-2284
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• 2020

The Nuclear Fuel Cycle Code "ANICCA" has been developed by SCK•CEN to answer particular questions about the Belgian nuclear fleet. However, the wide range of capabilities of the code make it also useful for international or regional studies that include advanced technologies and strategies of cycle. This paper shows the main features of the code and the facilities that can be simulated. Additionally, a comparison between several codes and ANICCA has also been made to verify the performance of the code by means of a simulation proposed in the last NEA (OECD) Benchmark Study. Finally, a case study of the Belgian nuclear fuel cycle phase out has been carried out to show the possible impact of the transmutation of the minor actinides on the nuclear waste by the use of an Accelerator Driven System also known as ADS. Results show that ANICCA accomplishes its main purpose of simulating the scenarios giving similar outcomes to other codes. Regarding the case study, results show a reduction of more than 60% of minor actinides in the Belgian nuclear cycle when using an ADS, reducing significantly the radiotoxicity and decay heat of the high-level waste and facilitating its management.

• 에너지공학
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• 제23권4호
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• pp.112-122
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• 2014

A main steam line break (MSLB) test at the ATLAS facility was simulated using the best-estimate thermal-hydraulic system code, MARS-KS. This has been performed as an activity at the third domestic standard problem for code benchmark (DSP-03) that has been organized by Korea Atomic Energy Research Institute (KAERI). The results of the MSLB experiment and the MARS input data prepared for the previous DSP-02 using the ATLAS facility were provided to participants. The preliminary MSLB simulation using the base input data, however, showed unphysical results in the primary-to-secondary heat transfer. To resolve the problems, some improvements were implemented in the MARS input modelling. These include the use of fine meshes for the bottom region of the steam generator secondary side and proper thermal-hydraulics calculation options. Other input model improvements in the heat loss and the flow restrictor models were also made and the results were investigated in detail. From the results of simulations, the limitations and further improvement areas of the MARS code were identified.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.520-545
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• 2010

This paper presents the neutronics modeling capabilities of the fast reactor simulation system SHARP, which ANL is developing as part of the U.S. DOE's NEAMS program. We discuss the three transport solvers (PN2ND, SN2ND, and MOCFE) implemented in the UNIC code along with the multigroup cross section generation code $MC^2$-3. We describe the solution methods and modeling capabilities, and discuss the improvement needs for each solver, focusing on massively parallel computation. We present the performance test results against various benchmark problems and ZPR-6 and ZPPR critical experiments. We also discuss weak and strong scalability results for the SN2ND solver on the ZPR-6 critical assembly benchmarks.

• 한국의학물리학회지:의학물리
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• 제13권3호
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• pp.149-155
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• 2002

최근 들어 방사선 수송이론, 컴퓨터 하드웨어 성능, 및 병렬 연산 기법의 발전에 힘입어, 몬테카를로 기반의 선량계산 기법을 임상에 적용할 수 있게 되었다. 임상적용을 위해 개발된 몬테카를로 기반 선량계산 코드간의 계산 소요 시간과 정확도를 비교할 목적으로 제13차 ICCR (International Conference on the use of Computers in Radiation Therapy, Heidelberg, Germany, 2000) 학술대회에서 벤치마킹 절차서가 제안되었다. 최근, 본원에서도 임상적용을 목표로 28개의 인텔 펜티움 프로세서로 구성된 Linux cluster 시스템을 구축하고, 여기에 몬테카를로 선량계산을 위한 BEAMnrc 코드를 설치하였다. 본 연구의 목적은 위에서 제안된 벤치마킹 절차를 수행하여 본원에서 구축한 몬테카를로 선량계산 시스템의 정량적 성능 평가를 시도하고자 하는 것이었다. 벤치마킹 절차는 크게 다음의 세 과정으로 구성되어 있다. a) 30.5 cm $\times$ 39.5 cm $\times$ 30 cm 의 팬톰(5 ㎣ voxels) 에 대한 통계적 불확정도 2%이내 결과를 얻기 위한 광자선 선량계산 속도. b) 위 팬톰에 대한 전자선의 선량계산 속도. c) 비균질 평판 매질로 구성된 팬톰내 광자선 및 전자선의 선량계산 결과를 EGSr/PRESTA 계산 결과와 비교 제시. 18 MV 광자선에 대해 선량계산 속도 평가 결과 5.5분이 소용되었다. 전자선의 경우, 실제 계산 시간은 광자선에 비해 약 10배 정도 빨랐으나, 병렬 연산을 처리하기 위해 소용되는 추가 시간 때문에 전체 계산에 소요되는 시간은 광자선과 비슷하였다. 본 원에서 사용한 몬테카를로 코드는 EGSnrc로써 EGS4의 개선 버전으로 이들 간의 정확도 비교는 큰 의미가 없을 것으로 판단된다. 하지만 두 계산 결과가 기대했던 바와 같이 매우 잘 일치하였다. 결론적으로, 본원에서 구축한 몬테카를로 치료계획시스템은 임상적용에 무리가 없을 것으로 판단하였다. 추후 본 시스템을 본원에서 사용하는 상용 치료계획시스템과 인터페이스를 개발하여, 통합환경을 구축함으로써, 몬테카를로 기반의 치료계획시스템의 임상적용과 관련된 연구들을 수행해 나갈 계획이다.