• Nuclear Engineering and Technology
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• v.54 no.5
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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.

• The KIPS Transactions:PartA
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• v.14A no.3 s.107
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• pp.167-172
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• 2007

VM(Virtual Machine) can be considered as a software processor which interprets the machine code. Also, it is considered as a conceptional computer that consists of logical system configuration. But, the execution speed of VM system is much slower than that of a real processor system. So, it is very important to optimize the code for virtual machine to enhance the execution time. Especially the optimizer for a virtual machine code on embedded devices requires the highly efficient performance to the ordinary optimizer in the respect to the optimized ratio about cost. Fundamentally, functions and basic blocks which influence the execution time of virtual machine is found, and then an optimization for them nay get the high efficiency. In this paper, we designed and implemented the optimizer for the virtual(or abstract) machine code(VMC) using profiling. Firstly, we defined the profiling information which is necessary to the optimization of VMC. The information can be obtained from dynamically executing the machine code. And we implemented VMC optimizer using the profiling information. In our implementation, the VMC is SIL(Standard Intermediate Language) that is an intermediate code of EVM(Embedded Virtual Machine). Also, we tried a benchmark test for the VMC optimizer and obtained reasonable results.

• Journal of Mechanical Science and Technology
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• v.20 no.12
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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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• v.16 no.5
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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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• v.49 no.7
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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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• v.51 no.2
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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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• v.52 no.10
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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.

• Journal of Energy Engineering
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• v.23 no.4
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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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• v.42 no.5
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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.

• Progress in Medical Physics
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• v.13 no.3
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• pp.149-155
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• 2002

Recent advances in radiation transport algorithms, computer hardware performance, and parallel computing make the clinical use of Monte Carlo based dose calculations possible. To compare the speed and accuracies of dose calculations between different developed codes, a benchmark tests were proposed at the XIIth ICCR (International Conference on the use of Computers in Radiation Therapy, Heidelberg, Germany 2000). A Monte Carlo treatment planning comprised of 28 various Intel Pentium CPUs was implemented for routine clinical use. The purpose of this study was to evaluate the performance of our system using the above benchmark tests. The benchmark procedures are comprised of three parts. a) speed of photon beams dose calculation inside a given phantom of 30.5 cm$\times$39.5 cm $\times$ 30 cm deep and filled with 5 ㎣ voxels within 2% statistical uncertainty. b) speed of electron beams dose calculation inside the same phantom as that of the photon beams. c) accuracy of photon and electron beam calculation inside heterogeneous slab phantom compared with the reference results of EGS4/PRESTA calculation. As results of the speed benchmark tests, it took 5.5 minutes to achieve less than 2% statistical uncertainty for 18 MV photon beams. Though the net calculation for electron beams was an order of faster than the photon beam, the overall calculation time was similar to that of photon beam case due to the overhead time to maintain parallel processing. Since our Monte Carlo code is EGSnrc, which is an improved version of EGS4, the accuracy tests of our system showed, as expected, very good agreement with the reference data. In conclusion, our Monte Carlo treatment planning system shows clinically meaningful results. Though other more efficient codes are developed such like MCDOSE and VMC＋＋, BEAMnrc based on EGSnrc code system may be used for routine clinical Monte Carlo treatment planning in conjunction with clustering technique.