• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3526-3539
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• 2022

Nuclear thermal-hydraulic system analysis codes have been developed to simulate nuclear reactor systems, which solve simplified governing equations by replacing source terms with constitutive relations for simulating entire reactor systems with low computational resources. For half a century, many efforts have been made for wider versatility and higher accuracy of system codes, but various factors can affect the code analysis results, and it was difficult to isolate these factors and interpret them individually. In this study, two system codes, RELAP5 and TRACE, which have many users and are highly reliable, are selected to analyze only the effects of constitutive relations. The influence of constitutive relations is analyzed using in-house platforms that replicate constitute relations of RELAP5 and TRACE equally to exclude factors that may affect analysis results, such as governing equation solvers and user effects. Among the various constitutive relations, the analysis is performed on the wall variables expected to have the most influence on the analysis results. Part 1 paper presents the methodology and wall friction model comparison, while Part 2 paper shows wall heat transfer comparison of the two selected codes.

• 전자공학회논문지
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• 제50권8호
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• pp.3-14
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• 2013

본 논문에서는 어떠한 이진 입력 이산 퀀텀채널(quantum channel)이 주어지더라도 대칭 용량을 달성할 수 있는 qubit(quantum bit)를 생성하기 위해, 극(polar) 퀀텀 채널 코딩이라 부르는 퀀텀 채널의 결합과 분리 형태를 제시한다. 현재의 용량은 동등 확률을 갖는 임의의 qubit 입력에 따라서 결정된다. 퀀텀채널의 분극은 대칭채널이 1에 근접하면 rate 1로 아니면 rate 0으로 전송하는 채널을 통해 퀀텀 데이터를 부분적으로 전송하는 퀀텀 오류정정 부호화에 아주 적합하다.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.344-349
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• 2016

Background: Carbon ion therapy has achieved satisfactory results. However, patients have a risk to get a secondary cancer. In order to estimate the risk, it is essential to understand particle transportation and nuclear reactions in the patient's body. The particle transport Monte Carlo simulation code is a useful tool to understand them. Since the code validation for heavy ion incident reactions is not enough, the experimental data of the elementary reaction processes are needed. Materials and Methods: We measured neutron production double-differential cross-sections (DDXs) on a carbon bombarded with 430 MeV/nucleon carbon beam at PH2 beam line of HIMAC facility in NIRS. Neutrons produced in the target were measured with NE213 liquid organic scintillators located at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. Results and Discussion: Neutron production double-differential cross-sections for carbon bombarded with 430 MeV/nucleon carbon ions were measured by the time-of-flight method with NE213 liquid organic scintillators at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. The cross sections were obtained from 1 MeV to several hundred MeV. The experimental data were compared with calculated results obtained by Monte Carlo simulation codes PHITS, Geant4, and FLUKA. Conclusion: PHITS was able to reproduce neutron production for elementary processes of carbon-carbon reaction precisely the best of three codes.

• Current Optics and Photonics
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• 제3권6호
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• pp.531-540
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• 2019

Spectral Amplitude Coding Optical Code Division Multiple Access (SAC OCDMA) is an advanced technique in asynchronous environments. This paper proposes design and implementation of a novel quantum logic gate (QLG) code, with code construction algorithm generated without following any code mapping procedures for SAC system. The proposed code has a unitary matrices property with maximum overlap of one chip for various clients and no overlaps in spectra for the rest of the subscribers. Results indicate that a single algorithm produces the same length increment for codes with weight greater than two and follows the same signal to noise ratio (SNR) and bit error rate (BER) calculations for a higher number of users. This paper further examines the performance of a QLG code based SAC-OCDMA system with NAND and direct detection techniques. BER analysis was carried out for the proposed code and results were compared with existing MDW, RD and GMP codes. We demonstrate that the QLG code based system performs better in terms of cardinality, which is followed by improved BER. Numerical analysis reveals that for error free transmission (10^-9), the suggested code supports approximately 170 users with code weight 4. Our results also conclude that the proposed code provides improvement in the code construction, cross-correlation and minimization of noises.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.275-283
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• 2019

Rare isotope beam facilities require shielding data in early stage of their design. There is much less shielding data on neutrons from the reactions between heavy ion beams and matter than the data on neutrons produced by protons. The purpose of the present work is to produce and thus increase the amount of shielding data on neutrons generated by high-energy heavy ion beams based on the RAON in-flight fragment facility. Calculations were performed with the computational Monte Carlo codes PHITS and MCNPX. The secondary neutron source terms were evaluated at 550 MeV/u for Ca, Kr, and Sn and at 400 MeV/u for U ions on a graphite target. Source terms and attenuation lengths were obtained by fitting the ambient dose equivalent inside an ordinary concrete shield.

• 한국통신학회논문지
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• 제28권6C호
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• pp.570-579
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• 2003

본 논문에서는 펄스 위치 변조를 사용하는 직접 검파 광통신 채널에서의 터보부호의 성능에 대하여 연구한다. 직접 검파기로서는 이상적인 광자 계수기를 사용하는 경우를 가정하며, 채널 잡음으로는 양자 잡음뿐 만 아니라 배경 잡음도 있는 경우를 가정한다. 즉 사용된 채널 모델은 M-ary PPM Poisson 채널이다. 이 채널에 터보 부호를 적용하기 위한 송수신기의 구조를 제시하고, 또한 제안된 터보 부호화 시스템을 위한 터보 복호 알고리즘을 유도한다. 복호 알고리즘의 유도는 기존의 AWGN 채널을 위한 터보 복호 알고리즘에서 가지 메트릭 연산 부분을 수정함으로써 얻어진다. 이론적인 비트 에러 확률 상한 값의 유도 및 컴퓨터 시뮬레이션의 실행을 통하여, 제안된 터보 부호화 방식의 성능을 분석하고 그 결과를 Reed-Solomon 부호 및 길쌈 부호의 성능과 비교한다.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4776-4797
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• 2022

In the one-dimensional thermal-hydraulic (TH) codes, a subcooled boiling model to predict the void fraction profiles in a vertical channel consists of wall heat flux partitioning, the vapor condensation rate, the bubbly-to-slug flow transition criterion, and drift-flux models. Model performance has been investigated in detail, and necessary refinements have been incorporated into the Safety and Performance Analysis Code (SPACE) developed by the Korean nuclear industry for the safety analysis of pressurized water reactors (PWRs). The necessary refinements to models related to pumping factor, net vapor generation (NVG), vapor condensation, and drift-flux velocity were investigated in this study. In particular, a new NVG empirical correlation was also developed using artificial neural network (ANN) techniques. Simulations of a series of subcooled flow boiling experiments at pressures ranging from 1 to 149.9 bar were performed with the refined SPACE code, and reasonable agreement with the experimental data for the void fraction in the vertical channel was obtained. From the root-mean-square (RMS) error analysis for the predicted void fraction in the subcooled boiling region, the results with the refined SPACE code produce the best predictions for the entire pressure range compared to those using the original SPACE and RELAP5 codes.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.961-970
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• 2012

Often, probabilistic fracture mechanics (PFM) approaches have been adopted to quantify the failure probabilities of Ni-base alloy components, especially due to primary water stress corrosion cracking (PWSCC), in a primary piping system of pressurized water reactors. In this paper, the key features of an advanced PFM code, PINEP-PWSCC (Probabilistic INtegrity Evaluation for nuclear Piping-PWSCC) for such purpose, are described. In developing the code, we adopted most recent research results and advanced models in calculation modules such as PWSCC crack initiation and growth models, a performance-based probability of detection (POD) model for Ni-base alloy welds, and so on. To verify the code, the failure probabilities for various Alloy 182 welds locations were evaluated and compared with field experience and other PFM codes. Finally, the effects of pre-existing crack, weld repair, and POD models on failure probability were evaluated to demonstrate the applicability of PINEP-PWSCC.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.323-334
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• 2013

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1218-1230
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• 2019

A sub-channel solver, named ${\underline{S}}teady$ and ${\underline{T}}ransient$ ${\underline{A}}nalyzer$ for ${\underline{R}}eactor$ ${\underline{T}}hermal$ hydraulics (START), has been developed using the homogenous model for two-phase conditions of light water reactors. The code is developed as a fast and accurate TH-solver for coupled and multi-physics calculations. START has been validated against the NUPEC PWR Sub-channel and Bundle Test (PSBT) database. Tests like single-channel quality and void-fraction for steady state, outlet fluid temperature for steady state, rod-bundle quality and void-fraction for both steady state and transient conditions have been analyzed and compared with experimental values. Results reveal a good accuracy of solution for both steady state and transient scenarios. Axially different values for turbulent mixing coefficient are used based on different grid-spacer types. This provides better results as compared to using a single value of turbulent mixing coefficient. Code-to-code evaluation of PSBT results by the START code compares well with other industrial codes. The START code has been parallelized with the OpenMP algorithm and its numerical performance is evaluated with a large whole PWR core. Scaling study of START shows a good parallel performance.