• Nuclear Engineering and Technology
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• 제37권5호
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• pp.469-478
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• 2005

The subchannel analysis capability of MARS, a multi-dimensional thermal-hydraulic system code, has been enhanced. In particular, the turbulent mixing and void drift models for the flow-mixing phenomena in rod bundles were improved. Then, the subchannel analysis feature was combined with the existing coupled system thermal-hydraulics (T/H) and 3D reactor kinetics calculation capability of MARS. These features allow for more realistic simulations of both the hot channel behavior and the global system T/H behavior. Using the coupled features of MARS, a coupled analysis of a main steam line break (MSLB) is carried out for demonstration purposes. The results of the calculations are very reasonable and promising.

• Nuclear Engineering and Technology
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• 제46권5호
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• pp.655-666
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• 2014

The CUPID code has been developed at KAERI for a transient, three-dimensional analysis of a two-phase flow in light water nuclear reactor components. It can provide both a component-scale and a CFD-scale simulation by using a porous media or an open media model for a two-phase flow. In this paper, recent advances in the CUPID code are presented in three sections. First, the domain decomposition parallel method implemented in the CUPID code is described with the parallel efficiency test for multiple processors. Then, the coupling of CUPID-MARS via heat structure is introduced, where CUPID has been coupled with a system-scale thermal-hydraulics code, MARS, through the heat structure. The coupled code has been applied to a multi-scale thermal-hydraulic analysis of a pool mixing test. Finally, CUPID-SG is developed for analyzing two-phase flows in PWR steam generators. Physical models and validation results of CUPID-SG are discussed.

• 한국통신학회논문지
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• 제38A권10호
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• pp.827-832
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• 2013

홀로그래픽 데이터 저장장치에서는 인접 심볼간 간섭이 2차원으로 발생하며, 인접 페이지간 간섭 또한 발생한다. 특히 멀티 레벨 홀로그래픽 데이터 저장장치의 경우, 한 픽셀이 나타내는 정보가 0과 1의 이진수가 아닌 그 이상의 정보를 더 저장하고 있기 때문에, 위와 같은 간섭들이 더 크게 발생할 수 있다. 본 논문에서는 4-레벨 홀로그래픽 데이터 저장장치에서의 성능 저하를 보완하기 위하여 2/3 변조 부호를 제안한다. 제안된 2/3 변조 부호는 비터비 검출 방법을 변조 방식에 적용하여 에러 정정 능력을 갖는다. 또한, 본 논문에서는 2/3 변조 부호를 위한 새로운 비터비 디코더를 제안한다. 제안된 비터비 검출기는 복호 시에 필요없는 상태에 대한 계산을 제거하여 복호 성능을 높인다. 제안된 비터비 검출기는 2/3 변조 부호에 대하여 기존의 비터비 검출기보다 더 뛰어난 성능을 보인다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.122-122
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.125.2-125.2
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• 2004

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.19-33
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• 2024

In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphon-breaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as / increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.

• 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.

• 전자공학회논문지
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• 제53권9호
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• pp.12-17
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• 2016

홀로그래픽 데이터 저장 장치(HDSS)는 페이지 단위로 저장 매체의 체적에 데이터를 저장 및 판독하고 2차원으로 데이터를 처리하기 때문에 데이터 전송 속도 및 저장 용량이 증가한다. 게다가, 멀티레벨 HDSS는 한 픽셀에 한 비트이상을 저장할 수 있다. 하지만 2차원으로 페이지를 처리하므로 기존의 데이터 저장 시스템과 달리 2차원으로 인접한 심볼 간 간섭(ISI) 및 인접 페이지 간 간섭(IPI)가 발생한다. 기존에 발표된 논문들은 멀티레벨 HDSS 환경에서 2차원 ISI 완화에 관한 연구에 초점을 두었지만 멀티레벨 HDSS 환경에서 2차원 ISI와 IPI를 동시에 완화하는 연구는 진행되지 않았다. 본 논문에서는 2차원 ISI 및 IPI를 동시에 완화하는 4-레벨 균형 변조부호를 제안하였다.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.981-988
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• 2018

An experiment was performed for the OECD/NEA ROSA-2 Project with the large-scale test facility (LSTF), which simulated a steam generator tube rupture (SGTR) accident due to a double-ended guillotine break of one of steam generator (SG) U-tubes with operator recovery actions in a pressurized water reactor. The relief valve of broken SG opened three times after the start of intact SG secondary-side depressurization as the recovery action. Multi-dimensional phenomena specific to the SGTR accident appeared such as significant thermal stratification in a cold leg in broken loop especially during the operation of high-pressure injection (HPI) system. The RELAP5/MOD3.3 code overpredicted the broken SG secondary-side pressure after the start of the intact SG secondary-side depressurization, and failed to calculate the cold leg fluid temperature in broken loop. The combination of the number of the ruptured SG tubes and the HPI system operation difference was found to significantly affect the primary and SG secondary-side pressures through sensitivity analyses with the RELAP5 code.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.980-992
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• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.