• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1417-1428
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• 2020

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MW_th. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MW_th.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4048-4056
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• 2023

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.54-59
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• 2019

The main objective of nuclear reactor safety is to maintain the nuclear fuel in a thermally safe condition with enough safety margins during normal operation and anticipated operational occurrences. In this research, core flow bypass is studied under the conditions of the unavailability of safety systems. As core bypass occurs, the core flow rate is assumed to decrease exponentially with a time constant of 25 s to new steady state values of 20, 40, 60, and 80% of the nominal core flow rate. The thermal hydraulic code PARET is used through these calculations. Reactor thermal hydraulic stability is reported for all cases of core flow bypass.

• Nuclear Engineering and Technology
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• 제16권3호
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• pp.155-160
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• 1984

표준 핵연료집합체나 최적 핵연료집합체로 구성된 원자력 1호기 원자로심의 열적여유도를 기존 열설계 방법과 통계적 열설계 방법을 이용하여 분석하였다. 통계적 열설계 방법은 노심내 운전변수들의 불확실도를 통계적으로 처리함으로써 기존 방법에 비하여 열적여유도를 증가시킨다. 계산을 위하여 정상상태와 과도시 열수력분석 전산코드인 COBRA-IV-i를 사용하였다. 계산결과 통계적 설계방법은 열적여유도를 크게 증가시키며, 표준 핵 연료집 합체는 물론 최적 핵 연료집 합체가 장전된 원자력 1호기의 열설계기준을 만족시키는 것으로 밝혀졌다. 그러나 기존 열설계 방법은 원자력 1호기 노심에 최적 핵연료집합체가 장전된 경우 열설계기준을 만족시키지 못하는 것으로 밝혀졌다.