• Nuclear Engineering and Technology
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• 제20권2호
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• pp.80-87
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• 1988

소형냉각재상실사고(SBLOCA)와 대형냉각재상실사고(LBLOCA)중에 노심의 Average Channel과 Hot Channel에서의 유량분포를 예측하였다. 아울러 REALP5/MOD2 코드를 사용하여 두 채널사이의 교차류고려여부가 실제사고 분석결과에 미치는 영향을 평가하였다. 현재까지 SBLOCA계산에서는 노심을 한개의 채널로 모델하는 것이 충분하다고 판단되어 왔으나 본 계산결과에 의하면 보수적인 계산을 위해서는 Hot Channel 모델링이 필요한 것으로 밝혀졌다. 그러나 LBLOCA Blowdown Phase존에서는 교차류의 고려 여부에 상관없이 Hot Channel 이 Average Channel보다 보수적 인 결과를 가져오며, 교차류의 영향도 미세한 것으로 판명되었다.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.358-369
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• 2002

Abundant information about analyses of single steam generator tube rupture (SGTR) events is available because of its importance in terms of safety. However, there are few literatures available on analyses of multiple steam generator tube rupture (MSGTR) events. In addition, knowledge of transients and consequences following a MSGTR event are very limited as there has been no occurrence of MSGTR event in the commercial operation of nuclear reactors. In this study, a postulated MSGTR event in an APR1400 is analyzed using thermal-hydraulic system code MARSI.4. The present study aims to examine the effects of affected steam generator selection. The main steam safety valve (MSSV) lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generate. (S/G) is affected. The comparison shows that the cases where two steam generators are simultaneously affected allow longer time for operator action compared with the cases where a single steam generator is affected. Furthermore, the tube ruptures in the steam generator where a pressurizer is connected leads to the shortest operator response time.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3449-3459
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• 2021

With the increasing availability of high-performance computing (HPC) platforms, uncertainty quantification (UQ) and sensitivity analyses (SA) can be efficiently leveraged to optimize design parameters of complex engineering problems using modeling and simulation tools. The workflow involved in such studies heavily relies on HPC resources and hence requires pre-processing and post-processing capabilities of large amounts of data along with remote submission capabilities. The NEAMS Workbench addresses all aspects of the workflows involved in these studies by relying on a user-friendly graphical user interface and a python application program interface. This paper highlights the NEAMS Workbench capabilities by presenting a semiautomated coupling scheme between Dakota and any given package integrated with the NEAMS Workbench, yielding a simplified workflow for users. This new capability is demonstrated by running a SA of a turbulent flow in a pipe using the open-source Nek5000 CFD code. A total of 54 jobs were run on a HPC platform using the remote capabilities of the NEAMS Workbench. The results demonstrate that the semiautomated coupling scheme involving Dakota can be efficiently used for UQ and SA while keeping scripting tasks to a minimum for users. All input and output files used in this work are available in https://code.ornl.gov/neams-workbench/dakota-nek5000-study.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1251-1260
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• 2019

The water-cooled WCLL blanket is one of the possible candidates for the blanket of the fusion power reactors. The plasma-facing first wall manufactured from the reduced-activation ferritic-martensitic steel Eurofer97 will be cooled with water at a typical pressurized water reactor (PWR) conditions. According to new estimates, the first wall will be exposed to peak heat fluxes up to $7MW/m^2$ while the maximum operated temperature of Eurofer97 is set to $550^{\circ}C$. The performed analysis shows the capability of the designed flat first wall concept to remove heat flux without exceeding the maximum Eurofer97 operating temperature only up to $0.75MW/m^2$. Several heat transfer enhancement methods (turbulator promoters), structural modifications, and variations of parameters were analysed. The effects of particular modifications on the wall temperature were evaluated using thermo-hydraulic three-dimensional numerical simulation. The analysis shows the negligible effect of the turbulators. By the combination of the proposed modifications, the permitted heat flux was increased up to $1.69MW/m^2$ only. The results indicate the necessity of the re-evaluation of the existing first wall concepts.

• Nuclear Engineering and Technology
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• 제26권4호
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• pp.562-569
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• 1994

이 연구에서는 피동형원자로의 과도현상을 분석하기 위한 KOTRAC 코드의 모델을 수정한 것이다. 이 코드에서 열수력학 모델로 도입하고 있는 mixture drift flux model은 피동형원자로와 같이 비상냉각수가 중력으로 주입되는 경우를 잘 모사할 수 있으나, 만일 가압기 밀림관 또는 수평관에서 상의 완전분리가 일어나게 될 때에는 증기상에서의 거의 영에 가까운 밀도로 인해 상당한 어려움이 존재하는 것이 밝혀졌다. 이 연구에서는 이러한 어려움을 극복하기 위해 일부 모델을 개선하였는데 가장 두드러진 것은 KOTRAC에서 사용하고 있는 flow distribution parameter를 Ishii 상관식으로 대체하여 코드를 수정하고 해석하였다. 이렇게 수정된 코드를 사용한 결과는 과도상태 해석코드인 RELAP5 /MOD3 계산결과와 비교적 잘 일치함을 볼 수 있었다.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.330-338
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• 2023

In this study, impacts of an enhanced-moderation Fuel Assembly (FA) named Truly Optimized PWR (TOP) lattice, which is modified based on the standard 17 × 17 PWR FA, are investigated in a natural circulation Soluble-Boron-Free (SBF) Small Modular Reactor (SMR). Two different TOP lattice designs are considered for the analysis; one is with 1.26 cm pin pitch and 0.38 cm fuel pellet radius, and the other is with 1.40 cm pin pitch and 0.41 cm fuel pellet radius. The NuScale core design is utilized as the base model and assumed to be successfully converted to an SBF core. The analysis is performed following the primary coolant circulation loop, and the reactor is modelled as a single channel for thermal-hydraulic analyses. It is assumed that the ratio of the core pressure drop to the total system pressure drop is around 0.3. The results showed that the reactor power could be increased by 2.5% and 9.8% utilizing 1.26/0.38 cm and 1.40/0.41 cm TOP designs, respectively, under the identical coolant inlet and outlet temperatures as the constraints.

• Ocean Systems Engineering
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• 제12권3호
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• pp.335-358
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• 2022

Hydrodynamic field alteration around a cylindrical pier using a curved vane is numerically investigated. The curved vane with various angles ranged from 10 to 220 degree is placed at the upstream of the cylindrical pier. Laminar flow is adopted in order to perform the steady-state analysis. It is found that the flow separation leads to the formation of four bubbles depending on the value of the curved vane angle. Two bubbles are located in the region between the rear of the curved vane and the leading surface of the cylindrical pier, while the remaining two bubbles are located at the wake zone behind the cylindrical pier. Numerical analysis is performed to reveal the hydrodynamic field and influence of curved vane on the formation and evolution of the bubbles. It is found that the center and size of the bubble depend mainly on the value of the curved vane angle. It is observed that the flow velocity vector shows clearly the alteration in the flow velocity direction especially at the leading surface and rear surface of the curved vane owing to the occurrence of flow separation and flow dissipation along the circumference of the vane.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1002-1012
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• 2024

To remove insoluble fission products, which could possibly cause reactor instability and significantly reduce heat transfer efficiency from primary system of molten salt reactor, a helium bubbling method is employed into a passive molten salt fast reactor. In this regard, two-phase flow behavior of molten salt and helium bubbles was investigated experimentally because the helium bubbles highly affect the circulation performance of working fluid owing to an additional drag force. As the helium flow rate is controlled, the change of key thermal-hydraulic parameters was analyzed through a two-phase experiment. Simultaneously, to assess the applicability of numerical model for the analysis of two-phase flow behavior, the numerical calculation was performed using the OpenFOAM 9.0 code. The accuracy of the numerical analysis code was evaluated by comparing it with the experimental data. Generally, numerical results showed a good agreement with the experiment. However, at the high helium injection rates, the prediction capability for void fraction of helium bubbles was relatively low. This study suggests that the multiphaseEulerFoam solver in OpenFOAM code is effective for predicting the helium bubbling but there exists a room for further improvement by incorporating the appropriate drag flux model and the population balance equation.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1698-1711
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• 2024

Heat pipe cooled reactors have gained attention as a potential solution for nuclear power generation in space and deep sea applications because of their simple design, scalability, safety and reliability. However, under complex operating conditions, a control strategy for variable load operation is necessary. This paper presents a two-dimensional transient characteristics analysis program for a heat pipe cooled reactor and proposes a variable load control strategy using the recuperator bypass (CSURB). The program was verified against previous studies, and steady-state and step-load operating conditions were calculated. For normal operating condition, the predicted temperature distribution with constant heat pipe temperature boundary conditions agrees well with the literature, with a maximum temperature difference of 0.4 K. With the implementation of the control strategy using the recuperator bypass (CSURB) proposed in this paper, it becomes feasible to achieve variable load operation and return the system to a steady state solely through the self-regulation of the reactor, without the need to operate the control drum. The average temperature difference of the fuel does not exceed 1 % at the four power levels of 70 %,80 %, 90 % and 100 % Full power. The output power of the turbine can match the load change process, and the temperature difference between the inlet and outlet of the turbine increases as the power decreases.