• Nuclear Engineering and Technology
• /
• 제54권5호
• /
• pp.1825-1834
• /
• 2022

Performing high-fidelity computational fluid dynamics (HF-CFD) to predict the flow and heat transfer state of the coolant in the reactor core is expensive, especially in scenarios that require extensive parameter search, such as uncertainty analysis and design optimization. This work investigated the performance of utilizing a multi-fidelity reduced-order model (MF-ROM) in PWR rod bundles simulation. Firstly, basis vectors and basis vector coefficients of high-fidelity and low-fidelity CFD results are extracted separately by the proper orthogonal decomposition (POD) approach. Secondly, a surrogate model is trained to map the relationship between the extracted coefficients from different fidelity results. In the prediction stage, the coefficients of the low-fidelity data under the new operating conditions are extracted by using the obtained POD basis vectors. Then, the trained surrogate model uses the low-fidelity coefficients to regress the high-fidelity coefficients. The predicted high-fidelity data is reconstructed from the product of extracted basis vectors and the regression coefficients. The effectiveness of the MF-ROM is evaluated on a flow and heat transfer problem in PWR fuel rod bundles. Two data-driven algorithms, the Kriging and artificial neural network (ANN), are trained as surrogate models for the MF-ROM to reconstruct the complex flow and heat transfer field downstream of the mixing vanes. The results show good agreements between the data reconstructed with the trained MF-ROM and the high-fidelity CFD simulation result, while the former only requires to taken the computational burden of low-fidelity simulation. The results also show that the performance of the ANN model is slightly better than the Kriging model when using a high number of POD basis vectors for regression. Moreover, the result presented in this paper demonstrates the suitability of the proposed MF-ROM for high-fidelity fixed value initialization to accelerate complex simulation.

• Nuclear Engineering and Technology
• /
• 제39권5호
• /
• pp.627-636
• /
• 2007

The present work is aimed at further discussing the effectiveness of dimensionless parameters recently proposed for the analysis of flow stability in heated channels with supercritical fluids. In this purpose, after presenting the main motivations for the introduction of these parameters in place of previously proposed ones, additional information on the theoretical bases and on the consequences of this development is provided. Stability maps, generated by an in-house program adapted from a previous application to boiling channels, are also shown for different combinations of the operating parameters. The maps are obtained as contour plots of an amplification parameter obtained from numerical discretization and subsequent linearization of governing equations; as such, they provide a quantitatively clear perspective of the effect of different boundary conditions on the stability of heated channels with supercritical fluids. In order to assess the validity of the assumptions at the basis of the in-house model, supporting calculations have been performed making use of the RELAP5/MOD3.3 computer code, detecting the values of the dimensionless parameters at the threshold for the occurrence of instability for a heated channel representative of SCWR proposed core configurations. The obtained results show reasonable agreement with the maps, supporting the applicability of the proposed scaling parameters for describing the dynamic behaviour of heated channels with supercritical fluids.

• Nuclear Engineering and Technology
• /
• 제53권12호
• /
• pp.3879-3891
• /
• 2021

A heat pipe residual heat removal system is proposed to be incorporated into the reactor driven subcritical (RDS) facility, which has been proposed by MIT Nuclear Reactor Laboratory for testing and demonstrating the Fluoride-salt-cooled High-temperature Reactor (FHR). It aims to reduce the risk of the system operation after the shutdown of the facility. One of the main components of the system is an air-cooled heat pipe heat exchanger. The alkali-metal high-temperature heat pipe was designed to meet the operation temperature and residual heat removal requirement of the facility. The heat pipe model developed in the previous work was adopted to simulate the designed heat pipe and assess the heat transport capability. 3D numerical simulation of the subcritical facility active zone was performed by the commercial CFD software STAR CCM + to investigate the operation characteristics of this proposed system. The thermal resistance network of the heat pipe was built and incorporated into the CFD model. The nominal condition, partial loss of air flow accident and partial heat pipe failure accident were simulated and analyzed. The results show that the residual heat removal system can provide sufficient cooling of the subcritical facility with a remarkable safety margin. The heat pipe can work under the recommended operation temperature range and the heat flux is below all thermal limits. The facility peak temperature is also lower than the safety limits.

• Nuclear Engineering and Technology
• /
• 제42권6호
• /
• pp.620-635
• /
• 2010

In this paper we describe current activities on the project Multi-Scale Modeling and Analysis of convective boiling (MSMA), conducted jointly by the Paul Scherrer Institute (PSI) and the Swiss Nuclear Utilities (Swissnuclear). The long-term aim of the MSMA project is to formulate improved closure laws for Computational Fluid Dynamics (CFD) simulations for prediction of convective boiling and eventually of the Critical Heat Flux (CHF). As boiling is controlled by the competition of numerous phenomena at various length and time scales, a multi-scale approach is employed to tackle the problem at different scales. In the MSMA project, the scales on which we focus range from the CFD scale (macro-scale), bubble size scale (meso-scale), liquid micro-layer and triple interline scale (micro-scale), and molecular scale (nano-scale). The current focus of the project is on micro- and meso-scales modeling. The numerical framework comprises a highly efficient, parallel DNS solver, the PSI-BOIL code. The code has incorporated an Immersed Boundary Method (IBM) to tackle complex geometries. For simulation of meso-scales (bubbles), we use the Constrained Interpolation Profile method: Conservative Semi-Lagrangian $2^{nd}$ order (CIP-CSL2). The phase change is described either by applying conventional jump conditions at the interface, or by using the Phase Field (PF) approach. In this work, we present selected results for flows in complex geometry using the IBM, selected bubbly flow simulations using the CIP-CSL2 method and results for phase change using the PF approach. In the subsequent stage of the project, the importance of effects of nano-scale processes on the global boiling heat transfer will be evaluated. To validate the models, more experimental information will be needed in the future, so it is expected that the MSMA project will become the seed for a long-term, combined theoretical and experimental program.

• 한국산학기술학회논문지
• /
• 제13권5호
• /
• pp.1969-1975
• /
• 2012

대형 상용 차량에 사용되는 에어 브레이크 시스템은 제동시 상대적으로 응답 시간이 길고 압력 손실이 크게 발생한다. 제동 시간은 파이프 시스템의 적절한 설계에 의해서 최소화 시킬 수 있다. 제동 시간은 시스템 압력의 증가, 에어 라인의 감소 및 재질의 변화에 대해 감소시킬 수 있다. 본 연구에서는 상용 트럭에 대한 리그를 구성하여 에어 브레이크의 특성에 대해서 고찰하였다. 해석 프로그램에 대한 신뢰성을 확보하기 위하여 실험 값과 비교하였고, 브레이크 시스템에 영향을 미치는 시스템 압력, 파이프의 재질 및 직경, 탱크와 챔버에서 압력에 대한 온도 영향을 검토하였다.

• Nuclear Engineering and Technology
• /
• 제52권10호
• /
• pp.2173-2182
• /
• 2020

Control valves are widely used to regulate fluid flux in nuclear power plants, and there are more than 1500 control valves in the primary circuit of one nuclear power plant. With their help, the flux can be regulated to a specific level of water or steam to guarantee the energy efficiency and safety of the nuclear power plant. The flux characteristics of the control valve mainly depend on the valve core shape. In order to analyze the effects of valve core shapes on flux characteristics of control valves, this paper focuses on the valve core shapes. To begin with, numerical models of different valve core shapes are established, and results are compared with the ideal flux characteristics curve for the purpose of validation. Meanwhile, the flow fields corresponding to different valve core shapes are investigated. Moreover, relationships between the valve core opening and the outlet flux under different valve core shapes are carried out. The flux characteristics curve and equation are proposed to predict the outlet flux under different valve core openings. This work can benefit the further research of the flux control and the optimization of the valve core for control valves in nuclear power plants.

• 인터넷정보학회논문지
• /
• 제10권3호
• /
• pp.113-129
• /
• 2009

기업의 업무 활동은 수많은 프로세스로 이루어져 있고 기업들이 업무를 처리하는 모든 활동 속에서 수많은 비즈니스 프로세스가 동시다발적으로 생성되고 처리된다. 전사적 업무환경은 업무의 처리를 간편하게 만든 동시에 더욱 복합적인 업무 형태를 만들어내었고 이에 따라 비즈니스 프로세스의 효율적 관리가 더욱 절실하게 되었다. 이에 기업 및 조직은 프로세스에 대한 개선과 새로운 비즈니스의 발견을 목표로 프로세스를 분석하고 비즈니스와 관련한 지식 도출을 위한 연구 개발에 집중하게 되었다. 프로세스를 분석하고 개선하기 위한 연구들로써 프로세스 마이닝, 프로세스 재발견, BPR(Business Process Reengineering) 등의 개념들이 등장하였고 실제 적용을 위한 연구들이 활발히 진행 중에 있다. 하지만 이러한 연구들은 대부분 프로세스 인스턴스들의 이력 정보에 대한 데이터웨어하우징을 통한 접근법을 따르고 있다. 이러한 접근법은 점점 복잡화되는 최근의 기업 환경과 수시로 변화하는 사용자의 요구를 충족시키기에 부족하다. 분석적인 측면을 고려하여 설계된 프로세스 인스턴스들은 효과적인 그룹화가 가능하고, 분석 도메인 내에서 사용자의 분석 요구가 변화하게 되는 경우에도 분석 비용을 줄일 수 있다. 본 연구에서는 기존 프로세스 모델링 방법론인 ICN을 확장하여 효과적인 프로세스 분석과 비즈니스 관련 지식 도출을 목표로 색채형 워크플로우 모델을 정의한다. 프로세스 인스턴스들에 대한 실행 정보를 통해 수동적인 프로세스 분석이 아닌, 프로세스 정의 시점부터 비즈니스 지식 도출을 목표로 하여 적극적이고 사용자 중심적인 프로세스 분석이 가능하게 한다.