• Nuclear Engineering and Technology
• /
• v.19 no.3
• /
• pp.198-204
• /
• 1987

A plant simulation code, MCSIM (Micro-Computer SIMulator), has been developed to simulate plant transient accidents for pressurized water reactors. Reactor coolant system is modeled using decoupled energy and momentum equations, drift flux two-phase flow model and integral momentum equation. A two-fluid pressurizer model is used to simulate the pressurizer dynamics. Pot Boiler model is used for steam generator, steady-state decoupled energy and momentum equations for secondary side system, and point kinetics equations for nuclear power calculation. For test of the present version of MCSIM, complete loss of flow and RCCA withdrawal accidents are calculated with MCSIM. The results are compared with those in FSAR of KNU 5 & 6.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.97-106
• /
• 2018

Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the important local power peaking components in nuclear reactors is axial power peaking, which continuously changes. The main challenge of nuclear reactor control during load-following operation is to maintain the AO within acceptable limits, at a certain reference target value. This article proposes a new robust approach to AO control of pressurized water reactors during load-following operation. This method uses robust feedback-linearization control based on the multipoint kinetics reactor model (neutronic and thermal-hydraulic). In this model, the reactor core is divided into four nodes along the reactor axis. Simulation results show that this method improves the reactor load-following capability in the presence of parameter uncertainty and disturbances and can use optimum control rod groups to maneuver with variable overlapping.

• Nuclear Engineering and Technology
• /
• v.37 no.1
• /
• pp.1-10
• /
• 2005

An overview of the reactor physics aspects of Generation Four(GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and ecoomics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.191-191
• /
• 1999

• Nuclear Engineering and Technology
• /
• v.48 no.2
• /
• pp.330-338
• /
• 2016

The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.121-122
• /
• 2005

• Nuclear Engineering and Technology
• /
• v.44 no.2
• /
• pp.161-176
• /
• 2012

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

• Biotechnology and Bioprocess Engineering:BBE
• /
• v.9 no.2
• /
• pp.86-92
• /
• 2004

In this paper miniaturized disposable micro/nanofluidic components applicable to bio chip, chemical analyzer and biomedical monitoring system, such as blood analysis, micro dosing system and cell experiment, etc are reported. This system includes various microfluidic components including a micropump, micromixer, DNA purification chip and single-cell assay chip. For low voltage and low power operation, a surface tension-driven micropump is presented, as well as a micromixer, which was implemented using MEMS technology, for efficient liquid mixing is also introduced. As bio-reactors, DNA purification and single-cell assay devices, for the extraction of pure DNA from liquid mixture or blood and for cellular engineering or high-throughput screening, respectively, are presented.

• Nuclear Engineering and Technology
• /
• v.43 no.1
• /
• pp.45-62
• /
• 2011

The Korean nuclear industry is developing a thermal-hydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE code adopts advanced physical modeling of two-phase flows, mainly two-fluid three-field models which comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or nonstructured meshes. The programming language for the SPACE code is C++ for object-oriented code architecture. The SPACE code will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWRs and the design of advanced reactors. This paper describes the overall features of the SPACE code and shows the code assessment results for several conceptual and separate effect test problems.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.475-480
• /
• 1997

In order to conduct the effective integration of computer-aided conceptual design for integrated nuclear power reactor, not only is a smooth information flow required, but also decision making fur both conceptual design and construction process design must be synthesized. In addition to the aboves, the relations between the one step and another step and the methodologies to optimize the decision variables are verified, in this paper especially, that is, scaling laws and scaling criteria. In the respect with the running of the system, the integrated optimization process is proposed in which decisions concerning both conceptual design are simultaneously made. According to the proposed reactor types and power levels, an integrated optimization problems are formulated. This optimization is expressed as a multi-objective optimization problem. The algorithm for solving the problem is also presented. The proposed method is applied to designing a integrated sub-critical reactors.