• Nuclear Engineering and Technology
• /
• 제49권1호
• /
• pp.71-81
• /
• 2017

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

• 한국에너지공학회:학술대회논문집
• /
• 한국에너지공학회 1994년도 춘계학술발표회 초록집
• /
• pp.130-137
• /
• 1994

As a part of developing engineering simulator for CANDU 6 nuclear power plants, present paper gives the tentative simulation results of reactor and turbine power transients including reactor-follow-turbine operation. One point kinetics equations are used for neutron dynamics, iodine and xenon loads. To calculate time-dependent high and low pressure turbine powers and grid frequency deviation, simple first order differential equations are used. In addition, control logics (reactor regulating system, demand power routine, and unit power regulator) used in the plant's process computers have been referenced.

• 대한전기학회논문지:시스템및제어부문D
• /
• 제54권9호
• /
• pp.567-569
• /
• 2005

NARX(Nonlinear AutoRegressive with eXogenous input) neural network was used for prediction of nuclear reactor behavior which was influenced by control rods in short-term period and also by the concentration of xenon and boron in long-term period in load following operations. The developed model was designed to predict reactor power, xenon worth and axial offset with different burnup states when control rods and boron were adjusted in load following operations. Data of the Korea Next Generation PWR were collected by ONED94 code. The test results presented exhibit the capability of the NARX neural network model to capture the long term and short term dynamics of the reactor core and the developed model seems to be utilized as a handy tool for the use of a plant simulation.

• Nuclear Engineering and Technology
• /
• 제41권10호
• /
• pp.1275-1284
• /
• 2009

A directly heated $SO_3$ decomposer for the sulfur-iodine and hybrid-sulfur processes has been introduced and analyzed using the computational fluid dynamics (CFD) code CFX 11. The use of a directly heated decomposition reactor in conjunction with a very high temperature reactor (VHTR) allows for higher decomposition reactor operating temperatures. However, the high temperatures and strongly corrosive operating conditions associated with $SO_3$ decomposition present challenges for the structural materials of decomposition reactors. In order to resolve these problems, we have designed a directly heated $SO_3$ decomposer using RA330 alloy as a structural material and have performed a CFD analysis of the design based on the finite rate chemistry model. The CFD results show the maximum temperature of the structural material could be maintained sufficiently below 1073 K, which is considered the target temperature for RA 330. The CFD simulations also indicated good performance in terms of $SO_3$ decomposition for the design parameters of the present study.

• Nuclear Engineering and Technology
• /
• 제53권8호
• /
• pp.2556-2563
• /
• 2021

Metaheuristic algorithms can work well in solving or optimizing problems, especially those that require approximation or do not have a good analytical solution. Particle swarm optimization (PSO) is one of these algorithms. The response quality of these algorithms depends on the objective function and its regulated parameters. The nonlinear nature of the pressurized light-water nuclear reactor (PWR) dynamics is a significant target for PSO. The two-point kinetics model of this type of reactor is used because of fission products properties. The proportional-integral-derivative (PID) controller is intended to control the power level of the PWR at a short-time transient. The absolute error (IAE), integral of square error (ISE), integral of time-absolute error (ITAE), and integral of time-square error (ITSE) objective functions have been used as performance indexes to tune the PID gains with PSO. The optimization results with each of them are evaluated with the number of function evaluations (NFE). All performance indexes achieve good results with differences in the rate of over/under-shoot or convergence rate of the cost function, in the desired time domain.

• Journal of Mechanical Science and Technology
• /
• 제18권6호
• /
• pp.990-1001
• /
• 2004

The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a Pressurized Water Reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure.

• Nuclear Engineering and Technology
• /
• 제47권5호
• /
• pp.534-545
• /
• 2015

In the very high temperature reactor (VHTR), which is a next generation nuclear reactor system, ceramics are used as a fuel coating material and graphite is used as a core structural material. Even if a depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change only slowly. This is because the thermal capacity of the core is so high. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel. The objectives of this study are to investigate the heat transfer characteristics of natural convection of a one-side heated vertical channel with inserted porous materials of high porosity and also to develop the passive cooling system for the VHTR. An experiment was carried out using a one-side heated vertical rectangular channel. To obtain the heat transfer and fluid flow characteristics of the vertical channel with inserted porous material, we have also carried out a numerical analysis using a commercial Computational Fluid Dynamics (CFD) code. This paper describes the thermal performances of the one-side heated vertical rectangular channel with an inserted copper wire of high porosity.

• Nuclear Engineering and Technology
• /
• 제33권4호
• /
• pp.397-408
• /
• 2001

A marine reactor shows very different thermal-hydraulic characteristics compared to a land- based reactor. Especially, study on the variation of flow field due to ship motions such as inclination, heaving and rolling is essential since the flow variation has great influence on the reactor cooling capability. In this study, the natural circulation characteristics of integral type marine reactor with modular steam generators were analyzed using computational fluid dynamics code, CFX-4, for inclined conditions. The numerical analyses are performed using the results of natural circulation experiments for integral reactor which are already conducted at Seoul National University. From the results, it was found that the flow rate in the ascending steam generator cassettes increases due to buoyancy effect. Due to this flow variation, temperature difference occurs at the outlets of the each steam generator cassettes. which is mitigated through downcomer by thermal mixing. Also, around the upper pressure header the flow from descending hot leg goes up to the ascending steam generator cassettes due to large natural circulation driving force in ascending steam generator cassettes. From this result, the increase of How rate in the ascending steam generator cassettes could be understood qualitatively.

• Nuclear Engineering and Technology
• /
• 제38권1호
• /
• pp.81-92
• /
• 2006

In this study, the core dynamics of a PWR reactor is identified online by a recursive least-squares method. Based on the identified reactor model consisting of the control rod position and the core average coolant temperature, the future average coolant temperature is predicted. A model predictive control method is applied to designing an automatic controller for the thermal power control of PWR reactors. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, this procedure for solving the optimization problem is repeated. The objectives of the proposed model predictive controller are to minimize both the difference between the predicted core coolant temperature and the desired temperature, as well as minimizing the variation of the control rod positions. In addition, the objectives are subject to the maximum and minimum control rod positions as well as the maximum control rod speed. Therefore, a genetic algorithm that is appropriate for the accomplishment of multiple objectives is utilized in order to optimize the model predictive controller. A three-dimensional nuclear reactor analysis code, MASTER that was developed by the Korea Atomic Energy Research Institute (KAERI) , is used to verify the proposed controller for a nuclear reactor. From the results of a numerical simulation that was carried out in order to verify the performance of the proposed controller with a $5\%/min$ ramp increase or decrease of a desired load and a $10\%$ step increase or decrease (which were design requirements), it was found that the nuclear power level controlled by the proposed controller could track the desired power level very well.

• Nuclear Engineering and Technology
• /
• 제41권10호
• /
• pp.1293-1306
• /
• 2009

In this paper, an infinite impulse response locally recurrent neural network (IIR-LRNN) is employed for modelling the dynamics of the Lead Bismuth Eutectic eXperimental Accelerator Driven System (LBE-XADS). The network is trained by recursive back-propagation (RBP) and its ability in estimating transients is tested under various conditions. The results demonstrate the robustness of the locally recurrent scheme in the reconstruction of complex nonlinear dynamic relationships.