• Nuclear Engineering and Technology
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• 제38권1호
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• pp.81-92
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• 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
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• 제51권6호
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• pp.1540-1553
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• 2019

A nuclear power plant (NPP) is a highly complex system-of-systems as manifested through its internal systems interdependence. The negative impact of such interdependence was demonstrated through the 2011 Fukushima Daiichi nuclear disaster. As such, there is a critical need for new strategies to overcome the limitations of current risk assessment techniques (e.g. the use of static event and fault tree schemes), particularly through simulation of the nonlinear dynamic feedback mechanisms between the different NPP systems/components. As the first and key step towards developing an integrated NPP dynamic probabilistic risk assessment platform that can account for such feedback mechanisms, the current study adopts a system dynamics simulation approach to model the thermal dynamic processes in: the reactor core; the secondary coolant system; and the pressurized water reactor. The reactor core and secondary coolant system parameters used to develop system dynamics models are based on those of the Palo Verde Nuclear Generating Station. These three system dynamics models are subsequently validated, using results from published work, under different system perturbations including the change in reactivity, the steam valve coefficient, the primary coolant flow, and others. Moving forward, the developed system dynamics models can be integrated with other interacting processes within a NPP to form the basis of a dynamic system-level (systemic) risk assessment tool.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3517-3527
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• 2021

New 80-MW (electric) ultra-long-life sodium cooled fast reactor core having inherent safety characteristics is designed with heterogeneous fuel assemblies comprised of driver and blanket fuel rods. Several options using upper sodium plenum and SSFZ (Special Sodium Flowing Zone) for reducing sodium void reactivity are neutronically analyzed in this core concept in order to improve the inherent safety of the core. The SSFZ allowing the coolant flow from the peripheral fuel assemblies increases the neutron leakage under coolant expansion or voiding. The Monte Carlo calculations were used to design the cores and analyze their physics characteristics with heterogeneous models. The results of the design and analyses show that the final core design option has a small burnup reactivity swing of 618 pcm over ~54 EFPYs cycle length and a very small sodium void worth of ~35pcm at EOC (End of Cycle), which leads to the satisfaction of all the conditions for inherent safety with large margin based on the quasi-static reactivity balance analysis under ATWS (Anticipated Transient Without Scram).

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.404-410
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• 1998

In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold leg temperature was very close to the technical specification limit of 298$^{\circ}C$ during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2540-2549
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• 2022

For a pressurized water reactor power plant, the reactor coolant pump (RCP) is a kernel component. And for a canned motor RCP, the rotor system's properties determines its safety. The liquid coolant inside the canned motor RCP fills clearance between the metal shields of rotor and stator, forming a lengthy clearance flow. The influence of inlet preswirl on rotordynamic coefficients of clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally investigated in this work. A quasi-steady state computational fluid dynamics (CFD) method has been used to investigate the influence of inlet preswirl. A vertical experiment rig has also been established for this purpose. Rotordynamic coefficients on different inlet preswirl ratios (IR) are obtained through CFD and experiment. Results show that the cross-coupled stiffness of the clearance flow would change significantly with inlet preswirl, but other rotordynamic coefficients would not change significantly with inlet preswirl. For the case of clearance flow between the stator and rotor cans, influence of inlet preswirl is not so significant as the IR is not large enough.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1310-1323
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• 2023

A reactor coolant pump (RCP) is essential for transporting coolant in the primary loop of pressurized water reactors. In the advanced passive reactor, the absence of a long pipeline between the steam generator and RCP serves as a transition section, resulting in a non-uniform flow field at the pump inlet. Therefore, the characteristics of the pump should be investigated under non-uniform flow to determine its influence on the pump. In this study, the pressure pulsation characteristics were examined in the time and frequency domains, and the sources of low-frequency and high-amplitude signals were analyzed using wavelet coherence analysis and numerical simulation. From computational fluid dynamics (CFD) results, non-uniform inflow has a great effect on the flow structures in the pump's inlet. The pressure pulsation in the pump at the rated flow increased by 78-128.7% under the non-uniform inflow condition in comparison with that observed under the uniform inflow condition. Furthermore, a low-frequency signal with a high amplitude was observed, whose energy increased significantly under non-uniform flow. The wavelet coherence and CFD analysis verified that the source of this signal was the low-frequency pulsating vortex under the steam generator.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 춘계학술발표회논문집(2)
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• pp.833-838
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• 1995

This study presents a method to predict plant-specific and operational history dependent scaling factors. Realistic and detailed approaches are taken to find scaling factors at reactor coolant. This approach begins with fission product release mechanisms and fundamental release properties of fuel-source nuclide such as fission product and transuranic nuclide. Scaling factors at various waste streams are derived from the predicted reactor coolant scaling factors with the use of radionuclide retention and build up model. This model makes use of radioactive material balance within the radioactive waste processing systems. According to input parameters of plant operation history, scaling factors predicted at reactor coolant and waste streams are well brought out the effects of plant operation history.

• Corrosion Science and Technology
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• 제10권5호
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• pp.167-174
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• 2011

Due to commercialization of nuclear power, most countries have taken interest in decontamination process of nuclear power plant and tried to develop a optimum process. Because open literature of the decontamination process are rare, it is hard to obtain skills on decontamination of foreign country and it is necessarily to develop proper chemical decontamination process system in Korea. In this study, applicable possibility in chemical decontamination for reactor coolant pump (RCP) was investigated for the various stainless steels. The stainless steel (STS) 304 showed the best electrochemical properties for corrosion resistance and the lowest weight loss ratio in chemical decontamination process with immersion type than other materials. However, the pitting corrosion was generated in both STS 415 and STS 431 with the increasing numbers of cycle. The intergranular corrosion in STS 431 was sporadically observed. The sizes of their pitting corrosion also increased with increasing cycle numbers.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1140-1153
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• 2016

Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.466-475
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• 2017

Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal-hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.