• Journal of the Korean Institute of Intelligent Systems
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• v.5 no.3
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• pp.52-63
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• 1995

A neuro-fuzzy algorithm is presented for nuclear reactor power control in a pressurized water reactor. Automatic reacotr power control is complicated by the use of control rods because of highly nonlinear dynamics in the axial power shape. Thus, manual shaped controls are usually employed even for the limited capability during the power maneuvers. In an attempt to achieve automatic shape control, a neuro-fuzzy approach is considered because fuzzy algorithms are good at various aspects of operator's knowledge representation while neural networks are efficinet structures capable of learning from experience and adaptation to a changing nuclear core state. In the proposed neuro-fuzzy control scheme, the rule base is formulated based ona multi-input multi-output system and the dynamic back-propagation is used for learning. The neuro-fuzzy powere control algorithm has been tested using simulation fesponses of a Korean standard pressurized water reactor. The results illustrate that the proposed control algorithm would be a parctical strategy for automatic nuclear reactor power control.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.752-759
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• 2017

The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions. The higher yield strength and better corrosion resistance of the nitrogen-alloyed Type 347 stainless steel might be a main cause of slower FCGR and more stable properties against changes in environmental conditions.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.311-322
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• 2012

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.924-933
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• 2015

The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional $^{235}U$ burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using $^{233}U$, $^{242}Pu$, $^{150}Nd$, and $^{133}Cs$ as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.71-81
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• 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.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.719-728
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• 2015

Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2556-2563
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• 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.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.172-177
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• 2021

The microstructure and texture of three 316L foils of 25 ㎛ thickness, which were subjected to different manufacturing process, were systematically characterized using advance analytical techniques. Then, the electrochemical property of the 316L foils in simulated pressurized water reactor (PWR) solution was analyzed using potentiodynamic polarization. The results showed that final rolling strain and annealing temperature had evident effect on grain size, fraction of recrystallization, grain boundary type and texture distribution. It was suggested that large final rolling strain could transfer Brass texture to Copper texture; low annealing temperature could limit the formation of preferable orientations in the rolling process to reduce anisotropy. Potentiodynamic polarization test showed that all samples exhibited good corrosion performance in the simulated primary PWR solution.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.838-848
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• 2015

One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC), which is a robust nonlinear controller, is presented.SMCis ameansto control pressurized water nuclear reactor (PWR) power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.42-46
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• 2010

The evaluation method for the failure frequency of reactor vessel under pressurized thermal shock(PTS) is developed using probabilistic fracture mechanics. The probabilistic reactor integrity evaluation code, named R-PIE code, is developed. The validity and uncertainty of the R-PIE code is investigated. The reactor failure frequencies under PTS for Kori-1 nuclear power plant and other type of domestic nuclear power plants are evaluated. The reference PTS temperature for domestic nuclear power plants is obtained for the rule making against PTS failure.