• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.452-459
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• 2023

The core power control is an important issue for the study of dynamic characteristics in China initiative accelerator driven subcritical system (CiADS), which has direct impact on the control strategy and safety analysis process. The CiADS is an experimental facility that is only controlled by the proton beam intensity without considering the control rods in the current engineering design stage. In order to get the optimized operation scheme with the stable and reliable features, the variation of beam intensity using the continuous and periodic control approaches has been adopted, and the change of collimator and the adjusting of duty ratio have been proposed in the power control process. Considering the neutronics and the thermal-hydraulics characteristics in CiADS, the physical model for the core power control has been established by means of the point reactor kinetics method and the lumped parameter method. Moreover, the multi-inputs single-output (MISO) logical structure for the power control process has been constructed using proportional integral derivative (PID) controller, and the meta-heuristic algorithm has been employed to obtain the global optimized parameters for the stable running mode without producing large perturbations. Finally, the verification and validation of the control method have been tested based on the reference scenarios in considering the disturbances of spallation neutron source and inlet temperature respectively, where all the numerical results reveal that the optimization method has satisfactory performance in the CiADS core power control scenarios.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.20 no.9
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• pp.2894-2900
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• 1996

The stream generator tubes represent an integral part of a major barrier against the fission product release to the environment. So, the rupture of these tubes could permit flow of reactor coolant into the secondary system and injure the safety of reactor coolant system. Therefore, if the crack was detected during In-Service Inspection of tubes the cracked tube should be evaluated by the pulgging criteria and plugged or not. In this study, the fracture mechanics evaluation is carried out on the thru-wall axial crack due to Primary Water Stress Corrosion Cracking in the roll transition aone of steam generator tube to help the assurence the integrity of tubes and estabilish the plugging criteria. Due to the Inconel which is used as tube material is more ductile than others, the plastic instability repture theory was used to calculate the critical and allowable crack length. Based on Leak Before Break concept the leak rate for the critical crack length and the allowable leak rate are compared and the safety of tubes was given.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.10
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• pp.1251-1259
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• 2013

In this study, high temperature design and creep-fatigue damage evaluation of a decay heat exchanger (DHX) in the decay heat removal systems of a sodium-cooled fast reactor (SFR) have been performed. Detail design and 3D finite element analysis have been conducted for the DHXs to be installed in active and passive decay heat removal systems in Korean Generation IV SFR, and the DHX installed in the STELLA-1(Sodium integral effect test loop for safety simulation and assessment) at KAERI (Korea Atomic Energy Research Institute). Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two Mod.9Cr-1Mo steel heat exchangers according to the elevated temperature design codes of ASME Section III Subsection NH and RCC-MR code. Code comparisons were made based on the creep-fatigue damage evaluation and issues on conservatisms of the design codes were discussed.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.110-124
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• 1993

A Nonlinear model-based Hybrid Controller (NHC) is developed which consists of the adaptive proportional-integral-feedforward (PIF) gains and variable structure controller. The controller has the robustness against modeling uncertainty and is applied to the trajectory tracking control of single-input, single-output nonlinear systems. The essence of the scheme is to divide the control into four different terms. Namely, the adaptive P-I-F gains and variable structure controller are used to accomplish the specific control actions by each terms. The robustness of the controller is guaranteed by the feedback of estimated uncertainty and the performance specification given by the adaptation of PIF gains using the second method of Lyapunov. The variable structure controller is incorporated to regulate the initial peak of the tracking error during the parameter adaptation is not settled yet. The newly developed NHC method is applied to the power tracking control of a nuclear reactor and the simulation results show great improvement in tracking performance compared with the conventional model-based control methods.

• 한국전산유체공학회:학술대회논문집
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• 2004.03a
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• pp.49-55
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• 2004

This paper addresses a numerical simulation of the flow and heat transfer in a simplified model of helically coiled tube steam generator using a general purpose computational fluid dynamic analysis computer code. The steam generator model is comprised of a cylindrical shell and helically coiled tubes. A cold feed water entered the tubes is heated up, evaporates. and finally become a superheated steam with a large amount of heat transferred continuously from the hot compressed water at higher pressure flowing counter-currently through the shell side. For the calculation of tube side two-phase flow field formed by boiling, inhomogeneous two-fluid model is used. Both the internal and external turbulent flows are simulated using the standard k-e model. The conjugate heat transfer analysis method is employed to calculate the conduction in the tube wall with finite thickness and the convections in the internal and external fluids simultaneously so as to match the fluid-wall-fluid interface conditions properly. The numerical calculations are peformed for helically coiled tubes of steam generator at an integral type pressurized water reactor under normal operation. The effects of tube-side inlet flow velocity are discussed in details. The results of present numerical simulation are considered to be physically plausible based on the data and knowledge from previous experimental and numerical studies where available.

• Tribology and Lubricants
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• v.21 no.6
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• pp.278-282
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• 2005

Deep groove ball bearing is installed in a control element of an integral nuclear reactor, where water is used as coolant and lubricant. This bearing is made of STS440C stainless steel for the raceways and the balls to use in radioactive environment and water. It is known that the retainer design affects ball bearing operability and endurance life, however there is no verified retainer design and material for water lubricated ball bearing. Four kinds of retainers are manufactured for the endurance test of water lubricated deep groove ball bearing. Three of them are commercially developed types and the other is designed for this research. It is verified that ball bearings with steel pressed and general plastic retainer can not survive to required life in the water, however bearings with machined type and cylinder type retainer can survive. This proves that one of the major design parameters for water lubricated ball bearing is retainer type and material. In this paper, experimental research of endurance test for water-lubricated ball bearing are reported.

• Journal of Institute of Control, Robotics and Systems
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• v.19 no.9
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• pp.839-845
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• 2013

A CSTR (Continuous Stirred Tank Reactor) is a highly nonlinear process with varying parameters during operation. Therefore, tuning of the controller and determining the transition policy of controller parameters are required to guarantee the best performance of the CSTR for overall operating regions. In this paper, a methodology employing the 2DOF (Two-Degree-of-Freedom) PID controller, the anti-windup technique and a fuzzy gain scheduler is presented for the temperature control of the CSTR. First, both a local model and an EA (Evolutionary Algorithm) are used to tune the optimal controller parameters at each operating region by minimizing the IAE (Integral of Absolute Error). Then, a set of controller parameters are expressed as functions of the gain scheduling variable. Those functions are implemented using a set of "if-then" fuzzy rules, which is of Sugeno's form. Simulation works for reference tracking, disturbance rejecting and noise rejecting performances show the feasibility of using the proposed method.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.3
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• pp.381-389
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• 2001

The objective of this paper is to evaluate the fracture resistance characteristics of SA508 Cl.1a to SA508 Cl.3 welds manufactured for the reactor coolant loop piping system of nuclear power plants. The effect of the crack plane orientation to the welding process orientation and the preheat temperature on the fracture resistance characteristics were discussed. Results of the fracture resistance test showed that the effect of the crack plane orientation to the welding process orientation of the fracture toughness is significant, while that of preheat temperature on the fracture toughness is negligible. The micro Vickers hardness test, the metallographic observation and the fractography analysis were conducted to analyse the crack jump phenomenon on the L-R crack plane orientation in the multi-pass welding zone. As these results, it is shown that the crack jump phenomenon was produced because of the inhomogeneity between welding beads and the crack plane orientation must be considered for the safety of the welding zone in the piping system.

• Journal of Institute of Control, Robotics and Systems
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• v.21 no.5
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• pp.482-489
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• 2015

CSTR (Continuous Stirred Tank Reactor) which plays a key role in the chemical plants exhibits highly nonlinear behavior as well as time-varying behavior during operation. The control of CSTRs in the whole operating range has been a challenging problem to control engineers. So, a variety of feedback control forms and their tuning methods have been implemented to guarantee the satisfactory performance. This paper presents a scheme of designing a nonlinear PID controller incorporating with a GA (Genetic Algorithm) for the temperature control of a CSTR. The gains of the NPID controller are composed of easily implementable nonlinear functions based on the error and/or the error rate and its parameters are tuned using a GA by minimizing the ITAE (Integral of Absolute Error). Simulation works for reference tracking and disturbance rejecting performances and robustness to parameter changes show the feasibility of the proposed method.

• Journal of Mechanical Science and Technology
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• v.18 no.12
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• pp.2158-2173
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• 2004

In this paper, the higher order terms in the crack tip stress fields are investigated macroscopically for more realistic assessment of structural material behaviors. For reactor pressure vessel material of A533B ferritic steel, effects of crack size and temperature have been evaluated using 3-point SENB specimens through a series of finite element analyses, tensile tests and fracture toughness tests. The T-stress, Q-parameter and q-parameter as well as the K and J-integral are calculated and mutual relationships are investigated also. Based on the evaluation, it has proven that the effect of crack size from standard length (a/W=0.53) to shallow length (a/W=0.11) is remarkable whilst the effect of temperature from -20$^{\circ}C$ to -60$^{\circ}C$ is negligible. Finally, the cleavage fracture toughness loci as a function of the promising Q-parameter or q-parameter are developed using specific test results as well as finite element analysis results, which can be applicable for structural integrity evaluation considering constraint effects.