• Proceedings of the KIEE Conference
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• 2005.10b
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• pp.104-106
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• 2005

In this work, a PWR reactor core dynamics is identified online by a recursive least squares method. Based on this 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 design an automatic controller for thermal power control in PWRs. 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, the procedure to solve the optimization problem is then repeated. The objectives of the proposed model predictive controller are to minimize both the difference between the predicted core coolant temperature and the desired one, and the variation of the control rod positions. Also, the objectives are subject to maximum and minimum control rod positions and maximum control rod speed. Therefore, the genetic algorithm that is appropriate to accomplish multiple objectives is used to optimize the model predictive controller. A 3-dimensional nuclear reactor analysis code, MASTER that was developed by Korea Atomic Energy Research Institute (KAERI), is used to verify the proposed controller for a nuclear reactor. From results of numerical simulation to check the performance of the proposed controller at the 5%/min ramp increase or decrease of a desired load and its 10% step increase or decrease which are 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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• v.50 no.7
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• pp.1173-1183
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• 2018

Transient fluid velocity and pressure fields in a pressurized water reactor (PWR) steam generator (SG) secondary side during the blowdown period of a feedwater line break (FWLB) accident were numerically simulated employing the saturated water flashing model. This model is based on the assumption that compressed water in the SG is saturated at the beginning and decompresses into the two-phase region where saturated vapor forms, creating a mixture of steam bubbles in water by bulk boiling. The numerical calculations were performed for two cases of which the outflow boundary conditions are different from each other; one is specified as the direct blowdown discharge to the atmosphere and the other is specified as the blowdown discharge to an extended calculation domain with atmospheric pressure on its boundary. The present simulation results obtained using the two different outflow boundary conditions were discussed through a comparison with the predictions using a simple non-flashing model neglecting the effects of phase change. In addition, the applicability of each of the non-flashing water discharge and saturated water flashing models for the confirmatory assessments of new SG designs was examined.

• Nuclear Engineering and Technology
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• v.20 no.4
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• pp.253-264
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• 1988

Our experience of neutron noise analysis in French-type 900 MWe pressurized water reactor (PWR) is presented. Neutron noise analysis is based on the technique of interpreting the signal fluctuations of ex-core detectors caused by core reactivity changes and neutron attenuation due to lateral core motion. It also provides advantages over deterministic dynamic-testing techniques because existing plant instrumentation can be utilized and normal operation of the plant is not disturbed. The data of this paper were obtained in the ULJIN unit 1 reactor during the start-up test period and the statistical descriptors, useful for our purpose, are power spectral density (PSD), coherence function (CF), and phase difference between detectors. It is found that core support barrel (CSB) motions induced by coolant flow forces and pressure pulsations in a reactor vessel were indentified around 8 Hz of frequency.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.24 no.2
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• pp.211-222
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• 2011

In this paper a structural analysis of the PWR(pressurized water reactor) canister with 102cm diameter is carried out to evaluate the structural safety of the canister for the impact force occurring at the moment of collision with the ground in the falling plumb down accident from the carriage vehicle which may happen during the canister handling at the spent nuclear fuel disposal repository. For this, a rigid body dynamic analysis of the canister is executed to compute the impact force using the commercial CAE system, RecurDyn, and a nonlinear structural analysis is performed to compute stresses and deformations occurring inside the canister for this computed impact force using the commercial FEM code, NISA. From these analysis results, the structural safety of the canister is evaluated for the falling plumb down accident from the carriage vehicle due to the inattention during the canister handling at the repository. The rigid body dynamic analysis performed assuming the canister as a rigid body shows that the canister falls plumb down to the ground in two types. And also it shows that early collision impact force is the biggest one and following impact forces decrease gradually. The height of the carriage vehicle in the repository is assumed as 5m in order to obtain the stable structural safety evaluation result. The nonlinear structural analysis of the canister is executed for the biggest early impact force. The structural analysis result of the canister shows that the structural safety of the PWR canister is not secured for the falling plumb down accident from the moving carriage vehicle because the maximum stresses occurring in the cast iron insert of canister are bigger than the yield stress of the cast iron.

• Corrosion Science and Technology
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• v.20 no.4
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• pp.189-195
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• 2021

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor's internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2458-2473
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• 2024

The turbulent flow in reactor pressure vessel (RPV) of pressurized water reactor (PWR) is important for the flow rate distribution at core inlet. Thus, it is vital to study the turbulent flow phenomena in RPV. However, the complicated fluid channel consisted of inner structures of RPV will block or refract the laser sheet of particle image velocimetry (PIV). In this work, the matched index of refraction (MIR) of sodium iodide (NaI) solution and acrylic was applied to support optical path for flow field measurements by PIV in the 1/10th scaled-down RPV model. The experimental results show detailed velocity field at different locations inside the scaled-down RPV model. Some interesting phenomena are obtained, including the non-negligible counterflow at the corner of nozzle edge, the high downward flowing stream in downcomer, large vortices above vortex suppression plate in lower plenum. And the intensity of counterflow and the strength of vortices increase as inlet flow rate increasing. Finally, the case of asymmetry flow was also studied. The turbulent flow has different pattern compared with the case of symmetrical inlet flow rate, which may affect the uniformity of flow distribution at the core inlet.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.12
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• pp.654-662
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• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.05a
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• pp.304-309
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• 2002

Fretting wear test in room temperature water was performed to evaluate the wear coefficient of Inconel 600,690 (Pressurized Water Reactor, PWR) and Alloy 800 (CANadian DeuteriumUranium, CANDU) steam generator (SG) tubes against ferritic and martensitic stainless steels. The main focus is to compare the wear behaviors between Alloy 800 and Inconel alloys. Test conditions are $10{\sim}30N$ of normal load, $200{\sim}450{\mu}m$ of sliding amplitude and 30Hz of frequency. The result indicated that the wear rate of Alloy 800 was higher than those of Inconel 690 at various test condition such as normal loads, sliding amplitudes etc. From the results of SEM observation, there was little evidence of plastic deformation layer that were dominantly formed on the worn surfaces of Inconel 690. Also, wear particles in Alloy 800 were released from contacting asperities deformed by severe plastic flow during fretting wear. Main cause of wear rate between Alloy 800 and Inconel 690 may be due to the difference of hardness between martensitic and ferritic stainless steel. The wear rate and wear mechanism of two tubes in room temperature water are discussed.

• Corrosion Science and Technology
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• v.11 no.4
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• pp.141-150
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• 2012

Stress corrosion cracks in Alloy 600 compact tension specimens tested at $325^{\circ}C$ in a simulated primary water environment of pressurized water reactor were analyzed by analytical transmission electron microscopy and secondary ion mass spectroscopy (SIMS). From a fine-probe chemical analysis, oxygen was found on the grain boundary just ahead of the crack tip, and chromium oxides were precipitated on the crack tip and the grain boundary attacked by the oxygen diffusion, leaving a Cr/Fe depletion (or Ni enrichment) zone. The oxide layer inside the crack was revealed to consist of a double (inner and outer) layer. Chromium oxides existed in the inner layer, with NiO and (Ni,Cr) spinels in the outer layer. From the nano-SIMS analysis, oxygen was detected at the locations of intergranular chromium carbides ahead of the crack tip, which means that oxygen diffused into the grain boundary and oxidized the surfaces of the chromium carbides. The intergranular chromium carbide blunted the crack tip, thereby suppressing the crack propagation.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.795-801
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• 1998

Metallic fatigue of Pressurized Water Reactor(PWR) materials is a generic safety issue for commercial nuclear power plants. It is very important to obtain the fatigue usage factor for component integrity and life extension. In this paper, fatigue usage was obtained at the inside surface of Kori unit 2, 3 and 4 RCP casing weld, based on the design transient. And it was intended to establish the procedure and the detailed method of fatigue evaluation in accordance with ASME Section III Code. According to this code rule, two methods to determine the stress cycle and the number of cycles could be applied. One method is the superposition of cycles of various design transients and the other is based on the assumption that a stress cycle correspond to only one design transient. Both method showed almost same fatigue usage in the RCP casing weld.