• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.71-76
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• 2019

Actual stress distributions in the nozzle of a pressure vessel may not be in plane strain condition, implying that the crack-tip constraint condition may be relaxed in the nozzle. In this paper, a methodology for evaluating the fracture toughness of the ASME Code is presented considering the relaxation of the constraint effect in the nozzle of the reactor pressure vessel. The crack-tip constraint effect is quantified by the T-stress. The equation, which represent the relation between the fracture toughness in the lower constraint condition and the plane strain fracture toughness, is derived using the T-stress. This equation is similar to the method for evaluating the fracture toughness of the Master Curve for low constraint conditions. As a result of evaluating the fracture toughness considering the constraint effect in the reactor inlet, outlet and direct injection nozzles using the proposed equation, it was confirmed that the fracture toughness in the nozzles is higher than the plane strain fracture toughness. Applying the proposed evaluation methodology, it is possible to reflect the relaxation of the constraint effect in the nozzles of the reactor pressure vessel, therefore, the safe operation area on the pressure-temperature limit curve can be prevented from being excessively limited.

• Journal of Powder Materials
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• v.1 no.1
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• pp.21-26
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• 1994

Aluminum was deposited on aluminum oxide powders using a fluidized bed reactor at atmospheric pressure. The aluminum oxide powders were irregular flakes with acute angles and the average particle size was 26 $\mu\textrm{m}$. The fluidized bed was formed by flowing argon gas at the velocity of 60 cm/sec. The optimal fluidization condition was obtained with the reactor designed to be tapered so that the fluid velocity decreases as the fluidizing gas goes up along the reactor. Aluminum was deposited by flowing TiBA(Triisobutylaluminum) evaporated at$250^{\circ}C$ through the fluidized bed reactor heated to 350~$450^{\circ}C$. The result from the analysis by XRD and EDAX confirmed the coating of aluminum and an SEM micrograph showed the conformality.

• Journal of Magnetics
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• v.21 no.4
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• pp.586-592
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• 2016

The magnetic-jack type CRDM withdraws or inserts a control rod assembly from/to the reactor core to control the core reactivity. The CRDM housings form not only the path of the electromagnetic field but also the pressure boundary of a nuclear reactor, and a periodic in-service inspection should be carried out if there are welded or flange jointed parts on the pressure boundary. The in-service inspection is a time-consuming process during the reactor refueling, and moreover it is difficult to perform the inspection over the reactor head. A magnetic motor housing is applied for the current SMART CRDM and has several welding joints, however a nonmagnetic motor housing with fewer or no welding joints may improve the operational efficiency of the nuclear reactor by avoiding or simplifying the in-service inspection process. Prior to the development, the magnetic field transfer efficiency of the nonmagnetic housing was required to be assessed. It was verified and optimized by the electromagnetic analysis of the lifting force estimation. Magnetic flux rings were adopted to improve the efficiency. In this paper, the design and optimization process of a nonmagnetic motor housing with the magnetic flux rings for the SMART CRDM are introduced and the analyses results are discussed.

• Korean Journal of Chemical Engineering
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• v.35 no.11
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• pp.2321-2326
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• 2018

We evaluated the optimal conditions for fluidization of nickel oxide (NiO) and its reduction into high-purity Ni during hydrogen reduction in a laboratory-scale fluidized bed reactor. A comparative study was performed through structural shape analysis using scanning electron microscopy (SEM); variance in pressure drop, minimum fluidization velocity, terminal velocity, reduction rate, and mass loss were assessed at temperatures ranging from 400 to $600^{\circ}C$ and at 20, 40, and 60 min in reaction time. We estimated the sample weight with most active fluidization to be 200 g based on the bed diameter of the fluidized bed reactor and height of the stocked material. The optimal conditions for NiO hydrogen reduction were found to be height of sample H to the internal fluidized bed reactor diameter D was H/D=1, reaction temperature of $550^{\circ}C$, reaction time of 60 min, superficial gas velocity of 0.011 m/s, and pressure drop of 77 Pa during fluidization. We determined the best operating conditions for the NiO hydrogen reduction process based on these findings.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.3
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• pp.505-513
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• 2002

During heatup and cooldown of pressurized water reactor, thermal stress was generated in the reactor pressure vessel (RPV) because of the temperature gradient. To prevent potential failure of RPV, pressure was required to be maintained below the P-T limit curves. In this paper, several methods for constructing the P-T limit curves including the ASME Sec. XI, App. G method were explained and the results were compared. Then, the effects of the various parameters such as flaw size, flaw orientation, cooldown rate, existence of chad, and reference fracture toughness, were evaluated. It was found that the current ASME Sec. XI App. G method resulted in the most conservative P-T limit curve. As the more accurate fracture mechanics analysis results were used, some of the conservatism can be removed. Among the parameters analysed, reference flaw orientation and reference fracture toughness curve had the greatest effect on the resulting P-T limit curves.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.543-550
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• 2012

The decrease in the fracture toughness of ferritic steels in a reactor pressure vessel is an important factor in determining the lifetime of a nuclear power plant. A surveillance program has been in place in Korea since 1979 to assess the structural integrity of RPV steels. In this work, the surveillance data were collected and analyzed statistically in order to derive the empirical relationship between the embrittlement and strengthening of irradiated reactor pressure vessel steels. There was a linear relationship between the yield strength change and the transition temperature shift change at 41 J due to irradiation. The proportional coefficient was about $0.5^{\circ}C$/MPa in the base metals (plate/forgings). The upper shelf energy decrease ratio was non-linearly proportional to the yield strength change, and most of the data lay along the trend curve of the US results. The transition regime temperature interval, ${\Delta}T_T$, was less than the US data. The overall change from irradiation was very similar to the US results. It is expected that the results of this study will be applied to basic research on the multiscale modeling of the irradiation embrittlement of RPV materials in Korea.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.1109-1112
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• 2017

The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1726-1746
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• 2022

This paper presents fracture mechanics analysis results for various cracks located at pressurized water reactor pressure vessel nozzle crotch corners taking into consideration constraint effect. Technical documents such as the ASME B&PV Code, Sec.XI were reviewed and then a fracture mechanics analysis procedure was proposed for structural integrity assessment of various nozzle crotch corner cracks under normal operation conditions considering the constraint effect. Linear elastic fracture mechanics analysis was performed by conducting finite element analysis with the proposed analysis procedure. Based on the evaluation results, elastic-plastic fracture mechanics analysis taking into account the constraint effect was performed only for the axial surface crack of the reactor pressure vessel outlet nozzle with cladding. The fracture mechanics analysis result shows that only the axial surface crack in the reactor pressure vessel outlet nozzle has the stress intensity factor exceeding the low bound of upper-shelf fracture toughness irrespectively of considering the constraint effect. It is confirmed that the J-integral for the axial crack of the outlet nozzle does not exceed the ductile crack initiation toughness. Hence, it can be ensured that the structural integrity of all the cracks is maintained during the normal operation.

• Nuclear Engineering and Technology
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• v.55 no.4
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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.