• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2720-2727
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• 2022

The Generation IV Lead-cooled fast reactor (LFR) take the liquid lead or lead-bismuth eutectic alloy (LBE) as the coolant of the primary cooling circuit. Combined with the natural characteristics of lead alloy and the design features of LFR, the system is the simplest and the number of equipment is the least, which reflects the inherent safety characteristics of LFR. The nuclear main coolant pump (MCP) is the only power component and the only rotating component in the primary circuit of the reactor, so the various operating characteristics of the MCP are directly related to the safety of the nuclear reactor. In this paper, various working conditions that may occur in the normal rotation (positive rotating) of the MCP and the corresponding internal flow characteristics are analyzed and studied, including the normal pump condition, the positive-flow braking condition and the negative-flow braking condition. Since the corrosiveness of LBE is proportional to the fluid velocity, the distribution of flow velocity in the pump channel will be the focus of this study. It is found that under the normal pump condition and positive-flow braking conditions, the high velocity region of the impeller domain appears at the inlet and outlet of the blade. At the same radius, the pressure surface is lower than the back surface, and with the increase of flow rate, the flow separation phenomenon is obvious, and the turbulent kinetic energy distribution in impeller and diffuser domain shows obvious near-wall property. Under the negative-flow braking condition, there is obvious flow separation in the impeller channel.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.373-379
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• 2017

Seismic probabilistic safety assessments are used to help understand the impact potential seismic events can have on the operation of a nuclear power plant. An important component to seismic probabilistic safety assessment is the seismic hazard curve which shows the frequency of seismic events. However, these hazard curves are estimated assuming a normal distribution of the seismic events. This may not be a strong assumption given the number of recorded events at each source-to-site distance. The use of a normal distribution makes the calculations significantly easier but may underestimate or overestimate the more rare events, which is of concern to nuclear power plants. This paper shows a preliminary exploration into the effect of using a distribution that perhaps more represents the distribution of events, such as the t-distribution to describe data. The integration of a probability distribution with potentially larger tails basically pushes the hazard curves outward, suggesting a different range of frequencies for use in seismic probabilistic safety assessments. Therefore the use of a more realistic distribution results in an increase in the frequency calculations suggesting rare events are less rare than thought in terms of seismic probabilistic safety assessment. However, the opposite was observed with the ground motion prediction equation considered.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.110-115
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• 2014

The nuclear power plant was launched by Kori unit 1 in 1978 years. Currently, 23 nuclear power plants have been operating in Korea since 1978 years. The localization was completed for most of the reactor facility from Hanbit(Youngkwang) unit 3&4. However, RCP(Reactor Coolant Pump) and MMIS(Man Machine Interface System) is an important technology that has been excluded from the scope of the technical transfer has been dependent on a specific overseas vendor. Recent success in RCP development through co-operation with government and industries. Developed RCP will be applied to Shin-Hanul unit 1&2 nuclear power plants. The RCP operates in high speed and high pressure condition and only rotating component in the NSSS(Nuclear Steam Supply System). Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can influence on the stability characteristics for entire RCS(Reactor Coolant System) loop, and can act as significant destabilizing forces. In this study, vibration evaluation of the pump shaft of development RCP estimated under normal operation and over speed conditions. In order to predict the vibration characteristics and dynamic behavior, modal analysis, critical speed analysis and unbalance response spectrum analysis were performed.

• Nuclear Engineering and Technology
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• v.37 no.2
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• pp.139-150
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• 2005

A robust control design procedure for a nuclear reactor has been developed and experimentally validated on the Penn State TRIGA research reactor. The utilization of the robust controller as a component of an autonomous control system is also demonstrated. Two methods of specifying a low order (fourth-order) nominal-plant model for a robust control design were evaluated: 1) by approximation based on the 'physics' of the process and 2) by an optimal Hankel approximation of a higher order plant model. The uncertainty between the nominal plant models and the higher order plant model is supplied as a specification to the ,u-synthesis robust control design procedure. Two methods of quantifying uncertainty were evaluated: 1) a combination of additive and multiplicative uncertainty and 2) multiplicative uncertainty alone. The conclusions are that the optimal Hankel approximation and a combination of additive and multiplicative uncertainty are the best approach to design robust control for this application. The results from nonlinear simulation testing and the physical experiments are consistent and thus help to confirm the correctness of the robust control design procedures and conclusions.

• Nuclear Engineering and Technology
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• v.36 no.5
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• pp.388-402
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• 2004

A phenomena identification and ranking table(PIRT) was developed for a main steam line break (MSLB) event for the Advanced Power Reactor-1400 (APR-1400). The selectee event was a double-ended steam line break at full power, with the reactor coolant pump running. The developmental panel selected the fuel performance as the primary safety criterion during the ranking process. The plant design data, the results of the APR-1400 safety analysis, and the results of an additional best-estimate analysis by the MARS computer code were used in the development of the PIRT. The period of the transient was composed of three phases: pre-trip, rapid cool-down, and safety injection. Based on the relative importance to the primary evaluation criterion, the ranking of each system, component, and phenomenon/process was performed for each time phase. Finally, the knowledge-level for each important process for certain components was ranked in terms of existing knowledge. The PIRT can be used as a guide for planning cost-effective experimental programs and for code development efforts, especially for the quantification of those processes and/or phenomena that are highly important, but not well understood.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.442-448
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• 2022

Intermediate-level radioactive waste (ILW) is not subject to legal approval for cave disposal in Korea. To solve this problem, a spent resin treatment device that separates ¹⁴C-containing resin from zeolite/activated carbon and desorbs ¹⁴C through a microwave device has been developed. In this study, we evaluated the radiological safety of the operators performing repair work in the event of a failure in such a device treating 1 ton of spent resin mixture per day. Based on the safety evaluation results, it is possible to formulate a design plan that can ensure the safety of workers while developing a commercialized device. When each component of the resin treatment device can be repaired from the outside, the maximum and minimum allowable repair times are calculated as 263.2 h and 27.7 h for the ¹⁴C-detached resin storage tank and zeolite/activated carbon storage tank, respectively. For at least 6 h per quarter, the worker's annual dose limit remains within 50 mSv/year; further, over 5 years, it remained within 100 mSv. At least 6 h of repair time per quarter is considered, under conservative conditions, to verify the radiological safety of the worker during repair work within that time.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.1085-1097
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• 2022

A printed circuit steam generator (PCSG) is being considered as the component for pressurized water reactor (PWR) type small modular reactor (SMR) that can further reduce the physical size of the system. Since a steam generator in many PWR-type SMR generates superheated steam, it is expected that dryout front oscillation can potentially cause thermal fatigue failure due to cyclic thermal stresses induced by the transition in boiling regimes between convective evaporation and film boiling. To investigate the fatigue issue of a PCSG, a reference PCSG is designed in this study first using an in-house PCSG design tool. For the stress analysis, a finite element method analysis model is developed to obtain the temperature and stress fields of the designed PCSG. Fatigue estimation is performed based on ASME Boiler and pressure vessel code to identify the major parameters influencing the fatigue life time originating from the dryout front oscillation. As a result of this study, the limit on the temperature difference between the hot side and cold side fluids is obtained. Moreover, it is found that the heat transfer coefficient of convective evaporation and film boiling regimes play an essential role in the fatigue life cycle as well as the temperature difference.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2488-2498
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• 2021

The wall film-wise condensation plays an important role in the heat transfer processes of heat exchangers, refrigerators, and air conditioner. In the field of nuclear engineering, steam condensation is often utilized in safety systems to remove the core decay heat under both transient and accident conditions. In particular, passive containment cooling system (PCCS), are designed to ensure containment safety under severe accident conditions. A computational fluid dynamics (CFD) scale analysis has been conducted to calculate the heat transfer rate of the PCCS. However, despite the increase in computing power, there are challenges in the long-term transient simulation of containment using CFD scale codes. In this study, a heat structure coupling between the CFD and system analysis codes was performed to efficiently analyze PCCS. In addition, the component unstructured program for interfacial dynamics (CUPID) was improved to analyze the condensation behavior of ternary gas mixtures. Thereafter, the condensation heat transfer on the primary side was calculated using the improved CUPID and CFD code, whereas that on the secondary side was simulated using MARS. Both the coupled codes were validated against the CONAN facility database. Finally, conjugate heat transfer simulations with wall condensation in the presence of non-condensable gases were appropriately performed.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.707-714
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• 2024

Different degrees of natural radioactivity found in quartz can have negative consequences on health. Quartz vein along the investigated Abu Ramad area, Egypt, had its natural radioactivity assessed. The HPGe spectrometer was used to determine the role played by the radionuclides ²³⁸U, ²³²Th, and ⁴⁰K in the gamma radiation that was emitted, and the results showed that these concentrations are 484.64 ± 288.4, 36.8 ± 13.1 and 772.2 ± 134.6 Bq kg^-1 were higher than the corresponding reported global limits of 33, 45, and 412 Bq kg^-1 for each radionuclide (²³⁸U, ²³²Th, and ⁴⁰K). Among the radiological hazard parameters, the excess lifetime cancer risk (ELCR) is estimated and it's mean value of ELCR (1.2) is higher than the permissible limit of 0.00029. The relationship between the radionuclides and the associated radiological hazard characteristics was investigated based on multivariate statistical methods including Pearson correlation, principal component analysis (PCA), and hierarchical cluster analysis (HCA). According to statistical research, the radioactive risk of quartz is primarily caused by the ²³⁸U, ²³²Thand ⁴⁰K. Finally, applying quartz to building materials would pose a significant risk to the public.

• Analytical Science and Technology
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• v.17 no.5
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• pp.416-422
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• 2004

Time-resolved laser induced breakdown spectroscopy (TRELIBS) has been developed and applied to the qualitative analysis of the component materials of nuclear power plant. The alloy samples used in this work were carbon steels (A106 Gr. B; A336 P11; A335 P22), stainless steels (type 304; type 316) and inconel alloys (Inconel 600; Inconel 690; Inconel 800). Carbon steels can be individually distinguished by the intensity ratio of chromium to iron and molybdenum to iron emission lines observed at the wavelength raging from 485 to 575 nm. Type 316 stainless steel can be easily differentiated from type 304 by identification of the molybdenum emission lines at an emission wavelength ranging from 485 to 575 nm: type 304 does not give any molybdenum emission lines, but type 316 does. The inconel alloys can be individually distinguished by the intensity ratio of Cr/Fe and Ni/Fe emission lines at the wavelength raging from 420 to 510 nm. TRELIBS has been proved to be a powerful analytical technique for direct analysis of alloys due to its non-destructivity and simplicity.