• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1289-1295
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• 2012

A process heat exchanger (PHE) in a nuclear hydrogen system is a key component for transferring the considerable heat generated in a very high temperature reactor (VHTR) to a chemical reaction that yields a large quantity of hydrogen. A performance test on a small-scale PHE prototype made of Hastelloy-X is underway in a small-scale gas loop at the Korea Atomic Energy Research Institute. Previous research on the high-temperature structural analysis of the small-scale PHE prototype had been performed using base material properties. In this study, a high-temperature structural analysis considering the mechanical properties in the weld zone was performed, and the obtained results were compared with those of the previous research.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.107-111
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• 2011

The release of a site and building from regulatory control is the final stage of the decommissioning process. The MARSSIM (Multi-Agency Radiation Survey and Site Investigation Manual) provides overall framework for conducting data collection for a final status survey to demonstrate compliance with site closure requirements. The KAERI carried out establishing a final status survey by using the guidance provided in the MARSSIM for of a site and building of the Korea Research Reactor. The release criteria for a site and building were set up based on these results of the site specific release levels which were calculated by using RESRAD and RESRAD-Build codes. The survey design for a site and building was classified by using the survey dataset and potential contamination. The number of samples in each survey unit was calculated by through a statistical test using the collected data from a scoping and characterization survey. The results of the final status survey were satisfied the release criteria based on an evaluation of the measured data.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.335-343
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• 2005

The Methodology of burnup calculation with EPMA test set up in this study. The spent fuel from PWR nuclear power plant was used as specimen. This $UO_2$ fuel with $3.2\%$ of enrichment had been irradiated up to 35,000 MWd/MTU(reference data). The burnup is very important factor for nuclear fuel to estimate all fuel behaviors in reactor. To measure amounts of fission products and actinides for the burnup calcualation, chemical analysis (destructive method) has been used but it mattes long experimental time and second radio-wastes. In this study, EPMA test was available to measure amount of fission products. Neodymium is able to be detected and quantified. It can be compared with the results from chemical analysis and ORIGEN-2 code calculation. Concentration of Nd from EPMA test showed good agreement with result of ORIGEN-2 code in the same burnup.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.484-492
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• 1994

Microstructures of Nb$_2$O$_{5}$-doped UO$_2$ pellets have been investigated during sintering under H$_2$ and $CO_2$ atmospheres. Pellets are sintered at 1$700^{\circ}C$ in H$_2$ atmosphere and at 130$0^{\circ}C$ in $CO_2$ atmosphere for 1 to 41 hr. The addison of Nb$_2$O$_{5}$ causes the formation of large pores, which shrink to some extent in H$_2$ atmosphere but very little in $CO_2$. Fine pores in the Nb$_2$O$_{5}$-doped UO$_2$ pellet are almost annihilated when sintered under H$_2$ atmosphere but little changed under $CO_2$ atmosphere. The increase in grain size due to Nb$_2$O$_{5}$ addition is much larger in H$_2$ atmosphere than in $CO_2$. Thus the enhancement of uranium diffusion in UO$_2$ due to the Nb$_2$O$_{5}$ addition is thought to be more significant in H$_2$ atmosphere. Microstructures of Nb$_2$O$_{5}$-doped UO$_2$ pellets sintered in H$_2$ atmosphere are discussed from the viewpoint of in-reactor performance. Possible defects formation due to Nb$_2$O$_{5}$ addition is discussed to explain the enhancement of uranium diffusion in H$_2$ and $CO_2$ atmospheres.> atmospheres.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.113-123
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• 2005

The removal of radioactive corrosion products from the reactor coolant through a magnetic filter system is one of the many approaches being investigated as a means to reduce radiation sources and exposures to the operational and maintenance personnel in a nuclear power plant. Many research activities in water chemistry, therefore, have been performed to provide a filtration system with high reliability and feasibility and are still in process. In this study, it was devised the magnetic filter system with permanent and electric magnets to remove the corrosion products in the coolant stream taking an advantage of the magnetic properties of corrosion particles. Permanent magnets were used for separation of corrosion products and electric magnets were utilized for flocculation of colloidal particles to increase in their size. Experiments using only permanent magnets, in the previous study, displayed the satisfactory outcome of filtering corrosion products and indicated that the removal efficiency was more than 90 $\%$ for above 5 $\mu$m particles. Experiments using electric magnets also showed the good performance of flocculation without chemical agents and exhibited that most corrosion particles were flocculated into larger aggregates about 5 $\mu$m and over in diameter. It is, thus, expected that the magnetic filter system with the arrangement of permanent and electric magnets will be an effective way for the removal of radioactive corrosion products with considerably high removal efficiency.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1115-1130
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• 2012

In nuclear power plants, the reactor pressure vessel (RPV) upper head control rod drive mechanism (CRDM) penetration nozzles are fabricated using J-groove weld geometry. Recently, the incidences of cracking in Alloy 600 CRDM nozzles and their associated welds have increased significantly. The cracking mechanism has been attributed to primary water stress corrosion cracking (PWSCC), and it has been shown to be driven by welding residual stresses and operational stresses in the weld region. The weld-induced residual stress is the main factor contributing to crack growth. Therefore, an exact estimation of the residual stress is important for ensuring reliable operation. This study presents the residual stress computation performed for an RPV CRDM penetration nozzle in Korea. Based on two and three dimensional finite element analyses, the effect of welding variables on the residual stress variation is estimated for sensitivity analysis.

• Nuclear Engineering and Technology
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• v.18 no.4
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• pp.273-280
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• 1986

The purpose of this study is to obtain the experimental data of the forced flow dryout heat flux in a heat generating debris bed which simulates the degraded nuclear reactor core after severe accident. An experimental investigation has been conducted of dryout heat flux in an inductively heated bed of steel particles with upward forced flow rising coolant circulation system under atmospheric pressure. The present observations were mainly focused on the effects of coolant mass flux, particle size, bed height, and coolant subcooling on the dryout heat flux The data were obtained when carbon steel particles in the size distribution 1.5, 2.5, 3.0 and 4.0 mm were placed in a 55 mm ID Pyrex glass column and inductively heated by passing radio frequency current through a multiturn work coil encircling the column. Distilled water was supplied with variation of mass flux from 0 to 3.5 kg/$\textrm{cm}^2$ s as a coolant in the tests, while the bed height was selected as 55 mm and 110 mm. Inlet temperature of coolant varied by 2$0^{\circ}C$ and 8$0^{\circ}C$. The principal results of the tests are: (1) Dryout heat flux increases with increase of upward forcing mass flux and particle size; (2) The dryout heat flux at the zero mass flux obviously depends on the Particle size as Previous studies; (3) The forced flow dryout heat flux in the shallow bed is somewhat higher than that in the deep bed,

• Analytical Science and Technology
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• v.14 no.4
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• pp.1075-1081
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• 2001

Instrumental Neutron Activation Analysis as a representative method of nuclear analytical technique, has advantages of non-destructive, simultaneous multi-element analysis with the characteristics of absolute measurement method. Up to date, $k_0$-quantitative method which is accurate, convenient and user-friendly, has been generalized world-widely. In this study, it is intented to introduce the general concept of $k_0$-method and to measure $k_0$-parameters for the future implementation to our NAA system. For this objectives, the definition of relevant factors for the quantitative analysis and the equation for the experimental determination of parameters such as $Q_0$(${\alpha}$) and f were summarized. Furthermore, a foundation for the $k_0$-standardization method was prepared through the measurement of ${\alpha}$ and f-value which depend on the specific character of irradiation hole at NAA#1-hole of HANARO research reactor.

• Journal of the Korean Society of Visualization
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• v.15 no.3
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• pp.20-26
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• 2017

The two-fluid equations are widely used to simulate two-phase flows in a nuclear reactor. For the two-fluid momentum equation, the wall and interfacial drag terms play an important role in predicting a two-phase flow behavior. Since the bubble density is much smaller than the water density, the bubble accelerates faster than the liquid in a nozzle. As a result, the bubble phase becomes faster than the liquid phase in the nozzle. In contrast, the opposite phenomena occur in the diffuser. The purpose of our study is to experimentally show these behaviors in an area-varying channel such as nozzle and diffuser. Experiments were made of turbulent bubbly flows in an area-varying horizontal channel. The velocities of the bubble and liquid phases were measured by the PIV technique. It was shown that the two-phase velocities were no longer close to each other in the area-varying regions. The bubble was faster than the liquid in the nozzle; in contrast, the bubble was slower than the liquid in the diffuser. Code simulations were also performed using the MARS code. By replacing the original wall drag model in the MARS code with Kim (1)'s wall drag partition model, we obtained the simulation results being consistent with experimental observations.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.1
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• pp.1-5
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• 2010

Acoustic emission(AE) has emerged as a powerful nondestructive tool to detect or monitor preexisting defects and leaks in the vessel structures. A Bragg grating based acoustic emission sensor system is developed. Various type of fiber Bragg grating sensor including the variable length of sensing part was fabricated and prototype sensor system was tested by using PZT pulser and pencil lead break sources. Two types of sensor attachment were used. First, the fiber Bragg grating sensor was attached fully to the surface using bonding agent. Second one is that one part of fiber was attached to the surface partly by bonding and the other part of fiber will be act as a cantilever. That is, the resonant frequency of the fiber Bragg grating sensor will depend on the length of sensing part. The final goal of the sensor system is to provide on-line monitoring of cracks or leaks in reactor vessel head penetration of nuclear power plants.