• Nuclear Engineering and Technology
• /
• v.51 no.7
• /
• pp.1729-1737
• /
• 2019

This paper presents neutronics design of VVER-1000 fuel assembly using burnable poison particles (BPPs) for controlling excess reactivity and pin-wise power distribution. The advantage of using BPPs is that the thermal conductivity of BPP-dispersed fuel pin could be improved. Numerical calculations have been conducted for optimizing the BPP parameters using the MVP code and the JENDL-3.3 data library. The results show that by using $Gd_2O_3$ particles with the diameter of $60{\mu}m$ and the packing fraction of 5%, the burnup reactivity curve and pin-wise power distribution are obtained approximately that of the reference design. To minimize power peaking factor (PPF), total BP amount has been distributed in a larger number of fuel rods. Optimization has been conducted for the number of BPP-dispersed rods, their distribution, BPP diameter and packing fraction. Two models of assembly consisting of 18 BPP-dispersed rods have been selected. The diameter of $300{\mu}m$ and the packing fraction of 3.33% were determined so that the burnup reactivity curve is approximate that of the reference one, while the PPF can be decreased from 1.167 to 1.105 and 1.113, respectively. Application of BPPs for compensating the reduction of soluble boron content to 50% and 0% is also investigated.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.21 no.5
• /
• pp.561-564
• /
• 2001

The cladding area of HANARO Research Reactor fuel rods should be checked not to have any defects larger than the size required at QA documents by using eddy torrent testing method doting fabrication process. To apply eddy current testing inspection to the fuel rods, encircling differential probes and standard specimen were designed and fabricated. The impedance of the fabricated probes was measured with impedance analyzer in order to cheek that the probe has a suitable impedance for the inspection frequency, and with this probe and MIZ-40A eddy current equipment, the detectability of this probes was investigated. The developed probes could detect artificial notch with 2mm length 10% depth of cladding thickness in cladding area. In addition, the probe was successfully applied to detect the defects in cladding area doting fabrication of the research reactor rods.

• Nuclear Engineering and Technology
• /
• v.35 no.5
• /
• pp.378-386
• /
• 2003

Recently failures of nuclear fuel rods in Korean nuclear power plants were reported and their failure causes have been investigated by using PIE techniques. Destructive and physico-chemical examinations reveal that the clad hydriding phenomena had caused the rod failures primarily and secondarily in each case. In this study, the basic mechanisms of the primary and the secondary hydriding failures are reviewed, PIE data such as cladding inner and outer surface oxide thickness and the restructuring of the fuel pellets are analyzed, and they are compared with the predicted behaviors by a fuel performance code. In addition, post-defected fuel behaviors are reviewed and qualitatively analyzed. The results strongly support that the hydriding processes, primary and secondary, played critical roles in the respective fuel rods failures and the secondary hydriding failure can take place even in the fuel rod with low linear heat generation rate.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 1998.04a
• /
• pp.454-460
• /
• 1998

The vibration behavior of fuel rods has been analyzed by FEM in consideration of axial force and support spring constants. The axial compression force on the fuel rod in reactor decreases with the fuel rod burnup, and its decrease makes the natural frequencies of fuel rod considerably increase. The change of support spring constant can contribute to the remarkable change of the mode shapes, but not greatly to the natural frequencies. The reaction forces of support springs are obtained from normalizing the lst mode with the max. 0.2 mm displacement. The calculated reaction forces are larger than the previous results obtained by disregarding the deflections of the support springs.

• Nuclear Engineering and Technology
• /
• v.55 no.9
• /
• pp.3241-3251
• /
• 2023

The potential for recriticality and sub-critical boron concentrations is analyzed during the relocation of the fuel rods in the assembly, which we call late phase of a severe accident, via coupling between MELCOR and whole-core Monte Carlo analyses by Serpent 2. The recriticality, initiated during the early phase, is found to maintain when the fuel assemblies containing intact fuel rods are submerged by the cooling water. It is also found that the effect of the negative reactivity insertion via remaining fission products in the fuel debris increases as the burnup increases. The sub-critical boron concentrations during the late phase are found to be 76~544 ppm lower than those during the early phase. Therefore, it can be concluded that the boron concentration that prevents recriticality not only during the early phase but also during the late phase is the sub-critical boron concentration during the early phase.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 2002.05a
• /
• pp.684-687
• /
• 2002

The authors have developed two devices which cuts the spend fuel rod to an optimal size and extracts fuel pellet from the pieces of cut fuel rods. These devices are so important to reduce radioactive wastes that some advanced countries developed their own methods and devices. The authors have benchmarked from these methods and devices. For spent fuel rod cutting, the tube cutting method has been chosen. some mechanical properties of the fuel tube and pellet has been carefully considered for an optimal cutting size. For fuel pellet extraction, a mechanically extracting method has been adopted. The existing chemical method have turned out to be inappropriate because it produced large amount of radioactive wastes, in spite of its high fuel recovery characteristics. The developed method has an advantage that it can be applied to other fuel rods that have different shapes and sizes. The two devices are set up and operated in the hot cell where people can not go in, so that the devices have been designed to be controlled remotely and modulated for easy maintenance. And the performance of the devices has been tested by using simulated fuel rod. From the experimental results, the devices are supposed to be useful for reducing radioactive wastes.

• Nuclear Engineering and Technology
• /
• v.26 no.1
• /
• pp.90-99
• /
• 1994

One of the major phenomena occurring in defective fuel rods is the oxidation of UO$_2$ fuel pellets from UO$_2$ to UO$_{2+}$x/ by the oxygen Produced from the dissociation of the steam in the Pellet-to-clad gap, which leads to the enhancement of fission gas release. In this paper, the oxidation kinetics of defective fuel rods was analyzed on the basis of operating conditions of the reactor and defective fuel rod itself. Oxidation kinetics of the fuel pellet was determined under the assumption that the gap is filled with the saturated steam of 150 atm and an enhancement factor for fission gas release was introduced to take into account the effect of fuel oxidation on fission gas release. Comparison with experimental data shows that the enhancement factor predicts well the increased fission gas release due to the oxidation of UO$_2$fuel pellets.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.16 no.11 s.116
• /
• pp.1115-1123
• /
• 2006

Nuclear fuel rods are exposed to axial flow in a reactor, and flow-induced-vibration due to the flow usually causes damage in the fuel rods. Thus a prior knowledge about dynamic behavior of a fuel rod exposed to the flow condition should be provided. This paper shows that dynamic characteristics of a nuclear fuel rod depend on axial flow velocity. Assuming small lateral displacement, the effects of uniform axial flow are investigated. The analytic results show that axial flow generally reduces fuel rod stiffness and raises its damping in normal condition. Also, the critical axial velocities which make the fuel rod behavior unstable were found. That is, solving generalized eigenvalue equation of the fuel rod dynamic system, the eigenvalues with positive real part are detected. Based on the simulation results, on the other hand, it turns out that the ordinary axial flow in nuclear reactors does not affect to stability of a nuclear fuel rod even in the conservative condition.

• The Transactions of The Korean Institute of Electrical Engineers
• /
• v.61 no.12
• /
• pp.1948-1952
• /
• 2012

The paper presents modal testing and analysis in order to obtain the dynamic characteristics of a middle spacer grids of a nuclear fuel rod. A spacer grid is one of the important structural elements supporting nuclear fuel rods. Such a fuel rod can be oscillated by its thermal expansion, neutron irradiation and etc. due to cooling water flow under the operation of a nuclear power plant. When the fuel rod vibrates, fretting wear due to repeated friction motion between the fuel rods and spacer grids can be occurred, and so the fuel rod is damaged. In this paper, through modal analysis and testing, natural frequencies and modes of a middle spacer grid were calculated, and the following conclusions were obtained. Firstly the numerical first-seven natural frequencies for spacer grids of a fuel rod having complicated structures have a small difference within 3.8% with experimental natural frequencies, and so the suitability of simulation results was verified. Secondly, experimental mode shapes for a middle spacer grid of a nuclear fuel rod were verified by obtaining lower non-diagonal terms through MAC(Modal Assurance Criteria), and were confirmed by the simulation modes.

• Nuclear Engineering and Technology
• /
• v.33 no.1
• /
• pp.73-82
• /
• 2001

Pressurized water reactor(PWR) fuel rods. which are continuously supported by a spring system called a spacer grid(SG), are exposed to reactor coolant at a flow velocity of up to 6-8 m/s. It is known that the vibration of 3 fuel rod is generated by the coolant flow, a so-called flow-induced-vibration(FIV), and the relative motion induced by the FIV between the fuel rod and the SG can wear away the surface of the fuel rod, which occasionally leads to its fretting failure. It is, therefore, important to understand the vibration characteristics of the fuel rod and reflect that in its design. In this paper, vibration analyses of the fuel rod with two different SGs were performed using both analytical and experimental methods. Updating of the finite element(FE) model using the measured data was performed in order to enhance confidence in the FE model of fuel rods supported by an SG. It was found that the modal parameters are very sensitive to the spring constant of the SG.