• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2001.11a
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• pp.245-251
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• 2001

A fretting wear tester is developed for experimental study on the fuel fretting problem of light water reactor. The feature of the developed tester is it can simulate the existence of gap between spring and fuel rod as well as different contacting force including the just-contact condition (0 N on the contact). Used are a servo-motor, an eccentric cylinder and lever mechanism for driving system. A spacer grid cell is constituted with four strap segments (each segment has a spring). This fretting wear tester can also be used as a fatigue tester of a spacer grid spring with the frequency of more than 10 Hz. It is required to simulate the frequency of the vibrating fuel rod due to flow-induced vibration in a reactor. In fretting wear test, up to two span-length of a fuel cladding tube can be accommodated. A specimen of cladding tube of one span-length is specially designed, which can be extended for two-span test. For .fatigue test, a device for clamping the spring fixture is installed additionally, Presently, the tester is designed for the condition of air environment and room temperature. The variation of the reciprocal distance is measured to check the stability of input force, which will be exerted to the cladding (for fretting wear. test) and the spring (for fatigue test) specimen.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.741-751
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• 2021

There has been a deep interest in trying to find better-performing fuel clad motivated by the desire to decrease the likelihood of the reactor barrier failure like what happened in Fukushima in recent years. In this study, the effect of move towards accident tolerant fuel (ATF) cladding as the most attracting concept for improving reactor safety is investigated for SMART modular reactor. These reactors have less production cost, short construction time, better safety and higher power density. The SiC and FeCrAl materials are considered as the most potential candidate for ATF cladding, and the results are compared with Zircaloy cladding material from reactor physics point of view. In this paper, the calculations are performed by generating PMAX library by DRAGON lattice physics code to be used for further reactor core analysis by PARCS code. The differential and integral worth of control and safety rods, reactivity coefficient, power and temperature distributions, and boric acid concentration during the cycle are analyzed and compared from the conventional fuel cladding. The rod ejection accident (REA) is also performed to study how the power changed in response to presence of the ATF cladding in the reactor core. The key quantitative finding can be summarized as: 20 ℃ (3%) decrease in average fuel temperature, 33 pcm (3%) increase in integral rod worth and cycle length, 1.26 pcm/℃ (50%) and 1.05 pcm/℃ (16%) increase in reactivity coefficient of fuel and moderator, respectively.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.249-257
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• 2015

A fuel assembly consists of fuel rods composed of pellets (UO2) and a cladding tube (Zircaloy). The role of the fuel rods in the reactor is to generate heat by nuclear fission, as well as to retain fission products during operation. A simulation method using a computer program was used to evaluate the safety of the nuclear fuel rods. This computer program has been called the fuel performance code. In the analysis of a light water reactor fuel rod, the gap conductance, which depended on the distance between the pellets and cladding tube, mainly influenced the thermomechanical behavior of the fuel rod. In this work, a 3D gap element was proposed to simulate the thermo-mechanical behavior of the nuclear fuel rod, considering the gap conductance. To implement the proposed 3D gap element, a 3D thermo-mechanical module was also developed using FORTRAN90. The asymmetric characteristics of the nuclear fuel rod, such as the MPS (missing pellet surface) and eccentricity, were simulated to evaluate the proposed 3D gap element.

• Journal of the Korean Society for Nondestructive Testing
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• v.16 no.1
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• pp.29-33
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• 1996

The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase & amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2003.05a
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• pp.82-87
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• 2003

Nuclear fuel with a big slenderness ratio is susceptible to flow-induced vibration under very severe conditions of high temperature, high flow and exposure to irradiation in nuclear reactor. The fuel assembly should, therefore, be designed to escape any resonance due to the vibration during the reactor operation, in particular, in case of the design changes. In addition, the amplitudes due to the grid vibration, the fuel rod vibration and the fuel assembly vibration should be minimized to reduce the grid-to-rod fretting wear. Fuel assembly vibration tests in air at room temperature and in water at high temperature have been performed to investigate fuel vibration behaviors. The frequency and damping during the test in air have been compared to those in water. Through the hydraulic test, the advanced assembly has been evaluated not to be susceptible to any resonance. In addition, the test data from the tests can be used to make fuel model and to evaluate grid-to-rod fretting wear.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3095-3105
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• 2022

Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7×7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9×9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1407-1424
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• 2024

In Pressurized Water Reactors, most of the failed fuel rods are often observed at the periphery of the fuel assembly, especially near the core baffle. The rod vibration-induced fretting wear is a significant failure mechanism strongly correlated with the coolant and support conditions. This paper presents a coupled vibration model of double-rod to predict the grid-to-rod fretting (GTRF) wear. A motion-dependent fluid force model is used to simulate the coolant cross flow, the gap constraints with asymmetric stiffness between spring and dimple on the vibration form, and the fretting wear are discussed. The results show the effect of the coupled vibration on the deterioration of wear, providing a sound theoretical explanation of some failure phenomena observed in the previous experiment. Exploratively, we analyze the impact of the baffle jet on the GTRF wear, which indicates that the high-velocity cross-flow will significantly affect the vibration forms while sharply changing the wear behavior.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3653-3664
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• 2021

In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO₂ benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1775-1782
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• 2023

Several reactor physics commissioning experiments were conducted to obtain the neutronic parameters at the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated the calculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuel reactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS first core up to the third core configuration, which has been done experimentally during these commissioning periods. The excess reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to the experiment data. The calculated integral reactivity of the control rod has an error of less than 1.0% compared to the experimental data. The calculated fuel reactivity value is consistent with the measured data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core model is in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.

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