• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.253.1-253
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.253.2-253
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.301.1-301
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• 1999

• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.347-354
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• 2009

Welding is required for a connection between two different components in the nuclear fuel of a pressurized water reactor. This work relies on a mechanical experiment and analytic results to investigate the structural integrity of nuclear fuel in a situation where some components are not welded to each other. A series of lateral vibration tests are performed in a test facility, and the test structures are examined in terms of dynamic behavior. In the tests, the displacement signal at every grid structure that sustains fuel rods is measured and processed to identify the dynamic properties. The fluid-elastic stability of the structure is also analyzed to evaluate susceptibility to a cross flow with an assumed conservative cross flow distribution. The test and analysis results confirm that the structural integrity can be maintained even in the absence of some welding connections.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.306-312
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• 2018

There is an increasing need in the United States and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations or interim storage sites. Under normal conditions, the Nuclear Regulatory Commission limits cladding temperature to $400^{\circ}C$ for high-burnup (>45 GWd/mtU) fuel, with higher temperatures allowed for low-burnup fuel. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at $400^{\circ}C$. Results were representative of the majority of US light water reactor fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1602-1611
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• 2021

A varying degree of eccentricity always exists in the initial configuration of a nuclear fuel rod. Its impact on traditional LWR fuel is limited as the radial gap closes relatively early during irradiation. However, the effect of misalignment is expected to be more relevant in rods with highly conductive fuels, large initial gaps and low conductivity filling gases. In this paper, we study similar characteristics in the experimental setup of two fuel disc irradiation campaigns carried out in the OECD Halden Boiling Water Reactor. Using the multi-dimensional fuel performance code OFFBEAT, we combine 2-D axisymmetric and 3-D simulations to investigate the effect of eccentricity on the fuel temperature distribution. At the same time, we illustrate how the advent of modern tools with multi-dimensional capabilities might further improve the design and interpretation of in-pile separate-effect tests and we outline the potential of such an analysis for upcoming experiments.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• Corrosion Science and Technology
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• v.2 no.6
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• pp.253-259
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• 2003

Zirconium based alloys have been extensively used as a cladding material for fuel rods in nuclear reactors, due to their low thermal neutron absorption cross-section, excellent corrosion resistance and good mechanical properties at high temperatures. However, a cladding material for fuel rods in nuclear reactors was contact water during long time at high-temperature, so it is necessary to improve the wear and corrosion resistance of the fuel cladding, At ambient environment, there are few data or paper on the characteristic of corrosion in chloride solution and acidic solution. The specimens used in this work are pure Zr and Zircaloy-4. Zircaloy-4 is a specific zirconium-based alloy containing, on a weight percent basis, 1.4% Sn, 0.2% Fe, 0.1% Cr. Pitting corrosion resistance of two alloys by ASTM G48 is higher than that of electrochemical method. Passive film formed on Zircaloy-4 is mainly composed of $ZrO_2$, metallic Sn, and iron species regardless of formation environments. Also, passive film formed on Zr alloys shows n-type semiconductic property on the base of Mott-Schottky plot.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.175-180
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• 1997

It has been known that eater-side corrosion of fuel rods in nuclear reactor is accompanied with the metallic loss of wall thickness and hydrogen pickup in the fuel dadding tube. The fuel dad corrosion is one of the major factors to be controlled to maintain the fuel integrity during reactor operation. An oxide later thickness measuring device equipped with ECT probe system was developed by KAERI, and whose performance test was carried out in NDT(Non-destructive Test) hot-cell or PIE(Post Irradiation Examination) Facility. At first, the calibration/performance test was executed for the unirradiated standard specimen rod fabricated with several kinds of plastic thin films whose thickness ore predetermined, and the result of which showed a good precision within 10% of discrepancy. And then, hot test us peformed for the irradiated fuel rod selectively extracted from J44 fuel assembly discharged from Kori Unit-2. The data obtained with this device were compared with the metallographic result obtained from destructive examination in PIEF hot-cell on the same fuel rod to verify the validity of the measurement data.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.375-388
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• 2024

In this study, to evaluate the shock and vibration load characteristics of used fuel, a sea transportation test was conducted using simulated fuel assemblies under normal transport conditions. An overall test data analysis was performed based on the measured strain and acceleration data obtained from cruise, rotation, acceleration, braking, depth of water, and rolling tests. In addition, shock response spectrum and power spectral densities were obtained for each test case. Amplification and attenuation characteristics were investigated based on the load path. The load was amplified as it passed from the overpack to the simulated used fuel-assembly. As a result of the RMS trend analysis, the fuel-loading position of the transportation package affected the measured strain in the fuel rod, and the maximum strains were obtained at the spans with large spacing. However, even these maximum strains were very small compared to the fatigue strength and the cladding yield strength. Moreover, the fuel rods located on the side exhibited a larger strain value than those at the center.