• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.363-373
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• 1991

The current rod internal pressure criterion permits fuel rods to operate with internal pressures in excess of system pressure only if internal overpressure does not cause the diametral gap enlargement. In this study, the generic allowable internal gas pressure not violating this criterion is estimated as a function of rod power. The results show that the generic allowable internal gas pressure decreases linearly with the increase of rod power. Application of the generic allowable internal gas pressure for the rod internal pressure design criterion will result in the simplication of the current design procedure for checking the diametral gap enlargement caused by internal overpressure because according to the current design procedure the cladding creepout rate should be compared with the fuel swelling rate at each axial node at each time step whenever internal pressure exceeds the system pressure.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.100-107
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• 1994

A statistical methodology is developed for calculating the nuclear fuel pod internal pressure of Korean PWR fuel in order to reduce over-conservatism of the current KAERI deterministic methodology. The developed statistical methodology employs the response surface method and Monte Carlo calculation. The simple regression equation for the rod internal pressure is derived by taking into account the various fuel fabrication-related and fuel performance model-related parameters. The validity of the regression equation is examined by the F-test, $R^2$-method and Cp-test The internal pressure predicted by the regression equation is in good agreement with that calculated by he computer code using the KAERI deterministic methodology. The distribution of the internal pressure from the Monte Carlo calculation is found to be normal. Comparison of the 95/95 rod internal pressure predicted by the developed statistical methodology with the maximum rod internal pressure by the deterministic methodology shows that the developed statistical methodology reduces significantly over-conservatism of the deterministic methodology.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.257-266
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• 1999

A simplified statistical methodology is developed in order to both reduce over-conservatism of deterministic methodologies employed for PWR fuel rod internal pressure (RIP) calculation and simplify the complicated calculation procedure of the widely used statistical methodology which employs the response surface method and Monte Carlo simulation. The simplified statistical methodology employs the system moment method with a deterministic approach in determining the maximum variance of RIP The maximum RIP variance is determined with the square sum of each maximum value of a mean RIP value times a RIP sensitivity factor for all input variables considered. This approach makes this simplified statistical methodology much more efficient in the routine reload core design analysis since it eliminates the numerous calculations required for the power history-dependent RIP variance determination. This simplified statistical methodology is shown to be more conservative in generating RIP distribution than the widely used statistical methodology. Comparison of the significances of each input variable to RIP indicates that fission gas release model is the most significant input variable.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1596-1609
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• 2019

The fuel rod performance has great importance for the safety and economy of an operating reactor. The fuel rod performance analysis code, which considers the thermal-mechanical response and irradiation effects of fuel rod, is usually developed in order to predict fuel rod performance accurately. The FRIPAC (${\underline{F}}uel$ ${\underline{R}}od$ ${\underline{I}}ntegral$ ${\underline{P}}erformance$ ${\underline{A}}nalysis$ ${\underline{C}}ode$) is such a fuel rod performance analysis code that has been developed recently by China Nuclear Power Technology Research Institute Co. Ltd. The code aims at the computational simulation of the Pressurized Water Reactor fuel rod behavior for both steady-state and power ramp condition. A brief overview of FRIPAC is presented including the computational framework and the main behavioral models. Validation of the code is also presented and it focuses on the fuel rod behavior including fuel center temperature, fission gas release, rod internal pressure/internal void volume, cladding outer diameter and cladding corrosion thickness. The validation is based on experimental data from several international projects. The validation results indicate that FRIPAC is an accurate and reliable fuel rod performance analysis code because of the satisfactory comparison results between the experimental measurements and the code predictions.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.242-242
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• 2002

Two MIMAS MaX fuel rods base-irradiated in a commercial PWR have been reinstrumented and irradiated at a test reactor. The fabrication data for two MOX roda are characterized together with base irradiation information. Both Rods were reinstrumented to be fitted with thermocouple to measure centerline temperature of fuel. One rod was equipped with pressure transducer for rod internal pressure whereas the other with cladding elongation detector. The post irradiation examinations for various items were performed to determine fuel and cladding in-pile behavior after base irradiation. By using well characterized fabrication and re-instrumentation data and power history, the fuel performance code, COSMOS, is verified with measured in-pile and PIE information. The COMaS code shows good agreement for the cladding oxidation and creep, and fission gas release when compared with PIE dad a after base irradiaton. Based on the re-instrumention information and power history measured in-pile, the COSMOS predicts re-instrumented in-pile thermal behaviour during power up-ramp and steady operation with acceptable accuracy. The rod internal pressure is also well simulated by COSMOS code. Therfore, with all the other verification by COSMOS code up to now, it can be concluded that COSMOS fuel performance code is applicable for the design and license for MaX fuel rods up to high burnup.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.175-180
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• 1998

The U-Zr metallic alloy with low swelling HT9 cladding is the candidate for the KALIMER fuel rod. The fuel rod should be able to maintain the structural integrity during its lifetime in the reactor. In a typical metallic fuel rod, load is mainly applied by internal gas pressure, and the deformation is primarily caused by creep of the cladding. The three-dimensional FEM modelling of a fuel rod is important to predict the structural behavior in concept design stage. Using the ANSYS code, the 3-D structure analyses were performed for various configuration, element and loads. It has been shown that the present analysis model properly evaluate the structural integrity of fuel rod. The present analysis results show that the fuel rod is expected to maintain its structural integrity during normal operation.

• Proceedings of the Materials Research Society of Korea Conference
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• 2002.05a
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• pp.164-164
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.252-252
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.285-285
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• 2002

• Journal of Drive and Control
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• v.12 no.4
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• pp.15-20
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• 2015

Unlike a typical metal spring, a gas spring uses compressed gas contained in a cylinder and compressed by a piston to exert a force. A common application includes automobiles where gas spring are incorporated into the design of open struts that support the weight of tail gate. They are also used in furniture such as office chairs, and in medical and aerospace applications. The gas spring works by the application of pressurized gas (nitrogen) contained in a cylinder. The internal pressure of the gas spring greatly exceeds atmospheric pressure. This differential in pressure exists at any rod position and generates an outward force on the rod, making the gas spring extend. In this paper, we investigated the dynamic characteristics of a gas spring on an automotive tail gate system.