• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.254-254
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• 2002

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.1044-1062
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• 2017

The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U-Mo fuel. Plate-type U-7 wt.% Mo/Al-5 wt.% Si, referred to as U-7Mo/Ale5Si, dispersion fuel with a uranium loading of $8.0gU/cm^3$, was selected to achieve higher fuel efficiency and performance than are possible when using $U_3Si_2/Al$ dispersion fuel. To qualify the U-Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U-7Mo/Al-5Si dispersion fuel ($8gU/cm^3$), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U-7Mo/Al-5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U-Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U-Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.453-454
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• 2005

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1213-1221
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• 2020

Chromium-nickel alloy 42XNM (XHM-1, Bochvalloy) is considered as a promising material for future generations of nuclear reactors, primarily as a material for the fuel elements shells in the development of accident tolerant fuel. However, as with most nickel-based alloys, 42ХNМ is characterized by a sharp decrease in plastic properties in the temperature range of (500-900)℃. This effect is enhanced by neutron irradiation. Preliminary tests of ring samples of 42XNM alloy (after irradiation as a part of the VVER-1000 control system) in the temperature range of ductility failure showed that the standard technique for processing tensile diagrams does not allow to evaluate the plastic properties correctly at low strains. Therefore, in this work, the alternative method for testing ring samples from materials with low plastic characteristics was developed. It was shown that the minimum value of the permanent strain of the irradiated 42XNM alloy in the temperature range of (500-1100)℃, determined by the alternative method, was ~1.6% at 750 ℃.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1330-1338
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• 2018

High-density UMo/Zr monolithic nuclear fuel plates have a promising application prospect in high flux research and test reactors. The solid state welding method called co-rolling is used for their fabrication. Hot co-rolling simulations for the composite blanks of UMo/Zr monolithic nuclear fuel plates are performed. The effects of coupon sizes and mechanical property parameters on the contact pressures between the to-be-bonded surfaces are investigated and analyzed. The numerical simulation results indicate that 1) the maximum contact pressures between the fuel coupon and the Zircaloy cover exist near the central line along the plate length direction; as a whole the contact pressures decrease toward the edges in the plate width direction; and lower contact pressures appear at a large zone near the coupon corner, where de-bonding is easy to take place in the in-pile irradiation environments; 2) the maximum contact pressures between the fuel coupon and the Zircaloy parts increase with the initial coupon thickness; after reaching a certain thickness value, the contact pressures hardly change, which was mainly induced by the complex deformation mechanism and special mechanical constitutive relation of fuel coupon; 3) softer fuel coupon will result in lower contact pressures and form interfaces being more out-of-flatness.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.187-198
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• 2009

In this paper, three power ramp tests performed on high burn-up Re-crystallized Zircaloy2 - UO2 BWR fuel rods (56 to 63 MWd/kgU) within the SCIP project are simulated with METEOR and ALCYONE 3D. Two of the ramp tests are of staircase type up to Linear Heat Rates of 420 and 520 W/cm and with long holding periods. Failure of the 420 W/cm fuel rod was observed after 40 minutes. The third ramp test consisted of a more standard ramp test with a constant power rate of 80 W/cm/min up to 410 W/cm with a short holding time. The tests were first simulated with the METEOR 1D fuel rod code, which gave accurate results in terms of profilometry and fission gas releases. The behaviour of a fuel pellet fragment and of the cladding piece on top of it was then investigated with ALCYONE 3D. The size and the main characteristics of the ridges after base irradiation and power ramp testing were recovered. Finally, the failure criteria validated for PWR conditions and fuel rods with low-to-medium burn-ups were used to analyze the failure probability of the KKL rodlets during ramp testing.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.820-828
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• 2018

The growing nuclear threat has amplified the need for developing diverse and accurate nuclear forensics analysis techniques to strengthen nuclear security measures. The work presented here is part of a research effort focused on developing a methodology for reactor-type discrimination of weapons-grade plutonium. To verify the developed methodology, natural $UO_2$ fuel samples were irradiated in a thermal neutron spectrum at the University of Missouri Research Reactor (MURR) and produced approximately $20{\mu}g$ of weapons-grade plutonium test material. Radiation transport simulations of common thermal reactor types that can produce weapons-grade plutonium were performed, and the results are presented here. These simulations were needed to verify whether the plutonium produced in the natural $UO_2$ fuel samples during the experimental irradiation at MURR was a suitable representative to plutonium produced in common thermal reactor types. Also presented are comparisons of fission product and plutonium concentrations obtained from computational simulations of the experimental irradiation at MURR to the nondestructive and destructive measurements of the irradiated natural $UO_2$ fuel samples. Gamma spectroscopy measurements of radioactive fission products were mostly within 10%, mass spectroscopy measurements of the total plutonium mass were within 4%, and mass spectroscopy measurements of stable fission products were mostly within 5%.

• 한국재료학회:학술대회논문집
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• 한국재료학회 2002년도 춘계 학술발표강연 및 논문개요집
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• pp.168-168
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.225-225
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.219.2-219
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• 2001