• Nuclear Engineering and Technology
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• 제38권3호
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• pp.301-310
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• 2006

Destructive methods were used for the burnup determination of U-Zr alloy nuclear fuel irradiated in the High-flux Advanced Neutron Application Reactor (HANARO) at KAERI. The dissolution rate of unirradiated U-Zr alloy fuel in $HNO_3$/HF mixtures was investigated for the experimental conditions of a different temperature, and initial concentrations of HF and $HNO_3$. The irradiated U-Zr alloy fuel specimen was dissolved in a mixed acid condition of 3 M HNO3 and 1 M HF at $90^{\circ}C$ for 8 hours under reflux. The total burnup was determined from measurement of the Nd isotope burnup monitors. The method includes U, Pu, $^{148}Nd,\;^P{145}Nd+^{146}Nd,\;^{144}Nd+^{143}Nd$ and total Nd isotopes determination by the isotope dilution mass spectrometric method (IDMS) using triple spikes $(^{233}U,\;^{242}Pu\;and\;^{150}Nd)$. The effective fission yield was calculated from the weighted fission yields averaged over the irradiation period. The results are compared with that obtained by the destructive -spectrometric measurement of the $^{137}Cs$ monitor.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.859-870
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• 2013

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ${\sim}60%^{235}U$; the mini-rods were irradiated to an average burnup of ${\sim}85%^{235}U$. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ~ 40% up to ~ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ~ 40% up to ~ 85%.

• 한국재료학회:학술대회논문집
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• 한국재료학회 2005년도 춘계학술발표대회 및 제8회 신소재 심포지엄 논문개요집
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• pp.183-183
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• 2005

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 추계학술논문요약집
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• pp.95-96
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• 2017

• 대한금속재료학회지
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• 제49권7호
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• pp.549-556
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• 2011

This study aimed to evaluate the performance of diffusion barriers in order to prevent fuel-cladding chemical interaction (FCCI) between the metallic fuels and the cladding materials, a potential hazard for nuclear fuel in sodium-cooled fast reactors. In order to prevent FCCI, Zr or V metal is deposited on the ferritic-martensitic stainless steel surface by physical vapor deposition with a thickness up to $5{\mu}m$. The diffusion couple tests using uranium alloy (U-10Zr) and a rare earth metal such as Ce-La alloy and Nd were performed at temperatures between 660~800$^{\circ}C$. Microstructural analysis using SEM was carried out over the coupled specimen. The results show that significant interdiffusion and an associated eutectic reaction ocurred in the specimen without a diffusion barrier. However, with the exception of the local dissolution of the Zr layer in the Ce-La alloy, the specimens deposited with Zr and V exhibited superior eutectic resistance to the uranium alloy and rare earth metal.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 춘계학술논문요약집
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• pp.73-74
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• 2017

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4084-4094
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• 2022

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. The first cycle of FAST testing was completed and four rodlets were removed for the initial post irradiation examination (PIE). The rodlet design and irradiation conditions were evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % fissions per initial metal atoms [%FIMA]), control rodlets, and a helium-bonded annular rodlet (4.7 %FIMA). Nondestructive PIE has been completed and includes visual inspection, neutron radiography and gamma scanning of the FAST capsules and rodlets. Radiography confirmed the integrity of the experiments, revealed that the annulus in the annular fuel was filled at a modest burnup (4.7 %FIMA), and indicated potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior, except for cesium in the He-bonded annular fuel. Future destructive PIE will be necessary to fully interpret the effects of accelerated irradiation on U-Zr metallic fuel behavior.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.734-743
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• 2017

In order to investigate the swelling behavior and fuel-cladding interaction mechanism of U-10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel-cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal-hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.395-400
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• 2003

사용후핵연료 금속전환체의 장기저장 안정성 확보를 위해 금속전환체의 주성분인 금속우라늄과 산화 안정화 후보물질인 Nb, Ti, Ni, Zr, Hf 등을 첨가한 모의금속전환체 합금을 제작하여 $200^{\circ}C~300^{\circ}C$ 온도구간에서 순수 산소분위기로 산화시험을 수행하였다. U-Nb, U-Zr, U-Ti 합금은 순수 금속우라늄보다 무게증가 측면에서의 산화저항성이 높았으나, U-Hf, U-Ni 합금의 경우에는 오히려 순수 금속우라늄보다 산화 저항성이 낮게 나타났다. 시편에 대한 미세성분 및 조직을 광학현미경, SEM, EPMA 등을 통해 분석하였다. 각 합금의 산화율 및 활성화에너지를 구한 결과 U-Nb 합금의 활성화에너지가 높고 산화 저항성이 가장 양호한 것으로 나타나 산화 저항성 후보물질로 선정하였다.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1300-1309
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• 2023

Simulated debris was synthesized using UO₂, Zr, and stainless steel and a heat treatment method under inert or oxidizing conditions. The primary U solid phase of the debris synthesized at 1473 K under inert conditions was UO₂, whereas a (U, Zr)O₂ solid solution formed at 1873 K. Under oxidizing conditions, a mixture of U₃O₈ and (Fe, Cr)UO₄ phases formed at 1473 K, whereas a (U, Zr)O_2+x solid solution formed at 1873 K. The leaching behavior of the fission products from the simulated debris was evaluated using two methods: the irradiation method, for which fission products were produced via neutron irradiation, and the doping method, for which trace amounts of non-radioactive elements were doped into the debris. The dissolution behavior of U depended on the properties of the debris and aqueous solution for immersion. Cs, Sr, and Ba leached out regardless of the primary solid phases. The leaching of high-valence Eu and Ru ions was suppressed, possibly owing to their solid-solution reaction with or incorporation into the uranium compounds of the simulated debris.