• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.745-752
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• 1995

고연소도 액체금속로용 금속연료를 개발하고자 U-l0wt%Zr 합금중 Zr 원소 대신에 X(:Si, Ta, Nb, W, Mo) 원소를 첨가한 U-7wt%Zr-3wt%X(:Si, Ta, Nb, W, Mo) 합금을 제조하여 미세조직에 미치는 합금원소 첨가의 영향을 조사하였다. 그 결과 U-7 wt%Zr-3wt%Si 합금을 제외한 모든 U-7wt%Zr-3wt%X(:Ta, Nb, W, Mo) 합금은 Matrix에 있어서 Laminar Structure를 그대로 유지하였다. U-7wt%Zr-3wt%Si 함금을 제외한 모든 U-7wt%Zr-3wt%X(:Ta, Nb, W, Mo) 합금의 주요한 상은 U-l0wt% Zr 합금과 마찬가지로 $\alpha$-U 및 $\delta$-UZr$_2$ 상이었다. U-7wt%Zr-3wt%X(:Ta, Nb, W, Mo) 합금은 U-l0wt%Zr 합금에 비해 Lamina Thickness가 크게 감소되었다. 특히 U-7wt%Zr-3wt%Mo 합금의 경우에 있어서는 U-l0wt%Zr 합금에 비해 1/3배 정도까지 Lamina Thickness가 크게 감소하였다. 이와 같은 합금원소 첨가에 의한 Laminar Structure의 미세화는 액체금속로강 금속연료내 Fission Gas의 Inter-connected Path가 보다 더 잘 형성됨으로 인해 Fission Gas Bubble에 대한 방출속도를 크게 증가시켜서 궁극적으로는 Fission Gas Bubble에 의한 Swelling을 저감시킬 것으로 기대된다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.177-182
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• 1997

금속연료가 연소할 때 발생하는 Fission Gas는 주로 직경방향으로의 Swelling을 일으켜 낮은 연소도의 원인이 되어 왔다. 따라서 본 연구에서는 높은 연소도를 갖는 금속연료의 개발을 목적으로 Fission Gas가 Plenum으로 쉽게 방출하는 조직인 Laminar Structure를 갖는 합금의 설계를 연구하였다. 또한 조사 후의 조직안정성을 예측하기 위해 열처리 후의 미세조직의 변화를 관찰하여 조직안정성을 시험하였다. U-10wt.%Zr 합금 중 Zr 원소 대신에 2wt.% 및 3wt.%의 W 또는 Mo을 첨가한 합금을 제조하여 합금원소 첨가의 영향에 따른 미세조직의 변화를 조사하였다. 그 결과 모든 조성의 합금은 Matrix에 있어서 Laminar Structure를 나타내었다. 또한 U-10wt.%Zr에 비해 2wt.% 및 3wt.%W의 W 또는 Mo를 첨가한 합금의 lamina Thickness가 철면 미세해짐을 확인하였다. 특히 U-7wt.%Zr-3wt.%W의 경우는 U-10wt.Zr에 비해 Laminar Thickness가 1/2배까지 감소되었다. 합금원소(W, Mo) 첨가에 의한 Laminar Thickness의 감소는 Fission Gas의 Inter-connected Path가 보다 잘 형성되게 하여 Gas의 방출속도를 증가시켜 Swelling을 감소시킬 것으로 생각된다. 열처리한 금속연료의 미세조직을 비교한 결과를 보면 합금원소(W, Mo)를 첨가한 합금을 50$0^{\circ}C$에서 1000시간동안 열처리한 것을 U-Zr 2원계 합금을 열처리한 것과 비교했을 때 약 1/3배 정도의 Laminar Thickness를 유지하는 것으로 보아 합금원소를 첨가하면 조사 후의 조직안정성에도 크게 기여할 것으로 기대된다.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2629-2639
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• 2021

Within the frame of the EMPIrE test, four monolithic mini-plates were irradiated in the ATR reactor. In two of them, the monolithic U(Mo) foil had been PVD-coated with Zr before the plate manufacturing. Extensive microstructural characterizations were performed on a fresh archive mini-plate, using Optical Microscopy (OM), Scanning Electron Microscopy (SEM) combined with Energy Dispersive Spectroscopy (EDS), Electron Backscattered Diffraction (EBSD) and Focused Ion Beam (FIB)/Transmission Electron Microscopy (TEM) with nano EDS. A particular attention was paid to the examination of the U(Mo) foil, the PVD coating, the cladding/Zr and Zr/U(Mo) interfaces. The Zr coating has a thickness around 15 ㎛. It has a columnar microstructure and appears dense. The cohesion of the cladding/Zr and Zr/U(Mo) interfaces seems to be satisfactory. An almost continuous layer with a thickness of the order of 100-300 nm is present at the cladding/Zr interface and corresponds to an oxidized part of the Zr coating. At the Zr/U(Mo) interface, a thin discontinuous layer is observed. It could correspond to locally oxidized U(Mo). This work provides a basis for interpreting the results of characterizations on EMPIrE irradiated plates.

• Proceedings of the Materials Research Society of Korea Conference
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• 2003.03a
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• pp.147-147
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• 2003

• Bulletin of the Korean Chemical Society
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• v.23 no.8
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• pp.1067-1072
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• 2002

A new polystyrene-divinylbenzene resin containing 2-(2-thiazolylazo)-5-dimethylamino-phenol (TAM) functional groups has been synthesized and its sorption behavior for nineteen metal ions, including Zr(Ⅳ),Hf(Ⅳ) and U(Ⅵ) has been investigated by batch and column methods. The chelating resin showed high sorption affinity for Zr(Ⅳ) at pH 1-5 and U(Ⅵ) at pH 4. Some parameters affecting the sorption of the metal ions have been detailed. The breakthrough and overall capacities were measured under optimized conditions. The overall capacities of Zr(Ⅳ), Th(Ⅳ) and U(Ⅵ), which showed higher than the other metal ions, were 0.90,0.84 and 0.80 mmol/g, respectively. The elution order of metal ions at pH 4 was evaluated as Zr(Ⅳ) > Th(Ⅳ) > U(Ⅵ) > Cu(Ⅱ) > Hf(Ⅳ) > W(Ⅵ) > Mo(Ⅵ) > In(Ⅲ) > Sn(Ⅳ) > Cr(Ⅲ) > V(Ⅴ) > Fe(Ⅲ). Quantitative recovery of most metal ions except Zr(Ⅳ) was achieved using 2M HNO3. Desorption and recovery of Zr(Ⅳ) was successfully performed with 2 M HClO4 and 2 M HCl.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.269.1-269
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• 2002

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1096-1108
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• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.198-203
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• 1998

액체금속로용 금속연료인 U-10wt.%Zr 합금의 기지조직은 공석조직으로 $\alpha$uranium과 UZr$_2$인 $\delta$상이 교대로 나타나는 층상조직을 이루고 있다. 잘 발달된 층상조직의 두께는 $\alpha$U이 40-8$\mu\textrm{m}$, $\delta$상인 UZr$_2$는 20-30$\mu\textrm{m}$로 $\alpha$U이 $\delta$상의 2-3배 정도 된다. 기지조직내에 나타나는 둥근 형태의 석출물 크기는 ø5-12$\mu\textrm{m}$이며, 응집된 석출물의 크기는 ø15-25$\mu\textrm{m}$이다. 석출물의 TEM SADP과 EDS 분석결과 순수한 $\alpha$Zr이 아니고 산소에 의하여 안정화되고 소량의 uranium을 함유한 Zr rich 상으로 $\alpha$Zr과 같은 hexagonal 결정구조를 갖는다. Rod 형태 및 사각형태의 석출물은 tetragonal 결정구조를 갖는 SiZr$_2$ 상이다.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.683-688
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• 2013

Candidate coating materials for re-usable metallic nuclear fuel crucibles, TaC, TiC, ZrC, $ZrO_2$, and $Y_2O_3$, were plasmasprayed onto a niobium substrate. The microstructure of the plasma-sprayed coatings and thermal cycling behavior were characterized, and U-Zr melt interaction studies were carried out. The TaC and $Y_2O_3$ coating layers had a uniform thickness, and high density with only a few small closed pores showing good consolidation, while the ZrC, TiC, and $ZrO_2$ coatings were not well consolidated with a considerable amount of porosity. Thermal cycling tests showed that the adhesion of the TiC, ZrC, and $ZrO_2$ coating layers with niobium was relatively weak compared to the TaC and $Y_2O_3$ coatings. The TaC and $Y_2O_3$ coatings had better cycling characteristics with no interconnected cracks. In the interaction studies, ZrC and $ZrO_2$ coated rods showed significant degradations after exposure to U-10 wt.% Zr melt at $1600^{\circ}C$ for 15 min., but TaC, TiC, and $Y_2O_3$ coatings showed good compatibility with U-Zr melt.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.207-213
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• 1997

The computational scheme on a irradiation temperature of U-10Zr fuel was established considering porosity formation, bond sodium infiltration and constituent redistribution. Thermotransport theory was adapted to model the redistribution phenomenon. As a results, the bond sodium seems to be logged in the outer region of fuel slug. The main driving force for constituent redistribution appears to be the Zr solubility change along to radial position of the fuel. It is evident that the heat of transport also has some contribution to the redistribution.