• Journal of Hydrogen and New Energy
• /
• v.24 no.4
• /
• pp.295-301
• /
• 2013

Long-term global energy-demand growth is expected to increase driven by strong energy-demand growth from developing countries. Fusion power offers the prospect of an almost inexhaustible source of energy for future generations, even though it also presents so far insurmountable scientific and engineering challenges. One of the challenges is safe handling of hydrogen isotopes. Metal hydrides such as depleted uranium hydride or ZrCo hydride are used as a storage medium for hydrogen isotopes reversibly. The metal hydrides bind with hydrogen very strongly. In this paper, we carried out a modeling and simulation work for absorption/desorption of hydrogen by ZrCo in a horizontal annulus cylinder bed. A comprehensive mathematical description of a metal hydride hydrogen storage vessel was developed. This model was calibrated against experimental data obtained from our experimental system containing ZrCo metal hydride. The model was capable of predicting the performance of the bed for not only both the storage and delivery processes but also heat transfer operations. This model should thus be very useful for the design and development of the next generation of metal hydride hydrogen isotope storage systems.

• Korean Journal of Metals and Materials
• /
• v.47 no.8
• /
• pp.474-481
• /
• 2009

The effects of final heat treatment on the mechanical and corrosion properties of a Zr alloy strip incorporating Nb were investigated. The chemical composition of the strip was Zr-1.49Nb-0.38Sn-0.20Fe-0.11Cr, and strip specimens were subjected to final heat treatment in a temperature range of $580{\sim}700^{\circ}C$. Tensile tests at room temperature and $316^{\circ}C$, along with corrosion tests in a simulated PWR loop and a 70 ppm LiOH solution environment at $360^{\circ}C$, were performed on the specimens. The mechanical properties of the strip were saturated when the specimens received final heat treatment at an elevated temperature of more than $640^{\circ}C$. However, the corrosion resistance of the strip in the simulated PWR loop and in the 70 ppm LiOH solution environment was improved with a decrease of the final annealing temperature. It is recommended that the alloy strip be finally heat-treated at a temperature of less than $620^{\circ}C$ for longer than 10 minutes in order to obtain fully recrystallized microstructures, and thereby attain enlarged tensile elongation, and to prevent the precipitation of ${\beta}-Zr$, which is known to deteriorate the corrosion resistance.

• Nuclear Engineering and Technology
• /
• v.42 no.2
• /
• pp.193-202
• /
• 2010

In order to study the cladding properties of zirconium after a loss-of-coolant accident (LOCA)-simulation oxidation and water quenching test, commercial Zircaloy-4 and two kinds of HANA claddings were oxidized at temperatures ranging from $900^{\circ}C$ to $1250^{\circ}C$ and exposed for 300 s, and then cooled to $700^{\circ}C$ before quenching. Microstructural observations were made to evaluate the matrix characteristics with the chemical compositions after the LOCA-simulation test. Ring compression testing was then performed to compare the ductile behaviour of the HANA and Zircaloy-4 claddings. An X-ray diffraction (XRD) analysis was carried out for temperatures ranging from room temperature to $1250^{\circ}C$ for the oxide layer to verify the oxide crystal structure at each oxidation temperature.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2003.11a
• /
• pp.421-425
• /
• 2003

The reductive extraction process is an important step to refine the TRU product from the electrorefining process for the preparation of transmutation reactor fuel. In this study, it was studied on the reductive extraction between the eutectic salt and Bi metal phases. The solutes were zirconium and the rare earth elements, where zirconium was used as a surrogate for the transuranic(TRU) elements. All the experiments were performed in a glove box filled with a argon gas. Li-Bi alloy was used as a reducing agent to reduce the high chemical activity of Li. The reductive extraction characteristics were examined using ICP, XRD and EPMA analysis. The reduction reaction was equilibrated within 3 hours after the Li addition. Three eutectic salt systems were compared and Zr was successfully separated from the rare earth elements in all the three salt systems.

• Journal of Surface Science and Engineering
• /
• v.36 no.3
• /
• pp.251-255
• /
• 2003

With increasing environmental demands in surface treatment of steel sheets, the passivation layers containing hexavalent chromium $(Cr^{+6})$ are being replaced by non-chromium or trivalent chromium compounds. After review on the various types of inorganic compounds, the zirconates was chosen as the candidate for alternative to sodium dichromate in the aspect of its barrier properties with excellent adhesion to organics. The ammonium zirconium carbonate (AZC) and sodium hexafluorozirconate (SFZ) could be reach $70-80\%$ level of CDC (cathodic dichromate) treatment by their single applications. But high porosity in the AZC layer and poor electrical conductivity of SFZ solution limit the single application of zirconate. Mixed composition of zirconates to compensate their inferiorities or incorporation of organic compounds to seal the porosity seems to be inevitable to match up the target level of Cr-free passivation of tinplate.

• Journal of Hydrogen and New Energy
• /
• v.15 no.4
• /
• pp.266-273
• /
• 2004

The objective of this study is an understanding of stress corrosion cracking of metals that is recognized to mostly limit the lifetime of the structural materials by comparing the features of delayed hydride cracking of zirconium alloys with those of stress corrosion cracking (SCC) of Ni-based alloys and hydrogen cracking of stainless steels. To this end, we investigated a dependence of delayed hydride cracking (DHC) velocity on the applied stress intensity factor and yield strength, and correlated a temperature dependence of the striation spacing and the DHC velocity. We reviewed a similarity of the features between the DHC of zirconium alloys, the SCC of Ni-based alloys and turbine rotor steels, and the hydrogen cracking of stainless steels and discussed the SCC phenomenon in metals with our DHC mode.

• Nuclear Engineering and Technology
• /
• v.46 no.5
• /
• pp.681-688
• /
• 2014

Hydride reorientation behaviors of PWR cladding tubes under typical interim dry storage conditions were investigated with the use of as-received 250 and 485ppm hydrogen-charged Zr-Nb alloy cladding tubes. In order to evaluate the effect of typical cool-down processes on the radial hydride precipitation, two terminal heat-up temperatures of 300 and $400^{\circ}C$, as well as two terminal cool-down temperatures of 200 and $300^{\circ}C$, were considered. In addition, two cooling rates of 2.5 and $8.0^{\circ}C/min$ during the cool-down processes were taken into account along with zero stress or a tensile hoop stress of 150MPa. It was found that the 250ppm hydrogen-charged specimen experiencing the higher terminal heat-up temperature and the lower terminal cool-down temperature generated the highest number of radial hydrides during the cool-down process under 150MPa hoop tensile stress, which may be explained by terminal solid hydrogen solubilities for precipitation, and dissolution and remaining circumferential hydrides at the terminal heat-up temperatures. In addition, the slower cool-down rate generates the larger number of radial hydrides due to a cooling rate-dependent, longer residence time at a relatively high temperature that can accelerate the radial hydride nucleation and growth.

• Korean Journal of Materials Research
• /
• v.29 no.8
• /
• pp.483-490
• /
• 2019

Currently, the Korean nuclear industry uses ZIRLO as material for nuclear fuel cladding(zirconium alloy). KEPCO Nuclear Fuel is in the process of developing a HANA alloy to enable domestic production of cladding. Cladding manufacture involves multistage heat treatments and pickling processes, the latter of which is vital for the removal of defects and impurities on the cladding surface. SMUT that forms on the cladding surface during such pickling process is a source of surface defects during heat treatment and post-treatment processes if not removed. This study analyzes ZIRLO, HANA-4, and HANA-6 alloy claddings to extensively study the SEM/EDS, XRD, and particle size characteristics of SMUT, which are second phase particles that are formed on the cladding surface during pickling processes. Using the analysis results, this study observes SMUT formation characteristics according to Nb concentration in Zr alloys during the washing process following the pickling process. In addition, this study observes SMUT removal characteristics on cladding surfaces according to concentrations of nitric acid and hydrofluoric acid in the acid solution.

• Korean Journal of Materials Research
• /
• v.8 no.6
• /
• pp.513-519
• /
• 1998

Zr-Si binary alloys containing 0.01 to O.lwt.%Si were prepared to investigate the effect of Si on the corrosion behavior of Zr. Corrosion test was performed in pure water at 36$0^{\circ}C$ under a pressure of 2660psi for 100days. The alloys containing 0.01 wt. % and 0.05wt. %Si had the black and uniform oxide film and didn't show the transition of corrosion rate. However. the alloys containing O.lwt.%Si had white oxide film and showed the trasition of corrosion rate at 70 days corrosion test. The weight gain increased with the increasing Si content from 0.01 to 0.1 wt.%. The variation of Si contents had no effect on changing the oxide structure but had significant effect on the electrical resistivity of oxide. The electrical resistivity decreased with increasing Si content. The fraction of precipitates in the Zr-Si binary alloys. identified as tetragonal $Zr_{3}$Si increased with increasing Si content. The increase of the volume fraction of precipitates is thought to be responsible for the increase of weight gain due to short circuit effect of precipitate.

• Proceedings of the Korea Association of Crystal Growth Conference
• /
• 1996.06b
• /
• pp.55-85
• /
• 1996

지르코니아 분말은 ZrO2 결정상이 온도변화에 따라 부피변화를 수반하는 상전이변태를 나타낸다. 단사정 ZrO2가 110$0^{\circ}C$에서는 정방정으로, 2$700^{\circ}C$ 내외에서는 입방정으로 결정구조가 가역적으로 변한다. 이 ZrO2에 금속산화물을 고용시키면 형석 (CaF2:Florite)형의 입방정 결정구조가 실온에서도 안정하게 존재하게 된다. 안정화제 산화물은 caO, MgO등 2가 산화물외에 3가 또는 4가의 금속산화물로서 Sc2O3, Y2O3, Sm2O3, Nd2O3, Gd2O3, Y2O3, CeO2 등이며 이들은 금속이온의 원자가가 변하기 쉬운 희토류 산화물이다. 안정화 지르코니아는 형석형 결정구조이며 결정화학적으로 보면 금속양이온이 산소이온에 대해서 정육면체형의 8배위를 하고 있다. 이때 이온반경비(양이온/음이온)에 따라 Zr+4자리와 O-2자리의 격자위치와 모양이 형성되므로 비틀어진 정육면체구조이건 이상적인 정육면체 형석구조를 이룬다. 이는 지르코니아의 결정상의 2상-3상인 부분안정화 지르코니아다결정체(PSZ : partially stabilized zirconia)이거나 단일상-2상인 정방정 지르코니아다결정체(TZP : tetragonal zirconia polycrystal)의 결정구조를 가지는데 기인한다. PSZ는 주로 MgO, CaO를 안정화제로 고용시켜 입방정 영역에서 소결하고 이를 다시 입방정과 정방정의 상 영역에서 열처리하여 입방정 입자내부에 정방정을 석출 형성시킨 것이며 TZP는 Y2O3 및 CeO2를 고용시켜 PSZ와 다르게 일반적인 상압소결한 정방정 결정상의 미립자이다. 산화지르코늄 분말은 지르콘사에서 열분해시킨 지르코늄소결.융해괴(caustic fusion clinker)를 산처리하여얻어진 지르코늄산용액(zirconyl acid solution : cloride, sulfide, nitride 등)으로부터 제조된다. 고순도 산화지르코늄은 용액 결정석출법에 의해 ZrOCl2.8H2O, 5ZrO2.3SO3.15H2O, ZrO(NO3)2.xH2O 등의 지르코늄 수화물만을 재결정화시킨 것으로부터 얻을 수 있으며 이 지르코늄염 수용액으로부터 입자미세구조를 효과적으로 제어하여 산화지르코늄 및 안정화 지르코니아 분말제조가 가능하다. 안정화 지르코니아 분말은 ZrO2와 안정화산화물의 고용을위하여 가열처리를 필요로 하며 일정온도에서 최적상태로 숙성하므로서 2가지 상(phase) 이상의 고용체를 가지게 된다. 안정화 지르코니아 분말은 고용처리온도를 낮추고 효과적으로 생성시키기 위해서는 지르코늄 및 안정화제염을 혼합하고 습식 직접합성하여 저온에서 고용체의 합해진상 영역을 생성시키는 것이다. 이는 지르코니아 원료분말의 미세구조를 제어하므로서 가능하며 이때 화학성분조성과 크기형태가 균일하게 분포된 입자분말을 얻을 수 있다.