• Journal of the Korean Ceramic Society
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• v.50 no.2
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• pp.122-126
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• 2013

The surface roughness of the inner and outer surfaces of a tube is an important requirement for nuclear fuel cladding. When an inner SiC clad tube, which is considered as an advanced Pressurized Water Cooled Reactor (PWR) clad with a three-layered structure, is fabricated by Chemical Vapor Deposition (CVD), the surface roughness of the substrate, graphite, is an important process parameter. The surface character of the graphite substrate could directly affect the roughness of the inner surface of SiC deposits, which is in contact with a substrate. To evaluate the effects of the surface roughness changes of a substrate, SiC deposits were fabricated using different types of graphite substrates prepared by the following four polishing paths and heat-treatment for purification: (1) polishing with #220 abrasive paper (PP) without heat treatment (HT), (2) polishing with #220 PP with HT, (3) #2400 PP without HT, (4) polishing with #2400 PP with HT. The average surface roughnesses (Ra) of each deposited SiC layer are 4.273, 6.599, 3.069, and $6.401{\mu}m$, respectively. In the low pressure SiC CVD process with a graphite substrate, the removal of graphite particles on the graphite surface during the purification and the temperature increasing process for CVD seemed to affect the surface roughness of SiC deposits. For the lower surface roughness of the as-deposited interlayer of SiC on the graphite substrate, the fine controlled processing with the completed removal of rough scratches and cleaning at each polishing and heat treating step was important.

• Journal of Mechanical Science and Technology
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• v.15 no.9
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• pp.1274-1280
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• 2001

The fretting wear behavior of the contact between Zircaloy-4 tube and Inconel 600, which are used as the fuel rod cladding and grid, respectively, in PWR nuclear power plants was investigated in air. In the study, number of cycles, slip amplitude and normal load were selected as the main factors of fretting wear. The results indicated that wear increased with load, slip amplitude and number of cycles but was affected mainly by the slip amplitude. SEM micrographs revealed the characteristics of fretting wear features on the surface of the specimens such as stick, partial slip and gross slip which depended on the slip amplitude. It was found that fretting wear was caused by the crack generation along the stick-slip boundaries due to the accumulation of plastic flow at small slip amplitudes and by abrasive wear in the entire contact area at high slip amplitudes.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• Resources Recycling
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• v.21 no.5
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• pp.18-30
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• 2012

Zirconium is one of the most important material used as cladding of fuel rods in nuclear reactors because of its high dimensional stability, good corrosion resistance and especially low neutron-absorbing cross section. However, Hf free nuclear grade Zr sponge is commercially produced by only three countries including USA, France and Russia. So, Zr has been thoroughly managed as a national strategic material in Korea. Most of the zirconium is used for Korean nuclear industry as nuclear fuel cladding materials manufactured from Hf free Zr alloy raw material. Also, there are some other applications such as alloying element and detonator. In this review, zirconium production and recycling technologies have been reviewed and current industrial status was also analyzed. And recent achievements in innovative reduction technologies such as electrolytic reduction process and molten oxide electrolysis were also introduced.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.153-162
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• 1980

The investigation of the fuel cladding temperature behavior and heat transfer mechanism during the reflooding phase of a LOCA plays an important role in performance evaluation of ECCS and safety analysis of water reactors. Reflooding experiments were performed with horizontal and vertical flow channels to investigate the effect of coolant flow channel orientation on rewetting process. Emphasis was mainly placed on the CANDU reactor which has horizontal pressure tubes in core, and the results were compared with those of vertical channel. Also to investigate the rewetting process visually, the experiments by using a rod in annulus and a quartz tube heated outside were performed. It can be concluded that the rewetting velocity in horizontal flow channel is clearly affected by flow stratification, however, the average rewetting velocity is similar to those in vertical flow channel for same conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.11-19
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• 2015

The production of nuclear fuel cladding tube is expected to increase with the nuclear power plant expansion. Zirconium(Zr) scrap that is generated during manufacturing is also expected to increase. Zr electrorefining experiment was carried out in the fluoride salt of LiF-KF-ZrF₄ using multiple electrode for scale up and improving throughput Zr electrorefiner develop-ment. The Zr reduction peak observed at-0.8 V(vs.Ni). Polarization behavior showed that the amount of applied current increases because of decreasing cell resistance as the number of cathode increases. Experimental results showed the highest recovery rate about 98% at lowest current density of 25.64 mA/cm² using 6 electrodes. XRD and TG analysis result show that pure Zr was recovered 99.92% and ICP analysis shows that lower impurity content than conventional impurity content of the Anode(97.8%). Electrorefining consumes energy about 7.15 kWh/kg less than 39.7% compared to the Kroll process using 6 electrode width of 20 mm and height of 65 mm. Because of increasing cell efficiency and recovery rate, using multiple cathode is determined as an efficient technique for scale up electrorefining Zr scrap.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.7 no.2
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• pp.113-119
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• 2006

Brazing is an important manufacturing process in the fabrication of Heavy Water Reactor fuel rods, in which bearing and spacer pads are joined to Zircaloy-4 cladding tubes. The physical vapor deposition(PVD) technique is currently used to deposit metallic Be on the surfaces of pads as a filler metal. Amorphous Zr-Be binary alloys which are manufactured by rapid solidification process are under developing to substitute the conventional PVD-Be coating. In the present study, brazed joint with PVD and amorphous alloys of $Zr_{1-x}Be_{x}(0.3{\le}x{\le}0.5)$ as filler metals are compared by mechanism, microstructure and hardness. The thickness of brazed joint with amorphous alloys became much smaller than that of PVD-Be. The erosion of base metal did not occur in the brazed joint with amorphous alloys. The brazing mechanism for PVD-Be seems to be Be diffusion into Zr-4 with capillary action resulting from eutectic reaction while that for amorphous alloys are associated with the liquid phase formation in the brazed joint. The brazed joint microstructure with PVD-Be consists of dendrite while that with amorphous alloys is globular. The $Zr_{0.7}Be_{0.3}$ alloy shows the smooth interface with little erosion in the base metal and is recommended a most suitable brazing filler metal for Zircaloy-4.

• Nuclear Engineering and Technology
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• v.21 no.2
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• pp.111-122
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• 1989

With the Zircaloy-4 tube oxidized in high temperature (1323 K) steam for 5, 10, 30 and 60 minutes, the expanding copper mandrel test was carried out over a temperature range of 673-l173k at $\varepsilon\;=\;3.0\times10^5S\;^1$. The oxidation parameters $(K_i)$ in the present study were linearly proportional to square root of time $(Ki= \delta_{kit})$ and their rate constants ($\delta_{ki}$) are 0.281, 2.82, and 2.313 for weight gain and thickness of Zr02 and $\alpha$(0) layer, respectively. Activation energy for high temperature (873-1073k) plastic deformation of Zircaloy-4 increases from 251 KJ/mol to 323 KJ/mol with increase in oxidation time from 5 minutes to 60 minutes due to the high strengthened Zr02. With the oxide layer thickness [K ; expressed in "Equivalent Cladding Reacted" (ECR,%)] and the yield stress obtained from the mandrel test, an empirical relation was derived as ($\sigma/C)^n=K^mexp$ (Q/RT) with n=6.9, m=5.7, C=0.155, 0.138, 0.051, and 0.046 MPa for Q=251, 258, 316, 323 KJ/mol, respectively.

• Resources Recycling
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• v.26 no.6
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• pp.29-37
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• 2017

Nuclear fuel cladding tube is fabricated by pilgering and annealing process. In order to remove impurity and oxygen layer on the surface, pickling process is carried out. When Zirconium(Zr) is dissolved and saturated in acid solution during the pickling process, all the waste acid including Zr is disposed. Therefore, $BaF_2$ is added into the waste acid to extract Zr and $Ba_2ZrF_8$ is subsequently formed. To recycle Zr by electrowinning process, $Ba_2ZrF_8$ is used as electrolyte, but it has high melting point ($1053^{\circ}C$). $ZrF_4$ should be added into $Ba_2ZrF_8$ to decrease the melting point. In this paper, it was investigated that $Ba_2ZrF_8$ was separated to $BaF_2$ and $ZrF_4$ by vacuum distillation. Firstly, $BaF_2$ with different particle size ($1{\mu}m$, $35{\mu}m$, $110{\mu}m$) was added into the waste acid and the respective precipitation property was estimated. $BaF_2$ obtained by vacuum distillation was shattered by ball-milling with different time. The precipitation efficiency was compared with $1{\mu}m$ of ${BaF_2}^{\prime}s$ one, which was not used as precipitation agent.

• Korean Journal of Materials Research
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• v.26 no.12
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• pp.681-687
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• 2016

Zirconium(Zr) nuclear fuel cladding tubes are made using a three-time pilgering and annealing process. In order to remove the oxidized layer and impurities on the surface of the tube, a pickling process is required. Zr is dissolved in HF and $HNO_3$ mixed acid during the process and pickling waste acid, including dissolved Zr, is totally discarded after being neutralized. In this study, the waste acid was recycled by adding $BaF_2$, which reacted with the Zr ion involved in the waste acid; $Ba_2ZrF_8$ was subsequently precipitated due to its low solubility in water. It is very difficult to extract zirconium from the as-recovered $Ba_2ZrF_8$ because its melting temperature is $1031^{\circ}C$. Hence, we tried to recover Zr using an electrowinning process with a low temperature molten salt compound that was fabricated by adding $ZrF_4$ to $Ba_2ZrF_8$ to decrease the melting point. Change of the Zr redox potential was observed using cyclic voltammetry; the voltage change of the cell was observed by polarization and chronopotentiometry. The structure of the electrodeposited Zr was analyzed and the electrodeposition characteristics were also evaluated.