• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.11 no.1
• /
• pp.55-61
• /
• 2013

The recovery of Zr from Zircaloy-4 (Zry-4) cladding hulls using a chlorination method was demonstrated for complete conversion of Zr into $ZrCl_4$. A chlorination reaction was performed by reacting Zry-4 hulls for 8 h under a 70 cc/min $Cl_2$ + 70 cc/min Ar flow at $380^{\circ}C$. The initial weight of the reactant (51.7 g) decreased to 0.49 g after 8 h of operation, which is only 0.95wt% of the initial weight. The weight of the total reaction products was 121.7 g with a high Zr purity of 99.80wt%. Fe and Sn were identified as major (0.18wt%) and minor (0.02wt%) impurities of the reaction products, respectively. It was also shown that Zr exhibited a high recovery ratio of 96.95wt% with a relatively small experimental loss of 2.34wt%. Observation of the reaction residues revealed that the chlorination reaction was dominant along the longitudinal direction, and surface oxide layers remained as reaction residues. The high purity and recovery ratio of Zr proposed the feasibility of the chlorination technique as an effective hull waste treatment method.

• Nuclear Engineering and Technology
• /
• v.52 no.4
• /
• pp.742-754
• /
• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

• Nuclear Engineering and Technology
• /
• v.52 no.3
• /
• pp.508-519
• /
• 2020

During transients or accidents, the reactor core is uncovered, and droplets entrained above the quench front collides with the uncovered fuel rod surface. Droplet impact cooling can reduce the peak cladding temperature. Besides zirconium-based cladding, versatile accidental tolerant fuel (ATF) claddings, including FeCrAl, have been proposed to increase the accident coping time. In order to investigate the effect of surface properties on droplet impact cooling of cladding surfaces, the droplet impact phenomena are photographed on the FeCrAl and zircaloy-4 (Zr-4) surfaces under different conditions. On the oxidized FeCrAl surface, the Leidenfrost phenomenon is not observed even when the surface temperature is as high as 550 ℃ with We > 30. Comparison of the impact behaviors observed on different materials shows that nucleate and transition boiling is more intensive on surfaces with larger thermal conductivity. The Leidenfrost point temperature (LPT) decreases with the solid thermal effusivity (${\sqrt{k{\rho}C_p}}$). However, the CHF temperature is relatively insensitive to the surface oxidation and Weber number. Droplet spreading diameter is analyzed quantitatively in the film boiling stage. Based on the energy balance a correlation is proposed for droplet maximum spreading factor. A mechanistic model is also developed for the LPT based on homogeneous nucleation theory.

• Nuclear Engineering and Technology
• /
• v.20 no.1
• /
• pp.54-64
• /
• 1988

This paper assesses the models in the SCDAP code using the results of the FLHT-2 test. Calculations show that the SCDAP correctly predicts Ire temperatures, oxidation front movement, overall hydrogen generation and peak generation rate, internal fuel rod pressure, and cladding rupture due to ballooning. A comparison of the calculated results with measured data shows that two phase level is underpredicted, and that radiation heat transfer and auto-catalytic reaction temperature of zircaloy are overpredicted. These models are recommended to be modified. The analysis also shows that the simulation of the gap in a fuel rod improves the code prediction on core damage progression.

• Journal of the Korean Society for Heat Treatment
• /
• v.10 no.3
• /
• pp.198-208
• /
• 1997

The-steady-state creep mechanism and behavior of Zircaloy-4 used as cladding materials in PWR have been investigated in air environment over the temp, ranges from 600 to $645^{\circ}C$ and stress ranges from 4 to $7kg/mm^2$. The stress exponents for the creep deformation of this alloy, n were decreased 4.81, 4.71, 4.64, and 4.56 at 600, 615, 630 and $645^{\circ}C$, respectively; the stress exponents decreased with increasing the temperature and got closer to about 5. The apparent activation energies, Q, were 62.1, 60.0, 57.9 and 55.4 kcal/mole at stresses of 4, 5, 6, $7kg/mm^2$, respectively; the activation energies decreased with increasing the stress and were close to those of volume self diffusion of Zr in Zr-Sn-Fe-Cr system. In results, it can be considered that the creep deformation for Zircaloy-4 was controlled by the dislocation climb over the ranges of this experimental conditions. Larson-Miller parameter, P, for the crept specimens was obtained as P=(T+460)(logt,+23). The failure plane observed by SEM slightly showed up intergranular fracture at this experiment ranges. However, it was essentially dominated by the dimple phenomenon, which was a characteristics of the transgranular fracture.

• Proceedings of the Korean Vacuum Society Conference
• /
• 1995.06a
• /
• pp.151-151
• /
• 1995

Nitrogen implantation process has been applied for improvement of wear resistance of Z Zircaloy-4 fuel cladding materials. Nitrogen was implanted at 120 ke V to a total do range of 1xHP ions/cm2 to 8xlO17 ions/cm2 at various temperatures of 298"C to 676"C. The m microstructure changes by nitrogen implantation were analyzed by using TEM, XRD 뻐d A AES, cmd then wear behavior was evaluated by ball-on-disc wear testings at various loads a and sliding velocity under unlubricated condition. Nitrogen implantation produced ZrNx nitride above 4.37x1017 ions!cm2 as well as heavy d dislocations, which enhanced microhardness of the implanted surface of up to 900 Hk from 2 200 Hk of unimplanted substrate. Hardness was also found to be increased with increasing i implantation temperature and enhanced up to OOOHk at 620 "C. the wear resistance was g greatly improved with increasing total ion do않 as well as implantation temperature. The effective enhancement of wear resistance at high dose and tem야ratures is believed d due to significant hardening associated with high degree of precipitation of Zr nitrides and g generation of prismatic dislocation I$\infty$ps.infty$ps.

• Nuclear Engineering and Technology
• /
• v.50 no.2
• /
• pp.237-245
• /
• 2018

Potential air ingress scenarios during accidents in nuclear reactors or spent fuel pools have raised the question of the influence of air, especially of nitrogen, on the oxidation of zirconium alloys, which are used as fuel cladding tubes and other structure materials. In this context, the reaction of zirconium with nitrogen-containing atmospheres and the formation of zirconium nitride play an important role in understanding the oxidation mechanism. This article presents the results of analysis of the interaction of the oxygen-preloaded niobium-bearing alloy $M5^{(R)}$ with nitrogen over a wide range of temperatures ($800-1400^{\circ}C$) and oxygen contents in the metal alloy (1-7 wt.%). A strongly increasing nitriding rate with rising oxygen content in the metal was found. The highest reaction rates were measured for the saturated ${\alpha}-Zr(O)$, as it exists at the metal-oxide interface, at $1300^{\circ}C$. The temperature maximum of the reaction rate was approximately 100 K higher than for Zircaloy-4, already investigated in a previous study. The article presents results of thermogravimetric experiments as well as posttest examinations by optical microscopy, scanning electron microscopy (SEM), and microprobe elemental analyses. Furthermore, a comparison with results obtained with Zircaloy-4 will be made.

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

• Nuclear Engineering and Technology
• /
• v.19 no.2
• /
• pp.122-129
• /
• 1987

Experiments and numerical analysis have been performed to investigate the effect of preoxidation by oxidizing Zircaloy-4 specimens at a higher temperature after a period of exposure at a lower temperature. The oxidation experiments were performed between $700^{\circ}C$ and 85$0^{\circ}C$ after Preoxidation at $650^{\circ}C$ in a steam and water mixture for 600 seconds and 1,800 seconds. As the thickness of preoxidized layer increased, the oxidation rate of preoxidized specimens at higher temperature became lower than that of as-received claddings. A transition region of oxidation rate exist in the preoxidized specimens, and the region disappeared rapidly as the oxidation temperature increased. This effect appeared more clearly at lower temperatures. According to the results of numerical analysis performed in this study, the growth rate of oxide layer thickness and weight gains were similarly affected by the thickness of preoxidized layer.

• Nuclear Engineering and Technology
• /
• v.41 no.2
• /
• pp.187-198
• /
• 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.