Oxidation Of Cladding Research Articles

Degradation behavior of chromium-coated zirconium cladding under 1200 oC steam oxidation according to the coating microstructure

In 2011, the Great East Japan Earthquake and tsunami caused a hydrogen explosion at the Fukushima Daiichi nuclear power plant, which exposed radioactive materials to the atmosphere and had a very negative impact on the nuclear power industry. Since then, efforts have intensified around the world to make nuclear power safer. Accident-tolerant fuel (ATF) is being developed to prevent the rapid oxidation of zirconium cladding, which directly causes hydrogen explosions. ATF research can be divided into two main approaches: changing the cladding material, or coating the surface of the cladding. Coatings are easier to commercialize and apply to existing nuclear power plants, so most vendors have focused on this approach. Chromium is a popular coating medium because of its superior properties such as a low oxidation rate and excellent adhesion. Therefore, research has been focused on suppressing the rapid oxidation of zirconium cladding in the event of an accident by adding a chromium coating with an appropriate thickness in terms of both economy and effectiveness. Even if the thickness of the coating is fixed, the penetration of oxidizing substances can be further delayed by improving the microstructure of the chromium coating, such as by reducing the grain boundary area. In this study, chromium-coated zirconium cladding tubes were fabricated by the arc ion plating process. The microstructure of the chromium coating was adjusted by varying the negative voltage (0–125 V), which in turn controlled the incident energy at which the chromium ionic particles hit the surface of the cladding tube. Experiments were then performed in 1200 °C steam environment to determine the optimal microstructure for high-temperature oxidation resistance. The development of various material degradation phenomena that occurred during 1200 °C steam oxidation was observed to identify the oxidation mechanism and the main factors of the microstructure that affect the zirconium oxidation rate.

Time in DNB experimental study on Cr coated zircaloy cladding

Since the Fukushima-Daiichi accident effort has been put in by the nuclear community to develop Accident Tolerant Fuel (ATF) cladding materials to reduce the risk of clad oxidation and, consequently, hydrogen generation and potential major accidents. Among designed materials, the use of a chromium coating on top of zirconium-alloy cladding showed reasonable results in preliminary tests, being a favorable cladding solution for Pressurized Water Reactors. This paper provides highly detailed temperature data obtained during time in departure from nucleate boiling (DNB) experiments under prototypical reactor conditions of temperature, pressure, mass flux, and heat flux profile on a Cr-coated zirconium-alloy cladded heater rod simulating the fuel pin. Data were obtained at mass fluxes from 1300 to 1700 kg/m2s, pressure of 16 MPa, and inlet temperature of 333 °C (subcooling degree of 14.4 °C). The change in dry time was obtained by increasing the spike in wall temperature necessary to shut down the power. These tests showed that increasing the spike in temperature to characterize the critical heat flux from 50 °C to 400 °C resulted in variations of the departure from nucleate boiling measured value within the uncertainty range of their measurements, hence, validating the use of a spike of 50 °C, a commonly used value in this kind of study, as an indicator of the critical heat flux. Within the dry region, the temperature was found to increase along the direction of the flow for a very short length, reaching a peak, from which it reduces linearly until the surface reaches a temperature that is no longer able to sustain a dry patch. It was noticed that the dry region can grow against the flow direction due to the cosine profile of the heat flux, and that small dryout events can occur without leading to departure from nucleate boiling, i.e., being fully suppressed. However, two consecutive events may sum up and cause the critical heat flux. Rewet time was found to increase with peak cladding temperature and decrease with mass flux. A maximum peak cladding temperature of 844 °C was found, with the surface remaining dry for approximately 14.1 s. The minimum film boiling temperature was measured during the rewet process, showing absolute values of around 580 °C, and difference between wall and inlet temperatures of about 240 °C. A post-inspection using microscope and X-ray images showed no damage to the cladding and heater structure, differently than what was noticed for the bare zircaloy in a previous study. This indicated that the presence of chromium coated cladding can improve the oxidation resistance of the cladding not only in short critical heat flux transients but also in longer events associated with major accidents, like a Loss of Coolant Accident.

Mixing of Hydrogen-Steam Buoyancy Jets Released into the Upper Atmosphere of the Containment Building

This study evaluates the spatial dilution of hydrogen concentration caused by steam-hydrogen buoyancy jets rising through the open top of the steam generator compartment during a loss-of-coolant accident in the OPR1000, the Korean Standard Nuclear Power Plant. The correlation of the concentration decay rate in the plume with relatively high buoyant flux was applied to estimate the hydrogen concentration in the rise distance of the buoyant jet. The MELCOR code was used to calculate the gas composition and discharge flow rate in the ruptured cold leg during the rapid cladding oxidation to determine the volume and buoyant fluxes that affect the mixing behavior. The concentration decay rate at the plume’s center decreases as the steam-hydrogen binary buoyant jet rises. Despite the assumed initial volume flux and simplified jet nozzle geometry, the decay rate correlation can assess conservatively the diluted hydrogen in a severe accident.

Analysis of FeCrAl cladding performance under loss-of-coolant accident conditions

Modeling developments and simulations of iron-chromium-aluminum (FeCrAl) cladding behavior under loss-of-coolant accident (LOCA) conditions have been performed using the BISON fuel performance code. First, we describe the modeling capabilities in BISON relevant to LOCA conditions, with an emphasis on new models added to the code in this work. The latter include updated models for FeCrAl cladding plasticity, burst failure and high-temperature oxidation. Subsequently, we describe BISON simulations of both FeCrAl and Zircaloy cladded fuel rods under the conditions of a postulated large-break LOCA (LB-LOCA) scenario in a BWR. The results indicate that the FeCrAl cladding will experience significantly lower ballooning, oxidation, and hydrogen production rates than Zircaloy cladding under the same conditions. However, the calculated burst time and temperature were similar for the Zircaloy and FeCrAl claddings. Finally, we perform extended simulations of fuel rod behavior for the conditions of a prolonged unmitigated LB-LOCA, including consideration of energy deposition due to the exothermic cladding oxidation reaction. The results point to a significantly reduced oxide thickness and hydrogen gas production for FeCrAl cladding compared to Zircaloy cladding.

Steam oxidation of Cr-coated zirconium alloy claddings at 1200 °C: Kinetics transition and failure mechanism of Cr coatings

The microscopic layered structural evolution, the diffusion behavior of O, Cr and Zr elements and oxidation kinetics of the Cr-coated zirconium alloy claddings during steam oxidation at 1200 °C were systematically studied. At the initial oxidation stage (within 1 h oxidation), the protective Cr2O3 layer formed on the Cr coating, and the interdiffusion between Cr and Zr resulted in a four-layers structure from outer to inner, Cr2O3, Cr, Cr-Zr diffusion layer and Zr substrate. At this stage, the oxidation kinetics followed parabolic law. With increasing oxidation time to 1.8 h, ZrO2 networks formed in the Cr coating due to the occurrence of the redox reaction between Zr and Cr2O3, which provided a pathway for the diffusion of oxygen toward substrate. The layered structure on the Cr coating side transformed from a four-layers to a five-layers structure with the formation of a ZrO2 layer in the substrate. Meanwhile, the oxidation kinetics became linear growth. The failure of Cr coatings resulting in the oxidation of Zr substrate accounts for the parabolic-to-linear kinetics transition.

Implementation of chrome oxidation model into best-estimate system code for accident tolerant fuel

A chrome oxidation model was developed/implemented into the MARS-KS 1.5 code to improve the prediction of accident tolerant fuel with chrome-coated cladding. Initially, a multi-layer cladding model was implemented, followed by investigation of the peak cladding temperature (PCT) during large break loss-of-coolant accidents with varying coating thicknesses. Results showed that increasing the coating thickness led to a decrease in blowdown PCT, while reflood PCT did not show a clear tendency. The chrome-coated cladding oxidation model was then developed/implemented, and validated by comparing weight gain due to oxidation using experimental data. The Parisi et al. (2020) model showed substantially lower weight gain, while the Wu et al. (2020) model predicted relatively well against the experimental data. However, there were still some scatters in oxidation data, indicating the need for developing a specific oxidation model for each Cr-coated cladding candidate for better prediction.

Eutectic reaction and oxidation behavior of Cr-coated Zircaloy-4 accident-tolerant fuel cladding under various heating rates

The Cr-coated zirconium alloy claddings are near-term approach for accident tolerant fuel (ATF) claddings. The performance of magnetron-sputtered Cr-coated zirconium alloys is investigated under conditions simulating severe accidents. The transient oxidation experiments were performed using a thermogravimetric analyzer (TGA) with five different heating rates (2–50 K/min) up to 1380 °C, which is above the Zr-Cr eutectic temperature. The weight gain of the Cr-coated specimens was found to be significantly lower than that of the uncoated reference up to approximately 1330 °C. However, after the occurrence of the Zr-Cr eutectic reaction and failure of the coating at 1330∼1380 °C, the oxidation rate rapidly increased and the weight gain rate of the Cr-coated specimens was 1.5 to 10 times higher than that of the uncoated specimens. The difference was greater at higher heating rates. These results suggest that the Cr-coating has a limited effect in preventing oxidation and degradation of cladding in case of eutectic reaction. The oxidation was faster at higher heating rate, and it may even cause runaway oxidation of cladding due to the temperature rise caused by oxidation heat. These findings are important for understanding the capabilities and degradation behaviors of Cr-coated cladding under severe accidental scenarios.

Qualification Analysis of Experimental Data for High-Temperature Oxidation of Russian Zirconium Alloys in Steam

The paper presents the results of qualification analysis of experimental data on high-temperature oxidation of the Russian zirconium alloys in steam, which include an assessment of the consistency of measurement results in tests, and numerical modeling of experiments with SOCRAT-V1/V2 code. Conclusions are provided on the possibility of using the SSC RIAR experimental programs for validation of severe accident codes, and the applicability of the SOCRAT-V1/V2 code for prediction of the high-temperature oxidation of the Russian zirconium claddings of fuel rods.

Local Disturbance of the Water-Chemical Regime as a Cause of Increased Oxidation of Cladding of VVER-1000 Fuel Elements

Local Disturbance of the Water-Chemical Regime as a Cause of Increased Oxidation of Cladding of VVER-1000 Fuel Elements

On the Thermodynamics of Aluminum Cladding Oxidation: Water as the Catalyst for Spontaneous Combustion

On the Thermodynamics of Aluminum Cladding Oxidation: Water as the Catalyst for Spontaneous Combustion

Hydride embrittlement resistance of Zircaloy-4 and Zr-Nb alloy cladding tubes and its implications on spent fuel management

• Mechanical tests of hydrided Zircaloy-4 and Zr-Nb alloy with Ring Compression Test. • EBSD analyses on texture of reactor-grade hydrided Zircaloy cladding tubes. • Abrupt ductile to brittle transition at 560 and 490 wppm for Zr-4 and Zr-Nb alloy. • Larger grain size of Zr-4 reduces interlink of inter-granular hydrides. • Reduced hydride interlink increases hydride embrittlement resistance. In the spent fuel storage phase, nuclear fuel cladding is subjected to increased embrittlement owing to a large amount of hydride precipitation. This study compares the differences in the hydrogen-induced cladding embrittlement of cold work stress-relief annealed (CWSR or SRA) Zircaloy–4 and Zr-Nb alloy cladding with ring compression test at the temperature of the spent fuel pool, which is approximately 40 °C. Experiments demonstrate that an abrupt ductile to brittle (DTB) transition occurs at the critical hydrogen content of 560 and 490 wppm for Zircaloy-4 and the tested Zr-Nb cladding tubes, respectively. Even beyond the critical hydrogen content, sufficiently high cladding ductility with the offset strain >10% is maintained up to ∼ 90 MWd/kgU for both cladding materials on the rod-average basis. Extensive EBSD analyses coupled to thermodynamic modeling demonstrate that this is primarily due to the slightly larger grain diameter of Zircaloy-4 tube, which reduces the number of available sites for inter-granular hydride precipitation. Reduced inter-granular hydride precipitation prevents the extent of hydride interlink, thereby improving the hydride embrittlement resistance. Nevertheless, the tested Zr-Nb alloy cladding presents an extened discharge burnup limit for abrupt DTB transition owing the reduced in-core cladding oxidation rate. The presented understanding of microstructural effect on hydride interlink and resulting embrittlement may provide a basis for understanding the general hydride embrittlement phenomena of Zircaloy cladding which include, but not limited to, wet storage, dry storage, and post-accident ductility of high burnup Zircaloy cladding.

Development of oxidation model for zirconium alloy cladding and application in the analysis of cladding behavior under loss of coolant accident

During Loss of Coolant Accident (LOCA) in Pressurized Water Reactor (PWR), the oxidation of zirconium alloy cladding has a significant impact on the core degradation process. However, empirical parabolic rate correlations applied in accident analysis codes only provide oxide thickness obtained by the hypothesis of stoichiometric zirconia, which has limitations for simulating complicated oxide behaviors under accident conditions. The diffusion model with moving boundaries proposed in previous studies can give the distribution of oxygen concentration. However, due to simplified assumptions cladding expansion was not properly considered. In this study, an oxidation model for zirconium alloy cladding considering oxygen diffusion, zirconium conservation and mass density variation with temperature and oxygen concentration is developed. The model is validated by experimental data and shows good agreement. To analyze cladding oxidation during LOCA transient, QUENCH-05 and QUENCH-SR test are simulated. The model can give reasonable simulation results. The simulation results of QUENCH-05 test confirm the fact that β-Zr plays an important role in maintaining the integrity of cladding under quenching process. The simulation results of QUENCH-SR test indicate that steam starvation can still lead to cladding embrittlement due to the consumption of β-Zr matrix and thus increase the risk of cladding failure.

Experimental study on thermo-chemical behavior of PHWR fuel channel under slumped fuel pin condition

The safe operation of the reactor is compromised under postulated accident events. The events may originate from single or multiple failures. LOCA, along with the unavailability of ECCS, is one such multiple failure event. Phenomena like fuel heat up, fuel slumping, clad oxidation, hydrogen generation and fission product release are about to take place. However, moderator cooling to the external surface of fuel channels limits the fuel heat up. The steam starvation in the channel limits hydrogen generation. The purpose of this study is to analyze the thermo-chemical behavior of a fuel channel of PHWR, assuming that the 37 fuel elements of a bundle are collapsed and collected in a tightly packed stack at the bottom of the pressure tube of the fuel channel. The temperature distribution is obtained under a pseudo-steady-state for the various components of the simulated fuel channel exposed to an oxidative environment. Experimental results have indicated that a significant temperature gradient gets established in radial, circumferential and axial directions in the fuel channel due to slumped fuel bundle configuration as well as hydrogen generation. Estimation from the heat balance studies shows that moderator functions as an active heat sink.

Oxidation Resistance and Stress Corrosion Cracking Susceptibility of 308L and 309L Stainless Steel Cladding Layers in Simulated Pressurized Water Reactor Primary Water

Exposure and slow strain rate tensile tests were conducted in a simulated pressurized water reactor (PWR) primary water to investigate the oxidation resistance and stress corrosion cracking (SCC) susceptibility of 308L and 309L stainless steel (SS) cladding layers. A double-layer structure oxide layer grown on 308L SS and 309L SS contained the Cr-enriched nanocrystalline internal layer and the Fe-enriched spinel oxide in the external layer. Ni-enrichment at the matrix/oxide boundary was observed. The internal oxide film on 309L SS was thicker and had a lower Cr content than that on 308L SS. Preferential dissolution of inclusions led to pits on 308L SS and 309L SS surfaces during the exposure tests. More inclusions in 309L would decrease its SCC resistance due to the pits’ ability to act as the SCC initiation site. 308L SS had a lower susceptibility to SCC than 309L SS in PWR primary water. Lower ferrite content and higher strength/hardness reduced the oxidation and SCC resistance of 309L SS cladding. The effect of ferrite on oxidation and SCC of the SS claddings is discussed.

Nuclear fuel rod cladding oxidation and hydrogen production model based on diffusion theory in a multiphase environment of ZrO2, α-Zr(O), and β-Zr at high temperatures (1273 K–1800 K)

This paper proposes a mathematical model for the oxidation process of zirconium under the theory of oxygen diffusion in Zircaloy. The model considers ZrO2, α-Zr(O), and β-Zr phases at high temperatures (1273 K–1800 K) in an equivalent fuel rod. The model also considers the heat transfer phenomenon, the decay heat after shutdown, the heat released by the oxidation reaction, the loss of coolant water in the core and the heat transported by the steam produced. A computer program was coded in the C++ environment. The accident scenario of a BWR short term station blackout was simulated with this model. The results are compared with the ones obtained using MELCOR and RELAP/SCDAP codes. The comparison yielded an approximate result for total hydrogen production at the end of the simulation, with a difference of −2.7% compared with RELAP/SCDAP, and a difference of −1.11% with MELCOR. With the present model it is possible to calculate the growth of ZrO2, α-Zr(O), and β-Zr phases through the cladding.

Fork-end heat pipe for passive air cooling of spent nuclear fuel pool

When a station blackout accident occurs in a nuclear power plant, active cooling of spent nuclear fuels stored in a water pool using pumps is not available. As a result, the stagnant water in the pool gradually heats up via decay heat from the spent fuels and evaporates, thus exposing the spent fuels to air. This rapidly overheats the spent fuels and produces hydrogen gas due to oxidation of the zircaloy cladding. Accumulation of hydrogen gas can result in harmful detonation in the spent fuel storage building similar to what occurred in the Fukushima accident. In this study, a new concept of a gravity-assisted heat pipe that removes decay heat of spent fuels in a water pool by natural convection of ambient air is proposed. In the design, the condenser at the top of the heat pipe consists of several finned tubes for passive air cooling while the evaporator at the bottom is a single rod submerged in the spent fuel pool, which is called fork-end heat pipe (FEHP). A theoretical analysis model of a FEHP was developed and a laboratory-scale specimen was manufactured and tested for validation. It was found that the developed theoretical analysis model reasonably predicted the heat transfer rate of the FEHP specimen at diverse operating conditions and the operating limit point where the heat transfer rate is maximized due to the counter-current flow limit phenomenon.

Accident tolerant fuels (FeCrAl Cladding & Coating) performance analysis in Boiling Water Reactor (BWR) by the MELCOR 1.8.6 UDGC

Since the accident at Fukushima, one major goal of reactor safety research has been the development of more accident tolerant technologies that can mitigate or delay fuel degradation during a Beyond Design Basis Accident (BDBA). One major effort has been focused on increasing the capability of the fuel to be more tolerant of damage during an accident, i.e., Accident Tolerant Fuel (ATF) materials. In this work, we present the development of a generic BWR plant model, the modification of MELCOR to model ATF materials and the use of ATF materials (specifically FeCrAl alloy) as a coating on Zircaloy cladding or as a substitute material for cladding and fuel assembly canister material and its effect on severe accident progression, specifically, a Station Blackout accident. The analysis indicates that significant fuel degradation via fuel heat-up, clad oxidation, and hydrogen generation was delayed up to an hour if FeCrAl alloy was used as a clad and canister material. And, combined with the passive safety systems (i.e., the Reactor Core Isolation Cooling system, RCIC), the extended operation of these systems delayed fuel degradation further. However, an adverse effect should be emphasized for the monolithic FeCrAl design—it generated more hydrogen than the designs based on the Zircaloy due to the high reaction rate at a high temperature of FeCrAl. The design of the FeCrAl-coated-Zircaloy avoids this defect. Therefore, it is a promising choice to combine some of the beneficial traits of both materials.

Quantification of the effect of Cr-coated-Zircaloy cladding during a short term station black out

Accident Tolerant Fuel (ATF) has the potential to improve the inherent safety of current reactors. In particular, an ATF Cr-coated Zirconium alloy (Cr-coated-Zr) cladding can be an effective short-term approach to improve the cladding oxidation resistance without changing the cladding base materials. In order to appropriately assess the effect of such an ATF concept it is important to quantify the performance of such a cladding improvement not only during normal operation but also during postulated accidents. In this work, we examine the effect of Cr-coated-Zr cladding (Cr-Zr) during a short-term station blackout (STSBO) accident as well as quantify the uncertainty in the use of ATF Cr-coated-Zr cladding. This is accomplished using a coated cladding model developed for the MELCOR severe accident systems code; i.e., MELCOR-1.8.6-UDGC model. First, we quantify the effect of the coated cladding using a set of base case simulations that show the improved oxidation resistance of the Cr-coated-Zr cladding with the UDGC model parameters. Then, we quantify the uncertainty in these simulations by a Monte-Carlo analysis that examines the effect of the ATF clad input parameters (e.g., oxidation rate and failure temperature) on key output parameters; i.e., timing of rapid hydrogen generation, clad temperatures and hot leg creep rupture. Finally, we use regression analysis methods to estimate the importance of these input parameters. These analyses show that the Cr-coated-Zr cladding can delay rapid clad oxidation and generation of hydrogen by 3600 s to 7200 s with associated delays in rapid clad temperature increases. This additional time can be of benefit for compensatory operator actions. In contrast there is little effect on the timing of hot leg creep rupture since this failure is mainly controlled by core decay heat. We also find that these results are not significantly affected by the uncertainty in input parameters, and based on the regression analysis, the high temperature Zr oxidation rate has the major influence on the selected output parameters for the Cr-coated-Zr cladding.

Kinetic Model of Incipient Hydride Formation in Zr Clad under Dynamic Oxide Growth Conditions.

The formation of elongated zirconium hydride platelets during corrosion of nuclear fuel clad is linked to its premature failure due to embrittlement and delayed hydride cracking. Despite their importance, however, most existing models of hydride nucleation and growth in Zr alloys are phenomenological and lack sufficient physical detail to become predictive under the variety of conditions found in nuclear reactors during operation. Moreover, most models ignore the dynamic nature of clad oxidation, which requires that hydrogen transport and precipitation be considered in a scenario where the oxide layer is continuously growing at the expense of the metal substrate. In this paper, we perform simulations of hydride formation in Zr clads with a moving oxide/metal boundary using a stochastic kinetic diffusion/reaction model parameterized with state-of-the-art defect and solute energetics. Our model uses the solutions of the hydrogen diffusion problem across an increasingly-coarse oxide layer to define boundary conditions for the kinetic simulations of hydrogen penetration, precipitation, and dissolution in the metal clad. Our method captures the spatial dependence of the problem by discretizing all spatial derivatives using a stochastic finite difference scheme. Our results include hydride number densities and size distributions along the radial coordinate of the clad for the first 1.6 h of evolution, providing a quantitative picture of hydride incipient nucleation and growth under clad service conditions.

Numerical simulation of the effects of localized cladding oxidation on LWR fuel rod design limits using a SLICE-DO model of the FALCON code

Abstract A methodology for evaluation of mechanical and thermal effects of localized non-axisymmetric oxidation in zircaloy claddings on LWR fuel reliability is proposed. To this end, the basic capabilities of the FALCON fuel behaviour code are used. Examples of methodology application to adjustment of selected operational limits for modern BWR fuel rods, to capture effects of the excess local oxidation, are presented. Specifically, the limiting rod internal pressure for the onset of cladding lift-off is reduced, depending on initial excess oxidation spot sizes. Also, the power limits for Anticipated Operational Occurrences are adjusted, to preclude fuel melting and cladding failure due to PCMI and PCI-SCC in the affected fuel rods.