Oxidation Of Cladding Research Articles

An implicit enthalpy formulation for oxidation of nuclear reactor fuel cladding by steam

The oxidation of the zircaloy cladding of fuel rods in nuclear reactors, important during a severe accident, is governed by the diffusion of oxygen in solid zirconium. The latter exists in many phases which depend on the content of oxygen and have different diffusion coefficients. Discontinuous changes in oxygen concentrations occur at the phase boundaries, which also move in the course of diffusion, obeying an equation similar to the Stefan condition in heat transfer problems with phase change. We have devised and implemented a scheme for 1-D problems using the finite-difference approach, which involves setting the diffusion coefficient to zero in ‘forbidden’ concentration ranges. The method is shown to be mathematically equivalent to the enthalpy formulation for the Stefan problem. With a boundary condition at the vapour/solid interface which takes into account steam diffusion and dissociation, the code is applied to analyse the phenomena of steam starvation and the dissolution of the oxide layer, which are relevant to severe accidents. Copyright © 2000 John Wiley & Sons, Ltd.

Small break LOCA analysis for optimization of safety injection system of Korean standard nuclear power plant

The Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3&4) design was based on ABB-Combustion Engineering's (ABB-CE's) standardized 3800 MWt Nuclear Steam Supply System (System 80). Despite the 25% core power difference between the two plants (3800 MWt for System 80, 2815 MWt for UCN 3&4), several design features have not been changed, especially the capacity of Safety Injection System (SIS). Focusing on this fact, the sensitivity study for UCN 3&4 SIS capacity on the behavior of the postulated small break Loss of Coolant Accident (LOCA) was performed to optimize the SIS capacity. Firstly, the sensitivity study for Safety Injection Tank (SIT) capacity on the postulated small break LOCA was performed by changing the size of SIT from 100% to 60% (100%, 85%, 75%, 60%, respectively) for the 0.5 ft 2 (465 cm 2) pump discharge leg break. Secondly, the sensitivity study of High Pressure Safety Injection (HPST) pump capacity on the behavior of small break LOCA was investigated by reducing its discharge flow rate of UCN 3&4 to 50% (100%, 85%, 75%, 70%, 60% and 50%, respectively) for the 0.05 ft 2 (46.5 cm 2 pump discharge leg break. Finally, the spectrum analysis was performed using the acceptable results of these two studies, i.e., the reduced SIT size and HPSI pump capacity to 70% of UCN 3&4 to retain sufficient margin for the small break LOCA. The results of this spectrum analysis show that the peak cladding temperature (PCT) and peak local cladding oxidation (PLO) for the limiting break size of 0.05 ft 2 (46.5 cm 2 pump discharge leg break were 1459°F (793°C) and 0.486% respectively, which are in compliance with the Emergency Core Cooling System (ECCS) acceptance criteria of 2200°F (1204°C) and 17% specified in 10 CFR 50.46. The resuts of this sensitivity study can be adapted to the optimal design of SIS for the future 2815 MWt Korean Standard Nuclear Power Plant and be contributed to the increase of economical gain.

Fuel oxidation and irradiation behaviors of defective BWR fuel rods

The fuel oxidation of UO 2 pellets in two types of defective Zircaloy-clad BWR fuel rods with small leaks has been examined along both pellet diametral and fuel rod axial directions. The post-defect irradiation time was a few months for the base-irradiated full length rod and several minutes for the power-ramped segment rod. No phase change to higher order oxides of U 40 9 or U 3O 8 was found, but hyperstoichiometric UO 2+ x with fluorite structure was still present for both fuels. The fuel oxidation significantly depended on the defect size and distance from the defect. The pellet volume-averaged O/M ratios at various axial locations were in the range of 2.02–2.06 for the base-irradiated fuel, and about 2.01 for the power-ramped fuel. The data revealed that pellet oxidation by steam proceeded notably even in a short period of several minutes and played a more important role for generating liberated hydrogen, which could cause secondary hydriding of Zircaloy cladding, in comparison with the inner wall oxidation of cladding.

Transient Behavior of Waterside Corroded PWR Fuel.

A behavior of waterside corroded PWR fuels under transient conditions was studied. In order to simulate the corroded condition, unpressurized test fuel rods having cladding oxidation up to 80 pm and hydrogen contents up to 344 ppm in maximum were fabricated and pulse irradiated in the Nuclear Safety Research Reactor (NSRR) belonged to the Japan Atomic Energy Research Institute. Major results obtained are: 1. The waterside corroded fuel rods having oxides to the magnitude of 50, 60, 70 and 80 μm did not fail in the energy deposition range from 256 to 279 cal/g.fuel. The value was exceeded the failure threshold of 260 cal/g. fuel for NSRR standard fuel. While, the reference (no oxide) fuel rods in the same energy deposition range failed by melt/brittle of the cladding followed by wrinkle deformation.2. From posttransient irradiation examination it was found that wrinkle deformation in the waterside corroded fuels occurred very little due to rather relatively low cladding surface temperature and small diametral residual strain than those in the reference fuels.

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The international Phébus Fission Product Programme investigated key phenomena occurring in light water reactor core meltdown accidents in a series of five in-pile experiments. Four of these tests focused on the degradation of fuel rod bundles, containing a central control rod, and on the resulting release of fission products, structural materials and actinides from the fuel rod bundle, their transport in the reactor coolant system (RCS) and their subsequent behaviour in the containment vessel. Various steam contents were used in the RCS, from highly oxidising conditions (in FPT0 and FPT1) to more reducing ones (in FPT2 and FPT3).The “early degradation phase” took place at the beginning of the Phébus driver core and fuel bundle heat-up phase, with a quasi-intact fuel bundle geometry. During this phase, the degradation of the control rod and the oxidation runaway due to the fast oxidation of the Zircaloy claddings of the fuel rods, were two major events which took place.The oxidation runaway locally increased the temperatures much above the temperatures resulting from the Phébus driver core heat transfer to the bundle and yielded a large hydrogen release, which amounted to 70–80% of the whole hydrogen production during the tests. The maximum hydrogen flow rates increased with increasing steam flow rates injected at the fuel bundle inlet.The failure mechanisms of silver–indium–cadmium (used in three tests) and boron carbide (used in one test) control rods involve eutectic interactions amongst the components of these control rods. Mechanical deformations of the control rod stainless steel cladding against the Zircaloy control rod guide tube are the main presumed mechanisms for the beginning of these eutectic formations. However, different post-failure scenarios can be postulated for the effect of control rod degradation on fuel bundle degradation for both types of control rods. The exothermic oxidation of the exposed boron carbide pellets led to the release of carbonaceous species (CO, CO2) as well as of additional hydrogen, but no significant methane release could be detected above the limits of detection.Overall, the results confirmed existing knowledge concerning early phase degradation phenomenology found in previous integral experiments such as CORA and QUENCH (Karlsruhe Institute of Technology) and Phébus SFD (IRSN Cadarache), and formed a sound basis for analysis of the late phase degradation subsequently observed. Quantitative analysis of boron carbide control rod degradation in FPT3 pointed to a need for improved modelling of chemical reactions involving this material, particularly its oxidation in steam; this has been studied in the BECARRE experiments conducted by IRSN in the International Source Term Programme, leading to better quantitative understanding and improved modelling in codes such as the European reference severe accident analysis code ASTEC.

Information on the evolution of severe LWR fuel element damage obtained in the CORA program

In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400°C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4C), which are leading to extensive low-temperature melt formation around 1200°C. Interrelations between those basic phenomena, resulting for example in cladding deformation (“flowering”) and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water (“quenching”) are determining the evolution paths of fuel element destruction, which are to be identified. A further important task is the abstraction from mechanistic and microstructural details in order to get a rough classification of damage regimes (temperature and extent), a practicable analytical treatment of the materials behaviour, and a basis for decisions in accident mitigation and management procedures.

Simulation with the HITO code of the interaction of Zircaloy with uranium dioxide and steam at high temperatures

In a hypothetical severe fuel damage accident the Zircaloy (Zry) cladding may react, on its internal face, with the UO 2 fuel and, on the external face, with steam. These reactions result in the formation of ZrO 2, various oxygen stabilized α-Zr(O) phases and a (U, Zr) alloy. The HITO (High Temperature Oxidation) code describes both internal and external oxidation of the Zry cladding. The diffusion equations of oxygen in the various reaction layers are solved by an implicit finite differences method. Experimental results of isothermal and transient experiments are used to check the code. Comparison between the experimental data and the results calculated with the HITO code shows a good agreement.

Post-irradiation examinations and composition of the residues from nitric acid dissolution experiments of high-burnup lwr fuel

An LWR UO 2 fuel pin cross section was investigated after irradiation up to 5.9% maximum burnup at 21 kW/m time-averaged linear heat rating. The residues of this fuel were analyzed after its dissolution in 7 M nitric acid. The section-averaged concentration of converted plutonium is 1.38% PuO 2. At the fuel surface, the maximum is 3.8% PuO 2 which results in 14% local burnup. The concentrations of the components in the platinum metal inclusions and the dissolved and submicroscopically precipitated fission products Xe, Cs, Nd and Zr were analyzed as a function of the relative fuel radius. The inner and outer Zircaloy-4 clad oxidation was investigated. Oxygen diffused from the fuel into the clad attaining 0.2% O at the inner clad surface. The residues from the dissolution experiments amount to 0.6% of the fuel by weight and contain about 82% fission products, which are mostly reprecipitated as oxides from the nitric acid solution. Only 10% of the residues are composed of the incompletely dissolved hexagonal Mo-Tc-Ru-Rh-Pd phase which was formed during the irradiation process.

Core heatup and thermal behavior of the upper plenum structures affected by free convective gas flows in PWRs

With regard to accident management actions, further improvements of reactor accident codes are required. For example, the description of the thermohydraulics used in such codes is not detailed enough. Its modeling features usually do not correspond to the advanced modeling of the solid structures, especially in the reactor core region. Therefore, significant efforts have been recently undertaken to couple codes for a detailed analysis of fuel rod behavior and advanced thermohydraulics in order to simulate free convective gas flows in the reactor pressure vessel and the reactor cooling system of a PWR. Following dryout of the core, these flows may have a significant effect on the further heatup of the core and the surrounding structures, especially in the upper plenum region. At the Institut für Kernenergetik und Energiesysteme (IKE), a coupling of the core meltdown analysis code MELSIM and the code FRECON for the description of two-dimensional gaseous free convective flows in porous media has been developed. In the paper, a representative calculation for a PWR high pressure case is presented. As a result, a large scale convective flow is established. This enhanced gas flow, compared to the case with once-through steam flow, due only to evaporation of the residual water, leads to a slower heatup of the core and to a later onset of oxidation and local core melting. On the other hand, a significant heatup of the surrounding structures occurs. Especially in the upper plenum, the melting temperatures of steel are eventually reached. During later stages of the accident sequence, limitations in clad oxidation may occur, owing to the increasing amount of hydrogen in the recirculating gas mixture.

LWR fuel rod behavior during severe accidents

Chemical interactions between UO 2 fuel and Zircaloy cladding up to 2350°C are described. UO 2/Zircaloy single effects tests have been performed with short LWR fuel rod segments in inert gas and under oxidizing conditions. The reaction kinetics of molten Zircaloy cladding with solid UO 2 fuel has been investigated with UO 2 crucibles containing molten Zircaloy. The UO 2/Zircaloy reactions obey parabolic rate laws. The oxygen uptake by solid Zircaloy due to chemical interaction with UO 2 occurs nearly as quickly as that from the reaction with steam or oxygen. To study the competing effects of the external and internal cladding oxidation under realistic boundary conditions and the influence of the uncontrolled temperature escalation due to the exothermic steam/Zircaloy reaction on the maximum cladding temperature, single rod and bundle experiments have been performed. Electrically heated fuel rod simulators, including absorber rod material (Ag, In, Cd alloy), guide tubes and grid spacers are used. The maximum measured cladding temperature during the temperature escalation was about 2200°C. The failure temperature of the absorber rods and the extent of bundle damage depends on the guide tube material (Zircaloy or stainless steel) and varies between 1200 and 1350°C. The molten materials and liquid reaction products can relocate and form large coherent lumps on solidification, which may result in complete blockage of the fuel rod bundle cross section. In the future, 7 × 7 bundle experiments of 2 m overall length will be performed in the new CORA facility to study, in addition, the influence of quenching on fuel rod integrity.

Oxidation of zircaloy cladding in air

Abstract Non-irradiated, stress-relieved Zircaloy-4 claddings were oxidized at 350, 400, 450 and 500 °C in air. Some samples were dipped in sea water, dried and heated. Weight gain measured as a function of time revealed that the oxidation has pre- and post-transition regions just like the reaction in high temperature steam and water, and that sea water salts enhanced the weight gain after the transition point. Optical microscopy was carried out on the post-test oxide layer and substrate metal structure. Rate constants of the oxidation were determined and proved to be close to those of the previous studies including the experiments in high temperature steam and water.

An assessment of oxygen diffusion during UO 2-zircaloy interaction

During high-temperature transients on zircaloy clad nuclear fuel rods, both external and internal cladding oxidation have been observed. The result of these oxidation processes is to embrittle the cladding, making it susceptible to brittle fracture. External oxidation reaction processes are relatively well understood, while the internal processes have received little attention. This paper examines the internal oxidation process which occurs as a result of oxygen transport from the UO 2 fuel pellets to the zircaloy cladding. An oxygen diffusion model has been derived which solves a coupled two-madia problem with a moving boundary in zircaloy due to alpha-beta phase change. Analytical results for the time and temperature dependence of such zircaloy oxidation were found to compare favorably with both in- and out-of-pile experimental data.

Probability Distributions of Peak-Clad Temperature and Cladding Oxidation Thickness in Loss-of-Coolant Accidents for a Typical Boiling Water Reactor

The probability distributions of the peak-clad temperature (PCT) and of the maximum cladding oxidation thickness supposed to occur in the hypothetical loss-of-coolant accidents (LOCAs) for a typica...

Oxidation reaction kinetics of Zircaloy-4 in an unlimited steam environment

Experimental studies using a “Gleeble” have been performed on reactor grade Zircaloy-4 tubing to determine the oxidation behavior in steam throughout the 871°C to 1482°C temperature range. Isothermal oxidation under unlimited steam conditions has been found to obey the parabolic oxidation law with a rate constant for total reacted metal of K p = 3.10 × 10 5 exp(−33, 370/RT (K)) (mg/cm 2 ) 2 /sec . This measured rate is significantly lower than that predicted by Baker-Just for temperatures above 1077°C and in very good agreement with that suggested by Klepfer for the temperature range where tetragonal zirconium oxide is stable. Oxidation data presented is also in good agreement with that given by Hobson and Kittenhouse for oxidation below 1316°C but significantly less for oxidation above 1316°C. Oxidation of Zircaloy-4 in steam has been observed to be independent of steam superheat temperature and only slightly dependent on steam flow. It is apparent that as long as steam is present and hydrogen can migrate away, oxidation rate is independent of Reynolds number or velocity. Only steam starvation and/or trapped hydrogen appear to lower the rate of oxidation of new Zircaloy cladding. Le comportement à l'oxydation dans la vapeur entre 871°C et 1482°C de tubes en Zircaloy 4 de qualité nucléaire a étéétudié expérimentalement en utilisant un “Gleeble”. L'oxydation isotherme dans des conditions de vapeur illimitées s'est révélée obéir à une loi d'oxydation parabolique avec une constante de vitesse K p = 3,10 × 10 5 exp(−33, 370/RT (K)) (mg/cm 2 ) 2 /sec . Cette vitesse mesurée est nettement inférieure à celle prévue par Baker-Just pour des températures supérieures à 1077°C et en très bon accord avec celle suggérée par Klepfer pour le domaine de température où la zircone quadratique est stable. Les données d'oxydation présentées sont également en bon accord avec celles de Hobson et Rittenhouse pour des oxydations inférieures à 1316°C mais nettement inférieures pour des oxydations au-dessus de 1316°C. L'oxydation du Zircaloy-4 dans la vapeur s'est révélée être indépendante de la température de surchauffe de la vapeur et seulement geu fonction du courant de vapeur. Il apparaft qu'aussi longtemps que la vapeur est présente et que l'hydrogène peut s'eliminer, la vitesse d'oxydation est indépendante du nombre de Reynolds ou de la vitesse. Seuls le séchage de la vapeur et/ou l'hydrogène piégé semblent abaisser la vitesse d'oxydation du nouveau Zircaloy pour gainage. Das Oxidationsverhalten von reaktorreinen Zircaloy-4-Rohren wurde in Wasserdampf zwischen 871 und 1482°C experimentell untersucht. Die isotherme Oxidation bei unbegrenzter Dampfkonzentration gehorcht einem parabolischem Oxidationsgesetz mit einer Geschwindigkeitskonstante für das gesamte reagierte Metall von K p = 3,10 · 10 5 exp(−q/RT) mg 2 /cm 4 · s mit q = 33370 cal/mol . DieserWertistwesentlichniedrigeralsdervonBakerundJustfürTemperaturenoberhalb l077°C vorausgesagte, steht aber in sehr guter Übereinstimmung mit dem von Klepter empfohlenen Wert für den Temperatur bereich, in dem tetragonales ZrO 2 stabil ist. Die mitgeteilten experimentellen Ergebnisse stimmen ferner mit denen von Hobson und Rittenhouse angegebenen Werten für die Oxidation unterhalb 1316°C gut überein, jedoch bedeutend weniger für die Oxidation oberhalb 1316°C. Die Zircaloy-4-Oxidation in Wasserdampf ist nach den Beobachtungen unabhängig von der Überhitzungstemperatur und nur wenig abhängig von der Strömung. Es ist ersichtlich, dass, solange Wasserdampf vorliegt und Wasserstoff abwandern kann, die Oxidationsgeschwindigkeit von der Reynolds-Zahl oder der Geschwindigkeit unabhängig ist. Nur ein Mangel an Wasserdampf und/oder abgeschiedener Wasserstoff scheinen die Oxidationsgeschwindigkeit neuer Zircaloy-Rohre zu erniedrigen.

Method for estimating the exposure time and temperature for zircaloy oxidation in steam

In a nuclear power plant, a potential risk in some low probability situations in severe accidents is air ingress into the vessel. Air is a highly oxidizing atmosphere that can lead to an enhanced core oxidation and degradation affecting the release of Fission Products (FP), especially increasing that of ruthenium. This FP is of particular importance because of its high radio-toxicity and its ability to form highly volatile oxides. Oxygen affinity is decreasing between Zircaloy cladding, fuel and ruthenium inclusions in the fuel. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues.A review of existing data in the field of Zircaloy-4 oxidation in air-containing atmosphere shows that this phenomenon is quantitatively well understood. The cladding oxidation process can be divided into two kinetic regimes separated by a breakaway transition. Before transition, a protective dense zirconia scale grows following a solid state diffusion-limited regime for which experimental data are well fitted by a parabolic time dependence. For a given thickness, which depends mainly on temperature and the extent of pre-oxidation in steam, the dense scale can potentially breakdown. In case of breakaway combined with oxygen starvation, cladding oxidation can then be much faster because of the combined action of oxygen and nitrogen through a complex self sustaining nitriding-oxidation process.A review of the pre-existing correlations used to simulate zirconia scale growth under air atmospheres shows a high degree of variation from parabolic to accelerated time dependence. Variations also exist in the choice of the breakaway parameter based on zirconia phase change or oxide thickness. Several correlations and breakaway parameters found in the literature were implemented in the MAAP4.07 Severe Accident code. They were assessed by simulation of the QUENCH-10 test, which is a semi-integral test designed to study fuel bundle exposure to steam first and then to air. This paper deals with the main results obtained with MAAP4.07 when simulating QUENCH-10.

Oxidation of nuclear reactor fuel cladding by steam: a moving boundary diffusion problem

The oxidation of the zircaloy cladding of fuel rods in nuclear reactors, important during a severe accident, is governed by the diffusion of oxygen in solid zirconium. The latter exists in many phases which depend on the content of oxygen and have different diffusion coefficients. Discontinuous changes in oxygen concentrations occur at the phase boundaries, which also move in the course of diffusion, obeying an equation similar to the Stefan condition in heat transfer problems with phase change. We have devised and implemented a novel scheme for 1-D problems using the finite-difference approach, which involves setting the diffusion coefficient to zero in forbidden concentration ranges. The method is shown to be mathematically equivalent to the enthalpy formulation for the Stefan problem. With a boundary condition at the vapour/solid interface which takes into account steam diffusion and dissociation, the code is applied to analyse the phenomena of steam starvation and the dissolution of the oxide layer, which are relevant to severe accidents.