Oxidation Of Cladding Research Articles

Comparison of Hydrogen Generation Rate between CORA-13 Test and MELCOR Simulation: Clad Solid-Phase Oxidation Models Using Self-Developed Code MYCOAC

To better understand the MELCOR oxidation and degradation models, past work compared the MELCOR model to a CORA experiment (CORA Test 13). These MELCOR analyses specifically focused on fuel bundle heatup and clad oxidation when compared to CORA test data. The comparison allowed the authors to investigate differences between hydrogen generation data and simulation results. Several potential reasons were considered for hydrogen generation rate differences, including MELCOR input power, heat transfer modeling, the clad solid-phase oxidation model, and the gaseous steam diffusion model. This work focuses on the possible uncertainty in the clad oxidation models used in MELCOR. First, the MELCOR nodalization approach for the CORA test was reviewed. Then, the temperature history and spatial variation were examined. One main focus was to consider other clad solid-phase oxidation models to compare the MELCOR models. This was accomplished by developing a separate model, MYCOAC, using MELCOR temperature predictions as input. Finally, the mass transfer resistance of steam diffusion to the clad surface was examined and found to be a small effect. While the Baker-Just solid-phase oxidation model showed better agreement with CORA data at low temperatures, the conclusion in this paper is that the oxidation models are not the major source of uncertainty in hydrogen generation rate differences. Future work will focus on heat transfer modeling of the CORA test.

Development of cladding oxidation analysis code [COAC] and application for early stage severe accident simulation of AP1000

Clad oxidation is an important phenomenon which can affect core degradation under severe accident conditions. The oxidation process always coincides with cladding material embrittlement. The energy released from the oxidation of cladding material can accelerate the accident process, and the hydrogen released from oxidation may threaten the integrity of vessel. As a consequence, the high temperature clad solid-phase oxidation model research and its code development play an important role in reactor safety analysis. To push the research on this region, a clad solid-phase oxidation analysis code (COAC) is developed by Xi'an Jiaotong University. To verify the quality of COAC, CORA-13 is used as a benchmark. The results show that COAC does produce expected values. COAC is then applied to AP1000 severe accident calculations. In this case, the station blackout accident (SBO) is modeled. The response of fuel rods under varying natural circulation conditions while coolant pump idle time is calculated. The calculations show that the oxidation of cladding can delay fracturing, and protect its integrity. Meanwhile, the performances of other oxidation models are explored in this paper.

Zirconium ignition in exposed fuel channel

A theoretical model based on simultaneous solution of the heat and mass transfer equations is developed for predicting the rate of thermo-chemical reaction between zirconium cladding and a hot steam environment. Ignition conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core are predicted based on the theory of metal combustion. A range of decay power, convective heat transfer coefficients, and initial temperatures leading to uncontrolled runaway cladding oxidation is identified. The model could be readily integrated as part of a fuel channel analysis code for predicting possible outcomes of different accident mitigation procedures in light water nuclear reactors under LOCA conditions.

Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5

The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents the analysis of loss of pool cooling accident scenarios and the discussion of mitigation measures for the SFP at a pressurized water reactor (PWR) NPP with the MAAP5 code. Analysis of uncompensated loss of water due to the loss of pool cooling with different initial pool water levels of 12.2m (designated as a reference case) and 10.7m have been performed based on a MAAP5 input model. Scenarios of the accident such as overheating of uncovered fuel assemblies, oxidation of claddings and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission products were predicted with the assumption that mitigation measures were unavailable. The results covered a broad spectrum of severe accident evaluations in the SFP. Furthermore, as important mitigation measures, the effects of recovering the SFP cooling system and makeup water in SFP on the accident progressions have also been investigated respectively based on the events of pool water boiling and spent fuels uncovery. Based upon the reference case, three cases with the recovery of SFP cooling system and three other cases with makeup water in SFP have been studied. The results showed that, severe accident might happen if SFP cooling system was not restored timely before the spent fuels started to become uncovered; spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate larger than the average evaporation rate defined as the division of pool water mass above the top of the spent fuel assemblies by the amount of time to start uncovering the spent fuel assemblies when no mitigation measures were taken. Although spent fuel assemblies were partly exposed due to the mass flow rate of makeup water being smaller than the average evaporation rate, continuous steam cooling and radiation heat transfer might maintain the spent fuels coolability as the actual evaporation was balanced by the makeup in a period of time of the order of several days. However, larger makeup rate should be guaranteed to ensure the long-term safety of SFP.

Cladding oxidation model development based on diffusion equations and a simulation of the monoclinic-tetragonal phase transformation of zirconia during transient oxidation

Zircaloy cladding oxidation is mostly represented by parabolic rate correlation. But the correlation approach is not suitable for long-term isothermal oxidation [4] or oxidation occurs under steam starvation conditions [5] and cannot obtain the detailed oxygen distribution which impacts the detailed mechanical behavior. To obtain the detailed oxygen distribution, a multi-phase diffusion problem with moving boundaries was introduced to simulate the cladding oxidation [9,10]. However, the hysteresis phenomenon related to the coexistence of monoclinic-tetragonal phases of zirconia which are very important to model the cladding oxidation during a LOCA, is not analyzed. In this study, a cladding oxidation model based on diffusion equations in the temperature range from 923K to 2098K which contains β-Zr, α-Zr, monoclinic-ZrO2, tetragonal-ZrO2, and cubic-ZrO2 is developed and the detailed oxygen distribution in the cladding could be obtained. It showed that the simulations of short-term and long-term isothermal oxidation, transient oxidation, and oxidation under steam starvation conditions were reasonable through comparing with the experimental data. We found that our model can give a reasonable simulation of the hysteresis phenomenon of monoclinic-tetragonal phase transformation during transient oxidation as well as a much better simulation of the hypothetical LOCA transient oxidation experiments [11] in ORNL than that by the code based on the parabolic rate correlation. This indicates that the developed model can accurately simulate the cladding oxidation during a LOCA transient.

Validation of ASTEC core degradation and containment models

Ruhr-Universitaet Bochum performed in a German funded project validation of in-vessel and containment models of the integral code ASTEC V2, jointly developed by IRSN (France) and GRS (Germany). In this paper selected results of this validation are presented.In the in-vessel part, the main point of interest was the validation of the code capability concerning cladding oxidation and hydrogen generation. The ASTEC calculations of QUENCH experiments QUENCH-03 and QUENCH-11 show satisfactory results, despite of some necessary adjustments in the input deck. Furthermore, the oxidation models based on the Cathcart–Pawel and Urbanic–Heidrick correlations are not suitable for higher temperatures while the ASTEC model BEST-FIT based on the Prater–Courtright approach at high temperature gives reliable enough results.One part of the containment model validation was the assessment of three hydrogen combustion models of ASTEC against the experiment BMC Ix9. The simulation results of these models differ from each other and therefore the quality of the simulations depends on the characteristic of each model. Accordingly, the CPA FRONT model, corresponding to the simplest necessary input parameters, provides the best agreement to the experimental data.

On the Use of Reduced-Moderation LWRs for Transuranic Isotope Burning in Thorium Fuel—II: Core Analysis

Multiple recycle of transuranic (TRU) isotopes in thermal reactors results in a degradation of the plutonium (Pu) fissile quality with buildup of higher actinides (e.g., Am, Cm, Cf), some of which are thermal absorbers. These phenomena lead to increasing amounts of Pu feed being required to sustain criticality and accordingly larger TRU content in the multirecycled fuel inventory, ultimately resulting in a positive moderator temperature coefficient (MTC) and void reactivity coefficient (VC). Because of the favorable impact fostered by use of thorium (Th) on these coefficients, the feasibility of Th-TRU multiple recycle in reduced-moderation (RM) pressurized water reactors (PWRs) and RM boiling water reactors (called RMPWRs and RBWRs, respectively) has been investigated. In this paper, Part II of two companion papers, the results of the single-assembly analyses of Part I are developed to investigate full-core feasibility. A large reduction in moderation is necessary to allow full actinide recycle. This increases the core pressure drop, which poses some thermal-hydraulic challenges, which are more pronounced if the design implementation is through retrofitting an existing PWR. For a given reactor cooling pump, the core flow rate is reduced. Despite this, it is possible to achieve feasible inlet and outlet temperatures and minimum departure from nucleate boiling ratio, for the reduction in moderation considered here. Reflood after loss-of-coolant accident is expected to be slower, which may lead to unacceptable peak clad temperatures and/or clad oxidation. Equilibrium cycles are presented for the RMPWR and RBWR, with a negative MTC and VC. However, the RMPWR may have positive reactivity when fully voided, and the hard spectrum makes it difficult to achieve an adequate shutdown margin, such that for the considered fuel designs, additional rod banks would be required.

Experimental results of the QUENCH-16 bundle test on air ingress

The out-of-pile bundle experiment QUENCH-16 on air ingress was conducted in the electrically heated 21-rod QUENCH facility at KIT in July 2011. It was performed in the frame of the EC supported LACOMECO program. The test scenario included the oxidation of the Zircaloy-4 claddings in air following a limited pre-oxidation in steam, and involved a long period of oxygen starvation to promote interaction with the nitrogen. The primary aim was to examine the influence of the formed oxide layer structure on bundle coolability and hydrogen release during the terminal flooding phase. QUENCH-16 was thus a companion test to the earlier air ingress experiment, QUENCH-10, which was performed with strongly pre-oxidized bundle. Unlike QUENCH-10, significant temperature escalation and intensive hydrogen release were observed during the reflood phase. Post-test investigations of bundle cross sections reveal residual nitride traces at various elevations. The nitrides were formed at upper bundle elevations characterized by steam starvation conditions. The external part of the oxide scale is of porous structure due to re-oxidation of nitrides during reflood. Relative thick internal oxide scales underneath this porous layer and residual nitrides were formed during reflood. At lower bundle elevations frozen partially oxidized melt was detected, relocated from upper elevations. Three contributors for the high hydrogen production during the reflood were recognized: re-oxidation of nitrides, secondary oxidation of residual cladding metal due to massive steam penetration through the porous oxide/nitride layer and melt oxidation.

The relationship between the amount of oxidation and activation energy on the steam oxidation reaction of Zircaloy-4 cladding

Abstract Zirconium-based alloys are widely used as cladding material for light-water reactors, and in the case of a loss-of-coolant accident (LOCA), oxidation of the cladding by high temperature steam plays important roles in fuel rod failure and hydrogen generation during the accident. In this study, considering that the oxidation rate is related to the activation energy of the oxidation reaction, the relationship between the amount of oxidation and the activation energy of cladding oxidation by steam was investigated by thermogravimetry. The oxidation rate of the specimen decreased with increasing heating rate. The activation energy of oxidation was evaluated based on a non-isothermal kinetics theory. It was found that the activation energy of oxidation depends on the specimen weight gain. The activation energies showed constant values of ∼300 and ∼180 kJ/mol in the oxidation range between 30 and 120 g/m 2 and above 120 g/m 2 , respectively. A master curve which expresses the amount of steam oxidation of Zircaloy-4 cladding was formulated based on the activation energies obtained.

Early phase fuel degradation in Phébus FP: Initiating phenomena of degradation in fuel bundle tests

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.

Transport and deposition in the Phébus FP circuit

In a nuclear reactor severe accident, fission products, after release from the fuel, along with related structural materials and actinides, are transported and partly deposited in the reactor coolant system, finally being injected into the containment. Both in the reactor coolant system and containment, they undergo physical and chemical processes that govern their behaviour and their airborne concentration in the containment atmosphere, which mainly determines the potential source term to the environment. The Phébus FP series provides a wide range of integral data on these processes, which are summarised in this account that covers the whole bundle geometry test series (FPT0 to FPT3). The transport in the circuit depends on the thermal hydraulic conditions (gas composition, temperature, flow rate). FPT0 and FPT1 were carried out under steam-rich conditions, while FPT2 and FPT3 were carried out with lower steam flows, leading to a period where hydrogen formed a large majority of the gas (∼90%) in the circuit during the main oxidation phase (oxidation of the Zircaloy cladding).The composition of the material transported in the circuit was dominated on average by the fission product noble gases, the volatile fission products Cs and Mo (except FPT0), the structural material Sn (from the cladding), the control rod materials Ag, In and Cd (FPT0/1/2) or B (FPT3), W and Re from the thermocouples, then other volatile fission products such as Te, Rb and I, the lower volatile Ba, and the fuel material U. For FPT2 and FPT3, whose main difference was in the control rod material, the proportions by mass are very similar if the control rod materials are excluded. Most of the materials were transported in the hot leg of the primary circuit as multi-component aerosols, with the notable exceptions of iodine (mainly in gaseous/vapour form) and Cd. In the cold leg, all materials are transmitted in aerosol form, with the notable exception of iodine in FPT3, where a large fraction reaches the containment in gaseous/vapour form. The size of aerosols is characterised by unimodal log–normal distributions; in FPT3 the results can also be interpreted in terms of bimodal distributions.Deposition is mainly concentrated where the temperatures of the wall and fluid decrease strongly, i.e. just above the bundle where the fluid cools from ∼2000°C to ∼700°C and, in the hot leg of the steam generator, from ∼700°C to ∼150°C. In the vertical line, simultaneously developing flows and changes in geometry (successive reductions in tube diameter) enhance deposition. The main processes in the circuit are: chemical transformation of vapours, vapour condensation onto structures as well as nucleation to form aerosols, or onto aerosols, aerosol agglomeration (primarily by diffusion), and thermophoretic deposition. The lower flows in FPT2/3 led to deposition being displaced upstream, e.g. onto the upper parts of the rods and also in the upstream part of the steam generator, where the presence of a large partial boron-rich blockage was deduced in FPT3. Revaporisation was observed in FPT1/2, notably for Cs, leading to delayed transport into the containment after reactor shutdown, also with some evidence for resuspension in FPT0. Post-test revaporisation experiments carried out externally on circuit samples from FPT1/3/4 aid in interpretation of the integral results. Also, the fact that temperatures fall from ∼2000°C at the exit of the bundle to ∼150°C at the exit to the containment, with fast transit times, implies that some reactions are kinetically limited so the composition at containment entry partly reflects that at higher temperatures. This applies particularly for iodine, where early presence of gaseous iodine is found in the containment (e.g. in FPT0, FPT1 and FPT3), and for carbonaceous gases resulting from B4C oxidation in FPT3, where the methane is lower than detection limits (reflecting high temperature bundle conditions, over 2000°C), whereas a non-negligible fraction would be expected under equilibrium, low temperature, containment conditions (typically about 150°C).

Analyses of the Phébus FPT3 Experiment Using the Severe Accident Codes ATHLET-CD, ICARE/CATHARE, and MELCOR

The aim of the Phébus Fission Product (FP) experimental program is to study the degradation phenomena and the behavior of the FPs released in the reactor coolant system and the containment building. The program consists of four in-pile bundle tests (FPT0, FPT1, FPT2, and FPT3), performed under different conditions concerning the thermal hydraulics and the environment of fuel rods, in particular, the amount of steam (strongly or weakly oxidizing atmosphere). The last test of this program, FPT3, was performed in November 2004 in Cadarache. During the FPT3 experiment, for the first time, boron carbide (B4C) was used as the absorber material instead of Ag-In-Cd, which was used in all the previous tests. Boron carbide is used in western-type pressurized water reactors, the EPR, boiling water reactors, and the VVER; consequently, assessing the effects of B4C on the main degradation phenomena and on gas release, as well as its impact on FP behavior is very important. This paper describes results from the Phébus FPT3 experiment, summarizes the test code modeling used in the different code applications, and reports the code results comparing some important experimental parameters, in particular regarding B4C control rod behavior. The severe accident codes used in these studies are Analysis of Thermal-Hydraulics of LEaks and Transients with Core Degradation (ATHLET-CD), ICARE/CATHARE, and MELCOR. The first part is an overview of the experimental results (boundary conditions, temperature evolutions, hydrogen and carbon compound releases coming from the oxidation of the Zircaloy claddings and the B4C absorber, and bundle degradation). The second part summarizes the code modeling used in the different code applications, in particular, those regarding absorber rod degradation and the oxidation process. The third part summarizes the code results comparing some important experimental parameters [thermal behavior, gas releases (H2, CO, CO2), and bundle degradation]. The conclusion focuses on the capabilities of the severe accident codes to simulate control rod behavior in a fuel rod assembly during the course of a severe accident transient.

Assessment of fuel rod performance codes under ramp scenarios investigated within the SCIP project

The behavior of advanced cladding materials under challenging conditions needs to be fully characterized, understood and modeled. This paper assesses the current predictability of fuel performance codes under loadings expected from pellet–clad mechanical interactions. A set of scenarios experimentally characterized within the SCIP project, were chosen so that a variety of materials and ramp power sequences could be examined.Four codes have been used in this study: ALCYONE V1.1, FALCON-PSI, FRAPCON-3 v3.3 and STAV7.3. Their predictions have been compared to data in terms of cladding oxidation, diameters and elongation. Predictability of clad oxidation was certainly scattered and while some codes showed reasonable accuracy, other results were notably deviated. As for diameters, most of the codes were capable of qualitatively capturing the axial profile, and showed consistency between diameters and hoop stress and strain predictions. Elongation estimates were generally poor, and were rather far from measurements in most cases (even the trends observed just vaguely followed by the codes).The results reported have been discussed in the light of the set of individual hypotheses and approximations made by modelers and codes regarding both boundary conditions (i.e., power histories, inlet coolant temperature, refabrication, etc.) and fuel and clad characterization (i.e., densification, rim porosity, materials properties, etc.). Additionally, code-to-code comparisons of some key variables (i.e., fuel temperature, contact pressure, hoop and axial stresses, etc.) highlighted systematic tendencies of the codes and supported the observations made.

Modelling of Zry-4 cladding oxidation by air, under severe accident conditions using the MAAP4 code

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.

Experimental and calculation results of the integral reflood test QUENCH-14 with M5® cladding tubes

The QUENCH-14 experiment investigated the effect of M5® cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (ISP-45) that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings. The PWR bundle configuration of QUENCH-14 with a single unheated rod, 20 heated rods, and four corner rods was otherwise identical to QUENCH-06. The test was conducted in principle with the same protocol as QUENCH-06, so that the effects of the change of cladding material could be observed more easily. Pre-test calculations were performed by the Paul Scherrer Institut (Switzerland) using the SCDAPSIM, SCDAP/RELAP5 and MELCOR codes. Follow-on post-test analyses were performed using SCDAP/RELAP5 and MELCOR as part of an ongoing programme of model validation and code assessment. Alternative oxidation correlations were used to examine the possible influence of the M5® cladding material on hydrogen generation, in comparison with Zircaloy-4. The experiment started with a pre-oxidation phase in steam, lasting ∼3000 s at ∼1500 K peak bundle temperature. After a further temperature increase to maximum bundle temperature of 2073 K the bundle was flooded with 2 g/s/rod water from the bottom. The peak temperature of ∼2300 K was measured on the bundle shroud, shortly after quench initiation. The electrical power was reduced to average value of 2 W/cm during the reflood phase to simulate effective decay heat level. Complete bundle cooling was reached in 300 s after reflood initiation. The development of the oxide layer growth during the test was essentially defined by measurements performed on the three Zircaloy-4 corner rods withdrawn successively from the bundle. The withdrawal of Zircaloy-4 and E110 corner rods after the test allowed a comparison of the different alloys in one test. One heated rod with M5 cladding was withdrawn after the test for a detailed analysis of oxidation degree and measurement of absorbed hydrogen. Post-test examinations showed neither breakaway cladding oxidation nor noticeable melt relocation between rods. Different from the QUENCH-14 (M5®) findings, the QUENCH-12 test with the E110 claddings performed under similar conditions had resulted in intensive breakaway effect at cladding and shroud surfaces during the pre-oxidation phase and local melt relocation on reflood initiation. The hydrogen production in QUENCH-14 up to reflood was similar to QUENCH-06 and QUENCH-12 bundle tests. During reflood 6 g hydrogen were released which is similar to QUENCH-06 (4 g) but much less than during quench phase of QUENCH-12 (24 g). The reason for the different behaviour of the Zr1%Nb cladding alloys is the different oxide scale properties of the materials.

Accelerated testing of HT-9 with zirconia coatings containing gallium using Raman Spectroscopy and XPS

Laser Raman spectroscopy and X-ray photoelectron spectroscopy were used to study the evolution of composition of oxide films in the presence of zirconia coatings on miniature HT-9 alloy specimens subjected to elevated temperature in air. The experiments expanded on previous efforts to develop a quick-screening technique for candidate alloys for cladding materials (HT-9) and actinide-based mixed oxide fuel mixtures (represented by the zirconia coating) by investigating the effect of both coating composition and annealing conditions on the high temperature reactions. In particular, the presence of the element Ga (a potential impurity in mixed oxide fuel) in the initial zirconia coating was found to accelerate the rate of oxide growth relative to that of yttria-stabilized zirconia studied previously. In addition, HT-9 samples that were subjected to different annealing conditions gave different results. The results suggest that the presence of Ga in a mixed oxide fuel will enhance the oxidation of HT-9 cladding under the conditions of this study, although the extent of enhancement is influenced by the type of annealing previously applied to the cladding material. Due to the multi-component layered structure of the oxidation products discovered in this fuel-cladding combination, it appears unlikely that Raman spectroscopy would be applicable as a stand-alone technique for the intended quick-screening approach to all but the simplest systems. Other techniques with better depth resolution, such as photoelectron spectroscopy, are required to follow the changes in composition of individual layers as they evolve during the oxidation process.

Oxidation of aluminum alloy cladding for research and test reactor fuel

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1–5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9–6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

Experimental investigation of the late phase of spent fuel pool accidents

Experimental programmes have been carried out in order to investigate the behaviour of nuclear fuel components in high-temperature air atmosphere, which characterises the main conditions of the late phase of spent fuel pool accidents. The tests provided new data on the oxidation of zirconium cladding in different atmospheres, on the oxidation and release of ruthenium from fuel pellets and on the integral behaviour of fuel bundles. The integral test confirmed that water injection into the spent fuel storage pool is the right measure to terminate a severe accident.

Behavior of Pre-hydrided Zircaloy-4 Cladding under Simulated LOCA Conditions

To promote a better understanding of high burn-up fuel rod behavior in a loss-of-coolant accident, laboratory-scale experiments were performed varying sample and test conditions with non-irradiated Zircaloy-4 claddings. Short test rods, fabricated with claddings having a wide range of hydrogen concentrations (about 100 to 1,450 ppm), were heated, isothermally oxidized at 1,220 to 1,500 K in steam flow, and quenched in flooding water. Axial shrinkage of the rods during the quench was restrained, controlling the maximum restraint load at four different levels. Test rods ruptured during the heat-up, and slight hydrogen concentration effects were seen on rupture temperature and strain. Depending primarily on the oxidized fraction of the cladding thickness, a part of claddings sustained circumferential cracking and fractured into two pieces during the quench. The fracture/no-fracture threshold of the oxidized fraction decreases as both the initial hydrogen concentration and axial restraint load increase. Consequently, when the restraint load is below 535 N, the fracture threshold is higher than 20% cladding oxidation, irrespective of the hydrogen concentration. This is sufficiently higher than the limit in the Japanese ECCS acceptance criteria.

GaInAsP Microcylinder (Microdisk) Injection Laser with AlInAs(Ox) Claddings

We propose and demonstrate a 1.5-µm-GaInAsP/AlInAs microcylinder laser which improves the fragility of the microdisk laser. The threshold current density was comparable to that reported for the 0.98-µm-GaInAs microcylinder laser. The selective oxidation of the AlInAs claddings will further reduce the threshold by strong optical confinement in the disk like structure and will simplify the large-scale integration of this type of device.