Elastic anisotropy of zirconium alloy fuel cladding

Some methods to determine the anisotropic elasticity coefficients of zirconium alloy fuel cladding are discussed together with the conventional elastic constants.A simplified method, which uses the f parameters, was proposed, and the validity and applicability of the method were also investigated. The integration method, which was originally proposed by Rosenbaum et al., was found to be in excellent agreement with the experimental values of our own twisting test from room temperature to 800°C. The proposed f parameter method was also found to agree well with the values obtained by the integration method or experiment, especially at high temperatures near 700°C. It became evident that the elastic property of the typical fuel cladding was roughly isotropic at room temperature, and that the elastic anisotropy monotonically increased with temperature. Some stress or strain distributions of the fuel cladding were also obtained using anisotropic elasticity constants. The stress induced in the fuel cladding with simulated ridge deformation was very little affected by the difference in texture, but was more influenced by the elastic constants employed.     

The elastic properties of zirconium alloy fuel cladding and pressure tubing materials

A knowledge of the elastic properties of zirconium alloys is required in the mathematical modelling of cladding and pressure tubing performance. Until recently, little of this type of data was available, particularly at elevated temperatures. The dynamic elastic moduli of zircaloy-2, zircaloy-4, the alloys Zr-1.0 wt% Nb, Zr-2.5 wt% Nb and Marz grade zirconium have therefore been determined over the temperature range 275 to 1000 K. Young's modulus and shear modulus for all the zirconium alloys decrease with temperature and are expressed by empirical relations fitted to the data. The elastic properties are texture dependent and a detailed study has been conducted on the effect of texture on the elastic properties of Zr-1.0 wt% Nb over the temperature range 275 to 775 K. The results are compared with polycrystalline elastic constants computed from single crystal elastic constants, and the effect of texture on the dynamic elastic moduli is discussed in detail.     

Gas-cooled fast breeder reactor designs with advanced fuel and cladding

This paper examines the potential impact of some alternative cladding and fuel materials being considered for the liquid metal fast breeder reactor (LMFBR) on the performance and design of large commercial gas-cooled fast breeder reactors (GCFRs). Mixed carbide fuel and Inconel 718 cladding material were examined. Another cladding alternative considered was silicon carbide (SiC), which presents some interesting possibilities in high-temperature performance. Design concepts based on the above fuel and claddings were examined and compared with a reference oxide/316 stainless steel design based on a commercial 4000 MW(th) [1500 MW(e)] system. Substantial benefits can be derived from a high-temperature cladding material such as Inconel 718 or ¹⁶Pe; core volume and steam generator heat transfer area could be reduced by 20% or more, and significant reductions in core inventory and doubling time are possible. Carbide fuels would reduce the number of fuel rods by 50% because of higher linear power, and doubling time would be lowered.     

Physicochemical interaction of oxide fuel with the cladding of the fuel elements of A fast reactor

Conclusions The investigations of fuel elements with mixed oxide fuel, used in BOR-60, revealed four types of corrosion damage to OKh16N15M3B steel cladding due to the action of fission products. It was shown that the general corrosion develops as a result of the interaction of the cladding with cesium with an oxygen potential created by the mixed oxide fuel. Precipitation of carbides, formed as a result of radiation-thermal aging of the steel, on grain boundaries leads to intercrystallite corrosion of the cladding in the presence of cesium. When mixed oxide fuel with the starting ratio O/M=1.98–2.00 is used in the fuel elements, iodide transport of the components of the steel, giving rise to intercrystallite corrosion of the cladding, occurs. It was demonstrated that chemical activity relative to the stainless steel, leading to corrosion damage to the cladding, is high.Translated from Atomnaya Énergiya, Vol. 56, No. 4, pp. 195–199, April, 1984.     

Axial strains in fuel cladding associated with creep and fast neutron irradiation

Commercially produced CANDU-PHWR (CANada Deuterium Uranium-Pressurized Heavy Water Reactor) Zircaloy-4 and Zr—2.5% Nb fuel cladding was biaxially creep-tested in the laboratory and in the WR-1 reactor. The axial strains measured have been interpreted, through a knowledge of the texture, to provide evidence supporting prismatic slip as the major process for axial contraction at high stresses and temperatures and that this is the same process which gives rise to axial lengthening of pressure tubes. At low stresses and temperatures, axial lengthening of fuel cladding appears to be associated with Coble creep in elongated and flattened grains.     

Stress biaxiality in high-burnup PWR fuel cladding under reactivity-initiated accident conditions

In order to promote a better understanding of failure mechanisms of high-burnup pressurized water reactor (PWR) fuels under reactivity-initiated accident (RIA) conditions, stress biaxiality in cladding has been estimated for the pellet-cladding (PC) mechanical interaction (PCMI) phase. The estimation was based on an analysis of the transient elongations of a pellet stack and a cladding tube measured in RIA-simulating experiments in the nuclear safety research reactor (NSRR) using the RANNS code. Stress biaxiality in the high-burnup PWR fuel cladding during the PCMI phase has been estimated to be 0.7–0.8, on average, at the mid-wall of the cladding. A comparison with fresh fuel test results and a sensitivity analysis showed that the effects of burnup and pulse width on cladding stress biaxiality are less than 10% for the investigated range. The present analysis also indicated that PC friction is strong, and that the cladding constraint on pellet stack elongation is significant irrespective of burnup. Therefore, it is recommended that strong PC friction be assumed, which is similar to the mechanical bonding condition, and that fuel pellets be treated as deformable materials in models of fuel behavior during the PCMI phase.     

Theoretical analysis of stress, strain and fracture for fast reactor fuel cladding under constant power and power-cycling conditions: Visco-elastic model of nuclear fuel pin behaviour

The strain distributions produced in collapsed cladding by the expansion of cracked pellets during a power increase have been calculated analytically to establish the effect of a soft metal interlayer. As the radial cracks in the pellet open, the tendency for the cladding to stretch preferentially over them is reduced by frictional sliding at the pellet/clad interface. The frictional forces opposing sliding are reduced by the interlayer, whilst the ability of the cladding to resist the friction forces without being locally deformed depends on its work-hardening coefficient. The efficacy of the interlayer is greatest for interlayers which require the least strain to bed them in.     

Fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

ABSTRACT

To evaluate the fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident (LOCA) conditions, laboratory-scale integral thermal shock tests were performed using the following advanced fuel cladding tubes with burnups of 73–85 GWd/t: M-MDA^TM, low-tin ZIRLO^TM, M5®, and Zircaloy-2 (LK3). In total eight integral thermal shock tests were performed for these fuel cladding tube specimens, simulating LOCA conditions including ballooning and rupture, oxidation, hydriding, and quenching. During the tests, the specimens were oxidized to 10%–30% equivalent cladding reacted (ECR) at approximately 1473 K and were quenched under axial restraint load of approximately 520–530 N. The effects of burnup extension and use of the advanced fuel cladding tubes on the ballooning and rupture, oxidation, and hydriding under LOCA conditions were inconsiderable. Further, the high-burnup advanced fuel cladding tube specimens did not fracture in the ECR values equal to or lower than the fracture limits of the unirradiated Zircaloy-4 cladding tube reported in previous studies. Therefore, it can be concluded that the fracture limit of fuel cladding tubes is not significantly reduced by extending the burnup to approximately 85 GWd/t and using the advanced fuel cladding tubes, though it slightly decreases with increasing initial hydrogen concentration.     

A creep rupture criterion for Zircaloy-4 fuel cladding under internal pressure

The usual criterion which limits the cladding strain to 0.01 to prevent the creep rupture under internal pressure seems too conservative for application to transport and interim storage. So we have analysed CEA’s data on this subject for CWSR Zircaloy-4 in order to find a less conservative criterion. Temperatures between 350 and 470 °C were studied for stresses between 100 and 550 MPa, according to the irradiation level from 0 to 9.5 × 10²⁵ n m⁻². Except for high stressed irradiated material (because of low ductility), the plastic instability appears as the major mechanism of rupture. For the unirradiated material, it is essentially due to the stress increase with strain. This instability is accelerated by annealing for the irradiated one at moderate or low stress. From these considerations, we propose a new rupture criterion for CWSR Zircaloy-4 cladding submitted to internal pressure, for both unirradiated and irradiated materials.     

Properties of zirconium base cladding materials-corrosion and hydrogen pickup

This paper describes first corrosion and hydrogen pickup by zircaloy in water and steam. Out-of-pile corrosion is: treated briefly, as irradiation has major effects on the process. Main emphasis is on hydrogen pickup in-pile, as experience with the Shippingport Atomic Power Station core has led to the conclusion that susceptibility to hydriding is the major factor in determining the ultimate limits on the useful life of zirconiumbase cladding, rather than loss of wall thickness. The principal alloys used or considered are discussed.     

Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, “Monju”. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and ¹³⁷Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.     

Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 10²⁶ n/m² as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.     

Separate-effect tests on zirconium cladding degradation in air ingress situations

In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident.A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available.New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program—ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the “as-received” state, or after pre-oxidation in steam. “Analytical” tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 °C up to 1500 °C. Systematic post-test metallographic inspections are performed.The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of the oxide scale degradation under air exposure, important features that have to be taken into account are highlighted.     

The effect of a zirconium strength differential on cladding collapse predictions

An investigation was made into the occurrence of a strength differential in the Zircaloy cladding of LWR fuels, and into the effect such a strength differential can have on the analytical predictions of cladding creep collapse during fuel densification. The strength differential, or SD, refers to the difference in the compressive and tensile yield strengths of a material. It was concluded that an SD in Zircaloy cladding can have a significant effect on cladding collapse predictions; inclusion of SD considerations in cladding creep down analysis can increase predicted collapse times by a factor of two.     

Behavior of Si impurity in Np-Am-MOX fuel irradiated in the experimental fast reactor Joyo

The irradiation behavior of uranium-plutonium mixed oxide fuels containing a large amount of silicon impurity was examined by post-irradiation examination. Influences of Si impurity on fuel restructuring and cladding attack were investigated in detail. Si impurity, along with Am, Pu and O were transported by spherical pores and cylindrical tubular pores to the fuel center during fuel restructuring of the Np-Am-MOX fuel, where a eutectic reaction of fuel and Si-rich inclusions occurred. After fuel restructuring of the Np-Am-MOX fuel, Si-rich inclusions without fuel constituents were agglomerated at fuel crack openings where shallow attacks on the inner wall of the cladding were seen. Such shallow attacks on the inner wall of the cladding were likewise observed near the location of fuel cracks in long-term steady-state irradiated MOX fuels. Evidence of these shallow attacks on the inner wall of the cladding remained after fuel restructuring in normal MOX fuel. However, grain boundary corrosion of the cladding inner wall at the opening of the fuel cracks was selective and was marked in MOX fuel at higher oxygen potential by the release of reactive fission products such as Cs and Te in comparison with other regions of cladding wall.     

Centrifugal vacuum gasting for fuel cladding tube blanks

Translated from Atomnaya Énergiya, Vol. 67, No. 1, pp. 35–37, July, 1989.