Unirradiated Zircaloy-4 Research Articles

Length scale effects of micro- and meso‑scale tensile tests of unirradiated and irradiated Zircaloy-4 cladding

Zircaloy-4 is an essential material for cladding structures within fission-based reactors. To explore the changes in properties measured on differing length scales, FIB-machined micro-scale tensile tests were performed on both irradiated and control groups of Zircaloy-4. This was correlated with tensile testing on femtosecond laser-machined meso‑scale specimens. Pronounced size effects were found when varying specimen geometry. Increases in tensile geometry size were associated with a reduction in measured yield stress for both irradiated and unirradiated samples. Meso-scale testing found strength and strain values similar to that of bulk-scale testing.

A study on the fracture pattern change of high-burnup fuel cladding failed by pellet-cladding mechanical interaction failure under reactivity-initiated accident conditions

ABSTRACT In a part of the historical reactivity-initiated accident (RIA) simulated tests on high-burnup fuels performed at the Nuclear Safety Research Reactor, the fuel failure caused by pellet-cladding mechanical interaction (PCMI) led to splitting into upper- and lower-part pieces or even fragmentation of the cladding tube. A massive release of fuel fragments accompanied this fracture pattern change from previously known axial cracks and thus identified as a potential concern in safety evaluation regarding core coolability. Dedicated out-of-pile mechanical tests were performed with unirradiated Zircaloy-4 cladding specimens to clarify the condition of such fracture pattern change. The specimens were pre-hydrided and subjected to loading with axial-to-hoop strain ratios of ℇz/ℇθ = 0.5–1.25, simulating the effects of hydrogen embrittlement and pellet-cladding mechanical bonding of high-burnup fuels, respectively. The results indicate that higher biaxiality of the loading and lower ductility (failure strain level) assist the fracture pattern change. This study proposes a conservative criterion that a PCMI failure splits the cladding tube into more than two pieces when strain ratio ℇz/ℇθ >0.75 and a concurrent hoop strain < 1.7% at the failure instant.

Hierarchical Bayesian modeling to quantify fracture limit uncertainty of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

For realizing a highly reliable fracture limit evaluation of fuel cladding tubes during loss-of-coolant accidents (LOCAs) in light-water reactors, we developed a method to quantify the fracture limit uncertainty of high-burnup advanced fuel cladding tubes. This method employs a hierarchical Bayesian model that can quantify uncertainty even with limited experimental data. The hierarchical Bayesian model was an extended version of the fracture probability estimation model developed in our previous study. The fracture limit uncertainty was quantified as a probability using the amount of oxidation (Equivalent cladding reacted: ECR) and the initial hydrogen concentration (the hydrogen concentration in the fuel cladding tubes before the LOCA-simulated tests) as explanatory variables. We divided the regression coefficients of this model into a hierarchical structure with an overall average term common to all types of fuel cladding tubes and a term representing differences among various types of fuel cladding tubes. This hierarchical structure enabled us to quantify the fracture limit uncertainty through the effective use of prior knowledge and data, even for high-burnup advanced fuel cladding tubes with a small number of data points. The fracture limits representing a 5% fracture probability with 95% confidence of the high-burnup advanced fuel cladding tubes evaluated by the hierarchical Bayesian model were higher than 15% ECR for the initial hydrogen concentrations of up to 700–900wtppm and restraint loads below 535N. These fracture limits were comparable to the limit of the unirradiated Zircaloy-4 cladding tube, indicating that the burnup extension and use of the advanced fuel cladding tubes do not significantly lower the fracture limit of fuel cladding tubes. Further, we proposed a method to reduce the fracture limit uncertainty by using non-binary data, instead of the binary data, depending on the condition of the fuel cladding tube specimens after performing the LOCA-simulated test, thereby increasing the amount of information in the data.

Effect of temperature and hydrogen concentration on the threshold stress intensity factor of radial delayed hydride cracking in fuel cladding

Three-point bending tests provide the ability to measure the propagation of axially oriented cracks in thin-walled fuel cladding in a radial outside-in direction. Via three-point bending, unirradiated Zircaloy-2 cladding containing 155 and 305 wppm hydrogen was mechanically loaded at different temperatures, sufficient to initiate delayed hydride cracking (DHC) whereupon the unloading rate was fast enough to arrest further propagation. The crack front dimensions were measured through fractography and implemented with the final load into a finite element model (FEM). Given the average crack length and final load, the FEM was able to back-calculate the stress intensity factor. Over a temperature range of 210–330 °C, around 6 MPa√m was calculated as the minimum threshold stress intensity factor for DHC. The trend shows that the ideal temperature for DHC with lowest threshold stress intensity factors may depend on the hydrogen concentration, and that the threshold stress intensity factors quickly increase with higher test temperatures. It has also been shown that creep affects the threshold stress intensity factor for optimal DHC less for lower hydrogen concentrations than for higher hydrogen concentrations. At low hydrogen concentration, the temperature of terminal solid solubility for precipitation (TSSP) is low enough so that creep-induced crack tip rounding plays a less significant role compared to the hydrogen kinetics.

Determining failure properties of as-received and hydrided unirradiated Zircaloy-4 from ring compression tests

An integrated experimental and computational approach was developed to determine the plastic strain at fracture and to investigate the fracture energy of as-received and hydrided unirradiated Zircaloy-4 nuclear cladding tubes in the hoop direction at room temperature from ring compression testing (RCT). This work builds on previous methods in the literature that were developed to obtain the mechanical properties of nuclear cladding before fracture (Young’s modulus, yield stress, and strain-hardening parameters) from experimental RCT results and extends the characterization of this material class to include material failure properties. A ductile damage approach was developed and implemented in a validated finite element model to predict material failure, evaluate fracture energy, and quantify the plastic strain at fracture initiation. The hydrogen content of the Zircaloy-4 specimens was varied up to 610 wppm when the failure became brittle. As expected, increasing hydrogen content caused embrittlement of the Zircaloy-4 samples tested and the plastic strain at fracture and fracture energy to decrease

Threshold stress intensity factor of delayed hydride cracking in irradiated and unirradiated zircaloy-4 cladding

The delayed hydride cracking (DHC) is becoming important because of its extended long-term dry storage and lower dry storage temperature than expected. However, it remains difficult to predict the DHC behavior of PWR spent fuel cladding due to the limited data of KIH (i.e., threshold stress intensity factor), especially for irradiated claddings. In this regard, the KIH for unirradiated and irradiated (49.641 GWd/tU) Zircaloy-4 claddings were experimentally measured using a pin-loaded-tension (PLT) technique and described by investigating the affected parameters. In addition, the results were compared with estimations from our previous studies. From the analysis, DHC is not expected to be a major concern during the predicted conditions of dry storage.

Understanding the role of hydrogen on creep behaviour of Zircaloy-4 cladding tubes using nanoindentation

The present article investigates the influence of hydrogen concentration on the creep performance of cold-worked stress-relieved unirradiated Zircaloy-4 cladding tube using nanoindentation technique. The as-received Zircaloy-4 tube is hydrided to the concentrations of 600 ppm and 900 ppm using gaseous hydrogen charging method. Constant load indentation creep tests are performed for a dwell period of 600 s in the temperature range of 300°C-500 °C at 1000 μN, 2000 μN, and 3000 μN. The impact of hydrogen is evaluated in terms of steady state power law creep exponent and activation energy. The power law creep exponent decreases with increase in hydrogen concentration, however, it remains fairly constant with increase in temperature up to 500 °C. Moreover, activation energy too decreases significantly with increase in hydrogen concentration. The mean stress exponent and activation energy are found to be 3.58 and 28.67 kJ/mol, respectively, for as-received sample.

Four-point-bend tests on high-burnup advanced fuel cladding tubes after exposure to simulated LOCA conditions

ABSTRACT To evaluate the fracture resistance of high-burnup advanced fuel cladding tubes during the long-term core cooling period following loss-of-coolant accidents (LOCAs), laboratory-scale four-point-bend tests were performed using the following advanced fuel cladding tubes with burnups of 73–84 GWd/t: low-tin ZIRLOTM, M5®, and Zircaloy-2 (LK3). Three four-point-bend tests were performed on the high-burnup advanced fuel cladding tube specimens subjected to the integral thermal shock tests which simulated LOCA conditions (ballooning and rupture, oxidation in high-temperature steam, and quench). During the four-point-bend tests, all the specimens that were oxidized at 1474 K to 9.9%–21.5% equivalent cladding reacted exhibited brittle fractures. The maximum bending moments were comparable to those of the conventional Zircaloy cladding tube specimens. Furthermore, the effects of oxidation and hydriding on the maximum bending moment were comparable between the high-burnup advanced fuel cladding tube specimens and the unirradiated Zircaloy-4 cladding tube specimens. Therefore, it can be concluded that the post-LOCA fracture resistance of fuel cladding tubes is not significantly reduced by extending the burnup to 84 GWd/t and using the advanced fuel cladding tubes, though it may slightly decrease with increasing initial hydrogen concentration in a relatively lower ECR range (< 15%), as observed for the unirradiated Zircaloy-4 cladding tubes.

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

ABSTRACTTo 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-MDATM, low-tin ZIRLOTM, 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.

Effects of hydride rim on the ductility of Zircaloy-4 cladding

High burn-up nuclear fuel cladding has a hydride rim owing to hydrogen diffusion under a temperature gradient. To simulate the high burn-up cladding, the hydride rim was formed in un-irradiated Zircaloy-4 cladding and ductility evaluation was conducted using ring compression test (RCT). The RCTs were conducted from room temperature to 300 °C for determining the ductile to brittle transition temperature (DBTT) of hydrided Zircaloy-4 cladding. Hydride rimmed specimens have lower ductility and DBTT than uniform hydrided specimens. The increased ductility at high temperatures is likely due to the increased ductility of the Zr matrix because the fractography at hydride rim still shows brittle fracture at those temperatures. The ductility of hydride rimmed specimens with radial hydrides are compared with those of irradiated Zr alloy claddings, and thus, it is concluded that hydride rimmed specimen can simulate the high burn-up Zr alloy cladding to some extent.

Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

ABSTRACT To evaluate the oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam, laboratory-scale isothermal oxidation tests were conducted using the following advanced fuel cladding tubes with burnups of up to 85 GWd/t: M-MDATM, low-tin ZIRLOTM, M5®, and Zircaloy-2 (LK3). These oxidation tests were performed in steam-flowing conditions at temperatures ranging from 1173 to 1473 K for durations between 120 and 4000 s, and the oxidation kinetics was evaluated. The oxidation kinetics of the high-burnup advanced fuel cladding tube specimens estimated by assuming the parabolic rate law was comparable to or slower than that of the unirradiated Zircaloy-4 cladding tube specimens reported in a previous study. It is considered that the protective effect of the corrosion layer hindered oxidation. Furthermore, no increase in the oxidation kinetics because of the pre-hydriding was observed. The onset times of the breakaway oxidations of these cladding tube specimens were comparable to those of the unirradiated Zircaloy-4 cladding tubes reported in previous studies. Therefore, it is considered that the burnup extension up to 85 GWd/t and the use of the advanced fuel cladding tubes do not significantly increase the oxidation kinetics and do not significantly reduce the onset time of the breakaway oxidation.

The effect of hydride morphology on the failure strain of stress-relieved Zircaloy-4 cladding with an outer surface pre-crack under biaxial stress states

ABSTRACT Hydride precipitates are considered to affect cladding integrity adversely during pellet-cladding mechanical interaction (PCMI) in a reactivity-initiated accident (RIA). This study aims to clarify the role of hydride precipitates in cladding failure under the biaxial stress condition. A displacement-controlled loading method was applied with a biaxial expansion-due-to-compression (biaxial-EDC) test apparatus to maintain a constant ratio of axial to hoop strains. The tests were conducted at room temperature on unirradiated stress-relieved Zircaloy-4 cladding tube samples with various hydrogen contents, pre-crack depths, and strain ratios. The amount and distribution of hydride precipitates (hydride morphology) were evaluated quantitatively and hydrogen content was measured to assess its effect on the decrease in outer surface hoop strain at failure (failure strain) of the samples. The decrease in failure strain of the hydrided sampleswas found to be more significant under lower strain ratios in the samples with shallower pre-crack. The failure strain of sample tended to be more sensitive to hydrogen content under the strain ratio with a higher axial component in the case of samples with hydrogen contents higher than ~150 wppm. This tendency might be explained by a scenario that considered the formation of micro-cracks in the hydride precipitates during the EDC test.

Release and speciation of carbon from Zircaloy-4 in anaerobic and highly alkaline conditions: Comparison of simple immersion and potentiostatic corrosion tests

ABSTRACTThe gas release and speciation of carbon species from irradiated and unirradiated Zircaloy-4 samples, representative for the fuel cladding as used in Belgian nuclear power plants, were studied in a saturated Ca(OH)2 solution in anaerobic conditions. This environment is relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level nuclear waste. To achieve this, we performed simple immersion and potentiostatic corrosion tests. Potentiodynamic polarization curves, recorded prior to the potentiostatic tests, revealed that irradiation seems to induce changes on the Zircaloy-4 corrosion behavior, such as a shift of the corrosion potential. Potentiostatic corrosion tests on unirradiated Zircaloy-4 provided a corrosion rate of ~54 nm/yr over a 7 day-experiment, whilst a corrosion rate of only ~4 nm/yr was calculated for the irradiated sample. Gas chromatography revealed that during simple immersion tests, which lasted 195 days, hydrogen, methane, ethane, and CO2 were produced, with methane being the major compound. Assuming that all carbon released from the metal was transformed into gaseous compounds, this yields to a corrosion rate ranging from 57 to 84 nm/yr for the irradiated sample. However, caution has to be taken on these corrosion rate and more tests should be performed to confirm these results.

Experimental and numerical investigation on thermo-mechanical behavior of fuel rod under simulated LOCA conditions

Abstract The data from out-of-pile single rod experiments can provide vital information about transient temperature variation and deformation rate of the cladding which can be extensively used to assess the flow blockage under Loss of Coolant Accidents (LOCA). With this motivation, in the present investigation experiments were conducted at low heating rates (2–8 K/s) to simulate a LOCA scenario. The un-irradiated Zircaloy-4 clad tube was internally heated using a tungsten heater with alumina pellets as a fuel rod simulator in an inert atmosphere. The ballooning initiation led to a significant decrease in temperature rise rate owing to increase in gap width between pellet and cladding. A maximum hoop strain of 93% at 1080 K in the α-phase and minimum hoop strain of 28% at 1340 K in the β-phase was observed. The Azimuthal Temperature Difference (ATD) over the clad tube was below 18 K for all tests and therefore ballooning was essentially symmetrical. Based on the experimental results, a burst criterion has been developed to predict the burst by a code named ‘TRAFR’ (Transient Response Analysis of Fuel Rod). The predictions based on plasticity and creep deformation models were compared with experimental results. At high temperature, the PLASTIC model predicted too low burst strain due to hardening effect. The burst strain predictions by the CREEP model were in good agreement with the experimental results in all temperature zones. The time-dependent behavior of creep phenomena led to a gradual rise in hoop stress until burst. The predictions with CREEP model can be improved by optimizing stress exponent value at high temperature in β-phase where superplastic behavior of Zircaloy-4 is governed by dislocation climb induced creep. However creep behavior is more complex and as such no simple law is established.

Experimental evaluation of nanoindentation as a technique for measuring the hardness and Young’s modulus of Zircaloy-2 sheet material

Experiments have been performed on a Zircaloy-2 sheet material to evaluate nanoindentation as a technique for measuring the Young’s modulus and hardness of the nuclear material. Room temperature transverse and longitudinal mechanical properties of an unirradiated Zircaloy-2 sheet material were measured with nanoindentation. The results were compared with those of tensile testing and Vickers microindentation obtained on the same material. Anisotropy evaluated with the conventional methods of microhardness measurement and tensile testing could be accurately confirmed by the nanoindentation measurement technique. The results show that the nanoindentation technique can be used to measure the Young’s modulus and hardness of small-volume samples of the material. The constraint factor, employed for converting hardness to yield strength, has also been determined and is reported for the material.

Mechanical Property Degradation of Unirradiated Zircaloy-4 Cladding After Creep Deformation

One of the limiting mechanisms of pressurized water reactor spent fuel cladding is creep owing to high temperature and rod internal pressure. Based on extensive studies, many countries have tentatively concluded that creep rupture is hard to occur under dry storage conditions and cannot severely degrade the integrity of the cladding if it meets the 400°C limitation owing to a self-limiting property. However, the changes in mechanical properties after creep deformation are not well understood due to the limited amount of relevant tests and analyses. In this regard, mechanical property degradation of unirradiated Zircaloy-4 cladding by creep deformation was investigated using a ring compression test and microscopic observation. In addition, the implication regarding spent fuel cladding integrity based on the test results is described.

Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.

Characterization of local hydride re-orientation in high burn-up PWR fuel rods induced by high pressure at high temperatures

ABSTRACTHydride re-orientation in high burn-up PWR fuel rods was induced by high pressure at high temperatures. The high-burnup specimens were sectioned from PWR rods taken from a 15×15 assembly of the H.B. Robinson (HBR) Unit 2 reactor. Out-of cell benchmark tests performed on unirradiated hydrided Zircaloy-4 specimens were conducted to determine the appropriate temperature, pressure, cooling rate, and number of cooling cycles for the reorientation of the irradiated in-cell specimens. The in-cell hydride reorientation tests were performed using highburnup fuel specimens under a hoop stress ≈145 MPa at 400ºC. The specimens were heated to the target temperature of 400ºC, held for 3 hours, cooled at 1ºC/min to 170ºC,and then heated at 1ºC/min to the target temperature again for five cycles. Post test metallographic examinations showed that a significant amount of radial hydrides were induced in the HBR fuel rods. The length of radial hydride was up to 60 µm. For unirradiated materials, the ductility of the radial hydride treated specimens is significantly reduced as compared to the as-hydrided specimens having the same hydrogen concentration (≈300 wppm in this work). The mechanical testing on irradiated fueled samples with hydride reorientation experiments have been performed, and will be reported separately in the near future.

Improved-EDC tests on the Zircaloy-4 cladding tube with an outer surface pre-crack

ABSTRACTIn order to investigate the failure behavior of fuel cladding under a reactivity-initiated accident (RIA) condition, biaxial stress tests on unirradiated Zircaloy-4 cladding tube with an outer surface pre-crack were carried out under room temperature conditions by using an improved Expansion-Due-to-Compression (improved-EDC) test method which was developed by Japan Atomic Energy Agency. The specimens with an outer surface pre-crack were prepared by using Rolling-After-Grooving (RAG) method. In each test, a constant longitudinal tensile load of 0, 5.0 or 10.0 kN was applied along the axial direction of specimen, respectively. All specimens failed during the tests, and the morphology at the failure opening of the specimens was similar to that observed in the result of post-irradiation examinations of high burnup fuel which failed during a pulse irradiation experiment. The longitudinal strain (ϵtz) at failure clearly increased with increasing longitudinal tensile loads and the circumferential strain (ϵtϑ) at failure significantly decreased in the case of 5.0 and 10.0 kN tests, compared with the case of 0 kN tests. From these tests, the data of cladding failure were obtained in the range of strain ratio (ϵtz/ϵtϑ) between about −0.6 and 0.7: this range of strain ratio covers the range between about 0.0 and 0.7 which is estimated in the case of RIA-simulated test. It is considered that the data obtained in this study can be used as a fundamental basis for quantifying the failure criteria of fuel cladding under a biaxial stress state.

A study on hydride reorientation of Zircaloy-4 cladding tube under stress

A study of the radial hydride formation in cold worked and stress relieved (CWSR) Zircaloy-4 cladding was performed using a ring tensile method and finite element analysis. Un-irradiated Zircaloy-4 cladding ring specimens charged with hydrogen were tested at various temperatures and loading conditions. Measured threshold stresses at loaded temperatures of 400°C, 335°C, 300°C, and 235°C turned out to be 60±5MPa, 68±5MPa, 75±6MPa, and 90±6MPa, respectively, demonstrating the tendency of decreasing threshold stress with increasing temperature. These stresses were lower than previously reported ones and close to the lower bound of collected data from the literature. These results reveal that high temperature preheating seems to affect an initiation of the reorientation. This implies that temperature increase caused by the vacuum drying of spent fuel assemblies for transportation can promote the hydride reorientation, leading to mechanical deterioration of the cladding.