Ductility Of Cladding Research Articles

Recrystallization and grain growth of Zr-Nb-Sn alloy in 400–500 °C and effect on hydride embrittlement

The recrystallization and grain growth of a reactor-grade Zr-Nb-Sn alloy were experimentally investigated in the temperature range of 400–500 °C. Electron backscatter diffraction (EBSD) analyses were conducted to characterize grain growth and texture change caused by annealing. Hardness measurements were conducted to obtain the recrystallization fractions and were compared with the machine learning-based recrystallization analyses provided by AZtecCrystal. With good agreement with the hardness-based recrystallization fraction, the machine-learning-based method provides classification of deformed (recrystallized and grown) and undeformed (non-recrystallized) grains, thereby revealing the local behavior of dynamic grain growth. Recrystallization occurs in 4 h at 450 °C, with distinct grain growth after 24 h. Mechanisms of intergranular hydride connectivity decrease with grain growth are elucidated. The reduced hydride connectivity with grain growth primarily occurs by the formation of intra-granular hydrides and homogenization of grain boundary energies. The reduced hydride connectivity significantly increases cladding ductility. For cladding with a significant hydrogen content (i.e., ∼600 wppm), the increase in cladding ductility (i.e., strain energy density (SED)) owing to the reduced hydride connectivity supersedes the background matrix ductility enhancement associated with grain growth.

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

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

Post-LOCA ductility of Cr-coated cladding and its embrittlement limit

Post-Loss of Coolant Accident (LOCA) ductility assessments of Cr-coated Zircaloy cladding was conducted in compliance with the United States Nuclear Regulatory Commission (U.S.NRC)’s guidelines. The Equivalent Cladding Reacted (ECR) limit for Cr-coated cladding oxidized on both sides, representing the embrittlement process of an outer-side coated cladding near the burst hole followed by ballooning, is found to be 13.8% at steam oxidation temperature of 1204°C. Ring Compression Test (RCT) induces peak tensile stress at the inner surface of the cladding specimen at which the crack starts to propagate from the brittle phases (ZrO2 and α(O)-Zr). Such characteristic of RCT is responsible for the decrease in the ECR limit for the claddings coated on the external surface as its oxygen-affected brittle phases are developed at the inner cladding surface. Outer-side Cr-coated cladding gives approximately 10 minutes of extra accident coping time at the fixed steam oxidation temperature 1204°C before it reaches the cladding embrittlement limit.

Effect of cooling rate on the residual ductility of Post-LOCA Zircaloy-4 cladding

Effects of cooling rates on mechanical behavior of steam-oxidized Zircaloy-4 are systematically investigated with the cladding material oxidized to the current ductility-based regulatory limits (17% ECR at 1204 °C). Three different cooling rates associated with distinctively different quenching environments are tested; ambient air cooled (<7 °C/s), boiling water (∼400–500 °C/s), and room temperature water (∼1600–2000 °C/s). Cooling rates of post-LOCA Zircaloy-4 have (1) limited effects on average cladding residual ductility for non-hydrided cladding, (2) pronounced effects on statistical variation of ring compression test results, and (3) effects on fracture strength reduction on pre-hydrided cladding (765 wppm). The limited effect of cooling rate on average residual ductility of non-hydrided cladding is due to the cooling-rate insensitive resulting phase thicknesses and their oxygen contents, both of which dictate the level of cladding residual ductility. Homogenizing prior-β phase with finer lamellar structures and reduced α incursion, fast cooling (i.e., 25 °C water quenching) effectively reduces statistical variation of RCT results. The presented results support regulatory validity of past experiments conducted irrespective of cooling rates for Zircaloy-4 with no pre-transient hydrogen content. The observed insensitivity of cooling rates on the ECCS criteria for non-hydrided Zircaloy-4 may represent a gap between ductility-based regulation and integral fuel rod behavior. Cooling rate effect is shown to become increasingly important for increasing pre-transient hydrogen embrittlement, as shown by the fracture strength changes.

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.

Ring compression tests on un-irradiated nuclear fuel rod cladding considering fuel pellet support

Ring compression tests on un-irradiated nuclear fuel rod cladding were conducted to analyze its mechanical behavior under pinching loads using appropriate boundary conditions, such as the consideration of fuel pellet support. The tested specimens included as-fabricated and artificially hydrogen-charged Zircaloy-4 cladding samples with a hydrogen content in a range of 285–470 ppm. Part of the samples were subjected to radial hydride treatment including a peak cladding hoop stress of 79 MPa and a peak cladding temperature of 400 °C to simulate SNF vacuum drying and to investigate treatment effects on the cladding response. Half of the tested rings were loaded with 20 mm long stainless steel pellets to analyze the impact of fuel pellet presence on the cladding loading capacity. The pellet-cladding gap width ranged from 60 to 180 μm. The other half of the rings was tested in empty state. The load-displacement curves obtained from ring compression tests conducted at room temperature on as-fabricated cladding exhibited a highly ductile material behavior. The presence of hydrogen in the cladding significantly embrittled the material, but unexpectedly, radial hydride treatment increased the cladding ductility. The ring compression tests conducted under pellet presence did not induce cladding cracking, even under extreme pinching loads. The results indicate that hydride-related material embrittlement likely does not cause nuclear fuel cladding failure when subjected to pinching loads, under the premise that a fuel pellet provides sufficient support to the cladding.

Fuel cladding behavior under rapid loading conditions

A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

Effects of hydride morphology on the embrittlement of Zircaloy-4 cladding

Spent nuclear fuel claddings discharged from water reactors contain hydrogen up to 800wppm depending on the burn-up and power history. During long-term dry storage, the cladding temperature slowly decreases with diminishing decay heat and absorbed hydrogen atoms are precipitated in Zr-matrix according to the terminal solid solubility of hydrogen. Under these conditions, hydrides can significantly reduce cladding ductility and impact resistance, especially when the radial hydrides are massively present in the material. In this study, the effects of hydride morphology on the embrittlement of Zircaloy-4 cladding were investigated using a ring compression test. The results show that circumferentially hydrided Zircaloy-4 cladding is brittle at room temperature but its ductility is regained substantially as the temperature goes above 150°C. On the other hand, radially hydrided cladding remains brittle at 150°C and micro-cracks developed in the radial hydrides can act as crack propagation paths. Fracture energy analysis shows that ductile to brittle transition temperature is low in between 25°C and 100°C in the former case, whereas it lies in between 200°C and 250°C in the latter case.

Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick’s law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

Effect of the Intermediate Cooling Process on the Behavior of the Zircaloy-4 Cladding After a High-Temperature Oxidation

Studies were conducted to investigate the effect of the intermediate cooling process on the thermal shock behavior of Zircaloy-4 fuel cladding under a simulated loss-of-coolant accident condition and to analyze the related mechanical and microstructural properties. The Zircaloy-4 specimen was oxidized at the desired temperature and time, then various cooling processes were applied such as the direct water quench, the intermediate cooling at 700°C for 200 and 2000 s, and the successive cooling from 950 to 700°C. The results showed that the direct water quenching without any intermediate cooling process reduced the cladding ductility in that it reduced the minimum equivalent cladding reacted from 20 to near 17%. Ring compression ductility decreased, and the minimum thickness of the prior-beta layer thickness that causes brittle failure increased from 0.3 to 0.4 mm in the case of the direct water quench condition. As the cooling rate increased, the size of the plate inside the prior-beta phase decreased so that it induced an increase in the residual dislocation density to result in a decrease of the cladding ductility. Additional oxidation effect during a slow cooling below 950°C had little influence on the cladding behavior.

Effect of radial hydrides on the axial and hoop mechanical properties of Zircaloy-4 cladding

The effect of radial hydrides on the mechanical properties of stress-relief annealed Zircaloy-4 cladding was studied. Specimens were firstly hydrided to different target hydrogen levels between 100 and 600 wt ppm and then thermally cycled in an autoclave under a constant hoop stress to form radial hydrides by a hydride reorientation process. The effect of radial hydrides on the axial properties of the cladding was insignificant. On the other hand, the cladding ductility measurements decreased as its radial hydride content increased when the specimen was tested in plane strain tension. A reference hydrogen concentration for radial hydrides in the cladding was defined for assessing the fuel cladding integrity based on a criterion of the tensile strength 600 MPa. The reference hydrogen concentration increased with the specimen (bulk) hydrogen concentration to a maximum of ∼90 wt ppm at the bulk concentration ∼300 wt ppm H and then decreased towards higher concentrations.

Effect of Cooling History on Cladding Ductility under LOCA Conditions

In a Loss-of-Coolant Accident of LWRs, heated and oxidized fuel cladding is slowly cooled and then quenched by re-flooding water. Effects of cooling rate during the slow-cooling process and quench temperature were investigated on post-quench ductility of the oxidized cladding. Unirradiated Zircaloy-4 cladding specimens were oxidized in steam at 1,373 or 1,473 K, cooled at a rate from 2 to 7 K/s, and finally quenched from temperatures ranging 1,073 to 1,373 K. Post-quench ductility was evaluated by ring-compression test, and microscopic properties were examined by metallurgical examination, hardness test, and oxygen analysis. Morphology of oxygen-rich α phase in the metallic prior-β phase layer changed depending on the cooling history. Area fraction of α phase region in the cross section obviously increased and post-quench ductility reduced with decrease in the quenching temperature. Since α phase region with low ductility can be preferential path for crack propagation, increase in area fraction of α phase region possibly decreases resistance for fracture, which results in ductility reduction. On the other hand, the area fraction was nearly constant irrespective of the rate of the slow cooling, and consequently, effect of the cooling rate on the post-quench ductility was negligible.

Effects of oxide and hydrogen on the behavior of Zircaloy-4 cladding during the loss of the coolant accident (LOCA)

Effects of surface oxide and absorbed hydrogen on the behavior of the loss of the coolant accident (LOCA) were investigated in this study. High temperature ballooning and thermal quench tests were performed for Zircaloy-4 cladding which had been prepared with up to 50 μm of oxide and 1000 ppm of hydrogen, respectively. In the high temperature ballooning test, the initially pressurized cladding was heated until a rupture. Threshold oxidation (ECR) of each condition was evaluated in the thermal quench test in which oxidized cladding at the LOCA temperature was quenched by water. Ring compression test was performed to assess the ductility of the quenched cladding The results showed that both the oxide and hydrogen affected the high temperature ballooning property due to the constraint of the α phase by the surface oxide as well as the expansion of the β phase by the absorbed hydrogen. In the quench test, the pre-oxide and absorbed hydrogen did not affect the high temperature oxidation whereas the threshold ECR decreased in the hydrogen charged cladding because the absorbed hydrogen increased the maximum oxygen solubility inside the residual β layer to reduce the cladding ductility.

Effect of Cooling History on Cladding Ductility under LOCA Conditions

Effect of Cooling History on Cladding Ductility under LOCA Conditions

Influence of Hydride Re-orientation on BWR Cladding Rupture under Accidental Conditions

Hydride precipitation along the radial-axial plane increases in high burn-up boiling water reactor (BWR) fuel claddings. The radially-oriented hydrides may have an important role during fuel behavior in a reactivity-initiated accident and may reduce ductility of the cladding under pellet-cladding mechanical interaction (PCMI) conditions. In order to promote a better understanding of the influence of the radial hydrides on cladding failure behavior under the PCMI conditions, tube burst tests were conducted for unirradiated BWR claddings charged with 200 to 650 ppm of hydrogen. About 20 to 30% of hydrides were re-oriented and precipitated along the radial-axial plane. The claddings exhibited large rupture openings with an axial crack at room temperature and 373 K. The crack penetrated through cladding wall preferentially along the radial hydrides, and radial cross section showed cladding failure in a brittle manner. However, reduction in residual hoop strain by precipitation of the radial hydrides was very small. It is accordingly expected that ductility of high burn-up BWR cladding is significantly reduced not only by precipitation of radial hydrides as far as hydrogen concentration and radial hydride fraction range in the present study.

Short-time creep and rupture tests on high burnup fuel rod cladding

Short-time creep and rupture tests were performed to assess the strain potential of cladding of high burnup fuel rods under conditions of dry storage. The tests comprised optimized Zry-4 cladding samples from fuel rods irradiated to burnups of up to 64 MWd/kg U and were carried out at temperatures of 573 and 643 K and at hoop stresses of about 400 and 600 MPa. The applied stresses were chosen to reach about 2% strain within an envisaged testing time of 3–4 days. The tests were followed by a low temperature phase at 423 K and 100 MPa to assess the long-term behaviour of the cladding ductility especially with regard to the higher hydrogen content in the cladding of the high burnup fuel. These tests showed that around 600 K, a uniform plastic strain of at least 2% is reached without cladding failure. The low temperature phase at 423 K for up to 5 days revealed no cladding failure under these conditions of reduced cladding ductility due to the increased hydrogen content.

Oxidation of Zircaloy-4 under High Temperature Steam Atmosphere and Its Effect on Ductility of Cladding

Oxidation of Zircaloy-4 under High Temperature Steam Atmosphere and Its Effect on Ductility of Cladding

Oxidation of Zircaloy-4 under High Temperature Steam Atmosphere and Its Effect on Ductility of Cladding

Zircaloy-steam reaction rate in flowing steam was measured in the temperature range of 900∼1,330°C. The reaction rate measured from the weight gain followed a parabolic law above 1,000°C, but it deviated from the law at 900 and 950°C. Growth rates of ZrO2 layer and ZrO2+ stabilized α-phase were also parabolic above 1,000°C. The parabolic rate law constants are represented as follows: Weight gain k w =0.468 exp(−40,710/RT) (g2/cm4.sec) ZrO2 k δ= 0.0215 exp(−35,860/RT) (cm2/sec) ZrO2 +stabilized α k ξ=0.306 exp(−38,970/RT) (cm2/sec). The ductility of zircaloy tubings oxidized in the temperature range of 950∼ 1,330°C was measured by a ring compression test at 100°C. It tends to decrease with increasing the amount of oxidation and the oxidation temperature except the case of 950°C, where the specimens become brittle even at less amounts of oxidation. From these experimental results, it is concluded that the regulation for the amount of the zircaloy oxidation in the acceptance criteria for LWR's LOCA analyses ...

Fast Reactor Fuel Performance Model Development

AbstractThe status of fast reactor mixed-oxide fuel performance model development, involving the OLYMPUS and CYGRO-F codes, is reviewed. Input information in several critical areas is examined including: the swelling, irradiation creep, and ductility of stainless-steel cladding; and the swelling and plasticity of mixed-oxide fuel. The predictions from the two codes are illustrated by parametric studies and application to high fluence/burnup ratio fuel rods of interest to the LMFBR program. The analytical studies indicate the major effects of stainless-steel swelling on fuel rod performance and the relationship between fuel density and cladding creep strain.