Zircaloy Cladding Research Articles

Effect of ATF Cr-coated-Zircaloy on BWR In-vessel Accident Progression during a Station Blackout

The deposition of protective coatings on nuclear fuel cladding has been considered as a near-term Accident Tolerant Fuel (ATF) concept that will reduce the high-temperature oxidation rate and enhance accident tolerance of the cladding while providing additional benefits during normal and transient. In this study, the performance of the proposed ATF concept of Cr-coated-Zircaloy is assessed using a generic Boiling Water Reactor MELCOR plant model considering a Short-term Station Blackout (STSBO) scenario. Simulation results indicate that the use of Cr-coated-Zircaloy as cladding and canister material mitigates the core degradation process as compared to the traditional Zircaloy cladding and canister design. The onset of fuel degradation and collapse is delayed by over thirty minutes, and the extent of fuel degradation is reduced. Specifically, the gross in-vessel hydrogen generation decreased by almost a factor of three. Although the eutectic reaction between Cr-coating and Zircaloy could cause an early failure of the coating, the improvement in the delay of fuel degradation is still notable. Additionally, a thicker coating is found helpful to obtain additional coping time and to decrease hydrogen generation. In addition to the eutectic formation that may compromise Cr-coated Zr, a different failure mode is identified for the Cr-coated-Zr when compared to Zircaloy; i.e., a complete melt of base material leads to component collapse before the coating is oxidized and consumed. These findings can help the industry focus on productive areas of research and development for accident-tolerant fuel concepts and enhancement of core safety margins.

Accident tolerant fuels (FeCrAl Cladding & Coating) performance analysis in Boiling Water Reactor (BWR) by the MELCOR 1.8.6 UDGC

Since the accident at Fukushima, one major goal of reactor safety research has been the development of more accident tolerant technologies that can mitigate or delay fuel degradation during a Beyond Design Basis Accident (BDBA). One major effort has been focused on increasing the capability of the fuel to be more tolerant of damage during an accident, i.e., Accident Tolerant Fuel (ATF) materials. In this work, we present the development of a generic BWR plant model, the modification of MELCOR to model ATF materials and the use of ATF materials (specifically FeCrAl alloy) as a coating on Zircaloy cladding or as a substitute material for cladding and fuel assembly canister material and its effect on severe accident progression, specifically, a Station Blackout accident. The analysis indicates that significant fuel degradation via fuel heat-up, clad oxidation, and hydrogen generation was delayed up to an hour if FeCrAl alloy was used as a clad and canister material. And, combined with the passive safety systems (i.e., the Reactor Core Isolation Cooling system, RCIC), the extended operation of these systems delayed fuel degradation further. However, an adverse effect should be emphasized for the monolithic FeCrAl design—it generated more hydrogen than the designs based on the Zircaloy due to the high reaction rate at a high temperature of FeCrAl. The design of the FeCrAl-coated-Zircaloy avoids this defect. Therefore, it is a promising choice to combine some of the beneficial traits of both materials.

High Temperature Anti-Oxidation Behavior and Mechanical Property of Radio Frequency Magnetron Sputtered Cr Coating

A dense and uniform Cr coating was fabricated on the zirconium alloys fuel claddings by radio frequency (RF) magnetron sputtering to improve the mechanical and anti-oxidation properties under 1200 °C steam environment. The phase composition and the micro and macro morphologies of the specimens were investigated by X-ray diffraction (XRD), energy dispersive spectroscopy (EDS), scanning electron microscope (SEM) and optical-microtopography (OM) analyses, and the high-temperature oxidation behavior was evaluated at 1200 °C steam environment for 3000 s. In this paper, there exists a positive correlation between thickness and Vickers hardness (HV), and a negative correlation between surface roughness and bonding force. Radial tensile was introduced to investigate the deformation-resistant performance, and the displacement of the Cr-coated specimen was as low as 2.32 mm, which was much lower than the uncoated zircaloy cladding (3.05 mm). Different thicknesses of Cr coatings were deposited to investigate the oxidation degrees of zircaloy cladding under a high-temperature steam environment. The optimal 6 μm Cr-coated zirconium alloys cladding exhibited an excellent anti-oxidation property, and the weight gain was as low as ~4.12 mg/cm2, which was almost one-third of the uncoated specimen.

Analysis of fuel rod behavior during loss-of-coolant accidents using the BISON code: Fuel modeling developments and simulation of integral experiments

In the first part of this work, we discussed BISON fuel performance code developments for Zircaloy cladding behavior under loss-of-coolant accident (LOCA) conditions and validation to separate-effects experiments. In this paper, we present modeling developments for UO2 fuel behavior during LOCAs and validation to integral experiments. Code developments are related in particular to modeling axial relocation of fuel fragments during cladding ballooning. Code validation is performed against three integral fuel rod LOCA tests from the Halden IFA-650 series, covering fresh and pre-irradiated fuel rods with and without significant axial fuel relocation observed. Calculated results are systematically compared to experimental data of fuel rod inner pressure evolution during the LOCA transients, time to cladding burst and post-test cladding diameter profile. Comparisons in terms of rod pressure evolution and time to burst appear adequate. Also, the ability to reproduce axial relocation of fuel fragments is demonstrated. Calculated cladding diameter profiles are shown to be considerably sensitive to the selection of the burst failure criterion, highlighting the difficulty to accurately evaluate cladding strains at burst during postulated LOCA scenarios with fuel performance codes.

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.

Irradiation-induced swelling of pure chromium with 5 MeV Fe ions in the temperature range 450–650 °C

A surface coating using pure chromium has been proposed to increase the accident tolerance of Zircaloy cladding in pressurized water reactors. However, there is not much irradiation experience with Cr alloys and especially pure Cr. In the present study, pure chromium was irradiated with 5 MeV Fe ions to 50 peak dpa (displacements per atom) at temperatures of 450, 500, 550, 600 and 650 °C. Then irradiation at the peak swelling temperature of 550 °C was conducted to 50, 100, 150 peak dpa. Swelling at 50 dpa was observed over the entire temperature range studied, 450–650 °C, but appeared to be decreasing strongly at the temperature boundaries of the experiment. After an initial transient of rapid swelling, chromium was observed to swell at a rate of ~0.03–0.04%/dpa (up to 120 local dpa), which is much lower than pure Fe at 0.2%/dpa. This low swelling rate was found to be relatively insensitive to dpa rate, which varied by a factor of ~2 over the depth of data collection. Swelling was observed to begin quickly with an incubation period less than 10 dpa. Self-organization in the form of void ordering was observed to be developing at 50 dpa, becoming better defined with increasing dose. The void alignment direction is determined to be the 〈111〉 axial direction.

Effect of initial coating crack on the mechanical performance of surface-coated zircaloy cladding

In this paper, the mechanical performance of cracked surface-coated Zircaloy cladding, which has different coating materials, coating thicknesses and initial crack lengths, has been investigated. By analyzing the stress field near the crack tip, the safety zone range of initial crack length has been decided. In order to determine whether the crack can propagate along the radial (r) or axial (z) directions, the energy release rate has been calculated. By comparing the energy release rate with fracture toughness of materials, we can divide the initial crack lengths into three zones: safety zone, discussion zone and danger zone. The results show that Cr is suitable coating material for the cladding with a thin coating while Fe–Cr–Al have a better fracture mechanical performance in the cladding with thick coating. The Si-coated and SiC-coated claddings are suitable for reactors with low power fuel elements. Conclusions in this paper can provide reference and guidance for the cladding design of nuclear fuel elements.

Dynamic PRA-Based Estimation of PWR Coping Time Using a Surrogate Model for Accident Tolerant Fuel

This study proposes an interpolation-based response surface surrogate methodology to manage a large number of scenarios in dynamic probabilistic risk assessment. It adopts the shape Dynamic Time Warping algorithm to cluster the interpolation neighborhood from time series sample data. The interpolation method was adapted from Taylor Kriging to allow a reduced-order model of the Taylor series. In order to demonstrate its applicability to complex issues in risk assessment for nuclear engineering, an example risk response surface to estimate emergency core cooling system (ECCS) criteria for triplex silicon carbide (SiC) accident-tolerant fuel was constructed. The response surface was exploited to estimate the cumulative failure probability of the fuel cladding structure due to the uncertainties in operator actions and safety systems. The functional failures were assessed based on a combination of individual layer failures computed by coupling Risk Analysis Virtual Environment software with a pressurized water reactor 1000-MW(electric) RELAP5 model and the in-house fuel performance assessment module. Results showed that SiC cladding failure probability spiked less than 1 min after a large-break loss-of- coolant accident whenever the current ECCS criteria for Zircaloy-4 (Zr-4) cladding was used. However, it still provides an increased safety margin of three orders of magnitude compared to Zr-4. This positive margin could be utilized to relax active ECCS requirements by allowing deviations of up to 450 s in its actuation time. The proposed surrogate methodology generated a response surface of SiC cladding failure probability reasonably well, with a significant savings of computation time. This methodology is expected to be useful in the analysis of system response with complex uncertainty sources.

In-situ deformation measurement of Zircaloy-4 cladding tube under various transient heating conditions using optical image analysis

In a loss of coolant accident (LOCA), fuel cladding deformation is an important factor in terms of maintaining coolable geometry. In order to simulate the cladding deformation behavior, experimental research and the development of a fuel performance code have been conducted. In addition, relevant experimental data concerning the major phenomena for validating the developed code are required. Therefore, in this study, an experimental apparatus named ‘DIMAT’ (Deformation In-situ Measurement Apparatus by image-analysis Technique) is developed to produce high temperature cladding tube deformation data. In order to simulate various accident conditions, an infrared (IR) furnace and pressure injection equipment were assembled. Additionally, the cladding specimen in the IR furnace was clearly recorded by application of high temperature resistant black spray paint and modification of a quartz tube. Some uncertainty factors that might arise due to the paint application were considered and evaluated. After the deformation behavior in a scenario was recorded, the in-situ hoop strain of the cladding was calculated through the developed image analysis method. To ensure accuracy of the established image analysis system, a validation experiment and uncertainty evaluation were performed. Finally, the burst experiments were carried out at low and high heating rates of 1, 14, and 28 °C/s, and consequently the in-situ high temperature strain rate was successfully measured in each experimental case.

Analysis of fuel rod behavior during loss-of-coolant accidents using the BISON code: Cladding modeling developments and simulation of separate-effects experiments

BISON is a finite-element based, multidimensional fuel performance code developed at Idaho National Laboratory. In this paper, we present BISON modeling developments for Zircaloy nuclear fuel cladding behavior under loss-of-coolant accident (LOCA) conditions and code validation to separate-effects experiments. Code developments include models for cladding high-temperature creep, crystallographic phase transition, high-temperature steam oxidation and failure due to burst. Code validation is performed against a substantial number of experimental tests for ballooning and burst behavior of Zircaloy claddings under LOCA conditions. Validation calculations are completed using 2D axisymmetric models. Additionally, we present a 3D calculation that demonstrates simulation of the cladding response in presence of azimuthal temperature variations. Calculated results are systematically compared to experimental data of cladding temperature, inner pressure and time at burst failure. Several of the presented BISON validation cases were developed in the framework of the IAEA Coordinated Research Project FUMAC. BISON modeling developments for UO2 fuel behavior under LOCA conditions and code validation to integral fuel rod experiments are presented in a complementary paper.

System code evaluation of near-term accident tolerant claddings during pressurized water reactor station blackout accidents

Following the Fukushima Daiichi nuclear accident in 2011, researches on Accident-tolerant fuels (ATFs) are currently of high interest in not only the nuclear industry but also governmental and international organizations. In this work, a quantitative evaluation of the performance of monolithic FeCrAl cladding and Cr-coated Zircaloy cladding has been performed for Pressurized Water Reactor (PWR) Station Blackout (SBO) accidents. A generic PWR model has been built in system thermal-hydraulics code TRACE based on the Surry Nuclear Power Station with counter-current natural circulation modelling capability for hotleg and steam generator U-tube components during the accidents. The base model results are then compared to MELCOR and RELAP simulations to verify the system component implementation in TRACE. Two PWR SBO scenarios were investigated, including: short-term SBO and long-term SBO with early reactor coolant pump (RCP) seal failure. These scenarios are defined to be very similar to the accidents studied in the State-of-the-Art Reactor Consequence Analysis (SOARCA) project. TRACE code is modified to reflect the oxidation kinetics of FeCrAl and Cr-coating. Larson-Miller creep rupture model is also implemented in TARCE using its built-in control systems to simulate the creep rupture of hotlegs. The comparison between the TRACE models with and without the counter-current flow modeling resulted in significant difference when comparing ATF cladding to Zircaloy for short term SBO, while it marginal impacted the performance during long term SBO with RCP seal failure. For short term SBO, both ATF cladding underwent hot leg creep rupture ~20 min after Zircaloy cladding. While Zircaloy and Cr-coated cladding had generated significant amount of hydrogen gas (>10 kg) before the creep rupture event, FeCrAl cladding had only generated <0.5 kg of hydrogen gas. For long term SBO with RCP seal failure, significant hydrogen generation and fuel melting was predicted before hot leg creep rupture for the ATF cladding while providing only 10–20 min additional coping time compared to Zircaloy.

Pellet-cladding mechanical interaction analysis of Cr-coated Zircaloy cladding

The thermal–mechanical response of Cr-coated Zr cladding during power ramp condition is of critical importance for a better understanding of its reliability and security in ATF designing. The pellet-cladding mechanical interaction (PCMI) performance of Cr-coated Zircaloy cladding was analyzed in this study using multi-step modeling strategy to capture the base-irradiation conditions of full-length fuel and the local transient PCMI responses during power ramp conditions. The traditional 1.5D fuel performance code, COPERNIC, and the commercial finite element (FE) code, ABAQUS, were combined to effectively reduce the computational demands while ensuring accuracy of local PCMI prediction. ABAQUS has been enhanced to support thermal, mechanical and irradiation behavior modeling of the coated cladding by adding a set of models representing fuel, cladding and coating in-pile behaviors using user-subroutines. Then, the PCMI responses of three Cr-coated claddings were analyzed under both steady and power ramp conditions. Parametric studies were carried out to demonstrate the effects of several key factors, such as coating thickness, burn-up level and transient power increment, etc., on PCMI behavior of the Cr-coated cladding, which may provide fuel designers and engineers a deeper understanding in thermal–mechanical responses of the Cr-coated cladding.

A three-dimensional axial fuel relocation framework with discrete element method to support burnup extension

Under normal operation, oxide nuclear fuel pellets subjected to high thermomechanical loading and irradiation damage can develop complex crack networks, separating the fuel pellet into a pile of fragments. During a Loss of Coolant Accident (LOCA), these pellet fragments, or fine powder especially when originating from the high-burnup fuel pellet periphery, may relocate axially into the ballooned region of the zircaloy cladding. In this work, the axial relocation process of fuel fragments and powder is modeled based on the three-dimensional Discrete Element Method (DEM) in the Abaqus software. A Voronoi tessellation technique is adopted to generate random convex polyhedrons in order to mimic a coarse fragmentation pattern. A parametric study in terms of the mean fragment size from 1.3 mm to 3.0 mm was conducted to investigate its effect on the axial profile of fuel mass fraction and filling ratio after fuel relocation. Then a sensitivity study of the randomness in fragmentation on axial fuel relocation was performed. For high-burnup fuels, in order to characterize the porous High Burnup Structure (HBS), the pellet rim was discretized into separate spherical shape particles. The effect of powder fraction on fuel relocation was demonstrated, and the results indicated that higher powder fraction can significantly increase the fuel mass fraction and filling ratio level in the ballooned region, which showed qualitative agreement with experimental observations. Finally, the verification against a one-dimensional theoretical relocation model was conducted to verify the robustness and high fidelity features of the DEM model in simulating axial fuel relocation under LOCA. The modeling approach serves as a mechanistic framework to support future burnup extensions for the oxide nuclear fuel.

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.

Plastic instability of mechanical bulging process for Zircaloy-4 cladding tube based on localized necking criterion

Plastic instability of mechanical bulging process, different from the traditional tube hydroforming, is developed for Zircaloy tube based on localized necking criterion. Based on the observation of published experiments of bulging, an elementary analyzed model of plastic deformation is proposed and applied for examining the variation of bulging process and its limitation. The developed model comprises two deformation regions, one is bulged part with friction effect and the other is flattened part but without friction effect. All mentioned process parameters, such as contact friction, material properties (strain hardening exponent and normal anisotropy), and process geometries conditions (tube radius and thickness), are taken into account for in such mechanical bulging model. Plastic instability in bulging is examined using localized necking criterion based on the proposed model. A published experiment was initially used to verify the present model. Then, the effects of process parameters on the variation of bulging limit are investigated. The developed model allowed us to better understand the failure mechanism through the localized necking criterion, and also can be defined as a sign of fracture for a bulged tube. Moreover, a proposal will be made to examine the limitation of bulging for practical needs.

Thermal creep behavior of CZ cladding under biaxial stress state

Thermal creep is a key property of zircaloy cladding. CZ developed by CGN is a new zircaloy used as PWR fuel cladding. This research is devoted to investigating the thermal creep behavior of CZ and build the thermal creep model of CZ. Twenty internal pressure creep tests were conducted, and the ranges of temperature and Tresca stress were 320–430 °C and 70–300 MPa, respectively. Real-time creep data were analyzed by separating primary creep and steady-state creep. Based on Soderberg model and creep test data, CZ thermal creep model is derived. As a whole, the mean value and the standard deviation of P/M of CZ saturated primary creep strain are very close to these from steady-state creep rate, however, the predictive effect of primary creep is less satisfactory. Four conditions, where there exists large deviation between predicted values and test data, are 320 °C and 300 MPa, 350 °C and 190 MPa, 380 °C and 160 MPa, 380 °C and 190 MPa, respectively. As primary creep was much smaller than steady-state creep in long-time operation, the thermal creep model built can be applied to predict the thermal creep behavior of CZ cladding.

Burst behavior of nuclear grade FeCrAl and Zircaloy-2 fuel cladding under simulated cyclic dryout conditions

A novel experiment to simulate cyclic dryout in boiling water reactors has been developed to better understand the performance of nuclear grade FeCrAl cladding in a BWR during dryout conditions caused by an Anticipated Operational Occurrence or Anticipated Transient Without SCRAM - both of which are Design Basis Accidents. Internally pressurized C26 M FeCrAl alloy cladding and Zircaloy-2 cladding were subjected to rapid 300°-650 °C thermal cycling in a steam environment; actual maximum temperatures were found to vary between materials but were always above 650 °C. In the range of 32–55 MPa hoop stress, Zircaloy-2 cladding burst within 1–16 cycles (about 100 s of dryout duration above 600 °C), while at 76 MPa hoop stress, C26 M cladding remained virtually undeformed after completing 54 cycles (over 1000 s of dryout duration above 600 °C). Higher temperature 300°-700 °C and 300°C–800 °C cycling experiments had to be performed to induce C26 M burst – failure occurred after 20 cycles in the former and during the first cycle in the latter. Zircaloy-2 and C26 M failure criteria were used to generate hoop stress specific dryout lifetimes. Overall, the simulated cyclic dryout experiments show that nuclear grade C26 M cladding has significantly enhanced survivability under dryout conditions relative to Zircaloy-2.

3D Analysis of a multipurpose modular reactor for small energy production

The Fixed Bed Nuclear Reactor (FBNR) is a modular reactor for small energy needs (70-120 MWel). FBNR is pressurized water reactor, but uses GEN-IV reactors fuel. UO2 kernels of 500 μm diameter with 25 μm thick SiC cladding are embedded in graphite matrix form the spherical fuel elements with a diameter of 15 mm and covered by 300 μm thick Zircaloy-4 cladding. FBNR has high level safety features and operates without On-site Refueling. Approximately 1.62 million spherical fuel elements are suspended in a core height of 200 cm and inner and outer diameters of 32 cm and 224 cm by upwards flowing light water coolant with 280°C entry and 340°C exit temperatures. The unit lattice geometry is dodecahedron with fuel element in center and surrounded with light water coolant-moderator.In the present work, 9% enriched UO2 fuel is used. Neutronic analysis has been performed with the MCNPX-2.7.0 code. As, it is not practical to tackle millions of fuel elements separately, at first run unit fuel cell calculations are performed in spherical geometry for one single lattice consisting of separate fuel, cladding and volume-equivalent moderator regions. In the second run, the fuel cell is homogenized. Infinite cell calculations have been executed for variable moderator/fuel (H/235U) ratios in order to determine under- and over-moderated criticality values. The criticality values for the homogenized cell geometry are slightly lower and remains <5% range, that is, on the conservative side, which is acceptable for the purpose of this study. This allows us to homogenize the entire core for the full 3-D reactor with core and reflector regions. Reactor criticality increases with increasing H/235U ratios in the, under moderated region, where H/235U = 18 is selected for further investigations. Temporal variation of the reactor criticality is pursued for full reactor power of 210 MWth (~70 MWel net) with keff = 1.36 at the beginning of life down to keff = 1.02 over 1260 days with a plant factor of 1.0. The final fuel burn up was found as 57.3 GWd/ton.

Measurement of Zircaloy-4 cladding tube deformation using a three-dimensional digital image correlation system with internal transient heating and pressurization

In this study, multi-dimensional (Multi-D) deformation of a Zircaloy-4 cladding tube under internal transient heating and pressurization conditions was measured by experimental setup named ‘Experimental setup to investigate high temperature deformation of cladding tube under internal pressurizing and heating (EIGEN)’ and three-dimensional digital image correlation (3D DIC) system. The ‘EIGEN’ experimental setup was designed to investigate high temperature deformation of the cladding tube under inert gas atmosphere. The 3D DIC system based on stereovision was further developed to be applied to high temperature by using high temperature resistant speckle pattern and bandpass filter. To evaluate measurement accuracy of the 3D DIC system, validation works of 3D DIC calibration were performed with proposed methodology. Additionally, a 3D finite element model (FEM) of the ‘EIGEN’ experiment was developed to evaluate the results of the 3D DIC. Finally, ‘EIGEN’ experiment was conducted under several fill pressure conditions such as 30, 50, 60, 70 bars. As a result, the established 3D DIC system could successfully measure in-situ and real-time large deformation such as ballooning and rupture as well as full-field multi-D deformation of the cladding tube at the high temperature under the inert atmosphere. Multi-D strains from the 3D DIC showed good agreement with calculated ones of the 3D FEM, and they were also compared with 3D laser scanning results for validation.

Deep neural network based prediction of burst parameters for Zircaloy-4 fuel cladding during loss-of-coolant accident

BackgroundUnderstanding the behaviour of nuclear fuel claddings by conducting burst test on single cladding tube under simulated loss-of-coolant accident conditions and developing theoretical cum empirical predictive computer codes have been the focus of several investigations. The developed burst criterion (a) assumes symmetrical deformation of cladding tube in contrast to experimental observation (b) interpolates the properties of Zircaloy-4 cladding in mixed α+ β phase (c) does not account for azimuthal temperature variations. In order to overcome all these drawbacks of burst criterion, it is reasoned that artificial intelligence technique may be a better option to predict the burst parameters. MethodsArtificial neural network models based on feedforward backpropagation algorithm with logsig transfer function are developed. ResultsNeural network architecture of 2-4-4-3, that is model with two hidden layers having four nodes in each layer is found to be the most suitable. The mean, maximum, and minimum prediction errors for this optimised model are 0.82%, 19.62%, and 0.004%, respectively. ConclusionThe burst stress, burst temperature, and burst strain obtained from burst criterion have average deviation of 19%, 12%, and 53% respectively whereas the developed neural network model predicted these parameters with average deviation of 6%, 2%, and 8%, respectively.