Pellet-cladding Mechanical Interaction Research Articles

Three-Dimensional Modeling of Thermal-Mechanical Behavior of Accident Tolerant Fuels

Considering the safety issues of the traditional UO2-Zr fuel, a variety of accident-tolerant fuel (ATF) candidates have been proposed in recent years. Among the several ATFs, U3Si2, and UN are the two promising candidates for fuel materials owing to their high thermal conductivity and high uranium density. The FeCrAl alloy and the SiC/SiC composite material are the two promising candidates for cladding owing to their high oxidation resistance and high strength. In order to quantitatively evaluate the performance of ATFs, this study summarizes the physical models of typical ATF cladding materials (FeCrAl and SiC) and pellet materials (UN and U3Si2). Then a three-dimensional non-linear finite element method is applied to simulate the thermal-mechanical behavior of several typical fuel-cladding combinations, including UO2-FeCrAl, UN-FeCrAl, U3Si2-FeCrAl, U3Si2-Zr, and U3Si2-SiC. The important physical quantities, such as the fuel centerline temperature, the deformation of the pellet and the cladding as well as the pellet-cladding mechanical interaction (PCMI) were studied. The fission gas release model was also verified and improved.

Multiphysics analysis of thorium-based fuel performance under reactor normal operating and accident conditions in a light water reactor

The ThO2 fuel has higher thermal conductivity and melting boiling point than the UO2 fuel, which is beneficial to the fast removal of heat and the improvement of fuel melt margin. In this paper, the material properties and thermodynamic behaviors of thorium-based fuel were firstly reviewed, and then the thermal physical properties and the fuel behavior models of Th0.923U0.077O2 fuel and Th0.923Pu0.077O2 fuel have been implemented in fuel performance analysis code FRAPCON and FRAPTRAN. Finally, the performances of Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and UO2 fuel under both normal operating conditions and transient conditions (both RIA and LOCA) are analyzed and compared. The thorium-based fuels are found to have lower fuel centerline temperature than that of the UO2 fuel under normal operating condition, and a delayed pellet-cladding mechanical interaction (PCMI) under both normal operating and transient conditions. Furthermore, the fission gas release, cladding strain and internal fuel energy under transient conditions are found to be lower too. Lastly, the cladding displacement and temperature under transient conditions are also compared. The thorium-based fuel was found to have a higher safety margin and accident resistance than the UO2 fuel under both normal operating condition and transient accident conditions.

Mechanical failure of high-burnup fuel rods with stress-relieved annealed and recrystallized M-MDA cladding under reactivity-initiated accident conditions

ABSTRACT To evaluate the effects of the hydride morphology and initial temperature of fuel cladding on the pellet-cladding mechanical interaction failure under reactivity-initiated accident (RIA) conditions, RIA-simulated experiments were performed on high-burnup fuels with stress-relieved annealed (SR) and recrystallized (RX) M-MDATM cladding at room and high (~280°C) temperatures. The results demonstrated that the failure-limit trend of RX-cladded fuels being lower than that of SR-cladded fuels for a similar hydrogen content holds up to at least about 700 wtppm. The observation of the fracture surfaces of failed RX cladding suggests a contribution of radially oriented hydrides to the crack formation and/or penetration, which coincides with the aforementioned failure-limit trend. The temperature effect, namely the failure-limit rise at a high temperature, is evident irrespective of the hydride morphology, while the degree of the temperature effect decreases as the hydrogen content increases.

Preliminary prediction for survival time of fuel rod under critical heat flux

In commercial water-cooled reactor safety analysis, the fuel rod is assumed to fail and release its fission gas inventory when it reaches critical heat flux (CHF). For this reason, CHF serves as a regulatory limit for all water-cooled reactors. However, according to existing in-pile experimental data, this assumption is conservative and may over-constrain the reactor core performance. With the advent of accident tolerant fuels (ATFs), particularly their ability to structurally withstand higher temperatures, the true margin to fuel failure after reaching CHF should be re-visited. To this end, preliminary fuel performance during CHF and Post-CHF is simulated and fuel rod survival time based on different cladding failure criteria is formulated. The post-CHF fuel performance is governed by cladding temperature, oxidation, pellet-clad mechanical interaction (PCMI) and hydride formation. The analysis includes impact of cladding temperature, fuel burnup and presence of Cr-coating as a near term ATF concept. The fuel performance analysis for fresh fuel implies that the cladding survival times of few days are possible and the cladding oxidation and plastic strain limit are the likely failure modes. For burned fuel, PCMI and hydride formation are the likely limiting failure modes that could decrease the survival time to less than a day. The addition of Cr coating reduces hydride formation and improves high temperature creep performance of the cladding and it can potentially prolong the survival times to several days. The estimated large survival times motivate further investigations, particularly, expansion of available data in areas of cladding mechanical performance and hydrogen pickup in the temperature range relevant to post-CHF film boiling regime to better inform the fuel performance predictions.

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.

Initial development of an RIA envelope for dispersed nuclear fuel

A reactivity insertion accident (RIA) is a design basis accident in which reactivity is rapidly injected as a result of a control rod ejection or blade drop scenario. The resultant power increase can result in fuel rod failure and subsequent release of radioactive material into the reactor’s primary system. For light water reactors, failure under RIAs is typically a result of fuel melt, pellet-cladding interaction, or rod over-pressurization. Fuel melt occurs when the fuel system is unable to transport heat out of the fuel system. Pellet-cladding interaction occurs when an aggressive fuel pellet expansion pushes on an embrittled cladding beyond its strain limit. Rod over-pressurization occurs when a fuel rod exceeds the critical heat flux and departs from nucleate boiling, causing the rod internal pressure to rapidly increase, exceeding the systems pressure, and balloon until it bursts. To prevent failure, regulators have developed safety requirements that limit the injected enthalpy based on the state of the fuel system. However, dispersed fuel could enable the removal or relaxation of regulator-imposed safety criteria. Dispersed fuel embeds fuel particles in a highly conductive metal. Dispersed particles can have a particle radius of 1 µm to greater than 100 µm. Traditional fuel systems are plagued by poor thermal conductivity, especially at higher burnups, thus reducing the ability of the fuel system to respond to an RIA. However, dispersed fuel could improve the ability of the fuel system to transport injected energy away from the fissile material and into the coolant more efficiently. Additionally, dispersing the fuel particles mitigates hard contact that typically occurs in the traditional zirconium-uranium dioxide fuel system, thus removing pellet-cladding mechanical interaction as a potential failure mechanism. This paper evaluates fuel particles dispersed in a metal matrix using the BISON fuel performance code to develop an initial failure threshold based on melt temperature of the fuel particle and/or cladding material as a function of fission density for beginning-of-life and end-of-life conditions. BISON results show that (1) reducing the size of the particles allows fission density to increase as a function of time, (2) particle-to-particle proximity must be considered to evaluate the limiting conditions, and (3) improving the fuel particle thermal conductivity improves thermal performance.

Mechanical failure of fresh nuclear grade iron–chromium–aluminum (FeCrAl) cladding under simulated hot zero power reactivity initiated accident conditions

The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiated by a rapid increase in reactivity which causes an increase in fission rate and fuel temperature. One potential mode of fuel system failure during RIA is pellet cladding mechanical interaction (PCMI) due to the rapid thermal expansion of the fuel pellet. Simulated PCMI experiments were performed by rapidly pressurizing iron-chromium-aluminum (FeCrAl) cladding tube samples using a hydraulic modified burst test system, causing the specimens to burst under a biaxial loading path. Deformation and rupture of the specimens were tracked with a telecentric lens and high-speed camera system. Outer surface strains were calculated using digital image correlation (DIC) on speckle patterns painted on the specimens’ outer surfaces. Experiments were conducted at approximately 275 °C, representative of hot zero power RIA conditions. The failure hoop strain was DIC-calculated between approximately 1.8–3.4%, corresponding to quasi-static energy depositions of approximately 110–260 calories per gram UO2 assuming initial pellet-cladding contact. These conditions are very similar to the proposed energy deposition limit of 150 calories per gram UO2 for unirradiated zirconium-based cladding in US Nuclear Regulatory Commission (NRC) Draft Regulatory Guide 1327. A very small strain rate dependence was observed in the data, with the magnitude of the failure hoop strain decreasing slightly with increasing strain rate. This observed dependence may be relevant because the FeCrAl cladding strain rate will be approximately 20% higher than in zirconium-based cladding due to the harder neutron spectrum and resultant shorter neutron generation time.

Analysis of cladding failure in a BWR fuel rod using a SLICE-DO model of the FALCON code

Cladding failure in a fuel rod during operation in a BWR is analyzed using a FALCON code-based model. Comparative calculation with a neighbouring, intact rod is presented, as well. A considerable ‘hot spot’ effect in cladding temperature is predicted with the SLICE-DO model due to a thermal barrier caused by the localized crud deposition. Particularly significant overheating is expected to occur on the affected side of the cladding of the failed rod, in agreement with signs of significant localized crud deposition as revealed by Post Irradiation Examination. Different possibilities (criteria) are checked, and Pellet-Cladding Mechanical Interaction (PCMI) is shown to be one of the plausible potential threats. It is shown that PCMI could lead to discernible concentrated inelastic deformation in the overheated part of the cladding. None of the specific mechanisms considered can be experimentally or analytically identified as an only cause of the rod failure. However, according to the current calculation, a possibility of cladding failure by PCMI cannot be excluded if the crud thickness exceeded 300 μm.

Development of Equivalent Beam Model of High Burnup Spent Nuclear Fuel Rods under Lateral Impact Loading

Spent nuclear fuel (SNF) is nuclear fuel that has been irradiated and discharged from nuclear reactors. During the whole management stages of SNF before it is, in the end, disposed in a deep geological repository, the structural integrity of fuel rods and the assemblies should be maintained for safety and economic reasons. In licensing applications for the SNF storage and transportation, the integrity of SNF needs to be evaluated considering various loading conditions. However, this is a challenging task due to the complexity of the geometry and properties of SNF. In this paper, a simple and equivalent analysis model for SNF rods is developed using model calibration based on optimization and process integration. The spent fuel rod is simplified into a hollow beam with a homogenous isotropic material, and the model parameters thus found are not dependent on the length of the reference fuel rod segment that is considered. Two distinct models with different interfacial conditions between the fuel pellets and cladding are used in the calibration to account for the effect of PCMI (Pellet-Clad Mechanical Interaction). The feasibility of the models in dynamic impact simulations is examined, and it is expected that the developed models can be utilized in the analysis of assembly-level analyses for the SNF integrity assessment during transportation and storage.

Research on performance enhancement of nuclear fuel with SiC cladding by using high thermal conductivity fuels

The engineering application of the fuel rod system consisting of traditional UO2 pellets and SiC accident tolerant fuel (ATF) cladding is challenged by high fuel operating temperatures, large fission gas release rates and high risks of brittle failure of the cladding. In this study, a preliminary research was carried out for performance enhancement of the fuel rods consisting of SiC cladding by replacing UO2 with high thermal conductivity fuels, including UO2/BeO, U3Si2 and UN/U3Si2. Thermo-mechanical modeling was conducted with the consideration of burnup effects for the three fuels with high thermal conductivity. An upgrade of the nuclear fuel analysis code FRAPCON4.0 was conducted to extend the simulation capability by implementing the new fuel models. Thermo-mechanical behavior of the original and improved fuel rods was simulated by the upgraded FRAPCON4.0. The simulation results show that using high thermal conductivity fuels can significantly reduce the fuel temperature and fission gas release. Furthermore, the fuel gaseous swelling is greatly mitigated due to the reduced fuel temperature, which is beneficial to prevent SiC cladding from brittle failure caused by pellet-cladding mechanical interaction. By comparison, UN/U3Si2 fuel presents the most promising and desired thermo-mechanical performance among all high thermal conductivity fuels investigated in this study.

Dynamic finite element analysis of segmented mandrel tests of hydrogenated E110 fuel cladding tubes

Pellet-cladding mechanical interaction (PCMI) occurs in the nuclear fuel rods under normal operation caused by pellet swelling and cladding creep, and in transients and during reactivity initiated accidents (RIA) because of the differences in thermal expansion. During normal reactor operation, hydrogen is generated by radiolysis and corrosion processes. Hydrogen easily dissolves to interstitial locations in the zirconium lattice, and at higher concentrations it forms various hydrides. The ductility of the cladding alloys depends on the amount of hydrogen absorbed.Cladding tube samples made of Russian Zr1%Nb alloy E110 were loaded radially by segmented tools (mandrels) to simulate the PCMI experimentally and evaluate the ductility of the cladding samples under various conditions. A finite element model was developed in MSC Marc and Mentat to simulate the mandrel tests using a non-linear plastic approximation. Numerical simulation of the mandrel tests was successful, the calculated and the measured force-displacement curves fit well. A simple numerical transformation was used to describe the different ductility of treated cladding samples. A relationship was found between the transformation parameter and the hydrogen content of the samples.

Fracture-mechanics-based evaluation of failure limit on pre-cracked and hydrided Zircaloy-4 cladding tube under biaxial stress states

ABSTRACTTo better understand the failure limit of fuel cladding during the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA), pre-cracked and hydrided cladding samples with base metal final heat-treatment status of cold worked (CW) and recrystallized (RX) were tested under biaxial stress conditions (strain ratios (εz/εθ) of 0 and 0.5). Displacement-controlled biaxial-expansion-due-to-compression (biaxial-EDC) tests were performed to obtain the hoop strain at failure (failure strain) of the samples. The conversion of the failure strains to J-integrals at failure (JIC) by finite-element analysis involving data of stress-relieved (SR) cladding specimens from our previous study revealed that the failure limit in the dimension of JIC unifies the effects of pre-crack depth. About 30–50% reduction in the JIC value was observed as the strain ratio increased from 0 to 0.5 irrespective of the annealing type, pre-crack depth, and hydrogen content. The rate of fractional decreases of JIC with increase of hydrogen content is in the order of CW>SR>RX, which are essentially independent of strain ratio for the CW and SR samples. The results were incorporated into the failure prediction model of the JAEA’s fuel performance code in the form of a correction factor that considers the biaxial loading effect.

Hot Isostatic Pressing (HIP): A novel method to prepare Cr-doped UO2 nuclear fuel

ABSTRACTThe addition of Cr2O3 to modern UO2 fuel modifies the microstructure so that, through the generation of larger grains during fission, a higher proportion of fission gases can be accommodated. This reduces the pellet-cladding mechanical interaction of the fuel rods, allowing the fuels to be “burned” for longer than traditional UO2 fuel, thus maximising the energy obtained. We here describe the preparation of UO2 and Cr-doped UO2 using Hot Isostatic Pressing (HIP), as a potential method for fuel fabrication, and for development of analogue materials for spent nuclear fuel research. Characterization of the synthesised materials confirmed that high density UO2 was successfully formed, and that Cr was present as particles at grain boundaries and also within the UO2 matrix, possibly in a reduced form due to the processing conditions. In contrast to studies of Cr-doped UO2 synthesised by other methods, no significant changes to the grain size were observed in the presence of Cr.

The effect of base irradiation on failure behaviors of UO2 and chromia-alumina additive fuels under simulated reactivity-initiated accidents: A comparative analysis with FEMAXI-8

This paper reports an analysis on the base irradiation behavior of the chromia-and-alumina-doped boiling water reactor (BWR) rod irradiated to 64 GWd/t in Oskarshamn-3, Sweden, and subjected to the reactivity-initiated-accident (RIA) test OS-1, which resulted in a fuel failure due to pellet-cladding mechanical interaction at the lowest fuel enthalpy increase in all the BWR tests ever performed. The inverse calculation which utilized post-irradiation examination data as its constraint conditions revealed that the OS-1 rod had very likely experienced higher cladding stress due to higher swelling rate during base irradiation than other BWR rods subjected to previous RIA tests. Thus, the OS-1 rod was prone to experience enhanced radial-hydride formation. The significant difference in the hoop-stress of more than 50 MPa discriminates the OS-1 rod from other BWR rods and supports the interpretation that enhanced radial-hydrides formation differentiates the failure behavior observed in the test OS-1 from the previous BWR-fuel tests.

Investigation of Pellet-Clad Mechanical Interaction in Failed Spent PWR Fuel

Investigation of Pellet-Clad Mechanical Interaction in Failed Spent PWR Fuel

Multiphysics modeling of thorium-based fuel performance with a two-layer SiC cladding in a light water reactor

The performance of thorium-based mixed oxide fuel ((Th,U)O2 fuel and (Th,Pu)O2 fuel) with a two-layer SiC cladding in a light water reactor have been investigated by finite element multiphysics modeling method. The material properties of specific thorium-based Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and SiC cladding were firstly reviewed and implemented into the multiphysics model. Then the performance of Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and UO2 fuel separately combined with Zircaloy cladding and two-layer SiC cladding have been investigated and compared under PWR normal operation conditions. Finally, the thorium-based fuel with Zircaloy cladding was found to decrease the fuel centerline temperature, especially for the Th0.923U0.077O2 fuel, but with a much earlier gap closure time. And the two-layer SiC cladding was found to effectively mitigate the pellet-cladding mechanical interaction (PCMI) but also greatly increase the fuel centerline temperature. So the combination of the thorium-based fuel with two-layer SiC cladding is expected to improve the reactor safety by keeping a moderate fuel centerline temperature and meanwhile greatly delaying the PCMI time.

Fuel performance optimization of U3Si2-SiC design during normal, power ramp and RIA conditions

Benefited from its higher mechanical strength and much-improved oxidation resistance at very high temperatures (>1200 °C), SiC cladding is one of the promising accident tolerant fuel (ATF) concepts. Considering the thermal conductivity degradation of SiC cladding after irradiation and its negligible creep rate, adopting conventional UO2 with SiC cladding can lead to aggravated fission gas release and elevated fuel centerline temperatures. In this work, U3Si2 fuel with its high thermal conductivity is considered in place of UO2, when adopting SiC cladding. First, the preliminary full core simulation for U3Si2-SiC was conducted to show its holistic fuel performance compared to the current UO2-Zircaloy fuel. Then the most limiting fuel rod was selected based on the calculated chemically vapor deposited (CVD) SiC failure risk from the full core simulation. For the limiting rod, fuel-clad gap optimization was conducted to minimize CVD SiC failure risk. Subsequently, power ramp history was appended to the normal power history and full core simulation analysis for updated U3Si2-SiC design was re-performed to ensure the effectiveness of the chosen gap width. For the limiting rod, reactivity-initiated accident (RIA) analysis at hot zero power (HZP) was analyzed to evaluate its performance and infer its potential failure modes. The results from normal operation and power ramp simulations implied that CVD SiC failure risk was mainly induced by the strong pellet-cladding mechanical interaction (PCMI) and the failure probability of CVD SiC can be effectively minimized to 2.0×10-7 through enlarging the fuel-clad as-fabricated gap width. The leading failure mode for U3Si2-SiC fuel system during RIA changed from fuel melting to Ceramic Matrix Composite (CMC) failure with increasing burnup.

Thresholds for failure of high-burnup LWR fuels by Pellet Cladding mechanical interaction under reactivity-initiated accident conditions

ABSTRACT To contribute to the future updating on the Japanese safety criteria for pellet/cladding mechanical interaction (PCMI) failure of light water reactor fuels under reactivity-initiated accident (RIA) conditions, this paper summarizes the recent important outcomes from research programs with the Nuclear Safety Research Reactor (NSRR). Applicability of current criteria, which are defined as a function of fuel burnup and possibility of introducing another parameter for new criteria were evaluated based on the results of the RIA-simulated pulse irradiation tests, post-test examinations, and supporting analytical work, such as the reevaluation of fuel enthalpies in earlier NSRR experiments. Failure-threshold curves based on cladding hydrogen content as a primary measure of fuel degradation have been proposed as a possible alternative that can be used to judge the occurrence of PCMI failure to ensure conservativeness in a more pertinent manner.

Integrating a multi-layer deformation model in FRAPTRAN for accident Tolerant fuel analysis

Fuel performance analysis capability is needed to assess proposed Accident Tolerant Fuel (ATF) designs. In this work, a deformation model was implemented in FRAPTRAN 2.0 to model a multi-layer ATF concept during both reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA). The modified FRAPTRAN (M−FRAPTRAN) performance was assessed against limited UO2-Zircaloy validation database and the performance was found to be comparable to FRAPTRAN 2.0 and experimental data. Finally, in order to demonstrate the added capabilities in M−FRAPTRAN, two-layer SiC clad design consisting of SiC/SiC ceramic matrix composite (CMC) and monolithic SiC (mSiC) was modeled. During RIA, the onset of pellet clad mechanical interaction can significantly raise the mSiC principal stresses. During LOCA, the SiC potential annealing behavior (recovery of swelling strain) has a significant influence on cladding stress distribution and outer region of mSiC cladding may undergo tensile stress, which can threaten the leak tightness of the cladding during LOCA.

Fracture behavior of recrystallized and stress-relieved Zircaloy-4 cladding under biaxial stress conditions

ABSTRACT Pellet-cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to failure of high-burnup fuel rods. The biaxial stress states of Zircaloy cladding tubes induced by PCMI loading presumably makes the tubes more susceptible to failure. To clarify the influence of the anisotropic mechanical properties of Zircaloy cladding tubes on their fracture behavior under biaxial stress conditions, biaxial tensile tests were performed, and the measured stresses and strains were converted to reduced parameters such as equivalent strain, equivalent stress, and stress triaxiality by using the anisotropic constants of the Hill yield function derived in our previous study. The minimum fracture strains of recrystallized (RX) and stress-relieved (SR) specimens were located where the stress ratio of axial to circumferential is 0.75. The equivalent plastic fracture strains tended to decrease monotonously with increasing stress triaxiality, which is a typical trend observed in ductile fracture, in the range of 0.65–0.78 for both specimens. In the case of SR specimens, however, the analysis with stress triaxiality did not reduce the fracture strains well to a single trend curve, showing that another influential factor is required such as an anisotropic fracture behavior or a plastic behavior deviated from that predicted by the yield function.