Rupture of spent fuel Zircaloy cladding in dry storage due to delayed hydride cracking

Delayed hydride cracking in the Zircaloy alloy has been considered as a possible degradation mechanism of spent nuclear fuel cladding in interim dry storage. Some recent in-core fuel failures indicated that a long axial crack developed in the cladding was a secondary failure by delayed hydride cracking. The aim of this study is to define the effects of hydride reorientation on the failure of Zircaloy cladding. Different hydride orientations, the amount of zirconium hydride and various cracking types, all have been considered for their effects on the crack growth and stability of the cladding, and have been thoroughly discussed in this paper. A finite element computer code, ANSYS, has been used in conjunction with the strain energy density theory. In summary, the crack propagation will be aggravated if the hydride orientation is shifted from the circumferential to the radial direction. For a larger crack length, the zirconium hydride plays an important role in affecting the crack growth because the strain energy density factor increases as the hydride approaches the crack tip. Furthermore, when thermal effects are considered, a compressive stress exists at the inner side of the cladding, while a tensile stress is found at the outer side of cladding, thus resulting in crack propagation from the outer side to the inner side of the cladding. These findings are in accordance with other experimental results in related literature.     

Post-irradiation examinations focused on fuel integrity of spent BWR-MOX and PWR-UO2 fuels stored for 20 years

Post-irradiation examinations (PIEs) of spent BWR-MOX and PWR-UO₂ fuel rods irradiated in commercial LWRs and stored for 20 years were carried out to evaluate fuel integrity during storage. Average burn-up of five fuel rods of the BWR-MOX fuel was about 20 MWd/kgHM and that of the PWR-UO₂ fuel was 58 MWd/kgHM. The PIE items included: (a) visual inspection of the cladding surface, (b) puncture test, (c) ceramographic observation on the pellet and cladding, (d) pellet density, (e) electron probe microanalysis of the pellet, (f) cladding tensile test, (g) hydrogen content and hydride orientation in the cladding, and (h) hydrogen redistribution in the cladding under temperature gradient. The PIE results showed no marked difference in the visual inspection, fission gas release, oxide layer thickness, pellet microstructure, and cladding mechanical properties or hydride orientation after storage. The result of the hydrogen redistribution experiment showed that hydrogen migration had little effect on the fuel integrity during dry storage. Hydrogen migration on the fuel rod for 40 years of storage was evaluated using the heat of transport obtained in the hydrogen redistribution experiment and calculated result showed that hydrogen migration had little effect on the fuel integrity during dry storage.     

Effect of hydride orientation on fracture toughness of Zircaloy-4 cladding

Hydrogen embrittlement is one of the major degradation mechanisms for high burnup fuel cladding during reactor service and spent fuel dry storage, which is related to the hydrogen concentration, morphology and orientation of zirconium hydrides. In this work, the J-integral values for X-specimens with different hydride orientations are measured to evaluate the fracture toughness of Zircaloy-4 (Zry-4) cladding. The toughness values for Zry-4 cladding with various percentages of radial hydrides are much smaller than those with circumferential hydrides only in the same hydrogen content level at 25 °C. The fractograghic features reveal that the crack path is influenced by the orientation of zirconium hydride. Moreover, the fracture toughness measurements for X-specimens at 300 °C are not sensitive to a variation in hydride orientation but to hydrogen concentration.     

A methodology for the evaluation of fuel rod failures under accident conditions

Abstract

Recent studies on the long-term behaviour of high-burnup spent fuel have shown that, under normal conditions of storage, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride cracking, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar regulatory rules have not yet been developed to address failures of fuel rod cladding that could potentially lead to reconfigured fuel geometry under hypothetical transport accidents. At issue is the effect on cladding ductility of potential changes in zirconium hydride morphology during dry storage. Recent studies have shown that above a certain level of cladding hoop stress, the decaying temperature history during dry storage can cause the hydrogen in solid solution to precipitate in the form of radial hydrides, which, depending on their relative concentration, can induce brittle failures in the cladding. From a US regulatory perspective such cladding failures, if they were to cause fuel reconfiguration, could invalidate the cask's criticality and shielding licensing analyses, which are based on coherent geometry. This paper describes a methodology for high-burnup spent fuel to determine the frequency of cladding failure and failure modes under drop accidents, considering end-of-storage spent fuel conditions. The degree to which spent fuel reconfiguration could occur during handling or transport accidents would depend to a large extent on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there are no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, this paper focuses on the development of a methodology for modelling and analysis that deals with this general problem on a generic basis. First, consideration is given to defining accident loading that is equivalent to the bounding hypothetical transport accident of a 9 m drop onto an essentially unyielding surface. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A model of material behaviour, with embedded failure criteria, for cladding containing various concentrations of circumferentially and radially oriented hydrides has been developed and implemented in a finite-element code. The hydride precipitation model, which describes the hydride structure of the cladding at the end of dry storage, and the hydride-dependent properties of high-burnup fuel cladding form the main input to the constitutive model. The third element in the overall process is to utilise this material model and its host finite-element code in the structural analysis of a transport cask subjected to bounding accident loading to calculate fuel rod failures and failure mode configurations. This requires detailed modelling of the transport cask and its internal structure, which includes the canister, basket, fuel assembly grids and fuel rods. The overall methodology is described.     

Radiological aspects and behavior of spent fuel considering long-term storage

For spent nuclear fuel management in Germany, the concept of dry interim storage in dual purpose casks before direct disposal is applied. Current operation licenses for storage facilities have been granted for a storage time of 40 years. Due to the current delay in site selection, an extension of the storage time seems inevitable. In consideration of this issue, GRS performed burnup calculations, thermal and mechanical analyses as well as particle transport and shielding calculations for UO₂ and MOX fuels stored in a cask to investigate long-term behavior of the spent fuel related parameters and the radiological consequences. It is shown that at the beginning of the dry storage period, cladding hoop stress levels sufficient to cause hydride reorientation could be present in fuel rods with a burnup higher than 55 GWd/tHM. The long-term behavior of the cladding temperatures indicates the possibility of reaching the ductile-to-brittle transition temperature during extended storage scenarios. Surface dose rates are 3 times higher when a cask is partially loaded with 4 MOX fuel assemblies. Due to radioactive decay, long-term storage will have a positive impact on the radiological environment around the cask.     

CANDU及RBMK压力管锆合金的氢致延迟断裂研究

采用紧凑拉伸试样(CT），在恒定载荷、不同氢含量、不同温度条件下，测量了CANDU堆和RBMK堆Zr-2．5Nb压力管材料氢致延迟开裂速率。用金相显微镜和扫描电镜观察断口及氢化物形貌，并测量临界应力场强度因子及开裂速率，对材料的微结构及氢化物分布进行分析。结果表明，氢致延迟断裂(DHC)生长呈阶梯状。与CANDU压力管比较，RBMK压力管的DHC开裂速率将近低一个数量级。其原因是RBMK压力管的屈服强度比CANDU压力管低得多。     

The performance of defected spent LWR fuel rods in inert and dry air storage atmospheres

A testing program using eight commercial PWR and BWR spent fuel rods was conducted to investigate their long-term stability under a variety of possible dry storage conditions. The objective of this project is to provide the Nuclear Regulatory Commission (NRC) with the information to confirm or establish spent-fuel, dry storage licensing positions regarding long-term, low-temperature ( <523 K) spent fuel rod behavior during dry storage, and for radioactive contamination arising from spallation of cladding crud. Until now, the testing program has included three interim nondestructive examinations and one destructive examination. This paper presents the results of the third examination conducted to determine any degradation in eight fuel rods after being subjected to 13168 h at temperature. During this examination, visual observations, diametrical measurements, and isotopic analysis of smears were used to assess the fuel rod behavior and particulate release.     

Assessment of aging of Zr---2.5Nb pressure tubes in CANDU reactors

In modern CANDU^™ nuclear generating stations, pressure tubes of cold-worked Zr---2.5Nb material are used in the reactor core to contain the fuel bundles and the heavy water (D₂O) coolant. The pressure tubes operate at an internal pressure of about 10 MPa and temperatures ranging from about 250°C at the inlet to about 310°C at the outlet. Over the expected 30 year lifetime of these tubes they will be subjected to a total fluence of approximately 3 × 10²⁶ n m⁻². In addition, these tubes gradually pick up deuterium as a result of a slow corrosion process. When the hydrogen plus deuterium concentration in the tubes exceeds the hydrogen-deuterium solvus, the tubes are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC). If undetected, such a cracking mechanism could lead to unstable rupture of the pressure tube. A fitness-for-service methodology has been developed which assures that this will not happen. A key element in this methodology is the acquisition of data and understanding—from surveillance and accelerated aging testing—to assess and predict changes in the DHC initiation threshold, the DHC velocity and the fracture toughness (critical crack length) as a function of service time. The most recent results of the DHC and fracture toughness properties of CANDU pressure tubes as a function of time in service are presented and used to suggest procedures for mitigation and life extension of the pressure tubes.     

Stress and temperature-dependent hydride reorientation of Zircaloy-4 cladding and its effect on the ductility degradation

A hydride reorientation can deteriorate the mechanical ductility of spent fuel cladding and make it more susceptible to failure. Therefore, an evaluation of the reorientation under dry storage conditions and their effects on the cladding ductility are critical issues in terms of the regulation criteria. In this work, biaxial stress was applied to Zircaloy-4 cladding by pressurizing Ar gas. The study showed that the hydride reorientation can occur at around 60 and 80 MPa at 400 and 300 °C, respectively. The ring compression test at room temperature showed that the ductility decreases with an increase in radial hydride quantity: Fl(45) and radial hydride continuity factor. In addition, a significant hydride reorientation can occur at high temperature conditions even if the hoop stress is equal to or less than 90 MPa which can bring a significant ductility degradation.     

The first step for delayed hydride cracking in zirconium alloys

Two models for delayed hydride cracking (DHC) in zirconium alloys are distinguished by their first step:

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The loading of a crack induces hydride precipitation. The hydride is postulated to create a hydrogen concentration gradient, where the bulk concentration is greater than that at the crack tip. This concentration gradient is taken as the driving force for diffusion of hydrogen to the crack tip, and subsequent hydride growth. This model is called the precipitate first model (PFM).

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The tensile stress at the crack tip induces a gradient in chemical potential that promotes the diffusion of hydrogen to the crack tip. Hydrides form if the hydrogen concentration reaches the solubility limit for hydride precipitation. The mechanism is postulated to create a hydrogen concentration gradient, where the bulk concentration is lower than that at the crack tip. The gradient in chemical potential is taken as the driving force for diffusion of hydrogen to the crack tip, and subsequent hydride growth. This model is called the diffusion first model (DFM).

The second model, DFM, is developed. This model is shown to describe the main features of the experimental observations of DHC, without invoking new phenomena, such as reduction in the solubility limit for precipitation of hydride, as required by the PFM.     

基于多物理场耦合的U3Si2燃料与双层SiC包壳组合的轻水堆燃料性能分析

基于COMSOL平台开发了一套基于多物理场全耦合的燃料性能分析程序,并通过径向功率分布模型对比验证了该程序的正确性与准确性;然后进一步分析了U₃Si₂燃料与双层SiC包壳组合、U₃Si₂燃料与锆合金包壳组合在反应堆正常运行工况下的性能,并与UO₂燃料与锆合金的组合进行了对比分析。计算结果发现U₃Si₂燃料与锆合金包壳组合相比UO₂燃料与锆合金的组合具有更低的燃料中心温度、裂变气体释放量及内压,但气隙闭合时间会提前;而U₃Si₂燃料与双层SiC包壳的组合相比U₃Si₂燃料与锆合金的组合具有更高的燃料中心温度、更大的裂变气体释放量及内压,且随着燃耗的增加,其燃料中心温度大幅增加,与锆合金包壳相比,双层SiC包壳能够有效延迟气隙闭合,缓解燃料与包壳的力学相互作用。      

Delayed hydride cracking in Zr-2.5Nb tube with the cooling rate and the notch tip shape

The objective of this study is to demonstrate the feasibility of the Kim’s delayed hydride cracking (DHC) model. To this end, this study has investigated the velocity and incubation time of delayed hydride cracking (DHC) for the water-quenched and furnace-cooled Zr-2.5Nb tubes with a different radius of notch tip. DHC tests were carried out at constant K_I of 20 MPa √m on cantilever beam (CB) specimens subjected to furnace cooling or water quenching after electrolytic charging with hydrogen. An acoustic emission sensor was used to detect the incubation time taken before the start of DHC. The shape of the notch tip changed from fatigue cracks to smooth cracks with its tip radius ranging from 0.1 to 0.15 mm. The DHC incubation time increased remarkably with the increased radius of the notch tip, which appeared more strikingly on the furnace-cooled CB specimens than on the water-quenched ones. However, both furnace-cooled and water-quenched CB specimens indicated little change in DHC velocity with the radius of the notch tip unless their notch tip exceeded 0.125 mm. These results demonstrate that the nucleation rate of hydrides at the notch tip determines the incubation time and the DHC velocity becomes constant after the concentration of hydrogen at the notch tip reaches terminal solid solubility for dissolution (TSSD), which agrees well with the Kim’s DHC model. A difference in the incubation time and the DHC velocity between the furnace-cooled and water-quenched specimens is attributed to the nucleation rate of reoriented hydrides at the notch tip and the resulting concentration gradient of hydrogen between the notch tip and the bulk region.     

Results of a long-term wet storage demonstration test with intact and operational defective LWR fuel rods

28 spent fuel rods — 18 intact and 10 operational defective rods — were included in the storage test program. Within 7 years the spent fuel rods were inspected four times. To characterize the spent fuel rods the following methods were applied during pool inspections: visual inspection, profilometry, eddy current testing, and oxide thickness recording.Summarizing the results of the intermediate and of the final inspection it has to be concluded that — as predicted — no change exceeding the detection limit could be found either at the intact or at the operational defective fuel rods. These results must be regarded as conservative because handling of the different spent fuel rods during inspection provided additional and atypical loads — especially for the operational defective spent fuel — in comparison with the long term storage of complete fuel bundles.The results of this carefully documented demonstration test has shown agreement with the theoretical analysis and with the overall experience available from pool storage that wet spent LWR-fuel storage can be performed without any problems even for extended periods of time.     

Comment on ‘The first step for delayed hydride cracking in zirconium alloys’ by G.A. McRae et al., J. Nucl. Mater. 396 (2010) 130-143

The aim of this paper is a reply to McRae et al.’s paper entitled “The first step for delayed hydride cracking (DHC) in zirconium alloys” claiming that the first step of DHC is hydrogen diffusion, not nucleation of hydrides as demonstrated by Kim’s new model. Despite the authors’ claim that the crack tip concentration is higher than the bulk concentration due to the stress gradient, their claim violates the thermodynamic principle that the stressed region should have a lower potential of hydrogen or lower hydrogen solubility than the unstressed region. Furthermore, it is demonstrated that the Diffusion First Model (DFM) proposed by the author is defective in terms of kinetics because hydrogen diffusion before hydride nucleation just governs the rate of hydride nucleation, neither the rate of hydride growth nor the crack growth rate (CGR).     

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

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

Research Program to Elucidate Outside-in Failure of High Burnup Fuel Cladding

Among a series of power ramp tests on 25 Zr-lined segment rods of burnup ranging from 43 to 61 GWd/t, five segment rods failed during the power ramp tests. One segment rod irradiated for 3 cycles (43 GWd/t) failed with a pinhole due to PCI/SCC. The rest of higher burnups failed with an axial crack on the outer surface. The failure threshold power tended to decrease as burnup increases.

Post irradiation examinations revealed increased cladding hydrogen absorption and its precipitates in the cladding outer rim after 4 and 5 cycle irradiations, in contrast to a uniform hydride distribution and a small hydrogen content after 3 cycle irradiation. Metallographic observations suggested an axial crack failure mode induced by the combined effects of high stress and hydrides precipitated in a radial direction during power ramp.

The axial crack failure during the power ramp is supposed to be initiated by a cracking of radial hydride formed by hydride re-distribution and re-orientation at the cladding outer rim and to propagate through a process of hydride concentration and precipitation at the crack tip. Research programs of experimental and analytical studies to clarify the conditions of such mechanism are on-going focusing on the hydrogen behavior and mechanical performance of the irradiated cladding.     

高燃耗对乏燃料贮存和运输的影响研究

本文对高燃耗对乏燃料包壳结构完整性的影响进行了分析。探讨了影响包壳结构完整性的重要温度限值,即燃料包壳温度限值、包壳溶解温度以及韧脆转变温度(DBTT)。给出了分析包壳结构完整性的方法,对拟在干式贮存设施内贮存超过20年的容器性能及贮存后运输时乏燃料组件的结构完整性进行了分析,并给出了相关建议。     

Elastic anisotropy of zirconium alloy fuel cladding

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

A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

In modern CANDU nuclear generating stations, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the fuel bundles and the heavy water (D₂O) coolant. The pressure tubes operate at an internal pressure of 10 MPa and temperatures ranging from 250°C at the inlet to 310°C at the outlet. Over the expected 30 year lifetime of these tubes, they would be subjected to a total fluence of 3×10²⁶ n m⁻². In addition, these tubes gradually pick up deuterium as a result of a slow corrosion process. When the hydrogen plus deuterium concentration in the tubes exceeds the hydrogen/deuterium solvus, the tubes are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC). If undetected, such a cracking mechanism could lead to unstable rupture of the pressure tube. The service life of the pressure tubes is determined, in part, by changes in the probability for the rupture of a tube. This probability is made up of the probability for crack initiation by DHC multiplied by the sum of the probabilities of break-before-leak and leak-before-break (LBB). A probabilistic model, BLOOM, is described which makes it possible to estimate the cumulative probabilities of break-before-leak and LBB. The probability of break-before-leak depends on the crack length at first leak detection and the critical crack length. The probability of a LBB depends on the shut-down scenario used. The probabilistic approach is described in relation to an example of a possible shut-down scenario. Key physical input parameters into this analysis are pressure tube mechanical properties, such as the crack length at first coolant leakage, the DHC velocity and the critical crack length. Since none of these parameters are known precisely, either because they depend on material properties, which vary within and between pressure tubes, and/or because of measurement errors, they are given in terms of their means and standard deviations at the different temperatures and pressures defined by the shut-down scenario.     

Beyond design basis accidents in spent-fuel pools — Generic Issue 82

Nuclear reactor plants include storage facilities for the wet storage of spent-fuel assemblies. The safety function of the spent-fuel pool (SFP) and storage racks is to cool the spent-fuel assemblies and maintain them in a subcritical array during all credible storage conditions and to provide safe means of loading the assemblies into shipping casks.Generic Issue 82 (GI-82) relates to the concern that for a postulated accident sequence that results in the loss of water from a light-water reactor (LWR) spent-fuel storage pool, a Zircaloy cladding fire could occur and propagate to older stored fuel. This issue was identified during hearings concerning SFP reracking amendments in the late 1970s when licensees were starting to use high-density storage racks. High-density racks are used to accommodate the storage of spent fuel in SFPs at reactor sites until such time as the Department of Energy (DOE) repository is available and spent fuel can be removed from the reactor sites. Maintaining a low-density storage configuration for recently discharged spent fuel would reduce the Zircaloy cladding fire probability by an order of magnitude, but at a greater cost for additional onsite storage space.The accident sequences that could result in water loss from the SFP, including beyond design basis earthquakes, various types of seal failures and dropped shipping casks, and the Zircaloy cladding fire issues have been studied by the NRC staff. The results of these studies are provided in NUREG-1353, “Regulatory Analysis for the Resolution of Generic Issue 82, Beyond Design Basis Accidents in Spent-Fuel Pools”. Although these studies conclude that most of the spent-fuel pool risk is derived from beyond design basis earthquakes, this risk is not greater than the risk from core damage accidents due to these beyond design basis earthquakes. Therefore, reducing the risk from spent-fuel pools due to events beyond the safe shutdown earthquake would still leave a comparable risk due to core damage accidents. The risk due to beyond design basis accidents in spent-fuel pools, while not negligible, is sufficiently low that the added cost involved with further risk reduction is not warranted.