Secondary hydriding criteria under irradiation conditions

Reliable criteria of secondary hydriding failures are important to assure safe operation of nuclear fuel in LWR power units. The present paper reviews available data on massive hydriding of zirconium-based claddings covering out-of-pile studies and in-pile tests in research reactors. Analyses of these experimental data give evidence that threshold conditions leading to the onset of massive hydriding are drastically changed under irradiation. The changes are caused mainly by irradiation damage of oxygen sublattice in ZrO₂ by fission fragments leaving the periphery of fuel pellets. The tests in research reactors provide a basis to develop a parametric dependency which relates the threshold of massive hydriding to composition of steam-hydrogen mixture, irradiation dose rate and temperature.     

The plutonium/hydrogen reaction: The pressure dependence of reaction initiation time and nucleation rate controlled by a plutonium dioxide over-layer

As part of an ongoing programme to quantify those parameters which influence the early stages of the plutonium hydriding reaction, the hydrogen pressure dependence of both plutonium hydriding initiation time (I_t) and hydriding nucleation rate (N_r) have been determined for plutonium covered in a reproducible dioxide over-layer. The data show that initiation time is inversely proportional to hydrogen pressure, while nucleation rate is proportional to hydrogen pressure. Both observations are consistent with a model of hydriding attack in which the dioxide over-layer acts as a diffusion barrier, controlling the flow of hydrogen from the gas phase to the oxide/metal interface. The low scatter and reproducibility of the experimental data set demonstrate the importance of synthesising well controlled and characterised oxide layers prior to determining these experimental parameters.     

破损燃料棒二次氢化行为观察与分析

燃料棒在堆内运行时,由于初次破口会导致包壳发生二次氢化现象,二次氢化是导致燃料棒发生严重破损的重要因素。针对实际工况下的破损燃料棒,在中国原子能科学研究院燃料与材料检验设施（303热室）上开展了相关辐照后检验,并采用热室金相手段,对燃料棒二次氢化行为进行了观察分析。结果表明：二次氢化破口有明显的氢化肿胀现象;氢化物分阶段从内壁扩散到外壁,并形成“日爆”现象;二次氢化部位芯块温度明显升高,并会导致芯块气孔迁移、芯块晶粒长大、柱状晶生长等现象发生;相比未破损棒,破损棒二次氢化部位水侧氧化膜厚度有增加现象,但仍处于正常范围内。     

The plutonium/hydrogen reaction: The pressure dependence of reaction initiation time

The hydrogen pressure dependence of the initiation time (I_t) of the plutonium hydriding reaction has been determined over a hydrogen pressure range of 10-1000 mbar for plutonium covered in a dioxide over-layer. The data show that hydriding initiation time is inversely proportional to hydrogen pressure. This observation is consistent with a model of hydriding attack in which the dioxide over-layer acts as a diffusion barrier, controlling the flow of hydrogen to the oxide/metal interface. The low scatter and reproducibility of the experimental data set illustrate the importance of synthesising a reproducible oxide layer prior to determining this experimental parameter.     

金属Ce氢化-氮化反应动力学

以金属Ce为原料,对Ce氢化-氮化动力学参数进行研究,分别考察了初始反应温度、原料气体初始压力和Ce片厚度等对金属Ce氢化-氮化反应的影响。研究结果表明:升高初始反应温度有利于缩短氢化过程的诱导时间,但对氢化反应速率无明显影响;提高H₂初始压力和降低片层厚度能明显加快氢化反应速率。此外,通过X射线衍射(XRD)和扫描电镜(SEM)等表征证明氢化铈可以与N₂反应得到氮化铈(CeN)和H₂,但反应需要较高的温度以克服活化能,增加初始反应温度和N₂初始压力可以提高反应速率。综合考虑,初始反应温度为350℃、N₂初始压力为60 kPa是氢化铈氮化的较优条件。     

High-pressure hydriding of Zircaloy cladding by the thermogravimetry and tube-burst techniques

Hydriding kinetics of modified Zircaloy claddings was studied by the thermogravimetric method at 400 °C and the tube-burst technique at 315 °C. Some specimens were prefilmed with a thin oxide layer by air oxidation on both the inner and outer surfaces which were either pickled or blasted. In the thermogravimetric test, the hydriding rates of bare cladding specimens (no oxide prefilm) were in the range 0.9-1.6 mg/cm² h with little effect of the surface treatment. Incubation times were less than 1 h or even zero. In the tube-burst test, immediate and extensive hydrogen uptake was observed for these non-coated specimens. On the other hand, the cladding specimens with oxide prefilm exhibited lower hydriding rates ranging from 0.01 to 0.05 mg/cm² h and incubation times increased to 42 h. In addition, no hydrogen uptake was observed for all oxide-coated specimens for 100-750 h.     

Ductile crack growth simulation from near crack tip dissipated energy

A method to calculate ductile tearing in both small scale fracture mechanics specimens and cracked components is presented. This method is based on an estimation of the dissipated energy calculated near the crack tip. Firstly, the method is presented. It is shown that a characteristic parameter G_fr can be obtained, relevant to the dissipated energy in the fracture process. The application of the method to the calculation of side grooved crack tip (CT) specimens of different sizes is examined. The value of G_fr is identified by comparing the calculated and experimental load line displacement versus crack extension curve for the smallest CT specimen. With this identified value, it is possible to calculate the global behaviour of the largest specimen. The method is then applied to the calculation of a pipe containing a through-wall thickness crack subjected to a bending moment. This pipe is made of the same material as the CT specimens. It is shown that it is possible to simulate the global behaviour of the structure including the prediction of up to 90-mm crack extension. Local terms such as the equivalent stress or the crack tip opening angle are found to be constant during the crack extension process. This supports the view that G_fr controls the fields in the vicinity near the crack tip.     

Bench Simulation of the Stages of an Anticipated Loss-of-Coolant Accident in a VVÉR Reactor System

A computational-experimental procedure for performing a bench simulation of the first and second stages of an accident are described. Experiments studying the behavior of fuel elements and model fuel assemblies under conditions simulating the first stage of an accident showed, for the first time, that the claddings of the most highly heat-stressed fuel elements bulge out and rupture. Tests performed on model fuel assemblies with 19 and 37 fuel elements with Zr + 1%Nb and É635 alloy cladding under the conditions characteristic for the second stage of an accident showed that the fuel-element cladding becomes deformed in the course of the tests. The claddings of the fuel-element simulators rupture at the heating stage in the temperature range 820–920°C. The blockages of the flow section of the model assemblies with Zr + 1%Nb and É635 alloy claddings were 42 and 70%, respectively.     

A model to describe the interaction between UO2 and zircaloy in the temperature range 1000–1700°C

During high temperature transients the Zircaloy claddings in a nuclear power reactor may chemically interact with their environment, either on the internal or on the external surfaces. The external oxidation due to the interaction with the coolant has been extensively studied and described in models. However, the internal oxide-metal interaction is less well known and has not yet been satisfactorily described in models. In this paper the internal oxidation, which results from the reduction of the UO₂ pellets by the Zircaloy cladding material is analyzed. A model is developed which solves only the oxygen diffusion problem in the five phases which are formed due to the UO₂/Zircaloy interaction, without taking into account zirconium and uranium diffusion. The oxygen diffusion coefficients of four of the phases are determined and the oxygen concentrations at some of the interfaces are adjusted in order that the interface movements be in accordance with the experimental data published in the literature for the temperature range 1000 to 1700°C. The model attempts only to simulate the kinetics of the system during the first stages of the process when all the interface movements obey parabolic rate laws.     

Development of 9Cr-ODS Martensitic Steel Claddings for Fuel Pins by means of Ferrite to Austenite Phase Transformation

For use as fuel cladding of liquid metal fast reactors, Fe-0.12C-9Cr-2W ODS martensitic steel claddings were developed by cold-rolling under the softened ferrite phase induced by slow cooling from austenite phase, subsequently by ferrite to austenite phase transformation to break up substantially elongated grains produced by cold-rolling at the final heat-treatment. The produced claddings showed noticeable improvement in tensile and creep rupture strength that are considerably superior to PNC-FMS and even austenitic PNC316 at higher temperature and extended time to rupture. The strength improvement is mainly attributed to titanium addition in ODS martensitic steels through its reduction of Y₂O₃ particle size and shortening inter-particles spacing. The behavior of oxide particle size reduction is associated with stoichiometry between Y₂O₃ and TiO₂.     

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.     

Pu2O3 and the plutonium hydriding process

The role of cubic Pu₂O₃ in the corrosion of PuO₂-coated Pu by H₂ was investigated. Experiments were conducted to demonstrate that nucleation of hydriding is promoted by formation of Pu₂O₃ sites in the oxide layer. The nucleation mechanism based on diffusion of hydrogen through the PuO₂ layer was evaluated and an alternative mechanism based on formation of catalytic Pu₂O₃ sites via the Pu-PuO₂ reaction is proposed. The possibility of active participation of other impurities and inclusions in the dioxide is also discussed.     

U-2.5Nb氢蚀初期动力学试验研究

利用带显微镜的气固反应系统,在线观察了U-2.5Nb氢蚀初期氢化物成核和生长过程,表征了U-2.5Nb氢蚀初期的动力学特征和反应速度,研究了温度、压力和膜厚对U-2.5Nb氢蚀初期行为特征的影响。结果表明,U-2.5Nb比未合金化铀的氢化速度大很多,抗氢蚀性能更差。在低温（<125℃）下,孕育期随反应温度的升高而变短,遵从Arrhenius关系,超过125℃,孕育期随反应温度的升高而变长,孕育期存在一最小值;孕育期随反应压力升高而变短,与反应氢压力成反比关系;膜厚对孕育期的影响极其明显,孕育期随氧化膜厚度增加而变长。孕育期越长的氢蚀,其表面的成核点越少,成核越不均匀,且易在局部形成大的氢化物,氢化速度亦越低。     

2D crack growth simulation with an energetic approach

It is well known that the tearing resistance curve J–Δa is not a material property. A recent approach, based on an energetic critical parameter to model ductile tearing propagation, is used to model 3D effects. The approach considered in this work aims to estimate the dissipated energy in the fracture process during ductile tearing represented by an intrinsic parameter G_fr. A fracture criterion, which accounts for the crack extension length, is defined and lies on a critical energy release rate, noted G_c, which is compared to G_fr. Previously, this parameter was obtained from a numerical local energy release rate, which handicaps the application field of the approach: a fine mesh for the whole propagation area was needed and the criterion allowed only to model 1D propagation. A new manner to estimate G_c is then proposed in this article, based on the J plastic part variation, which allows to model 2D propagation by defining a local criterion. This new calculation method is validated on a CT specimen made in Tu52b ferritic steel, by comparing the results obtained from the two methods of G_c calculation. Then, the 2D crack growth case is studied, by modelling the propagation in a ring, loaded in compression. It is shown that a 3D effect, such as tunnel effect, could be successfully represented with this approach.     

Nuclear Characteristics of Molten-Salt-Cooled D-D Fusion Reactor Blankets

To consecutively decompose ¹⁴CO₂ into carbon (¹⁴C) through its reaction with Hz, an apparatus using microwave discharge and its conditioning were investigated. The reaction produces CO as an intermediate, and proceeds in the two steps of (1) “CO₂+H₂ → CO+H₂O” and (2) “CO+H₂+C_n → C_n+1+H₂O”, where C_n denotes the carbon already deposited on the wall of the discharge tube. Preliminary dispersion of carbon to the wall of the discharge tube by sputtering of a graphite particle was effective to promote the reaction. Two silica discharge tubes (6 mm O.D., 4 mm I.D., and 150 mm length each) were connected in series to proceed the former reaction in the first discharge tube and the latter one in the second one. When a 1:3 mixture of CO₂ and H₂ (total pressure 0.67kPa) was passed through the discharge tubes at a linear gas velocity of approximately 30mm/s and discharged for 60 h under microwave of 30–40 W supplied from two 2,450MHz power generators (200 W each), more than 90% of CO₂ was converted into CO in the 1st tube and about 23% of the CO was then decomposed into carbon in the 2nd tube. However, about 50% of the CO escaped from the tube without being decomposed, and about 0.5% and 1% of the carbon fed were hydrogenated into CH₄ and C₂H₂, respectively. The rest about 25% which was not confirmed was probably evacuated from the 2nd tube as microparticles of carbon. To completely decompose CO₂ into carbon, additional discharge tubes are necessary downstream of the 2nd tube.     

蒸汽发生器管子—管板接头设计的计算方法

本文主要介绍液压胀接管子-管板接头设计的一种理论计算方法。该方法是:将接头的整个胀接过程分为四个阶段,以此为数学模形,用弹性理论和塑性理论分阶段对管子或管子和管板的应力、应变和位移进行分析,从而估算胀管压力、变形、残余应力和拔脱力等。该方法可用于设计蒸汽发生器和钢制管壳式换热器。文中指出,对核蒸汽发生器的管子-管板接头来说,采取液压胀后再局部滚胀比只进行液压胀更好。     

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

ABSTRACT

To evaluate the fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident (LOCA) conditions, laboratory-scale integral thermal shock tests were performed using the following advanced fuel cladding tubes with burnups of 73–85 GWd/t: M-MDA^TM, low-tin ZIRLO^TM, M5®, and Zircaloy-2 (LK3). In total eight integral thermal shock tests were performed for these fuel cladding tube specimens, simulating LOCA conditions including ballooning and rupture, oxidation, hydriding, and quenching. During the tests, the specimens were oxidized to 10%–30% equivalent cladding reacted (ECR) at approximately 1473 K and were quenched under axial restraint load of approximately 520–530 N. The effects of burnup extension and use of the advanced fuel cladding tubes on the ballooning and rupture, oxidation, and hydriding under LOCA conditions were inconsiderable. Further, the high-burnup advanced fuel cladding tube specimens did not fracture in the ECR values equal to or lower than the fracture limits of the unirradiated Zircaloy-4 cladding tube reported in previous studies. Therefore, it can be concluded that the fracture limit of fuel cladding tubes is not significantly reduced by extending the burnup to approximately 85 GWd/t and using the advanced fuel cladding tubes, though it slightly decreases with increasing initial hydrogen concentration.     

A simple coincidence method of deuterium profiling using the DHe, α)H reaction

A simple standard coincidence arrangement is described for D(³He, α)H depth profiling experiments of deuterium in solids. Both reaction products are detected in coincidence, the high-energy protons being observed in transmission through a foil target. The energy of the α-particles is converted into the corresponding depth scale, whereas the local deuterium concentration is calculated from the yield of the α-particle spectrum. The measurement of α-particles in coincidence with protons allows a reduction of background arising from Rutherford scattering of ³He and other reaction products. For samples of deuterium implanted into Ta₂O₅ and Er, the method allows a reduction of the backscattering yield by a factor of more than 10³. The residual background is due to accidental coincidences. It can be even more reduced using an accurate experimental geometry. The background reduction is largest for samples with a low content of deuterium, allowing measurements of deuterium profiles of 9 × 10¹³ D⁺/cm² implanted into Ta₂O₅ at an energy of 8 keV. This corresponds to a maximum deuterium concentration of about 5 × 10⁻⁴ D per Ta₂O₅. This method is not restricted to thin films, but it allows measurements of deuterium profiles in a thick sample, e.g. of an implantation profile in a 0.4 mm silicon wafer.     

Application of Electron–Positron Annihilation for Determining the Thickness of the Barrier Layer on VVÉR Fuel-Element Cladding

The possibility of using the nuclear-physical method of positron annihilation to determine the thickness of a barrier layer consisting of zirconium on the inner surface of VVÉR fuel-element claddings is examined. The basic principles and possibilities of the method and the results of investigations are presented.     

Improvement in hydriding property of LaNi4.8Al0.2 alloy encapsulated by SiO2 sol

LaNi_4.8Al_0.2 alloy particles encapsulated by SiO₂ matrix were prepared by the sol gel method. Scanning electron microscope (SEM) imaging was applied to determine the silica network outside the encapsulated alloy. The hydriding kinetics, pulverization and poisoning characteristics of LaNi_4.8Al_0.2 alloy were investigated before and after being encapsulated by silica. The results reveal that the hydriding properties of encapsulated alloy are excellent. The hydrogenation rate of encapsulated alloy is faster than that of the original alloy. The quantities of hydrogen stored by the encapsulated and original alloy are 169.3 Nml/g and 147.1 Nml/g, respectively. The LaNi_4.8Al_0.2 alloy particles are broken up into powder after 10 times hydrogen absorption/desorption cycles, while the encapsulated alloy do not show any breakdown after 30 times hydrogen adsorption/desorption cycles. The quantities of hydrogen absorbed by original alloy particles are less than 8.2 Nml/g in H₂-14.4% CO and 18.6 Nml/g in H₂-12.8% CO₂, while the quantities of hydrogen absorbed by encapsulated alloy agglomerations are 84.5 Nml/g in H₂-14.4% CO and 168.9 Nml/g in H₂-12.8% CO₂. These results clearly indicate that the pulverization and poisoning resistance properties of LaNi_4.8Al_0.2 alloy are evidently enhanced after being encapsulated by silica network.