The growth of oxide and oxygen-stabilized alpha layers in steam-oxidized zircaloy

Experimental investigations of the oxidation of Zircaloy in steam at high temperatures suggest temperature gradients exist across the oxide and oxygen-stabilized α layers even when specimens are exposed under nominally isothermal conditions. This paper presents a simple model that permits one to calculate the ratio of the thickness, of the oxide to oxygen-stabilized α layers in the presence of temperature gradients as well as under truly isothermal exposure conditions. The shape of the oxide to oxygen-stabilized α thickness ratio curve as a function of temperature was found to be in excellent agreement with oxidation kinetics data that were used to derive a scaling factor for the model. Variations in the temperature dependence of this ratio from independent measurements can be reproduced if it is assumed that temperature differences on the order of 10°C exist between the oxide layer and the oxygen-stabilized α layer. Metallographic evidence is presented that suggests the rate-controlling oxidation step occurs in the vicinity of the interface between the oxide and oxygen-stabilized a layers     

Evaluation of Thermocouple Fin Effect in Cladding Surface Temperature Measurement during Film Boiling

Thermocouple fin effect on surface temperature measurement of a fuel rod has been studied at elevated wall temperatures under film boiling condition in a reactivity initiated accident (RIA) situation. This paper presents an analytical equation to evaluate temperature drops caused by the thermocouple wires attached to cladding surface. The equation yielded the local temperature drop at measuring point depending on thermocouple diameter, cladding temperature, coolant flow condition and vapor film thickness. The temperature drops by the evaluating equation were shown in cases of free and forced convection conditions. The analytical results were compared with the measured data for various thermocouple sizes, and also with the estimated maximum cladding temperature based on the oxidation layer thickness in the cladding outer surface.

It was concluded that the temperature drops at above 1,000°C in cladding temperature were around 120 and 150°C for 0.2 and 0.3 mm diameter Pt-Pt-Rh thermocouples, respectively, under a stagnant coolant condition. The fin effect increases with the decrease of vapor film thickness such as under forced flow cooling or at near the quenching point.     

N36锆合金氧化层微观形貌及润湿特性实验研究

研究了国产N36锆合金包壳在600、700℃和800℃常压下形成的氧化层微观形貌和表面润湿特性。对N36锆合金样件进行氧化,并测量了氧化层厚度和表面接触角。对样件表面进行扫描电子显微镜（SEM）观测获得样件的表面微观形貌,利用能谱仪（EDS）对样件表面进行局部扫描获得了成分元素种类和含量分布,分析了氧化温度和氧化时间对于N36锆合金表面润湿性的影响规律。结果表明,氧化后的样件表面润湿性增强,氧化层表面裂纹的尺寸、深度、内部结构都会影响表面润湿性。随着氧化温度升高,裂纹尺寸有增加的趋势。在同一氧化温度下,随着氧化时间的增长,样件表面裂纹的尺寸和数量都有增加的趋势。本文研究有助于深入了解N36锆合金包壳材料表面氧化的微观特性。     

The effect of oxidation-and-quenching process during a LOCA on the behavior of the oxidation and embrittlement of Zircaloy-4 cladding under reheating transients

Two-sided oxidation tests, ring compression tests and semi-integral quench tests on Zircaloy-4 cladding specimens were conducted under temperature transient conditions simulating a post-quench reheat transient in order to evaluate the effect of high-temperature oxidation and quenching during a loss-of-coolant accident (LOCA) on the behavior of the oxidation and embrittlement of the cladding under a loss of long-term core-cooling condition. Test specimens prepared from non-irradiated Zircaloy-4 cladding tube were oxidized at a temperature between 1173 and 1473 K in steam flow and quenched by soaking the specimen in room temperature water. Re-heating tests were performed on the specimens in steam flow at a temperature between 1173 and 1473 K. The suppression of oxide layer growth and weight gain was observed under certain reheating-after-quenching conditions. Nevertheless, it seemed that the temperature transients including quenching-and-reheating process did not significantly affect the embrittlement of cladding. It was found that the embrittlement behavior of cladding during the temperature transients including quenching-and-reheating process could be dealt with on the basis of the Equivalent Cladding Reacted (ECR) based on the Baker–Just correlation.     

Analysis of stresses in GCFR cladding under normal operating conditions

A modified version of the LIFE-III code, LIFE-GCFR, and classical stress analysis techniques have been employed to calculate the stresses in GCFR cladding under normal reactor operating conditions. Several types of loadings on the cladding which occur during normal operation have been considered. These include fuel-cladding mechanical interaction, thermal stresses induced by radial and axial temperature gradients, and stresses induced by swelling gradients. The combined and individual effects of these loadings as well as the effect of creep on cladding stresses have been assessed. Results obtained from this study have provided input to the experimental GCFR cladding development work at Argonne National Laboratory.     

High Temperature Deformation Behavior of Gradually Pressurized Zircaloy-4 Tubes

In order to obtain fundamental perspectives on fuel cladding deformation behavior under changing temperature and pressure conditions in a hypothetical loss-of-coolant accident of PWR, Zircaloy-4 tube burst test was conducted in both air and 99.97% Ar atmospheres. The tubes were directly heated by ac-current and maintained at various temperatures, and pressurized gradually until rupture occurred. Rupture circumferential strains were generally larger in Ar gas than in air and attained a maximum around 1,120 K in both atmospheres. Some tubes tested in air produced axially-extended long balloons, which proved not to be explained by such properties or ideas as effect of cooling on strain rate, superplasticity, geometrical plastic instability and stresses generated by surface oxide layer. A cause of the long balloon may be obtained in the anisotropy of the material structure. But even a qualitative analysis based on this property can not be made due to insufficient data of the anisotropy.     

Cr涂层锆合金包壳模拟LOCA试验研究

2011年日本福岛核事故暴露传统锆合金燃料包壳在失水事故(LOCA)工况下的安全性问题。为了探究新型Cr涂层锆合金包壳在LOCA工况下的性能,本研究针对物理气相沉积(PVD)工艺涂覆的12～15μm厚度Cr涂层Zr-1Nb合金包壳管,开展模拟LOCA工况下的高温蒸汽氧化-淬火试验,氧化温度为1200℃和1300℃,单面氧化时间为10 min和20 min,淬火温度约800℃,之后对淬火后试样进行环压测试。结果发现,在研究条件下,Cr涂层未出现剥落,涂层完整;Cr涂层锆合金包壳外表面形成较为致密Cr₂O₃层,抑制O原子扩散至锆合金基体,阻止锆合金基体被氧化为ZrO₂层和α-Zr(O)层,环压测试发现淬火后包壳保持良好塑性。研究表明,在本测试工况下Cr涂层锆合金包壳相比传统锆合金包壳具有更强的抗LOCA事故能力。     

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

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

钠滴氧化燃烧特性的实验研究

液态钠泄漏和燃烧是钠冷快堆在运行中一多发的常见事故。本文主要针对液态钠滴在不同初始温度(140~370℃)和氧浓度(4%~21%)条件下的氧化燃烧行为进行实验研究。实验通过1套钠滴燃烧装置和高速摄像机使钠滴的氧化燃烧行为可视化。实验结果表明:钠滴的初始温度和氧浓度越低,钠滴表面产生的柱状氧化物越长;在相同氧浓度条件下,钠滴初始温度越高,越易着火燃烧;钠滴初始温度在200℃以下时很难点燃,当有扰动破坏了表面的氧化层结构时,钠滴也会逐渐燃烧;钠滴初始温度在140~370℃的条件下,氧浓度≥12%时,钠滴能燃烧充分,最高温度基本可升至600~800℃;氧浓度12%时,燃烧并不充分,燃烧的最高温度均在600℃以下。这些结果对柱状流及雾状钠火的研究有重要的指导意义。     

Evaluation of fracture resistance of ruptured,oxidized, and quenched Zircaloy cladding by four-point-bend tests

To evaluate fracture resistance of light water reactor fuel rods under loss-of-coolant accident (LOCA) and post-LOCA cooling conditions, four-point-bend tests were performed on non-irradiated Zircaloy cladding samples that were ruptured, oxidized in high-temperature steam, and quenched in flooding water. The bend test methodology was designed to apply a uniform bending moment to the entire rupture region and to generate tensile stress on the ruptured side. The fracture bending moment of the cladding decreased with oxidation temperature and hydrogen concentration as well as oxidation amount. Comparison with bending moments estimated from design basis seismic ground motion indicated that the cladding is unlikely to be fractured by seismic loads during post-LOCA cooling if high-temperature oxidation is kept below 15% equivalent cladding reacted, the oxidation limit of the Japanese LOCA criteria.     

The contribution of non-wetting droplets to direct cooling of the fuel during PWR post-LOCA reflood

During the reflood of a Pressurised Water Reactor (PWR) following a loss of coolant accident, precursory cooling prior to the arrival of the rewetting front is of vital importance in limiting the rise in cladding temperature before rewet. This precursory cooling is achieved by a flow of superheated vapour, with entrained saturated drops, which evaporate into the vapour and act as a heat sink. In this paper we investigate a complementary mechanism; the direct cooling of the cladding by the drops themselves. Cladding temperatures are such that wetting by these droplets does not occur. On the contrary, droplets bounce off a vapour cushion formed during the ∼10 ms or so that they are in close proximity to the cladding. Using a combination of previous experimental correlations and recent CFD calculations, we estimate the rate of heat removal from the cladding surface as a result of the droplet impingement. Thus, we estimate the heat removed as a result of one impingement and estimate the total rate of heat removal by estimating the number droplets impinging on the cladding per unit surface area. The heat extracted by those droplets is found to be about 1/10 of the heat extracted by single-phase vapour under typical reflood conditions. Though there significant uncertainties in these estimates, it does seem that direct cooling by droplets, not generally incorporated in analyses of reflood, could actually be making a significant contribution to keeping cladding temperatures down to acceptable levels.     

The effects of deformation on the high-temperature steam oxidation of zircaloy-2

An experimental study was made of the effects of deformation on the oxidation of Zircaloy-2 reactor fuel cladding in flowing steam in the temperature range 700 to 1300°C. The kinetics and mechanism of oxygen penetration and embrittlement were examined using gravimetric and metallographic methods. Tensile deformation during oxidation resulted in cracking of the growing oxide and local oxidation of the metal at the base of the crack. The number of points of local attack was inversely related to the temperature of oxidation. The depth of local penetration of the metal (in excess of the uniform depth of penetration for undeformed cladding) was greatest when the deformation rate was slow. The maximum depth of local attack found in these experiments was equivalent to an increase in total oxygen penetration of about twice that due to uniform diffusion. The possible relevance of these results to the analysis of reactor loss-of-coolant accidents is discussed.     

Development of Computer Code PRECIP-II for Calculation of Zr-Steam Reaction

A code PRECIP-II has been developed for the prediction of Zircaloy cladding oxidation under loss-of-coolant accident conditions of a light water reactor. This code is an extended, improved version of SIMTRAN-I produced by S. Malang. Main improvements have been made on the treatment of boundary conditions in a cooling period, which enabled the calculation to take into account the structure change of β-phase, i.e. α-phase precipitation. Modifications were made on the values of oxygen solubilities and diffusivities so that the values of isothermal reaction rate constants derived from calculations may match to the experimental data. Then, calculations and experiments of oxidation were carried out along various types of temperature histories, making comparison between the results of calculation and measurement on weight gain, thicknesses of oxide and α-phase layers. The results were that the differences between the calculated and experimental data lie for the most part within ±10%.     

Mechanical characteristics of fuel rod claddings in transport conditions

Abstract

A synthesis on the mechanical characteristics of unirradiated and irradiated fuel rod claddings was performed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to have reference data for the assessment of the safety demonstrations in normal and accident conditions of transport required by the procedure of package licensing. Indeed, the transport conditions correspond to a range of cladding temperatures (200–550°C) which is only partly covered by the data acquired within the framework of the safety demonstration relative to the reactor normal operating conditions, especially beyond 400°C. This work concerned Zircaloy-4 cladding material (Zry-4) and M5^TM. Data about mechanical properties (elastic and ductile properties, creep behaviour), oxidation (in reactor and under air during transport), hydrides and fracture toughness have been collected and synthesised. The laws presented in the document can be used to obtain orders of magnitude of oxide layer thickness, hydrogen content and creep deformation rate. The following phenomena which could influence the mechanical behaviour of the cladding were more particularly studied: oxidation which could become very important during transport in case of cladding temperatures of ～500°C; creep for which only a few data ～500°C are available and which depends in particular on the internal pressure of the rods, the cladding oxidation and the presence of the hydrides; and recrystallisation of Zry-4 at ～500°C, which could have consequences on the mechanical properties of the cladding after cooling during the storage. For other topics of interest for the study of the mechanical behaviour of the cladding, such as the fracture toughness for example, it was identified that the data available is scarce.     

CUPIDON-code decrivant le comportement thermique et mecanique d'un crayon combustible de reacteur a eau sous pression au cours d'un accident de depressurisation

In the scope of the PHEBUS experimental program to be performed in Cadarache on the behaviour of PWR's fuel assemblies under loss of coolant accidental conditions, a computer code has been developed to help designing the experimental rods and to contribute to the definition of the test runs.This code, dubbed CUPIDON, deals only with the thermal and mechanical behaviour of the rods as well as the oxidation of the cladding outside surface; it does not include any thermohydraulic subroutine. Rather, it is coupled with the RELAP code for providing necessary input data such as coolant temperatures and pressures and cladding-to-coolant heat transfer coefficients. It is restricted to a single, non irradiated, rod of short length as representing the PHEBUS experimental conditions.It is a two dimensional code, using a finite difference resolving technique. It calculates the radial thermal profile across each section of the rod, the stress and creep rate to which the cladding is submitted and the rate of formation of the oxide layer on the surface of the cladding under steady state and transient conditions. As cladding plastic strain input data, it is using the EDGAR-ZY experimental results.     

In situ XRD analysis of the oxide layers formed by oxidation at 743 K on Zircaloy 4 and Zr–1NbO

Two alloys, having different oxidation behaviour (Zy4 and Zr–1NbO), have been investigated during oxidation at high temperature (743 K) and low oxygen pressure (10 kPa) by in situ X-ray diffraction (XRD). Tetragonal phase content and ‘pseudo-stresses’ on the monoclinic phase have been measured as a function of the oxide layer thickness. The tetragonal phase contents are similar for both alloys and decreased with the oxide layer thickness. Pseudo-stresses were much more compressive on Zr–1NbO alloy, with limited changes at the corrosion kinetics transition. On cooling, the tetragonal fractions do not change, while ‘pseudo-stresses’ decreased in different ways for the two alloys. With respect to stress analysis, no correlation was found between ‘pseudo-stresses’ and tetragonal phase content. In addition, due to the thermoelastic properties of the highly anisotropic phases of the zirconia, large internal thermal stresses are expected to develop during any temperature changes. The orders of magnitude of them are similar to the stresses induced by swelling during oxidation from Zr to ZrO₂.     

A model for uniform Zircaloy clad corrosion in pressurized water reactors

A phenomenological water-side corrosion model for Zircaloy fuel cladding for pressurized water reactors (PWRs) is considered. The model acounts for the breakaway transition in the Zircaloy oxidation rate that takes place in an isothermal condition and the changes that occur during reactor operation, i.e. the dependence of oxide growth on fast neutron flux and cladding oxide layer thickness. Closed-form analytical solutions of the oxidation kinetics equations are obtained. The corrosion kinetics model is coupled to PWR thermal and hydraulic models which assume a subchannel that is either a closed single channel or a multichannel which accounts for coolant cross-flow and coolant enthalpy mixing. Both single-phase forced convection and subcooled nucleate boiling are accounted for in the thermal-hydraulic models. The model calculates the coolant temperature at the axial midplane of each axial segment of the fuel rod. When an oxide layer is present, the temperature at the metal-oxide interface is determined. This temperature in turn is used to determine the oxide growth via the Arrhenius temperature dependence of the Zircaloy oxidation rate. The predictions of the model have been compared with the measured cladding oxide data obtained in PWRs. The data for a given rod were obtained at various burn-ups (at the end of reactor cycles) and various axial positions of the rod. Our evaluations show that the model predicts the measured data satisfactorily; however, the deviations are discussed. The model has been used to study the effect of core loading patterns on cladding oxide growth. Our analyses show that core nuclear design is an important factor for water-side corrosion of fuel rods.     

Laser cladding of Colmonoy 6 powder on AISI316L austenitic stainless steel

Stainless steels are widely used in nuclear power plant due to their good corrosion resistance, but their wear resistance is relatively low. Therefore, it is very important to improve this property by surface treatment. This paper investigates cladding Colmonoy 6 powder on AISI316L austenitic stainless steel by CO₂ laser. It is found that preheating is necessary for preventing cracking in the laser cladding procedure and 450 °C is the proper preheating temperature. The effects of laser power, traveling speed, defocusing distance, powder feed rate on the bead height, bead width, penetration depth and dilution are investigated. The friction and wear test results show that the friction coefficient of specimens with laser cladding is lower than that of specimens without laser cladding, and the wear resistance of specimens has been increased 53 times after laser cladding, which reveals that laser cladding layer plays roles on wear resistance. The microstructures of laser cladding layer are composed of Ni-rich austenitic, boride and carbide.     

Study of spray cooling of a pressure vessel head of a boiling water reactor

The present paper deals with a theoretical analysis of the spray cooling of a Reactor Pressure Vessel (RPV) head in a Boiling Water Reactor (BWR). To this end a detailed computational model has been developed. The model predicts the trajectories, diameters and temperatures of subcooled droplets moving in saturated vapor. The model has been validated through comparison with experimental data, in which droplet temperatures were measured as functions of the distance that they cover in saturated vapor from the moment they leave the sprinkler outlet to the moment they impact on the RPV head inner wall. The calculations are in very good agreement with measurements, confirming the model adequacy for the present study. The model has been used for a parametric study to investigate the influence of several parameters on the cooling efficiency of the spray system. Based on the study it has been shown that one of the main parameters that govern the temperature increase in a subcooled droplet is its initial diameter. Comparisons are also made between conclusions from the theoretical model and observations made through flow and temperature measurements in the plant (Forsmark 1 and 2). One of these observations is that the rate at which the RPV head temperature decreases on the way down from hot to cold standby is constant and independent of the sprinkling flow rate as long as the flow rate is above a certain minimum value. Accordingly, the theoretical model shows that if one assumes that the cooling of the RPV head is through a water film built on the inner wall due to sprinkling, the heat removal rate is only very weakly dependent on the sprinkling flow rate.     

Effects of Coolant Flow on Light Water Reactor Fuel Behaviors during Reactivity Initiated Accident

This paper describes the in-pile experimental results to study the influences of coolant flow on fuel behaviors under reactivity initiated accident (RIA) conditions performed in the Nuclear Safety Research Reactor (NSRR). A single PWR type test fuel rod was irradiated by a large neutron pulse in the NSRR to simulate a prompt power excursion of RIA's. The effects of coolant flow were studied at a coolant flow velocity of 0.3~1.8m/s and a coolant temperature of 20~90°C under the atmospheric pressure. It was found that the cooling conditions had considerable influences on fuel thermal behaviors under prompt heat-up. The increase of coolant flow velocity and subcooling enhanced heat transfer coefficient at cladding surface during film boiling, which resulted in large decrease of maximum cladding temperature and film boiling duration, and consequently in the increase of fuel failure threshold energy. The data tendencies are summarized and the influences of coolant flow are discussed with some computer analyses.