Effect of microstructure on irradiation creep and growth of zircaloy pressure tubes in power reactors

Cold-worked zircaloy pressure tubes fabricated by eight different routes and irradiated in four power reactors have diametral strain rates varying from ?1 to 11 × 10^?29 m²/n (>1 MeV) and elongation rates varying from 5 to 16 × 10^?29 m²/n (>1 MeV). X-ray diffraction and optical metallography show large differences in crystallographic texture, dislocation density and grain shape amongst the eight types of tubes. A quantitative correlation of the elongation and diametral rates of seven types of tubes with their texture, grain shape and dislocation density is developed assuming additive creep and growth components. The derived magnitudes of the creep and growth components in the pressure tube agree with published uniaxial creep and growth data for cold-worked zircaloy. The growth rate is approximately proportional to dislocation density and increases in a given direction as the grains become more elongated in that direction and as the proportion of basal plane normals in that direction decreases. The creep rate is insensitive to dislocation density. For internally pressurized tubes the diametral creep rate increases and the axial creep rate decreases as the proportion of basal plane normals in the radial direction increases. The model developed to describe the total long term dimensional changes of the pressure tubes successfully predicts the diametral and longitudinal rates for tubes which experienced end loads during operation.     

Mechanical Properties of Neutron Irradiated Fuel Cladding Tubes

There was few post irradiation examination data on the mechanical properties of domestic fuel cladding tubes used for light water reactors, then those data obtained abroad have been often used in the fuel design or fuel performance codes. Although, many reports discussed the deformation mechanism of the tube, almost all the data were not obtained from irradiated specimens but unirradiated ones. In recent years, systematic post irradiation examinations on domestic fuel elements used in Japanese light water reactors and the related studies were performed.

This report first summarizes briefly the crystallographic texture which characterizes the properties of Zircaloy fuel cladding tubes, followed by an explanation of basic properties such as elasticity, plasticity, creep and fatigue. Finally, the up-to-date results are introduced.     

锆合金辐照蠕变和生长的宏介观模型研究进展

锆合金作为核反应堆堆芯的主要结构材料之一,在服役过程中会发生辐照蠕变和生长行为,严重影响其使用可靠性。预测锆合金的辐照蠕变和生长是保障反应堆安全运行的关键。本文聚焦于两类锆合金构件,包括压水堆用锆合金包壳管及重水堆用Zr-2.5Nb压力管,分别从宏介观尺度详细综述了其辐照变形预测模型。针对适用于包壳管的宏观经验模型及介观力学模型,分别描述了两类模型的特征,重点介绍了介观力学模型的研究现状及最新进展,并对其未来的发展方向提出建议;针对适用于压力管的宏观经验模型,包括加拿大CANDU压力管辐照变形计算方程(简称C6方程)及秦山重水堆压力管辐照变形计算方程(简称秦山方程),分别描述了两类方程的具体形式,并介绍了C6方程国产化的进展。     

Mössbauer spectroscopy of tin in unirradiated and neutron irradiated Zircaloys

Mössbauer spectroscopy of the 23.9 keV γ-rays in ¹¹⁹Sn nuclei was applied to study Zircaloy-2, Zircaloy-4, and other tin-bearing zirconium-based alloys of interest to nuclear power technology. Zircaloys are extensively used in nuclear reactors as fuel cladding. In CANDU reactors, Zircaloys are also used as major structural components such as calandria tubes, and were used until the late 1970's as pressure tubes (now replaced by Zr–2.5Nb alloy). Unirradiated specimens of these alloys, as well as radioactive specimens, both neutron-irradiated in high-flux test reactors and extracted from nuclear power-reactor components after many years of service, were examined. The obtained spectra consistently showed tin in substitutional solid solution in α-Zr, whereas no evidence was found of metallic Sn or intermetallic Zr₄Sn precipitates. In oxide scrapes removed from Zircaloy-2 pressure tube of one of CANDU reactors, where the alloy was exposed for about 10 years to pressurized heavy water coolant at temperatures of ∼280°C, a considerable fraction of tin was found in the Sn(IV) state, in the form that coincides with the state of tin in stannic oxide, SnO₂. The same form of tin was identified in filterable deposits in the primary heavy water coolant of CANDU reactors. For comparison, in Zircaloy heated in air, SnO₂ was formed only at temperatures above 500°C.     

Investigation of in-reactor creep and irradiation growth of zirconium-2.5 wt% niobium channel tubes

We summarize the diametral creep results obtained in the MR reactor of the Kurchatov Institute of Atomic Energy on zirconium-2.5 wt% niobium pressure tubes of the type used in RBMK-1000 power reactors. The experiments that lasted up to 30 000 h cover a temperature range of 270 to 350°C, neutron fluxes between 0.6 and 4.0 ×10¹³ n/cm² · s (E > 1 MeV) and stresses of up to 16 kgf/mm². Diametral strains of up to 4.8% have been measured. In-reactor creep results have been analyzed in terms of thermal and irradiation creep components assuming them to be additive. The thermal creep rate is given by a relationship of the type ε_th = A₁ exp [(A₂ + A _3σt) T] and the irradiation component by ε_rad = A_4σtø(T − A₅), where T = temperature, σ_t = hoop stress, ø = neutron flux and a₁ to A₅ are constants. Irradiation growth experiments carried out at 280° C on specimens machined from pressure tubes showed a non-linear dependence of growth strain on neutron fluence up to neutron fluences of 5 × 10²⁰ n/cm². The significance of these results to the elongation of RBMK reactor pressure tubes is discussed.     

Effect of ion irradiation and indentation depth on the kinetics of deformation during micro-indentation of Zr-2.5%Nb pressure tube material at 25 °C

Micro-indentation creep tests were performed at 25 °C on radial-normal samples cut from Zr-2.5Nb CANDU pressure tube material in both the as-fabricated condition and after irradiation with 8.5 MeV Zr⁺ ions. The average indentation stress, and hence the yield stress, was found to increase with decreasing indentation depth and with increasing levels of ion irradiation. The activation energy of the indentation creep rate and hence the, activation energy of the obstacles that limit the rate of dislocation glide, was independent of indentation depth but increased from ΔG₀ = 0.185 to 0.215 μb³ with increasing ion irradiation damage. The magnitude of the activation energy indicates that ion irradiation introduces a new type of obstacle into the microstructure which reduces the low temperature indentation creep rate of Zr-2.5Nb pressure tubes. This is supported by TEM images showing that Zr⁺ ion irradiation produces small, nanometer size, dislocation loops which act as obstacles to dislocation glide and thus influence both the yield stress and the activation energy of the low-temperature thermal creep of Zr-2.5Nb pressure tube material. These findings suggest that neutron irradiation will have similar effect upon yield stress and low-temperature thermal creep as the Zr⁺ ion irradiation since both create similar crystallographic defects in Zr-2.5Nb pressure tubes.     

Prediction of fuel channel—graphite gas-gap behaviour in RBMK reactors

To maintain thermal contact between the fuel assembly and the graphite moderator, RBMK design reactors employ graphite split rings, which are alternatively tight on the pressure tube or tight on the graphite brick central bore. The split in the graphite rings allows a helium/nitrogen gas mixture to flow up the fuel channel. This prevents oxidation of the graphite and can be sampled to detect pressure tube leaks. The initial clearance between the rings and pressure tube or graphite brick is approximately 2.7 mm (1.35 mm each side). Due to material property changes of the pressure tubes and graphite during operation of the reactor, the size of the clearance between the rings and the pressure tube/brick, called the “gas-gap”, varies. Closure of these gaps has been identified as a possible safety case issue by reactor designers and by independent reviews carried out as part of TACIS reviews and as part of the Ignalina Safety Analysis Report. The reasons for this are that gas-gap closure would cause the pressure tube to be tightly gripped by the graphite bricks via the split rings, which could lead to:

• Extra loading on the upper pressure tube zirconium/steel transition joint, particularly during shut down and emergency transients.

• Splitting of the graphite brick, leading to loss of thermal contact between the pressure tube and graphite. As approximately 5.6% of the heat in graphite-moderated reactor is generated within the moderator through neutron and gamma-heating, loss of thermal contact would result in higher graphite temperatures, accelerating the rate of graphite expansion and hence increasing the loading of the core radial restraint.

• Graphite debris may become lodged in inter-brick gaps, leading to increased axial pressure tube loading during shut down and emergency transients.

The authors have carried out deterministic assessments based on the Ignalina RBMK-1500 reactors in Lithuania, modelling the behaviour of the graphite under irradiation and have predicted graphite bore diameter changes that are in good agreement with the measurements of graphite bore diameters taken at Ignalina Nuclear Power Plant (NPP). A probabilistic model has been developed using the actual results of the deterministic calculations with non-linear graphite behaviour. Statistical analysis of the measurements of tube and graphite diameters taken from Units 1 and 2 at Ignalina NPP has been carried out. Further work has been carried out to try to determine the uncertainty inherent in the predictions of the gas-gap closure from the calculations. The overall objective of the studies is to aid prediction of the gas-gap closure process, and help to identify a suitable monitoring strategy for gas-gap closure that could be used for any RBMK reactor.     

Neutron irradiation studies on low density pan fiber based carbon/carbon composites

Carbon has been extensively used in nuclear reactors and there has been growing interest to develop carbon-based materials for high-temperature nuclear and fusion reactors. Carbon-carbon composite materials as against conventional graphite material are now being looked into as the promising materials for the high temperature reactor due their ability to have high thermal conductivity and high thermal resistance. Research on the development of such materials and their irradiation stability studies are scant. In the present investigations carbon-carbon composite has been developed using polyacrylonitrile (PAN) fiber. Two samples denoted as Sample-1 and Sample-2 have been prepared by impregnation using phenolic resin at pressure of 30 bar for time duration 10 h and 20 h respectively, and they have been irradiated by neutrons. The samples were irradiated in a flux of 10¹² n/cm²/s at temperature of 40 °C. The fluence was 2.52 × 10¹⁶ n/cm². These samples have been characterized by XRD and Raman spectroscopy before and after neutron irradiation. DSC studies have also been carried out to quantify the stored energy release behavior due to irradiation. The XRD analysis of the irradiated and unirradiated samples indicates that the irradiated samples show the tendency to get ordered structure, which was inferred from the Raman spectroscopy. The stored energy with respect to the fluence level was obtained from the DSC. The stored energy from these carbon composites is very less compared to irradiated graphite under ambient conditions.     

A high-temperature longitudinal strain rate equation for Zr-2.5 wt% Nb pressure tubes

The pressure tubes in CANDU reactors are horizontal. Thus, if, during a postulated loss-of-coolant accident, the pressure tube temperature should rise sufficiently, the self-weight of the pressure tubes together with the weight of the fuel could cause the pressure tubes to sag. Since any pressure tube deformation would control how the core heat is transferred to the surrounding moderator, which is a large heat sink, the accurate prediction of this sag is essential. Most CANDU reactors have pressure tubes of cold-worked Zr-2.5 wt% Nb. A longitudinal strain rate equation was developed for this material using four-point bend tests. This strain rate equation was successful in predicting the longitudinal strain, due to bending, in specimens for which the temperature was ramped at 1°C/s and 5°C/s.     

Friction Pressure Drop and Heat Transfer Coefficient of Two-Phase Flow in Helically Coiled Tube Once-Through Steam Generator for Integrated Type Marine Water Reactor

Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report.     

Critical heat flux of water in vertical round tubes at low pressure and low flow conditions

An experimental study on critical heat flux (CHF) has been performed for water flow in vertical round tubes under low pressure and low flow (LPLF) conditions to provide a systematic data base and to investigate parametric trends. Totally 513 experimental data have been obtained with Inconel-625 tube test sections in the following conditions: diameter of 6, 8, 10 and 12 mm; heated length of 0.31.77 m; pressure of 106951 kPa; mass flux of 20277 kg m⁻² s⁻¹; and inlet subcooling of 50654 kJ kg⁻¹, thermodynamic equilibrium critical quality of 0.3231.251 and CHF of 1081598 kW m⁻². Flow regime analysis based on Mishima & Ishii’s flow regime map indicates that most of the CHF occurred due to liquid film dryout in annular-mist and annular flow regimes. Parametric trends are examined from two different points of view: fixed inlet conditions and fixed exit conditions. The parametric trends are generally consistent with previous understandings except for the complex effects of system pressure and tube diameter. Finally, several prediction models are assessed with the measured data; the typical mechanistic liquid film dryout model and empirical correlations of (Shah, M.M., 1987. Heat Fluid Flow 8 (4), 326–335; Baek, W.P., Kim, H.G., Chang, S.H., 1997. KAIST critical heat flux correlation for water flow in vertical round tubes, NUTHOS-5, Paper No. AA5) show good predictions. The measured CHF data are listed in Appendix B for future reference.     

Electric conductivity of neutron-irradiated aluminum alloys

The effect of ionizing radiation, temperature, and products of the radiolysis of water on the materials of fuel elements used in nuclear reactors during operation is examined. The effect of irradiation in a reactor on the electric conductivity and the change of the linear dimensions of the alloys SAV-1 and AMG-2 are investigated. It can be concluded on the basis of the relative elongation that the aluminum alloys studied can be used to fluence 10²⁰ cm⁻² without appreciable loss of reliability.     

CATHENA: A thermalhydraulic code for CANDU analysis

This paper describes the Canadian algorithm for thermal hydraulic network analysis (CATHENA) transient, thermalhydraulics code developed for the analysis of postulated upset conditions in CANDU®¹ reactors. The core of a CANDU reactor consists of a large number of horizontal pressure tubes containing fuel bundles. As a result of the unique design of the CANDU reactor, the CATHENA thermalhydraulic code has been developed with a number of unique modelling capabilities. The code uses a one-dimensional, two-fluid, nonequilibrium representation of two-phase flow. Some of the unique features of the CATHENA code are the one-step semi-implicit numerical method used and the solid heat transfer modelling capability that allows horizontal fuel bundles to be represented in detail. The code has been used in the design and analysis of CANDU-3, CANDU-6 and CANDU-9 reactors. The code has also been used for the design and analysis of the multiple applied lattice experimental (MAPLE) class of reactors and for the analysis of thermalhydraulic experimental programs conducted by Atomic Energy of Canada Limited (AECL).     

Anisotropy of irradiation creep of Zr-2.5wt%Nb and Zircaloy-2 alloys

Stress relaxation in bending tests have been used to determine the creep anisotropy of Zr-2.5wt%Nb and Zircaloy-2 alloys during fast neutron irradiation at 570 and 320 K. The ratios of creep rates from longitudinal and transverse sections were found to be larger in Zr-2.5wr%Nb materials than in Zircaloys at both temperatures, and larger at 320 K than at 570 K for both alloys. The creep anisotropy at 570 K for both alloys can be related to crystallographic texture by a model in which the strains are produced by slip of dislocations of the type $\text{1}\text{3}$〈1120〉, 65% on prism planes and 35% on basal planes; at 320 K the ratios of prism to basal slip are 84/16 in Zr-2.5wt%Nb and 73/27 in Zircaloy-2. The latter difference is attributed to alloying. An important implication of these results is that the reduction of elongation rates in CANDU¹ power reactors due to compressive end load is less than was previously believed.     

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.     

Irradiation swelling, creep and thermal-stress analysis of LWR fuel elements, computer code isune-2

The irradiation swelling, creep, and thermal-stress analysis of light-water reactor (LWR) oxide (UO₂) fuel elements is analysed. The analysis is based on the basic physical and mathematical assumptions and the experimental data of the fuel and cladding (or canning) materials. In the analysis, the nuclear, physical, metallurgical, and thermo-mechanical properties of the fuel and cladding materials under irradiation environment are examined carefully. The objectives of the paper are mainly (1) to formulate and carry out the irradiation swelling, irradiation creep, and thermal-stress analysis of fuel elements for LWR power reactors, and (2) to develop a computer code which will facilitate the computations for fuel element design, safety analysis, and economic optimization of the power reactors. In a general procedure of the analysis, the irradiation swelling, irradiation creep, temperature distribution, etc. in the fuel and cladding of the oxide fuel elements during the reactor in operation are studied. Some theoretical models and empirical relations (on the basis of accepted experimental data) for irradiation swelling and creep in the fuel and irradiation creep in cladding materials are postulated and developed. Some analytical and empirical relations (based on test results) for heat generation and temperature distribution in the fuel during fuel restructuring are derived. The fuel restructure is, in general, divided into the central void, columnar grain, equiaxed grain, and unaffected grain zones (or regions) after a sufficiently long period for the fuel elements to be irradiated (or operated). From these relations derived for irradiation swelling, irradiation creep, and temperature distribution in the fuel and cladding, together with the well-known strain-stress, incompressibility, compatibility, and stress equilibrium equations, the irradiation swelling, creep, and thermal-stress analysis for the LWR fuel elements can be carried out.From the analytical results obtained, a computer code, ISUNE-2 (which is in the sequence of computer code ISUNE-1 and -1A developed and used previously for liquid-metal fast breeder reactor fuel element design and safety and economic analysis), can be developed. With some reliable experimental data (measured during fuel elements in operation) as input, the computer code may predict various cases of LWR (oxide or carbide) fuel elements in operation. The general scope and resulting contribution of this paper is to provide a realistic analysis and a reliable operating LWR fuel element code for use by nuclear power utilities to predict the fuel element behavior in power reactors. The fuel element design, safety analysis, and economic optimization depend largely on the fuel element behavior in the power reactors.     

Structural integrity assessment of pressure tubes for Wolsong Unit 1 based on operational experiences

Pressure tube integrity has been considered as a key issue since the first operation of the CANDU reactor. Wolsong Unit 1 has been in service since 1983 and subjected to inspections three times covering 44 tubes. The in-service inspections revealed that a major portion of inspected tubes was in contact with calandria tubes. This is likely to increase the probability of blister formation which is a potential threat to pressure tube integrity. Fortunately, the inspection results indicated that no tube has been affected by blister formation so far. Nevertheless, to reduce the undue risk of blister formation the utility has decided to conduct spacer relocation every year until the entire core is covered. On the other hand, LBB analysis of pressure tubes using AGS performance measured at Wolsong Unit 2 indicated that the operational safety margin was marginal when using 15-year operational data. This raises the concern of pressure tube integrity at Wolsong Unit 1 which has a more than 15-year operation. This paper presents the overall integrity evaluation of Wolsong Unit 1 pressure tubes considering AGS performance test results and operating experience data.     

Life Time Estimation for Cladding Tube Cracking Caused by Absorber Swelling of PWR RCCA Rodlets

Intergranular cracks of cladding tubes had been observed at the tips of the rodlets of PWR rod cluster control assemblies (RCCAs). Because RCCAs are important core components, an investigation was carried out to estimate their service life time.

(1) As it is essential to know the effect of slumping of the neutron absorber for the life time estimation, tests on absorber material were carried out. Both the dynamic and static stresses of the absorber are sufficiently small compared with its mechanical characteristics and it is concluded that slumping does not occur.

(2) The crack was initiated at the inner surface of the cladding tubes and Sipush et al. obtained an intergranular fracture surface in tensile tests of similar material at a very slow strain rate in an argon gas atmosphere. Therefore the mechanism of the intergranular crack of the cladding tube is not IASCC but irradiation assisted cracking (IAC) caused by an increase in hoop strain due to the swelling of the absorber and a decrease in elongation due to neutron irradiation.

(3) The crack initiation limit of cylindrical shells made of low ductile material and subjected to internal pressure is determined in relation to the uniform strain of the material and is in accordance with that of the RCCA rodlets in an actual plant.

From the above investigation, the method of estimating the life time and countermeasures for its extension are obtained.     

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.     

Modeling of condensation in horizontal tubes

Condensation in horizontal tubes plays an important role for example in the determination of the operation mode of horizontal steam generators of VVER reactors or passive safety systems for the next generation of nuclear power plants. Two different approaches (HOTKON and KONWAR) for modeling this process have been undertaken by Forschungszentrum Jülich (FZJ) and the University for Applied Sciences Zittau/Görlitz (HTWS) and implemented into the 1D-thermohydraulic code ATHLET, which is developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH for the analysis of anticipated and abnormal transients in light water reactors. Although the improvements of the condensation models are developed for different applications between generators and emergency (VVER steam generators-emergency condenser of the SWR1000) with very different operation conditions (e.g. the temperature difference over the tube wall in HORUS is up to 30 K and in NOKO up to 250 K, and the heat flux density in HORUS is up to 40 kW/m² while that in NOKO is up to 1 GW/m²) both models are now compared and assessed by Forschungszentrum Rossendorf FZR e.V. Therefore post-test calculations of four selected HORUS experiments were performed with ATHLET/KONWAR. It can be seen that the calculations with the extension KONWAR as well as HOTKON significantly improve the agreement between computational and experimental data.