Zirconium Alloy Tubes Research Articles

Depth profiles of oxygen and nitrogen concentrations in weld metals produced by TIG welding Study of welding of zirconium alloy tubing (4th report)

Summary This paper describes an investigation of the depth profiles of oxygen [O] and nitrogen [N] in TIG weld metals of Zircaloy‐2 tubing by two methods: a calculation‐based method for chemical analysis of [N] in weld metals before and after etching to determine the [N] content in the weld metal surface layer, and a method for direct analysis by SIMS (secondary ion mass spectrometry). The results obtained from the evaluations made by the two methods may be summarised as follows. The [N] content of the weld metal produced in a welding atmosphere with a high PN2 is much higher in the outer part, exponentially decreasing towards the centre. The [N] content of the weld metal outer part decreases with a decreasing PN2 in the welding atmosphere, its depth profile changing to become flat for weld metals produced in a PN2 = 0.4 Pa welding atmosphere. The depth profile of [O] differs significantly from that of [N].

Investigation of swelling during seal‐welding of tubing and technology for its prevention: Study of welding of zirconium alloy tubing (1st report)

Summary Swelling is liable to occur during welding of tubing which has some internal gas volume. The present study aims to clarify this phenomenon experimentally and to develop a technology for its prevention during seal‐welding of Zircaloy‐2 tubing. Swelling occurs inside the weld metal when the pressure differential inside and outside reaches a critical value as the internal pressure increases. Swelling then develops towards the outside as the pressure increases.

TIG welding phenomena in a varying atmosphere and properties of welds: Study of welding of zirconium alloy tubing (2nd report)

TIG welding phenomena in a varying atmosphere and properties of welds: Study of welding of zirconium alloy tubing (2nd report)

時間的に変化する雰囲気中でのＴＩＧ溶接現象と溶接部の性質 ジルコニウム合金管の溶接に関する研究（第２報）

時間的に変化する雰囲気中でのＴＩＧ溶接現象と溶接部の性質 ジルコニウム合金管の溶接に関する研究（第２報）

Analysis of Palladium Coatings to Remove Hydrogen Isotopes from Zirconium Fuel Rods in Canada Deuterium Uranium-Pressurized Heavy Water Reactors: Thermal and Neutron Diffusion Effects

This paper reports that, in pressurized heavy water nuclear reactors of the type standardly used in Canada (Canada deuterium uranium-pressurized heavy water reactors), the zirconium alloy pressure tubes of the core absorb deuterium produced by corrosion reactions. This deuterium weakens the tubes through hydrogen embrittlement. Thin palladium coatings on the outside of the zirconium are analyzed as a method for deuterium removal. This coating is expected to catalyze the reaction D{sub 2} + 1/2O{sub 2} {r reversible} D{sub 2}O when O{sub 2} is added to the annular (insulating) gas in the tubes. Major reductions in the deuterium concentration and, hence, hydrogen embrittlement are predicted. Potential problems such as plating the tube geometry, neutron absorption, catalyst deactivation, radioactive waste production, and oxygen corrosion are shown to be manageable. Also, a simple set of equations are derived to calculate the effect on diffusion caused by neutron interactions. Based on calculations of ordinary and neutron flux induced diffusion, a palladium coating of 1 {times} 10{sup {minus}6} m is recommended. This would cost approximately $60,000 per reactor unit and should more than double reactor lifetime. Similar coatings and similar interdiffusion calculations might have broad applications.

A Comparison of X-Ray and Neutron Diffraction Measurement of Lattice Parameters for the Determination of Residual Stress and Chemical Composition in Zirconium Alloy Tubes

AbstractIntergranular residual stresses can exist in zirconium alloys, especially when there is a large distribution of grain orientations. The stresses result from the anisotropic plasticity and thermal expansion of the hexagonal close-packed crystal structure of α-zirconium. Apart from complicating the characterisation of materials using lattice parameter measurements, the intergranular stresses can significantly affect material behaviour, especially in nuclear reactor environments, and there is therefore a great deal of interest in their measurement.The effects of specimen preparation and surface relaxation on X-ray diffraction measurements of lattice parameters of zirconium alloys have been investigated by comparing bulk neutron diffraction with X-ray diffraction on identical materials. The results show that: (i) intergranular or interphase residual stresses exist in dual-phase Zr-2.5Nb pressure tubes; (ii) the stresses normal to the surface of an X-ray diffraction specimen are not relieved completely when there are intergranular residual stresses in the material. One can conclude that intergranular stresses have to be considered when determining chemical compositions from lattice parameter measurements and also when measuring macroscopic residual stress using X-ray diffraction.

A probabilistic approach to leak-before-break in CANDU pressure tubes

In the candu R reactor, the coolant passes through the core in zirconium alloy pressure tubes. A few of these pressure tubes have leaked at cracks near the rolled joint where the pressure tube is attached to the end fitting. A probabilistic methodology, and associated computer code (called marathon), has been developed to calculate the time from first leakage to unstable fracture in a probabilistic format. The methodology explicitly uses material property distributions, and allows the risk associated with leak-before-break to be estimated. A model of the leak detection system is included to calculate the time between leak detection and unstable fracture. The sensitivity of the risk to changing reactor conditions allows the optimization of reactor management after leak detection. Preliminary material property distributions show the probability of unstable fracture is very low, and that ample time is available to shut down the unit and locate the leaking tube.

ORCHID?A computer simulation of the reliability of an NDE inspection system

CANDU pressurized heavy water reactors contain several hundred horizontally-mounted zirconium alloy pressure tubes. Following a pressure tube failure, a pressure tube inspection system called CIGARette was rapidly designed, manufactured and put in operation. Defects called hydride blisters were found to be the cause of the failure, and were detected using a combination of eddy current and ultrasonic scans. A number of improvements were made to CIGARette during the inspection period. The ORCHID computer program models the operation of the delivery system, eddy current and ultrasonic systems by imitating the on-reactor decision-making procedure. ORCHID predicts that during the early stage of development, less than one blistered tube in three would be detected, while less than one in two would be detected in the middle development stage. However, ORCHID predicts that during the late development stage, probability of detection will be over 90%, primarily due to the inclusion of axial ultrasonic scans (a procedural modification). Rotational and axial slip could severely reduce probability of detection. Comparison of CIGARette's inspection data with ORCHID's predictions indicate that the latter are compatible with the actual inspection results, though the numbers are small and data uncertain. It should be emphasized that the CIGARette system has been essentially replaced with the much more reliable CIGAR system.

Release of oxides of nitrogen in etching Zr+1% Nb tubes in a mixture of nitric and hydrofluoric acids

Toxic oxides of nitrogen NO/sub x/ are released on etching fuel-pin tubes made of zirconium alloys, which adversely effect the working conditions and produce air pollution. In this paper, the authors examine the effects of some technological parameters on the release of these oxides during zirconium-alloy tube etching in a mixture of nitric and hydrofluoric acids, and they also examine ways of reducing the yield of NO/sub x/ during the etching. The experiments were performed with a thermostatic teflon reactor. The region that was examined and the variation intervals are shown. The authors used experiment planning with the scheme for a 2/sup 5-1/ fractional factor experiment (FFE). A planning matrix for the 2/sup 5-1/ FFE and the measurement results are presented. It is shown that the release of the NO/sub x/ directly into the gas and the accumulation in the solution vary considerably with the etching conditions. The regression equations and the ranking diagram presented here provide scope for reducing the release of oxides of nitrogen during etching zirconium-alloy tubes.

In-reactor deformation of zirconium alloy tubes at 330 K

Internally pressurized tubes of Zircaloy-2 and Zr-2.5 wt% Nb were irradiated in a fast neutron flux of 1.75 × 10 17 n/( m 17· s) at about 330 K for up to 27000 h. Periodic measurement of diameters and lengths have shown that strain rates are still changing after 27000 h and that the ratio of transverse to axial strains is consistent with that calculated from the crystallographic texture. Annealed Zircaloy-2 showed smaller strains than either cold worked Zircaloy-2 or cold worked Zr-2.5 wt% Nb at equivalent stresses.

Anisotropy of irradiation creep and growth of zirconium alloy pressure tubes

Irradiation creep and growth anisotropy factors calculated from crystallographic texture and grain shape measurements, together with measured axial and transverse strain rates of zirconium alloy pressure tubes in CANDU ∗ ∗ CANada Deuterium Uranium. power reactors, have been used to calculate proportions of creep and growth. Using the irradiation creep and growth rates derived for cold drawn Zr-2.5%Nb and cold drawn Zircaloy-2 pressure tubes, we have defined ideal textures and grain shapes resulting in minimum axial and transverse strain rates. For the conditions of stress, temperature and fast neutron flux under which CANDU pressure tubes operate, proportions of growth are too small compared to the creep to get the ideal of zero axial and transverse strains. The closest approach to this ideal is when basal poles are in the radial/transverse plane, regardless of grain shape of alloy. The most favourable grain shape is plate shaped with the plates oriented in the axial/radial plane.

Fracture Control in the Nuclear Power Industry

AbstractThe majority of the world's commercial power reactors are similar to U.S.-designed pressurized-water (PWR) or boiling-water (BWR) designs in which the nuclear core resides in a large, heavy-walled ferritic steel pressure vessel. Fracture control of these vessels combines the quantitative approach of linear elastic fracture mechanics (LEFM) with qualitative materials testing: lower-bound fracture toughness data and conservative applied stress intensities are used to estimate maximum permissible flaw sizes, and Charpy properties are related to minimum permitted service temperature in the steel containment. The Canadian designed CANDU–PHW reactors employ hundreds of small-diameter zirconium alloy pressure tubes, rather than a single large vessel, as the primary containment for the pressurized coolant. Fracture control philosophy is based on the demonstrated tolerance of pressure tubes to artificial flaws, proven in a large, continuing experimental program. Pressure tubes have been shown to “fail-safe...

Relationship between contractile strain ratioR and texture in zirconium alloy tubing

The crystallographic textures of zirconium alloy tubing used as cladding in nuclear reactor fuel are commonly characterized by the quantitative texture numbers F (Kaellstroem) and f/sub r/ (Kearns) which are derived from the direct and inverse pole figures. The texture numbers of zircaloy 2 and 4 tubes have been correlated experimentally with the value of the contractile strain ratio R which is a measure of the plastic anisotropy of the tube. The correlations were based on the results of 20 different tubing lots. The f/sub r/-R correlation shows much less data scatter than the F-R correlation. By assuming a simple plastic deformation model for zirconium alloys the following relations between texture and anisotropy are obtained: F = R-1/R+1 and f/sub r/ = R/R+1. The theoretically derived relations are in good agreement with the experimental data. The procedure of correlating texture with plastic anisotropy is not limited to zirconium alloy tubing, but should be equally applicable to textured sheet and plate materials and other alloys with a limited number of slip systems.

Fracture behaviour of zirconium alloy pressure tubes for Canadian nuclear power reactors

The fracture behaviour of zirconium alloy pressure tubes was measured using internally pressurised specimens containing artificial flaws and three-point bend specimens loaded in impact. Four materials were tested: cold-worked Zircaloy-2 and Zr-2·5 wt% Nb, heat treated Zr-2·5 wt% Nb and a recently-developed complex alloy called XL, also cold-worked. Critical crack lengths pertinent to reactor operating conditions exceed fifteen times tube wall thickness and are large enough to claim ‘leak-before-break’ as a realistic safety criterion. Fracture modes during crack extension are discussed; these are influenced by temperature and hydrogen but are always ductile at reactor operating temperatures ( > 513 K ). Relevant aspects of delayed hydrogen cracking are discussed: effects are highly localised at stress concentrations and will not affect conclusions about tube fracture behaviour.

Neutron Radiation Damage in Zirconium Alloys

Exposure of zirconium alloys to fast neutron irradiation produces defect clusters and dislocation loops of dimensions and distribution which are determined by fluence, temperature and alloy composition; particularly the amount of alloy addition in solid solution in the h.c.p. α-Zr phase. Here I describe the current status of CRNL work on the TEM examination of the damage produced at ≈575 K which is the operating temperature for zirconium alloy pressure tubes and fuel sheathing in CANDU-PHW reactors. Specimens were prepared as in ref. (1) and were examined in a Hitachi 11A microscope fitted with a modified HK3A ±30° double axis tilt stage (2).

A simple method for determining the pole figure of zirconium alloy tubing

The paper describes a simple inclined section method for determining the pole figure of heavily drawn zirconium alloy tubing from a single texture run on a composite specimen consisting of up to 24 tube sections. The technique can be used for other materials and results in considerable savings in time and effort. The problem of obtaining reliable random intensity levels for a composite specimen is overcome by integrating the total diffracted intensity during the texture run. This data can be used to calculate a random intensity level for the particular specimen and set of experimental conditions used in the texture run. The integrated intensity values can also be used to calculate both the orientation parameter and the fraction of prism planes oriented for slip. The calculation is simple and gives relatively accurate results.

Fracture of zirconium alloys in iodine vapour

Zirconium alloy tubing in the cold-worked condition has proved susceptible to short-term brittle transgranular cracking when exposed to iodine vapour over the temperature range 250–510 °C at tensile hoop stresses close to the 0.2% proof level. Thermal annealing reduced the cracking susceptibility of Zircaloy fabricated by cold-drawing but did not completely remove it after 850 °C, 1 h; in contrast a 650 °C, l h anneal was sufficient to prevent cracking in tube-reduced Zircaloy. Textural differences arising from the forming processes are considered to be responsible for this behaviour. Since the failures resulted from metal-gas interactions, the dominant mechanism is suggested to be one of mechanical fracture rather than electrochemical attack, a conclusion supported by results of fractographic examination.

Nuclear reactor channel with zirconium alloy pressure tube at optimum operating temperature: Application to a reactor cooled with organic liquid

A confrontation with problems relating to the development of a pressure tube for an organic cooled heavy water moderated reactor (ORGEL type) lead us to make a preliminary study of a channel which would utilise known zirconium alloys in a safe temperature range. This concept has not been further developed. Hydrogen absorption (with consequent embrittlement) in zirconium alloys in contact with coolants releasing hydrogen, such as organic liquids, increases rapidly with the operating temperature. In the proposed solution the control of the hydrogen absorption is acheived by limiting the temperature of the pressure tube. The proposed solution also avoids the use of zirconium alloys at a too low temperature. By means of two insulating layers, one outside the fule element and the other between the pressure tube and the moderator, the temperature of the pressure tube is kept low enough to avoid hydrogenation, while at the same time remaining above the ductile-brittle transition temperature of the material. The thermal insulatant on the fuel element also acts as its supporting structure, and largely protects the cans against mechanical stresses. Lastly, some solutions are presented, together with the relevant considerations, particularly in relation to reactors cooled with organic liquid.

The development of texture in zirconium alloy tubes

The crystallographic texture of a large number of zirconium alloy tubes has been related to the fabrication history. The texture developed is controlled by the direction of major strain in the final stages of fabrication. A production route can now be specified to achieve the desirable texture of basal plane normals oriented radially in the tube.

Ultrasonic inspection of tubes

The paper deals with the principles of ultrasonic methods of detecting flaws and wall thickness variations in tubes of diameter 0.5 in to 5.5 in with bore/wall-thickness ratios of about 25:1. It pays particular attention to the best probe arrangement. A tube testing tank embodying these principles and particularly designed for versatility in testing pressure tubes and fuel element sheaths of zirconium alloys has been installed. The instruments are mainly an assemblage of commercially available ultrasonic testing equipment feeding either a pen recorder or facsimile recorder according to the detail required from the inspection. Some examples are given of tests on zirconium alloy tubes, the records of inspection having been correlated with metallographic examination. Complete ultrasonic examination can be done with the present instruments at speeds of a few feet per minute, the exact speed depending on the diameter of the tube, but there are possibilities for increasing the rate of inspection