Oxygen Concentration In Sodium Research Articles

ИССЛЕДОВАНИЕ ГЕТТЕРНОЙ ОЧИСТКИ НАТРИЯ ОТ КИСЛОРОДА

The article focuses on a comprehensive analysis of the new concept of a combined sodium oxygen purification module. The combined purification module includes a combination of two modules: chemisorption purification with a soluble getter and high-temperature getter purification with a high-temperature getter. A special zirconium alloy in a finely dispersed granular form can be used as a high-temperature getter. Other materials (e.g. titanium and various alloys) can be used and must be tested accordingly. At the first stage of the work, research was carried out on the module for chemisorption purification of sodium from oxygen at the Protva-1 stand, which showed the effectiveness of purification using the proposed getter materials, at the second, research on the module for high-temperature getter purification. For 20 hours of operation of the chemisorption module, the oxygen concentration in sodium was reduced by 12 ppm (3 g oxygen), while 4.5 g of magnesium reacted (the initial mass of magnesium was 25 g). According to the results of experiments in the sodium loop, the decrease in the concentration of dissolved oxygen in sodium during purification with a zirconium getter for 20 h was about 50 ppm. Two promising options for the development of a combined module for the purification of sodium from oxygen are proposed. Experimental justifications of selected chemisorbent, getter and filtering material are carried out. The main part of the study is analysis of chemisorption efficiency, purification process dynamics and influence on it of temperature factor for selected chemisorbent.

Evaluation of cobalt free coatings as hardfacing material candidates in sodium-cooled fast reactor and effect of oxygen in sodium on the tribological behaviour

The feedback produced by operating Sodium-cooled Fast Reactors (SFRs) has shown the importance of material tribological properties. Where galling or adhesive wear cannot be allowed, hardfacing alloys, known to be galling-resistant coatings, are usually applied on rubbing surfaces. The most used coating is the cobalt-base alloy named Stellite 6® because of its outstanding friction and wear behaviour. Nevertheless, cobalt is an element which activates in the reactor leading to complex management of safety during reactor maintenance and mainly decommissioning. As a consequence, a collaborative work between CEA, EDF and FRAMATOME has been launched for selecting promising cobalt-free hardfacing alloys for the 600 MWe Sodium-cooled Fast breeder reactor project named ASTRID. Several nickel-base alloys have been selected from literature review then deposited by Plasma Transferred Arc or Laser Cladding on 17Cr austenitic stainless steel 316L(N) according to RCC-MRx Code (AFCEN Code). Among the numerous properties required for qualifying their use as hardfacing alloys in SFR, good corrosion behaviour and good friction and wear behaviour in sodium are essential. First results on these properties are shown in this article. Firstly, the corrosion behaviour of all coatings was evaluated through exposure tests in purified sodium for 5000 h at 400 °C. All coatings showed an acceptable corrosion behaviour in sodium. Finally, the friction and wear properties of one alloy candidate, NiCrBSi alloy, were studied in sodium in a dedicated designed facility. The influence of the oxygen concentration in sodium on the friction and wear properties was evaluated.

Specifics of high-temperature sodium coolant purification technology in fast reactors for hydrogen production and other innovative applications

In creating a large-scale atomic-hydrogen power industry, the resolution of technological issues associated with high temperatures in reactor plants (900°C) and large hydrogen concentrations intended as long-term resources takes on particular importance. The paper considers technological aspects of removing impurities from high-temperature sodium used as a coolant in the high-temperature fast reactor (BN-HT) 600MW (th.) intended for the production of hydrogen as well as other innovative applications. The authors examine the behavior of impurities in the BN-HT circuits associated with the mass transfer intensification at high temperatures (Arrhenius law) in different operating modes. Special attention is given to sodium purification from hydrogen, tritium and corrosion products in the BN-HT. Sodium purification from hydrogen and tritium by their evacuation through vanadium or niobium membranes will make it possible to develop compact highly-efficient sodium purification systems. It has been shown that sodium purification from tritium to concentrations providing the maximum permissible concentration of the produced hydrogen (3.6Bq/l according to NRB-99/2009) specifies more stringent requirements to the hydrogen removal system, i.e., the permeability index of the secondary tritium removal system should exceed 140kg/s. Provided that a BN-HN-type reactor meets these conditions, the bulk of tritium (98%) will be accumulated in the compact sodium purification system of the secondary circuit, 0.6% (∼ 4·104Bq/s), will be released into the environment and 1.3% will enter the product (hydrogen). The intensity of corrosion products (CPs) coming into sodium is determined by the corrosion rate of structural materials: at a high temperature level, a significant amount of corrosion products flows into sodium. The performed calculations showed that, for the primary BN-HT circuit, the amount of corrosion products formed at the oxygen concentration in sodium of 1mln–1 exceeds 900kg/yr with fuel element claddings made of EP-912-VD steel and 464kg/yr with molybdenum alloy claddings. For the secondary circuit, the amount of corrosion products totals 263kg/yr for each loop. Taking into account the high-temperature experiments which demonstrated high efficiency of retaining corrosion product suspensions by the strainers located in the low-temperature area, it is proposed to cool sodium to the required temperature alongside the corrosion products retention on the mass transfer surfaces, including strainers. It is shown that, by using 30% of the power required to produce hydrogen with 50% efficiency, the BN-HT is capable of producing about 0.6·106m3 of hydrogen per 24 hours which is sufficient for modern large-scale enterprises processing medium-grade crude oil or the implementation of other technologies.

Development of an Electrochemical Oxygen Sensor for Liquid Sodium Using a Yttria Stabilized Zirconia Electrolyte

An accurate means of measuring oxygen concentration in sodium is crucial to characterizing the corrosion of sodium-cooled fast reactor (SFR) materials. Significant improvements with regards to lifetime/sensitivity/temperature threshold/etc. can be made to current Galvanic-cell oxygen sensing technology, so extensive research has been conducted at the University of Wisconsin-Madison to address these issues. Electrochemical oxygen sensors using yttria-stabilized zirconia (YSZ) electrolytes were designed and tested in liquid sodium up to 400[°C]. These sensors produced an oxygen-dependent signal, and this signal followed the expected functional form, but the lifetime and reproducibility of the sensors were inadequate for commercial use. Sensor performance was analyzed and temperature effects on ionic conductivity, non-bulk thermodynamics, and YSZ-based issues were quantified to study deviations from theory in both slope and signal offset.

Oxygen Interactions between Sodium and Uranium-Plutonium Oxide Fuel

The oxygen concentration of sodium in equilibrium with uranium-25% plutonium oxide fuel was measured at temperatures of 650 to 900 deg C for fuels with oxygen-to-metal ratios of 1.90, 1.94, and 1.97. The oxygen concentration in sodium at the three-phase Na -- MO/sub 2-x/ Na/sub 3/MO/sub 4/ (M = uranium plus plutonium) equilibrium has also been measured for the same temperature range. This three-phase equilibrium was established by reacting mixed-oxide fuel with sodium to which Na/sub 2/O had been added. The oxygen concentrations in sodium were determined by the vanadium-wire equilibration method. The oxygen concentration in sodium at the three-phase equilibrium varied from ~ 0.1 to 0.4 ppM in the temperature range investigated. Oxygen concentrations in EBR-II primary sodium, which have been measured by the same method, vary from 0.4 to 0.9 ppM oxygen depending on the cold-trap operation. These values indicate that the sodium--fuel reaction product, i.e., Na/sub 3/MO/sub 4/, is stable in sodium at the temperatures (<1000 deg C) and oxygen levels present in EBR-II. (auth)

Solubility of oxygen in vanadium

Three techniques have been employed to establish the temperature dependence of the terminal solubility of oxygen in vanadium in the range 200 °–750 °C. Internal friction measurements, which indicate the amount of oxygen in solution in a body-centered cubic metal, were used to determine the phase boundary to 400 °C. Saturation concentrations at the higher temperatures were determined from oxygen distribution experiments in the VONa system. The low oxygen partial pressures, e.g., 10 −34-10 −47 atm, required for vanadium-vanadium oxide phase equilibrium were obtained by controlling the oxygen concentration in liquid sodium at levels of 0.03 to 20 ppm. The oxygen partial pressures at the three-phase sodium-vanadium-vanadium oxide equilibrium were calculated from measured oxygen concentrations in sodium and from available thermodynamic data for the binary sodium-oxygen system. Gaseous oxidation experiments conducted in a Sieverts' apparatus confirmed that sodium interactions or complex oxidation formation did not affect the previous results. The oxygen concentration of α-vanadium, which is in equilibrium with an oxide designated V 9O, is adequately expressed by the equation log C 0 o ( wt.%) = 0.705 − 515( 1 T )(° K −1) for the temperature range investigated.

Calibration of Electrochemical Oxygen Meters in Sodium Using Uranium

Methods using uranium to remove or getter oxygen quantitatively from sodium have been used to calibrate the potential of the meter as a function of the oxygen concentration in sodium. The oxygen meters were constructed with sodium‐sodium oxide reference electrodes and electrolytes. The calibration methods depended on the deviation of these meters from ideal behavior. The methods involved the summation of the weight gains of uranium, i.e., the weights of oxygen removed from the sodium, the derivative of the weight gains for a given potential change as a function of the number of repetitive oxygen removal steps, and the time dependence of the potential during the removal of oxygen from sodium by uranium. The behavior of a meter depended on the purity of the electrolyte used. High‐purity electrolytes gave the theoretical dependence of potential on concentration but impure tubes showed marked deviations from theoretical behavior at low concentrations at temperatures of 500°C.

The determination of oxygen concentration and mass transfer coefficients in sodium using uranium

Uranium in sodium at 500 to 650 °C has been found to react with oxygen at a rate proportional to the oxygen concentration in the sodium. The rate of this reaction is controlled by the mass transfer of oxygen in the sodium and not by that in the uranium oxide formed as an adherent reaction product. Methods are discussed whereby uranium can be used to determine oxygen concentrations and mass transfer coefficients in sodium.