Thermodynamic and experimental study of UC powders ignition

Mixed plutonium and uranium carbide (UPuC) is considered as a possible fuel material for future nuclear reactors. However, UPuC is pyrophoric and fine powders of UPuC are subject to temperature increase due to oxidation with air and possible ignition during conditioning and handling. In a first approach and to allow easier experimental conditions, this study was undertaken on uranium monocarbide (UC) with the aim to determine safe handling conditions for the production and reprocessing of uranium carbide fuels. The reactivity of uranium monocarbide in oxidizing atmosphere was studied in order to analyze the ignition process. Experimental thermogravimetric analysis (TGA) and differential thermal analysis (DTA) revealed that UC powder obtained by arc melting and milling is highly reactive in air at about 200 °C. The phases formed at the various observed stages of the oxidation process were analyzed by X-ray diffraction. At the same time, ignition was analyzed thermodynamically along isothermal sections of the U-C-O ternary diagram and the pressure of the gas produced by the UC + O₂ reaction was calculated. Two possible oxidation schemes were identified on the U-C-O phase diagram and assumptions are proposed concerning the overall oxidation and ignition paths. It is particularly important to understand the mechanisms involved since temperatures as high as 2500 °C could be reached, leading to CO(g) production and possibly to a blast effect.     

Hydrogen permeability in Kh18N10T stainless steel from plasma glow-discharge

Conclusions In the entire temperature range under consideration (520–1000°K), the radiation-induced desorption rate is higher than the thermal desorption rate.In the temperature range below 700°K, the magnitude of hydrogen permeability is determined by the radiation-induced desorption rate that depends on temperature according to the equationIn the 700–1000°K range this rate is independent of temperature, and is equal to 7.10^–3 cm/sec.Above 1000°K, the rates of the radiation-induced desorption, thermal desorption, and diffusional discharge are comparable, and the exponential temperature dependence of flux would be disrupted.Translated from Atomnaya Énergiya, Vol. 59, No. 4, pp. 269–273, October, 1985.     

Thermodynamic Properties of U0.718Ce0.282O2±x

An experimental study of the structure and thermodynamic properties of a sample of U_0.718Ce_0.282O_2±x , which is an analogue of uranium–plutonium oxide, is performed to study the effect of adding cerium oxide on the properties of oxide fuel. The experimental specimens were obtained by powder metallurgy. x-Ray diffraction analysis of sintered tablets showed the formation of an equilibrium one-phase solid substitution solution based on UO₂ with substoichiometric composition. The potentiometric method of the emf of a solid-electrolyte galvanic cell in the range 850–1050°C and solid-phase coulometric titration with working temperature 1000°C were used to study the dependence of the oxygen potential of uranium-cerium oxide on the oxygen/metal ratio and the temperature. The concentration dependences of the enthalpy and the entropy of dissolution of oxygen in uranium-cerium oxide are calculated and constructed and it is determined that these dependences vary strongly at near-stoichiometric composition. The results can be used to analyze the properties of uranium–plutonium oxide fuel.     

Application of K-type nuclear emulsion in personnel fast-neutron dosimetry

Techniques for monitoring individual neutron dosage in the 0.5–15 MeV neutron energy range are described. K-type nuclear photographic emulsion of Soviet manufacture is used as detector. The emulsion is placed in a special correcting stack to ensure that the number of particle tracks in the emulsion per dose unit will be practically independent of neutron energy in the range mentioned. The method offers a sensitivity of 4 · 10^–4 tracks per Po-Be neutron, and the accuracy attainable in measuring the monthly critical tolerance dose is ±20%In conclusion, the authors appreciate this opportunity to express theirheartfelt gratitude to A. Yu. Deberdeev and M. D. Deberdeev for their invaluable assistance in preparing the emulsion film, to the members of the team headed by S. I. Lyubomilov for processing the materials obtained, and to V. S. Martynova for her efficient inspection of the films.     

85Kr Extraction from Water Samples

A system for extracting krypton from a sample of natural water preextracted under field conditions is described. A carbon trap at –76°C is used to precipitate inert gases. The radiation of ⁸⁵Kr is detected using a 3.5-liter multiwire proportional counter, placed on the ground and in a gallery at a depth of 500 m water equivalent in a low-background counter. The background of the counter enclosed in shielding with a built-in anticoincidence ring in the ⁸⁵Kr -particle energy range is 0.8 counts/min on the surface and 2 counts/h in the underground laboratory.Various schemes for degassing and removing krypton from water samples are examined – heating, bubbling, and vacuum degassing. The extraction coefficients for dissolved gases are as follows: 88 ± 2% for extraction by heating water, 95 ± 3% for bubbling, and 70 ± 10% for vacuum degassing.Preliminary measurements show that the ⁸⁵Kr concentration in different natural waters in Krasnodarsk krai is 130–1080 decays/min·mmole.     

Determination of the235U concentration in the aqueous coolant of the first circuit of a nuclear reactor

Conclusions A technique was developed for determining the²³⁵U concentration in the aqueous coolant of the first circuit of a nuclear reactor: detection limit 3·10^–12 g/cm³. Using the method in the IVV-2M reactor has shown that with this technique, an operational monitoring of the uranium concentration in the coolant and in the fluids washed from the surface of the first circuit, as well as monitoring other qqueous samples, is possible.Lavsan, which is directly irradiated in a liquid sample and electrochemically etched, can be recommended as a detector. The optimal conditions of etching 180-m-thick lavsan (after irradiation with thermal neutrons to a flux of (1–2)·10¹⁶ cm^–2) are: 30% aqueous KOH solution, a temperature of (70±0.2)°C, an electric field strength of 20 kV/cm, a frequency of 4 kHz, and an etching time of 100 min.Translated from Atomnaya Énergiya, Vol. 61, No. 5, pp. 334–338, November, 1986.     

Gamma and neutron dosimetry in nuclear reactors by means of colored polyvinyl alcohol films

It is shown that polyvinyl alcohol films containing methylene blue which become discolored under the effects of radiation are suitable for gamma and neutron radiation dosimetry in nuclear reactors. The degree of discoloration varies linearly with the dose in the 10⁴–10⁶ rad range and, over a broad range of values, is practically independent of variations in linear energy transfer, dose rate, and temperature. Films containing boric acid in addition to the dye can be used for recording thermal-neutron doses in the 10¹²–10¹⁴ neutrons/cm² range. The measurement error is within ±10% in every case.Translated from Atomnaya Énergiya, Vol. 19, No. 3, pp. 273–276, September, 1965.     

Use of concretes for high-temperature shielding of nuclear reactors

The authors have studied the possibility of using chromite and chomotte heat-resistant concretes for the thermal shields of reactors. They observe neutron fluxes of various intensities (up to 10¹³ neutrons/cm²·sec, with spectrum similar to fission spectrum), absorbed by shields of these materials. They compute the transmission of neutrons and of fluxes of gamma quanta and the heat emission in the shielding. They calculate the temperatures in the shielding for various neutron fluxes, concrete thicknesses and cooling conditions. They perform a statistical calculation of the temperature stresses for shielding constructed of heat-resistant ferroconcrete.It was established that nuclear reactor shields can be made from heat-resistant ferroconcrete when the neutron fluxes on the concrete are up to 10¹³ neutrons/cm²·sec, for temperatures up to 1000–1100° C and temperature differences of up to 900° C.Translated from Atomnaya Énergiya, Vol. 19, No. 6, pp. 524–529, December, 1965Report read by G. I. Budker at the International Conference on High-Energy Accelerators (Frascati, Italy).     

Temperature dependence of erosion of alloys of vanadium and niobium during bombardment with helium ions

Conclusions It has been established that maximal erosion of vanadium and its alloys is observed at 700–900°K, the corresponding range for niobium alloys being 900–1100°K. The maximal value of the erosion coefficients of alloys for vanadium and alloys V+25% Zr+C, Nb+4.2% Mo+Zr, and Nb+1.1% Zr+C is 1.5±0.7, 0.6±0.3, 0.4±0.2, and 0.15±0.07, respectively.Translated from Atomnaya Énergiya, Vol. 42, No. 1, pp. 13–15, January, 1977.     

Radiation-Induced Change in the Properties of Isotropic Structural Graphite

Three types of samples of isotropic graphite with different grain density and size were irradiated in a BOR-60 reactor up to neutron fluence (1.7–2.8)·10²⁶ m^–2 (E > 0.18 MeV) at 360–400°C. After irradiation, the change in the dimensions, resistivity, linear thermal expansion coefficient and dynamic elastic modulus were investigated. It was determined that the density in the range 1.67–1.76 g/cm³ results in an increase of the maximum weight and depth of volume shrinkage of isotropic fine-grain graphite. An equation was proposed for fitting the temperature dependence of the critical neutron fluence in the range 380–780°C for the experimental graphite samples.     

Physical properties of an alumino-silicate waste form for cesium and strontium

Nuclear fuel reprocessing will be required to sustain nuclear power as a baseload energy supplier for the world. New reprocessing schemes offer an opportunity to develop a better strategy for recycling elements in the fuel and preparing stable waste forms. Advanced strategies could create a waste stream of cesium, strontium, rubidium, and barium. Some physical properties of a waste form containing these elements sintered into bentonite clay were evaluated. We prepared samples loaded to 27% by mass to a density of approximately 3 g/cm³. Sintering temperatures of up to 1000 °C did not result in volatility of cesium. Instead, the crystallinity noticeably increased in the waste form as temperatures increased from 600 to 1000 °C. Assemblages of silicates were formed. Significant water evolved at approximately 600 °C but no other gases were generated at higher temperatures.     

X-ray diffraction of UC-UC2 and UC-UN alloys at elevated temperatures

Unit-cell dimensions of U-C-N alloys were measured at temperatures of 760 to 2250 °C with an X-ray diffractometer. Lattice dimensions of the face-centered cubic phase in both the UC-UC₂ and UC-UN solutions are found to be linear functions of the UC mole fraction. In carbon saturated body-centered tetragonal α-UC₂ between 1000 and 1400 °C, a small but distinct decrease in rate of volume change (slope) occurs with increasing temperature. The data suggest that the C/U ratio in α-UC₂ at 1000 °C and below is 1.97± 0.03, a value in excess of that prevailing in carbon-saturated α-UC₂ at higher temperatures.     

Influence of the neutron flux density on the radiation embrittlement of VVér-440/213 vessel materials

In work on minisamples of the fifth complex of the No. 3 unit of the Kola nuclear power plant it is shown that for neutron fluence 410²³ m^–2 (operation for approximately 10 yr), neutron flux density 310¹⁵ sec^–1m^–2 and copper content 0.03% and 0.09% in the metal the shifts of the cold-brittleness temperature are 50 and 120°C, respectively. Under the same irradiation conditions but with neutron flux density 310¹⁶ sec^–1m^–2, this shift for standard samples is 50°C. These results attest to the state of the vessel material at a given moment in time.Translated from Atomnaya Énergiya, Vol. 97, No. 3, pp. 177–182, September, 2004.     

Radiation dosimeters based on thermoluminescence measurements in aluminum phosphate glass (IKS dosimeters)

Glass dosimeters useful in beta-gamma dosimetry, slow neutron dosimetry, and the dosimetry of high-energy charged particles in the 0.02 to (1–2)·10⁶ rad range have been developed. The dosimeters are capable of storing and retaining information for an unusually long interval (as long as a month in a 150°C environment). These glasses are not excited by daylight, but daylight does exert a deexcitation effect: 26 to 38% of the stored light sum is dissipated by deexcitation in a period of 40 days.The effective atomic number of the optimum glass recipes is 11 to 13. A fitter of 0.6mm Sn+0.5 mm Al helps counteract the "hardness variation" in the range from 40 keV on higher, within an error of ±20%. The glass dosimeters are usable repeatably.Translated from Atomnaya Énergiya, Vol. 15, No. 1, pp. 48–52, July, 1963     

Mössbauer study and structural characterization of UO2–Gd2O3 sintered compounds

Samples of UO₂and up to 10 wt% of Gd₂O₃ were prepared by solid-state reaction under a reducing atmosphere, in a thermal path comprising ramps and dwell times in the temperature range of 900–1750 °C. The sintered material was analyzed by X-ray diffraction and ¹⁵⁵Gd Mössbauer spectroscopy. The results showed that for samples annealed up to 900 °C, the gadolinium sesquioxide remained unreacted. However, when the temperature was increased to 1300 °C, a solid-state reaction took place forming mixed oxides. For the more severe sintering condition, at 1750 °C, gadolinia left urania partially unreacted producing a material consisting of two compositions, UO₂ (with no dissolved gadolinium) and (U, Gd)O₂. The proposed heating cycle provided pellets free from Gd₂O₃ phase and may be used by the nuclear fuel industry as a suitable sintering process.     

Temperature dependence of liquid Sn sputtering by low-energy He and D bombardment

Absolute sputtering yields of liquid tin from 240 to 420 °C due to irradiation by low-energy helium and deuterium have been measured. For ion energies ranging from 300 to 1000 eV, temperature enhancement of liquid tin sputtering was noted. These measurements were obtained by IIAX (the Ion-surface InterAction eXperiment) using a velocity-filtered ion beam at 45° incidence to sputter material from a liquid tin target onto deposition monitors. Sputtering yields from 500 eV ion bombardment at 45° incidence increase from 0.1 ± 0.03 and 0.019 ± 0.008 Sn particles/ion at room temperature, for He⁺ and D⁺ ions respectively, to 0.30 ± 0.12 and 0.125 ± 0.05 Sn particles/ion for 380 °C. Temperature enhanced sputtering has been seen in other liquid metals (namely lithium, tin-lithium, and gallium) using both ion beam and plasma irradiation.     

Behavior of mechanical properties of nickel-alloyed reactor pressure vessel steel under neutron irradiation and post-irradiation annealing

The effect of neutron irradiation and post-irradiation thermal annealing on tensile and impact properties of Cr–Ni–Mo steel used for WWER-1000 reactor pressure vessel (RPV) manufacturing was studied. A gap in yield stress and ultimate tensile stress fluence dependence at the fluence range of 0–3×10²³ neutrons m⁻² was observed while ductile-to-brittle transition temperature (DBTT) was continuously increasing with damage dose. The post-irradiation annealing recovery of tensile properties was found to be higher than the one of impact properties. Over-recovery of tensile properties due to 460 and 490°C post-irradiation annealings were observed. The annealing effectiveness of WWER-440 and WWER-1000 grades was compared. Nickel was supposed to affect both the radiation sensitivity and the post-irradiation residual DBTT shift of WWER-1000 type steel.     

Computational-Experimental Investigation of the Wigner Energy and the Physical Properties of Plunger Graphite in the Safety and Control System of an RBMK Reactor

The total amount of stored energy (Wigner energy) and the physical-mechanical properties of the graphite plunger, which exhausted its total service life (6 yr), in the No. 2 unit of the Kursk nuclear power plant were estimated experimentally. The results showed that the total accumulated energy was 180–220 cal/g ((8–10)·10⁵ J/ kg). The real tempearture of the graphite plunger was found to be much lower – 70–80°C compared with the computed value 180–200°C. The energy was nonuniformly distributed over the cross section and azimuth of the plunger.Measurements showed that the thermal conductivity of the graphite in the plunger is low (no greater than 14–15 W/(m·K) at the measurement temperature 70°C) and that the temperature dependence is clearly nonmonotonic and contains stages with accelerated variation followed by moderation. These stages of nonmonotonic behavior correlate with stages where energy is released in experiments with linear heating of the irradiated graphite samples.     

Formation of porosity in uranium under the action of thermal cycling

In accelerated thermal cycling with a cycle of 50 sec period, considerable changes appear in uranium after 50–1000 cycles, depending upon the temperature range of the cycle. Cycling in the temperature range of the -phase (with heating up to between 550 and 600 °C) produces in texturcd uranium (containing about 0.1% carbon) a directional deformation and porosity, accompanied by a drop in density. After 5000 cycles, the drop in density amounts to 8% Thermal cycling with = ß = -transformations produces a pronounced distortion of the original shape of uranium speclmens and intense porosity formation, with a considerable drop in density, which attains 30% after 1000 cycles.