Interaction of Metal Melt with Zirconium Dioxide Refractories in an External Trap

The results of an investigation of the interaction of metal melt of reactor materials with zirconium dioxide refractories (ceramic, concrete) in an atmosphere with various oxygen contents at temperatures up to 2600 K are presented. The experiments showed that the melt Zr + 2.5% Nb at 2273 K does not give rise to erosion of the ceramic; at 2500–2600 K, it permeates and dissolves the ceramic. Erosion of concrete starts at the melting point of the alloy. A melt of reactor steel does not interact with refractories. Under certain conditions, capillary permeation of steel into refractories occurs. At temperatures above 2500 K, a steel–20% Zr melt gives rise to erosion of zirconium dioxide concrete.The results are used to analyze the operation of a refractory protective layer of an EPR trap.     

Interaction of melt components with zirconium dioxide ceramic

Conclusions The interaction rate of zirconium dioxide ceramic at ∼2300 K depends on the composition of the gas environment (partial pressure of oxygen), the ratio of the oxide and metal phases of the melt (especially iron oxides and zirconium), and the porosity. The temperature distribution affects the region where the zirconium dioxide interacts with the oxides. The region of the ceramic penetrated by the oxides of nickel, titanium, and chromium is larger under isothermal conditions than for a temperature gradient of 100 deg/mm. These research results can be used to estimate the mass of the ceramic trap for a major accident in air. In this case erosion of the ceramic is determined by the iron oxide. The experimental data also show that the mass of a ceramic trap should be at least five times larger than the mass of iron oxide. This relationship is independent of time, but it can be changed due to the superposition of effects from the interaction of all the melt materials with the ceramic. Ceramic samples 20 mm diameter and 10 mm thick did not break down in a temperature gradient of 100 deg/mm at a heating rate of 200 deg/min with natural cooling. Because the erosion of the two types of zirconium dioxide ceramic by metal oxide melts was the same, it may be possible to use cheaper ceramics with calcium oxide stabilizers. Materials Research Center, Technology Novosiberian Department, United Institute of High Temperatures of the Academy of Sciences (NITs TIV NO IVTAN). Translated from Atomnaya énergiya, Vol. 81, No. 6, pp. 468–471, December, 1996.     

Interaction of the metallic components of melt in a reactor core with zirconium dioxide ceramic

Conclusions The results permit drawing the following conclusions: the penetration depth of the melt into the ceramic during the experiment was equal to approximately 0.12–0.2 mm in the case of a mixed melt of the steel and zirconium and approximately 0.35–0.4 mm in the case of the pure-zirconium melt; the crucible ceramic does not undergo erosion under the action of both types of melts; in the case of an interaction with the zirconium melt, a zone of softening forms in the ceramic on account of the reduction of zirconium dioxide to ZrO_0.35; zirconium and to a lesser degree iron in the melt are partially oxidized, primarily on account of diffusion transfer of oxygen from the ceramic into the melt, in the process of the interaction with the zirconium-dioxide based ceramic in an inert medium; very little iron is transferred from the melt into the ceramic; under the conditions of an inert medium the zirconium is the main corroding component of the melt; for a comparatively low content of zirconium in the melt steel + zirconium the zirconium dioxide based ceramic is quite highly resistant to the action of the melt; and, the zirconium dioxide based ceramic is highly resistant to the action of the melt under these conditions as compared with the refractory zirconium dioxide concrete [1, 2] and construction concrete [8]. Scientific-Research Center TIV Industrial Association "Institute of High Temperature," Russian Academy of Sciences. Translated from Atomnaya énergiya, Vol. 79, No. 6, pp. 454–458, December, 1995.     

Behavior of zirconium-dioxide ceramic under the operating conditions of an external trap

The results of an experimental investigation of the penetration and interaction of a melt of iron oxides, which are one of the components with significant mass and corrosive power, with zirconium dioxide ceramic with about 20% porosity are reported. It is concluded on the basis of the data obtained that such a ceramic has functional possibilities for use in the external trap of a nuclear power plant. 2 figures, 8 references. NITs TIV OIVT of the Russian Academy of Sciences. Translated from Atomnaya énergiya, Vol. 87, No. 1, pp. 48–53. July, 1999.     

Effect of Reactor Irradiation on the Electric Parameters of Alumina-Rich Ceramics

Samples of commercial electric-grade ceramic before and after exposure to rays (21 Gy·sec) and a thermal neutron flux (3·10¹⁶ sec^–1·m^–2) at temperatures 300–900 K in the frequency range 0.3–30 kHz are investigated. The changes in electric conductivity and dielectric losses are discussed on the basis of a model of a disordered dielectric with electrically charged, locally uncompensated, structural defects. Correlations are made between the number of structural defects and the total internal surface area.     

Apparatus for fine purification of inert gases from oxygen and water vapor

The presence of oxides in apparatuses using sodium and alloys of sodium with potassium strongly increases corrosion of the loop, and, in individual cases, may cause blockage of the passage. In such apparatuses it is necessary to have over the surface of the metal, an atmosphere of inert gases containing the minimal amounts of oxygen and water vapor, to protect the metal from oxidation. The inert gas is used also as a separating medium between the metal and the sensitive elements of the apparatus in measuring the pressure.Argon, nitrogen, and helium may be used as shielding gases. Argon is the most convenient of them for experimental apparatuses due to its high specific gravity in comparison with air. This makes it possible to open separate parts of the apparatus for a short time (oxidation of the metal does not occur then). For industrial apparatuses, it is advantageous to use nitrogen. It is apparently not advantageous to use helium due to its fluidity, low specific gravity, and high cost. These inert gases contain a definite amount of oxygen and water vapor and may themselves be a source of contamination of the metal with oxides. Thus, pure argon contains 0.05–0.1 vol. % of oxygen according to specification data and technical nitrogen contains 0.5–1 vol. %. A special apparatus was constructed for purifying these gases to remove oxygen and water vapor.     

The effect of radiation on the properties of silicone rubber

As a result of -radiation, silicone rubber undergoes a series of changes due to radiation cross-linking. The modulus of elasticity increases linearily with a dose up to 150–200 megarads. The vitrification temperature (– 120 to- 125 ° C) hardly changes up to 100 megarads and at 270 megarads it is –110 to –115 ° C. With cross-linking, the rate of crystallization and degree of crystallinity decrease. The melting point falls from – 35 ° C for the initial material to – 55 ° C for rubber Irradiated with a dose of 40 megarads. A dose of 100 megarads practically completely eliminates crystallization. This dose produces rigid rubbers with a modulus of 200– 250 kg/cm² and a high degree of frost resistance, (Tg –125 ° C), but with a very small breaking stretch (15–20%) with a strength of 30–40 kg/cm².     

Reactivity Behavior During RBMK Startup

During startup of an RBMK reactor, the reactivity varies from –(4–7)_eff to 0–0.1_eff. Positive reactivity is introduced locally – by extracting control rods. Since the physical dimensions of an RBMK reactor are large, a local change in the properties can produce a large change in the spatial distribution of the neutron flux in the core. The possible range of variation of the reactivity of a subcritical and a critical reactor with one control rod extracted is analyzed for the actual states of the power-generating units of a nuclear power plant with RBMK reactors. It is shown that the extraction of some rods in an RBMK reactor in subcritical and critical states can increase the reactivity by 1_eff or more.     

The high-density Z-pinch as a pulsed fusion neutron source for fusion nuclear technology and materials testing

The dense Z-pinch (DZP) is one of the earliest and simplest plasma heating and confinement schemes. Recent experimental advances based on plasma initiation from hair-like (10s m in radius) solid hydrogen filaments have so far not encountered the usually devastating MHD instabilities that plagued early DZP experimenters. These encouraging results along with the debut of a number of proof-of principle, high-current (1–2 MA in 10–100 ns) experiments have prompted consideration of the DZP as a pulsed source of DT fusion neutrons of sufficient strength (S _N 10¹⁹ n/s) to provide uncollided neutron fluxes in excess ofI _w = 5–10 MW/m² over test volumes of 10–30 liters or greater. While this neutron source would be pulsed (100s ns pulse widths, 10–100 Hz pulse rate), giving flux time compressions in the range 10⁵–10⁶, its simplicity, near-term feasibility, low cost, high-Q operation, and relevance to fusion systems thatmay provide a pulsed commercial end-product, e.g., inertial confinement or the DZP itself, together create the impetus for preliminary consideration as a neutron source for fusion nuclear technology and materials testings. The results of a preliminary parametric systems study (focusing primarily on physics issues), conceptual design, and cost vs. performance analyses are presented. The DZP promises an inexpensive and efficient means to provide pulsed DT neutrons at an average rate in excess of 10¹⁹ n/s, with neutron currents I_w10 MW/m² over volumes V_exp 30 liter using single-pulse technologies that differ little from those being used in present-day experiments.Work supported by U.S. DOE.     

A pulsed neutron generator

The authors give the results of physical investigations on a neutron generator with spark-ion source, yielding pulsed-neutron fluxes from the reactions D+ D and D+ T, with mean yields ~7·10⁶ and ~10⁹ per pulse respectively. The total pulse length is ~100–250sec and the potential across the accelerator gap is ~110 kV.Translated from Atomnaya Énergiya, Vol. 18, No. 4, pp. 336–342, April, 1965     

Thermochemical Stability, Radiation Testing, Fabrication, and Reprocessing of Mononitride Fuel

A thermodynamic analysis and experimental investigations have shown that mononitride fuel is thermochemically stable up to 1973–2073 K, at which temperature the equilibrium vapor pressure of nitrogen does not exceed 4.5·10^–7–2.1·10^–6 MPa. It is concluded on the basis of a generalization of the data from radiation testing of mononitride fuel with burnup up to 9–10% in fast and 16.8% in thermal reactors with lineal power density from 400 to 1300 W/cm that it should operate reliably in fuel elements with helium and liquid-metal sublayers. The requirement for the impurity (oxygen and carbon) content in it is formulated. When both oxygen and carbon impurities are present simultaneously in mononitride, the mass fraction of each should not exceed 0.15%. The methods for fabricating mononitride fuel are determined by the final product of the reprocessing of irradiated fuel. Consequently, methods for fabricating mixed nitride fuel from oxides and metals are now being developed.     

Method for measuring the radiation-induced electrical conductivity in ferroelectrics with pulsed irradiation

Conclusions Under the assumptions that the volume and surface RIE are power-law functions of the dose rate and Ohm's law is valid for the FE with irradiation, and equation was derived for the leakage current owing to RIE and the electric field generated by the pyroelectric effect in the FE. In this case the parameters characterizing the RIE can, be found by solving a system of four equations, which are constructed based on experimental results for samples of material with different dimensions. In practice the number of unknown can be reduced by using specially constructed samples of the materials (this work) or radiation collimators, which enables irradiation of the required region of the sample.The experimental results obtained with PKR-7M piezoelectric ceramic showed that for dose rates (0.25–5.2)·10⁷ Gy·sec^–1 [the corresponding dose range is (0.19–3.6)×10³ Gy] both the volume and surface RIE can be represented by a linear function of the dose rate with proportionality coefficients k_v=1.2·10^{–11 –1}·m^–1·Gy^–1·sec and k_s=0.53·10^{–11 –1}·Gy^–1·sec.The upper limit of the residual RIE of PKR-7M with a maximum dose rate per pulse of 5.2·10⁷ Gy·sec^–1 equals =10^{10–6 –1}·m^–1.Translated from Atomnaya Énergiya, Vol. 63, No. 4, pp. 261–264, October, 1987.     

AgInCd control rod failure in the QUENCH-13 bundle test

The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO₂ pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI–Switzerland and AEKI–Hungary.Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g of H₂ during the pre-reflood phases). Posttest examinations of bundle structures revealed the presence of only little relocated AgInCd melt in the form of rivulets, mainly in the coolant channels surrounding the control rod simulator and at axial elevations between the third (0.55 m) and first spacer grids (−0.1 m).Results of QUENCH-13 on the onset of absorber rod failure are in agreement with CORA results of nine experiments each containing one or more AgInCd/stainless steel/Zircaloy-4 control rod assemblies. Bundle degradation triggered by early melt formation was, however, more pronounced in the CORA experiments with maximum bundle temperatures of 2300 K (compared to 1800 K in QUENCH-13). Consequently, QUENCH-13 allowed studying the initiation of absorber rod failure by eutectic reactions of SS-Zr, and later on of AgInCd-Zr, as well as the redistribution of the absorber material within the test bundle. Furthermore, input data for modeling of aerosol release during severe accidents are considered as benefits of the experiment.     

Study of the deactivation mechanisms for some constructional steels by secondary-ion mass spectrometry

Conclusions Dense corrosion layers on steels 48TS, 2Kh13, and 12Kh18N10T obtained under conditions simulating the first loop in a boron-regulated reactor are depleted in the alloying elements at the surface. Treatment with boric permanganate (pH7.4) leads to selective dissolution of the iron and partial dissolution of the iron-rich surface layer of the model oxide. Triloncitrate recipes dissolve the oxide layers efficiently when these are enriched in iron but are ineffective in dissolving chromium-bearing oxide layers.Because of these features, the performance in removing model oxide by treatment with these transformable deactivating recipes decreases in the series 48TS, 2Kh13, and 12Kh18N10T. Therefore, a basic problem in devising effective method of deactivating constructions containing low-alloy, stainless, and chromium steels is the prevision in the deactivating solution of conditions favoring relatively uniform dissolution of the oxides of iron (nickel) and the oxides of chromium. None of the recipes considered above satisfies this requirement. Further research is in hand on the possibility of realizing these conditions.Translated from Atomnaya Énergiya, Vol. 53, No. 3, pp. 171–174, September, 1982.     

γ-Ray spectra excited in inelastic scattering of fast neutrons on manganese,aluminum, iron,copper, tin,and antimony

The study of inelastic scattering of fast neutrons is an important problem of both theoretical and practical interest. From the theoretical point of view the importance of this work lies in the possibility of obtaining data concerning levels in stable nuclei. The practical value arises in connection with the important role played by inelastic scattering of neutrons in fast-neutron reactors as well as the fact that the extension of reactor theory to fast-neutron reactors requires data on the spectra of inelastically scattered neutrons [1, 2]. In this connection the necessity for developing a neutron spectrometer for fast neutrons and-spectroscopy for inelastic neutron scattering is obvious. In the last 5–7 years a great deal of work has been devoted to this problem.The present work reports on measurements of-ray spectra excited in inelastic scattering of 2.8 Mev neutrons by manganese, aluminum, iron, copper, tin and antimony. The measurements were carried out with a scintillation spectrometer consisting of an NaI(Tl) crystal, a FEU-1B photomultiplier and a 50-channel pulse-height analyzer with a magnetic-drum memory. The spectrometer resolution was 6.5–7% for-rays from Co⁶⁰.-Rays of the following energies (Mev) were found: in manganese 0.97, 1.41, 1.92, 2.3; in aluminum 0.84, 1.00, 1.80, 2.16; in iron 0.84, 1.25, 1.46, 1.70; in copper 0.63, 0.78. 0.96, 1.12, 1.38, 1.46, 1.72, 2.03; in tin 0.84, 1.16, 1.50, 1.80, 2.16; in antimony 1.04, 1.50, 1.84, 2.16.Abbreviated version of a paper appearing in the Ukrainian Journal of Physics.The authors wish to take this opportunity to thank L. M. Beliaev and G. F. Dorbrzkanskii of the Institute of Crystallography, Academy of Sciences, USSR for making the NaI(TI) crystal and for kndly allowing us to use it in carrying out the present work.     

Heats of formation of UFe2 and U6Fe

UFe₂ was obtained by heating stoichiometric amounts of uranium and iron. The heat of formation (–H_298°K ⁸) was found from the difference between the heat of dissolution of UFe₂ and that of a stolchiometric mixture of its components; taking into account the admixture correction, the value was 7.7 ± 0.3 kcal/mole. The heat of formation of U₆Fe was calculated, the value being 3.9 ± 1.2 kcal/mole.Translated from Atomnaya Énergiya, Vol. 13, No. 6, pp. 572–575, December, 1962     

Effects of working fluid and shaft rotation speed on the performance of HTR plants and the size of CBC turbo-machine

Investigated are the effects of the molecular weight of the working fluid, reactor exit temperature, and shaft rotation speed on the size and number of stages of the turbo-machine as well as the performance of high temperature reactor (HTR) plants with actively cooled reactor pressure vessel and direct or indirect Closed Brayton Cycles (CBCs). The present analyses for working fluids of helium (4 g/mol) and the 15 g/mol He–Xe and He–N₂ binary mixtures are performed for a reactor thermal power of 600 MW, shaft rotation speed of 3000–9000 rpm, and reactor exit temperature from 973 K to 1223 K. For the plants with indirect CBCs, the analyses assume a temperature pinch of 50 K in the IHX. Results show that the CBC compression ratio is relatively low (2.6 for He and He–Xe and 3.2 for He–N₂), increases very little with increasing the reactor exit temperature and near the maximum thermal efficiency of the plant. For the plants with a direct helium CBC, the thermal efficiency increases from 42% to 51% as the reactor exit temperature increases from 973 K to 1223 K, respectively, versus 37% to 47% for the plants with indirect He-CBC. The HTR plants with indirect He–Xe and He–N₂ CBCs and operating at a turbine inlet temperature of 1123 K have slightly higher thermal efficiencies (45.9% and 45.8%) than the He plant with indirect CBC (45.6%), generating 1.6 MWe more electrical power. The molecular weight of the working fluid has a very small effect on the plant thermal efficiency, but significantly reduces the size and number of stages of the CBC turbo-machine. Increasing the shaft rotation speed also decreases the size and number of stages of the CBC turbo-machine.     

Large-scale ECOKATS experiments: Spreading of oxide melt on ceramic and concrete surfaces

Large-scale ECOKATS experiments are performed to study spreading of an oxide melt on ceramic and concrete surfaces. The oxide melt generated by a thermite reaction was composed of 41 wt.% Al₂O₃, 24 wt.% FeO, 19 wt.% CaO and 16 wt.% SiO₂. This melt was selected as the most appropriate simulation of a corium melt because of its wide freezing range of approx. 450 K. Despite a rather low liquidus temperature, the attempt to measure melt viscosity failed. As spreading of high-temperature oxide melts is nearly isothermal during the early phase of motion, i.e., only thin thermal boundary layers will develop, the melt viscosity can be estimated from a two-dimensional spreading experiment, ECOKATS-V1, on a ceramic substrate by approximate self-similar solutions. To further study the influence of the gas release from the substrate caused by thermal erosion of the underlying concrete by a corium melt on spreading, a large amount of the oxide melt was released into a 2.6 m long and 0.29 m wide channel leading into a 3 m × 4 m rectangular surface. Spreading on a concrete substrate is influenced by the gas release from the decomposed concrete, which changes viscosity. A viscosity increase by a factor of 3.6 was estimated from spreading in the concrete channel.     

Absorption and outgassing of helium during He+ ion bombardment of niobium

Conclusions The capute and outgassin of helium during bombardment by He⁺ ions of energy up to 15 KeV of a niobium target witha temperature of 290–1800°K have been studdied. It has been shown that as the temperature during irradation, T_b, is increased the coefficient of gaseous emission, , increases, while the number of injected atoms, N₀, and the capture coefficient decrease. For temperature greater than 1500°K practically 100% of the bombarding atoms are released to the vacuum. As the dose is increased, and N₀ increase and falls. An increase in to unity indicates saturation of saturation of theniobium by helium. The irradiation dose at which this saturation sets in and its level decrease with the temperature T_b.As a result of an analysis of the outgassing spectra, it has been shown thatthe release of helium occurs in three stages, each of which is due to one of the following causes: diffusion of individual atoms dissolved in the lattice; relase of helium from gas bubbles located in the volume of the metal; release of gas due to bursting of the shells of surface blisters. The last process has a basic role in outgassing at large irradiation doses.Translated from Atomnaya Énergiya, Vol. 38, No. 3, pp. 152–155, March, 1975.     

Fusion Safety Studies in Russia in 1996

The paper seeks to provide a summary report of observations and results of some Russian fusion safety studies performed in 1996. Release of tritium and helium from neutron irradiated beryllium at relatively high neutron fluences has a burst nature. With the growth of the beryllium temperature-increase rate to 90 K/s, the temperature of tritium burst release decreases from 800 to 450–500°C and for helium decreases from 1200 to 500°C. Characterization of carbon and tungsten dust produced in experiments simulating plasma disruptions revealed that dust particle distribution of sizes for graphites and carbon fiber composites has a bimodal nature with maxima in the range of 0.01–0.03 and 2–4 m for composite UAM and in the range of 0.14–0.18 and 2–4 m for graphite MPG-8. Chemical reactivity of beryllium with air was studied as well. A mathematical model for beryllium weight gain under its chemical interaction with air at temperatures of 700–800°C as a function of beryllium porosity, temperature, and interaction duration was developed.