Uranium monocarbide

Uranium monocarbide is of interest as a possible nuclear fuel and material for nuclear thermoelectrical transformations. In order to accurately define the effect of the conditions for preparing uranium monocarbide (UC) on its composition, studies were made which established the optimum regime for preparing UC with a stoichiometric composition.Studies were made of the conditions for sintering and hot pressing of UC powder and also conditions for sintering UC + U alloys, giving specimens with a porosity of about 5%. The specific gravity of UC powder determined by a pycnometer is 12.97 ± 0.09 g/cm³, microhardness of the phase-923 ± 56 kg/mm².The thermal conductivity of UC in the range 100–700 °C varies from 0.028 to 0.04 cal/cm. sec.deg; the mean thermal coefficient of linear expansion in the range 20–1500 ° is 11.6 · 10^–6. Specimens of UC subjected to cyclic heat-treatment in the range 200–1000 ° withstood 500 cycles without failure. Specimens of UC + U withstood more than 1000 cycles.     

Lithium nitrate as a fusion reactor coolant fluid?: A thermochemical assessment

In connection with possible application as a coolant fluid in fusion reactors, molten LiNO₃ and LiNO₃/LiNO₂ mixtures, and their mixtures with Na/K nitrates, have been evaluated with respect to their high temperature stability, their ability to reversibly dissolve and release tritium and, in a very limited sense, their corrosiveness toward structural alloys. The results of the primarily thermochemical evaluation indicate that with respect to thermal/radiolytic decomposition and corrosivity LiNO₃/LiNO₂ may be suitable for application up to R 700^°K; however, with respect to reversibility/irreversibility of tritium release, it appears unsuitable at all likely operating temperatures (R 675–800°K).     

Hydrogen and helium trapping in tungsten under single and sequential irradiations

The trapping of hydrogen and helium in polycrystalline tungsten irradiated with 500 eV He⁺, H⁺ and D⁺ ions, individually or sequentially, has been measured by thermal desorption spectroscopy. Specimens irradiated with 500 eV He⁺ at 300 K show three He release peaks in the vicinity of ∼500, ∼1000, and ∼1200 K. The helium is thought to form He vacancy complexes or bubbles. Increasing the specimen temperature to 700 K does not significantly affect the trapping behavior of He. Sequential He⁺-D⁺ irradiation at 300 K results in the elimination of He release above 800 K. Instead, both D and He were released in the range 400-800 K. This is interpreted as interstitial D and He released from the near surface. Sequential He⁺-D⁺ irradiation at 700 K resulted in a reduced single He peak at ∼1000 K with very little release observed below 800 K; no D was trapped for irradiations at 700 K. Sequential D⁺-He⁺ irradiations at 300 K show that He trapping occurs in much the same manner as for the He⁺-only case while D retention is reduced at the near surface. Sequential D⁺-He⁺ irradiations at 700 K indicate that pre-irradiation with D⁺ has little or no effect on the subsequent trapping behavior of He.     

Lifetime of paramagnetic E1 centers in quartz rocks

Conclusions In quartz rocks E₁ centers are formed in both paramagnetic and nonparamagnetic states under the effect of the natural background of ionizing radiation.Nonparamagnetic E₁ centers go over into the paramagnetic state under the effect of heat on quartz (thermal activation of E₁ centers).The rate of thermal activation depends on the absolute temperature of the specimen.Two subtypes of E₁ centers differing as to the rate of thermal activation were detected.By quasilinear extrapolation in the coordinates (C_E–C₀)/(C_M–C₀)-t^1/2 it was found that thermal activation of E₁ centers of the first type, determined from the initial segments of the process, whould be completed at the usual temperature of 293°K (20°C) in 0.8·10⁸ yr.Annealing of E₁ centers of the first subtype in natural occurrence (at 293°K) is completed in 6.4·10⁸ yr. The time for the completion of the thermal activation and annealing of E₁ centers of the second subtype is roughly an order of magnitude longer than in the case of the first subtype.As a result of a long experiment (3 yr) it was established that at 423°K (150°C) the method of quasilinear extrapolation inables the thermal activation time to be estimated more accurately than does the barrier-activation theory usually employed.Translated from Atomnaya Énergiya, Vol. 49, No. 1, pp. 31–33, July, 1980.     

Measurement of the Characteristics of the Ventilation Pipe of the Filtration System for the Steam-Gas Medium of a Nuclear Power Plant

The dependence of the heat power of a pipe (air–air heat exchanger) on the flow rate and the temperature of the heating air is investigated experimentally. The temperature fields along the height of the pipe are measured. In the experimental velocity range 1.2–9.5 m/sec and heating-air temperature range 120–280°C, the heat power depends strongly on the temperature and negligibly on the flow rate of the heating air. Increasing the cold-air flow rate increases the thermal power in direct proportion.     

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.     

Influence of both reactor irradiation and helium isotopes upon the mechanical properties of titanium and α-titanium alloys

Conclusions Reactor irradiation of titanium and its -phase alloys containing³He and irradiation with⁴He ions leads to increased strength and reduced plasticity in the temperature range 20–400°C. The influence of the two types of irradiation upon these materials is similar. At 20–600°C the irradiated materials conserve an adequate plasticity reserve (>6%). Viscous destruction of the initial and irradiated samples of the materials examined takes place in mechanical testing. At the temperature (500–600°C) at which recrystallization of nonirradiated materials begins, total recovery of the strength properties and noticeable recovery of the plasticity are observed in the irradiated samples. Titanium alloys are inclined to high-temperature radiation-induced embrittlement to a lesser degree than austenitic stainless steels.Translated from Atomnaya Énergiya, Vol. 53, No. 4, pp. 237–242, October, 1982.     

Dynamic Monte-Carlo modeling of thermal desorption of H2 from H irradiated graphite

Thermal desorption of hydrogen molecules from H⁺ irradiated graphite is studied using dynamic Monte Carlo simulation. The purpose of this study is to understand the experimentally observed phenomena that the thermal desorption of H₂ from the graphite exhibits sometimes single desorption peak, sometimes double peaks, and even three desorption peaks under certain circumstances. The study result reveals that the fluence of pre-implanted H⁺, the concentration of trap sites, porosity, and mean crystallite volume are important parameters in determining the number of desorption peaks. It is found that low implantation fluence and high concentration of trap sites easily lead to the occurrence of single desorption peak at around 1000 K, and high implantation fluence and low concentration of trap sites favor the occurrence of double desorption peaks, with a new desorption peak at around 820 K. It is also found that small porosity of graphite and large crystallite volume benefit the occurrence of single desorption peak at around 1000 K while large porosity of graphite and small crystallite volume facilitate the occurrence of double desorption peaks, respectively, at around 820 and 1000 K. In addition, experimentally observed third desorption peak at lower temperature is reproduced by simulation with assuming the graphite containing a small concentration of solute hydrogen atoms.     

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.     

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.     

Uranium mononitride: Heat capacity and thermal conductivity from 298 to 1000 °K

The heat capacity and the thermal diffusivity of uranium mononitride were measured by a laser flash method at temperatures ranging from 298 to 1000 °K. The samples were arc-melted UN having nearly zero porosity and sintered UN having porosity of 10.1%. The heat capacity of UN was represented by C_p = 12.08 + 2.548 × 10 ^{− 3}T − {−1.252 × 10⁵T⁻² cal/mol · deg K (298–1000 °K)}. From the heat capacity data, entropy, enthalpy and the Gibbs energy function of UN were calculated. The thermal conductivities of arc-melted UN, calculated from the heat capacity and the thermal diffusivity data, at 350 and 1000 °K were 0.031 and 0.045 cal/ cm · sec · deg K, respectively. The results agreed reasonably well with those of Moore et al. obtained at lower temperatures.     

Design and experiment of hot gas duct for the HTR-10

A horizontal coaxial double-tube hot gas duct is a key component connecting the reactor pressure vessel and the steam generator pressure vessel for the 10 MW High Temperature Gas-cooled Reactor—Test Module. Hot helium gas from the core outlet flows into the steam generator through the liner tube, while helium gas after being cooled returns to the core through a passage formed between the inner tube and the duct pressure vessel. Thermal insulation material is packed into the space between the liner tube and the inner tube to resist heat transfer from the hot helium to the cold helium. The thermal compensation structure is designed in order to avoid large thermal stress because of different thermal expansions of the duct parts under various conditions. According to the design principal of the hot gas duct, the detailed structure design and strength evaluation for it has been done. A full-scale duct test section was then made according to the design parameters, and its thermal performance experiment was carried out in a helium test loop. With helium gas at pressure of about 3.0 MPa and a temperature over 900 °C, the continuous operation time for the duct test section lasted 98 h. At a helium gas temperature over 700 °C, the cumulative operation time for the duct test section reached 350 h. The duct test section also experienced 20 pressure cycles in the pressure range of 0.1–3.4 MPa, 18 temperature cycles in the temperature range of 100–950 °C. Thermal test results show an effective thermal conductivity of the hot gas duct thermal insulation is 0.47 W m⁻¹ °C⁻¹ under normal operation condition. In addition, a hot gas duct depressurization test was carried out; the test result showed that the pressure variation occurred on the liner tube was not more than 0.2 MPa for an assumed maximum gas release rate.     

Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U

Release behavior of tritium from the graphite tiles used at dome top and inner dome wing in JT-60U was investigated by the thermal desorption method in dry argon, argon with oxygen and water vapor, or argon with hydrogen. It was found that approximately 20-40% of total tritium is left in graphite even after heating to the high temperature above 1000 °C in dry argon. The residual tritium could be removed by exposing the graphite tile to oxygen with water vapor or hydrogen at the high temperature above 1000 °C. The tritium retention of the dome top tile was quantified as 84-30 kBq/cm². The inner dome wing tile had a steep tritium distribution from 8 to 0.1 kBq/cm². It is observed that a measurable amount of tritium existed in the deep site of the graphite tile.     

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.     

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.     

Penetration of hydrogen isotopes through the superalloy ÉI698 at high pressures and temperatures

Conclusions We determined the parameters of the processes of penetration and diffusion of the hydrogen isotopes through the superalloy ÉI698 in the 600–1050°K range at pressures up to 100 MPa. It was shown that the penetrating flux is in proportional to p in the 0.1–100 MPa pressure range.The diffusion coefficient of hydrogen (deuterium) does not depend on the concentration of the dissolved hydrogen (deuterium) that attains a value of 1.3·10⁶ cm³/cm³ Me at a pressure of 100 MPa and a temperature of 1050°K. The concentration of the dissolved hydrogen is not critical with respect to hydrogen-embrittlement of the ÉI698 alloy subjected to loading for a period of 600 h under an internal pressure of 100 MPa. Based on the obtained values of the penetration rates of hydrogen and deuterium through the ÉI698 alloy, the decrease in the quantity of the gas in the vessel can be taken into account properly when carrying out experimental studies on the processes of mu-catalysis.Translated from Atomnaya Énergiya, Vol. 65, No. 6, pp. 395–399, December, 1988.     

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).     

Thermo-physical properties and transient heat transfer of concrete at elevated temperatures

The objective of this study is to produce our own experimental data of physical properties of domestic concrete used in Korean NPPs, and to study on the thermal behavior of concrete exposed to high temperature conditions. The compressive strength and chemical composition of the concrete used in the Yonggwang NPP units 3 and 4 were analyzed. The chemical composition of Korean concrete is similar to that of US basaltic concrete. The thermal properties of the concrete, such as density, conductivity, diffusivity, and specific heat were also measured with a wide temperature range of 20–1100 °C. Most thermo-physical properties of concrete decrease with an increase in temperature except for the specific heat, and particularly the conductivity and the diffusivity are a 50% lower at 900 °C as compared with the values at room temperature. The specific heat increases until 500 °C, decreases from 700 to 900 °C, and then increases again when temperature is above 900 °C. In this work, we also have performed CORCON analysis and MCCI experiments to simulate a transient thermal behavior of concrete exposed to high temperature conditions. The measured maximum downward heat flux to the concrete specimen was estimated to be about 2.1 MW m⁻² and the maximum erosion rate of the concrete to be 175 cm h⁻¹ with maximum erosion depth of about 2 cm. In the CORCON analysis, it is found that the concrete compositions have an important effect upon concrete erosion.     

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.     

Separation of boron isotopes using isotopic exchange between BF3 and a complex of BF3 with phenetole

The thermal dissociation of the complex of BF₃ with phenetole in the temperature region 172–182° C was studied. It was shown that at water concentrations no greater than 0.01–0.02%, the completeness of thermal dissocation is sufficient to obtain highly concentrated B¹⁰. The influence of additions of phenol, which also forms a complex with BF₃, upon thermal dissociation of the complex was studied. Experiments on the separation of boron isotopes were conducted on a laboratory column 0.94 m high and 16 mm in diameter. The values of BETT for 15, 25, and 35° C were calculated according to the degree of dissociation in the standard state. The results of the experiments on a laboratory column confirmed the data on thermal dissociation, obtained on a model apparatus.Translated from Atomnaya Énergiya, Vol.22, No.4, pp.297–302, April. 1967.