Analysis of relative release rates of radionuclides from the RBMK-1500 reactor fuel elements

Specific activities (concentrations) of fission products (FP) and activation products in spent fuel elements of the RBMK-1500 reactor were calculated using SCALE 5 computer code. Different burnup (5.1–21.0 MWd/kg) fuel assemblies were experimentally investigated. Activities of radionuclides present in the coolant water of storage cases of defective fuel elements were experimentally measured and analyzed. Experimental results provide a basis for a quantitative analysis of radionuclide release from spent fuel of the RBMK-1500 reactor. Relative release rates of radionuclides from the fuel matrix were assessed based on a comparison of experimental results with theoretical calculations. On the basis of analysis results released fission and activation products can be divided into several groups according to their release rates from fuel; this can be generalized for radionuclides with similar chemical properties.     

Radiation following the ejection of radionuclides into the atmosphere

Translated from Atomnaya Énergiya, Vol. 69, No. 3, pp. 161–164, September, 1990.     

Complex radiation monitoring of the fuel distribution in vibration-packed fuel elements

Conclusions The monitoring procedure, based on the combined use of a number of radiation methods, gives the possibility of obtaining complete information about the distribution of fuel materials in a fuel element. It is expedient to use the -absorption method in conjuction with passive gamma-scanning and computer tomography.Parametric modeling of the monitoring process on a computer allows the characteristics of the monitoring apparatus to be optimized, and its errors to be analyzed.The MNMG-1M apparatus can be used for monitoring the distribution of vibration-packed fuel in fuel-element rods at both the stage of development and research and also in the conditions of their commerical production.Translated from Atomnaya Énergiya, Vol. 59, No. 1, pp. 22–27, July, 1985.     

Growth of the interaction layer around fuel particles in dispersion fuel

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl_x. The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel.     

Ambient dose equivalent conversion coefficients for radionuclides exponentially distributed in the ground

Conversion coefficients of radionuclide deposition density to the ambient dose equivalent rate at 1 m height above ground were calculated for exponentially distributed sources in the ground. First, Monte Carlo transport simulations assuming exponential distributions in the ground were performed to obtain ambient dose equivalent for mono-energetic gamma-ray sources having different relaxation depths; next, on the basis of the simulated data, conversion coefficients for radionuclides were composed considering recent nuclear decay data. The ambient dose equivalent rates were then compared to the effective dose rates for reference adults and a new-born baby as well as to air kerma rates quoted from previous studies. It was confirmed that the ambient dose equivalent sufficiently overestimates effective doses, independently of age, for sources exponentially distributed in the ground. Furthermore, the air kerma was found to also overestimate the effective doses for all ages in the same conditions. In order to verify the computed conversion coefficients, the ratio of ambient dose equivalent to air kerma obtained by simulation was compared to the ratios measured at hundreds of locations in Japan which have been contaminated with radioactive cesium after the accident at the nuclear power plant in Fukushima Prefecture, Japan, in 2011; a good agreement was observed.     

Assessment of the models of urania fuel oxidation in steam-rich atmosphere

The urania fuel oxidation model for a steam-rich atmosphere developed by Cox et al. using the extensive experimental database has been subsequently used widely but with some inconsistencies in implementations. They are listed and evaluated in this work to help improve the existing models as well as for future model development. The comparison of the equilibrium stoichiometry deviation, calculated using various models, is also given. Small differences between the equilibrium constants used for steam dissociation are found. In one application, the original model for steam dissociation was rewritten in terms of component partial pressures for simplification. It is shown that the modified model differs from the original model by a multiplicative factor (1 ? p _O2 /P _tot). A more complex equation that preserves total pressure is derived. However, the effect on the calculated fuel oxidation is much smaller than the effect of the uncertainty of the equilibrium stoichiometry deviation. The discrepancies in stoichiometry deviations and the estimate of the surface area of a pellet may lead to a double underprediction of fuel oxidation rate.     

Analysis of critical heat flux during subcooled boiling for finned fuel elements

Experimental and analytical studies were performed to determine the critical heat flux (CHF) during subcooled boiling on finned fuel elements. Tests were conducted in a vertical, concentric-annulus test section consisting of a glass tube containing a finned heater element with either six, eight, or ten longitudinal fins. The phenomena leading to CHF are described and the parametric trends are discussed.A two-dimensional finite-element heat transfer model using the Galerkin method was used to analyse the experimental data to obtain CHF values. A dimensionless correlation was derived to predict the CHF values during subcooled boiling. Over 90% of the predicted CHF values agreed with those obtained from the two-dimensional analysis within ±30%.     

Radiostability of fuel plates used in SM-2 reactor fuel elements

Conclusions The data cited imply a high structural and dimensional stability on the part of plate type fuel elements used in the SM-2 reactor. The linearity of the volume increase with burnup allows us to assess changes in the physical, hydraulic, and heat-transfer parameters of the core. We note that a process of layer separation resulting in a loss of pressuretightness sets-in in some fuel elements with 30% uranium burnup, and this could be involved in failures during cooling or cooldown. Checking out this last point would enable us to settle on a criterion for the limiting allowable burnup. It is highly probable that this criterion will be related to the physics of the reactor, rather than to the degree of radiation damage of the reactor materials.Translated from Atomnaya Énergiya, Vol. 24, No. 5, pp. 432–435, May, 1968.     

The effect of precipitation on the long-term release of radionuclides from used fuel

Solubility-limited dissolution models can be used to determine source terms for the safety assessment of used nuclear fuel disposal in geological vaults. These models usually assume that the solubility of the used-fuel matrix is constant. However, spatial inhomogeneities can lead to solubility gradients and, hence, cause precipitation downstream from a dissolving used-fuel matrix. In this paper, the effect of precipitation on the used-fuel dissolution rate has been investigated. Precipitation of uranium-containing solids has been found to enhance the dissolution rate of used fuel, especially at long times. The enhancement factor has been calculated for various vault conditions.     

Sorption of short-lived radionuclides in a layer of sorbent with spherical granules

"Radon" Moscow Scientific-Production Organization. Translated from Atomnaya Énergiya, Vol. 74, No. 3, pp. 210–214, March, 1993.     

Effect of fuel burnup on the effectiveness of ultrasonic detection of leaky fuel elements in VVER-1000 fuel assemblies

The effect of fuel burnup in VVER-1000 fuel elements on the utilization effectiveness of ultrasonic detection of leaky fuel elements is examined. It is determined that the limitations of this method are due to the interaction of fuel-element cladding with the fuel pellets. Threshold for fuel burnup in VVER-1000 fuel elements with E-110 alloy cladding, determining the application limits of ultrasonic detection of leaky fuel elements in fuel assemblies, is determined.