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.     

Modelling the activity of I in the primary coolant of a CANDU reactor

A mathematical treatment has been developed to describe the activity levels of ¹²⁹I as a function of time in the primary heat transport system during constant power operation and for a reactor shutdown situation. The model accounts for a release of fission-product iodine from defective fuel rods and tramp uranium contamination on in-core surfaces. The physical transport constants of the model are derived from a coolant activity analysis of the short-lived radioiodine species. An estimate of 3×10⁻⁹ has been determined for the coolant activity ratio of ¹²⁹I/¹³¹I in a CANDU Nuclear Generating Station (NGS), which is in reasonable agreement with that observed in the primary coolant and for plant test resin columns from pressurized and boiling water reactor plants. The model has been further applied to a CANDU NGS, by fitting it to the observed short-lived iodine and long-lived cesium data, to yield a coolant activity ratio of ∼2×10⁻⁸ for ¹²⁹I/¹³⁷Cs. This ratio can be used to estimate the levels of ¹²⁹I in reactor waste based on a measurement of the activity of ¹³⁷Cs.     

Non-dimensional analysis of the transient flow rate and the coolant temperature in a reactor coolant system under different pump failures

The mathematical models are developed to solve the non-dimensional transient flow rates in two loops and a reactor core under different power failures of reactor coolant pumps. Comparison of the experimental results of the Qinshan Nuclear Power Plant and the test results of the nuclear ship reactor in Japan under one pump power failure shows an excellent agreement. The non-dimensional flow rates are determined by the established non-dimensional parameters λ, ?, and α. Under the sequential power failure of two pumps, the non-dimensional flow rates are determined by the established λ, ?, α, and ΔT parameters. λ, ?, α, and ΔT are four important non-dimensional parameters in the prediction of flow transients. λ indicates the resistance coefficient ratio of the single loop to the reactor core, ? indicates the fluid inertia ratio of the reactor core to the single loop, α indicates the ratio of the initial kinetic energy of the single loop coolant fluid to the effective initial kinetic energy of the reactor coolant pump, and ΔT means the non-dimensional time interval of the sequential power failure of two pumps. The effects of λ, ?, α, and ΔT on the non-dimensional flow rates and the temperature change are investigated.     

An improved method for reactor coolant pump abnormality monitoring using power line signal analysis

An improved method to detect the reactor coolant pump (RCP) abnormality is suggested in this work. The monitoring parameters that are acquired from power line signal analysis are motor torque, motor speed and characteristic harmonic frequencies. The combination of Wigner–Ville Distribution (WVD) and feature area matrix comparison method is used for abnormality diagnosis. For validation of the proposed method, the test was performed during cool-down phase and heat-up phase in nuclear power plant (NPP) by cross-comparison with RCP vibration monitoring system (VMS). Using pump internal inspection results, the diagnosis prediction is verified.