EXPERIMENTAL STUDIES ON DEVELOPMENT OF THE METHOD FOR DETERMINING THE AMOUNT OF IMPURITATED GASES IN NUCLEAR FUEL

The article is devoted to an issue of estimating the impurity gas amount in nuclear fuel in the aspect of the distracting contribution from released gases to the total pressure inside ampoule of the device in the simulating a severe accident with core melting. The paper presents a method based on measuring the pressure and temperature of gas in a closed values of the fuel elements during the fuel melting. The correctness of the developed methodology is confirmed by the results of experiments on the melting of fuel in a pulsed graphite reactor IGR with the implementation of a controlled neutron pulse.

Modeling of thermal processes in the contact zones of fuel elements

In this work, using two specific examples, a general approach to the mathematical modeling of thermal processes in the contact zones of fuel elements in the development and optimization of various technological processes, systems and devices is considered. In the first example, a mathematical model of heat transfer in the contact zone (metal-hybrid thermal interface) between the heat-generating element and the heat-dissipating radiator is considered. In the second case, the thermal process in the processing of materials with a bonded diamond tool in the contact zone "diamond grain – binder – processed material" is considered and analyzed. The general approach to modeling thermal processes in the contact zones of various fuel elements makes it possible to optimize the parameters of technological processing modes and the correct operating conditions for products and systems

Features of methods for monitoring the fuel cladding tightness in lead-cooled fast breeder reactors

To timely detect failed fuel elements, a reactor plant should be equipped with a fuel cladding tightness monitoring system (FCTMS). In reactors using a heavy liquid-metal coolant (HLMC), the most efficient way to monitor the fuel cladding tightness is by detecting gaseous fission products (GFP). The article describes the basic principles of constructing a FCTMS in liquid-metal-cooled reactors based on the detection of fission products and delayed neutrons. It is noted that in a reactor plant using a HLMC the fuel cladding tightness is the most efficiently monitored by detecting GFPs. The authors analyze various aspects of the behavior of fission products in a liquid-metal-cooled reactor, such as the movement of GFPs in dissolved and bubble form along the circuit, the sorption of volatile FPs in the lead coolant (LC) and on the surfaces of structural elements, degassing of the GFPs dissolved in the LC, and filtration of cover gas from aerosol particles of different nature. In addition, a general description is given of the conditions for the transfer of GFPs in a LC environment of the reactor being developed. Finally, a mathematical model is presented that makes it possible to determine the calculated activity of reference radionuclides in each reactor unit at any time after the fuel element tightness failure. Based on this model, methods for monitoring the fuel cladding tightness by the gas activity in the gas volumes of the reactor plant will be proposed.

Исследование поверхностного окисления циркония активным тепловым методом

The possibility of testing the surface oxidation of zirconium has been investigated by the method of one-way active thermal non-destructive testing based on the analysis of radiation temperatures. The emissivity of the oxidized surface of zirconium samples obtained at different annealing temperatures was estimated at various stages and heating temperatures in the infrared wavelength range. It is shown that there is a principal possibility to remotely test the oxidation process of zirconium alloys used in the nuclear industry for the manufacture of fuel elements operating in the core of nuclear reactors.

Analysis of different factors on the features of thermal destruction of a fuel pin

The presented work studies the influence of various factors that affect the specific features of fuel pins melting. For this purpose, fuel pins with different geometries and energy release are considered. Numerical simulation of melting is carried out using a program module for calculating the destruction of fuel rods. Comparison with theoretical calculations is made. The analysis of the convergence of calculations with respect to the time step value and the number of calculated cells along the radius and height is carried out. As a result of work with the use of numerical methods, the characteristic times of destruction of fuel elements during an accident with a loss of coolant flow rate (an accident of the ULOF type) and the dependence of weight loss on time are obtained under various conditions.

On the Development of Tolerant Fuel --- a New Generation Fuel

The concept of tolerant fuel is considered as applied to water-cooled power reactors. The concept is based on eliminating the steam-zirconium reaction. For this, two work areas, i.e., using the physical and the thermodynamic barriers, were considered. Physical barrier presupposes exclusion of contact between water and zirconium, and the thermodynamic barrier (the most radical method) envisages replacement of the alloy containing zirconium with other materials inert to water when exposed to high temperature in the reactor core (∼ 1200 °C). Consequences of the most devastating accidents at the nuclear power plants in the world were discussed: Three Mile Island, Chernobyl and Fukushima. The latest accident in Japan brought to the fore the concept of tolerant nuclear fuel, i.e., being resistant to accidents. Work orientation in creating the tolerant fuel is indicated. Main attention is paid to materials and technologies applied to tolerant fuel. General requirements to safety analysis of the reactor facility fuel system currently developed in the Russian Federation and abroad, as well as current safety criteria for fuel elements, under design-based accidents are presented. Procedure for calculating justification of the safety criteria fulfillment for fuel elements under design-basis accidents is briefly considered. Main characteristics of the new generation materials under development for reactor cores as applied to tolerant fuel are presented. Based on comparing the proposed materials as the tolerant fuel for the fuel element claddings, composite materials based on the heat-resistant SiC/SiC ceramic system could be recommended, and as far as fuel materials are concerned --- materials with increased density, uranium capacity and thermal conductivity values, i.e., nitride fuel and fuel made of uranium silicide

Impact of Radiation-Induced Microstructures on the Integrity of Spent Nuclear Fuel (SNF) Elements in Long-Term Storage

The current safety concept provides for a period in the range of 40 years for interim storage of spent fuel elements. Since the requirement for proof of safety for to up to 100 years arises, the integrity of the spent fuel elements in prolonged interim storage and long-term repositories is becoming a critical issue. In response to this safety matter, this study aims to assess the impact of radiation-induced microstructures on the mechanical properties of spent fuel elements, in order to provide reliable structural performance limits and safety margins. The physical processes involved in radiation damage and the effect of radiation damage on mechanical properties are inherently multiscalar and hierarchical. Damage evolution under irradiation begins at the atomic scale, with primary knock-on atoms (PKAs) resulting in displacement cascades (primary damage), followed by the defect clusters leading to microstructural deformations. In this context, we have developed and applied a multiscale simulation methodology consistent with the multistage damage mechanisms and the corresponding effects on the mechanical properties of spent fuel cladding and its integrity. Within the improved hierarchical modelling sequence, the effect of the radiation field on the fuel element cladding material (Zircalloy-4) is assessed using Monte Carlo methods. A molecular dynamics method is employed to model damage formation by PKAs and primary damage defect configurations. The formation of clusters and evolution of microstructures are simulated by extending the simulation sequence to a longer time scale with the kinetic Monte Carlo (KMC) method. Transferring the calculated radiation-induced microstructures into macroscopic quantities is ultimately decisive for the structural/mechanical behaviour and stability of the cladding material, and thus for long-term integrity of the spent fuel elements. Results of the multiscale modelling and simulations as well as a comparison with experimental results will be presented at the conference session.

STUDY OF MATERIALS AND WELDING USED IN THE CONSTRUCTION OF RACKS FOR STORAGE OF FUEL ELEMENTS AT NUCLEAR PLANTS

This paper showed the technological innovations and the necessary requirements for the welding of ASTM A240 TP316L Austenitic Stainless Steel in the construction of racks used in the storage of Fuel Elements inside nuclear power plants. It presents the development of welding processes using coated electrode, SMAW and as addition metal rods EAS 2-IG / ER 308L. This study is divided into two stages, the preparation of technical documentation and the development of methods and manufacturing processes used in the qualification of welding processes. A sequence was outlined based on real situations used by nuclear component manufacturing companies, meeting the physical and mechanical properties required by the nuclear classification standards and their regulations. The results showed that the welding processes were satisfactory, that the destructive and non-destructive tests showed that there was no discontinuity in the surface and defects in the volume of the welding and that, in the present study, safety in the project for the operation of a Nuclear Power Plant was demonstrated.