Modeling of elastic-plastic deformation of the ring parts as applied to the analysis of the results of testing the material of the fuel-element cladding

An analytical formula for a smooth description of the tension diagram of EK-181 steel and a method for rearranging the diagram when changing the direction of deformation are proposed for the first time. The process of straightening a quarter of an annular sample and further stretching is numerically modeled. It is shown that the conditional yield strength of the material of the straightened sample is 7.5% less than the actual conditional yield strength of steel. It is shown that the test for pure bending of a cantilever sample in the form of a semicircle with the processing of the bending diagram (by analogy with GOST 3565–80 for torsion) provides an estimate of the conditional yield strength which is 32% higher than the actual yield strength. The possibility of numerical reconstruction of the tension diagram from the diagram of pure bending of a cantilevered semi-ring sample is proved. It is shown that this procedure really gives the value of the conditional yield strength of steel EK-181 with a tolerance for the residual deformation of 0.2%. The analysis of the test procedure for the rings of fuel element cladding and the proposed algorithm for determination of the conditional yield stress of the ring material is carried out. Attention is drawn to the arbitrariness of the choice of the designed load on the two-stage diagram of the diametrical tension of the ring and to the lack of scientific substantiation of the possibility of determining the yield stress on the second part of the diagram. It is shown that this method in the current form contradicts GOST for tensile testing due to the absence of a base with a uniform stress state on the ring. Therefore, the considered method is not recommended for determining the values of the conditional yield strength suitable for strength calculations.

IMPROVEMENT OF FUEL ELEMENT SHELL CONTROL METHODS TO INCREASE NUCLEAR REACTOR SAFETY

The analysis of the existing methods of control of the surface of the fuel element cladding material was carried out, which showed that their use for detecting surface and internal defects, such as local inhomogeneities, micro- and macropores, various cracks, axial looseness, etc. is characterized by low efficiency, is a laborious process that requires additional surface treatment, material of the fuel elements cladding. In addition, the investigated methods of controlling the surface of the fuel element cladding material make it possible to visually identify only rough external cracks and large slag inclusions, small cracks and non-metallic inclusions invisible under the slag layer. It is proposed to assess the quality of the surface of the shell material in case of its damage and destruction, the use of a computational apparatus based on the method of the theory of fractals. It is proposed to use the fractal properties of the shell material structure and a quantitative fractal value – the fractal dimension, which makes it possible to determine the degree of filling of the volume of the shell material structure during fuel element depressurization. A mathematical model of damage to the structure of the fuel element cladding material is developed depending on the simultaneous effect of high temperature and internal pressure caused by the accumulation of nuclear fuel fission products between the nuclear fuel pellet and the inner surface of the fuel element cladding, taking into account the fractal increases in the geometric parameters of the fuel element cladding. It is shown that damaged structures of the fuel rod cladding material depend on the pressure and temperature inside the fuel rod cladding, as well as the fractal increase in geometric parameters, such as: volume and surface area, outer and inner diameters, height and cross-sectional area, cladding length and height of nuclear pellets, gap between the inner surface of the cladding and nuclear fuel. A criterion for assessing the integrity of the fuel rod cladding is determined, which depends on the change in geometric values in the event of damage and destruction of the structure of the fuel rod cladding material. Practical recommendations are given on the use of the proposed method for monitoring the tightness of the fuel element cladding for processing information obtained from the computational module of the system for monitoring the tightness of the cladding for the automated process control system of the nuclear power plant power unit, which makes it possible to detect the depressurization of fuel elements at an earlier stage in comparison with the standard procedure.

Approximate model for evaluation of thermal-hydraulic transients associated with rapid power increase in research nuclear reactor

A simple model, for the estimation of changes in the nuclear fuel element cladding temperature as well as the conditions of the formation of the onset of nucleate boiling, is proposed. The results of this estimation are sufficient to assess the effect of the transient with the peak of the reactor's power on the device's safety, without the need for time-consuming thermal calculations. The basic parameters determined using the proposed model are the maximum wall temperature of the device in a hot spot, the time constant of the wall temperature change, and the course of changes in the onset of nucleate boiling ratio in time. The model may be used for investigating the thermal behavior of thin heat-generating and water-cooled elements (such as fuel elements or uranium irradiation targets) during rapid power rise. The results of the temperature estimation with the proposed model were tested considering the hot spot in the MR-6 type nuclear fuel. The SN code with coupled kinetics and thermal-hydraulics, developed in the MARIA reactor was used to validate the results. The maximum cladding temperature, the thermal time constant and the onset of nucleate boiling ratio parameter simulated by the SN code and the proposed scheme appeared to be very consistent.

The Forecasting of Corrosion Damage of Structural Materials during Dry Long-Term Storage of RD BN-350 SNF with CC-19 SFA

There are results of long-term thermal aging of samples of irradiated and nonirradiated FA jacket and nonirradiated fuel element cladding at a temperature range from 300 to 550°C in argon, to 600°C in air. Materials have been studied before and after thermal tests. The forecast estimation of expected corrosion damage of barrier material at the radionuclide release from spent fuel assemblies of BN-350 reactor into environment during dry storage for 50 years was carried out.

The Influence of Thermal Shocks on the Thermophisics Properties of the Zircaloy-4

This paper presents the researches on the influence of thermal shocks on the heat transfer properties of the zircaloy-4, now used in fuel element cladding of third generation nuclear reactors. Thermal shock testing was performed using solar energy at high temperatures, up to 1350�C, with 1, 3 and 6 thermal cycles of 60s. The determination of the thermal diffusivity of the tested samples was made by the flash method at 350 �C, the operating temperature of the third-generation nuclear reactors.