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The RBMK-1500 reactor at the Ignalinsk Atomic Power Plant was designed for fuel of 2% enrichment. In the first unit, after removing all the additional absorbers intended to compensate for the excess reactivity in reactor startup, the depth of fuel burnup is 18-19 MW.day/kg. The mean operational reserve of reactivity is 36 manual control rods.Analysis of the causes of the accident at the Chernobyl Atomic Power Plant has implicated design deficiencies of the control rods and a nonoptimal uranium-graphite ratio, as a result of which the steam reactivity % was (4-5)~ [1]. After the accident, measures to increase reactor safety were taken at all atomic power stations with RBMK reactors, including the Ignalinsk plant [2, 3]. In the first stage, the steam reactivity was reduced to 1/3 by loading 52-54 additional absorbers and increasing the operational reactivity reserve to 55 rods. This eliminated the possibility of uncontrollable increase in reactor power (on the basis of instantaneous neutrons) in he case of dehydration of the active zone.Increasing the number of absorbers in the active zone considerably reduces the depth of fuel burnup and impairs the economic characteristics of the fuel cycle. The depth of fuel burnup is decreased to approximately 14 MW-day/kg, i.e., by 25 %. In addition to direct economic losses due to reduction in fuel burnup, spent-fuel storage becomes a problem, because the accelerated rate of fuel-rod replacement leads to rapid filling of the storage tanks.In the second stage, fuel of enrichment 2.4% was introduced at RBMK-1000 reactors. This restored the design burnup depth and considerably improved the economic characteristics of the fuel cycle. However, in RBMK-1500 reactors, the accompanying increase in graphite temperature prevents increase in the initial enrichment. Calculations show that, despite reduction in the permitted thermal power to 4200 MW, the increase in nonuniformity of energy liberation on transition to fuel of 2.4% enrichment leads to disruption of the operational limits. Hence, other approaches are required to improve the economic performance of RBMK-1500 reactors while maintaining safe operation.In 1987, an intensive search for a more economical means of reducing c%, other than the introduction of additional absorbers, began at the Kurchatovskii Institute Russian Scientific Center (RSC) and the Scientific-Research and Design Institute of Energy Technology (SRDIET). Around 30 different designs of the fuel assembly and fuel and casing materials were considered [4, 5]. The addition of erbium to uranium dioxide proved most promising. Staff at the Ignalinsk plant had the idea od using erbium as the material for the additional absorbers and rods loaded in the fuel assembly. However, subsequent research showed that the only practical approach with a real economic impact (without loss of safety) is to place the erbium in the fuel. Note that non-Russian research on the use of erbium in PWR was published at about the same time (the late 1980s and early 1990s) [6, 7]. PROPERTIES OF ...

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Recommendations for calculating the thermal creep properties of uranium dioxide for fuel element serviceability analysis programs are developed on the basis of a physical model. The deformation processes in the model include diffusion and diffusion-controlled motion of dislocations. It is shown on the basis of the analysis of the thermodynamics of point defects in ionic crystals that the diffusion of ions is controlled by a vacancy mechanism and that the diffusion coefficient depends on the temperature and oxygen coefficient. The model includes the influence of temperature, stress, fuel density, grain size, and oxygen ratio on the creep rate. The relations obtained in the this work have made it possible to improve by approximately a factor of 10 the agreement between the calculations and experimental data as compared with the empirical relation used previously to describe the characteristics of creep.The characteristics of creep of a fuel core are largely determined by the stress-strain state of the cladding. The relations for calculating them are an integral and important part of modern computer codes used for analyzing the serviceability of fuel elements. At the present time, primarily empirical relations recommended by the Matpro library of the properties of reactor materials [1] are used to calculate the creep rate of oxide fuel. However, these relations do not properly take account of the effect of the deviation from stoichiometry on the creep rate and activation energy. The computational results differ from the experimental data by more than a factor of 100. Since they do not follow from physical ideas about creep the Matpro library recommendations are unsuitable not only for extrapolation but also for interpolation of experimental data.A review of the experimental data has shown that the thermal creep rate of uranium dioxide under compression is a linear function of the stress for σ ≤ 30-40 MPa. A power-law dependence with exponent 4-5 is observed at higher stresses. In the linear range, the creep rate is inversely proportional to the squared grain size. The activation energy of creep is close to the activation energy of uranium diffusion by the vacancy mechanism. Therefore it can be stated that at low stress creep is controlled by the diffusion mechanism and at high stress by dislocation climb processes. These two mechanisms operate in parallel, but their contribution to the total deformation depends strongly on the acting stresses. A generalization of the theory of diffusion creep for polycrystals with different methods of accommodation of grains is presented in [2] together with relations for creep by the dislocation climb mechanism. Therefore, in a wide stress range the creep rate can be described by a sum of two terms [2-4]:

This is a report of a study of the effect of alloy additives on the properties of fuel under conditions typical of water cooled reactors. The behavior of uranium oxide fuel with added mixtures of the oxides of aluminum, silicon, niobium, and iron during reactor irradiation of experimental fuel elements is investigated in the MIR research reactor. The feasibility of using aluminum-silicate additives for improving the operating characteristics of fuel pellets under reactor irradiation conditions is demonstrated.One of the approaches for improving the characteristics of fuel elements is to optimize the microstructure of the fuel (grain size, porosity) by means of alloy additives [1].The work reported here was done for the purpose of studying the effect of alloy additives on the properties of fuel under the operating conditions for fuel elements in water cooled reactors. The behavior of uranium oxide fuel with added mixtures of the oxides of aluminum, silicon, niobium, and iron during reactor irradiation of experimental fuel elements was investigated in the MIR research reactor.An experiment with an abrupt change in the power of VVER-1000 fuel elements was conducted in 1991 in the PVK-2 loop of the MIR reactor using a technique developed at the NIIAR [2]. An irradiation device ( Fig. 1) with 72 experimental fuel elements (Fig. 2) was loaded into channel 2-6 in the assemblies for units Nos. 1, 5, 6, and 4 with 18 fuel elements apiece. The characteristics of the fuel elements are listed in Table 1.During the experiments, the device with the test fuel elements is mounted in the channel of a loop assembly in which the parameters of the coolant are maintained at levels corresponding to the VVER. The reactor is brought to a power sufficient to ensure attainment of the required initial conditions of the experiment in the loop channel. The control rods closest to the loop channel lie below it. After all the parameters are stabilized and equilibrium states are reached at this power level, the power is abruptly increased by the required amount. This is done by extracting the nearest control rods with compensation of the resulting positive reactivity by inserting control rods in other sites within the core.Test Procedure. The variation in the parameters during the experiment is illustrated in Fig. 3. The abrupt power increase took place over 13 min. Then the fuel elements were tested at the higher power for another 10.5 h, after which the reactor was stopped to unseal the fuel elements in another loop channel. According to the readings of the shell seal control system, all the fuel elements in this experimental device remained intact. A maximum amplitude of the power increase for the fuel