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At the present stage of development of nuclear power it is expected that the operating conditions for nuclear power plants will become more complicated. In addition, reactors are required to be more reliable and economical. Reliability is substantiated on the basis of computational and experimental investigations of the strength and thermophysical characteristics of the fuel elements in different reactor-operation regimes, most of the work involving computational analysis. It is obvious that the quality of the computational substantiation depends on the adequacy of the modeling of numerous and interrelated processes by the computational apparatus employed [1].The available computational programs for modeling thermophysical processes in nuclear fuel elements, for example, different versions and modifications of the PIN code [2] and the thermophysical code RET [3], cannot be used for adequate substantiation of the fuel element reliability, as a whole, since they do not consider mechanical factors. Such factors include viscosity, plasticity of the materials of the cladding and the fuel, the stress gradients in the fuel and the cladding, and the mechanical interaction of the fuel with the cladding. These factors determine the strength and reliability criteria for the fuel elements, and they also strongly influence the thermophysical processes. The results of modeling as well as experimental data show that approximately half of the change in the diameter of the cladding and in the elongation of a fuel element are caused by the interaction of the fuel with the cladding. The rest of the change is due to creep of the cladding under the pressure differential and the radiation-induced growth. When the mechanical processes are neglected, an error is made in the calculation of the fuel-cladding gap and therefore in the temperature, and this error increases with burnup.A more accurate description of the coupled thermophysical and thermomechanical problem is presented in [4]. However, when the Monte Carlo method is used to describe the interaction of the fuel with the cladding, the computer time increases substantially. Moreover, the model presented neglects the changes in the structure of the fuel and some thermophysical processes. In the present paper we describe a method, implemented in the PULSAR-2 computational code, that is intended for modeling thermophysical, strength, and reliability characteristics of BBI~R and RBMK fuel elements in the case when the reactors operate in quasistationary, transitional, and maneuvering modes.General Formulation of the Problem. The calculations were performed on a full-scale model of a cylindrical fuel element with a given construction. In modeling the operation of the fuel element, the following processes are taken into account: thermoelastic expansion of the fuel and cladding; creep of the fuel and cladding; plastic deformation of the cladding when the fuel element reaches nominal power and during maneuvering of the power; swelling of the fuel; radiation-induced additional sintering of...
At the present stage of development of nuclear power it is expected that the operating conditions for nuclear power plants will become more complicated. In addition, reactors are required to be more reliable and economical. Reliability is substantiated on the basis of computational and experimental investigations of the strength and thermophysical characteristics of the fuel elements in different reactor-operation regimes, most of the work involving computational analysis. It is obvious that the quality of the computational substantiation depends on the adequacy of the modeling of numerous and interrelated processes by the computational apparatus employed [1].The available computational programs for modeling thermophysical processes in nuclear fuel elements, for example, different versions and modifications of the PIN code [2] and the thermophysical code RET [3], cannot be used for adequate substantiation of the fuel element reliability, as a whole, since they do not consider mechanical factors. Such factors include viscosity, plasticity of the materials of the cladding and the fuel, the stress gradients in the fuel and the cladding, and the mechanical interaction of the fuel with the cladding. These factors determine the strength and reliability criteria for the fuel elements, and they also strongly influence the thermophysical processes. The results of modeling as well as experimental data show that approximately half of the change in the diameter of the cladding and in the elongation of a fuel element are caused by the interaction of the fuel with the cladding. The rest of the change is due to creep of the cladding under the pressure differential and the radiation-induced growth. When the mechanical processes are neglected, an error is made in the calculation of the fuel-cladding gap and therefore in the temperature, and this error increases with burnup.A more accurate description of the coupled thermophysical and thermomechanical problem is presented in [4]. However, when the Monte Carlo method is used to describe the interaction of the fuel with the cladding, the computer time increases substantially. Moreover, the model presented neglects the changes in the structure of the fuel and some thermophysical processes. In the present paper we describe a method, implemented in the PULSAR-2 computational code, that is intended for modeling thermophysical, strength, and reliability characteristics of BBI~R and RBMK fuel elements in the case when the reactors operate in quasistationary, transitional, and maneuvering modes.General Formulation of the Problem. The calculations were performed on a full-scale model of a cylindrical fuel element with a given construction. In modeling the operation of the fuel element, the following processes are taken into account: thermoelastic expansion of the fuel and cladding; creep of the fuel and cladding; plastic deformation of the cladding when the fuel element reaches nominal power and during maneuvering of the power; swelling of the fuel; radiation-induced additional sintering of...
Our objective in the present work is to test a new computational code capable of simulating the spatial dynamics of the escape of radioactive fission products from unsealed fuel elements into the ftrst-loop coolant. To this end, we employed a problem that essentially consists of estimating the scales of variation of the t31I radioactivity in the first-loop coolant under conditions of a transient process in a VVt~R-440 core. The transient process consists of a deformation of the spatial field of energy release during the motion of a reactivity regulating rod. Situations in which one unsealed fuel element can be located in different fuel assemblies of the core were studied. An adequate numerical simulation of this complicated transient process, which depends on many parameters, requires a simultaneous solution of the neutron, thermophysical, and radiation problems.At present, calculations ordinarily performed for the purpose of substantiating new reactor designs for nuclear power plants are performed using codes developed at the beginning of the 1990s, in which the computational methods presented in, specifically, [1][2][3] are implemented. A special feature of these computational methods is that they are oriented toward timeand space-averaged parameters, such as, energy release, temperature, and physicochemical properties of the medium. Furthermore, the models do .not take into account important mechanisms that affect the escape of fission products from unsealed fuel elements during transient processes with a rapid variation of the temperature of a fuel element. These limitations were overcome in the VVERRAD code, which is briefly presented in the present paper. The code is a further development of a computational method now employed for estimating the radioactivity of the coolant in the first loop of a VVI~R reactor [1, 2].The VVERAD program consists of three modules, the main one being a block that calculates the coolant radioactivity due to the escape of fission products from unsealed fuel elements. The other two modules are auxiliary modules and are used for calculating the nonstationary heat-release and temperature fields in the reactor core on the basis of relatively simple mathematical models. To perform precise calculations and for modeling problems in which the limitations of the models are substantial, the blocks for calculating the energy release and temperature are not used. Under these conditions, these blocks are replaced by prescribing initial data in the form of two-dimensional (coordinate, time) arrays of values (heat release and temperature), which are obtained in separate, accurate neutron-physical and thermophysical calculations. This organization of the problem makes it possible to reduce substantially the time required for preliminary (ordinarily numerous) computational variants.Let us consider in greater detail the computational models implemented in the main module VVERRAD. Models that make it possible to perform detailed calculations of the yield of gaseous fission products from the fuel m...
Al~traet. Over the past 20 years computer modelling of fuel performance has developed into a well-established procedure, which has been valuable in the understanding and the improvement of fuel rod behaviour. The range of computer models presently applied to cylindrical ceramic fuels is reviewed. A critical appraisal is made of the numerical techniques used for mechanical and thermal analysis. The necessity for benchmark calculations is emphasized and various approaches for model validation are discussed. A number of special topics are chosen for deeper discussion. These include: improved description of cladding deformation and the estimation of failure; the analysis of stress concentrations in the cladding; fission gas analysis; and chemical modelling. Finally some speculation is made on the future of fuel rod modelling.
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