Methodological program techniques for performing neutron-physical calculations of fast-reactor cores with a description of the main codes TRIGEX, JARFR, GEFEST, MMKKENO, and ModExSys are described. The results of verification and assessment of the methodological errors on the basis of a test model of BN-600 are presented. Transport effects and mesh errors in the reaction rate distributions in the core, lateral screen, and in-reactor storage area are evaluated. An integral assessment of the computational accuracy is performed by comparing with the experimental data obtained by γ-scans of BN-600 fuel assemblies. It is shown that calculations describe the experimental data over the core with an error no worse than 5%. In the lateral screen and the in-reactor storage area, the discrepancy between the calculations and experiment does not exceed 20-30%.Computational modeling of the neutron field is the main method of determining the energy release in a fast-reactor core, where, as a rule, there is no apparatus for tracking the power density. Correspondingly, computational accuracy is one of the key elements of the operational reliability of fuel elements and assemblies. Another aspect of the problem is substantiating the conservatism of fast-reactor core design -design limits or margins affecting the technical-economic characteristics. Experience in operating BN-600 makes it possible to evaluate the accuracy of the design and operational programs under real conditions of a commercial medium-power fast reactor and to take it into account in subsequent development work.Such an analysis is based on the measurement of the power density in BN-600 by means of γ-scanning. Eleven such experiments were completed in the time period from physical startup of the reactor in 1980 and up to 2008. The present article analyzes the last experiments performed in 2003-2006 in the course of transitioning from a 01M1 core to a 01M2 core with maximum burnup 11.1% [1]. This upgrade was an important next step in making use of the serviceability margin of reactor fuel assemblies, which required greater attention to monitoring the characteristics of the core [2].Program and Constants System for Neutron-Physical Calculations of Fast Reactors. A consistent system for performing neutron-physical calculations of fast reactors is now available. It includes the program systems TRIGEX [3], JARFR [4], GEFEST [5], and MMKKENO [6] as well as the BNAB constants library [7,8]. It should be noted that the base
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Computational tracking of BN-600 operation is described. The high quality of computational tracking is largely due to the nature of a fast reactor, in this case . Unlike reactors with a thermal neutron spectrum, in a fast reactor, because the prompt and delayed fission neutrons as well as the absorbed neutrons are almost in the same energy range as the fast neutrons, a computational cell can be confidently homogenized and the reactor is strongly coupled to the neutron field. These are the reasons why the behavior of the reactor can be successfully predicted by means of computational programs which are based on the diffusion approximation neglecting the anisotropy of the interaction of the neutrons and the heterogeneity of the medium.The GEFEST system of computer programs for performing three-dimensional neutron-physical calculations was developed to validate the safe operation of sodium-cooled fast reactors with uranium and uranium-plutonium fuel [1]. This system is used to determine the neutron field and the energy release at the points of a computational model of a reactor (about 20000) taking account of the real position of the control rods. This information is especially important when adequate information from measuring instruments is unavailable. Aside from assessment of the integral and local characteristics of the reactor, taking account of the burnup and the displacement of the control rods during reactor operation, the GEFEST system makes it possible to reconstruct the history of each fuel assembly, control rod, and other structures loaded into the core.The GEFEST system consists of a collection of program-independent modules, which solve the problem in an autonomous operating regime. The modular structure makes it possible to investigate in detail the state of the reactor, for example, to determine the reactivity or kinetics of the reactor and other properties. Each module can be replaced by another one, intended for solving the same problem but using new knowledge. The integral parts of the system are the fuel archive and the working files. The certificate data and the basic computational characteristics of the fuel assemblies and the control rods are stored in the fuel archive. Intermediate computational results are contained in the working files.First Version of the GEFEST System with the BNAB-78 Library and ARAMAKO Constants Preparation System. The program JAR and then URAN were used to calculate the first microruns of the BN-600 reactor. At the end of 1987, a three-dimensional program system, called GEFEST, for performing operational calculations together with the earlier two-dimensional code was installed in the Beloyarskaya nuclear power plant. The initial version of the system of programs was intended to perform calculations on the ES-1055 computer. When PCs appeared a version of the system was developed for them. Aside from the computational part, it included means for performing analysis and for graphical representation of data. In 1992, the GEFEST system with the ARAMAKO-S1 constants prepara...
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