According to the International Agency on Atomic Energy, 297 research reactors are currently in operation around the world [1]. More than 100 have been in use for over 30 years. Therefore, there is a pressing need to decommission and/or reconstruct them. More than 99% of the total activity of decommissioned reactors is induced activity. The long-term induced activity is of practical interest, since the reactor is held for a considerable time prior to decommissioning or reconstruction. One-dimensional models cannot provide the three-dimensional distribution of induced activity and do not permit the determination of the total activity of the reactor materials.In the present work, two-dimensional calculations of the long-term induced activity (disregarding the activity of the fuel) are conducted for materials from the IRT research reactor at Moscow Engineering-Physics Institute (MEPI), which went into operation in 1967.The two-dimemional model is shown in (R, z) geometry in Fig.

The future development of nuclear power requires the development of a new of generation of reactors that meet the modern international reliability and safety requirements. An example of such reactors is the promising power-generating unit with enhanced safety with a 1800 MW(th) water-moderated water-cooled BBER-500 reactor [1]. This plant employs the principle of self-shielding and is a further development of the BBR (PWR) plants, widely used in the world, with the traditional loop arrangement of the first loop.After power plants have reached their service life, they must be decommissioned [2, 3]. According to estimates made by the IAEA, by the year 2010 approximately 200 of the currently operating power generating units with an equivalent electric power of 1000 MW each will be decommissioned and dismantled [4].The disassembly costs are equal to 20-30% of the total construction costs of the power plants [3]. All components, included in the dismantling process will be radioactive to a greater or lesser degree. For accident-free standard operation, the induced activity exceeds 99% of the total activity neglecting the activity of the fuel [5, 6]. Long-lived induced activity is of practical interest, since disassembly is performed at least no earlier than two years after the reactor is shut down.Our objective in this work is to make a two-dimensional investigation of the long-lived induced activity of the structures and materials of a BBER-500 reactor after decommissioning.The BBER-500 reactor consists of the following basic units: a thick-wall metallic vessel, the top block with the safety and control rod drives, intravessel units (shaft, recess), and a core with an equivalent radius of 158 cm and a height of 355 cm.The reactor vessel consists of a vertical vessel, whose inner surface is covered with a 0.7 cm thick stainless anticorrosion facing. In the horizontal plane, passing through the center of the core, the equivalent thickness of the recess is 15.5 cm, the thickness of the shaft is 6.5 cm, and the thickness of the reactor vessel walls is 19.25 cm. The shielding beyond the vessel consists of concrete. A dry side shielding, consisting of serpentinite concrete with a thickness of 73.85 cm with a lining, followed by the standard concrete shielding, is installed on the right side. in the radial direction, at the level of the core.The service life of the reactor is taken to be 50 years with a power utilization coefficient of 0.8. The construction of the main unit of the BBER-500 reactor is planned for the Leningrad nuclear power plant.The construction described above corresponds to the layout of the BBER-500 reactor displayed in Fig, 1. An approximation along the radius r in the horizontal plane is displayed in Fig. 2. Here. r = 0 is the center of the core.The induced activity in the (r, z) geometry was calculated using the AKTIVATSIYA-2 program-constant system [7, 8], which includes the KASKAD-1 program [9, 10] with the DLC-23/CASK library of constants [11]. This system is designed for computational investigati...

Interest in reference experiments continues to increase. Such experiments are the basis for verification, validation, improvement, and correction of the computational programs, employed in radiation and transport problems, and libraries of radiation-matter intersection cross sections [1, 2]. Full-scale experiments, performed on real nuclear-power plants, are of special interest.With the further development and improvement of personal computers, reference calculations play an independent role together with experiments. Such calculations, performed with verified and tested programs and constants in good approximations, assessed on experiments, provide specialists more systematic differential data than an expensive experiment.Any description of the reference information should exclude equivocal interpretations of data, it should make it possible to reproduce the results, and its errors should be carefully assessed [3].The present work is devoted to a full-scale macroscopic (integral) reference experiment and calculation, oriented toward determination of the dose rate of activational -r-radiation in the shielding for the example of the first power-generating unit of the Armenian nuclear power plant with a BBER-440 reactor after the plant stopped operating. The reactor was put online in January 1977 and shut down on February 25, 1989.A BBER-440 reactor with a thermal power of 1375 MW (Fig. 1) consists of a 244 cm high cylindrical core with an equivalent radius of 144 cm, a 14 cm thick metal vessel, structures in the vessel (shafts with a bottom, cage with a compartment), and shielding consisting of serpentinite and standard concrete. The last 43.6 cm of the serpentinite concrete are borated (with boron carbide). The serpentinite concrete is poured into a metal structure with a 1.2 cm thick wall. A steel-foil thermally-insulating cylindrical part of the vessel is also secured on the metal structure. The nuclear concentration of the main elements and impurities for different zones of the power-generating unit were taken from the data of [4] and are presented in Table 1. The average monthly thermal power is presented in Table 2. The measurements were performed in a hollow experimental 70 mm in diameter channel with 3 mm thick steel walls, which was arranged in a radial direction in the serpentinite and standard concrete in a vertical plane at the level of the center of the core (see Fig. 1). The channel was provided beforehand during the construction of the power-generating unit. In the scheme represented in the two-dimensional (R, z) geometry of the reactor, the origin of the z-axis is located at the center of the core, the axis R lies in the horizontal plane passing through the center of the core. The average volume intensity of the neutrons in the core in the nominal regime was taken to be equal to 6.8.1012 cm-3-sec -1. The real distribution of the neutron density was chosen along the vertical direction in the core (Fig. 2), and a uniform radial distribution, with an error of several percent, was assumed.The radiation...

Analysis of the sensitivity of radiation field functionals to the parameters of the transport equation makes it possible to solve diverse problems of radiation physics. Those problems include evaluation of the shielding calculation error due to the indeterminacies of the initial parameters (radiation-matter interaction cross sections, assigned shielding characteristics, etc.), investigation of the laws of radiation propagation in the medium, determination of the input parameter that most affects the results, quantitative estimates of how small design changes affect the results, Issuance of practical recommendations for improving the quality of nuclear power plant design [1]. The last two are issues that are associated in one way or another with the process of logical optimization of the designs and radiation shielding of nuclear power plants. The functions of the relative sensitivity (in other words, the sensitivity profiles) serve as the initial information for optimization.In order to obtain the initial information for optimization of the VVI~R-440 reactor configuration, we have made a multigroup analysis of the sensitivity for the base model of reactor radiation shielding in the radial direction, using the example of the second unit of the Armyansk Nuclear Power Plant [2]. The chosen one-dimensional model of shielding in a cylindrical geometry corresponds to the reactor cross section in the horizontal plane passing through the center of the reactor core. The structure of the model and the size of the spatial zones are given in the first three columns of Table 1: the compartment is arbitrarily joined to the reactor cage.The radiation field and the conjugate functions were calculated from the one-dimensional ANISN program [3] in the PsS8 approximation of the scattering indicatrix and the radiation flux density, using the 171-group cross-section library VITAMIN-C [4]. Multigroup sensitivity coefficients were calculated from the modified program SWANLAKE-NG [5]. Some of the results are given as examples in Table 1 and Figs. I and 2. As the radiation field functionals we considered the fastneutron flux r t with an energy higher than 100 keV, which determines the radiation damage to the reactor vessel, the lowenergy neutron flux ~t/~, recorded with allv detector, in the vessel and in the first 50 cm of the structural concrete, which is responsible for the induced activity of the reactor materials. That induced activity must be known in order to solve the problems of decommissioning the reactor, the equivalent dose behind the radiation shielding being N. Table 1 also gives the coefficients of the relative sensitivity of the given functionals to the interaction cross section (to a change in the density of the materials of the plant) as well as for the individual spatial zones and for the entire configuration. The energy dependences of the functions of the relative sensitivity are given as an example in Figs. 1 and 2.The data in Table 1 indicate that for a fast-neutron flux r in the reactor vessel the sensitivity to t...