MCNP calculation of the neutron-physical characteristics of a reactor and radiation protection in a nuclear power facility for use in spacecraft

Abstract

MCNP [1] is a high-precision computer code that makes it possible to perform Monte Carlo simulation of neutron, γ-ray, and electron transport in systems with a complicated three-dimensional geometry. The code was developed over a period of many years, during which the range of solvable problems expanded. MCNP employs an ENDF/B [2] based library of nuclear constants with continuous variation of the particle energy. The library includes data on the neutron and photon interactions, activation, and neutron thermalization for different material temperature. The geometric objects are given by surfaces, such as planes, spheres, cylinders, and cones, making it possible to describe complex systems in detail, for example, an electricity generating channel (EGC) consisting of several electricity-generating elements with multilayer collector packets, gas-removing apparatus, gaps, emitters, and fuel kernels. Elements of periodicity in radial-azimuthal and vertical directions of the core can be prescribed. Examples of the prescription of complex geometries are shown in Figs. 1 and 2.Reactor Calculation. The MCNP code was used to determine the neutron multiplication coefficients with estimation of the maximum reactivity and the effectiveness of the control organs and to calculate the regulatory characteristics of the control organs, emergency situations associated with the reactor falling into water, transverse and longitudinal compression of the reactor caused by the reactor striking a solid surface, and other situations. To obtain the regulatory characteristics, the orientation angle of the absorbing lining relative to the reactor core was changed by a small amount (Fig. 3). Emergency situations which are associated with transverse compression of the reactor, caused by impact against a solid surface, were simulated by ellipses for the circles forming the lateral surface of the core, vessel, and lateral reflector. The lattice of EGCs was tightly squeezed and the peripheral EGCs were redistributed into the open spaces appearing in the core. The ratio of the semiaxes of the ellipse was varied from 1.1 to 1.4. A free space in the connecting crosspieces of the EGCs was chosen for the longitudinal compression of the reactor.The MCNP code was used to calculate the energy release in the reactor. Here, it possible to estimate the total energy release from various reactions, for example, nuclear fission, (n, γ), and (n, 2n), as well as from individual nuclear reactions. In the computational model, the fuel kernels of the electricity generating elements were divided into very thin layers (Fig. 4). A lattice over the height of the fuel was created in order to simplify the formulation of the model. The flux of neutrons and γ rays was calculated in 26 neutron energy groups (Fig. 5). The reactivity change over a reactor run is simulated with the program complex for isotopic kinetics ORIGEN2 (USA) [3]. The Monteburns program (USA) is used to link the program systems [4].Radiation Protection. Large computational capacity is necessary in order...

Minimize fission power peaking factor in radial direction of water-cooled and water-moderated thermionic conversion reactor core

“…The MNCP code (MCNP 1997) and the ENDF/B-6 evaluated nuclear data library (ENDF/B-VI 1994) were used to model and calculate the core, with the neutron thermalization (Krotov and Son'ko 2009) taken into account.…”