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...
The results of computational studies on choosing radiation protection for planetary-surface nuclear power plants are present. Protection on the base of a 0.4-1.5 MW(t) YaEU-100 thermionic space reactor was considered for a Martian nuclear power plant and a 0.36 and 0.6 MW(t) YaEU-25 reactor was considered for a lunar reactor. The mass/size characteristics of the radiation protection were obtained for different arrangements of the nuclear power plant on the planet -directly on the surface with protection delivered or an embankment consisting of local soil and in a shaft prepared beforehand.The computational investigations of the choice of radiation protection for planetary-surface nuclear power plants which were performed at the Physics and Power Engineering Institute are presented. Protection based on the 0.4-1.5 MW(t) YaEU-100 thermionic space reactor was considered for a Martian nuclear power plant and 0.36 and 0.6 MW(t) YaEU-25 and -50, respectively, reactors were considered for a lunar nuclear power plant. Different variants of radiation protection were considered for both planets: natural or artificial burial of the nuclear power system, embankment of the nuclear power plant, and a radiation protection block consisting of materials delivered from Earth.The characteristics of radiation transfer and the choice of an optimal configuration for the radiation protection were analyzed using the domestic codes RAPID [1] and KASKAD [2] (two-dimensional geometry). The final (verification) calculations of the radiation fields in the reactor, shielding, and surrounding space (atmosphere and soil on Mars and the Moon) were performed with the MCNP code [3] using ENDF/BVI constants [4].Protection for a Martian Nuclear Power Plant. The radiation protection must provide on a scientific base 1 km from the nuclear power plant a yearly dose of reactor radiation not exceeding 2 cSv (Fig. 1). The angle in the direction of the base protected from the nuclear power plant radiation was 90°.The particulars of radiation transfer in the Martian atmosphere, whose density is approximately 100 times lower than that on Earth and which is 95% carbon dioxide, should be noted [5]. The Martian atmosphere, which essentially does not attenuate the direct reactor radiation, does strongly scatter it, which makes it impossible to use shadow protection. Calculations performed to determine the influence of the radiation scattered in the Martian atmosphere on the radiation environment showed that a hemisphere with radius at least 11 km makes a substantial contribution to the yearly equivalent dose. For an unshielded reactor, the fraction of the radiation scattered in the atmosphere is 16% for neutrons and 2% for photons ( Table 1). The radiation scattered by the ground also makes an appreciable contribution. Since the dose of the scattered radiation from an unprotected reactor exceeds the prescribed levels, circular shielding is required. Shaping the protection could substantially decrease its mass. Figure 2 illustrates the need for shaping. This figu...
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