Vol 120, No 3 (2016)

An approach to solving the three-dimensional transport equation is described. Directional probabilities, which make it possible to analyze axial leakage numerically, are defined. The problems of modeling the boundary conditions for the neutron flux under ordinary operating conditions of reactors and under the conditions of dewatering are discussed. The results obtained show that if the effect of the spectrum of the two-dimensional problem of a VVER cell is taken into account, then the axial leakage changes, especially near the reflector. The axial leakage also increases if the cell contains dewatered channels, but only if the boundary conditions are modeled in the form of mirror reflection on a real hexahedral boundary. This effect vanishes if the equivalent cylindrical cell approximation is used and the boundary conditions are modeled in the form of isotropic reflection.

The concentration of neutron radiation far from a source is discussed. A new method is proposed for solving this problem by choosing from among the scattered neutrons the ones that have acquired the required direction of flight and energy. The material along which such a neutron will travel must possess a small neutron cross section at the given energy. Materials consisting of isotopes with a large resonance cross section, for example, ⁵⁶Fe, meet the requirements. The method discussed for concentrating high-energy neutrons makes it possible to increase their concentration more than 1000-fold in the detection area far from the source compared with isotropic propagation away from the source in all directions with equal probability.

The results of investigations of swelling and in-reactor creep of the ferrite-martensite class steels EP-450, EP-823, and EI-852 irradiated in a BN-350 reactor at 305–700°C to damaging dose in the range 20–89 dpa are presented. The characteristics of radiation creep of three of the ferrite-martensite steels are close despite the differences of the chemical composition and heat-treatment. The relation B = 4.864·10^–2 + 1.45·10^–3T describes the average modulus of radiation creep of the experimental steel in the interval 305–550°C.

The impact of research building B at the Bochvar All-Russia Research Institute for Inorganic Materials (VNIINM) on the subsoil (geological medium) is evaluated. It is shown that there is no radioactive contamination of the soils or underground water. The test results are used to substantiate the radiation safety of decommissioning the building. The network of observational wells that was created as part of engineering and environmental research is intended for monitoring the status of the underground water at the site of VNIINM during and after the decommissioning period for building B.

The mathematical modeling of nuclear fuel burnup is based on the solution of a system of ordinary differential equations of the Bateman type. The particularities of the systems obtained are examined. Examples of possible computational inaccuracies of the ORIGEN2 code are presented. New approaches to the solution of isotopic kinetics problems with a complete basis of elements in the fission product yield, including short-lived elements, are examined.

Methods for trapping the ¹⁴C in waste gases by means of a combined technology are examined. The wastes gases from pyrochemical production can be purified by removing ¹⁴C by passing the gases at rates to 5 m³/h through a 0.5 m high layer of Ba(OH)₂·8H₂O granules in a 0.14 m in diameter apparatus after dilution with atmospheric air. The purification of the waste gases from hydrometallurgical production by removal of ¹⁴C (50 m³/h) is best accomplished by passage through a 0.28 m in diameter packed tower with 2.7–4 m high packing irrigated by an alkali solution which is regenerated by means of calcium hydroxide. The waste gases from pyrochemical production can also be passed through a packed tower. In this case the ¹⁴C is fixed in insoluble calcium carbonate and is placed in a storage facility and subsequently buried. The proposed technologies make it possible to purify the waste gases formed during the reprocessing of spent uraniumplutonium nitride nuclear fuel in a reprocessing module used in an on-site nuclear fuel cycle by removing ¹⁴C to a level admissible for air in production enclosures.