Vol 123, No 6 (2018)

A special version of the IR code is described. The purpose of this code is to generate recommendations for the reactor operator when operating in a maneuvering regime, automate boron regulation, and study the maneuvering characteristics of VVER-1000 fuel loads. The results of modeling of operation in a daily load schedule are presented for the example of a stationary run of VVER-TOI. The impact of variation of the parameters of power maneuvers (run, operating time a reduced power, loading rate) on the boron regulation parameters is analyzed.

A scenario is examined for the development of a beyond design basis accident with loss-of-power for internal needs and functional failures/delays in the actuation of mobile means for emergency cooling RBMK in application to the No. 3 unit of the Kursk NPP. It is shown that actions taken by operating personnel to control pressure in the circulation loop, primarily with the aid of the valves of fast reduction setup for dumping steam in the turbine condensers are effective. The sequence of these actions makes it possible to prevent collapse of the channel tubing at high pressure, preserve the integrity of the reactor space, and effectively limit the accidental emission of radionuclides into the environment.

The accident in the No. 4 unit of the Chernobyl NPP destroyed the reactor and enclosures and resulted in atmospheric emissions of fission products, irradiated fuel, moderator, and other radioactive materials. The large-scale destruction could have been due to the chemical interaction of the coolant (water) and moderator (graphite) with carbon monoxide and hydrogen forming at temperatures above 1000 K. The results of calculations of the composition of a thermodynamically equilibrium mixture in the system of elements C–H–O and measurements of the characteristic graphite-water interaction time at ~1500 K are presented and this time is compared with the characteristic development time of the accident at the Chernobyl NPP.

The first reactor tests and post-reactor studies of fuel elements with cladding made from the zirconium alloys E-636M, E-635opt, and E-635M1, developed at VNIINM, have been conducted. The purpose of the studies was to search for alloys of the type E-635 that surpass the base alloy in terms of resistance to corrosion and hydration. The studies make it possible to recommend the alloys E-635M and E-635opt for use as cladding of the fuel elements in extended-life cores.

A Monte Carlo method implemented in the PRIZMA-DSP code to calculate multiplying systems, including weakly coupled systems, is described. In the code, the basis of the method is a sequential calculation of a prescribed chain of active generations, which is described by, first and foremost, the dominant ratio of the system. The fission points of the running generation serve as a source for constructing neutron trajectories, new fission points arising on a trajectory are transmitted unchanged to calculate the next generation, and so on. In the calculation of the active generation, K_eff and made-to-order linear-fractional functions are evaluated. In order for the distribution to become established, a passive calculation with prescribed chain length is made before the active calculation. In addition, after the calculation of each generation the operation of random mixing of the entire set of fission points and their distribution over the nuclei is performed in order to obtain a uniform distribution. At the start of the calculation, a special rejection procedure performs a pre-set number of preliminary computational iterations so that an arbitrary zeroth distribution over all nuclei goes to a distribution obtained via a characteristic function.

The first author’s name should read A. K. Valter.