The possibility of using the Monte Carlo method for high-speed calculations of the fast-neutron fluence on a VVER-1200 vessel during some time interval is substantiated. In the model developed, the flow attenuation behind the core is described using the MCU-PD code in a computational regime at low power with an external source set alternately in the FA at the periphery of the core. The distribution of a time-varying source of fast neutron over the core is determined, in the diffusion approximation by means of the DYN3D engineering code, for actual reactor operation. The source irregularity along the FA profile is described in the x and y directions in the P-1 approximation.

A study is performed of the theory of coherent states’ applicability to solving statistical problems in the reactor physics. A mathematical description of the theory is provided in the formalism of both overlap integrals and operators of the creation and annihilation of phonons. The Poisson distribution is compared to coherent states, and the applicability of the formalism of the theory of coherent states to the physics of reactors is analyzed.

244Cm is a very important nuclide due to the high specific heat release and it may become reasonable to recycle the spent nuclear fuel to extract this isotope. In this work, the dynamics of 244Cm isotope production for the fuel assembly of the WWER-1200 reactor with an enrichment of 4.95% and burnup up to 70 MW · day/kg was calculated using Serpent Monte Carlo code. A similar calculations was performed for MOX-based fuel assembly with the same irradiation characteristics. It has been shown that production of 244Cm in high burnup uranium dioxide fuel, and especially in MOX fuel, reaches a high value, which can cause problems during spent fuel management. The main problems associated with the curium relates to the residual heat and neutron activity.

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction β for 239Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

Current work is aimed at the analysis of the fission products decay influence during fuel reloading, when calculating the accumulated fission products activity for the VVER-1200 reactor fuel campaign. The Bateman problem solution based technique was used for calculations, within the framework of the two fissile nuclides approximation. The fission products producing process for the VVER-1200 reactor stationary campaign is considered, taking into account the reactor shutdown periods for refueling and without taking them into account (instant reload approximation). It was shown, that the instant reload approximation for fission products activity calculations gives the similar accurate result, as calculations with taking into account the shutdown periods. The results can be used to significantly simplify the calculations of fission product activity accumulation in nuclear power reactors.

To study the kinetics of subcritical systems and determine the optimal conditions for the transmutation of longlived radioactive waste in the neutron spectrum of ADS-systems the “Yalina” research nuclear facility was created at Joint Institute for Power and Nuclear Research – Sosny (Minsk, Belarus). The main safety indicator of a subcritical system (active zone reactivity) was measured for a “Yalina-Thermal” assembly via three independent methods: inverse multiplication, probabilistic and impulse ones. For the inverse multiplication method, the neutron flux density was monitored during assembly loading. For a fuel load of 285 EK-10 rods the neutron multiplication was M = 22.3±0.6, and the effective neutron multiplication coefficient was keff = 0.9551± 0.0016. The probabilistic method (Feynman-alpha method), based on measuring fluctuations in the neutron density level within a system with a fission chain reaction, gave the ratio of the variance to the average counting rate value D/n = 1.779±0.005, which corresponds to keff = 0.9597 ±0.0003. The pulse method is aimed at studying the neutron flux behavior of after the neutron pulse injection into the breeding system. Measurements were held with the same setup, used in the Feynman-alpha method. The measured decay constant of instantaneous neutrons is α = –670±0.7 1/s, which corresponds to keff = 0.9560±0.0001. The effective multiplication factor keff of the subcritical assembly “Yalina-Thermal”, obtained via three different independent methods, is around average value of keff = 0.9569 ± 0.0018. The methods considered can be used for subcritical level monitoring for ADS-systems and research nuclear facilities.

Present work describes Monte-Carlo calculations of the neutron field and minor actinide transmutation reaction rates within the Yalina-Thermal sub-critical assembly of the Joint Institute for Power and Nuclear Research – Sosny of the National Academy of Sciences of Belarus. The computer model of the facility was prepared for the corresponding calculations via MCU-PD and MCNP Monte-Carlo codes. The model neutron characteristics estimations were performed as well as the nuclear safety analysis. The up-to-date ENDF B/VIII, JEFF 3.3 and JENDL 4.0 nuclear data libraries were used during research.

The Republic of Belarus is a participant in global efforts for nuclear non-proliferation, ensuring the preservation of nuclear materials and countering nuclear terrorism. In accordance with the Decree of the President of the Republic of Belarus No. 124 dated March 29, 2011, the Joint Institute for Power and Nuclear Research – Sosny of National Academy of Science of Belarus is the operator (operating organization) of a number of critical and subcritical nuclear facilities, radiation installations and fresh nuclear fuel storage, containing highly enriched uranium fuel. A nuclear security culture is an important necessary constituent of the nuclear and radiation security. The authors of this paper will report on the approach the Institute is taking in regard to advancing nuclear security program, what are the Institute’s goals and objectives, and what steps are being taken to support effective the physical protection system at the Scientific Institution “JIPNR - Sosny” and another operating organization in the Republic of Belarus. The range of measures aimed at reducing the likelihood of internal violators activity, early detection of potential internal violators and preventing from causing harm are presented.

Current experimental and computational studies of the Yalina subcritical setup are reviewed. The main characteristics of the Yalina Thermal facility and its neutron physics parameters are given. Computer modeling of the experimental setup is done using the Monte Carlo MCU-PD and MCNP codes for neutron physics and the updated ENDF/b-VIII, JENDL 4.0, and JEFF 3.3 nuclear data libraries. Rates of the transmutation reaction for 129 I, 237 Np, and 241 Am in experimental channels are calculated.

Accelerator-driven systems (ADS) are considered to be promising for energy generation and transmutation of the long-live radiotoxic nuclides contained in spent nuclear fuel. The Yalina assembly has the same conceptual construction with a neutron generator coupled to a subcritical core. The major difference between ADS and Yalina experiments is that the neutron energy spectrum of the Yalina core is thermal. The simulation of the Yalina-Thermal experimental program was conducted via new neutron-physics code SuperMC developed by the FDS Team in China. SuperMC is a CAD-based Monte Carlo program for integrated simulation of nuclear systems by making use of hybrid Monte Carlo and deterministic methods and advanced computer technologies. The calculations were held with the use of the up to date nuclear data libraries ENDF/B-VII.1, ENDF/B-VIII, FENDL 3.0, JEFF 3.3, JENDL 4.0. The effective multiplication constant, k eff , and the fraction of delayed neutrons, β eff , were calculated for the 285 fuel rod core configuration of the Yalina-Thermal subcritical assembly. Evaluation of the Yalina-Thermal experimental program using SuperMC code has been made for the first time in this work.

Results are presented from calculating the neutron radiation of spent nuclear fuel (SNF) from stationary campaigns of VVER-1000 and VVER-1200 reactors due to spontaneous fission and (α, n) reactions. An analytical dependence is obtained for the average number of neutrons produced in spent nuclear fuel per α-particle on the particle’s energy. The neutron radiation of VVER-1200 spent nuclear fuel is several times higher than that of a VVER-1000 reactor, due primarily to the increased number of 244Cm isotopes produced in high-burnup fuel.

Due to the effective development of ion-track technology, it became possible to produce porous templates with large areas, which are of interest for mass production of nanostructures. Given that the template parameters often define properties of the resulting nanostructures and nanosystems, a reliable method for non-destructive testing is needed for a rapid control of template parameters. Such method could be ellipsometry, allowing for a single measurement to collect statistical information from a large area and to save time for certification. In order to adapt the ellipsometry method for controlling the parameters of ion-track patterns, the first studies of SiO2/Si templates with low porosity were carried out. Using the standard model of the interaction of a polarized light beam with a layered structure of silicon oxide on silicon, the basic parameters of the pores were determined by means of mathematical transformations and subsequently compared with the results of scanning electron microscopy.

The functional form of nuclei decay for a finite number of particles within sub-Poisson distribution terms and time scale discretisation were considered. The mathematical apparatus for particles’ birth and death model with respect to the theory of a nuclear thermal point-reactor was made. During research the fundamental curves, the exponential dependence of neutron generation lifetime and state decay possibility were obtained for a thermal reactor breeding medium. Main parameter of a breeding (neutron multiplication factor K) was obtained through mathematical functions considered.

Current work presents general principles for constructing a mathematical apparatus within the framework of a sub-Poisson distribution that describes the neutron generations emergence. On the basis of these principles, analytical expressions were obtained for describing the point reactor breeding medium main parameters. The corresponding dependencies for generation lifetime and appearance time were formed.

In the presented work the residual heat of spent fuel from WWER-1000 and WWER-1200 reactors after preliminary storage in spent fuel pool is investigated. The selection of nuclides with a significant contribution to the residual heat was made. Their absolute and relative contributions depending on time are considered. Comparison of residual heat generation of spent fuel from a WWER-1000 reactor with a burnup of 40 MW . day / kg and a WWER-1200 reactor with a burnout of 70 MW . day / kg after storage in the spent fuel pool was made.