A variable-geometry beam-shaping assembly for accelerator-based BNCT

Although the beam shaping assemblies (BSAs) for reactor-based Boron Neutron Capture Therapy (BNCT) facilities are typically of a single design, the accelerator beams with the possibility to provide neutron spectrum to give characteristics which are optimum for different treatment sites and tumor depth generally may require a fine-tuning procedure which can be undertaken with variable-geometry BSA. In this study, a special geometry is proposed for use with a hybrid photoneutron source equipped with drill-chuck type head. Both the neutron spectrum and epithermal neutron flux can be treated by changing the BSA geometry

Development of thorium-containing nuclear fuel cycle

The results providing advantages of thorium-232 as a reproducing nuclide in comparison with uranium-238 as a part of nuclear fuel of new generation reactors are presented. The explanation of the effects which were found earlier in numerical simulation of parameters of open thorium - plutonium nuclear fuel cycle is offered. Scientific and technical solutions allow considering the possibility of including thorium-232 in the fuel of nuclear reactors, which are based on existing design solutions, and beginning the design of new generation materials: a new generation of fuel rods and fuel assemblies, where the isotope uranium-238 will be completely replaced with thorium-232

Влияние распределения гадолиний содержащих твэлов реактора ВВЭР-1000 на выгорание

Различный состав и пространственное распределение выгорающих поглотителей по-разному влияют на размножающие и теплофизические свойства реактора. В работе создана расчетная 3D модель ТВС реактора с выгорающими поглотителями. В качестве выгорающих поглотителей используются традиционно-применяемые Gd2O3-содержащие твэлы и твэлы с AmO2. Проведенные расчеты позволили исследовать нейтронно-физические свойства реактора и нивелировать, возникающие офсеты энерговыделения в переходных режимах эксплуатации

Analysis constants for database of neutron nuclear data

At present there is a variety of experimental and calculation nuclear data which arerather entirely presented in the following evaluated nuclear data libraries: ENDF (USA), JEFF(Europe), JENDL (Japan), TENDL (Russian Federation), ROSFOND (Russian Federation).Libraries of nuclear data, used for neutron-physics calculations in programs: Scale (OrigenArp),MCNP, WIMS, MCU, and others. Nevertheless all existing nuclear data bases, includingevaluated ones, contain practically no information about threshold neutron reactions on {232}Thnuclei; available values of outputs and cross-sections significantly differ by orders. The workshows necessity of nuclear constants corrections which are used in the calculations of grids andthorium storage systems. The results of numerical experiments lattices and storage systemswith thorium

Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work

Particle emission from irradiated VVER-1200 fuel with Am burnable absorber

For long-term and safe operation of the reactor, nuclear fuel is modified by doping with various homogeneous compounds and heterogeneous inclusions. Such modified fuel has demonstrated satisfactory performance when irradiated at elevated temperatures and high burnup. However, the issues of radiation safety when handling modified fuel remain less studied. Elements of low and medium atomic mass are often targets for (α, n) reactions, so their use as alloying additives, as well as the use of α-emitting additives, can complicate the radiation situation at the stages of the nuclear fuel cycle. In this work, the neutron components of the radiation characteristics of UO2 with Gd2O3 and AmO2 additives were analyzed. Americium has been investigated as a possible alternative to gadolinium. In fuel containing AmO2, the neutron yield is higher compared to Am-free fuel and is formed mainly by (α, n) reactions in AmO2 in fresh fuel and spontaneous fission of 244Cm nuclides in spent fuel. The research was carried out with the aim of developing procedures and regulations for handling new fuel during its manufacture and after irradiation in the reactor. This work contributes to the study of the neutronic and radiation characteristics of Am-containing fuel, which has the potential for use in modern reactors. Calculations were performed using verified computational codes SOURCES-4C and WIMS-D5B