Particularities of spatial kinetics of hybrid thorium reactor installation containing the long neutron source based on magnetic trap

In this work, we study the features of the spatial kinetics of installation as a hybrid thorium reactor with an elongated plasma neutron source based on a magnetic trap. The active zone of the installation under study consists of an assembly of hexagonal fuel blocks of a unified design and a long solenoid with a high-temperature plasma column passing through the axial region of the core. Combining engineering expertise in creating nuclear reactors with a physics-technical potential for obtaining high-temperature plasma in a long magnetic trap we ensure the solution of the multidisciplinary problem posed. These studies are of undoubted practical interest, since they are necessary to substantiate the safety of operation of such hybrid systems. The research results will allow optimizing the active zone of the hybrid system with leveling the resulting offset radial and axial energy release distributions. Results of our study will be the basis for the development of new and improvement of existing methods of criticality control in related systems such as "pulsed neutron source - subcritical fuel assembly"

Particularities of spatial kinetics of hybrid thorium reactor installation containing the long neutron source based on magnetic trap

In this work, we study the features of the spatial kinetics of installation as a hybrid thorium reactor with an elongated plasma neutron source based on a magnetic trap. The active zone of the installation under study consists of an assembly of hexagonal fuel blocks of a unified design and a long solenoid with a high-temperature plasma column passing through the axial region of the core. Combining engineering expertise in creating nuclear reactors with a physics-technical potential for obtaining high-temperature plasma in a long magnetic trap we ensure the solution of the multidisciplinary problem posed. These studies are of undoubted practical interest, since they are necessary to substantiate the safety of operation of such hybrid systems. The research results will allow optimizing the active zone of the hybrid system with leveling the resulting offset radial and axial energy release distributions. Results of our study will be the basis for the development of new and improvement of existing methods of criticality control in related systems such as "pulsed neutron source - subcritical fuel assembly"

ПОЛУЧЕНИЕ КОМПЛЕКСНЫХ ПРЕПАРАТОВ НА ОСНОВЕ ЛИПОСОМАЛЬНОЙ ФОРМЫ СТРЕПТОКИНАЗЫ И ИХ ФАРМАКОКИНЕТИЧЕСКИЕ ХАРАКТЕРИСТИКИ

It was obtained that liposomal streptokinase (SK) with a hydrodynamic diameter of ~70 nm and a zeta-potential of –6.2 mV contains 14.1 % wt of drug. The complex formulations based on liposomal SK include “associated” and “free” SK in the ratios of 20/80, 40/60 and 50/50. The in vivo experiments on rats showed an increase in the elimination half-time from 1.8 to 31.9 min and in the time to reach the maximum concentration of streptokinase from 15 to 45 min. The decrease in the elimination rate constant by a factor of 18 compared with SK was also found. The optimal ratio of “associated” and “free” SK in the complex formulation was 40 and 60 % respectively. It was used to obtain liposomal fibrin-specific form of thrombolytic with similar physico-chemical and pharmacokinetic parameters. Получена липосомальная форма стрептокиназы (СК) с гидродинамическим диаметром ~70 нм, дзета-потенциалом –6,2 мВ и степенью включения вещества 14,1 %, на основе которой приготовлены комплексные препараты, содержащие «связанную» и «свободную» СК в соотношениях 20/80, 40/60 и 50/50. Для них в эксперименте in vivo на крысах показано увеличение периода полувыведения от 1,8 до 31,9 мин и времени достижения максимальной концентрации стрептокиназы от 15 до 45 мин, а также уменьшение константы элиминации в ~18 раз по сравнению с нативной СК. Оптимальное соотношение «связанной» и «свободной» СК в комплексном препарате составило 40 и 60 % соответственно, что было использовано для получения липосомальной фибрин-специфичной формы тромболитика, который обладает практически такими же физико-химическими и фармакокинетическими параметрами.

Solution of neutron-transport multigroup equations system in subcritical systems

An iteration method has been implemented to solve a neutron transport equation in a multigroup diffusion approximation. A thermoelectric generator containing plutonium dioxide, used as a source of thermal and electric power in spacecraft, was studied. Neutron yield and multigroup diffusion approximation data was used to obtain a continuous and group distribution of neutron flux density spectra in a subcritical multiplying system. Numerical multigroup approaches were employed using BNAB-78, a system of group constants, and other available evaluated nuclear data libraries (ROSFOND, BROND, BNAB, EXFOR and ENDSF). The functions of neutron distribution in the zero iteration for the system of multigroup equations were obtained by approximating an extensive list of calculated and experimental data offered by the EXFOR and ENDSF nuclear data libraries. The required neutronic functionals were obtained by solving a neutron transport equation in a 28-group diffusion approximation. The calculated data was verified. The approach used is more efficient in terms of computational efforts (the values of the neutron flux density fractions converge in the third iteration). The implemented technique can be used in nuclear and radiation safety problems

Peculiarities of the radiation formation in dispersed microencapsulated nuclear fuel

A computational study has been performed for various options of the thorium reactor core loading. Neutronic studies of fuel have been conducted, its isotopic composition has been calculated, and the alpha emitters and the sources of neutron and photon radiation in the microencapsulated nuclear fuel have been analyzed. The studies had the purpose of developing the methodology used to estimate the radiation characteristics of nuclear fuel with a complex inner structure. Emphasis is placed on calculating the quantitative and spectral composition of the neutrons formed as the result of (a, n) reactions on small- and average-mass nuclei. The ratio of the quantity of the neutrons resulting from the (a, n) reactions to the quantity of the neutrons formed as the result of spontaneous fission has been calculated for fuel with heterogeneous and homogeneous arrangements of fissionable and structural elements. The developed tools will make it possible to estimate the neutron radiation dose, to revise the traditional fresh and spent fuel handling procedures, and to estimate, using the Rossi alpha method, the neutron multiplication factor in deeply subcritical systems. The neutron yield and spectrum were calculated using an analytical model and verified codes such as WIMS-D5B, ORIGEN-APP, SOURCES-4C and SRIM-2013

Power density dynamics in a nuclear reactor with an extended in-core pulse-periodic neutron source based on a magnetic trap

The article examines the features of the spatial kinetics of an innovative hybrid nuclear power facility with an extended neutron source based on a magnetic trap. The fusion-fission facility under study includes a reactor plant, the core of which consists of an assembly of thorium-plutonium fuel blocks of the HGTRU reactor of a unified design and a long magnetic trap that penetrates the near-axial region of the core. The engineering solution for the neutron plasma generator is based on an operating gas-dynamic trap based on a fusion neutron source (GDT-FNS) developed at the Novosibirsk G.I. Budker Nuclear Physics Institute of the Siberian Branch of the Russian Academy of Sciences. The GDT-FNS high-temperature plasma pinch is formed in pulse-periodic mode in the investigated hybrid facility configuration, and, at a certain pulse rate, one should expect the formation of a fission wave that diverges from the axial part of the system and propagates throughout the fuel block assembly in a time correlation with the fast D-D neutron pulse source. In these conditions, it is essential to study the fission wave propagation process and, accordingly, the power density distribution formation within the facility blanket. The paper presents the results of a study on the steady-state and space-time performances of neutron fluxes and the power density dynamics in the facility under investigation. The steady-state neutronic performance and the space-time fission wave propagation were simulated using the PRIZMA software package developed at FSUE RFNC-VNIITF

Fusion-fission hybrid reactor facility: neutronic research

The authors investigate the neutronic characteristics of the operating mode of a hybrid nuclear-thermonuclear reactor. The facility under study consists of a modified core of a high-temperature gas-cooled thorium reactor and an extended plasma neutron source penetrating the near-axial region of the core. The proposed facility has a generated power that is convenient for the regional level (60–100 MW), acceptable geometric dimensions and a low level of radioactive waste. The paper demonstrates optimization neutronic studies, the purpose of which is to level the resulting offsets of the radial energy release field, which are formed within the fuel part of the blanket during long-term operation and due to the pulsed operation of the plasma D-T neutron source. The calculations were performed using both previously developed models and the SERPENT 2.1.31 precision program code based on the Monte Carlo method. In the simulation, we used pointwise evaluated nuclear data converted from the ENDF-B/VII.1 library, as well as additional data for neutron scattering in graphite from ENDF-B/VII.0, based on the S (α, β) formalism

Fusion-fission hybrid reactor facility: neutronic research

The authors investigate the neutronic characteristics of the operating mode of a hybrid nuclear-thermonuclear reactor. The facility under study consists of a modified core of a high-temperature gas-cooled thorium reactor and an extended plasma neutron source penetrating the near-axial region of the core. The proposed facility has a generated power that is convenient for the regional level (60–100 MW), acceptable geometric dimensions and a low level of radioactive waste. The paper demonstrates optimization neutronic studies, the purpose of which is to level the resulting offsets of the radial energy release field, which are formed within the fuel part of the blanket during long-term operation and due to the pulsed operation of the plasma D-T neutron source. The calculations were performed using both previously developed models and the SERPENT 2.1.31 precision program code based on the Monte Carlo method. In the simulation, we used pointwise evaluated nuclear data converted from the ENDF-B/VII.1 library, as well as additional data for neutron scattering in graphite from ENDF-B/VII.0, based on the S (α, β) formalism