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"

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

Dynamics of population of gadolinium-156 nuclei energy levels during neutron pumping of isotope-modified gadolinium oxide

The possibility of transformation of energy of fast and epithermal neutrons to energy of coherent photon radiation at the expense of a neutron pumping of the active medium formed by nucleus with long-living isomerous states is theoretically described. The channel of the nucleus formation in isomeric state as a daughter nucleus resulting from the nuclear reaction of neutron capture by a lighter nucleus is taken into consideration for the first time. The analysis of cross sections dependence of radiative neutron capture by the nuclei of gadolinium isotopes Gd155 and Gd156 is performed. As a result, it is stated that the speed of Gd156 nuclei formation exceeds the speed of their “burnup” in the neutron flux. It is provided by a unique combination of absorbing properties of two isotopes of gadolinium Gd155 and Gd156 in both thermal and resonance regions of neutron energy. The possibility of excess energy accumulation in the participating medium created by the nuclei of the pair of gadolinium isotopes Gd155 and Gd156 due to formation and storage of nuclei in isomeric state at radiative neutron capture by the nuclei of the stable isotope with a smaller mass is shown. It is concluded that when the active medium created by gadolinium nuclei is pumped by neutrons with the flux density of the order of 1013 cm-2·s-1, the condition of levels population inversion can be achieved in a few tens of seconds. The wave length of the radiation generated by the medium is 0.0006 nm

Subcriticality control elements in a reactor system with an extended plasma source of neutrons with regard for temperature

Materials have been selected for the shim rods and burnable absorbers to compensate for the excessive reactivity of the facility’s blanket part and to provide for the possibility of reactivity control in conjunction with a plasma source of neutrons. Burnable absorber is a layer of zirconium diboride (ZrB2) with a thickness of 100 μm applied to the surface of fuel compacts. Boron carbide (B4C) rods installed in the helium flow channels and used to bring the entire system into a state with keff = 0.95 have been selected as the shim rod material. Throughout its operating cycle, the facility is subcritical and is controlled using the neutron flux from the plasma source. Verified codes, WIMS-D5B (ENDF/B-VII.0) and MCU5TPU (MCUDВ50), as well as a modern system of constants were used for the calculations. The facility’s neutronic performance was simulated with regard for the changes in the inner structure and temperature of the microencapsulated fuel and fuel compact materials caused by long-term irradiation and by the migration of fission fragments and gaseous chemical compounds

Conceptual approaches and methods of treating the irradiated potential nuclear fuel

Various types of reactors are applied in the world nuclear power industry. One of the ways to implement the existing trend to increase the effectiveness of using nuclear fuel in nuclear power engineering is the growth of its burn-up fraction. The relevance of the research is caused by the problem related to increase of burn-up fraction depth both of normal nuclear fuel and new types of fuel, as well as by the necessity to develop the conceptually new approaches to handling such fuel in storage systems and transport means. The main aim of the research is to optimize the system parameters and schemes of handling in «dry» storage of spent fuel irradiated in uranium-graphite reactor of channel type. The methods. The research and numerical experiments were carried out with the assistance of verified computer codes programs based on the Monte Carlo method (MCU5TPU and Scale5), modern evaluated nuclear data library (ENDF/B-VIII, JENDL-3.3, JEFF 3.0, EXFOR, ROSFOND) and multi-group approximation. The sharing of the precise calculation program code of MCU and Scale allowed carrying out the verification of the obtained results of numerical experiments. The results. The authors have carried out the computational studies of the neutron-physical characteristics of the system of «dry» storage of spent nuclear fuel irradiated in the uranium-graphite reactor. Practical recommendations for optimizing the system parameters and schemes of handling and placement of spent fuel in a «dry» storage were developed. The system parameters and handling schemes in the process of «dry» storage of spent fuel optimized due to the alternating layers of placing fuel with different burn-up and enrichment. multiplying syste

Conceptual approaches and methods of treating the irradiated potential nuclear fuel

Various types of reactors are applied in the world nuclear power industry. One of the ways to implement the existing trend to increase the effectiveness of using nuclear fuel in nuclear power engineering is the growth of its burn-up fraction. The relevance of the research is caused by the problem related to increase of burn-up fraction depth both of normal nuclear fuel and new types of fuel, as well as by the necessity to develop the conceptually new approaches to handling such fuel in storage systems and transport means. The main aim of the research is to optimize the system parameters and schemes of handling in «dry» storage of spent fuel irradiated in uranium-graphite reactor of channel type. The methods. The research and numerical experiments were carried out with the assistance of verified computer codes programs based on the Monte Carlo method (MCU5TPU and Scale5), modern evaluated nuclear data library (ENDF/B-VIII, JENDL-3.3, JEFF 3.0, EXFOR, ROSFOND) and multi-group approximation. The sharing of the precise calculation program code of MCU and Scale allowed carrying out the verification of the obtained results of numerical experiments. The results. The authors have carried out the computational studies of the neutron-physical characteristics of the system of «dry» storage of spent nuclear fuel irradiated in the uranium-graphite reactor. Practical recommendations for optimizing the system parameters and schemes of handling and placement of spent fuel in a «dry» storage were developed. The system parameters and handling schemes in the process of «dry» storage of spent fuel optimized due to the alternating layers of placing fuel with different burn-up and enrichment. multiplying syste