Nuclear Energy and Technology

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List of Papers (Total 26)

Verification on application program generation and loading for safety systems of nuclear power plants based on the reverse engineering method

The article describes an automated verification method used for application software of control safety systems based on the TPTS-SB equipment. Verification is performed by comparing two mathematical models (oriented graphs): one obtained by processing the original design data, i.e., graphical functional diagrams, and the other formed by reversing the program code loaded from the...

Comparison of spallation reaction models based on multiple-criteria decision analysis

The paper presents the results of a comparative evaluation of the predictive ability of seventeen spallation reaction models (CEM02, CEM03, Phits/jam, Cascade/ASF, Phits/Bertini, Bertini/Dresner, Cascade-4, INCL4/Abla, INCL4/smm, geant4/binary, Isabela/smm, geant4/Bertini, Isabela/Abla, INCL4/Gemini, CASCADeX-1.2, Isabel/Gemini, Phits/jqmd) for the interaction reactions of high...

Comparison of two key analysis methods for the seismic stability of equipment on the example of a ventilation unit

Results of calculation seismic resistance analysis of light equipment of nuclear power plants performed on the example of a ventilation unit using two most common analytical techniques - linear spectral analysis and direct dynamic methods - are discussed. The basic concepts, assumptions and limitations of the linear spectral method are described. Examples are given of specific...

Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA...

On the necessity and the role of descriptors of neutron activated structural and shielding materials of nuclear installations for future decommissioning

Existing situation in nuclear industry is characterized with simultaneous development of the following two processes: design and construction of new generation of nuclear installations and decommissioning of installations of older generations. Significant amounts of radioactive wastes generated during the decommissioning phase are determined both for the first and the second...

A technique for detection of WWER fuel failures by activity of Xe radionuclides during reactor operation

Fuel failures during operation of Nuclear Power Plants (NPPs) may lead to substantial economic losses. Negative effects of reactor operation with leaking fuel in the core may be reduced if fuel failures are detected in due time of the cycle. At present time, the ratio of the normalized release rates of 131I and 134I is used to detect fuel failures in WWERs during steady state...

The SIMCO containment code applied to modeling hydrogen distribution in containments of nuclear power facilities

The article gives a general description of the SIMCO calculation code designed to simulate thermohydraulic and physico-chemical processes in containments of nuclear power facilities. The authors present a calculation technique based on a physico-mathematical model in lumped parameters. As a numerical solution method, the modified semi-implicit SIMPLER procedure is used. The code...

Simulate the effect of integral burnable absorber on the neutronic characteristics of a PWR assembly

This article examines the effect of an integral burnable absorber (IBA) on the neutronic characteristics of Pressurized Water Reactor (PWR) to provide possible improvements for the fuel management. MCNPX code was used to design a three dimensional model for PWR assembly. The designed model has been validated by comparing the output data with a previously published data. MCNPX...

Estimation influence of boric acid drop entrainment to its accumulation in the VVER reactor in the case of accident

Process of boric acid mass transfer during accidents accompanied with rupture of circulation pipelines in VVER reactors of new generation equipped with passive safety systems are examined. Results of calculation of variation of boric acid concentration in VVER-TOI reactor in case of accident development process are presented. Positive effects of boric acid droplet entrainment on...

Development of the technique for determination the rate of oxidation of structural steels in heavy liquid metal coolants

The article considers the main methods for studying the process of oxidation of structural steels and evaluating their corrosion resistance in heavy liquid metal coolants (HLMC) under static and dynamic conditions. It is shown that the main disadvantage of these methods is the impossibility of evaluating the results in real time. The authors propose a new method for the...

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...

Measurement of the spent fuel rod cladding temperature during the in-pile testing at 500–900°C

This paper deals with the problem of measuring the VVER-1000 burnup fuel cladding temperature in a 500–900°C range in the process of experiments in a channel of the MIR research reactor to obtain data on the fuel element behavior under the influence of the parameters typical of the maximum design-basis loss-of-coolant accident (LOCA). Studying the burnup fuel cladding deformation...

Elaboration of approach to nuclear energy systems assessment by criterion of sustainable development

The paper describes the approach to the assessment of nuclear energy systems based on the integral indicator characterizing the level of their sustainability and results of comparative assessment of several nuclear energy system options incorporating different combinations of nuclear reactors and nuclear fuel cycle facilities. The nuclear energy systems are characterized by...

Flow-accelerated corrosion rate and residual life time estimation for the components of pipeline systems at nuclear power plants based on control data

As of today, large volumes of data related to non-destructive operational control are accumulated on NPPs. For ensuring safe operation of power units, optimization of scope and scheduling operational control it is necessary to continue development of guidance documents, software products, methodological guidance and operational documentation (Baranenko et al. 1998, Gulina et al...

About chemical form and binding energy of 14C in irradiated graphite of uranium-graphite nuclear reactors

Issues associated with handling irradiated graphite of uranium-graphite nuclear reactors are examined. It is demonstrated that selection of approaches, methods and means for handling irradiated graphite are determined by the form of occurrence and binding energy of long-lived 14C radionuclide with graphite crystalline lattice. The purpose of the present study is the determination...

Regulation of the temperature in the ampoule channel with natural circulation of coolant

It has been shown by calculations that it is possible to extend considerably the capabilities for control of temperature conditions in an ampoule channel with natural coolant circulation, using the proposed hydraulic circuit layout, on samples during irradiation in the SM-3 reactor reflector cell by changing the circulation circuit geometry through the arrangement of a bypass...

Minimize fission power peaking factor in radial direction of water-cooled and water-moderated thermionic conversion reactor core

The paper investigates the possibility for reducing the radial power peaking factor kr inside the core of a water-cooled water-moderated thermionic converter reactor (TCR). Due to a highly nonuniform power density, the TCR generates less electric power and the temperature increases in components of the thermionic fuel elements, leading so to a shorter reactor life. A TCR with an...

Brittle fracture resistance of reactor pressure vessel steels in the initial state

The authors investigate the influence of chemical and structural inhomogeneity on the brittle fracture resistance (BFR) of VVER vessel materials in the initial state (without irradiation). It is proposed to replace the brittle fracture resistance assessment using the critical brittleness temperature TC for the BFR assessment using the brittle-viscous transition temperature TT...

Radiation-induced separation and accumulation of electric charge in supercapacitors

In current sources with a radioactive isotope (CSRI), nuclear energy is directly converted into electricity due to the separation of electric charges during the decay of radioactive isotopes. It was previously shown that asymmetric supercapacitors can be used as CSRI prototypes if, after being exposed to pulsed reactor irradiation, the electric charge on their plates increases to...

Axial dispersion and mixing of coolant gas within a separate-effect prismatic modular reactor

Multiphase Reactors Engineering and Applications Laboratory performed gas phase dispersion experiments in a separate-effect cold-flow experimental setup for coolant flow within heated channels of the prismatic modular reactor under accident scenario using gaseous tracer technique. The separate-effect experimental setup was designed on light of local velocity measurements obtained...

A modern data measurement system to study and test thermionic heat to electricity converters

Studies and tests are conducted to determine the performance of thermionic nuclear power plants (TNPP) a stage in which is pre-irradiation testing of laboratory thermionic converters (TIC) with flat and cylindrically shaped electrodes using test facilities fitted with automated data measurement systems (DMS). The TIC volt-ampere characteristics (VAC) are measured in the DMS...

Development and study of a microwave reflex-radar level gauge of the nuclear reactor coolant

The article considers the design of a microwave reflex-radar level gauge of the nuclear reactor coolant. The main advantage of the reflex-radar measurement principle is that it does not affect the accuracy of measuring the level of bubbles present, coolant condensation and boiling, changes in its pressure as well as temperature and density. In addition, the measuring transmitter...

Quantitative evaluation of the plutonium proliferation resistance

The mathematical model presented in (Kulikov et al. 2018) can be used for the quantitative evaluation of the plutonium proliferation resistance. This requires the warm-up process of an implosion nuclear explosive device (NED) with a different structure to be analyzed with respect to various heat removal conditions and the option to be identified in which the NED remains...

Reactivity margin evaluation software for WWR-c reactor

The WWR-c reactor reactivity margin can be calculated using a precision reactor model. The precision model based on the Monte Carlo method (Kolesov et al. 2011) is not well suited for operational calculations. The article describes the work on creating a software package for preliminary evaluations of the WWR-c reactor reactivity margin. The research has confirmed the possibility...

Application of diffusion approximation in the calculations of reactor with cavities

The importance of calculation of radiation fields inside in-reactor cavities is associated with the necessity to simulate the emergency modes in fast breeder reactors (FBR), as well as reactor states with different coolant levels in special dedicated channels of passive feedback devices in lead-cooled fast reactors (LFR) of BREST type or in sodium cavities in sodium-cooled fast...