Subjects -> ENERGY (Total: 414 journals)
    - ELECTRICAL ENERGY (12 journals)
    - ENERGY (252 journals)
    - ENERGY: GENERAL (7 journals)
    - NUCLEAR ENERGY (40 journals)
    - PETROLEUM AND GAS (58 journals)
    - RENEWABLE ENERGY (45 journals)

NUCLEAR ENERGY (40 journals)

Showing 1 - 37 of 37 Journals sorted alphabetically
Atom Indonesia     Open Access  
Bulletin of the Atomic Scientists     Hybrid Journal   (Followers: 8)
CNL Nuclear Review     Partially Free  
Eksplorium : Buletin Pusat Pengembangan Bahan Galian Nuklir     Open Access  
EPJ Nuclear Sciences & Technologies     Open Access   (Followers: 3)
Fusion Science and Technology     Hybrid Journal   (Followers: 4)
Ganendra : Majalah IPTEK Nuklir     Open Access  
Hyperfine Interactions     Hybrid Journal   (Followers: 1)
IEEE Transactions on Sustainable Energy     Hybrid Journal   (Followers: 13)
International Journal of Advanced Nuclear Reactor Design and Technology     Open Access  
International Journal of Critical Infrastructure Protection     Hybrid Journal   (Followers: 4)
International Journal of Nuclear Energy Science and Engineering     Open Access   (Followers: 5)
International Journal of Nuclear Law     Hybrid Journal   (Followers: 3)
International Journal of Nuclear Safety and Security     Hybrid Journal   (Followers: 1)
International Journal of Nuclear Security     Open Access   (Followers: 1)
Journal of Nuclear Energy Science & Power Generation Technology     Hybrid Journal   (Followers: 2)
Journal of Nuclear Engineering & Technology     Full-text available via subscription   (Followers: 3)
Journal of Nuclear Science and Technology     Hybrid Journal   (Followers: 2)
Journal of Power Technologies     Open Access   (Followers: 6)
Journal of Radiation Research     Open Access   (Followers: 3)
Journal of the Physical Society of Japan     Hybrid Journal   (Followers: 2)
Kerntechnik     Full-text available via subscription  
Majalah Ilmiah Teknologi Elektro : Journal of Electrical Technology     Open Access   (Followers: 1)
Nano Energy     Open Access   (Followers: 11)
Nanomaterials and Energy     Hybrid Journal   (Followers: 2)
Nuclear Energy and Technology     Open Access   (Followers: 3)
Nuclear Engineering and Technology     Open Access   (Followers: 5)
Nuclear Materials and Energy     Open Access   (Followers: 1)
Nuclear Science and Engineering     Hybrid Journal   (Followers: 7)
Nuclear Science and Techniques     Full-text available via subscription  
Nuclear Technology     Hybrid Journal   (Followers: 5)
Nucleus     Open Access  
Nukleonika     Open Access  
Radiation Detection Technology and Methods     Hybrid Journal   (Followers: 1)
Tri Dasa Mega : Jurnal Teknologi Reaktor Nuklir     Open Access  
Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir     Open Access  
World Journal of Nuclear Science and Technology     Open Access   (Followers: 4)
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Nuclear Energy and Technology
Number of Followers: 3  

  This is an Open Access Journal Open Access journal
ISSN (Online) 2452-3038
Published by Pensoft Homepage  [58 journals]
  • Computational and experimental justification for increasing the
           performance of the regenerative heat exchanger in the steam generator
           blowdown system of the AES-2006 project (RU V-392M)

    • Abstract: Nuclear Energy and Technology 8(4): 297-302
      DOI : 10.3897/nucet.8.97652
      Authors : Sergey V. Yaurov, Andrey V. Borovoy, Andrey V. Yudin, Mikhail V. Bolgov, Aleksandr D. Danilov : The article discusses the design and operation modes of the regenerative heat exchanger (RHE) in the steam generator (SG) blowdown and drainage system (LCQ) at Novovoronezh NPP-II 1 and 2 (Project AES-2006). The results of mathematical modeling of the RHE operating modes are presented in order to identify the causes of its low efficiency. Based on the results of the commissioning of the SG blowdown and drainage system at NvNPP-II 1, as well as the thermohydraulic calculations of the RHE operating modes, the authors put forward assumptions regarding changes in the rerouting of the piping (Volnov et al. 2017, Yaurov et al. 2017). According to their proposals, the RHE piping was upgraded at NvNPP-II 2. The upgrading in the RHE piping was implemented first at NvNPP-II 2 at the stage of installing the systems and, after the expected result was confirmed, it was applied in April 2020 at NvNPP-II 1. In addition, the authors carried out a comparative analysis of the results of testing the thermohydraulic characteristics of RHEs of the blowdown and drainage system for NvNPP-II 1 (before upgrading, after upgrading in scheduled maintenance 2020) and NvNPP-II 2. These improvements made it possible to achieve more efficient operation of the RHE in the SG blowdown and drainage system and the system as a whole. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:15:14 +020
  • Nuclear energy system modelling application package: functional
           overview and examples

    • Abstract: Nuclear Energy and Technology 8(4): 289-295
      DOI : 10.3897/nucet.8.97651
      Authors : Andrei A. Andrianov, Ilya S. Kuptsov, Tatiana A. Osipova, Anastasia A. Spiridonova, Olga N. Andrianova, Tatiana V. Utianskaya : The paper provides a brief description of the functionality of the nuclear energy system modelling application package (NESAPP), including a list of modelling objects, key assumptions and areas of application. NESAPP consists of the following main modules: NUDAPS (a module for calculating thermal neutron cross-sections, resonance integrals and one-group neutron cross-sections, and associated uncertainties), NUCLEX (a module for calculating the evolution of the nuclide composition and characteristics of nuclear fuel in reactors and at the nuclear fuel cycle front-end and back-end steps), NUCAB (a module for adjusting isotopic composition and blending), FANES (a module for analysing material flows and integrating data in nuclear energy evolution scenarios), ECNES (a module for assessing economic performance metrics for the nuclear energy evolution scenarios). Each of the modules is a calculation tool that can be used as independent or integrated into the software for technical and economic modelling of nuclear energy systems. Various calculation models are implemented in the modules, allowing users to evaluate the methodological component of the calculation uncertainty in scenario modelling studies and the functionality for assessing the impact of initial data uncertainties on the resulting indicators. The authors also provide some examples of applying NESAPP. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:15:00 +020
  • Thermohydraulic studies of alkali liquid metal coolants for
           justification of nuclear power facilities

    • Abstract: Nuclear Energy and Technology 8(4): 281-288
      DOI : 10.3897/nucet.8.96568
      Authors : Yulia A. Kuzina, Aleksandr P. Sorokin, Valery N. Delnov, Nataliya A. Denisova, Georgy A. Sorokin : The paper presents and discusses the results of experimental and computational studies obtained by the authors on hydrodynamics and heat exchange in fuel assemblies of the alkali liquid metal cooled fast reactor cores, and experimental data on hydrodynamics of flow paths in the heat exchanger and reactor header systems. Investigation results are presented on in-tank coolant circulation obtained using a well-developed theory of approximation simulation of the nonisothermic coolant velocity and temperature fields in the fast neutron reactor primary circuit and demonstrating stable stratification and thermal fluctuations in the coolant. Results are presented from experimental and computational simulation of the alkali liquid metal boiling process based on fuel assembly models during an emergency situation caused by an operational occurrence involving simultaneous loss of power for all reactor coolant pumps and the reactor scram rod failure. Objectives are formulated for further studies, achieving which is essential for the evolution of the liquid metal technology, as dictated by the need for the improved safety, environmental friendliness, reliability and longer service life of nuclear power facilities currently in operation and in the process of development. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:14:47 +020
  • On the concept of “effective delayed neutron fraction”

    • Abstract: Nuclear Energy and Technology 8(4): 275-279
      DOI : 10.3897/nucet.8.96567
      Authors : Anatoly G. Yuferov : The article considers methodological issues related to the conceptual and terminological apparatus of the dynamics of nuclear reactors. Based on an elementary analysis of the standard point reactor kinetics equations, the author shows that it is necessary to clarify the physical meaning of the parameter β included in the equations, which is traditionally interpreted as the “effective delayed neutrons fraction” (EDNF). It follows directly from the kinetics equations that the parameter β, which appears in these equations as the EDNF, is, from the point of view of the neutron balance, the fraction of prompt neutrons consumed for the generation of delayed neutron precursors (DNPs), and, from the point of view of the DNP balance, the DNP yield per prompt neutron in a single fission event. With these interpretations taken into account, the role of the β parameter is considered in situations related with its adjustment by multiplying it by the “delayed neutron efficiency factor” and with the establishment of the actual fractions of prompt and delayed neutrons. In particular, it is shown that: the statement “if the delayed neutron fraction is β, then the prompt neutron fraction is equal to 1 – β”, used in the problems of analyzing the nuclear reactor dynamics as a starting position, cannot be considered applicable to any reactor conditions; an increase in the β parameter by multiplying it by the “delayed neutron efficiency factor” leads, contrary to traditional interpretations, not to an increase but to a decrease in neutron reproduction in a supercritical reactor. The proposed clarifications are appropriate both in terms of more adequate descriptions of processes in nuclear reactors and in relation to the formulations of nuclear safety requirements. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:14:31 +020
  • Investigation of algorithms for suppressing xenon oscillations in
           a VVER-1200 reactor

    • Abstract: Nuclear Energy and Technology 8(4): 267-273
      DOI : 10.3897/nucet.8.96566
      Authors : Denis A. Soloviev, Artsrun G. Khachatryan, Yevgeny V. Chernov, Rashdan T. Al Malkawi : This paper presents the results of numerical studies of various algorithms for suppression of xenon offset and power distribution oscillations in the core of a VVER-1200 reactor. The purpose of the research is to select an algorithm that minimizes the amount of liquid radioactive wastes during water exchange in the primary circuit of a nuclear power plant. For this, several algorithms for xenon oscillations suppression were considered. The first algorithm considered was an algorithm for suppression of xenon oscillations, which uses regulation due to AWP only, without utilization of any additional regulation. The second algorithm considered was an algorithm based on the use both AWP and boron regulation. In this algorithm suppression of xenon oscillations was carried out with the help of accelerated initiation of the work of the AWP by changing the boric acid concentration with constant second circuit pressure of the NPP and by utilization of the second control rods group. Last algorithm considered was algorithm based on the use of temperature control for accelerated initiation of the work of the AWP. In this algorithm, xenon oscillations suppression was carried out by changing coolant temperature at the reactor inlet caused by pressure change in the secondary circuit in the normal operation margins, and by involving the second group of control rods. It was shown that the best way to suppress xenon offset and power distribution oscillations in terms of minimization of radioactive liquid wastes amount is the algorithm with accelerated initiation of the AWP due to temperature regulation, with elimination of temperature regulation after minimizing of current axial offset value deviation from the nominal one. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:14:19 +020
  • Simulating a lead-cooled reactor campaign using the EUCLID/V1 code

    • Abstract: Nuclear Energy and Technology 8(4): 261-265
      DOI : 10.3897/nucet.8.96565
      Authors : Aleksandr A. Belov, Valery P. Bereznev, Galina S. Blokhina, Dmitry P. Veprev, Dmitry A. Koltashev, Vladimir S. Potapov, Olga I. Chertovskikh, Aleksey V. Shershov : The paper presents the results of the development of the EUCLID/V1 integrated dynamic code designed to analyze and justify the safety of fast neutron reactor facilities with a liquid-metal coolant, in terms of simulating the reactor campaign. The relevance of this study lies in the need to simulate the behavior of the core at any time during the campaign. It lets us to obtain a full dataset for subsequent simulations of the reactor dynamic conditions (including transient states or accidents). The authors have developed a fuel archive to store calculated data in HDF5 format, created a computational model editor to generate input data in the fuel archive format, and also provided an example of computing the campaign of a lead-cooled fast reactor for three core design models shown in this paper. The main array of fuel assemblies was simulated as a single unit in the first model, as three units in the second model, and in the third every single assembly was unique. In addition, the authors have shown changes in the total masses of actinides in the core, revealed that the different core models have an insignificant effect on the evolution of the total masses of actinides, and given the fuel assembly burnup values for the three core models. For the third model, the largest difference between the minimum and maximum burnup values was obtained with an almost identical average over the fuel assemblies. The reactivity margin over time for the three core models was presented. It was shown that the values and behavior of the reactivity margin during the three micro-campaigns are almost equal. From the fourth to the sixth cycle, the reactivity margin value for the third core model was lower than for the first and the second ones. Finally, the authors conclude that it is desirable to evaluate the behavior of the reactivity margin for lead-cooled fast reactor campaigns based on the detailed model of the core. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:14:03 +020
  • Comprehensive analysis of proliferation protection of uranium due to
           the presence of 232U and its decay products

    • Abstract: Nuclear Energy and Technology 8(4): 253-260
      DOI : 10.3897/nucet.8.96564
      Authors : Gennady G. Kulikov, Anatoly N. Shmelev, Vladimir A. Apse, Evgeny G. Kulikov : For a comprehensive assessment of the protection of uranium against proliferation due to the presence of uranium-232 in it, the authors of the article propose and substantiate an integral protection criterion for this material. The criterion is based on the physical barriers against the proliferation of uranium created by uranium-232, namely: (1) the radiolysis of uranium hexafluoride, which hinders attempts to re-enrich uranium and, as a result, a significant critical mass; (2) hard γ-radiation, which leads to incapacity and death of those who try to handle this material without radiation protection; (3) increased heat release, which disables the components of a nuclear explosive device; and (4) a significant source of neutrons that causes predetonation and thereby reduces the energy yield of a nuclear explosive device. These barriers appear at various stages of uranium handling not only in the indicated order but also act simultaneously, mutually reinforcing one another. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:13:50 +020
  • Carrying out calculations of radiation safety during unloading and
           disassembly of cores of spent removable parts of reactors with liquid
           metal coolant of submarines

    • Abstract: Nuclear Energy and Technology 8(4): 247-251
      DOI : 10.3897/nucet.8.96563
      Authors : Elena V. Devkina, Igor R. Suslov, Vladimir A. Chernov : The results of calculations conducted to substantiate radiation safety while handling spent removable sections (SRS) of reactors with a liquid metal coolant (LMC) for nuclear submarines (NS) are presented in the article. The spent removable sections of reactors with liquid metal coolant for nuclear submarines are the sources of intense neutron and gamma radiation. Shielding should ensure the dose rate level for neutrons and gamma radiation which is not exceeding the values allowed for transportation of nuclear materials established by the NP-053-04 therefore it will attenuate emission of neutrons and gamma quanta by several orders of magnitude. A homogeneous model of the reactor core was used for calculations. Sources of neutrons and photons in the spent nuclear fuel (SNF) of the SRS, sources of photons in the reactor control devices and in construction materials (the reactor vessel and grids of fuel rods) have been taken into consideration while conducting the calculations. The computer code MCNP-4B was employed to calculate dose rates for neutrons and photons. In most cases direct calculations of dose rates for neutrons and secondary gamma-quanta using the MCNP-4B code provided acceptable results with admissible methodical errors. For the tasks with sources of gamma quanta direct calculation using the MCNP-4B brought unsatisfactory results due to strong attenuation. Various methods were applied to reduce dispersion: the first one is to assign importance to the cells and the second one is the method of weight windows iteration. Values of dose rates were obtained with acceptable errors. The results of the calculations provided necessary information to conduct operations to unload spent nuclear fuel from the SRS. The results of performed calculations were also used in the design and manufacturing of the shielding. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:13:37 +020
  • Lead-bismuth cooled reactors: history and the potential of development.
           Part 2. Prospects for development

    • Abstract: Nuclear Energy and Technology 8(4): 237-246
      DOI : 10.3897/nucet.8.96562
      Authors : Vladimir M. Troyanov, Georgy I. Toshinsky, Vladimir S. Stepanov, Vladimir V. Petrochenko : The article presents the main provisions of the concept of the design of the SVBR-100 civilian reactor that meets the requirements for Generation IV nuclear technologies, which is being developed on the basis of a critically analyzed experience in developing and operating lead-bismuth-cooled reactor plants. The authors describe the current status of the project and the prospects for the use of such reactor plants in the nuclear power industry after demonstrating their reliability and safety in the operating conditions of a pilot commercial power plant. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:13:20 +020
  • The effect of errors in the neutron flux density on the uncertainties
           of nuclear concentrations of nuclides arising during the calculation of
           fuel burnup in cells with different neutron spectra

    • Abstract: Nuclear Energy and Technology 8(4): 231-235
      DOI : 10.3897/nucet.8.96559
      Authors : Aleksandr N. Pisarev, Valery V. Kolesov, Dmitry V. Kolesov : Computational studies have been carried out showing the complex time dependence of uncertainties in nuclear concentrations of various nuclides arising from the propagation of the neutron flux density errors in the burnup calculation process in cells with different neutron spectra on the above errors. It is found that these uncertainties not only depend on the burnup time in a complex way, but also depend on the spectrum of the cell. The variants of the cell with thermal and fast neutron spectra were considered. The calculations were performed using the VisualBurnOut program (Kolesov et al. 2009), which makes it possible to estimate these uncertainties arising due to errors in the input parameters of the burnup problem (reaction rates, neutron flux density, etc.). The influence of the number of calculated burnup points on the results of burnup calculations by the Monte Carlo method was investigated. Uncertainties arising in nuclear concentrations at intermediate calculation steps due to errors in nuclear concentrations appearing at the previous step were taken into account in the calculations. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:13:05 +020
  • Physical feasibility of minor actinides transmutation in a
           two-component nuclear energy system in Russia

    • Abstract: Nuclear Energy and Technology 8(4): 225-230
      DOI : 10.3897/nucet.8.93664
      Authors : Andrey A. Kashirskii, Andrey Yu. Khomiakov, Elena A. Rodina : A transition to a two-component nuclear power structure with a reactor fleet consisting of thermal and fast reactors as envisioned in the Russian nuclear power development strategy to 2050 and outlook to 2100 will require optimal spent nuclear fuel and radioactive waste management solutions. A core issue in this regard is managing the long-lived minor actinide (MA) inventory that affects overall nuclear power ecological safety. The study examines several options for homogenous MA (Am and Np) transmutation using modern calculation codes with MA transmutation rate and material balances taken into account. Results demonstrate that if fast reactor installed capacity reaches 92 GWe by 2100 there would not be any need for dedicated MA-burners as the MA issue would be gradually resolved within the two-component nuclear energy system by the end of the century. HTML XML PDF
      PubDate: Tue, 13 Dec 2022 15:12:43 +020
  • Experimental study of using microwave reflex-radar level gauges
           for liquid metal coolants

    • Abstract: Nuclear Energy and Technology 8(3): 219-223
      DOI : 10.3897/nucet.8.94540
      Authors : Vladimir I. Melnikov, Tatyana A. Bokova, Vadim V. Ivanov, Aleksandr R. Marov, Natalia A. Lobaeva, Anatoly S. Kvashennikov, Pavel A. Bokov, Nikita S. Volkov : The article presents the results of work aimed at solving the problem of measuring the coolant level in miscellaneous tanks of liquid-metal-cooled reactor plants, mainly of an integral layout with a free level of the primary coolant. The choice of relevant measuring means and methods is limited by the extreme parameters of the liquid metal coolant (LMC) and operating conditions. Traditional measuring means are practically unsuitable; therefore, measuring the HLMC level is a complex technical task. Based on this review, they propose and describe a method of pulsed microwave reflectometry as the most promising in terms of combining the characteristics of reliability, accuracy and ease of use. The results of the experimental study demonstrated the efficiency of the level gauge, which worked according to this method, for measuring the level of lead-bismuth coolant in the control tank under conditions close to natural ones. An analysis of the results confirmed the possibility of using this method to control the level of melts of various metals as applied to HLMC reactor plants. Using the device for measuring the level, which works according to the proposed method, it is possible to control the level of melt of various metals in tanks in real time without the need to move various parts of the sensitive element of the level gauge while maintaining the tightness of the circuit. This device is applicable for various nuclear power plants, accelerator-controlled systems, research reactors and experimental facilities with liquid metal coolants. HTML XML PDF
      PubDate: Tue, 27 Sep 2022 13:01:07 +030
  • Experience of using loose parts monitoring systems at Novovoronezh

    • Abstract: Nuclear Energy and Technology 8(3): 203-209
      DOI : 10.3897/nucet.8.94106
      Authors : Alexey V. Voronov, Mikhail T. Slepov : In VVER reactor plants, it is impossible to completely exclude the appearance of loose, loosely fixed and foreign objects in the main circulation circuit. Operational experience shows that early detection and estimation of the parameters of such incidents can provide the time required to eliminate or minimize damage to the main equipment of the reactor plant. For this reason, most modern power units with pressurized water reactors (PWR, VVER) are equipped with a loose parts monitoring system (LPMS). At the units under construction, these systems are laid down as standard ones; the power units put into commercial operation in the Soviet period were also equipped with them. The requirements for them are established by international standards. Ongoing research work in this area is aimed at determining the root cause of the acoustic anomaly and the localization of its epicenter. Also, no less significant are the works aimed at determining the mass of a loose object (LO). The most precise definition of this parameter will make it possible to have an idea of the nature of the LO before its withdrawal from the primary circuit and to conclude about whether this object is accidentally found or it is a detached part of the steam generators, main circulation pumps, internal devices or shut-off and control valves. HTML XML PDF
      PubDate: Tue, 27 Sep 2022 13:00:54 +030
  • Computational and experimental studies into the hydrodynamic operation
           conditions of container filters for ion-selective treatment

    • Abstract: Nuclear Energy and Technology 8(3): 197-202
      DOI : 10.3897/nucet.8.94105
      Authors : Oleg L. Tashlykov, Ilya A. Bessonov, Artem D. Lezov, Sergey V. Chalpanov, Maxim S. Smykov, Gleb I. Skvortsov, Victoria A. Klimova : Formation of radioactive waste (RW) is specific to the NPP operation. Liquid radioactive waste (LRW) forms in the process of the reactor plant operation, and in decontamination of equipment, rooms and overalls. The radionuclides found mostly in vat residues are 134, 137 Cs in the form of ions and 60Co and 54Mn isotopes in the form of chelates including substances used for equipment decontamination. Among the well-known conditioning techniques, selective sorption provides for the greatest reduction of LRW amounts. The efficiency of using the amount of the filter material can be increased by supplying the treated medium simultaneously to several sorbent layers. The paper presents computer simulation results for three proposed options of improved container filter designs for ion-selective treatment differing in the ways used both to separate the treated water flows and to deliver these to the sorbent layers. The improved efficiency of the sorption processes in the proposed designs was estimated using computer simulation in SolidWorks Flow Simulation. Three sorbent grades from NPP Eksorb were used for the study. A series of experimental studies of the flow through the sorbent layer was undertaken to determine the hydraulic resistance of the studied samples. The obtained experimental data was added to the Solidworks Flow Simulation engineering database for simulation of the earlier presented designs. Representative parameters of the flow inside of container filters were obtained as a result of the simulation. HTML XML PDF
      PubDate: Tue, 27 Sep 2022 13:00:40 +030
  • Lead-bismuth cooled reactors: history and the potential of development.
           Part 1. History of development

    • Abstract: Nuclear Energy and Technology 8(3): 187-195
      DOI : 10.3897/nucet.8.93908
      Authors : Vladimir M. Troyanov, Georgy I. Toshinsky, Vladimir S. Stepanov, Vladimir V. Petrochenko : The article is devoted to the history of the creation of lead-bismuth-cooled reactor units (RUs) for nuclear-powered submarines (NPSs), which were developed in the absence of the necessary knowledge and experience, as well as under strict deadlines for completing work, which practically excluded the possibility of carrying out related full-scale scientific research. This led to a number of failures at the stage of developing this unique technology, the causes of which were later identified and eliminated. The authors explain the reasons for choosing a lead-bismuth eutectic alloy as a coolant, outline the main scientific and technical problems solved in the course of developing a lead-bismuth-cooled reactor unit, including those related to the coolant and corrosion resistance of steels, consider issues of ensuring radiation safety during work related to the release of polonium, ensuring the reliability of steam generators, incidents and accidents that occurred during the period of operation and ways to eliminate their causes. HTML XML PDF
      PubDate: Tue, 27 Sep 2022 13:00:27 +030
  • The concept of a thermionic reactor-converter with evaporative
           heat transfer

    • Abstract: Nuclear Energy and Technology 8(3): 179-185
      DOI : 10.3897/nucet.8.93907
      Authors : Pavel A. Alekseev, Georgiy E. Lazarenko, Vladimir A. Linnik, Aleksandr P. Pyshko : As a result of the analytical studies of the designs of thermionic reactor-converters, four groups of technical solutions have been identified that differ in the method of heat transfer from the fuel to the emitters of the thermionic converter: one option with direct in-core transfer (combining the fuel cladding with the emitter) and three options with thermionic converters taken away from the reactor core, in which case the heat is removed either by heat pipes (common or individual for each fuel element) or is arranged based on the principle of a steam chamber. The article describes the advantages and disadvantages for each of these methods. It is shown that at present the most developed design remains the version with in-core power conversion and, in the future it will be based on the steam chamber since the ingress of gaseous fission products into the inter-electrode gap as well as the influence of fuel swelling on the inter-electrode gap size are excluded and it ensures constant temperature and heat flux density on the surface of all emitters of the thermionic converters, which makes it possible to select the optimal operating parameters for them. A model of a thermionic reactor-converter with a steam chamber containing a reactor core and a zone of thermionic converters has been developed in which the fuel element of the reactor core and the power generating channels of the thermionic converter are separated in space, covered with a capillary porous structure and interconnected by a honeycomb capillary porous spacer plate to provide for circulation of the liquid metal coolant and to let its steam pass through. Neutronic calculations have demonstrated the possibility of a duration for the reactor campaign in excess of ten years following the nuclear safety regulations when a gadolinium oxide coating is applied to the surface of the fuel rods and the reactor vessel in the area of the reactor core. The assessment of thermal and electrical parameters shows that, due to the constant temperature and heat flux density on the surface of all emitters and optimization of the power conversion process for all the thermionic converters, one can expect to reach the maximum efficiency of 20%. HTML XML PDF
      PubDate: Tue, 27 Sep 2022 13:00:16 +030
  • Real-time temperature field recovery of a heterogeneous reactor based
           on the results of calculations in a homogeneous core

    • Abstract: Nuclear Energy and Technology 8(3): 211-217
      DOI : 10.3897/nucet.8.94107
      Authors : Vyacheslav S. Kuzevanov, Sergey K. Podgorny : Advanced pressurized water reactors are the main part of a new generation of nuclear power plant projects under development that provide cost-effective power production for various needs (Yemelyanov et al. 1982, Klimov 2002, Boyko et al. 2005, Baklushin 2011, Bays et al. 2019, Nuclear Technology Review 2019). The innovative technologies are aimed at improving the safety and reliability as well as at reducing the cost of NPPs. At the same time, improvements in design, technological and layout solutions are focused primarily on the reactor core. Assessments of the efficiency of these improvements are preceded by numerical simulations of the processes in the core, in particular heat generation and sink, with account for the difference between the study object and the standard version tested in operational practice. The authors of the article propose a method for calculating the temperature field in the core of a heterogeneous reactor (using the example of a pressurized water reactor), which makes it possible to quickly assess the level of temperature safety of various changes in the core and has the necessary speed for analyzing transients in real time. This method is based on the energy equation for an equivalent homogeneous core in the form of a heat equation that takes into account the main features of the simulated heterogeneous structure. The procedure for recovering the temperature field of a heterogeneous reactor uses the analytical relation obtained in this work for the heat sink function, taking into account inter-fuel element heat leakage losses. Calculations of temperature fields in the model of the PWR type reactor (The Westinghouse Pressurized Water Reactor Nuclear Plant 1984) were carried out in stationary and transient operating modes. The calculation results were compared with the results of CFD simulation. The area of competing use of the temperature field recovery method was indicated. HTML XML PDF
      PubDate: Tue, 27 Sep 2022 10:37:46 +030
  • 14C in tree rings in the vicinity of the nuclear facility
           deployment areas

    • Abstract: Nuclear Energy and Technology 8(3): 173-177
      DOI : 10.3897/nucet.8.93905
      Authors : Evgeny I. Nazarov, Aleksandr V. Kruzhalov, Maksim E. Vasyanovich, Aleksey A. Ekidin, Vladimir V. Kukarskikh, Ekaterina V. Parkhomchuk, Aleksey V. Petrozhitskii, Vasily V. Parkhomchuk : 14C is naturally and artificially occurred radionuclide presented in atmosphere. 14C is produced during the operation of a nuclear reactor of any type, enters the atmosphere and became a part of carbon cycle. The article presents the results of measuring the concentration of 14C in the tree rings of 10 pines in the area of the Beloyarsk NPP (BelNPP) and the Institute of Nuclear Materials (INM), Zarechny. The sampling site, located 1200 m east of the INM, was selected based on long-term observations of meteorological parameters. The measurements were carried out using the accelerator mass spectrometer of the Budker Institute of Nuclear Physics, Novosibirsk. The influence of the operation of nuclear installations on the concentration of 14C in the atmospheric air is demonstrated. The range of values for the concentration of carbon-14 in the sample ranged from 116.0 ± 4.4 to 192.0 ± 8.5 pMC. HTML XML PDF
      PubDate: Tue, 20 Sep 2022 10:32:59 +030
  • Neutron background of composite low-enriched uranium fuel of the
           IVG.1M research reactor

    • Abstract: Nuclear Energy and Technology 8(3): 167-172
      DOI : 10.3897/nucet.8.93895
      Authors : Ruslan A. Irkimbekov, Aleksandr D. Vurim, Sergey V. Bedenko, Artur S. Surayev, Galina A. Vityuk : IVG.1M is a research pressurized water reactor designed to use high-enriched fuel. As part of the core conversion program, the reactor will be switched to a new low-enriched composite uranium fuel. Further operation of the reactor is determined by the availability of fresh fuel to replace the core after the next campaign and the possibility of ensuring safe storage of irradiated spent nuclear fuel (SNF) unloaded from the core. The SNF storage conditions are assessed in terms of ensuring nuclear and radiation safety. Radiation safety of the research reactor fuel storage is achieved, first of all, by solving problems of protection against γ-radiation, while neutron radiation, as a rule, is not considered due to its significantly lower intensity compared to γ-radiation. As for the new low-enriched fuel of the IVG.1M reactor, which is characterized by a set of elements with low and medium atomic masses, on which the (α, n) reaction is possible, the assessment of the neutron component is a necessary procedure to ensure safe fuel storage. The authors of the article propose a procedure for calculating the neutron component of the radiation characteristics of fresh and irradiated composite fuel of the IVG.1M reactor, and also estimate the (α, n)-component. The results of the research will be useful in selecting SNF storage and transportation technologies as well as in providing scientific justification for the possibility of using neutron radiation to control burnup. The research was carried out using verified computational codes MCNP5 and Sources-4C, high-precision experimental EXFOR and computational ENDSF data, as well as evaluated nuclear data libraries. HTML XML PDF
      PubDate: Tue, 20 Sep 2022 10:32:48 +030
  • Comparison of methods for calculating the neutronic characteristics of
           a VVER-1200 fuel assembly

    • Abstract: Nuclear Energy and Technology 8(3): 161-165
      DOI : 10.3897/nucet.8.93894
      Authors : Aleksey V. Lavronenko, Vyacheslav G. Savankov, Ruslan A. Vnukov, Elena A. Chistozvonova : This article presents the results of neutronic calculations of a VVER-1200 fuel assembly carried out using the multi-purpose three-dimensional continuous-energy Monte Carlo particle transport code Serpent 2. The study compares neutronic characteristics during the fuel burnup process (1) with and (2) without fuel cooling. In the first option, the FA fuel campaign was simulated with 30-day cooling periods between reactor campaigns. The second option assumed simulating the FA fuel campaign without fuel cooling. In the course of the study, the authors determined the infinite neutron multiplication factors as well as the fuel burnup dependence of the concentrations of xenon, samarium and gadolinium nuclides. In addition, it should be noted that no differences were found in the change in the concentration of gadolinium isotopes, the discrepancy in the values of the multiplication factor, and the accumulation of samarium isotopes during the campaign. HTML XML PDF
      PubDate: Tue, 20 Sep 2022 10:32:36 +030
  • Export prospects of fast reactors desined in Russia with closed nuclear
           fuel cycle facilities

    • Abstract: Nuclear Energy and Technology 8(3): 153-159
      DOI : 10.3897/nucet.8.80757
      Authors : Nikolay Vladimirovich Gorin, Vladimir Petrovich Kuchinov, Andrey Vladimirovich Krivtsov, Alexander Igorevich Orlov, Vladimir Vladislavovich Shidlovskiy, Daria Borisovna Matveeva : The inevitability of switching to carbon-free energy to withstand the climate change is no longer disputed by anyone today. There is no alternative to this, and the scientific community is forming an appropriate understanding of the need for the development of nuclear energy as carbon free energy source. Solutions are already being discussed at the level of the President and the Government of Russia. In this regard, the article shows that such a solution is possible only based on a new technological platform – two-component nuclear power with the development of technologies of fast reactors with a closed fuel cycle. At the same time the prevailing view in the public opinion of Russia, and not only in it, is that climate change problem can be solved only at the expense of solar and wind energy. This attitude needs to be changed, because without the understanding and support of society, it is impossible to achieve a wide spread of fast reactors with closed fuel cycle technologies. It is concluded that in order to promote a new technological platform in commercial energy and ensure the export prospects of fast reactors of Russian design with closed nuclear fuel cycle facilities, it is necessary to attract representatives of business circles and large energy businesses to the number of supporters of such development by demonstrating the profitability of solutions in the medium and long term, implemented in the case of the use of Russian technologies of fast reactors with the closure of the nuclear fuel cycle. HTML XML PDF
      PubDate: Tue, 20 Sep 2022 10:32:24 +030
  • TES-3 – transportable nuclear power plant mounted on
           self-propelled tracked vehicles

    • Abstract: Nuclear Energy and Technology 8(2): 139-142
      DOI : 10.3897/nucet.8.89356
      Authors : Natalia Yu. Naumenko, Inna M. Mokhireva : Until the mid-1950s, scientists and engineers at the Obninsk Institute of Physics and Power Engineering (IPPE) had worked out a number of unique projects that served as the foundation for the development of domestic and world nuclear power engineering. The list of these projects includes, in particular, TES-3, the first mobile nuclear power plant, which has become a symbol of small-scale nuclear power engineering, a historical achievement of Russian scientists, and part of the heritage of the City of Peaceful Atom. TES-3, a demonstration and experimental plant, being one of the possible nuclear power sources for remote areas, was a mobile power-generating unit consisting of four tracked platforms with a reactor unit equipped with a water-cooled and water-moderated reactor with a 1.5 MW turbogenerator. The “self-propelled uranium-fueled machine” was created in record-breaking time due to the scale and cooperation of the project participants under the scientific guidance of the Laboratory V staff. The plant showed reliability in operation, good controllability, safety and maintainability. Over the entire operating period in the power generation mode, TES-3 worked for about 1300 hours without any radiation accidents. After the completion of the first fuel campaign in 1965, the reactor was shut down, but the idea of mobile low-capacity large-component nuclear power plants was further developed in the form of mobile nuclear power plants of the next generation. HTML XML PDF
      PubDate: Tue, 28 Jun 2022 19:45:44 +030
  • On the scalability of the operating capacity of hydrogen

    • Abstract: Nuclear Energy and Technology 8(2): 143-152
      DOI : 10.3897/nucet.8.83223
      Authors : Alexandr V. Avdeenkov, Sergey G. Kalyakin, Sergey L. Soloviev, Huong Duong Quang : One of the main factors in the capacity of passive autocatalytic recombiners (PARs) is its productivity or the hydrogen removal rate. In this work it was demonstrated that regardless of the type of a recombiner, the hydrogen removal rate is mostly determined by the catalytic surface area and the molar density of hydrogen at the inlet. It means that the performance of a recombiner should obey geometric and physical scalability. Geometric scalability is characterized by the retention of the specific (per unit area of the catalytic surface) hydrogen removal rate with increasing the size of the recombiner by increasing the inlet section while maintaining the height and design of the catalytic unit. Physical scalability is characterized by maintaining the hydrogen removal rate of the recombiner at a constant input molar density of hydrogen in an air-hydrogen environment while simultaneously changing the input temperature and pressure. For a numerical demonstration of scalability, several calculations were performed with different initial hydrogen concentrations, external conditions and amounts of catalytic elements. It was shown that, regardless of the number of catalytic plates in the recombiner, the specific removal rate of hydrogen will remain unchanged and that under different external conditions (temperature, pressure), in case they correspond to the same inlet hydrogen density, the hydrogen removal rate does not change. HTML XML PDF
      PubDate: Tue, 28 Jun 2022 16:00:09 +030
  • Innovation needs in nuclear reactor safety and risk

    • Abstract: Nuclear Energy and Technology 8(2): 77-90
      DOI : 10.3897/nucet.8.82296
      Authors : Francesco D’Auria, Romney B. Duffey : After three quarters of a century using nuclear fission to produce energy, Nuclear Reactor Safety and Risk constitutes an established technological sector. A key feature is continuous updating following new discoveries and progress in knowledge, resulting in extensive and elaborate safety methodologies, which are still not internationally accepted, generally applicable or technically consistent. Each country developed its own methods, guides, traditions and requirements to deal with evolving design, safety, siting and licensing issues. There is a clear parallel in societal risk perception between nuclear radiation exposure in accidents and viral infection in pandemics and the fear of the “unknown”. Unfortunately, over the last 20–30 years the declining introduction of electricity by nuclear fission in the countries that contributed most to its earliest development also has broken the bond between new scientific advancements and improvements of existing safety methodologies. By looking at the origins and fundaments of nuclear technology, we consider the following topics of both deterministic and probabilistic interest: a) Loss of Coolant analysis; b) nuclear fuel accident performance weaknesses; c) role of containment and ultimate heat sinks; d) residual risk and emergency system deployment, and e) independent and risk informed decision making assessment. As a key outcome, we propose modifying the traditional licensing methodology, and the use of active and/or passive systems by being subsumed into a broader Engineered Safety Features Management process. Furthermore, we emphasize the need of connecting the As Low As Reasonably Achievable principle with the analyses to demonstrate the safety of nuclear installations minimizing the need for excessive “paper” safety analyses and licensing efforts. HTML XML PDF
      PubDate: Mon, 27 Jun 2022 19:57:06 +030
  • The BFS complex – a unique facility to justify the neutronic
           parameters of the new generation fast reactor cores

    • Abstract: Nuclear Energy and Technology 8(2): 97-105
      DOI : 10.3897/nucet.8.83655
      Authors : Sergey M. Bednyakov, Andrey V. Gulevich, Vladimir G. Dvukhsherstnov, Dmitry A. Klinov, Igor P. Matveenko, Gennadiy M. Mikhailov, Mikhail Y. Semenov : The BFS complex comprising two fast critical facilities – BFS-1 and BFS-2 – is a unique experimental base for research into fast reactor physics, reactor safety, core optimization, justification of the closed fuel cycle parameters. The critical facilities have the same pitch of the core lattice, they are loaded with the same materials for core simulations but they differ in size. Over 60 years of the BFS operation, IPPE specialists have gained considerable experience in operating the facilities and carrying out experiments. More than 150 critical assemblies have been studied in BFS. HTML XML PDF
      PubDate: Mon, 27 Jun 2022 18:46:51 +030
  • Possibility for using a low-enriched target to produce 99Mo in the
           MAK-2 research channel of the VVR-ts reactor

    • Abstract: Nuclear Energy and Technology 8(2): 133-137
      DOI : 10.3897/nucet.8.89351
      Authors : Aleksander S. Zevyakin, Valery V. Kolesov, Artem V. Sobolev, Oleg Yu. Kochnov : Thermal-hydraulic calculations have been conducted with respect to the active part of the MAK-2 loop facility of the VVR-ts research reactor for the 99Mo production. The computational studies were undertaken both for the case of using a highly 235U enriched target and for a low-enriched target. The calculation was performed for the actual technical characteristics of the research channel. The power density for the two simulated cases was obtained in the course of a preliminary neutronic calculation and selected for the most heated channel. The problem is solved for the steady-state mode of the channel coolant flow and takes into account the dependence of the thermophysical parameters of materials on temperature. The volumetric temperature distribution in the channel was obtained in the process of the calculation. The calculation results present the maximum temperatures of the target materials for the 99Mo production. An analysis of the obtained results has shown that the maximum temperatures of the aluminum sleeve and the target filling materials do not exceed the critical values. For the analyzed calculation cases, the maximum coolant temperature is localized at a point near the sleeve wall surface and does not reach the boiling temperature for a given pressure. The study has therefore shown that it is possible to reduce the 235U enrichment of the target filling fissile material to 19.7%, provided the average density of the mixture and the amount of 235U in the target remain the same. At the same time, the amount of the medicinally important 99Mo generated will not practically change, which will lead to reduced capital costs for a highly enriched mixture of the target matrix. HTML XML PDF
      PubDate: Mon, 27 Jun 2022 15:59:36 +030
  • Ultrasonic monitoring of the VVER-1000 FA form change

    • Abstract: Nuclear Energy and Technology 8(2): 127-132
      DOI : 10.3897/nucet.8.89350
      Authors : Aleksandra V. Voronina, Sergey V. Pavlov, Sergey V. Amosov : A procedure has been developed to determine the geometrical parameters of fuel assemblies (FA) by an ultrasonic pulse-echo technique used for all types of light-water reactor FAs. The measurement of geometrical parameters is achieved through the pairwise installation of ultrasonic transducers opposite the FA spacer grid faces at a distance of not more than a half of the transducer acoustic field near-region length such that the acoustic axes of the pairwise transducers are parallel to each other. The advantages of the presented technique is that it enables monitoring of any FA modifications, including the VVER reactor assemblies with a different number of spacer grids. The paper presents a mathematical model of the acoustic path developed in a geometrical acoustics approximation and its verification results. The model was used for computational and experimental studies of the ultrasonic test technique, and engineering formulas have been developed to calculate the errors of the transducer-measured distance to the FA surface. A code has been developed to simulate the FA form change monitoring and can be used to design new monitoring systems. The developed technique to determine the VVER-1000 FA geometrical parameters was introduced at units 1 and 2 of the Temelin NPP, the Czech Republic, for the TVSA-T FA form change monitoring. The successful use of the proposed technique makes it possible to recommend it for use in inspection benches at other NPPs. HTML XML PDF
      PubDate: Mon, 27 Jun 2022 15:59:06 +030
  • Evaluation of transmutation rate of some LLFP in experimental fast
           reactor JOYO

    • Abstract: Nuclear Energy and Technology 8(2): 91-96
      DOI : 10.3897/nucet.8.78428
      Authors : Naima Amrani, Ahmed Boucenna, Ahmed Abdelghafar Galahom : A transmutation process of three long-lived fission products (79Se, 99Tc and 107Pd) in the experimental fast reactor JOYO is postulated. The possibility of increasing the transmutation rate utilizing the high neutron flux present in the JOYO reactor by loading neutron-moderating subassemblies in the reflector zone has been investigated. A cluster of reflector subassemblies was replaced with beryllium or zirconium hydride (ZrH1.65) moderated subassemblies. These moderated subassemblies surrounded one central test subassembly that would contain the three long-lived fission products (LLFP) simultaneous and without isotopic separation. ChainSolver 2.34 code is used to calculate the transmutation rates. In this study, the new characteristics of LLFP transmutation in a fast reactor using moderator materials were shown for future applications. HTML XML PDF
      PubDate: Mon, 27 Jun 2022 13:23:31 +030
  • Proliferation protection of uranium due to the presence of U-232 decay
           products as intense sources of hard gamma radiation

    • Abstract: Nuclear Energy and Technology 8(2): 121-126
      DOI : 10.3897/nucet.8.87814
      Authors : Gennady G. Kulikov, Anatoly N. Shmelev, Vladimir A. Apse, Evgeny G. Kulikov : The objectives of the article are (1) to show the nuclear and physical causes of hard γ-quanta in the U-232 decay chain, (2) to propose tactics for handling uranium containing U-232, and (3) to assess the efficiency of its protective γ-barrier against uncontrolled proliferation. The authors show the general picture of the decay chains of U-232 nuclide transformations, on which the protection of uranium from its uncontrolled proliferation is based. During the decay of nuclei, their emission of α- or β-particles is only the first stage of the most complex process of rearrangement of both the internal structure of the nucleus itself, which consists in the rearrangement of the neutron and proton shells and the levels of its excitation, and in the rearrangement of the electron shells of the atom. As a rule, the daughter nucleus is in a highly excited state, which is removed by the emission of hard γ-quanta and internal conversion electrons. After the second case, the remaining excitation of the atom is removed by the emission of characteristic γ-quanta and Auger-electrons with characteristic γ-quanta. In addition, explanations are given for the quantum-mechanical reasons for the hard γ-radiation of Tl-208 and Bi-212, which complete the U-232 decay chain. The authors also proposed a tactic for handling uranium containing uranium-232. Since the hard γ-quanta of Tl-208 and Bi-212 appear only at the end of the U-232 decay chain, after its chemical purification from its decay products, U-232 itself does not pose a radiation hazard; therefore, at this time it is advisable to conduct all necessary operations for transporting the material to the plant, fabricating uranium-based fuel containing U-232, and transporting this fuel to the nuclear facility where it will be used. HTML XML PDF
      PubDate: Mon, 27 Jun 2022 13:01:02 +030
  • Identifying the key development areas for small nuclear power

    • Abstract: Nuclear Energy and Technology 8(2): 115-120
      DOI : 10.3897/nucet.8.87811
      Authors : Sergey L. Soloviev, Denis G. Zaryugin, Sergey G. Kalyakin, Sergey T. Leskin : The paper considers the key characteristics of the small nuclear power plant (SNPP) modular design, demonstrates the possibility for reducing the construction cost and time for this class of plants due to factory fabrication, the effect of series manufacturing, and less redundant safety systems. It has been shown that it is possible to extend considerably the fields of application for nuclear technologies thanks to modularity and the possibility of ensuring high safety indicators. Potential applications for SNPPs have been analyzed, including power supply to remote (Arctic) territories, switchover from (renovation of) coal-based electricity generation, high-potential heat and hydrogen production for commercial consumers, and other applications. Rationale has been provided for most typical consumer requirements that define the greatest efficiency of the SNPP application in the given field. The need has been shown for developing and introducing a new technology platform for the SNPP-based nuclear power to decarbonize globally the world economy thanks to expanding greatly the application of nuclear power technologies in addition to the technology platform currently developed for the CNFC with fast reactors (addressing the objective of fuel supply and waste recycling) and the controlled nuclear fusion technology platform (addressing the objective of long-term global energy supply). The new platform needs to be based on an extensive international cooperation involving the formation of international consortiums. It has been proposed that a test site be set up to elaborate hydrogen (heat) production technologies for an individual commercial consumer (captive production) and other technologies for the practical use of SNPPs based on a pilot demonstration nuclear power plant. HTML XML PDF
      PubDate: Mon, 27 Jun 2022 13:00:21 +030
  • Results of validation and cross-verification of the ROK/B design code
           on the problem of loss of cooling in the spent fuel pool

    • Abstract: Nuclear Energy and Technology 8(2): 107-113
      DOI : 10.3897/nucet.8.87809
      Authors : Ruslan M. Sledkov, Valery Ye. Karnaukhov, Oleg Ye. Stepanov, Mark M. Bedretdinov, Igor A. Chusov : The procedures of validation and cross-verification of the newly developed computational code ROK/B are described. The main problem solved using the ROK/B code is the substantiation by calculation of the coolant density in the spent fuel pool (SFP) and the temperature regime of the fuel assemblies during a protracted shutdown of the cooling systems (break in the supply of cooling water). In addition to the above, it is possible to use the ROK/B code to carry out calculation of an accident with the discharge of the coolant from the SFP with simultaneous prolonged shutdown of the cooling systems. The ROK/B code allows carrying out calculations for various types of designs of the fuel assemblies and VVER reactors, in particular, VVER-1000, VVER-1200 and VVER-440 power units with single- and two-tiered fuel assembly arrangement, with clad pipes in racks (for compacted assemblies storage) and pipes without cladding, with cased assemblies and caseless ones. During fuel reloading, a high level of the coolant is maintained, which makes it possible to do “wet” transportation of the assemblies from the reactor to the SFP. The mathematical model for heat and mass transfer calculation, including the boiling coolant model, implemented in the ROK/B code, includes: the motion equation, equations for calculating the enthalpy along the height of the fuel section of a fuel assembly with natural circulation of coolant within the channel containing the fuel assembly (lifting section) and in the inter-channel space (lowering section), the equation of mass balance between the channels of the racks with assemblies and in the inter-assembly space and the amount of evaporated (and outflowed) water, the heat balance equation for a fuel rod in a steam environment. The system of equations is supplemented by closing relations for calculating the thermal physics properties of water and steam, fuel and cladding materials, as well as the coefficients of heat transfer from the wall to the steam, hydraulic resistance and density of the steam-water mixture in the channels, and the heat released in the reaction of steam with zirconium. Validation of the computational code was carried out on the basis of the data of the ALADIN experiment performed by German specialists and the data of JSC OKB Gidropress. Cross-verification of the ROK/B code was carried out in comparison with the results of calculation using the KORSAR/GP and SOKRAT/B1 codes. Based on the results of the validation, it has been concluded that the deviation of the ROK/B results from the experimental data is not more than 2 to 10% (10% for the option with a fuel rod power of 20 W). Based on the results of cross-verification, it has been concluded that the discrepancy between the ROK/B results and the calculation results for the KORSAR/GP and SOKRAT/B1 codes is not more than 0.5% (for SOKRAT/V1) and less than 10% (for KORSAR/GP). HTML XML PDF
      PubDate: Mon, 27 Jun 2022 12:59:49 +030
  • Simulating the fuel cycle of a lead-cooled fast reactor

    • Abstract: Nuclear Energy and Technology 8(1): 71-76
      DOI : 10.3897/nucet.8.83062
      Authors : Aleksey V. Balovnev, Vladimir K. Davydov, Andrey P. Zhirnov, Andrey V. Moiseev, Evgenii O. Soldatov : The development of nuclear power with fast reactors is associated with the implementation of a closed nuclear fuel cycle (CNFC). In this regard, one actual task is to simulate the stages of the fuel cycle with study of the neutron-physical characteristics of the core. The design of a reactor for operation in the closed nuclear fuel cycle mode is impossible without the using of verified and certified software packages for calculating fast reactors, capable of simulating all stages of the operation of the reactor facility and the fuel cycle. For the calculations, the FACT-BR software package was used, which has all the necessary capabilities to simulate the operation of the reactor in the closed nuclear fuel cycle mode, taking into account the stages of fuel storage and refabrication. The article presents a technique for modeling the fuel cycle, implemented for the operation of fast reactors with a lead coolant. To demonstrate methodology, a closed nuclear fuel cycle was simulated for the BREST-OD-300 and BR-1200 reactors for the design life. The article describes the scenarios in which the calculation of the burnup of reactor was carried out. In the considered scenarios, it is assumed that the unloading of fuel at the end of the micro campaign is conducted according to the maximum burnup. During the computational modeling the ranges of changes in fuel density and enrichment, reactivity margin, breeding ratio and isotopic composition of plutonium were determined. HTML XML PDF
      PubDate: Fri, 18 Mar 2022 10:46:20 +020
  • Mathematical simulation of an automatic steam turbine control

    • Abstract: Nuclear Energy and Technology 8(1): 63-69
      DOI : 10.3897/nucet.8.83146
      Authors : Maksim A. Trofimov, Yevgeny G. Murachev, Aleksandr A. Rogoza, Nikolay D. Yegupov : The paper considers the construction of a mathematical model for an electrohydraulic system to control automatically the Т-63-13,0/0,25 product manufactured by JSC Kaluga Turbine Plant. Mathematical simulation of control systems makes it possible to improve considerably the quality of control, that is, the accuracy and reliability of such systems, as well as to accelerate greatly the development and calculation of the control system and the parameters of its individual components. The T-63-13,0/0,25 mathematical model of the ASTCS allows estimating the effects of design parameters during any load dropping (in a range of 0 to 100%) and the quality of control for the monitored parameters both in the process of operation as part of an isolated power system (generator output, frequency) and an integrated power system (generator output). A mathematical representation has been developed in the model for the control units, the T-63-13,0/0,25 product model, and the electronic controlling part of each of the control units. It has been proposed that pulse-width modulation be used to control the synchronous motors which makes it possible to control the synchronous machine shaft speed by changing the supply voltage frequency. To this end, the control system’s model uses a frequency converter which is proposed to be used in the real control system. The developed control system with one adjustable steam extraction in the T-63-13,0/0,25 steam turbine is coupled and autonomous, that is, each of the two meters for the turbine’s controlled parameters has effect on both steam distribution systems such that a deviation for one of the controlled parameters does not lead to excitations in the other. HTML XML PDF
      PubDate: Fri, 18 Mar 2022 10:40:00 +020
  • SNF processing electrochemical operations: liquid-metal and salt
           medium purification

    • Abstract: Nuclear Energy and Technology 8(1): 55-61
      DOI : 10.3897/nucet.8.82620
      Authors : Andrey S. Shchepin, Andrey M. Koshcheev, Ivan V. Kuznetsov, Maya Yu. Kalenova, Irina M. Melnikova : The paper investigates the process of regeneration of a liquid metal medium used in the pyroelectrochemical reprocessing of spent mixed uranium-plutonium nitride fuel produced by a fast neutron reactor. The investigation concerns the interaction of liquid cadmium with sludge formed during the anodic dissolution of ceramic nitride pellets in a 3LiCl-2KCl melt medium as well as the possibility of its purification by filtration from individual metal fission products. Anode sludge is represented by fission products of the platinum group, zirconium, molybdenum and technetium. It was determined by scanning electron microscopy that the metal product is composed of several intergrowth phases. It was found that upon contact of a polymetallic alloy simulating anode sludge with a melt, the liquid metal phase is saturated to 0.025 wt% of Pd, 0.01 wt% of Rh for 50 hours at 500 °C, while zirconium forms an insoluble dispersed intermetallic compound ZrCd3. Powders of molybdenum and technetium, which are not wetted with cadmium, can be completely removed using a filter mesh of plain weaving of the P-200 type. It is also possible to remove zirconium from anodic cadmium by filtration. The filtration efficiency of ruthenium and palladium powders did not exceed 54.3 and 13.1 wt%, respectively, due to partial dissolution and thinning of particles, which will lead to saturation of the liquid metal phase and the need to purify it by alternative methods. HTML XML PDF
      PubDate: Fri, 18 Mar 2022 10:29:38 +020
  • Comparison of the minor actinide transmutation efficiency in models of
           a fast neutron uranium-thorium fueled reactor

    • Abstract: Nuclear Energy and Technology 8(1): 49-53
      DOI : 10.3897/nucet.8.82757
      Authors : Valery V. Korobeinikov, Valery V. Kolesov, Aleksandr V. Mikhalev : In terms of nuclear raw materials, the issue of involving thorium in the fuel cycle is hardly very relevant. However, in view of the large-scale nuclear power development, the use of thorium seems to be quite natural and reasonable. The substitution of traditional uranium-plutonium fuel for uranium-thorium fuel in fast neutron reactors will significantly reduce the production of minor actinides, which will make it attractive for the transmutation of long-lived radioactive isotopes of americium, curium and neptunium that have already been and are still being accumulated. Due to the absence of uranium-233 in nature, the use of thorium in the nuclear power industry requires a closed fuel cycle. At the initial stage of developing the uranium-thorium cycle, it is proposed to use uranium-235 instead of uranium-233 as nuclear fuel. Studies have been carried out on the transmutation of minor actinides in a fast neutron reactor in which the uranium-thorium cycle is implemented. Several options for the structure of the core of such a reactor have been considered. It has been shown that heterogeneous placement of americium leads to higher rates of its transmutation than homogeneous placement does. HTML XML PDF
      PubDate: Fri, 18 Mar 2022 10:29:22 +020
  • Radioecological monitoring and its role in ensuring the safety of
           nuclear power plants

    • Abstract: Nuclear Energy and Technology 8(1): 43-48
      DOI : 10.3897/nucet.8.82619
      Authors : Sergey V. Fesenko, Natalia I. Sanzharova, Yevgeny I. Karpenko, Nizametdin N. Isamov, Vladimir K. Kuznetsov, Aleksey V. Panov, Pavel N. Tsygvintsev : The article presents methodological approaches to the organization of radioecological monitoring in the regions where nuclear power plants are located. The analysis of the monitoring results at the Beloyarsk, Kursk, Leningrad and Rostov NPPs showed that the contribution of the natural radiation background to the public exposure dose is within a narrow range from 3.13 to 4.16 mSv per year, and the dose from the existing technogenic contamination varies from 0.47 μSv (Rostov NPP) up to 150 μSv per year (Beloyarsk NPP). The variability of the exposure doses is determined by the influence of natural climatic conditions and by differences in characteristics of contamination sources, including differences in electricity generation technologies. The technogenic radiation background in the area of the Beloyarsk NPP is determined by environmental contamination as a result of previous activities, whereas in the areas of the Leningrad NPP and the Kursk NPP it is associated with Chernobyl fallout (91 and 14 μSv per year, respectively). The contribution of NPPs to the existing technogenic radiation background varies from 1% (Rostov NPP) to 10–11% (Kursk and Beloyarsk NPPs). HTML XML PDF
      PubDate: Thu, 17 Mar 2022 10:29:07 +020
  • Phenomenology of acoustic standing waves as applied to the
           VVER-1200 reactor plant

    • Abstract: Nuclear Energy and Technology 8(1): 37-42
      DOI : 10.3897/nucet.8.82755
      Authors : Gennady V. Arkadov, Vladimir I. Pavelko, Vladimir P. Povarov, Mikhail T. Slepov : The insufficiently studied issues of acoustic standing waves (ASW) in the main circulation circuits of the VVER reactor plants are considered. For a long time no proper attention has been given to this phenomenon both by the researchers and NPP experts. In general, generation of ASWs requires the acoustic inhomogeneities of the medium in the planes perpendicular to the direction of propagation of the longitudinal wave, in which a jump in acoustic resistance occurs, this is shown by the authors based on an example of the wave equation solution (D’Alembert equation) for a certain function of two variables. The ASW classification has been developed based on the obtained experimental material, 6 ASW types have been described, and their key parameters have been specified. The amplitude distributions have been plotted for all major ASW types proceeding from the phase relations of signals from the pressure pulsation detectors and accelerometers installed on the MCC pipelines. The nature of these distributions is general and they are valid for all VVER types. For the first time the globality of all lowest ASW types is identified. Four attribute properties of the ASWs have been formulated. The first attribute is the regular ASW temperature dependences, which is the source of the diagnostic information in the process of heating/cooling of the VVER unit. The linear experimental dependences of the ASW frequencies on coolant temperature have been obtained. The frequencies, at which the MCC resonant excitation due to coincidence of the ASW frequencies with the RCP rotational frequency harmonics, have been found experimentally. The ASW energy, which origin has resulted from the RCP operation, is estimated. The RCP operation can be presented as continuous generation of pressure pulsations, which fall onto the acoustic path inhomogeneities in the form of a traveling wave and generate a standing wave after reflection from them. HTML XML PDF
      PubDate: Thu, 17 Mar 2022 10:28:52 +020
  • Nuclear-optical converters for detecting intense neutron

    • Abstract: Nuclear Energy and Technology 8(1): 31-36
      DOI : 10.3897/nucet.8.82558
      Authors : Pyotr B. Baskov, Gleb V. Marichev, Vyacheslav V. Sakharov, Vladimir A. Stepanov : In the design of nuclear-optical converters (NOC) for detecting intense neutron fields (fluxes over 1015 cm–2·s–1), it is proposed to use hybrid gas ionization chambers (IC), in which electrical and optical neutron detecting methods are combined. For hybrid ICs, a technology is proposed for obtaining radiation-resistant and mechanically strong radiator materials capable of operating at temperatures of up to 1000 °C. This technology is based on solid-phase boron diffusion saturation of steel. It is shown that, at thermal neutron fluxes of 1×1010 n/(cm2·s) and higher, the integral intensity of argon luminescence as a result of ionization by α-particles and 7Li ions from layers of boride phases is sufficient for detection. The combination of optical and radiation properties of multicomponent fluoride glasses makes it possible to use them as condensed active substances of NOCs. Choosing the elemental and isotopic composition, it becomes possible to use fluoride glasses for multichannel neutron detection as well as to significantly simplify the procedure for separating gamma and neutron components of radiation under conditions of intense radiation fluxes. It has been experimentally shown that in irradiation with a neutron flux of 1×1017 n/(cm2·s), the intensity of Nd IR luminescence in glasses based on zirconium fluoride (ZBLAN) increases in the presence of Gd, which interacts with neutrons. HTML XML PDF
      PubDate: Thu, 17 Mar 2022 10:28:37 +020
  • Fusion-fission hybrid reactor facility: neutronic research

    • Abstract: Nuclear Energy and Technology 8(1): 25-30
      DOI : 10.3897/nucet.8.82294
      Authors : Sergey V. Bedenko, Igor O. Lutsik, Anton A. Matyushin, Sergey D. Polozkov, Vladimir M. Shmakov, Dmitry G. Modestov, Vadim V. Prikhodko, Andrey V. Arzhannikov : 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. HTML XML PDF
      PubDate: Thu, 17 Mar 2022 10:26:56 +020
  • Evaluation of the permissible 99Mo activity in the KL-15 cask in the
           design of transportation and process scheme

    • Abstract: Nuclear Energy and Technology 8(1): 21-24
      DOI : 10.3897/nucet.8.82239
      Authors : Vladimir V. Fomichev, Denis A. Pakholik, Oleg Yu. Kochnov, Nikita V. Kuznetsov, Mikhail V. Kharitonov, Vyacheslav V. Nichugovsky : The demand for the use of radioactive isotopes in medicine is increasing with each coming year necessitating the increased output of radionuclide products. One of the most widely spread radionuclides used in medicine is technetium-99m (99mТс) (Feasibility of producing molybdenum-99 2015, NEA 2012, The Supply of Medical Radioisotopes 2015). The very short 99mТс life (6-hour half-life) requires its production directly on the site of medical treatment. This is achieved using molybdenum-technetium generators (Kodina and Krasikova 2014, Technical Reports No. NF-T-5.4. 2013, Technetium-99 Generator 2021) loaded with molybdenum-99 (99Мо), which uninterruptedly decays (half-life of 66 hours) yielding 99mTc. Close attention must be paid in the course of production of molybdenum-technetium generators to radiation safety during transportation of 99Мо on the territory of the manufacturing facility. The main measure for ensuring radiation safety during transportation of 99Мо is the application of special packaging kits. The Karpov Institute of Physical Chemistry JSC uses a wide range of packaging kits of types A and B for transportation of radioactive materials on the territory of the manufacturer with design features providing the required level of radiation safety. In particular, the KL-15 shipping cask loaded/unloaded from the top is used for onsite transportation of 99Мо for charging molybdenum-technetium generators. The maximum permissible activity of 99Мо is not specified in the passport of the KL-15 cask. Planned construction of a radionuclide production shop in accordance with GMP requirements will require the increase of output of target radionuclides by several times. The above considerations necessitated the evaluation of the maximum permissible activity of 99Мо planned to be transported in KL-15 casks. No other type of standard casks can be used because of their outside dimensions prohibiting the unloading of 99Мо inside the “hot” chamber. Calculation and experimental evaluation of permissible 99Мо activity during transportation inside the KL-15 cask was performed. The paper presents the calculated evaluation of the maximum permissible activity of 99Мо in a KL-15 cask to ensure the radiation exposure of personnel of group A working with the cask not exceeding the established level at the enterprise (80 μSv per shift) and not requiring the use of additional measures and means of protection. The results of the work allow us drawing the conclusion that the KL-15 cask ensures the required level of radiation safety with up to 241 Ki of 99Мо loaded in the cask. HTML XML PDF
      PubDate: Thu, 17 Mar 2022 10:26:25 +020
  • Minor actinides transmutation in pressurized water reactors. 1.
           Multiple recycling of minor actinides on the example of one VVER reactor

    • Abstract: Nuclear Energy and Technology 8(1): 13-19
      DOI : 10.3897/nucet.8.80502
      Authors : Yury A. Kazansky, Gleb W. Karpovich : This article explores the possibilities and conditions of combustion in a pressurized water reactor of its own accumulated minor actinides (MA). The simplest computational model is used: an infinitely extended medium with the distribution and composition of all materials of the fuel assembly of the reactor core, similar to VVER-1200, with uranium dioxide having an initial 235U enrichment of 4.95%. The burnup model is presented in the form of iterations, each of which simulates a fuel campaign lasting 4 years without refueling. At the start of the cycle, special fuel rods are loaded with minor actinides extracted from the reprocessed SNF of the VVER-1200 reactor. After the end of the fuel campaign, all the MAs are removed from the SNF and used in a new iteration. As a result of calculations, it was found that the MA mass in the cycle after 3–7 iterations (depending on the number of fuel elements allocated for the placement and accumulation of MAs) tends to an equilibrium state (regardless of the MAs added every 4 years). In other words, the fuel rods allocated for loading MAs play the role of a kind of furnace, into which, in each iteration, MAs from the previous iteration accumulated in the given reactor are loaded. After several iterations, the burned MA mass converted into fission products is compared with the incoming one. The inclusion of MAs in this way into the fuel cycle converts at least 86% of MAs into fission products without affecting the power generation of the nuclear power plant. It is important that MAs are temporarily unloaded from the reactor after the next iteration in order to remove fission products and to add a new portion of MAs. After stopping the reactor operation, about 16% of the total amount of MAs generated for the entire history of the reactor’s life is discharged into the storage facility. The initial fuel composition in the fuel rods allocated for loading MAs differs from the others only in the amount of MAs and the mass of 238U. The simplified computational model used in this work (without annual overloads of the reactor) influenced the burnup depth and, naturally, the duration of operation, i.e., the k∞ value becomes less than 1 after 1056 days instead of the actual 1460 days with annual fuel overloads. This affected the average fuel composition and, consequently, the neutron spectrum, and could affect the main result of the work, i.e., the number of burned-out MAs in different iterations. Additional calculations, taking into account the annual overloads of the reactor, showed that the change in the spectral composition had little effect on the amount of MAs at the end of the fuel campaign (within 2%). It turned out that the replacement of 238U with minor actinides in fuel rods, the number of which is less than 10, leads to a loss of reactivity. When the number of fuel rods for loading MAs is more than 10, the reactivity increases, giving hope for burning up MAs accumulated in several reactors. HTML XML PDF
      PubDate: Tue, 15 Mar 2022 10:26:01 +020
  • Neutronics and burnup analysis of VVER-1000 LEU and MOX assembly
           computational benchmark using OpenMC Code

    • Abstract: Nuclear Energy and Technology 8(1): 1-11
      DOI : 10.3897/nucet.8.78447
      Authors : Md. Imtiaj Hossain, Yasmin Akter, Mehraz Zaman Fardin, Abdus Sattar Mollah : A handful of computational benchmarks that incorporate VVER-1000 assemblies having low enriched uranium (LEU) and the mixed oxide (MOX) fuel have been put forward by many experts across the world from the Nuclear Energy Agency. To study & scrutinize the characteristics of one of the VVER-1000 LEU & MOX assembly benchmarks in different states were considered. In this work, the VVER-1000 LEU and MOX Assembly computational-benchmark exercises are performed using the OpenMC software. The work was intended to test the preciseness of the OpenMC Monte Carlo code using nuclear data library ENDF/B-VII.1, against a handful of previously obtained solutions with other computer codes. The kinf value obtained was compared with the SERPENT and MCNP result, which presented a very good similarity with very few deviations. The kinf variation with respect to burnup upto 40 MWd/kgHM was obtained for State-5 by using OpenMC code for both the LEU and MOX fuel assembly. The depletion curves of isotope concentrations against burnup upto 40 MWd/kg/HM were also generated for both the LEU and MOX fuel assembly. The OpenMC results are comparable with those of benchmark mean values. The neutron energy vs flux spectrum was also generated by using OpenMC code. Based on the OpenMC results such as kinf, burnup, isotope concentrations and neutron energy spectrum, it is concluded that the OPenMC code with ENDF/B-VII.1 nuclear data library was successfully implemented. It is planned to use OpenMC code for calculation of neutronics and burnup of the VVER-1200 reactor to be commissioned in Bangladesh by 2023/2024. HTML XML PDF
      PubDate: Mon, 14 Mar 2022 10:25:45 +020
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