Latest Articles from Nuclear Energy and Technology Latest 57 Articles from Nuclear Energy and Technology https://nucet.pensoft.net/ Thu, 28 Mar 2024 17:21:28 +0200 Pensoft FeedCreator https://nucet.pensoft.net/i/logo.jpg Latest Articles from Nuclear Energy and Technology https://nucet.pensoft.net/ Towards a uniform description of recombiners performance by a consistent CFD approach with the use of a detailed mechanism of hydrogen oxidation https://nucet.pensoft.net/article/122353/ Nuclear Energy and Technology 10(1): 33-39

DOI: 10.3897/nucet.10.122353

Authors: Alexander V. Avdeenkov, Oleg I. Achakovskii, Vladimir V. Ketlerov, Sergei L. Soloviev, Quang Huong Duong

Abstract: For a consistent CFD substantiation of the recombiner performance, a detailed mechanism of hydrogen and oxygen recombination is used. The detailed mechanism of chemical kinetics (multi-step recombination reaction) makes it possible to claim universality, both in the numerical justification of the recombiner performance and in the justification of the flameless recombination threshold and makes it possible to justify the method for optimizing the recombiner to improve its characteristics. The models developed based on this approach were applied to both flat and cylindrical catalytic elements, which are used in FR and RVK recombiners, respectively. As part of the numerical studies, the detailed recombination mechanism was verified, namely the temperature distribution along the catalytic elements was compared and the performance of catalytic elements was compared as well. Good agreement was obtained between the calculated and experimental data. The approach considers not only the mechanism of surface recombination of hydrogen and oxygen on platinum, but also the mechanism of recombination in the gas phase. This makes it possible to calculate the onset of intense combustion outside the catalytic plates, which is a sign of volumetric ignition of the hydrogen-air environment. The concentrations at which such ignition is possible were obtained at different contents of water vapor in the medium. Thus, the proposed approach and the created models make it possible to fully describe the performance of recombiners of distinct designs without the use of additional experimental data, which is extremely necessary when justifying the hydrogen explosion safety of nuclear power plants.

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Research Article Tue, 26 Mar 2024 19:49:13 +0200
Numerical simulation of fluid dynamics and mixing in headers of sodium-air heat exchangers https://nucet.pensoft.net/article/122352/ Nuclear Energy and Technology 10(1): 27-32

DOI: 10.3897/nucet.10.122352

Authors: Timur R. Smetanin, Vasilii V. Pakholkov, Sergey A. Rogozhkin, Sergey F. Shepelev

Abstract: The paper presents the results of a numerical simulation for sodium fluid dynamics and mixing in the tubing system of an air-cooled heat exchanger (AHX), which is a part of the emergency cooldown system (ECS) of sodium fast reactors (SFRs). Non-uniform sodium flows in the AHX tubing system may lead to the mixing of different-temperature sodium flows, temperature fluctuations and tube breaks. It was found in the course of investigating accidents involving breaks in the PFR and Phénix reactor AHX tubing systems that the failure was caused by the metal temperature fluctuations (Cruickshank and Judd 2005). The numerical simulation used three- and one-dimensional computer codes. It has been found that the calculations of the AHX sodium flow rate distribution with a practically acceptable accuracy can be performed using a one-dimensional code. The factors that influence the non-uniform distribution of sodium flows in the AHX tubing system have been analyzed. Calculations have been performed for the AHX sodium flow distributions and for the mixing of different-temperature sodium flows in the AHX outlet header. The results are presented from calculating the amplitude of sodium fluctuations near the AHX header walls. The effect from shutting down several modules on the non-uniform flow distribution and temperature fluctuations in the AHX has been investigated. Approximations of numerical solutions have been obtained for the sodium flow distribution as a function of the number of the modules shut down.

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Research Article Tue, 26 Mar 2024 19:49:03 +0200
Computational simulation of the heat and mass transfer processes occurring in the containment of Novovoronezh NPP II’s units 1 and 2 https://nucet.pensoft.net/article/122292/ Nuclear Energy and Technology 10(1): 19-25

DOI: 10.3897/nucet.10.122292

Authors: Sergey L. Soloviev, Andrey V. Shishov, Vladimir P. Povarov, Sergey V. Yaurov

Abstract: The paper presents information on the key approaches to the design of the containment ventilation system for units 1 and 2 of Novovoronezh NPP II (NPP-2006 project). The authors have developed a CFD model for the containment of Novovoronezh NPP II’s units 1 and 2, which includes the key structural components and the basic equipment installed within the containment. A series of the containment air temperature measurements was undertaken during power operation of the units. Based on the measured temperature values, a series of calculations was undertaken to determine the air temperature field inside the containment. It is revealed that when ensuring the design characteristics of the cooling capacity of the ventilation system stages, the design parameters of the containment air, wall and equipment temperature are achieved. In addition, with proper mixing of the containment air, it is possible to significantly reduce the average air temperature in the most “hot” rooms. Based on the calculation results, causes have been identified for the low efficiency of the ventilation system, and specific measures have been proposed for increasing significantly the system capacity. The proposed approach to determining the characteristics of ventilation systems using modern methods of three-dimensional computational hydro-gas dynamics makes it possible to optimize and modernize existing ventilation systems, as well as to assess the efficiency of ventilation at the design stage of nuclear power plants. The developed and proposed CFD model makes it possible to do this at the modern level without resorting to bench/experimental modeling issues.

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Research Article Tue, 26 Mar 2024 17:40:45 +0200
Computer simulation of thermal processes involving Sr and Ca radionuclides in the process of heating radioactive graphite in an air atmosphere https://nucet.pensoft.net/article/116661/ Nuclear Energy and Technology 9(4): 273-279

DOI: 10.3897/nucet.9.116661

Authors: Nikolay M. Barbin, Stanislav A. Titov, Dmitry I. Terentiev, Anton M. Kobelev

Abstract: The paper presents the results from a thermodynamic analysis of the behavior of Sr and Ca radionuclides in the process of heating radioactive graphite in an air atmosphere. The TERRA software package was used for the thermodynamic analysis in a temperature range of 300 to 3600 K to determine the possible composition of the ionized, gaseous and condensed phases. It has been found that strontium is in the form of condensed SrCl2(c) and gaseous SrCl2 in a temperature range of 300 to 1600 K, and in the form of gaseous SrCl2, SrO, SrCl and Sr and ionized SrCl+, Sr+ and SrO+ when the temperature is increased from 1600 to 3600 K. Calcium is in the form of condensed CaCl2(c), CaUO4(c), CaO(c) and gaseous CaCl2 in the temperature interval between 300 and 2100 K, and in the form of gaseous Ca, CaCl and CaO and ionized Ca+, CaO+ and CaCl+ when the temperature is increased from 2100 to 3600 K. The paper determines the key reactions within individual phases and among condensed, gaseous and ionized phases. The equilibrium constants of their reactions have been calculated. Based on the results obtained, dependence plots are presented for the Sr and Ca radionuclide distribution by phases.

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Research Article Tue, 19 Dec 2023 17:56:52 +0200
Coupled modeling of neutronics and thermal-hydraulics processes in LFR under SG-leakage condition https://nucet.pensoft.net/article/99154/ Nuclear Energy and Technology 9(2): 85-92

DOI: 10.3897/nucet.9.99154

Authors: Yurii S. Khomyakov, Valerii I. Rachkov, Iurii E. Shvetsov

Abstract: The lead cooled reactor BREST-OD-300 is developing as a part of Russian federal project “PRORYV”. Two- circuit scheme is used in the reactor for heat removal. An inherent risk of two- circuit reactor is the potential danger of water steam ingression in the core in the case of large leakage in steam generator initiated, for example, Steam Generator Tube Rupture (SGTR). Reactor power and temperature response on vapor penetration to the core is studied, but pressurization effects are not in the purview of the paper. The 3D multi-physics (neutronics + thermal-hydraulics) UNICO-2F code was developed for study of SGTR accident. The code calculates unsteady 3D space dependent distributions of coolant velocity, pressure and temperature, space distributions of vapor concentration and heat release density in the core and 3D temperature distributions in the fuel pins. Guillotine rupture of one tube in Steam Generator (SG) is considered as initial event of the accident. It is shown that even with the most conservative assumptions reactivity insertion due to vapor ingress in the core causes small increase of power in level and as a result maximum cladding temperature continue to stay well below safe operation design limit in the entire transient. Hypothetical option of simultaneous tube rupture in few SG belonging to different loops is also analyzed. It is demonstrated that even in the case of simultaneous large leak in two SG the transient stays mild and temperature in the core after two small oscillations is stabilized at acceptable level. In the long term the analysis confirmed the high level of reactor self-protection against SGTR accident.

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Research Article Tue, 20 Jun 2023 10:42:45 +0300
A spatial dynamic model of the SHELF-M reactor facility with fuel and coolant temperature feedbacks https://nucet.pensoft.net/article/102912/ Nuclear Energy and Technology 9(1): 71-76

DOI: 10.3897/nucet.9.102912

Authors: Denis A. Plotnikov, Aleksey L. Lobarev, Ivan N. Krivoshein, Pavel B. Kuznetsov, Anastasia N. Ivanyuta

Abstract: The evolution of nuclear power is inseparably linked with the development of breakthrough solutions in the field of economic development of new territories. A pressing issue in this connection nowadays is generation of power for remote and hard-to-reach areas with decentralized power supply. To resolve this issue, JSC NIKIET is developing a version of the SHELF-M modular water-cooled water-moderated reactor facility as a source of power for offshore installations, including the Arctic coast areas, as well as localities with practically no power and transport infrastructure. One of the stages in justifying the safety of the reactor facility operation is to investigate the behavior of the reactor facility in dynamic transient modes at various power levels. To this end, a spatial dynamic model has been developed for the reactor facility with fuel and coolant temperature feedbacks. The dynamic model development process is a complex task that includes both preparation of constants for subsequent calculations and generation of the reactor neutronic and thermophysical models. The paper describes the development stages of the SHELF-M reactor facility spatial dynamic model and the results of coupled neutronic and thermophysical calculations for transients using the developed dynamic model of the reactor. Shim rod movement in the cold and hot states of the SHELF-M reactor facility is considered as transients.

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Research Article Fri, 17 Mar 2023 18:37:40 +0200
Study into the physical chemistry and technology of alkali liquid metal coolants for nuclear and thermonuclear power plants https://nucet.pensoft.net/article/101761/ Nuclear Energy and Technology 9(1): 43-49

DOI: 10.3897/nucet.9.101761

Authors: Aleksandr P. Sorokin, Yuliya A. Kuzina, Radomir Sh. Ashadullin, Viktor V. Alekseev

Abstract: It is shown that, as the result of developing alkali liquid metal coolants, including sodium, eutectic sodium-potassium alloy, lithium and cesium, the scientific basis has been established for their application in nuclear power. The paper presents data from investigations of thermophysical, neutronic and physicochemical properties and characteristics of various alkali liquid metal coolants, the content of solid-phase and dissolved impurities in coolants, mass transport of impurities in circulation circuits with alkali liquid metal coolants, development of systems for removal of impurities, and control of the content of impurities in alkali liquid metal coolants. Alkali liquid metal coolants are considered as a part of a system that includes a structural material in contact with the coolant, and a gas space that compensates for the thermal expansion of the coolant. The state of the system is defined by the physicochemical properties of the system’s components. And the coolant and the structural materials also represent subsystems consisting of a base material, a coolant and impurities contained both in the material and in the coolant. It has been shown that each alkali liquid metal coolant has its own set of impurities that define its technology. It depends on the physicochemical properties of the solution of the structural material impurities and components in the coolant. Objectives have been formulated for investigating further alkali liquid metal coolants, as stemming from the need to improve the efficiency, environmental friendliness, reliability and safety, and for extending the life of nuclear power plants in operation or under design. Alkali liquid metals are promising candidate materials for being used in thermonuclear power not only as the coolant but also as the tritium breeding medium. These include, first of all, lithium and its eutectic alloy with lead (17 at. % of lithium). The possibility for using lithium or a lithium-lead alloy as a coolant in the blanket of the international thermonuclear power reactor is compared.

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Research Article Fri, 17 Mar 2023 18:36:52 +0200
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) https://nucet.pensoft.net/article/97652/ 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

Abstract: 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.

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Research Article Tue, 13 Dec 2022 15:15:14 +0200
Thermohydraulic studies of alkali liquid metal coolants for justification of nuclear power facilities https://nucet.pensoft.net/article/96568/ 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

Abstract: 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.

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Research Article Tue, 13 Dec 2022 15:14:47 +0200
Experimental study of using microwave reflex-radar level gauges for liquid metal coolants https://nucet.pensoft.net/article/94540/ 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

Abstract: 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.

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Research Article Tue, 27 Sep 2022 13:01:07 +0300
Experience of using loose parts monitoring systems at Novovoronezh NPP https://nucet.pensoft.net/article/94106/ Nuclear Energy and Technology 8(3): 203-209

DOI: 10.3897/nucet.8.94106

Authors: Alexey V. Voronov, Mikhail T. Slepov

Abstract: 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.

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Research Article Tue, 27 Sep 2022 13:00:54 +0300
Computational and experimental studies into the hydrodynamic operation conditions of container filters for ion-selective treatment https://nucet.pensoft.net/article/94105/ 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

Abstract: 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.

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Research Article Tue, 27 Sep 2022 13:00:40 +0300
Lead-bismuth cooled reactors: history and the potential of development. Part 1. History of development https://nucet.pensoft.net/article/93908/ 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

Abstract: 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.

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Research Article Tue, 27 Sep 2022 13:00:27 +0300
Real-time temperature field recovery of a heterogeneous reactor based on the results of calculations in a homogeneous core https://nucet.pensoft.net/article/94107/ Nuclear Energy and Technology 8(3): 211-217

DOI: 10.3897/nucet.8.94107

Authors: Vyacheslav S. Kuzevanov, Sergey K. Podgorny

Abstract: 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.

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Research Article Tue, 27 Sep 2022 10:37:46 +0300
Results of validation and cross-verification of the ROK/B design code on the problem of loss of cooling in the spent fuel pool https://nucet.pensoft.net/article/87809/ 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

Abstract: 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).

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Research Article Mon, 27 Jun 2022 12:59:49 +0300
Mathematical simulation of an automatic steam turbine control system https://nucet.pensoft.net/article/83146/ 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

Abstract: 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.

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Research Article Fri, 18 Mar 2022 10:40:00 +0200
Possibility of simulating natural circulation in fast neutron reactors using a light water test facility https://nucet.pensoft.net/article/78625/ Nuclear Energy and Technology 7(4): 349-355

DOI: 10.3897/nucet.7.78625

Authors: Viktor I. Slobodchuk, Dmitry A. Uralov, Ekaterina A. Avramova

Abstract: The paper evaluates the possibility of modeling the heat transfer phenomena in a liquid-metal coolant using a light water test facility. It considers the natural circulation of the coolant in the upper plenum of the fast-neutron reactor. The sodium-cooled BN-1200 reactor was selected as the reactor installation to be modeled. The development of novel reactor designs must be based on the results of experimental studies. Some problems of modeling thermohydraulic processes in BN type reactors are studied by using sodium test facilities. Experimental studies of natural convection processes using light water test facilities can be considered as a good alternative to those using sodium test facilities. To validate the model, the similarity theory and the “black box” method were used and their principles and applicability were analyzed. Using the “black box” method makes it possible to avoid detailed modeling of such components as the reactor core and heat exchangers, replacing them by a simplified representation of these components to simulate the integral characteristics of the existing real life equipment. The paper considers the basic criteria which determine the similarity of the thermohydraulic processes under study. The governing criteria of similarity were estimated based on the fundamental differential equations of natural convection heat transfer. Based on these criteria, a set of dimensionless values was obtained which show the correlation between the model parameters and the characteristics of the reactor facility. Besides, generalized relationships were derived which can be used to estimate the scaling factors for calculating the key values of the reactor facility based on the model parameters. These relationships depend on the thermal-physics parameters of the working fluids, the geometrical scale value and the ratio of the thermal power of the model to that of the reactor facility, i.e., model-to-reactor thermal power ratio. The conditions under which it is possible to model sodium coolant by light water with adequate accuracy were analyzed. An example is given of the numerical values of the scaling factors for one of the reference light water test facilities. The paper uses the experience of a number of foreign researchers in this field, in particular, the accepted assumptions which do not result in serious loss in modeling accuracy. According to the available estimates, the assumptions used do not result in considerable losses in accuracy. Thus, the natural circulation of the sodium coolant in the upper plenum of the fast-neutron reactor can be simulated with adequate accuracy by using light water test facilities.

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Research Article Fri, 17 Dec 2021 17:32:23 +0200
Studies of critical heat fluxes in small diameter channels https://nucet.pensoft.net/article/78624/ Nuclear Energy and Technology 7(4): 341-348

DOI: 10.3897/nucet.7.78624

Authors: Vladimir I. Belozerov

Abstract: The paper presents the results of experimental studies of critical heat flows in vertical small diameter channels, when the coolant moves from bottom to top, which were carried out in the Obninsk branch of MEPhI in the 1970s of the last century but have not become widespread due to the lack of demand for their practical use. Nowadays, the interest in such works is manifested, first of all, in the development of compact plants and devices, particularly in nuclear power engineering. As a coolant, water, Freon-12 and 96% ethyl alcohol were used. High velocities of underheated liquid at high heat fluxes on the channel wall lead to the so-called “fast crisis” of heat transfer. In this case, the magnitude of the heat flux depends mainly on the parameters of the coolant flow in the wall zone rather than the flow core. The “slow crisis” is mainly observed at high vapor concentrations, relatively low mass flow rates and in an annular-dispersed flow. The value of the critical heat flow in this case depends mainly on the flow parameters in the core, which are probably close to the average coolant flow parameters. The conditions in the near-wall region are also largely determined by the flow in the core. High heat transfer coefficients in a flow moving at high speed usually result in a much smaller and slower rise in the wall temperature. Sometimes a DNB heat flux can occur bypassing the boiling process. In the core of a VVER-type reactor operating in its nominal mode, surface boiling is present in a number of fuel rods. Probably, surface boiling will also be present in transportable and small-size nuclear power plants. Therefore, an important task is to conduct relevant research in this area.

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Research Article Fri, 17 Dec 2021 17:32:02 +0200
Analysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors https://nucet.pensoft.net/article/78368/ Nuclear Energy and Technology 7(4): 311-318

DOI: 10.3897/nucet.7.78368

Authors: Artavazd M. Sujyan, Viktor I. Deev, Vladimir S. Kharitonov

Abstract: The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.

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Research Article Fri, 17 Dec 2021 16:00:03 +0200
Investigations of regularities in the accumulation of hydrogen-reduced slags in circulation circuits with lead-containing coolants https://nucet.pensoft.net/article/74154/ Nuclear Energy and Technology 7(3): 245-252

DOI: 10.3897/nucet.7.74154

Authors: Vladimir V. Ulyanov, Mikhail M. Koshelev, Vladlena S. Kremlyova, Sergey E. Kharchuk

Abstract: The paper presents a computational analysis of regularities in the accumulation of slags during the interaction of lead and lead-bismuth coolants with oxygen gas. Oxidation of lead-containing coolants will cause the formation of lead oxide, while the formation of bismuth oxide is unlikely. Dosed supply of oxidizing gas to lead-containing coolants makes it possible to oxidize, selectively, chromium and nickel to their oxides without the slag formation from solid lead oxide. Regularities were studied which are involved in the lead oxide formation during the interaction of lead-containing coolants with oxygen gas. It has been found that, in the process of interacting with oxygen gas, a lead-bismuth alloy is oxidized 1.7 times as intensively as lead, this being explained by the presence of bismuth in the alloy. Bismuth is oxidized more intensively than both lead and the lead-bismuth alloy. The inert gas overpressure during depressurization does not prevent air oxygen from entering the circuit, and the dependence of the nitrogen and oxygen flow into the circuit on the argon flow out of the loop is close to linear regardless of the circuit state (cold, without coolant; heated, without coolant; heated, with circulating coolant). Oxygen is a chemically active impurity and is absorbed by the circuit; it is therefore important to control nitrogen in the gas spaces of the reactor and research plant circuits with lead-containing coolants. This will make it possible to signal, in a timely manner, the ingress of oxygen into the circuit and to take measures required to avoid or reduce the scale of the slag formation from lead oxides.

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Research Article Thu, 23 Sep 2021 15:14:28 +0300
Experience of using the NALCO 1392 scale inhibitor in the circulating water supply system of the Novovoronezh NPP https://nucet.pensoft.net/article/68940/ Nuclear Energy and Technology 7(2): 85-89

DOI: 10.3897/nucet.7.68940

Authors: Dmitry M. Dronov, Aleksandr V. Gontovoy, Yelena N. Sarkisyan, Natalya V. Karandeeva

Abstract: Power facilities use large amounts of water to cool steam in the steam turbine condensers, and lubricating oils, gas and air in turbine sets. The key requirement for the quality of cooling water is to ensure normal vacuum in condensers. Cooling water must not form mineral and biological deposits and corrosion products in the system. Deposits of mineral salts in the condenser tube system, as well as in auxiliary cooling systems, lead to deterioration in heat exchange and a major decrease in the cost effectiveness of the power equipment operation, and require the heat-exchange equipment to be periodically cleaned. The source water used for cooling is normally taken from nearby water bodies (large rivers or lakes). Circulating water supply systems are used most commonly: these systems use repeatedly the same water inventory for cooling, and require only small amounts of water added to make up for evaporation losses. Coolers, in this case, are cooling towers, spray pools and evaporation ponds. The water chemistry should ensure the operation of equipment without any damage to its components or the loss of efficiency caused by the corrosion of the internal surfaces as well as without scale and sludge formation. It is exactly when using circulating water supply that a stabilizing treatment program is the most practicable way to ensure a cost-effective and environmentally friendly mode of operation. To inhibit scaling processes on the heat-exchange surfaces of the turbine condenser tubes at the Novovoronezh NPP’s unit 5, the cooling water was treated with the NALCO 1392 inhibitor. The results of the NALCO 1392 inhibitor pilot tests in the circulating water supply system (with a cooling pool) are presented.

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Research Article Mon, 21 Jun 2021 10:46:47 +0300
Selection of a turbulence model to calculate the temperature profile near the surface of VVER-1000 fuel assemblies in the NPP spent fuel pool https://nucet.pensoft.net/article/68939/ Nuclear Energy and Technology 7(2): 79-84

DOI: 10.3897/nucet.7.68939

Authors: Aleksandra V. Voronina, Sergey V. Pavlov

Abstract: The paper considers the problem of selecting a turbulence model to simulate natural convection near the surface of a VVER-1000 fuel assembly unloaded from the reactor by computational fluid dynamics (CFD simulation) methods. The turbulence model is selected by comparing the calculated data obtained using the Ansys Fluent software package with the results of experimental studies on the natural convection near the surface of a heated vertical plate immersed in water, which simulates the side face of the VVER-1000 fuel assembly in a first approximation. Two-parameter semi-empirical models of turbulence, k-ε and k-ω, are considered as those most commonly used in engineering design. The calculated and experimental data were compared based on the excessive temperature of the plate surface and the water temperature profiles in the turbulent boundary layer for convection modes with a Rayleigh number of 8∙1013 to 3.28∙1014. It has been shown that the best agreement with experimental data, with an average deviation not exceeding ~ 8%, is provided by the RNG k-ε model which is recommended to be used to simulate natural convection near the surface of VVER-1000 FAs in the NPP spent fuel storage pool.

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Research Article Mon, 21 Jun 2021 10:44:55 +0300
Investigation of the critical heat flux in small-diameter channels https://nucet.pensoft.net/article/65754/ Nuclear Energy and Technology 7(1): 73-78

DOI: 10.3897/nucet.7.65754

Authors: Vladimir I. Belozerov, Aleksandr S. Gorbach

Abstract: The paper describes experimental studies into the hydrodynamics and heat exchange in a forced water flow in small-diameter channels at low pressures. The timeliness of the studies has been defined by the growing interest in small-size heat exchangers. Small-diameter channels are actively used in components of compact heat exchangers for present-day engineering development applications. The major difficulty involved in investigation of heat-transfer processes in small-diameter channels consists in the absence of common methodologies to calculate coefficients of hydraulic resistance and heat transfer in a two-phase flow. The channel size influences the heat exchange and hydrodynamics of a two-phase flow as one of the determining parameters since the existing internal scales (vapor bubble size, liquid droplet diameter, film thickness) may become commensurable with the channel diameter, this leading potentially to different flow conditions. It is evident that one cannot justifiably expect a change in the momentum and energy transfer regularities in single-phase flows as the channel size is reduced for as long as the continuum approximation remains valid. The authors have analyzed the experiments undertaken by Russian scientists to investigate the distribution of thermal-hydraulic parameters in channels with a small cross-section in the entire variation range of the flow parameters in the channel up to the critical heat flux conditions when the wall temperature increases sharply as the thermal load grows slowly. The experimental critical heat flux data obtained by Russian and foreign authors has been compared.

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Research Article Tue, 30 Mar 2021 19:39:02 +0300
Development of a methodological approach for the computational investigation of the coolant flow in the process of the sodium cooled reactor cooldown https://nucet.pensoft.net/article/65442/ Nuclear Energy and Technology 7(1): 61-66

DOI: 10.3897/nucet.7.65442

Authors: Denis V. Didenko, Dmitry Ye. Baluyev, Oleg L. Nikanorov, Sergey A. Rogozhkin, Sergey F. Shepelev, Andrey A. Aksenov, Maksim N. Zhestkov, Aleksandr Ye. Shchelyaev

Abstract: A methodological approach has been developed for the computational investigation of the thermal-hydraulic processes taking place in a sodium cooled fast neutron reactor based on a Russian computational fluid dynamics code, FlowVision. The approach takes into account the integral layout of the reactor primary circuit equipment and the peculiarities of heat exchange in the liquid metal coolant, and makes it possible to model, using well-defined simplifications, the heat and mass exchange in the process of the coolant flowing through the reactor core, and the reactor heat-exchange equipment. Specifically, the methodological approach can be used for justification of safety during the reactor cooldown, as well as for other computational studies which require simulation of the integral reactor core and heat-exchange equipment. The paper presents a brief overview of the methodological approaches developed earlier to study the liquid metal cooled reactor cooldown processes. General principles of these approaches, as well as their advantages and drawbacks have been identified. A three-dimensional computational model of an advanced reactor has been developed, including one heat-exchange loop (a fourth part of the reactor). It has been demonstrated that the FlowVision gap model can be applied to model the space between the reactor core fuel assemblies (interwrapper space), and a porous skeleton model can be used to model the reactor’s heat-exchange equipment. It has been shown that the developed methodological approach is applicable to solving problems of the coolant flow in different operating modes of liquid metal cooled reactor facilities.

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Research Article Tue, 30 Mar 2021 19:38:56 +0300
Forecast of the thermal regime of an underground storage facility for heat-generating materials under mixed convection conditions https://nucet.pensoft.net/article/64361/ Nuclear Energy and Technology 7(1): 1-8

DOI: 10.3897/nucet.7.64361

Authors: Pavel V. Amosov

Abstract: The paper presents the results of a study based on numerical simulation methods of the thermal regime of an underground facility for long-term spent nuclear fuel storage in the version of a built-in reinforced concrete structure. A multiphysical computer model was constructed in a two-dimensional setting by means of the COMSOL software. The mathematical model was based on the continuity equations, Navier-Stokes equations and the general heat transfer equation. The conditions of mixed convection were taken into account in the ‘incompressible ideal gas’ approximation, in which the thermophysical properties of air were a function of temperature only. For two parameters of the model, the following values were taken: the air flow rates providing the velocity at the inflow boundary = 0.01, 0.03 and 0.05 m/s, and the effective heat conductivity coefficients of the material of the built-in structure = 1.0 and 2.0 W/(m×K). Numerical experiments were performed for a period of up to 5 years of fuel storage. Special emphasis was given to the fundamental difference between the non-stationary structure of the velocity fields forecasted in the model of an ‘incompressible ideal gas’ and the ‘frozen’ picture of aerodynamic parameters in the model of an incompressible fluid. An analysis was made of the dynamics of spatial temperature field distributions in different areas of the model. It was shown that the criterion temperature control requirements were met when the facility was operated under conservative ventilation conditions in terms of the air flow rate and the heat conductivity coefficient of the built-in structure material. The dynamics of heat flows directed into the rock mass through the base and from the surface of the built-in structure into the air was analyzed. The heat flow dominance from the structure surface was also noted. Finally, the influence of the effective heat conductivity coefficient of the built-in structure material and the air flow rate on the values of heat flows directed into the air and rock mass was demonstrated.

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Research Article Tue, 23 Mar 2021 19:35:19 +0200
Thermodynamics of equilibrium states and approaches to analyzing the mass transport in metal-oxide systems https://nucet.pensoft.net/article/60300/ Nuclear Energy and Technology 6(4): 261-268

DOI: 10.3897/nucet.6.60300

Authors: Olga V. Lavrova, Aleksandr Yu. Legkikh

Abstract: Analysis of corrosion processes has a major role in justifying the reliability and safety of developed nuclear reactors of a new generation with heavy liquid metal coolants. An approach has been developed which allows practical conclusions to be made with respect to the processes in the given metal-oxide system based on analyzing state diagrams for these systems in the "oxidation potential – temperature" coordinates. The proposed approach relies on a long-term experience of experimental and computational studies concerned with the interaction of various steel grades with molten lead and lead-bismuth, as well as with the transport of metal impurities within these molten metals. The oxidation potential of a metal-oxide system is measured in experimental studies using oxygen activity sensors developed and manufactured at IPPE. The applicability of the proposed approach to analyzing the processes of mass transport in iron-oxygen, lead-oxygen, sodium-oxygen, and iron-water vapor systems has been demonstrated.

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Research Article Fri, 20 Nov 2020 10:52:58 +0200
Operating experience and ways to improve the performance of the service water supply system at the Novovoronezh NPP II (Units 1 and 2) https://nucet.pensoft.net/article/60461/ Nuclear Energy and Technology 6(4): 253-260

DOI: 10.3897/nucet.6.60461

Authors: Vladimir P. Povarov, Dmitry B. Statsura, Dmitry Ye. Usachev

Abstract: The operating experience of Novovoronezh NPP II-1 shows that, in the summer period, the temperature of the cooling water exceeds the design value: this indicates the insufficient performance of the service water supply system. The main factor that has a negative impact on the performance of this system is the formation of carbonate deposits on the cooling tower filler. At Novovoronezh NPP II-1, the cooling tower water distribution system was cleaned from carbonate deposits by the method of combined vibration and aerohydraulic impact. The tested method of cleaning the filler cannot be considered optimal, since the main stage that determines the entire cleaning duration is the assembly/disassembly of the cooling tower filler. It is necessary to continue research on the choice of a strategy for controlling the carbonate deposition rate, taking into account the revealed influence of the design features of the main cooling water pipelines and pipelines of the cooling tower water distribution system on the mechanism of deposit formation in the peripheral spraying area. As compensating measures to ensure the required temperature regime of the turbine plant equipment at Novovoronezh NPP II-1, it is practiced during the summer period to put the standby heat exchangers of the lubrication system and the standby pump of the nonessential services cooling water system into parallel operation. This solution is fraught with the risk of an unplanned decrease in the electrical load if this equipment is turned off in the event of a malfunction. To increase the operating stability of Novovoronezh NPP II-1 and -2 in the summer period, it is proposed to carry out a number of measures aimed at mitigating the negative consequences caused by the elevated service water temperature. Equipment upgrade options are evaluated, e.g., by installing an additional pump for the turbine building services cooling system and (or) laying an additional pipeline to supply part of the makeup water from the Don River directly to the suction pipelines of the pumps of the turbine building services cooling system.

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Research Article Fri, 20 Nov 2020 10:52:25 +0200
Analysis of system characteristics of a reactor with supercritical coolant parameters https://nucet.pensoft.net/article/60296/ Nuclear Energy and Technology 6(4): 243-247

DOI: 10.3897/nucet.6.60296

Authors: Anton S. Lapin, Aleksandr S. Bobryashov, Victor Yu. Blandinsky, Yevgeny A. Bobrov

Abstract: For 60 years of its existence, nuclear energy has passed the first stage of its development and has proven that it can become a powerful industry, going beyond the 10% level in the global balance of energy production. Despite this, modern nuclear industry is capable of producing economically acceptable energy only from uranium-235 or plutonium, obtained as a by-product of the use of low enriched uranium for energy production or surplus weapons-grade plutonium. In this case, nuclear energy cannot claim to be a technology that can solve the problems of energy security and sustainable development, since it meets the same economic and ‘geological’ problems as other technologies do, based on the use of exhaustible organic resources. The solution to this problem will require a new generation of reactors to drastically improve fuel-use characteristics. In particular, reactors based on the use of water cooling technology should significantly increase the efficiency of using U-238 in order to reduce the need for natural uranium in a nuclear energy system. To achieve this goal, it will be necessary to transit to a closed nuclear fuel cycle and, therefore, to improve the performance of a light-water reactor system. The paper considers the possibility of using a reactor with a fast-resonance neutron spectrum cooled by supercritical water (SCWR). The SCWR can be effectively used in a closed nuclear fuel cycle, since it makes it possible to use spent fuel and discharge uranium with a small amount of plutonium added. The authors discuss the selected layout of the core with a change in its size as well as the size of the breeding regions (blankets). MOX fuel with an isotopic plutonium content corresponding to that discharged from the VVER-1000 reactor is considered as fuel. For the selected layout, a study was made of the reactor system features. Compared with existing light-water reactors, this reactor type has increased fuel consumption due to its improved efficiency and nuclear fuel breeding rate up to 1 and above.

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Research Article Wed, 18 Nov 2020 18:00:02 +0200
Basic models and approximation for the engineering description of the kinetics of the oxide layer of steel in a flow of heavy liquid metal coolant under various oxygen conditions https://nucet.pensoft.net/article/59068/ Nuclear Energy and Technology 6(3): 215-234

DOI: 10.3897/nucet.6.59068

Authors: Alexandr V. Avdeenkov, Oleg I. Achakovsky, Vladimir V. Ketlerov, Vladimir Ya. Kumaev, Alexander I. Orlov

Abstract: The article presents the results of corrosion processes, kinetics and changes in the oxide layer modeling using MASKA-LM software complex. The complex is intended for a numerical simulation of three-dimensional non-stationary processes of mass transfer and interaction of impurity components in a heavy liquid metal coolant (HLMC: lead, lead-bismuth). The software complex is based on the numerical solution of coupled three-dimensional equations of hydrodynamics, heat transfer, formation and convective-diffusive transport of chemically interacting components of impurities. Examples of calculations of mass transfer processes and interaction of impurity components in HLMC, formation of protective oxide films on the surfaces of steels are given to justify the coolant technology.

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Research Article Mon, 16 Nov 2020 09:51:00 +0200
Experimental studies into the performance of the lead coolant axial pump wet ends to justify main circulation pumps for the HMLC reactor plant circuits https://nucet.pensoft.net/article/57736/ Nuclear Energy and Technology 6(3): 143-147

DOI: 10.3897/nucet.6.57736

Authors: Aleksandr V. Beznosov, Pavel A. Bokov, Aleksandr V. Lvov, Tatyana A. Bokova, Nikita S. Volkov, Aleksandr R. Marov

Abstract: The paper presents the results of the studies to justify the design solutions for the main circulation pumps of the heavy liquid-metal cooled reactor plant circuits. A substantial difference has been shown in the performance of pumps for the heavy liquid-metal coolant transfer. The studies have confirmed the qualitative difference in the cavitation performance of coolants, the state of the gases and vapors they contain, the influence of supply and discharge devices, and the effects of the impeller blade section performance and geometry and the hub-tip ratio on the pump performance. The studies were performed based on NNSTU’s lead-cooled test facilities with the coolant temperature in a range of 440 to 550 °C and the coolant flow rate of up to 2000 t/h. The outer diameter of the impellers and the straightening devices was about 200 mm, and the thickness of the flat 08Kh18N10T-steel blades was 4.0 mm and that of the airfoil blades was up to 6.0 mm. The pump shaft speed changed in a stepped manner from 600 rpm to 1100 rpm after each 100 rpm. The studies were conducted to justify the engineering and design solutions for pumps as applied to conditions of small and medium plants with fast neutron lead cooled reactors currently under investigation at NNSTU (BRS-GPG). The experimental results can be recommended for use to design other HLMC transfer pumps.

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Research Article Fri, 11 Sep 2020 10:00:02 +0300
On the control of coolant parameters in long-life facilities https://nucet.pensoft.net/article/57734/ Nuclear Energy and Technology 6(3): 137-141

DOI: 10.3897/nucet.6.57734

Authors: Yury G. Cherednichenko, Oleg E. Levin

Abstract: When nuclear power plants with heavy coolants are brought to operating mode as well as during their operation, it is necessary to control and maintain the oxygen content in the coolant within the specified limits. As a rule, the oxygen content in metal melts is controlled by sensors based on solid oxygen-ionic electrolytes. The article presents an analysis of the methodological aspects of dissolved oxygen control in non-isothermal circulating loops with metal coolants, using such sensors. It is shown that in the presence of dissolved loop wall materials and suspensions of their various oxides in the coolant, control over the values of the oxygen activity and concentration calculated for a pure coolant is in general unjustified. The authors present the experimental results of the distribution of oxidation potentials along the loop depending on the coolant temperature, obtained during long-term tests of cladding samples in a lead melt in two circulation facilities – SM2-M and TsU1-M – which differ in principal methods for maintaining specified oxygen conditions. In the low temperature region, the experimental values of the oxidation potential in both facilities are lower than those calculated for pure lead, which leads to a difference by two or more times of the calculated oxygen concentrations for the regions of the loop with Тmin and Тmax, i.e., the so-called oxygen ‘non-isoconcentration’ is observed along the loop. In deoxidation mode during hydrogen ejection into the coolant, the oxidation potential in the loop changes in a complex way, and it makes no sense to talk about the oxygen concentration. It is concluded that in long-life facilities, the coolant parameters for oxygen must be controlled not by the calculated oxygen activity or concentration but by the oxidation potential in the maximum temperature region. To obtain the correct values of the oxidation potential, measurements should be carried out in temperature-stable modes of throughout the facility.

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Research Article Fri, 11 Sep 2020 10:00:01 +0300
Conceptual issues of the cold filter trap development for the sodium coolant purification in fast-neutron reactors https://nucet.pensoft.net/article/55220/ Nuclear Energy and Technology 6(2): 105-111

DOI: 10.3897/nucet.6.55220

Authors: Viktor V. Alekseev, Yuliya A. Kuzina, Aleksandr P. Sorokin

Abstract: The paper presents the results of studying the peculiarities of heat and mass exchange in cold traps for the sodium purification of impurities in fast reactor circuits both in dedicated test areas simulating various trap components (isothermal sump, nonisothermal sump, filters, final cooling area) and in trap prototype models. As a result, a scientific rationale has been formed for developing traps of a unique design for various reactors. The impurity capacity of the traps is three to four times as high as that of the best foreign counterparts. Tests have shown these to be highly efficient in purifying sodium of oxygen and hydrogen and much less efficient in sodium purification of corrosion products and carbon. Taking into account the leakage of radioactive sodium during operation of the BN-600 reactor primary circuit traps, a decision was made to install the purification system in the reactor tank to improve the safety of the large fast reactor. It was resolved to exclude the accumulation of hydrogen in the primary circuit traps in nominal conditions. Two trap designs, with argon and sodium cooling, are discussed. It has been shown that operation of the reactor purification system with argon cooling will require 20 trap replacements during the reactor operating life and seven replacements if the deposition of hydrogen into the primary circuit cold traps is excluded. The sodium-cooled version of the trap built in the reactor tank has the same overall dimensions as the argon-cooled trap. The cooling sodium circulates in two trains: outside the jacketed working space body (up to 30% of the flow rate) and in the coil inside of the working space (up to 70% of the flow rate). Updates have been proposed to the trap design based on the calculations using the codes simulating the in-trap processes of heat and mass exchange.

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Research Article Thu, 25 Jun 2020 15:43:25 +0300
Correlations for calculating the transport and thermodynamic properties of lead-bismuth eutectic https://nucet.pensoft.net/article/55232/ Nuclear Energy and Technology 6(2): 125-130

DOI: 10.3897/nucet.6.55232

Authors: Igor A. Chusov, Vladimir G. Pronyayev, Grigory Ye. Novikov, Nikolay A. Obysov

Abstract: The paper presents recommended correlations for calculating the thermodynamic and transport properties of Pb-Bi eutectic (44.5% Pb + 55.5% Bi), namely: density, dynamic viscosity, specific heat, thermal conductivity, surface tension, specific electrical resistance, and local speed of sound as a function of temperature. These correlations are based on calculated data presented in 39 experimental studies performed in our country and abroad and published during the period from 1923 to 2015. The authors had information on 1103 experimental points; however, a direct assessment was performed on 1076 points. The main difficulty in processing the data was that the experiments considered in the work were performed at different times using a variety of measurement methods, non-unified methods of statistical processing, varying degrees of eutectic purity, etc. The basis of the data estimation technique was the modified least square method, which made it possible to take into account the errors of the experimental data involved. The paper gives the error values of the proposed correlations and the temperature ranges of their applicability. The paper was prepared based on the results of the work of the Thermodynamic Data Center (TDC INPE NRNU MEPhI) of Rosatom State Corporation.

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Research Article Thu, 25 Jun 2020 12:40:37 +0300
A localization method for loose parts monitoring system of VVER reactor plants https://nucet.pensoft.net/article/51252/ Nuclear Energy and Technology 6(1): 29-35

DOI: 10.3897/nucet.6.51252

Authors: Ivan V. Maksimov, Vladimir V. Perevezentsev

Abstract: As operational experience shows, it can hardly be excluded that some detached or loosened parts and even foreign objects (hereinafter referred to as the ‘loose parts’) may appear in the main circulation loop of VVER reactor plants. Naturally, the sooner such incidents are detected and evaluated, the more time will be available to eliminate or at least minimize damage to the reactor plant main equipment. The paper describes a method for localizing the impact of loose parts located in the coolant circulation circuit of a VVER reactor plant. To diagnose malfunctions of the reactor plant main equipment, it is necessary to accurately determine the place where the acoustic anomaly occurred. Therefore, if some loose parts make themselves felt, it is important to track the path of their movement along the main circulation circuit as well as their location using physical barriers. The method is based on the representation of the surface, along which an acoustic wave travels, as a 3D model of the reactor plant (RP) main circulation circuit. The model has the form of a graph in which the vertices characterize the control points on the RP surface and the edges are the distances between them. The method uses information about the acoustic wave velocity and the time difference of arrivals (TDOAs) of the signal received by various sensors. It is shown that, when the effect is received by more than three sensors, along with an estimate of the impact coordinate, it becomes possible to estimate the average acoustic wave velocity. To determine time of arrival, the signal dispersion change point detection method is used. Provided that the average size between the control points on the RP surface was 300 mm, the average localization error was about 600 mm. The developed algorithm can be easily adapted to any VVER reactor plant. The obtained deviation values are acceptable for practical use.

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Research Article Fri, 27 Mar 2020 18:14:22 +0200
A study on the kinetics of bismuth oxide reduction by hydrogen as applied to the technology of removing hydrogen from circulation circuits with heavy liquid metal coolants https://nucet.pensoft.net/article/48425/ Nuclear Energy and Technology 5(4): 331-336

DOI: 10.3897/nucet.5.48425

Authors: Igor I. Ivanov, Vasily M. Shelemetyev, Radomir Sh. Askhadullin

Abstract: As part of the project on developing methods for removing hydrogen and tritium from the circulation circuits of reactor plants with heavy liquid metal coolants, the authors studied the kinetics of bismuth oxide reduction by hydrogen in the temperature range of 425–500 °C and hydrogen concentrations of 25–100 vol.%. The kinetic characteristics of the test reaction were determined by continuous measurements of the water steam (reaction product) concentration in mixtures of hydrogen with helium that passed through a heated reaction vessel with a sample of bismuth oxide. The water steam concentration was measured by a thermal-conductivity detector. The obtained time dependences of the bismuth oxide reduction degree (with varying reaction conditions) were processed by the affine time transformation method. It was also found that the reduction process ran in kinetic mode. The reduction mechanism is the same in the entire temperature range. The limiting reaction stage is the adsorption of hydrogen on the surface of the bismuth oxide sample. The time dependence of the reduction degree is in good agreement with Avrami-Erofeev equation with n = 1. The reaction activation energy is 92.8 ± 1.9 kJ/mol. The reduction reaction rate is directly proportional to the concentration of hydrogen in its mixture with an inert gas.

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Research Article Tue, 10 Dec 2019 14:34:06 +0200
Calculations of research reactor thermal hydraulics based on VVER-440 fuel assamblies https://nucet.pensoft.net/article/48397/ Nuclear Energy and Technology 5(4): 317-321

DOI: 10.3897/nucet.5.48397

Authors: Thi Zieu Chang Doan, Georgy E. Lazarenko, Denis G. Lazarenko

Abstract: Having thoroughly analyzed the design features of VVER-type pressurized water reactors and VVR-type research reactors, the authors propose a design of a research reactor with low-enriched fuel based on deeply updated VVER-440 fuel assemblies. The research reactor is intended to solve a wide range of applied problems in nuclear physics, radiation chemistry, materials science, biology, and medicine. The calculated thermal hydraulics confirms the correctness of the fundamental approaches laid down in the reactor design. An equivalent reactor core model in the form of a thick-walled cylinder was considered, and the radial power density distribution was obtained. According to the heat power level, five groups of FAs were identified. For each group, the coolant mass flow rate was calculated, which ensures alignment with the outlet coolant temperature. The coolant flow regime was also estimated. It turned out that for the first row of FAs, the flow regime is in the transition region, while for the other rows the flow regime is laminar. A test by the Gr.Pr≥1.105 criterion showed its conformity (the calculated value was 1.96.106), indicating the transition to a viscous-gravitational regime. The FE surface overheating was calculated relative to the mixed coolant average temperature. The axial coolant flow temperature distribution is the same in all the FAs, the change in power is compensated by the corresponding change in the coolant flow. The maximum coolant overheating on the FE wall relative to the flow core is observed in the central FAs, reaching 31 °C, the boiling margin is about 15 °C. The estimates showed a significant dynamic pressure margin during natural thermal-convective circulation. By calculation, the values of the FE surface overheating during the reactor normal operation were obtained. An approximately 15-degree surface overheating margin relative to the saturation curve is shown, which guarantees the absence of cavitation wear of the FE claddings. In general, the performed calculations confirmed the correctness of the approaches laid down in the reactor design and made it possible to specify the core thermal hydraulics necessary for further developing the concept.

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Research Article Tue, 10 Dec 2019 14:33:21 +0200
A study into the modes of the VVER-1000 RCP starting in an earlier inoperative loop https://nucet.pensoft.net/article/48393/ Nuclear Energy and Technology 5(4): 305-311

DOI: 10.3897/nucet.5.48393

Authors: Iliya Ye. Bragin, Vladimir I. Belozerov

Abstract: To simulate the mode of the RCP starting in an earlier inoperative loop, KORSAR/GP, a code supporting coupled numerical modeling of neutronic and thermal-hydraulic transients in a VVER reactor plant in operating and emergency conditions, was chosen as the computational tool. Studying these modes using thermal-hydraulic codes makes it possible to analyze the course of transients and certain emergency processes without using commercial test procedures, which contributes to laying the groundwork for addressing the issues involved in ensuring the reliability, operating safety and efficiency of nuclear power plants. Increased requirements to the safety of NPPs identify the need for avoiding excessive conservatism in the analysis based on which requirements to safety systems are formulated, as well as for enhancing the knowledge of the regularities of thermal-hydraulic transients based on advanced computer programs (or codes) designed for improved computational analysis of non-stationary thermal hydraulics in the water-cooled reactor circulation circuits in emergency and transient modes relying on inhomogeneous non-equilibrium mathematical models of two-phase flows and on a detailed description of the physical transient regularities. The purpose of the study is to analyze computationally the starting of a VVER-1000 RCP in an earlier inoperative loop at different reactor plant power values. To do this, one requires to develop the VVER-1000 reactor primary circuit computational pattern to model the transient taking place as one RCP is started, to conduct a further analysis, and to compare the key monitored reactor coolant and core parameters (power, temperature, flow rate, etc.).

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Research Article Tue, 10 Dec 2019 14:32:18 +0200
Temperature field in gas-cooled reactor core in transient conditions under different approaches to mass flow profiling https://nucet.pensoft.net/article/48392/ Nuclear Energy and Technology 5(4): 297-303

DOI: 10.3897/nucet.5.48392

Authors: Vyacheslav S. Kuzevanov, Sergey K. Podgorny

Abstract: Positive effect of profiling the gas-cooled reactor core within the framework of the GT-MHR project was investigated in (Podgorny and Kuzevanov 2017, Kuzevanov and Podgorny 2017, 2018). The necessity arises to supplement already implemented analysis of equilibrium conditions of core operation with investigation of effects of profiling on the temperature field in transient modes of reactor core operation. The present paper is dedicated to the investigation of development of transients in gas-cooled nuclear reactor core subject to the implementation of different principles of core profiling. Investigation of transients in reactor core represents complex problem, solution of which by conducting direct measurements is beyond the resources available to the authors. Besides the above, numerical simulation based on advanced CFD software complexes (ANSYS 2016, 2016a, 2016b, Shaw 1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994) is also fairly demanding in terms of required computer resources. The algorithm for calculating temperature fields using the model where the reactor core is represented as the solid medium with gas voids was developed by the authors and the assumption was made that heat transfer due to molecular heat conductivity can be described by thermal conductivity equation written for continuous medium with thermal physics parameters equivalent to respective parameters of porous object in order to get the possibility of obtaining prompt solutions of this type of problems. Computer code for calculating temperature field in gas-cooled reactor in transient operation modes was developed based on the suggested algorithm. Proprietary computation code was verified by comparing the results of numerous calculations with results of CFD-modeling of respective transients in the object imitating the core of gas-cooled nuclear reactor. The advantage of the developed computer code is the possibility of real-time calculation of evolution of conditions in complex configurations of gas-cooled reactor cores with different channel diameters. This allows using the computer code in the calculations of transients in loops of reactor facility as a whole, in particular for developing reactor simulators. Results are provided of calculations of transients for reactor core imitating the core of gas-cooled nuclear reactor within the framework of GT-MHR project performed for different approaches to profiling coolant mass flow. Results of calculations unambiguously indicate the significant difference of temperature regimes during transients in the reactor core with and without profiling and undeniable enhancement of reliability of nuclear reactor (Design of the Reactor Core 2005, International Safeguards 2014, Safety of Nuclear Power Plants 2014) with profiling of coolant mass flow in the reactor core as a whole.

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Research Article Tue, 10 Dec 2019 14:31:52 +0200
Heat transfer intensification in emergency cooling heat exchanger and dry cooling towers on nuclear power plant using air-water mist flow https://nucet.pensoft.net/article/47972/ Nuclear Energy and Technology 5(4): 281-287

DOI: 10.3897/nucet.5.47972

Authors: Akram H. Abed, Sergey E. Shcheklein, Valery M. Pakhaluev

Abstract: Advanced nuclear power plants are equipped with passive emergency heat removal systems (PEHRS) for removing the decay heat from reactor equipment in accidents accompanied by primary circuit leakage to the final heat absorber (ambient air). Herein, the intensity of heat dissipation to air from the outer surface of the heat exchanger achieved by buoyancy induced natural convection is extremely low, which need to a large heat exchanger surface area, apply different types of heat transfer intensification including (grooves, ribs and extended surfaces, positioning at higher altitudes, etc.). The intensity of heat removal is also strongly dependent on the ambient air temperature (disposable temperature head). Construction of nuclear power plants in countries with high ambient temperatures (Iran, Bangladesh, Egypt, Saudi Arabia, and others) which are characterized by a high level of ambient temperature imposes additional requirements on the increase of the heat exchange surfaces. The experimental investigation results of heat transfer intensification by a low energy ultrasonic which supply a fine liquid droplet (size ~3 µm) in the cooling air are presented in the present paper. In such case, the heat transfer between the surface and cooling flow involves the following three physical effects: convection, conductive heat transfer, and evaporation of water droplets. The last two effects weakly depend on the ambient air temperature and provide an active heat removal in any situation. The investigation was performed using a high-precision calorimeter with a controlled rate of heat supply (between 7800 and 12831 W/m2) imitating heated surface within the range of Reynolds numbers from 2500 to 55000 and liquid (water) flow rates from 23.39 to 111.68 kg·m-2·h-1. The studies demonstrated that the presence of finely dispersed water results in a significant increase in heat transfer compared with the case of using purely air-cooling. With a fixed heat flux, the energy efficiency increases with increasing water concentration, reaching the values over 600 W·m-2·C-1 at 111.68 kg·m-2·h-1, which is 2.8 times higher than for air cooling. With further development of research in order to clarify the optimal areas of intensification, it is possible to use this technology to intensify heat transfer to the air in dry cooling towers of nuclear power plants and thermal power plants used in hot and extreme continental climates.

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Research Article Tue, 10 Dec 2019 14:30:43 +0200
Experimental studies into the dependences of the axial lead coolant circulation pump performance on the pump straightening device parameters https://nucet.pensoft.net/article/39340/ Nuclear Energy and Technology 5(3): 237-240

DOI: 10.3897/nucet.5.39340

Authors: Aleksander V. Beznosov, Aleksander V. Lvov, Pavel A. Bokov, Tatyana A. Bokova, Nikita S. Lukichev

Abstract: The paper presents the results of experimental studies into the dependences of the axial pump performance (delivery rate, head, efficiency) in lead coolant on the parameters of the straightening device (SD) installed downstream of the impeller (the SD inlet flow angle and the number of the SD blades with a variable impeller speed change). The studies were performed as applied to the operating conditions of small and medium plants with lead cooled fast neutron reactors with horizontal steam generators (BRS GPG). The designs of such plants are being matured at Nizhny Novgorod State Technical University (NNSTU). The experiments were conducted on the FT-4 NGTU test bench at the lead coolant temperatures in a range of 440 to 500 °C. The number of the test blades was five and eight, and the SD inlet flow angle was 22, 24, 28, and 32°. The tests were also performed without an SD (with the SD dismantled). The shaft speed of the NSO-01 NGTU pump, with changeable SDs installed into its rotating assembly, was varied in a range of 600 to 1100 rev/min with a step of 100 rev/min. The SD sleeve diameter was 82 mm, the SD blade diameter and height were 213 mm and 80 mm respectively, and the maximum lead coolant flow rate during the studies was up to ~ 1650 t/h. The NSO-01 NGTU pump performance was determined with four changeable straightening devices and with no SD, the pump shaft speed being 600 to 1100 rev/min, as the circulation circuit hydraulic resistance changed owing to the movement of the wedge in the valve installed in it. The tests were performed with the impeller designed and supplied by NNSTU (D = 213 mm, dsl = 82 mm, the blade number is four, and the blade angle is 28°). The obtained results are recommended for use to design axial heavy liquid metal coolant pumps.

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Research Article Wed, 25 Sep 2019 10:40:47 +0300
Experimental studies of temperature pulsations during the process of mixing non-isothermal coolant flows in nuclear reactor equipment components https://nucet.pensoft.net/article/39319/ Nuclear Energy and Technology 5(3): 225-229

DOI: 10.3897/nucet.5.39319

Authors: Sergey M. Dmitriev, Alexandr V. Mamaev, Renat R. Ryazapov, Aleksey Ye. Sobornov, Andrey V. Kotin, Dmitry Ye. Bescherov, Mikhail A. Bolshukhin

Abstract: One of the most important scientific and technical tasks of the nuclear power industry is to assure the reactor equipment life and reliability under random temperature pulsations. High-intensity temperature pulsations appear during the process of mixing non-isothermal coolant flows. Coolant thermal pulsations cause corresponding, sometimes very significant, fluctuations in the temperature stresses of the heat-exchange surface metal, which, added to static loads, can lead to fatigue failure of equipment components. The purpose of this work was to conduct an experimental study of the temperature and stress-strain states of a pipe sample under the influence of local stochastic thermal pulsations caused by the mixed single-phase heat coolant flows. To solve the set problems, an experimental section was created, which made it possible to simulate the process of mixing non-isothermal coolant flows accompanied by significant temperature pulsations. The design of the experimental section allowed us to study the thermohydraulic and life characteristics of pipe samples made of austenite steel (60×5 mm). Some tools were developed for measuring the pipe sample stress-strain state and the coolant flow temperature field in the zone of mixed single-phase media with different temperatures. The measuring tools were equipped with microthermocouples and strain sensors. As a result, we obtained experimental data on temperature pulsations, time-averaged temperature profiles of the coolant flow in the mixing zone as well as statistical and spectral-correlation characteristics of thermal pulsations. Based on the results of measuring the relative strains, the values of fatigue stresses in the mixing zone were calculated. In addition, some devices and methods were elaborated to measure the temperature and stress-strain states of the pipe sample under the influence of local stochastic thermal pulsations. The developed experimental section provided thermal-stress loading of the metal surface at a high level of alternating stress amplitudes causing rapid damage accumulation rates. The results were included in the database to verify the method for assessing the fatigue life of structural materials for nuclear power plants as applied to austenite steel 12Cr18Ni10Ti under the influence of random thermal cyclic loads.

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Research Article Wed, 25 Sep 2019 10:39:53 +0300
Development of the model to determine the fuel temperature field in a two-dimensional problem statement https://nucet.pensoft.net/article/39318/ Nuclear Energy and Technology 5(3): 219-224

DOI: 10.3897/nucet.5.39318

Authors: Vladimir A. Gorbunov, Natalya B. Ivanova, Nikita A. Lonshakov, Yaroslav V. Belov

Abstract: Water-cooled water-moderated reactors (VVER) are widely used at Russian nuclear power plants. The VVER reactor core is formed by fuel assemblies consisting of fuel rods. The fuel in fuel rods is uranium dioxide. The safety of the reactor operation is ensured through stringent requirements for the maximum nuclear fuel temperature. Calculation of temperature fields within the reactor core requires associated problems to be solved to determine the internal energy release in fuel based on neutronic characteristics. Dedicated software for such calculations is not available to a broad range of users. At the present time, there are numerical thermophysical modeling packages available for training or noncommercial applications which are used extensively, including Elcut, Flow Vision, Ansys Fluent, and Comsol Multiphysics. Verification of the obtained results is becoming an important issue in building models using these calculation packages. An analytical solution was obtained as part of the study for the fuel temperature field determination. A program was developed in MathCAD based on this solution. A model was developed in Comsol Multiphysics to determine the fuel temperature field with constant thermophysical properties in a two-dimensional problem statement. The numerical model was verified using the analytical solution. The influence of the number of the grid nodes on the solution accuracy was established. The analytical solution can be used to determine the fuel temperature field at any radial coordinate of the reactor. The temperature field determination model developed in MathCAD can be used to verify numerical models of the fuel temperature field determination developed in dedicated packages.

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Research Article Wed, 25 Sep 2019 10:39:18 +0300
Outcomes of the “steady-state crisis” experiment in the MIR reactor channel https://nucet.pensoft.net/article/39288/ Nuclear Energy and Technology 5(3): 207-212

DOI: 10.3897/nucet.5.39288

Authors: Aleksandr V. Alekseev, Oleg I. Dreganov, Aleksey L. Izhutov, Irina V. Kiseleva, Vitaly N. Shulimov

Abstract: To license nuclear fuel for nuclear power plants, data on the behavior of fuel elements (FE) under design-basis accidents are required. These data are obtained during tests of fuel assemblies (FA) and single fuel elements in research reactor channels followed by post-test studies in protective chambers. A reactivity-initiated accident (RIA) with an unauthorized release of CPS rods from the reactor core leads to a pulsed channel power increase. This accident can proceed according to two scenarios: without a critical heat flux (CHF) on the fuel element jacket at the final stage and with a dry heat flux. To date, a series of experiments have been carried out according to the first scenario in the MIR reactor channel and the corresponding data on the behavior of fuel elements have been obtained. An urgent task for today is to prepare and conduct reactor experiments according to the second scenario. The main experimental parameter that determines the behavior and final state of the studied fuel elements is their temperature. No experimental data were found on the critical heat flux for the rod bundles in the low coolant mass flow rate region (experiments in the MIR reactor channel can be conducted in the range of 200–250 kg/(m2s)). The available data are in the extrapolation range. The “steady-state crisis” experiment was conducted to obtain data on the critical heat flux value within the specified coolant mass flow rate range in the MIR reactor channel. The test object was a jacket fuel assembly composed of three shortened VVER-1000 fuel rods with a length of 1230 mm (the fuel part length = 1000 mm) installed in a triangular grid at a pitch of 12.75 mm, which is a cell of the VVER-1000 core. This assembly configuration is used for in-pile tests to study the behavior of fuel elements under emergency conditions. The in-pile testing results are presented. The paper shows the possibility of detecting the start and development of a dry heat flux based on the readings of thermocouples located inside the FE kernel. As a result, the directly measured test parameters were used to determine the critical heat flux value.

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Research Article Wed, 25 Sep 2019 10:38:27 +0300
Ontologies and databases on thermophysical properties of nuclear reactor materials https://nucet.pensoft.net/article/36476/ Nuclear Energy and Technology 5(2): 145-153

DOI: 10.3897/nucet.5.36476

Authors: Igor A. Chusov, Pavel L. Kirillov, Vladimir G. Pronyaev, Nikolay A. Obysov, Grigoriy E. Novikov

Abstract: The study is dedicated to the information technologies for storage, systematization and distribution of thermophysical data for nuclear power engineering. The general trend existing in the areas involving wide use of scientific data is the shifting from conventional databases to the development of a consolidated infrastructure capable of overcoming sharply growing volumes of scientific data with continuously increasing complexity of the data structure due to the expansion of the range of materials. The above infrastructure ensures interoperability, including data exchange and dissemination. The general principle of data management for thermophysical properties of the nuclear reactor materials based on the subject-oriented ReactorThermoOntology (RTO) is suggested in the present paper. The ontology includes a unified glossary of all concepts, expanded through logical connections and axioms. The suggested RTO ontology combines the terms typical for reactor materials, their characteristics, as well as all types of information entities determining textual, mathematical and computer structures. In the coded form, the ontology becomes the control add-in capable to integrate heterogeneous data. Its most important feature is the possibility of its permanent expansion, which is necessary with introduction of new materials and terms related to them, e.g. nanostructures characteristics. Beside the ontology, description of the reactor materials, the possible scenarios for the use of the ontology during the phases of design, operation and integration of autonomous resources, primarily databases, are examined in the paper. The use of Big Data technology with diverse variations of logical structures of the data is suggested as the most efficient tool for data integration. Joint use of the technologies which before were applied separately, such as exchange standard in the form of the structured text documents, data control based on the ontology and platform for the work with big data, allows the conversion of multiple existing primary resources (databases, files, archives, etc.) to the standard JSON text format for the subsequent semantic integration.

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Research Article Fri, 21 Jun 2019 15:00:03 +0300
Diagnostics of the critical heat flux state of a VVER reactor based on a channel steaming model https://nucet.pensoft.net/article/36475/ Nuclear Energy and Technology 5(2): 139-144

DOI: 10.3897/nucet.5.36475

Authors: Svetlana A. Kachur

Abstract: The purpose of the study is to develop a model for predicting the process of a critical heat flux state with the VVER reactor core channel steaming. The model describes the dynamics of the nuclear reactor behavior in conditions of uncertainty, which are typical of abnormal situations, based on information on the process of heat exchange in the core process channels. The use of the proposed model leads to an increase in the speed of response due to a simplified procedure to calculate the parameters of the heat exchange process in the reactor core. The quality of the reactor state assessment is improved through the prediction of the heat exchange process parameters and determination of the critical heat flux parameters in the core prior to the onset of surface boiling the potentiality of which is not predicted in modern VVER in-core monitoring systems. A modification of the mathematical model has been proposed which offers the simplest possible way of using the advantages of neural networks in diagnostics. The model can be used to develop systems for diagnostics of in-core anomalies and systems for adaptive control of the VVER-type reactor thermal power.

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Research Article Fri, 21 Jun 2019 15:00:02 +0300
A study into the dependence of the cladding-fuel pellet gap conductance on burn-up and the effects on the reactor core neutronic performance https://nucet.pensoft.net/article/35579/ Nuclear Energy and Technology 5(2): 97-102

DOI: 10.3897/nucet.5.35579

Authors: Sergey B. Vygovsky, Fedor V. Gruzdov, Rashdan T. Al Malkawi

Abstract: This paper presents the results of the research to study the dependence of the VVER-1000 (1200) cores neutronic characteristics on the cladding – fuel pellet gap conductance coefficient in the process of the fuel burn-up. The purpose of the study was to determine more accurately the dependence of the cladding – fuel pellet gap conductance coefficient on the fuel burn-up as shown in the Final Safety Report for the Bushehr NPP and to determine the extent of the effects this dependence had on the spatial distribution of the neutron field, on the xenon accumulation rate, and on the kinetic and dynamic behavior of the reactor facility. The paper presents the results of calculating the parameters using which the heat engineering safety of the reactor core is monitored in the process of the fuel burn- up (for a generalized fuel load of a VVER-1000) during the transition to an 18-month nuclear fuel cycle. This paper also includes the results of a numerical research to determine the cladding – fuel gap conductance coefficient depending on the fuel burn-up. These results have shown that, in reality, the gap conductance coefficient dependence on the burn-up does not affect greatly the steady-state characteristics. At the same time, it affects to rather a great extent the xenon accumulation rate, specifically in the event of an extended fuel life. In conditions of maneuvering (load following) modes accompanied by the xenon processes in the reactor core. These facts should be into consideration in design of engineering codes, that used to support the operation of the VVER-1000 (1200) and full-scale simulators.

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Research Article Fri, 17 May 2019 09:45:02 +0300
Gas-cooled nuclear reactor core shaping using heat exchange intensifiers https://nucet.pensoft.net/article/34294/ Nuclear Energy and Technology 5(1): 75-80

DOI: 10.3897/nucet.5.34294

Authors: Vyacheslav S. Kuzevanov, Sergey K. Podgorny

Abstract: The need to shape reactor cores in terms of coolant flow distributions arises due to the requirements for temperature fields in the core elements (Safety guide No. NS-G-1.12. 2005, IAEA nuclear energy series No. NP-T-2.9. 2014, Specific safety requirements No. SSR-2/1 (Rev.1) 2014). However, any reactor core shaping inevitably leads to an increase in the core pressure drop and power consumption to ensure the primary coolant circulation. This naturally makes it necessary to select a shaping principle (condition) and install heat exchange intensifiers to meet the safety requirements at the lowest power consumption for the coolant pumping. The result of shaping a nuclear reactor core with identical cooling channels can be predicted at a quality level without detailed calculations. Therefore, it is not normally difficult to select a shaping principle in this case, and detailed calculations are required only where local heat exchange intensifiers are installed. The situation is different if a core has cooling channels of different geometries. In this case, it will be unavoidable to make a detailed calculation of the effects of shaping and heat transfer intensifiers on changes in temperature fields. The aim of this paper is to determine changes in the maximum wall temperatures in cooling channels of high-temperature gas-cooled reactors using the combined effects of shaped coolant mass flows and heat exchange intensifiers installed into the channels. Various shaping conditions have been considered. The authors present the calculated dependences and the procedure for determining the thermal coolant parameters and maximum temperatures of heat exchange surface walls in a system of parallel cooling channels. Variant calculations of the GT-MHR core (NRC project No. 716 2002, Vasyaev et al. 2001, Neylan et al. 1994) with cooling channels of different diameters were carried out. Distributions of coolant flows and temperatures in cooling channels under various shaping conditions were determined using local resistances and heat exchange intensifiers. Preferred options were identified that provide the lowest maximum wall temperature of the most heat-stressed channel at the lowest core pressure drop. The calculation procedure was verified by direct comparison of the results calculated by the proposed algorithm with the CFD simulation results (ANSYS Fluent User’s Guide 2016, ANSYS Fluent. Customization Manual 2016, ANSYS Fluent. Theory Guide 2016, Shaw1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994).

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Research Article Thu, 11 Apr 2019 11:04:22 +0300
Investigation of the small break conditions in the primary circuit of a VVER-1000 reactor https://nucet.pensoft.net/article/33982/ Nuclear Energy and Technology 5(1): 47-52

DOI: 10.3897/nucet.5.33982

Authors: Vladimir I. Belozerov, Mikhail M. Zhuk, Anna M. Terekhova

Abstract: Modes with violation of the reactor plant cooling conditions on the primary circuit side of a VVER reactor were simulated using the TRAC-PD2 and Open FOAM thermohydraulic codes (TRAC-PD2 1981, OpenFOAM User Guide Version 1.6. 2009, OpenFOAM Programmer’s Guide Version 1.6. 2009) based on energy and mass conservation equations for the three-dimensional unsteady flow of a two-phase mixture. Coupled simulation of the dynamics of neutronic and thermohydraulic processes (TRAC-PD2 1981, OpenFOAM User Guide Version 1.6. 2009, OpenFOAM Programmer’s Guide Version 1.6. 2009, Bolshagin et al. 2009, Galanin 1971, Weinberg and Wigner 1961, Ovchinnikov and Semenov 1988, Report LA-UR-03-1987) aims to improve the qualitative understanding and the quantitative presentation of their effects on safety. Investigating these modes using the above thermohydraulic codes makes it possible to analyze the course of transients and certain emergency processes without using the industrial testing method, this providing the basis for solving the problems of ensuring the reliability, operational safety and efficiency of nuclear power plants. A modern nuclear reactor is a complex system studying and calculating which requires more than the use of simple theoretical models. Thermohydraulic calculations are an essential part of most engineering and technological development works in nuclear power. Since, in conditions of an NPP, no technologically conventional way can be used to verify and update the results and findings of an a priori analysis on the basis of commercial tests, investigations based on codes are used in some cases as the tools to study and predict the parameters of thermohydraulic processes in the reactor’s circulation circuit. The main purpose of the study is to calculate and investigate, based on codes, modes with violation of the reactor plant cooling conditions on the primary circuit side of a VVER reactor in order to determine if calculated parameters conform to the acceptance criteria established by regulatory documents.

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Research Article Wed, 20 Mar 2019 10:50:21 +0200
Vertical steam generators for VVER NPPs https://nucet.pensoft.net/article/33980/ Nuclear Energy and Technology 5(1): 31-38

DOI: 10.3897/nucet.5.33980

Authors: Mikhail Yu. Egorov

Abstract: Steam generators for NPPs are the important large-sized metal consuming equipment of nuclear power installations. Efficiency of steam generator operation determines the overall service life of the whole nuclear facility. The main aim of the current study is to analyze advantages and shortcomings of horizontal and vertical types of steam generator design. This analysis is aimed at the development of recommendations for designing advanced steam generators for future Russian units of NPPs with VVER reactors of increased power. Design solutions and fifty-year experience of operation of 400 steam generators of horizontal type accepted in Russia and of vertical type applied by Westinghouse, Combustion Engineering, Siemens, Mitsubishi, Doosan were analyzed within the framework of the present study. Advantages and drawbacks of both types of equipment determining the development of conditions of the operating processes were also identified and systematized. Currently NPPs equipped with VVER are characterized with extended surface area of containment shells due to the application of four-loop design configuration and horizontal-type steam generators. It was established that steam generator equipment of horizontal type is characterized by such inherent disadvantages of design, technological and operational nature as the following: 1) small height and volume of the vapor space above the evaporation surface reducing separation capabilities and the capacity of the equipment as a whole; 2) impossibility of organizing separate single-phase pre-boiling section. Because of the above, horizontal steam generators with dimensions permissible for railroad transportation and, for VVER-1200 with reactor vessel diameter equal to 5 m, by water transport as well, have exhausted the possibilities for further significant increase of the per unit electric power. The demonstrated advantages of vertical-type steam generators were as follows: 1) absence of stagnant zones within the second cooling circuit, and, consequently, of hold-ups in them; 2) uniformity of heat absorption efficiency of the heating surface ensuring, as well, improved conditions for moisture separation; 3) high degree of moisture removal from steam-water mixture due to the combination of moisture separating elements of chevron and swirl-vane types; 4) increased temperature drop with parameters of generated steam elevated by 0.3 – 0.4 MPa. Conclusion was made on the advisability of introduction of steam generators with vertical-type layout in the Russian nuclear power generation. Practical tasks that need to be addressed in order to ensure introduction of vertical steam generators at NPPs with high-power VVER reactors were formulated.

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Research Article Wed, 20 Mar 2019 10:44:17 +0200
The SIMCO containment code applied to modeling hydrogen distribution in containments of nuclear power facilities https://nucet.pensoft.net/article/31892/ Nuclear Energy and Technology 4(4): 279-285

DOI: 10.3897/nucet.4.31892

Authors: Vyacheslav I. Dorovskikh, Sergey L. Dorokhovich, Aleksey A. Zajtsev, Valery A. Levchenko, Igor N. Leonov

Abstract: 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 was examined using analytical and qualitative tests. A comparison of the numerical and analytical solutions showed good agreement. The code was verified using the experimental data obtained at the NUPEC installation (Japan). Based on the results of testing and verification, it was concluded that, in general, physico-mathematical code models adequately describe the processes of heat/mass transfer in the containment. Therefore, this SIMCO code version can be used to analyze the totality of thermophysical and physico-chemical processes in nuclear power facilities with containments, including the transfer of hydrogen/steam/air mixtures.

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Research Article Thu, 13 Dec 2018 10:42:00 +0200
Development and study of a microwave reflex-radar level gauge of the nuclear reactor coolant https://nucet.pensoft.net/article/31857/ Nuclear Energy and Technology 4(3): 185-190

DOI: 10.3897/nucet.4.31857

Authors: Vladimir I. Melnikov, Vadim V. Ivanov, Ivan A. Teplyashin, Mikhail A. Timonin

Abstract: 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 design is quite simple. In this level gauge, a microwave waveguide made as a coaxial line is used as a transducer (measuring probe). The probe consists of a steel pipe with an external diameter of 20 mm and a central electrode: it is located vertically and immersed in a controlled coolant. The probe wave resistance is 50 ohms. The device electrical diagram is presented. The oscillograms of the received signals and the basic relationships explaining the level gauge operation are given. The signals of the coaxial measuring probe are studied in a fluid with a variable dielectric constant. The results of an experimental study of the level gauge operation in a water coolant at high parameters are given: at pressures up to 10 MPa and temperatures up to 310 °C. It is shown that the device maintains its functional stability under these conditions. The level gauge’s readings practically need not be corrected when the coolant’s thermophysical properties change. The device is intended for use in the control and management systems of nuclear power plants as well as in fuel reprocessing plants.

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Research Article Fri, 7 Dec 2018 10:39:30 +0200
Axial dispersion and mixing of coolant gas within a separate-effect prismatic modular reactor https://nucet.pensoft.net/article/27346/ Nuclear Energy and Technology 4(3): 167-178

DOI: 10.3897/nucet.4.27346

Authors: Ibrahim A. Said, Mahmoud M. Taha, Vineet Alexander, Shaoib Usman, Muthanna H. Al-Dahhan

Abstract: 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 by using hot wire anemometry. The measurements consist of pulse-response of gas tracer that is flowing through the mimicked riser channel using air as a carrier. The dispersion of the gas phase within the separate-effect riser channel was described using one-dimensional axial dispersion model. The axial dispersion coefficient and Peclet number of the coolant gas phase and their residence time distribution within were measured. Effect of heating intensities in terms of heat fluxes on the coolant gas dispersion along riser channels were mimicked in the current study by a certain range of volumetric air flow rate ranging from 0.0015 to 0.0034 m3/s which corresponding to heating intensity range from 200 to 1400 W/m2. Results confirm a reduction in the response curve spreads is achieved by increasing the volumetric air velocity (representing heating intensity). Also, the results reveal a reduction in values of axial dispersion coefficient with increasing the air volumetric flow rate.

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Research Article Fri, 7 Dec 2018 10:37:00 +0200
Analysis of mass transfer processes in a reactor during a loss-of-coolant accident https://nucet.pensoft.net/article/30777/ Nuclear Energy and Technology 4(2): 149-154

DOI: 10.3897/nucet.4.30777

Authors: Aleksey Kulikov, Andrey Lepyokhin, Vitaly Polunichev

Abstract: The purpose of the work was to optimize the parameters of the spillage system equipped with a gas pressure hydroaccumulator for a ship pressurized water reactor in a loss-of-coolant accident. The water-gas ratio in the hydroaccumulator and the hydraulic resistance of the path between the hydroaccumulator and the reactor were optimized at the designed hydroaccumulator geometric volume. The main dynamic processes were described using a mathematical model and a computational analysis. A series of numerical calculations were realized to simulate the behavior dynamics of the coolant level in the reactor during the accident – by varying the optimized parameters. Estimates of the minimum and maximum values of the coolant level were obtained: depending on the initial water-gas ratio in the hydroaccumulator at different diameters of the flow restrictor on the path between the hydroaccumulator and the reactor. These results were obtained subject to the restrictive conditions that, during spillage, the coolant level should remain above the core and below the blowdown nozzle. The first condition implies that the core is in safe state, the second excludes the coolant water blowdown. The optimization goal was to achieve the maximum time interval in which these conditions would be satisfied simultaneously. The authors propose methods for selecting the optimal spillage system parameters; these methods provide the maximum time for the core to be in a safe state during a loss-of-coolant accident at the designed hydroaccumulator volume. Using these methods, it is also possible to make assessments from the early stages of designing reactor plants.

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Research Article Mon, 26 Nov 2018 16:10:00 +0200
Components of small and medium sized HLMC reactor plant circuits https://nucet.pensoft.net/article/30524/ Nuclear Energy and Technology 4(2): 87-92

DOI: 10.3897/nucet.4.30524

Authors: Aleksander Beznosov, Tatyana Bokova, Pavel Bokov

Abstract: Small and medium sized lead and lead-bismuth cooled reactors currently under development in Russia are Generation IV reactors. This paper presents a review and new scientific and engineering solutions which are in line with the evolutionary development of small and medium sized reactor plants with heavy liquid metal coolants (HLMC). A growing interest in small and medium sized reactor plants for transpolar applications, as well as for regional and other NPPs, and the emerging trend towards the substitution of coal-fired boiler stations for small modular reactors initiate R&D on new designs and operational solutions for fast neutron HLMC reactor plants. Such solutions are based on unique domestic experience of building and operating ground prototype test facilities and series lead-bismuth cooled reactor plants, as well as nuclear power units for various applications. These solutions provide for improved properties of advanced HLMC reactors, primarily in economic and safety terms, as compared to other small and medium sized reactor plants. Theoretical and experimental work was undertaken at Nizhny Novgorod State Technical University (NNSTU) for justifying small and medium sized reactor plant designs with horizontal steam generators (BRS-GPG). Nonconventional scientific and engineering solutions have been considered aimed to improve the cost effectiveness and safety of HLMC NPP units, including for the localization of a potentially dangerous severe accident of the “intercircuit steam generator break” type. The review and integrated research results are presented which make it possible to justify nonconventional engineering solutions for the BRS-GPG reactor plant (reactor circuit circulation pattern, steam generator type, reactor circuit heat removal in standby and emergency modes, etc.).

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Research Article Mon, 26 Nov 2018 16:01:00 +0200
Estimation influence of boric acid drop entrainment to its accumulation in the VVER reactor in the case of accident https://nucet.pensoft.net/article/29844/ Nuclear Energy and Technology 4(1): 65-71

DOI: 10.3897/nucet.4.29844

Authors: Andrej V. Morozov, Anna V. Pityk, Sergej V. Ragulin, Azamat R. Sahipgareev, Aleksandra S. Soshkina, Aleksandr S. Shlyopkin

Abstract: 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 the processes of acid accumulation and crystallization in the reactor core are demonstrated. The obtained results allow formulating the conclusion on the possibility of these processes in the reactor core which may lead to the disruption of heat removal from fuel pins. Review of available published reference data on physical properties of boric acid solutions (density, viscosity, thermal conductivity) is given. It is established that available information is of too general nature and fails to cover the whole range of parameters (acid temperature, pressure and concentration) typical for potential emergency situation on NPP equipped with VVER reactor. Necessity of experimental study of processes of droplet entrainment under parameters typical for VVER emergency operation conditions, as well as investigation of thermal physics properties of boric acid within wide range of acid concentration values is required.

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Research Article Thu, 18 Oct 2018 10:11:35 +0300
Measurement of the spent fuel rod cladding temperature during the in-pile testing at 500 – 900°C https://nucet.pensoft.net/article/29838/ Nuclear Energy and Technology 4(1): 21-26

DOI: 10.3897/nucet.4.29838

Authors: Oleg I. Dreganov, Vitalij N. Shulimov, Irina V. Kiselyova, Aleksandr V. Alekseev

Abstract: 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 pattern requires measurements of the cladding temperature with no (thermal, mechanical and other) impacts on the cladding in the maximum deformation region. For dynamic experiments in the MIR reactor channel with fuel testing in a vapor-gas environment, a cladding temperature measuring unit has been developed, in which the cladding is not subjected to external impacts in the maximum deformation region. In the process of being installed into the spacer grid, the thermoelectric transducer (TET) has its hot junction forced against the cladding making it possible to prevent the external impact on the cladding. The thermometric characteristic of the TET attachment, which is associated with the impact of the grid as such on its thermal condition, was studied using a laboratory facility. This technique was used in an in-pile experiment to study the fuel cladding deformation pattern.

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Research Article Wed, 17 Oct 2018 10:49:36 +0300
Regulation of the temperature in the ampoule channel with natural circulation of coolant https://nucet.pensoft.net/article/28724/ Nuclear Energy and Technology 4(1): 1-6

DOI: 10.3897/nucet.4.28724

Authors: Tatiana Osipova, Vladimir Starkov, Vitaly Uzikov

Abstract: 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 heat removal line formed in the upper part above the flow limiter as compared to control only by changing the thermal conductivity of the gas gap in the channel body (through changing the gas pressure or composition). The ampoule channel test conditions, layout and simulation model for thermal-hydraulic analysis using the RELAP5/MOD3.2 code are presented. An investigation was conducted to study the effects of the bypass cooling circuit on the temperature conditions during irradiation of samples in an ampoule channel. The bypass flow rate change is achieved by varying the passage area of the flow limiter orifice. Options have been considered for filling the channel body gas gap with helium and a helium mixture. The calculation showed that the heat removed by the bypass line could reach 40% of the total heat released in the channel. With helium used in the channel body gap, the temperature conditions during irradiation are adjusted in a broader range (200–330 °С) than with a gas mixture of a lower thermal conductivity (279–330 °С), the major temperature variation taking place with the flow limiter orifice area being less than 0.2–0.3 cm2. Any further increase in the flow limiter orifice area does not lead to a major temperature change in the coolant flowing about the samples.

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Research Article Tue, 25 Sep 2018 11:49:04 +0300