2026-10-09T17:10:06Z https://nucet.pensoft.net/oai.php

10.3897/nucet.4.29453 2018-09-25 nucet

Minimize fission power peaking factor in radial direction of water-cooled and water-moderated thermionic conversion reactor core Alekseev,Pavel Krotov,Aleksei Ovcharenko,Mikhail Linnik,Vladimir Thermionic converter reactor thermionic fuel element power peaking factor lattice pitch genetic algorithm optimization Nuclear Energy and Technology 4(1): 7-11 The paper investigates the possibility for reducing the radial power peaking factor kr inside the core of a water-cooled water-moderated thermionic converter reactor (TCR). Due to a highly nonuniform power density, the TCR generates less electric power and the temperature increases in components of the thermionic fuel elements, leading so to a shorter reactor life. A TCR with an intermediate neutron spectrum has its thermionic fuel elements (TFE) arranged inside the core in concentric circles, this providing for a nonuniform TFE spacing and reduces kr. The water-cooled water-moderated TCR under consideration has a much larger number of TFEs arranged in a hexagonal lattice with a uniform pitch. Power density flattening in a core with a uniform-pitch lattice can be achieved, e.g., through using different fuel enrichment in core or using additional in-core structures. The former requires different TFE types to be taken into account and developed while the latter may cause degradation of the reactor neutronic parameters; all this will affect the design’s economic efficiency. It is proposed that the core should be split into sections with each section having its own uniform lattice pitch which increases in the direction from the center to the periphery leading so to the radial power density factor decreasing to 1.06. The number of the sections the core is split into depends on the lattice pitch, the TFE type and size, the reflector thickness, and the reactor design constraints. The best lattice spacing options for each section can be selected using the procedure based on a genetic algorithm technology which allows finding solutions that satisfy to a number of conditions. This approach does not require the reactor dimensions to be increased, different TFE types to be taken into account and developed, or extra structures to be installed at the core center. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.29453 https://doi.org/10.3897/nucet.4.29453 https://nucet.pensoft.net/article/29453/ https://nucet.pensoft.net/article/29453/download/pdf/ en

10.3897/nucet.4.28724 2018-09-25 nucet

Regulation of the temperature in the ampoule channel with natural circulation of coolant Osipova,Tatiana Starkov,Vladimir Uzikov,Vitaly SM research reactor natural-circulation ampoule channel investigation results irradiation temperature conditions power density. Nuclear Energy and Technology 4(1): 1-6 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.28724 https://doi.org/10.3897/nucet.4.28724 https://nucet.pensoft.net/article/28724/ https://nucet.pensoft.net/article/28724/download/pdf/ en

10.3897/nucet.4.29838 2018-10-17 nucet

Measurement of the spent fuel rod cladding temperature during the in-pile testing at 500 – 900°C Dreganov,Oleg Shulimov,Vitalij Kiselyova,Irina Alekseev,Aleksandr Laboratory facility (LF) experimental fuel element (EFE) electroheated fuel element simulator (EHFES) cladding spacer grid (SG) thermoelectric transducer (TET) hot junction temperature heat-up rate MIR reactor loss-of-coolant accident (LOCA). Nuclear Energy and Technology 4(1): 21-26 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.29838 https://doi.org/10.3897/nucet.4.29838 https://nucet.pensoft.net/article/29838/ https://nucet.pensoft.net/article/29838/download/pdf/ en

10.3897/nucet.4.29842 2018-10-17 nucet

Elaboration of approach to nuclear energy systems assessment by criterion of sustainable development Usanov,Vladimir Kviatkovskii,Stepan Andrianov,Andrey Nuclear energy system sustainable development closed nuclear fuel cycle fast reactors. Nuclear Energy and Technology 4(1): 27-33 The paper describes the approach to the assessment of nuclear energy systems based on the integral indicator characterizing the level of their sustainability and results of comparative assessment of several nuclear energy system options incorporating different combinations of nuclear reactors and nuclear fuel cycle facilities. The nuclear energy systems are characterized by achievement of certain key events pertaining to the following six subject areas: economic performance, safety, availability of resources, waste handling, non-proliferation and public support. Achievement of certain key events is examined within the time interval until 2100, while the key events per se are assessed according to their contribution in the achievement of sustainable development goals. It was demonstrated that nuclear energy systems based on the once-through nuclear fuel cycle with thermal reactors and uranium oxide fuel do not score high according to the integral sustainable development indicator even in the case when the issue of isolation of spent nuclear fuel in geological formation is resolved. Gradual replacement of part of thermal reactors with fast reactors and closing the nuclear fuel cycle results in the achievement of evaluated characteristics in many subject areas, which are close to maximum requirements of sustainable development, and in the significant enhancement of the sustainability indicator. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.29842 https://doi.org/10.3897/nucet.4.29842 https://nucet.pensoft.net/article/29842/ https://nucet.pensoft.net/article/29842/download/pdf/ en

10.3897/nucet.4.29850 2018-10-17 nucet

Flow-accelerated corrosion rate and residual life time estimation for the components of pipeline systems at nuclear power plants based on control data Baranenko,Valery Gulina,Olga Salnikov,Nikolaj Flow-accelerated corrosion (FAC) thickness gauging evaluation of FAC rate bends welding residual lifetime. Nuclear Energy and Technology 4(1): 35-42 As of today, large volumes of data related to non-destructive operational control are accumulated on NPPs. For ensuring safe operation of power units, optimization of scope and scheduling operational control it is necessary to continue development of guidance documents, software products, methodological guidance and operational documentation (Baranenko et al. 1998, Gulina et al. 2013, Recommendation (NSAC-202L-R4) 2013). Approaches are examined to assessment of the rate of erosion-corrosion wear (flow-accelerated corrosion - FAC) according to the data of operational control. The present study was performed based on the data of thickness gauging of different elements of pipelines of NPPs with different types of reactor. Further development of ideas exposed in (Baranenko et al. 2016) allowed revealing specific features of ECW processes on straight sections, bends and in the zones adjacent to weld joints of pipelines of NPPs equipped with VVER and RBMK reactors. Presence of the process of deposition of corrosion products on internal surfaces of pipeline walls results in the fact that residual lifetime of elements nominally increases due to deposition. However, real wall thickness under the layer of deposits is unknown just as the initial wall thickness is unknown as well. Investigation implemented in the present study is aimed at the substantiation of the methodology of calculation of FAC rate according to the data of operational control for the purpose of drawing calculation results closer to the reality keeping conservatism. Uniform approach to the assessment of FAC rate in the examined elements of pipelines was developed. Methodologies for evaluation of correction coefficients taking into account dimensional technological tolerances, special features of geometry of the element, as well as effect of deposits on the results of thickness measurements were suggested based on the data of operational control and industry standards. The implemented studies demonstrated efficiency of the developed procedures for pipeline welding zones. Analysis of known and newly developed procedures was performed for bends and ranking of these procedures according to the criterion of “conservatism of evaluation of residual lifetime” was executed. Introduction of correction coefficients allows enhancing conservatism of calculations of lifetime characteristics as compared with calculations performed on the basis of nominal values of thicknesses; the result depends on the type and dimensions of the element, its geometry, as well as on the type of reactor. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.29850 https://doi.org/10.3897/nucet.4.29850 https://nucet.pensoft.net/article/29850/ https://nucet.pensoft.net/article/29850/download/pdf/ en

10.3897/nucet.4.29853 2018-10-17 nucet

Development of seismic protection system for design extension conditions Galiev,Ilnar Chernyaev,Alexey Bibik,Stanislav Seismic protection system maximum design earthquake reactor shutdown seismic station safety VVER data transmission. Nuclear Energy and Technology 4(1): 43-50 Development of seismic protection system for design extension conditions (SPS DEC) is suggested for enhancing safety of operation of NPP located on the territories with unfavorable seismic conditions. The idea of the system consists of the creation of a network of seismic stations arranged at a certain distance from the NPP and equipped with data transmission system. In case of detection by seismic sensors of movements of the ground with magnitude of vibrations exceeding a certain preset value, seismic stations transmit over radio-channel a signal indicating exceedance of the setting before the seismic wave reaches the NPP. This allows initiating transition of the reactor to subcritical operation mode prior to the beginning of destruction of equipment and reactor building. Ensuring reliable protected communication is achieved by simultaneous use of three radio-channels arranged in accordance with TETRA, WiMAx and LTE standards, as well as by the application of appropriate cryptography, authentication and data protection methods for preventing data corruption. Analysis was performed for determining optimal distance between seismic stations and the NPP and optimal number of these stations was determined for the following two options of arrangement of seismic stations: radial arrangement surrounding the NPP and arrangement along the direction towards the place with the highest probability of earthquake incipience. Layout was suggested of multilevel majoritarian data processing for excluding false triggering of the system. Conclusions were formulated on the enhancement of safety of NPP operation and significant reduction of probability of emergency situations due to the generation of anticipatory signals to reactor shutdown systems in case of earthquakes with intensity exceeding maximum design earthquake. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.29853 https://doi.org/10.3897/nucet.4.29853 https://nucet.pensoft.net/article/29853/ https://nucet.pensoft.net/article/29853/download/pdf/ en

10.3897/nucet.4.29855 2018-10-17 nucet

About chemical form and binding energy of 14C in irradiated graphite of uranium-graphite nuclear reactors Bespala,Evgeny Pavliuk,Alexander Zagumennov,Vladimir Kotlyarevskiy,Sergey Uranium-graphite reactors irradiated graphite binding energy binding strength radionuclide radioactive carbon processing decontamination. Nuclear Energy and Technology 4(1): 51-56 Issues associated with handling irradiated graphite of uranium-graphite nuclear reactors are examined. It is demonstrated that selection of approaches, methods and means for handling irradiated graphite are determined by the form of occurrence and binding energy of long-lived 14C radionuclide with graphite crystalline lattice. The purpose of the present study is the determination of possible chemical compounds in which 14C can be found and assessment of fastness of its binding in the structure of irradiated graphite. Indigent and foreign experience of handling graphite radioactive wastes was analyzed, calculations and measurements were performed. Information was provided on the channels of accumulation of 14C in the structure of reactor graphite and it was demonstrated that the largest quantities of the radionuclide in question are generated according to the reaction 14N(n, p)14C. Here, most part of radioactive carbon is generated on 14N nuclei found in the form of impurities in non-irradiated graphite and in the composition of gas used for purging nuclear reactor in the process of operation. 14C radionuclide generated according to 14N(n, p)14C nuclear reaction is localized in the near subsurface graphite layer (in the near subsurface layer of pores) at the depth of not more than 50 nm. Analysis was performed of possible chemical compounds which may incorporate radioactive carbon. It was established that the form of occurrence is determined by the operational properties of specific graphite element in the reactor core. 14C binding energy in the structure of irradiated graphite was evaluated and depth of its penetration in the structure was calculated. It was established that selective extraction of this radionuclide is possible only under elevated temperatures in weakly oxidizing environment which is explained by the binding energy reaching up to 800 kJ/mole in the process of chemical sorption of 14C on the surface of graphite and depth of its occurrence equal to ~ 70 nm in the course of ion implantation. It was demonstrated that radioactive carbon generated according to 13C(n, γ)14C nuclear reaction is uniformly distributed among graphite elements and possesses binding energy ~477 kJ/mole. Its selective extraction is possible only under the condition of destruction of graphite crystalline lattice and organization of the process of isotopic separation. The obtained results allow recommending the most efficient methods of handling irradiated graphite during decommissioning uranium-graphite reactors. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.29855 https://doi.org/10.3897/nucet.4.29855 https://nucet.pensoft.net/article/29855/ https://nucet.pensoft.net/article/29855/download/pdf/ en

10.3897/nucet.4.29858 2018-10-17 nucet

Converting ENDF libraries into relational format Yuferov,Anatoliy ENDF libraries conversion relational format. Nuclear Energy and Technology 4(1): 57-63 The article considers the issues of converting the ENDF format systems of constants to relational databases. This conversion can become one of the tools facilitating the development and operation of factual information, techniques and algorithms in the field of nuclear data and, therefore, increasing the efficiency of the corresponding computational codes. The work briefly examines an infological model of ENDF libraries. The possible structure of tables of the corresponding relational database is described. The proposed database schema and the form of tables take into account the presence of both single and multiple properties of the isotopes under consideration. Consideration is given to the difference in organizational requirements for transferring constants from relational tables to programs and performing a visual analysis of data in tables by a physicist-evaluator. The conversion algorithms and results are described for the ROSFOND-A and ENDF/B-VII.1 libraries. It is shown that performing calculations directly in the DBMS environment has its advantages in terms of simplifying programming and eliminating the need to solve a number of problems on data verification and validation. Possible approaches are indicated to ensure operation of inherited software together with nuclear data libraries in the relational format. Some terminological refinements are proposed to facilitate constructing an infological model for ENDF format. The conversion programs and the ENDF/B-VII.1 library in the relational format are available on a public site. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.29858 https://doi.org/10.3897/nucet.4.29858 https://nucet.pensoft.net/article/29858/ https://nucet.pensoft.net/article/29858/download/pdf/ en

10.3897/nucet.4.29859 2018-10-17 nucet

Nuclear perspectives at exhausting trends of traditional energy resources Uliyanin,Yury Kharitonov,Vladimir Yurshina,Daria Nuclear energy fuel burn-up natural uranium nonrenewable conventional energy resources hydrocarbons production dynamics production rate production peak nuclear breeders. Nuclear Energy and Technology 4(1): 13-19 For the first time the analytical relationship was established between the nuclear energy generation worldwide and supply of NPPs with natural uranium, as conventional resources are expected to deplete by the end of this century. Forecast results include the dynamics of a potential increased shortage of conventional energy resources, such as hydrocarbon fuels (coal, oil, natural gas) and natural uranium, in the course of time due to a growing energy demand (at the rate of 1 to 2% per year), on the one hand, and the depletion of nonrenewable resources, on the other hand. The forecast is based on the current geological data on extractable hydrocarbon and uranium resources, and a mathematical model for the dynamics of nonrenewable resources production. The forecast shows that, with the present-day paradigm of handling the produced conventional energy sources, the reserves of these will be significantly depleted by the end of this century, and their production peaks are expected to be reached by the mid-century. In the event of state-of-art NPP designs, the dynamics of the installed capacity will follow the dynamics of the natural uranium depletion, and the NPP contribution to the supply of energy for the needs of humankind will go down while increasing at the same time the total shortage of conventional energy sources. By 2100, however, the contribution of nuclear power (based on thermal neutrons) to primary sources may reach 10%, since hydrocarbons will be depleted at a higher rate than uranium. Meanwhile, this amount of nuclear energy will be negligible, as compared to the demand for primary energy, after the 2040s even at the smallest possible rate of growth in demand (1%/year). A growing spread between the increasing energy demand and the decreasing supply of exhaustible conventional energy resources necessitates the evolution of nuclear fuel breeding (breeding of 239Pu from 238U and, possibly, 233U from 232Th) no later than the 2030s. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.29859 https://doi.org/10.3897/nucet.4.29859 https://nucet.pensoft.net/article/29859/ https://nucet.pensoft.net/article/29859/download/pdf/ en

10.3897/nucet.4.29837 2018-10-18 nucet

Solution of neutron-transport multigroup equations system in subcritical systems Shamanin,Igor Bedenko,Sergey Nesterov,Vladimir Lutsik,Igor Prets,Anatoly Subcritical system neutron distribution function neutron transport multigroup diffusion approximation. Nuclear Energy and Technology 4(1): 79-85 An iteration method has been implemented to solve a neutron transport equation in a multigroup diffusion approximation. A thermoelectric generator containing plutonium dioxide, used as a source of thermal and electric power in spacecraft, was studied. Neutron yield and multigroup diffusion approximation data was used to obtain a continuous and group distribution of neutron flux density spectra in a subcritical multiplying system. Numerical multigroup approaches were employed using BNAB-78, a system of group constants, and other available evaluated nuclear data libraries (ROSFOND, BROND, BNAB, EXFOR and ENDSF). The functions of neutron distribution in the zero iteration for the system of multigroup equations were obtained by approximating an extensive list of calculated and experimental data offered by the EXFOR and ENDSF nuclear data libraries. The required neutronic functionals were obtained by solving a neutron transport equation in a 28-group diffusion approximation. The calculated data was verified. The approach used is more efficient in terms of computational efforts (the values of the neutron flux density fractions converge in the third iteration). The implemented technique can be used in nuclear and radiation safety problems. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.29837 https://doi.org/10.3897/nucet.4.29837 https://nucet.pensoft.net/article/29837/ https://nucet.pensoft.net/article/29837/download/pdf/ en

10.3897/nucet.4.30387 2018-10-18 nucet

Development of the technique for determination the rate of oxidation of structural steels in heavy liquid metal coolants Ivanov,Konstantin Ivanov,Konstantin Lavrova,Olga Salaev,Sergey Askhadullin,Radomir Mass transfer coolant lead lead-bismuth structural steel chemical-spectral analysis oxidation of steels oxide film oxygen oxygen activity sensor oxidation kinetics. Nuclear Energy and Technology 4(1): 73-78 The article considers the main methods for studying the process of oxidation of structural steels and evaluating their corrosion resistance in heavy liquid metal coolants (HLMC) under static and dynamic conditions. It is shown that the main disadvantage of these methods is the impossibility of evaluating the results in real time. The authors propose a new method for the experimental determination of the oxidation rate of steels in molten HLMCs, which makes it possible to measure the reaction rate without depressurizing the plant, by periodically injecting air doses of a known volume and monitoring its response time. As additional data in the research methodology, the results of a chemical-spectral analysis of the coolant and slags were provided after the experimental campaign completion as well as a metallographic analysis of steel samples to determine the oxide coating thickness and its comparison with the calculated value for the integral oxygen assimilation by the system. To implement the methodology, a laboratory facility was proposed, equipped with an oxygen thermodynamic activity (TDA) sensor of the RF-IPPE design. The sensor is certified by Gosstandart of Russia (certificate RU. 31.002 A No. 15464), registered in the State Register of Measuring Instruments (No. 25282-03) and approved for use in the Russian Federation. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30387 https://doi.org/10.3897/nucet.4.30387 https://nucet.pensoft.net/article/30387/ https://nucet.pensoft.net/article/30387/download/pdf/ en

10.3897/nucet.4.29844 2018-10-18 nucet

Estimation influence of boric acid drop entrainment to its accumulation in the VVER reactor in the case of accident Morozov,Andrej Pityk,Anna Ragulin,Sergej Sahipgareev,Azamat Soshkina,Aleksandra Shlyopkin,Aleksandr VVER emergency operation mode boric acid accumulation droplet entrainment thermal physics properties of boric acid solutions density viscosity thermal conductivity. Nuclear Energy and Technology 4(1): 65-71 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.29844 https://doi.org/10.3897/nucet.4.29844 https://nucet.pensoft.net/article/29844/ https://nucet.pensoft.net/article/29844/download/pdf/ en

10.3897/nucet.4.30524 2018-11-26 nucet

Components of small and medium sized HLMC reactor plant circuits Beznosov,Aleksander Bokova,Tatyana Bokov,Pavel Small and medium sized reactor plant key components engineering solutions heavy liquid metal coolants intercircuit SG break Nuclear Energy and Technology 4(2): 87-92 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.). info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30524 https://doi.org/10.3897/nucet.4.30524 https://nucet.pensoft.net/article/30524/ https://nucet.pensoft.net/article/30524/download/pdf/ en

10.3897/nucet.4.30525 2018-11-26 nucet

Computational model and physical and technical factors determining the plutonium proliferation resistance Kulikov,Yevgeny Kulikov,Gennady Apse,Vladimir Shmelev,Anatoly Geraskin,Nikolay Plutonium plutonium-238 proliferation resistance nuclear explosive device explosive cryogenic temperatures Nuclear Energy and Technology 4(2): 93-97 Since the closed nuclear fuel cycle suggests that plutonium is extracted from irradiated fuel and is recycled in nuclear reactors as part of the loaded fuel, proliferation resistance of fissile materials (plutonium) is becoming a problem of a practical significance. It is important to understand to what extent the physical and technical properties of fissile materials are capable to prevent these from being diverted to nonenergy uses. This paper considers the term ”proliferation resistance” from a physical and technical point of view with no measures taken for the physical protection, accounting and control of nuclear materials. Thus, proliferation resistance of plutonium means that it is technically impossible to fabricate a nuclear explosive device (NED) of the implosion type due to the overheating of the device’s components and the resultant NED failure. The following conclusions have been made. The assessment of the plutonium proliferation resistance is not justified where it relies on the analysis of an implosion-type NED excluding the use of modern heat-resistant and heat-conducting chemical explosives (CE) which are inaccessible. Consideration of the asymptotic temperature profile in the NED components is not justified enough for the development of plutonium proliferation resistance recommendations. No options enabling the slowdown of the NED warm-up process have been exhausted for analyzing the physical and technical factors that determine the proliferation resistance of plutonium. General conclusion. The underlying rationale in a fundamental monograph by Dr. G. Kessler proved to be insufficiently valid, which has led to an unfounded inference as to the status of the plutonium proliferation resistance. The development of the procedures used and other factors taken into account are expected to increase the requirements to the content of the 238Pu isotope in plutonium for ensuring its proliferation resistance. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30525 https://doi.org/10.3897/nucet.4.30525 https://nucet.pensoft.net/article/30525/ https://nucet.pensoft.net/article/30525/download/pdf/ en

10.3897/nucet.4.30527 2018-11-26 nucet

Lead reactor of small power with metallic fuel Khorasanov,Georgiy Samokhin,Dmitriy Zevyakin,Aleksandr Zemskov,Yevgeniy Blokhin,Anatoliy Fast reactor hard neutron spectrum metallic uranium-plutonium fuel natural lead coolant americium-241 Nuclear Energy and Technology 4(2): 99-102 The possibility for obtaining a hard neutron spectrum in small reactor cores has been considered. A harder spectrum than spectra in known fast sodium cooled and molten salt reactors has been obtained thanks to the selection of relatively small core dimensions and the use of metallic fuel and natural lead (natPb) coolant. The calculations for these compositions achieve an increased average neutron energy and a large fraction of hard neutrons in the spectrum (with energies greater than 0.8 MeV) caused by a minor inelastic interaction of neutrons with the fuel with no light chemical elements and with the coolant containing 52.3% of 208Pb, a low neutron-moderating isotope. An interest in creating reactors with a hard neutron spectrum is explained by the fact that such reactors can be practically used as special burners of minor actinides (MA), and as isotope production and research reactors with new consumer properties. With uranium oxide fuel (UO2) substituted by metallic uranium-plutonium fuel (U-Pu-Zr), the reactors under consideration have the average energy of neutrons and the fraction of hard neutrons increasing from 0.554 to 0.724 MeV and from 18 to 28% respectively. At the same time, the one-group fission cross-section of 241Am increases from 0.359 to 0.536 barn, while the probability of the 241Am fission increases from 22 to 39%. It is proposed that power-grade plutonium resulting from regeneration of irradiated fuel from fast sodium cooled power reactors be used as part of the fuel for future burner reactors. It contains unburnt plutonium isotopes and some 1% of MAs which transmutate into fission products in the process of being reburnt in a harder spectrum. This will make it possible to reduce the MA content in the burner reactor spent fuel and to facilitate so the long-term storage conditions for high-level nuclear waste in dedicated devices. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30527 https://doi.org/10.3897/nucet.4.30527 https://nucet.pensoft.net/article/30527/ https://nucet.pensoft.net/article/30527/download/pdf/ en

10.3897/nucet.4.30662 2018-11-26 nucet

ROCOCO: A constants supply system for Monte-Carlo reactor calculation Zherdev,Gennady Kislitsyna,Tamara Nikolayev,Mark Nuclear data radiation field Monte-Carlo calculation combination of detailed and group descriptions of cross-section energy dependence Nuclear Energy and Technology 4(2): 103-109 The ROCOCO system (Kislitsyna and Nikolayev 2016) is designed to supply constants for Monte-Carlo calculations of both neutron fields and the gamma fields they generate. The initial database of nuclear data used in the system is the Russian national library of evaluated neutron data (ROSFOND) (Nikolayev 2006, Zabrodskaya et al. 2007, RUSFOND 2017). ROCOCO is specific in that it enables the estimator to optimize the level of detail when describing neutron cross-section energy dependences. Cross-sections of key fuel, structural and coolant materials can be described in such detail as the evaluated data permits; cross-sections of secondary nuclides (minor actinides, fission products, etc.) can be described in a 299-group BNAB approximation (Manturov et al. 1996) with regard for the resonance self-shielding by subgroup method or without regard for self-shielding altogether. The energy dependence of gamma rays is described in a 127-group P5 approximation (Koshcheyev et al. 2014). Optimizing the level of detail makes it possible to reduce to a great extent the counting time with no major effect on the result and its error. Where desired, in the process of calculating the energy release in neutron reactions or in gamma-quanta formation matrices, contributions from the decay of radionuclides formed in these reactions (with a half-life of less than three years) can be taken into account. The energy dependence of the elastic scattering anisotropy is described in detail, or in the event of a group or subgroup description of cross-sections, by defining 33 boundaries of 32 equiprobable cosine intervals of the scattering angle. The thermalization effects in calculations of neutron fields are taken into account either in an ideal gas approximation or using 72-group thermalization matrices built based on thermalization files contained (if any) in the ROSFOND library. It should be noted that the system contains descriptions of detailed dependences of elastic scattering cross-sections and angular distributions on all multi-isotope elements; the relationship between the scattering angle and the energy loss in this case is determined with the use of the energy-dependent effective atomic weight. The system’s programs are written in the FORTRAN language. The system is easily integrated in Monte-Carlo codes. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30662 https://doi.org/10.3897/nucet.4.30662 https://nucet.pensoft.net/article/30662/ https://nucet.pensoft.net/article/30662/download/pdf/ en

10.3897/nucet.4.30663 2018-11-26 nucet

Power coefficient of reactivity: definition, interconnection with other coefficients of reactivity, evaluation of results of transients in power nuclear reactors Kazansky,Yury Slekenichs,Yanis Nuclear power plants power reactivity coefficient temperature reactivity coefficients nuclear reactor dynamics Nuclear Energy and Technology 4(2): 111-118 It is assumed by the authors of the present paper that with growing contribution of nuclear power in the production of electricity, nuclear power plants will be used to a higher degree in a manoeuvrable mode of operation rather than in the base-load mode. In other words, change of power from the nominal level to that of coverage of auxiliary loads will be becoming quite common and not so rare event as scheduled reactor shutdowns for fuel reloading or preventive works. There exist well-known problems in the use of nuclear reactors in the manoeuvrable operation mode, which include the task shared by all types of nuclear reactors. It is advisable to have a unified indicator weakly power-dependent and fairly easy to measure, which would make it possible to formulate the judgement about the nature of the transient processes within the entire power range and to assess the reactivity required for changing the power level by the preset value. Power reactivity coefficient (PRC) can be used as such indicator. Analysis was made of existing definitions and understanding of PRC in relevant references. It turned out that there is no generally accepted definition of the PRC. Based on the performed study, the following definition was suggested: the PRC is the ratio of the low reactivity introduced into the reactor to the power increment at the end of the transient process. It is assumed here that variation of reactivity is dependent on the energy released in nuclear fission but is not related to the changes of reactivity induced by feedback signals in the automatic reactor power control system. Analysis of the relationship between the PRC and temperature coefficients and technological parameters associated with the steady-state control program was performed taking the above suggested definition into account. PRC calculations were performed using the simplest model of VVER-1000 type power reactor. It was found that PRC is weakly power-dependent. The purpose of the present study is to investigate dependence of PRC on the temperature reactivity effects and on the technological parameters associated with the steady-state control program of the power unit, using the example of VVER-1000. Effects of PRC on the static and dynamic power reactor operation modes are analyzed. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30663 https://doi.org/10.3897/nucet.4.30663 https://nucet.pensoft.net/article/30663/ https://nucet.pensoft.net/article/30663/download/pdf/ en

10.3897/nucet.4.30677 2018-11-26 nucet

Small nuclear power plants for power supply in arctic regions: assessment of spent nuclear fuel radioactivity Naumov,Vadim Gusak,Sergey Naumov,Andrey Arctic regions of Russia small nuclear power plants reactors spent nuclear fuel fuel cycle radioactivity Nuclear Energy and Technology 4(2): 119-125 The purpose of the present study is the investigation of mass composition of long-lived radionuclides accumulated in the fuel cycle of small nuclear power plants (SNPP) as well as long-lived radioactivity of spent fuel of such reactors. Analysis was performed of the published data on the projects of SNPP with pressurized water-cooled reactors (LWR) and reactors cooled with Pb-Bi eutectics (SVBR). Information was obtained on the parameters of fuel cycle, design and materials of reactor cores, thermodynamic characteristics of coolants of the primary cooling circuit for reactor facilities of different types. Mathematical models of fuel cycles of the cores of reactors of ABV, KLT-40S, RITM-200M, UNITERM, SVBR-10 and SVBR-100 types were developed. The KRATER software was applied for mathematical modeling of the fuel cycles where spatial-energy distribution of neutron flux density is determined within multi-group diffusion approximation and heterogeneity of reactor cores is taken into account using albedo method within the reactor cell model. Calculation studies of kinetics of burnup of isotopes in the initial fuel load (235U, 238U) and accumulation of long-lived fission products (85Kr, 90Sr, 137Cs, 151Sm) and actinoids (238,239,240,241,242Pu, 236U, 237Np, 241Am, 244Cm) in the cores of the examined SNPP reactor facilities were performed. The obtained information allowed estimating radiation characteristics of irradiated nuclear fuel and implementing comparison of long-lived radioactivity of spent reactor fuel of the SNPPs under study and of their prototypes (nuclear propulsion reactors). The comparison performed allowed formulating the conclusion on the possibility in principle (from the viewpoint of radiation safety) of application of SNF handling technology used in prototype reactors in the transportation and technological process layouts of handling SNF of SNPP reactors. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30677 https://doi.org/10.3897/nucet.4.30677 https://nucet.pensoft.net/article/30677/ https://nucet.pensoft.net/article/30677/download/pdf/ en

10.3897/nucet.4.30771 2018-11-26 nucet

Potential of application of IRT-T research reactor as the solution of the problem of graphite radwaste disposal Pavliuk,Alexander Kotlyarevskiy,Sergey Bespala,Evgeny Bespala,Yuliya Irradiated reactor-grade graphite uranium-graphite reactor research reactor IRT-T Nuclear Energy and Technology 4(2): 127-133 Aspects of handling irradiated graphite during decommissioning uranium-graphite reactors (UGR) of different types were investigated. It was demonstrated that handling reactor graphite is complicated by the presence in the composition of graphite of long-lived radionuclides, especially 14C, which may get entrained in biological cycles since carbon constitutes one of the main components of biological chains. Practical implementation of the process of selective separation of 14С can significantly reduce potential danger represented by graphite radioactive wastes due to the reduction of graphite activity as related to the isotope in question, as well as due to the reduction of the leaching rate by separating 14С isotope which is the most weakly bound within the graphite structure. Conclusion was formulated that analytical measurement methodologies and calculation methods allow reliably estimating only the total quantity of 14C accumulated in graphite, the contribution of 14C accumulation channel from 13C(n, γ)14C reaction, as well as the total contribution of 14N(n, p)14C reaction on nitrogen impurities and on nitrogen contained in purge gas. Method was suggested for estimating the values of contributions of different channels of accumulation on nitrogen impurities and nitrogen contained in purge gas using IRT-T research reactor (Tomsk, Tomsk Region). Parallel irradiation of batches of samples of non-irradiated (fresh) reactor-grade graphite contained in different gaseous media constitutes the basis of the study. Algorithm was suggested for calculating contributions of all channels of 14C accumulation according to the results of measurements to be obtained in the proposed studies. Recommendations were formulated on the use of all brands of graphite applied for manufacturing elements of graphite stacks of uranium-graphite reactors designed in Russia for determining selectively separated fraction of 14C for all types of graphite radioactive wastes by the companies in the RF which operated (are operating) the uranium-graphite reactors. Time of exposure of samples of irradiated graphite in the GEK-4 horizontal experimental channel of the IRT-T reactor was calculated and was found to be equal to ~ 10 days. Methodology was suggested for conducting a series of experiments for determining the values of contributions of 14C accumulation channels in the irradiated reactor graphite. The methodology suggested can be applied for determining fraction of selectively separated 14C in irradiated graphite elements of practically all uranium-graphite nuclear reactors, including reactors operated abroad Russia, under the condition of maintaining carbon dioxide gas atmosphere in one of the irradiated containers. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30771 https://doi.org/10.3897/nucet.4.30771 https://nucet.pensoft.net/article/30771/ https://nucet.pensoft.net/article/30771/download/pdf/ en

10.3897/nucet.4.30774 2018-11-26 nucet

Analysis of a severe beyond design basis accident for the EGP-6 reactor of the Bilibino NPP. Radioactive source term determination Mukhamadeev,Ruben Parafilo,Leonid Baranaev,Yury Suvorov,Albert Severe beyond design basis accident heat exchange crisis fuel damage dynamic fission products yields radioactive source term Nuclear Energy and Technology 4(2): 135-142 Analysis was performed of dynamic phase of severe accident of the EGP-6 reactor of the Bilibino NPP, due to uncontrolled reactivity insertion initiated by withdrawal of two pare of automatic control rods with followed by full failure of reactor emergency protection system. This initial event leads to promt increasing of reactor core power up to 450% of nominal value with short period, coupled with rise of temperature of fuel, pressure and temperature of coolant. These factors lead to crisis of heat exchange with subsequent ruptures tubes of fuel assemblies and coolant blow down into graphite stack. All its lead to rise of pressure in reactor shell and damage of it, outflow of steam-water mixture through up-reactor area to ventilation system, communication corridors and reactor hall and further – to atmospheric release. Transient processes were calculated using code RELAP5/Mod3.2. It was considered stages of processes of fuel damage and evaluated dynamic of a number and degree of damaged fuel assembles. They were grouped on burn-up and for each group it was performed analysis of dynamic of damage values. Further it was considered processes of yield of fission products from damaged fuel with models, based on experimental data on yield of fission products from fuel material, used in assembles of Bilibino NPP fuel type (fuel tubes with steel cladding, where fuel material is grits of uranium dioxide in magnesium), under condition of severe accident, especially performed in SSC IPPE. Transport of fission products with steam and air up to release points was evaluated with models, based on experimental data of fission product transport through graphite stack under conditions of severe accident, also especially performed in SSC IPPE. Evaluation of source term was performed in accordance with accident dynamic and assumed modes of release for conservative and most possible approaches. It was noted good self-protection property of EGP-6 reactor under severe beyond design basis accident condition. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30774 https://doi.org/10.3897/nucet.4.30774 https://nucet.pensoft.net/article/30774/ https://nucet.pensoft.net/article/30774/download/pdf/ en

10.3897/nucet.4.30775 2018-11-26 nucet

Analysis of acoustic leak signals for enhancing sensitivity of control due to the creation of effective diagnostic indicators Leskin,Sergey Shvetsov,Dmitry Trykov,Evgeny Puzakov,Aleksey control of leak tightness of equipment dispersion of acoustic signal modeling coolant leaks method of principal components additional diagnostic indicators pattern identification Nuclear Energy and Technology 4(2): 141-147 Acoustic leak control systems (for instance, SAKT) are used at present for controlling leak tightness of equipment and pipelines, as well as for detecting in timely manner coolant leaks from the primary cooling circuit of nuclear reactor installations (NRI) during operation of power unit on different power levels in the modes of normal operation and during disturbances of normal operation. Time averaged dispersion of acoustic signal is used as the main diagnostic indicator for detecting leaks in these systems. Sensitivity of this indicator is determined by the exceedance by the signal of the preset threshold value which is defined in accordance with the background. Here, background values of acoustic signal depend on the operational modes of the equipment and do not allow in many cases determining coolant leak during early stages of leak development. New approach to the formation of diagnostic indicators for detecting loss of sealing in the circuit during early stage of development of coolant leak is suggested. Methodology for obtaining diagnostic indicators is based on the processing in different frequency bands of acoustic signal accompanying coolant leakage from the pipeline using the method of principal components. Efficiency of the developed methodology of coolant leak detection is illustrated by processing acoustic signals for experimental facility modeling coolant leakage in case of loss of sealing of the circuit. Even in the presence of significant acoustic background sensitivity of the method allows detecting leaks with significantly lower flow rates (up to five times smaller) than the conventional processing of acoustic signals. Implementation of the developed methodology will not require significant expenditures for upgrading already existing leak control systems operated at present on different NPPs. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30775 https://doi.org/10.3897/nucet.4.30775 https://nucet.pensoft.net/article/30775/ https://nucet.pensoft.net/article/30775/download/pdf/ en

10.3897/nucet.4.30777 2018-11-26 nucet

Analysis of mass transfer processes in a reactor during a loss-of-coolant accident Kulikov,Aleksey Lepyokhin,Andrey Polunichev,Vitaly Accident coolant leak coolant level core safe state optimization spillage system reactor hydroaccumulator design analysis Nuclear Energy and Technology 4(2): 149-154 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30777 https://doi.org/10.3897/nucet.4.30777 https://nucet.pensoft.net/article/30777/ https://nucet.pensoft.net/article/30777/download/pdf/ en

10.3897/nucet.4.30779 2018-12-07 nucet

Brittle fracture resistance of reactor pressure vessel steels in the initial state Anosov,Nikolay Skorobogatykh,Vladimir Gordyuk,Lyubov Mikheev,Vasiliy Pogorelov,Yegor Shamardin,Valentin Brittle fracture resistance assessment critical brittleness temperature ductile-to-brittle transition temperature standard deviation conservative estimates of the VVER-1000 vessel life. Nuclear Energy and Technology 4(3): 155-161 The authors investigate the influence of chemical and structural inhomogeneity on the brittle fracture resistance (BFR) of VVER vessel materials in the initial state (without irradiation). It is proposed to replace the brittle fracture resistance assessment using the critical brittleness temperature TC for the BFR assessment using the brittle-viscous transition temperature TT. Consideration was given to calibration charts used for studying the TT dependence on the grain size and heat treatment. A comparison of the TC and TT values in the experimental industrial 15H2NMFA-A steel billets shows that the TC values are significantly lower than the TT values: – at the lower level of conservatism, the difference between TC and TT is 22 °C; – at the upper level of conservatism, this difference is 24 °C. The array data on the critical brittleness temperature and the ductile-to-brittle transition temperature of impact test samples of 15H2NMFAA (for VVER-1000) and 15H2NMFA grade 1 (for VVER-1200) steels were statistically processed. The industrial shell samples were manufactured at the “Energomashspetsstal” plant (Kramatorsk, Ukraine). It was found that, in the metal of VVER-1000 vessel surveillance specimens with the copper content – less than 0.06%, heat treatment has a significant effect on the TT value, which changes from –99 to –28°C; – from 0.07 to 0.12%, heat treatment has a significant effect on the TT value, which changes from –60 to –40°C. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30779 https://doi.org/10.3897/nucet.4.30779 https://nucet.pensoft.net/article/30779/ https://nucet.pensoft.net/article/30779/download/pdf/ en

10.3897/nucet.4.30780 2018-12-07 nucet

Radiation-induced separation and accumulation of electric charge in supercapacitors Stepanov,Vladimir Chernov,Vladimir Parshikov,Yury Lebedev,Viktor Kharanzhevsky,Yevgeny Supercapacitor neutron irradiation radiation-induced electric charge Nuclear Energy and Technology 4(3): 163-166 In current sources with a radioactive isotope (CSRI), nuclear energy is directly converted into electricity due to the separation of electric charges during the decay of radioactive isotopes. It was previously shown that asymmetric supercapacitors can be used as CSRI prototypes if, after being exposed to pulsed reactor irradiation, the electric charge on their plates increases to several coulombs as a result of internal induced activity. In this paper, the electric charge separation and accumulation in supercapacitors were studied directly in the process of neutron irradiation. The study was focused on the electrophysical characteristics of cylindrical supercapacitors with an organic electrolyte produced by JSC “ELEKOND”. A comparison of symmetric and asymmetric supercapacitors showed that an effective charge accumulation occurs in the asymmetric capacitors: it is independent of the neutron flux density and determined by the absorbed radiation dose. The electrical voltage between the plates of a symmetrical supercapacitor with a capacity of 100 F during irradiation up to an absorbed dose of 50 Gy reaches 1.24 mV. When asymmetric supercapacitors are irradiated with the same dose, a significant increase in the potential difference up to 1.15 V is observed during irradiation and for a long time afterwards (1.5·105 s) due to the electric charge redistribution (~ 5·10–3 C) in the electrolyte and carbon particles with the formation of a double electrical layer. The post-radiation increase in the capacity of asymmetric supercapacitors is ~ 5 mF. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30780 https://doi.org/10.3897/nucet.4.30780 https://nucet.pensoft.net/article/30780/ https://nucet.pensoft.net/article/30780/download/pdf/ en

10.3897/nucet.4.27346 2018-12-07 nucet

Axial dispersion and mixing of coolant gas within a separate-effect prismatic modular reactor Said,Ibrahim Moharam,Mahmoud Alexander,Vineet Usman,Shaoib Al-Dahhan,Muthanna Prismatic modular reactor gaseous tracer technique axial dispersion coefficient Peclet number Nuclear Energy and Technology 4(3): 167-178 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.27346 https://doi.org/10.3897/nucet.4.27346 https://nucet.pensoft.net/article/27346/ https://nucet.pensoft.net/article/27346/download/pdf/ en

10.3897/nucet.4.31867 2018-12-07 nucet

A modern data measurement system to study and test thermionic heat to electricity converters Kirillov,Andrey Yarygin,Valeriy Thermionic converter data measurement system volt-ampere characteristic thermocouples pulse mode Nuclear Energy and Technology 4(3): 179-183 Studies and tests are conducted to determine the performance of thermionic nuclear power plants (TNPP) a stage in which is pre-irradiation testing of laboratory thermionic converters (TIC) with flat and cylindrically shaped electrodes using test facilities fitted with automated data measurement systems (DMS). The TIC volt-ampere characteristics (VAC) are measured in the DMS jointly with the measured test section and experimental test facility temperature fields. The structure and the characteristics of a DMS based on products from ICP DAS Co., Ltd are presented. A developed VAC measurement program providing the operator with a convenient graphic interface and enabling adjustment of the measurement parameters has been considered. The VAC recording errors in the process of measurements have been determined using TIC simulators. The error in the VAC diffusion portion on a simulator (with a current of less than 3 A) is not more than 1%. Thanks to the use of modern components, the developed DMS offers extended functional capabilities for measuring the thermocouple signals in an experimental electrophysical test facility. The DMS structure provides for the convenience of scaling (through a larger number of measuring channels) and makes it possible to add modules from other manufacturers. The experience of operating this DMS will be used to develop the DMS for an in-pile test system designed for similar functions. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31867 https://doi.org/10.3897/nucet.4.31867 https://nucet.pensoft.net/article/31867/ https://nucet.pensoft.net/article/31867/download/pdf/ en

10.3897/nucet.4.31857 2018-12-07 nucet

Development and study of a microwave reflex-radar level gauge of the nuclear reactor coolant Melnikov,Vladimir Ivanov,Vadim Teplyashin,Ivan Timonin,Mikhail Microwave reflex-radar level gauge high pressure water coolant measuring probe nuclear reactor power-generating equipment Nuclear Energy and Technology 4(3): 185-190 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31857 https://doi.org/10.3897/nucet.4.31857 https://nucet.pensoft.net/article/31857/ https://nucet.pensoft.net/article/31857/download/pdf/ en

10.3897/nucet.4.31859 2018-12-07 nucet

Quantitative evaluation of the plutonium proliferation resistance Kulikov,Yevgeny Kulikov,Gennady Apse,Vladimir Shmelev,Anatoly Geraskin,Nikolay Plutonium plutonium-238 proliferation resistance nuclear explosive device explosive cryogenic temperatures Nuclear Energy and Technology 4(3): 191-195 The mathematical model presented in (Kulikov et al. 2018) can be used for the quantitative evaluation of the plutonium proliferation resistance. This requires the warm-up process of an implosion nuclear explosive device (NED) with a different structure to be analyzed with respect to various heat removal conditions and the option to be identified in which the NED remains operational for the longest time possible. The fraction of the 238Pu isotope with which, even in this case, the NED will prove to be operational only for quite a short time can be regarded as sufficient for the plutonium with such composition to be considered a proliferation resistant material. The purpose of the paper is to evaluate in quantitative terms the content of 238Pu in plutonium for ensuring its proliferation resistance and to identify the factors which influence significantly this evaluation. The data, procedures and findings from earlier works on the topic, as well as the authors’ own estimates and calculations were used for the study. It has been shown that the important factors involved in the plutonium proliferation resistance evaluation are the NED technology level and the required NED lifetime. Depending on the required lifetime, tougher requirements can be introduced with respect to the 238Pu content both from the standpoint of low-technology and high-technology NEDs. With a lifetime of five hours taken as the guide-mark (a NED is unlikely to be finally assembled, transported and used for such a short time), it is only plutonium containing 55% of 238Pu that can be considered a proliferation resistant fissile material. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31859 https://doi.org/10.3897/nucet.4.31859 https://nucet.pensoft.net/article/31859/ https://nucet.pensoft.net/article/31859/download/pdf/ en

10.3897/nucet.4.31862 2018-12-07 nucet

Reactivity margin evaluation software for WWR-c reactor Belyavtsev,Ivan Starkov,Sergey WWR-c reactor reactivity margin artificial neural networks software package generalized approximation theorem Nuclear Energy and Technology 4(3): 197-201 The WWR-c reactor reactivity margin can be calculated using a precision reactor model. The precision model based on the Monte Carlo method (Kolesov et al. 2011) is not well suited for operational calculations. The article describes the work on creating a software package for preliminary evaluations of the WWR-c reactor reactivity margin. The research has confirmed the possibility of using an artificial neural network to approximate the reactivity margin based on the reactor core condition. Computational experiments were conducted on training the artificial neural network using the precision model data and real reactor measured data. According to the results of the computational experiments, the maximum relative approximation error ∆k/k for fuel burnup was 3.13 and 3.56%, respectively. The mean computation time was 100 ms. The computational experiments showed it possible to construct the artificial neural network architecture. This architecture became the basis for building a software package for evaluating the WWR-c reactor reactivity margin – REST API based web-application – which has a convenient user interface for entering the core configuration. It is also possible to replenish the training sample with new measurements and train the artificial neuron network once again. The reactivity margin evaluation software is ready to be tested by the WWR-c reactor personnel and to be used as a component of the automated reactor refueling system. With minor modifications, the software package can be used for reactors of other types. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31862 https://doi.org/10.3897/nucet.4.31862 https://nucet.pensoft.net/article/31862/ https://nucet.pensoft.net/article/31862/download/pdf/ en

10.3897/nucet.4.31863 2018-12-07 nucet

Application of diffusion approximation in the calculations of reactor with cavities Seleznev,Evgeny Bereznev,Valery Fast breeder reactor diffusion approximation cavities calculations. Nuclear Energy and Technology 4(3): 203-209 The importance of calculation of radiation fields inside in-reactor cavities is associated with the necessity to simulate the emergency modes in fast breeder reactors (FBR), as well as reactor states with different coolant levels in special dedicated channels of passive feedback devices in lead-cooled fast reactors (LFR) of BREST type or in sodium cavities in sodium-cooled fast reactors (SFR). The Last Flight (LF) method (Bell and Glesston 1974, Davison 1960, DOORS3.2 1988, Mynatt et. al. 1969, Rhoades and Childs 1988, Rhoades and Sipmson 1997, SCALE 2009, Voloschenko et. al. 2012), or the method of the unscattered component is widely known and is commonly used in computer codes based on the method of spherical harmonics for obtaining solution in a gas medium at a certain distance from the calculated volume domain (DORT (Rhoades and Childs 1988), TORT (Rhoades and Sipmson 1997) and others (SCALE 2009)). The practice of its application (DOORS3.2 1988) demonstrated that acceptable results are achievable at considerable distances from the surface separating dense and gas media (more than two meters). Obtaining high-quality solution is not guaranteed for cavities within the calculation area. In addition, it is desirable to implement the cavities calculation methodology within the framework of the approximations used in reactor calculations introducing certain specific features. In particular isotropy of the neutron flux density and the necessity of forced introduction of a “conditional” calculation cell on the boundary surface of the void cavity are assumed in the diffusion approximation. If the LF method is oriented on the connection of the source point with the detection point, then it is necessary to determine in the calculation of neutron field in the cavities the neutrons escaping the surface area of the source and neutrons reaching a certain surface area of the cavity. In order to solve the problem, the authors suggested using the approximate solution presented in the paper. Thus, an algorithm for calculation of in-reactor cavities using the diffusion approximation was developed and implemented by the authors. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31863 https://doi.org/10.3897/nucet.4.31863 https://nucet.pensoft.net/article/31863/ https://nucet.pensoft.net/article/31863/download/pdf/ en

10.3897/nucet.4.31865 2018-12-07 nucet

Capabilities of gamma-spectrometry methods for on-line monitoring of nitride SNF pyrochemical reprocessing Kanashov,Boris Smirnov,Valery Kadilin,Vladimir Ibragimov,Renat Dedenko,Grigory Vlasik,Konstantin Rudenko,Vladimir Glagovsky,Eduard Lupar,Yevgeny Poletov,Grigory Lomtev,Yevgeny Smirnov,Aleksandr Khrunov,Vladimir Pyrochemical process mixed nitride uranium-plutonium spent nuclear fuel (MUPN SNF) pyroelectrochemical refining fission products on-line monitoring of nuclear materials gamma-spectrometry experimental facility Nuclear Energy and Technology 4(3): 211-215 The paper reports the first test results for detectors of various types and equipment of gamma-spectrometry channels under external radiation originating from pyrochemical reprocessing of spent mixed nitride uranium-plutonium (MNUP) fuel. Testing was carried out on a solid-state detector with a CdZnTe crystal, a scintillation detector with a LaBr3crystal, and an ionization chamber based on compressed xenon. Simulated external gamma-radiation was created by means of a Co-based scattered gamma-radiation source. The paper also describes an experimental facility and a measurement technique, and presents the facility testing results for the above three detectors. The solid-state detector was proved to have the best performance. However, achieving the design characteristics of the gamma-spectrometry channel requires new solutions for protection and collimation of gamma-radiation produced by a real MNUP SNF reprocessing facility. What is meant here is the influence of the detectors’ geometry on the configuration of the protective collimator which is proposed to be used in real conditions. Thus, if a Xe-based detector is used, the calculated mass of the protective collimator is 900 kg, while it is possible to use less massive protection for the other detectors. In addition, when manufacturing neutron shielding for detectors based on CdZnTe and LaBr3, it is necessary to consider the neutron radiation factor in MNUP SNF processing. It is possible to surround the collimator with a moderating layer (for example, polyethylene) and create inside it a skin from a thermal neutron absorber (for example, based on cadmium). info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31865 https://doi.org/10.3897/nucet.4.31865 https://nucet.pensoft.net/article/31865/ https://nucet.pensoft.net/article/31865/download/pdf/ en

10.3897/nucet.4.31861 2018-12-07 nucet

ROCOCO system of combined neutron constants – current status and results of testing using geometrical module of the MMK code Zherdev,Gennady Kislitsyna,Tamara Nikolayev,Mark New ROCOCO combined system of neutron constant preparation further refinement introduction in the practice of calculations Monte-Carlo method results of comparative calculations potential of development Nuclear Energy and Technology 4(3): 217-222 Results of studies aimed at the further refinement of the ROCOCO system (routine for calculation and organization of combined constants including cross-sections in group and subgroup representation with detailed description of energy dependence of neutron cross-sections) (Zherdev et al. 2018, Kislitsina and Nikolaev 2016) are presented in the paper. Inclusion of this system as a physical module into a set of Monte Carlo calculation codes with OOBG geometric module from the MMK code (Zherdev et al. 2003) is discussed. OOBG module is designed for calculation of neutron multiplication systems with heterogenous cores arranged as hexagonal grids with different degrees of complexity. The name ROCOCO-MMK was assigned to the complex. Results of testing the complex in the calculations of multi-zone neutron multiplication systems (including those with zones containing neutron moderator, zones with close composition but with different temperature, etc.) are described. Accounting for the dependence of constants for one and the same nuclide in the zones with different compositions and temperatures required substantial modernization of routines for preparation of constants for calculation described in (Zherdev et al. 2018). Algorithm for preparation of subgroup constants was modified, methodology for taking into account resonance self-screening of cross-sections within the range of unresolved resonances was improved, and other changes were introduced in the process of this modernization. Results of calculations are compared with data obtained using the MCNP-5 precision program (MCNP 1987), which is linked to the same library of evaluated neutron data ROSFOND as that used in ROCOCO. The ROCOCO-MMK includes procedures for registering different neutron flux functionals (also based on ROCOCO data), which allowed including it in the SCALA computation complex (Zherdev et al. 2003, Zherdev 2005), and performing step-by-step calculation of evolution of fuel nuclide composition during the fuel residence campaign. Directions for further development of the system are outlined in conclusion and, in particular, some possibilities of using the created software for further improvement of methods for preparation of few-group constants for calculations in diffusion approximation are examined. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31861 https://doi.org/10.3897/nucet.4.31861 https://nucet.pensoft.net/article/31861/ https://nucet.pensoft.net/article/31861/download/pdf/ en

10.3897/nucet.4.31871 2018-12-13 nucet

Comparison of two key analysis methods for the seismic stability of equipment on the example of a ventilation unit Sobolev,Artem Danilov,Pavel Zevyakin,Aleksandr Kurkov,Sergej Seismic stability linear spectral method dynamic analysis method reduced stresses accelerogram finite element model Nuclear Energy and Technology 4(4): 243-249 Results of calculation seismic resistance analysis of light equipment of nuclear power plants performed on the example of a ventilation unit using two most common analytical techniques - linear spectral analysis and direct dynamic methods - are discussed. The basic concepts, assumptions and limitations of the linear spectral method are described. Examples are given of specific calculation cases when the method in question is not applicable in the generally accepted formulation. In particular, the phase difference and, possibly, accelerations (displacements) must be taken into consideration in the calculations of extended spatial structures for mutually remote boundary conditions. Another example are the reservoirs not completely filled with liquids. In such case waves may be formed in the liquid and taking them into account is not possible in the linear spectral method. Specific features are examined of application of the dynamic analysis method including the input data, approaches and methodologies required for synthesizing the calculated accelerograms. A sequence of operations performed during synthesizing calculated accelerograms is provided, materials are provided containing the description of the mathematical apparatus applied for deriving the final mathematical relations for calculating response spectra and the calculation relations as such are given. The concept of the damping coefficient is explained, its influence on the calculated results and the approaches to its determination are demonstrated. Options with complete absence of damping and with absolute damping are discussed. A real ventilation set applied in active ventilation systems of nuclear power plants was accepted as the test model. Results calculated for the detailed finite-element model of the ventilation unit using the Zenith-95 software package are presented. These results include the distribution of the calculated reduced stresses. Analysis of the results obtained using the two methods demonstrated overestimation of calculated results by the linear spectral method as compared to those obtained by the dynamic analysis method, which means that the former method underestimates the equipment’s resistance to seismic effects. In addition, the dynamic method shows additional areas in the ventilation unit where significant reduced stresses are found while the linear spectral method ignores these areas. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31871 https://doi.org/10.3897/nucet.4.31871 https://nucet.pensoft.net/article/31871/ https://nucet.pensoft.net/article/31871/download/pdf/ en

10.3897/nucet.4.31873 2018-12-13 nucet

Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh Hossain,Ismail Akbar,Mohammad Velkin,Vladimir Shcheklein,Sergey Seismic hazard parameters IAEA seismic acceleration and damage peak ground acceleration Nuclear Energy and Technology 4(4): 251-256 Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31873 https://doi.org/10.3897/nucet.4.31873 https://nucet.pensoft.net/article/31873/ https://nucet.pensoft.net/article/31873/download/pdf/ en

10.3897/nucet.4.31875 2018-12-13 nucet

On the necessity and the role of descriptors of neutron activated structural and shielding materials of nuclear installations for future decommissioning Bylkin,Boris Engovatov,Igor Kozhevnikov,Alexey Sinyushin,Dmitry Nuclear facilities NPP unit decommissioning neutron induced activity radioactive waste radiation shielding concrete descriptor Nuclear Energy and Technology 4(4): 257-262 Existing situation in nuclear industry is characterized with simultaneous development of the following two processes: design and construction of new generation of nuclear installations and decommissioning of installations of older generations. Significant amounts of radioactive wastes generated during the decommissioning phase are determined both for the first and the second types of installations by the induced activity of neutron irradiated structural and shielding materials. Concentration of the so-called radioactivity-hazardous nuclides in primary building and construction materials is the most important characteristics determining the resulting levels of induced activity. Values of these concentrations for the same type of material extracted from different geological deposits may differ by one or two orders of magnitude. Information about concentrations of radiation-hazardous elements in radiation shielding materials is fragmented and, as a rule, unsuitable for practical application. The purpose of the present study was to substantiate the necessity of compiling and recording the data on the concentrations of radioactivity-hazardous nuclides for building and structural materials for nuclear installations during the phases of design, operation and decommissioning. Three types of shielding concrete compositions were selected for the investigation. Concentrations of radioactivity-hazardous nuclides were mainly obtained by neutron activation technique. Neutron transport calculations were performed in one-dimensional cylindrical geometry at the core mid-plane according to usual core-vessel-shielding model of modern VVER reactor unit including 2-m thick concrete shield. Both transport and activation calculations were undertaken using modules of SCALE system. The obtained results allow estimating neutron-induced activation levels in the material as the function of irradiation time, amounts and categories of radioactive waste and their evolution during the decay time from 1 to 100 years. It was established that neutron-induced activity of shielding concrete strongly depends on the actual concentrations of radioactivity-hazardous nuclides in the concrete including ‘trace’ concentrations (other factors being the same). It was also shown that failure to take such concentrations into account may lead to the underestimation of neutron-induced activation levels and amounts of radioactive wastes and their category. The obtained results confirmed the necessity of compiling and maintaining data records on the concentrations of radioactivity-hazardous nuclides for materials used in structural and shielding materials of nuclear installations. Proposals were formulated on the potential accumulation of information, composition and formatting of descriptors of chemical composition of shielding and structural materials of nuclear installations. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31875 https://doi.org/10.3897/nucet.4.31875 https://nucet.pensoft.net/article/31875/ https://nucet.pensoft.net/article/31875/download/pdf/ en

10.3897/nucet.4.31889 2018-12-13 nucet

A technique for detection of WWER fuel failures by activity of Xe radionuclides during reactor operation Kalinichev,Peter Evdokimov,Igor Likhanskii,Vladimir WWER fuel rod fuel failure fission products technique coolant activity iodine radionuclides radioactive noble gases Nuclear Energy and Technology 4(4): 263-270 Fuel failures during operation of Nuclear Power Plants (NPPs) may lead to substantial economic losses. Negative effects of reactor operation with leaking fuel in the core may be reduced if fuel failures are detected in due time of the cycle. At present time, the ratio of the normalized release rates of 131I and 134I is used to detect fuel failures in WWERs during steady state operation. However, based on the activity of iodine radionuclides, it is not always possible to detect the fuel failure. This situation may occur in case of a small defect in cladding of a leaking fuel rod or for high burnup fuel if the defect is overlapped by the surface of the fuel pellet. If it is so, fuel deposits may be the dominant contributor to iodine activity, and the fuel failure may not be noticeable. In PWRs, fuel failures are detected by activity of radioactive noble gases. Noble gases are not adsorbed on cladding inner surface, as distinct from iodine radionuclides. Release of noble gases from the leaking fuel rod may be considerable even though defect in cladding is small. In this paper, a technique is proposed for detection of fuel failures at WWER reactors by activity of radioactive noble gases in the primary coolant. It is shown that radioactive noble gases may be a more sensitive indicator of fuel failures than iodine radionuclides. Detection of fuel failures is based on monitoring of the ratio between 133Xe and 135Xe activity. Some examples of practical applications are given. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31889 https://doi.org/10.3897/nucet.4.31889 https://nucet.pensoft.net/article/31889/ https://nucet.pensoft.net/article/31889/download/pdf/ en

10.3897/nucet.4.31891 2018-12-13 nucet

Systemic competitiveness of nuclear energy sources Usanov,Vladimir Nuclear power business competitiveness electricity cost capital market funding mechanisms computational model financially integrated system Nuclear Energy and Technology 4(4): 271-277 Possibilities are analyzed for improving the commercial attractiveness of nuclear electricity generation in market conditions. A model is presented in which a financially integrated electricity generating system comprising several units of one technological type, rather than a single unit, is subject to an economic analysis. Issues have been considered involved in the calculation of the electricity cost in such systems and their construction. It has been shown that the calculated unit cost of the electricity generated in a financially integrated nuclear energy system with the number of units being more than one, provided it is financed by shareholders and creditors, can be lower as compared with the cost of the electricity generated by power units, not integrated economically, of the same capacity under the same investment conditions. The effect is achieved thanks to the short-term crediting component in the electricity cost the funds on which can be returned, at a time, for a smaller number of units (even for only one), as electricity is produced by all units in the system. The results of the calculations for nuclear energy sources and combined-cycle plants using the developed model make it possible to conclude that the switch from economic models of individual nuclear units to models of integrated energy systems can bring the calculated economic performance of nuclear power closer to (or better) the performance of fossil-fuel energy sources. If achieved, this may increase the commercial attractiveness of nuclear power and contribute to a growth in the public and private investments in nuclear power business. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31891 https://doi.org/10.3897/nucet.4.31891 https://nucet.pensoft.net/article/31891/ https://nucet.pensoft.net/article/31891/download/pdf/ en

10.3897/nucet.4.31892 2018-12-13 nucet

The SIMCO containment code applied to modeling hydrogen distribution in containments of nuclear power facilities Dorovskikh,Vyacheslav Dorokhovich,Sergey Zajtsev,Aleksey Levchenko,Valery Leonov,Igor SIMCO code physico-mathematical model lumped parameters analytical test verification hydrogen distribution Nuclear Energy and Technology 4(4): 279-285 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31892 https://doi.org/10.3897/nucet.4.31892 https://nucet.pensoft.net/article/31892/ https://nucet.pensoft.net/article/31892/download/pdf/ en

10.3897/nucet.4.30379 2018-12-13 nucet

Simulate the effect of integral burnable absorber on the neutronic characteristics of a PWR assembly Galahom,Ahmed PWR IBAs MCNPX code Gd2O3 Power Thermal neutron flux Nuclear Energy and Technology 4(4): 287-293 This article examines the effect of an integral burnable absorber (IBA) on the neutronic characteristics of Pressurized Water Reactor (PWR) to provide possible improvements for the fuel management. MCNPX code was used to design a three dimensional model for PWR assembly. The designed model has been validated by comparing the output data with a previously published data. MCNPX code was used to analyze the radial thermal neutron flux and the radial power distribution through PWR assembly with and without IBA. Gadolinium is burnable absorber material that was used in the IBA rods. The gadolinium element suppressed the power in the regions where they were distributed. The existence of IBA rods has a large effect on the Kinf. This effect decreases gradually with burnup due to the degradation of gadolinium. The gadolinium isotopes degradation was analyzed with burnup. Different numbers of IBA rods were investigated to optimize the suitable number that can be used in the PWR assembly. The gadolinium effect on the concentration of 135Xe and 149Sm resulting from the fission process was analyzed. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.30379 https://doi.org/10.3897/nucet.4.30379 https://nucet.pensoft.net/article/30379/ https://nucet.pensoft.net/article/30379/download/pdf/ en

10.3897/nucet.4.31868 2018-12-13 nucet

Verification on application program generation and loading for safety systems of nuclear power plants based on the reverse engineering method Belonosov,Mikhail Kishkin,Vladimir Korolev,Sergey A. Verification reverse engineering code generation safety systems controller mathematical model instrumentation tools Nuclear Energy and Technology 4(4): 223-228 The article describes an automated verification method used for application software of control safety systems based on the TPTS-SB equipment. Verification is performed by comparing two mathematical models (oriented graphs): one obtained by processing the original design data, i.e., graphical functional diagrams, and the other formed by reversing the program code loaded from the controller. The vertices in both graphs are functional blocks of mathematical and logical operations; the edges are connections between them. The constructed mathematical models undergo a comparison, covering the vertices and edges of the graphs as well as the memory cells and values of constants. The equivalence of mathematical models proves the correspondence between the program code and the initial set of design functional diagrams. The proposed automated verification method makes it possible to prove that no distortion is introduced into the program during the process of converting graphical functional diagrams into the program code with its subsequent translation and loading into the controller. It is postulated that any distortions will be detected during the verification procedure, which is performed every time after loading the code into the controller. The solution provides an acceptable speed when large volumes of vector graphics stored in a relational database are processed, and makes it possible to visualize the verification results. The proposed method is implemented in the GET-R1 instrumentation tools for TPTS-SB and is used in designing and verifying the application software of the safety systems at the Belarusian NPP. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31868 https://doi.org/10.3897/nucet.4.31868 https://nucet.pensoft.net/article/31868/ https://nucet.pensoft.net/article/31868/download/pdf/ en

10.3897/nucet.4.31869 2018-12-13 nucet

Comparison of spallation reaction models based on multiple-criteria decision analysis Andrianov,Andrey Korovin,Yury Kuptsov,Ilya Konobeyev,Aleksandr Andrianova,Olga Нigh-energy nuclear reactions nuclear data multiple-criteria decision analysis methods uncertainty Nuclear Energy and Technology 4(4): 229-234 The paper presents the results of a comparative evaluation of the predictive ability of seventeen spallation reaction models (CEM02, CEM03, Phits/jam, Cascade/ASF, Phits/Bertini, Bertini/Dresner, Cascade-4, INCL4/Abla, INCL4/smm, geant4/binary, Isabela/smm, geant4/Bertini, Isabela/Abla, INCL4/Gemini, CASCADeX-1.2, Isabel/Gemini, Phits/jqmd) for the interaction reactions of high-energy protons with natPb nuclei using the most popular methods of multiple-criteria decision analysis (MAVT/MAUT, AHP, TOPSIS, PROMETHEE). Multiple-criteria decision analysis methods are used extensively to support decision-making in various fields of knowledge, including nuclear physics and engineering, when aggregating conflicting criteria with due account for the expert and decision-maker opinions. Four factors of computational and experimental agreement (R, D, F, H), most commonly used in this field of knowledge, have been employed as the criteria, which, having been aggregated as part of applying respective multiple-criteria decision analysis methods, make it possible to estimate the integral measure of the computational model effectiveness and to rank the models, using this as the basis, depending on the degree of their predictive ability. It has been demonstrated that the ranking results obtained using different multiple-criteria decision analysis methods show a good agreement. Using a stochastic approach to the generation of weights, the models were ranked in conditions with the absence of data on the significance of individual agreement factors. Recommendations are presented for using the multiple-criteria decision analysis methods to address tasks involved in the preparation of nuclear data in conditions of a multiple-factor evaluation of discrepancies between calculations and experiment. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31869 https://doi.org/10.3897/nucet.4.31869 https://nucet.pensoft.net/article/31869/ https://nucet.pensoft.net/article/31869/download/pdf/ en

10.3897/nucet.4.31870 2018-12-13 nucet

Safe development of nuclear power technologies in the Arctic: prospects and approaches Sarkisov,Ashot Antipov,Sergey Smolentsev,Dmitry Bilashenko,Vyacheslav P. Kobrinsky,Mikhail N. Sotnikov,Vladimir A. Shvedov,Pavel A. Arctic region nuclear icebreaking fleet Low-Power Nuclear Power Plant development forecast radiation safety sea areas mathematical modeling Nuclear Energy and Technology 4(4): 235-241 The demands for nuclear power technologies in the Arctic for solving social and economic problems of the state can only be satisfied if adequate strategies of their safe handling at all stages from design to decommissioning are defined, methodological approaches and mathematical models for predicting and minimizing adverse environmental impacts of potential emergency situations at such facilities are developed, and scientifically-based results yielded within a decision-making support system for the elimination of such emergencies are applied. Special relevance of these requirements is determined by unique features of the Arctic nature and its role in the generation of climatic and hydrological processes in the World Ocean. Main results and generalized conclusions based on the analysis of radiological consequences of the large-scale application of nuclear power industry for the benefit of economic development of the Arctic region are provided in the present paper. The analysis was performed within the framework of the complex research project “Development of the methodological approaches and mathematical models to access the environmental impact of the possible accidents at the floating nuclear power objects, model calculations of the radiation propagation in the Arctic aquatic territories in case of emergency situations”. The increasing demand for the low-power nuclear power plants for the benefit of development of remote areas, the technological and economic advantages of such power plants as well as minimal possible environmental consequences of the hypothetic accidents resulted in the qualitative changes in the attitude towards their usage. Estimation was made of the scale of application of nuclear power and results were obtained of numerical modeling of distribution of reactivity in case of accidents. The conclusion was drawn on the necessity to adhere to the low-power nuclear energy generation development strategy based on the modular design concept. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2018 Research Article text/html info:doi:10.3897/nucet.4.31870 https://doi.org/10.3897/nucet.4.31870 https://nucet.pensoft.net/article/31870/ https://nucet.pensoft.net/article/31870/download/pdf/ en

10.3897/nucet.5.33976 2019-03-20 nucet

Determination of 18-month fuel cycle parameters for the purpose of fuel costs minimization at the basis of use constructions of fuel assemblies in VVER-1200 reactors Hashlamoun,Taha Mohd Vygovsky,Sergey Leskin,Sergey Duman,A. Safa PROSTOR fluence VVER-1000 fuel-reloading scheme 18-month fuel cycle effective days natural days fuel enrichment burn-up operating costs reactor vessel Nuclear Energy and Technology 5(1): 9-15 This article presents the results of research, that were focused on determining the optimal parameters of the extension of (reactor life-time) reactor fuel cycle in order to reduce the total operating costs of nuclear power plants during the transition from 12-month reactor fuel cycle to 18-month fuel cycle. The relevance of the research is related to the fact that, in recent years, there is a transition at all operating nuclear power plants VVER-1000 (1200) from 12-month reactor fuel cycle to extended 18-month fuel cycle. At the same time, represent the interests to solve the problem of conservation the extension of reactor life-time while reducing the number of loaded fuel assemblies with fresh fuel assemblies, which would reduce the total operating, and fuel costs. Search for solutions of this problem is associated with mandatory implementation of all requirements for the safe operation of the reactor and the reduction of the maximum fast neutron fluence on the reactor vessel in comparison with its value at the operating nuclear power plants. In the present work, with using the program PROSTOR software complex researched the neutron-physical characteristics of the core at the nominal parameters of the VVER-1200 reactor through the implementation of various fuel cycle strategies. The article developed various schemes of fuel-reloading for an 18-month fuel cycle with a different number of fuel assemblies. The article carries out a comparative analysis of the main parameters in the core for fuel-reloading schemes options of an 18- and 12-month fuel cycle with each other. Determine the minimum amount of fuel assemblies and provide the necessary duration of the reactor life-time for 18-month fuel cycle with using the extension of reactor life-time by reducing the power at the end of the reactor cycle to 70% of the nominal power. In the article, the arrangements of fuel assemblies were developed to provide limitations of local power by volume of the core, which reduce the fluence of fast neutrons on the reactor vessel in comparison with the projected value of the fluence. This article shows that the 18-month fuel cycle for the VVER-1200 reactor is more economical than the 12-month fuel cycle. These studies were carried out for the VVER-1200 reactor at the power of 100% of the nominal. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.33976 https://doi.org/10.3897/nucet.5.33976 https://nucet.pensoft.net/article/33976/ https://nucet.pensoft.net/article/33976/download/pdf/ en

10.3897/nucet.5.33977 2019-03-20 nucet

Conceptual potential of a pyroelectrochemical technology for the thorium engagement in the fast neutron fuel cycle Naumov,Valery Thorium oxide fuel thorium fuel cycle options pyroelectrochemical processes electrochemical behavior of thorium fuel reprocessing Nuclear Energy and Technology 5(1): 17-22 The use of thorium in combination with plutonium in nuclear power generation offers a solution to the problem of reducing the accumulation of long-lived transplutonium nuclides. Along with this, the existing uranium fuel cycle (UFC) has such disadvantage as the vulnerability to unauthorized use of nuclear materials. The thorium fuel cycle (TFC) is devoid of these drawbacks. The engagement of thorium in nuclear power is possible provided the availability of an appropriate technology for reprocessing irradiated thorium. A fuel cycle based on thorium oxide may not differ in principle from the already developed pyrochemical fuel cycle involving uranium and plutonium oxides. Thorium oxide is most commonly obtained in compact state by electrolysis of molten salts from thorium-containing electrolytes. The most thorough studies of physical and chemical and electrochemical behavior of thorium in molten haloids of alkali and alkaline-earth metals were conducted in the 1960ies and the 1970ies. Since extensive experimental material has been accumulated by now for justification of the use of pyroelectrochemical and chemical processes for regeneration of fuel in molten salts, then it has also been proposed that technologies for fuel reprocessing in molten chlorides of alkali metals should be applied resulting in a crystalline product that can be used for the fuel element fabrication. Unlike uranium and plutonium, thorium behavior in molten salt environments is less complex. In molten salts, thorium exists predominantly in the form of Th4+, and the mixture of uranium and thorium dioxides with ThO2 content reaching up to 50 % can be obtained by electrolysis of molten salts. Therefore, the existing amount of knowledge about the chemistry of thorium allows regarding the use of pyrochemical processes in production of thorium oxide as highly promising, and the available data on the physical and chemical properties of thorium and its compounds in high-temperature molten salts makes it possible to state that the pyroelectrochemical technology can be potentially used in production and reprocessing of thorium fuel. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.33977 https://doi.org/10.3897/nucet.5.33977 https://nucet.pensoft.net/article/33977/ https://nucet.pensoft.net/article/33977/download/pdf/ en

10.3897/nucet.5.33978 2019-03-20 nucet

Peculiarities of the radiation formation in dispersed microencapsulated nuclear fuel Bedenko,Sergey Knyshev,Vladimir Kuznetsova,Mariya Lutsik,Igor Shamanin,Igor Thorium reactor isotopic composition alpha particle transport neutron source spectrum of radiation sources Nuclear Energy and Technology 5(1): 23-29 A computational study has been performed for various options of the thorium reactor core loading. Neutronic studies of fuel have been conducted, its isotopic composition has been calculated, and the alpha emitters and the sources of neutron and photon radiation in the microencapsulated nuclear fuel have been analyzed. The studies had the purpose of developing the methodology used to estimate the radiation characteristics of nuclear fuel with a complex inner structure. Emphasis is placed on calculating the quantitative and spectral composition of the neutrons formed as the result of (a, n) reactions on small- and average-mass nuclei. The ratio of the quantity of the neutrons resulting from the (a, n) reactions to the quantity of the neutrons formed as the result of spontaneous fission has been calculated for fuel with heterogeneous and homogeneous arrangements of fissionable and structural elements. The developed tools will make it possible to estimate the neutron radiation dose, to revise the traditional fresh and spent fuel handling procedures, and to estimate, using the Rossi alpha method, the neutron multiplication factor in deeply subcritical systems. The neutron yield and spectrum were calculated using an analytical model and verified codes such as WIMS-D5B, ORIGEN-APP, SOURCES-4C and SRIM-2013. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.33978 https://doi.org/10.3897/nucet.5.33978 https://nucet.pensoft.net/article/33978/ https://nucet.pensoft.net/article/33978/download/pdf/ en

10.3897/nucet.5.33980 2019-03-20 nucet

Vertical steam generators for VVER NPPs Egorov,Mikhail Vertical type steam generator horizontal type steam generator light water nuclear reactor natural circulation variable cyclic thermal stresses pre-boiling section heat exchange steam-water mixture temperature difference separators of chevron and swirl-vane type containment the four-loop layout. Nuclear Energy and Technology 5(1): 31-38 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.33980 https://doi.org/10.3897/nucet.5.33980 https://nucet.pensoft.net/article/33980/ https://nucet.pensoft.net/article/33980/download/pdf/ en

10.3897/nucet.5.33981 2019-03-20 nucet

Optimization models of a two-component nuclear energy system with thermal and fast reactors in a closed nuclear fuel cycle Andrianov,Andrey Kuptsov,Ilya Osipova,Tatyana Andrianova,Olga Utyanskaya,Tatyana Thermal reactors fast reactors closed nuclear fuel cycle UOX MOX MNUP optimization uncertainty forecasting MESSAGE Nuclear Energy and Technology 5(1): 39-45 The article presents a description and some illustrative results of the application of two optimization models for a two-component nuclear energy system consisting of thermal and fast reactors in a closed nuclear fuel cycle. These models correspond to two possible options of developing Russian nuclear energy system, which are discussed in the expert community: (1) thermal and fast reactors utilizing uranium and mixed oxide fuel, (2) thermal reactors utilizing uranium oxide fuel and fast reactors utilizing mixed nitride uranium-plutonium fuel. The optimization models elaborated using the IAEA MESSAGE energy planning tool make it possible not only to optimize the nuclear energy system structure according to the economic criterion, taking into account resource and infrastructural constraints, but also to be used as a basis for developing multi-objective, stochastic and robust optimization models of a two-component nuclear energy system. These models were elaborated in full compliance with the recommendations of the IAEA’s PESS and INPRO sections, regarding the specification of nuclear energy systems in MESSAGE. The study is based on publications of experts from NRC “Kurchatov Institute”, JSC “SSC RF-IPPE”, ITCP “Proryv”, JSC “NIKIET”. The presented results demonstrate the characteristic structural features of a two-component nuclear energy system for conservative assumptions in order to illustrate the capabilities of the developed optimization models. Consideration is also given to the economic feasibility of a technologically diversified nuclear energy structure providing the possibility of forming on its base a robust system in the future. It has been demonstrated that given the current uncertainties in the costs of nuclear fuel cycle services and reactor technologies, it is impossible at the moment to make a reasonable conclusion regarding the greatest attractiveness of a particular option in terms of the economic performance. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.33981 https://doi.org/10.3897/nucet.5.33981 https://nucet.pensoft.net/article/33981/ https://nucet.pensoft.net/article/33981/download/pdf/ en

10.3897/nucet.5.33982 2019-03-20 nucet

Investigation of the small break conditions in the primary circuit of a VVER-1000 reactor Belozerov,Vladimir Zhuk,Mikhail Terekhova,Anna Mode pipe break failure of heat removal reactor plant pressure stop valve departure from nucleate boiling safety factor high-speed dump valves with discharge to atmosphere (HSDV-A) Nuclear Energy and Technology 5(1): 47-52 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.33982 https://doi.org/10.3897/nucet.5.33982 https://nucet.pensoft.net/article/33982/ https://nucet.pensoft.net/article/33982/download/pdf/ en

10.3897/nucet.5.33984 2019-03-20 nucet

Infological models of the ENDF-format nuclear data Yuferov,Anatoliy ENDF libraries infologic modeling relational format Nuclear Energy and Technology 5(1): 53-59 Issues involved in the infologic modeling of the ENDF-format nuclear data libraries for the purpose of converting ENDF files into a relational database have been considered. The transfer to a relational format will make it possible to use standard readily available tools for nuclear data processing which simplify the conversion and operation of this data array. Infological models have been described using formulas of the “Entity (List of Attributes)” type. The proposed infological formulas are based on the physical nature of data and theoretical relations. This eliminates the need for a special notation to be introduced to describe the structure and the content of data, which, in turn, facilitates the use of relational formats in codes and solution of nuclear data evaluation problems. The concept of nuclear informatics has been formulated based on relational DBMS technologies as one of the tools for solving the “big data” problem in modern science and technology. The organizational and technological grounds for the transfer of ENDF libraries to a relational format are presented. Requirements to the nuclear data presentation formats supported by relational DBMS are listed. Peculiarities of the infological model construction, conditioned by the hierarchical nature of nuclear data, are identified. The sequence for the ENDF metadata saving is presented, which can be useful for the verification and validation (testing of the structural and syntactical validity and operability) of both source data and the procedures for the conversion to a relational format. Formulas of infological models are presented for the cross sections file, the secondary neutron energy distributions file, and the nuclear reaction product energy-angle distributions file. A complete array of infological models for ENDF libraries and the generation modules of respective relational tables are available on a public website. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.33984 https://doi.org/10.3897/nucet.5.33984 https://nucet.pensoft.net/article/33984/ https://nucet.pensoft.net/article/33984/download/pdf/ en

10.3897/nucet.5.33985 2019-03-20 nucet

Study of hydrogen generation and radionuclide release during wet damaged oxide spent fuel storage Gayazov,Artem Komarov,Sergey Leshchenko,Anton Revenko,Ksenia Smirnov,Valery Zvir,Elena Ilyin,Pavel Teplov,Vadim Fire and explosion safety damaged SNF VVER-440 radiolysis hydrogen gaseous fission products UO2 dissolution Nuclear Energy and Technology 5(1): 61-66 The paper describes the outcomes of the experiments to study hydrogen and gaseous fission products accumulation during simulations of the wet damaged VVER-440 SNF storage in air-tight canisters with the water drained and no drying conducted. Physical and chemical processes occurring during the damaged oxide SNF storage in wet environment are discussed. The experiments were carried out in two stages: 1) preliminary soaking of fine fuel particles in water in an air-tight canister, 2) water draining and keeping the wet SNF in the air-tight canister. The experiments were conducted one after another using the same SNF canister and differing only in the SNF soaking temperature, i.e. 25 and 80 °С. The radionuclide release into the liquid phase during the SNF storage under water was studied. Uranium and cesium isotopic concentrations were found to reach steady values when the SNF is kept under water for more than a month. The kinetics of hydrogen and gaseous fission product accumulation in the gaseous phase during wet storage of the spent fuel in the air-tight canister with the water drained coincide for both experiments. The kinetics demonstrate an abrupt decrease of the hydrogen and gaseous fission product accumulation rate in 46 hours. The data obtained can be applied for development and verification of the damaged SNF behavior models during SNF storage in wet environment under radiolysis. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.33985 https://doi.org/10.3897/nucet.5.33985 https://nucet.pensoft.net/article/33985/ https://nucet.pensoft.net/article/33985/download/pdf/ en

10.3897/nucet.5.32239 2019-03-20 nucet

Optimization of two opposing neutron beams parameters in dynamical (B/Gd) neutron cancer therapy Haidar,Nassar Accelerator Based Modulated Neutron Sources One-Speed Neutron Diffusion Two Opposing Neutron Beams Dynamical NCT Pareto Optimization. Nuclear Energy and Technology 5(1): 1-7 We demonstrate how the therapeutic utility index and the ballistic index for dynamical neutron cancer therapy (NCT) with two opposing neutron beams form a nonlinear optimization problem. In this problem, the modulation frequencies ω and ϖ of the beams and the relative time advance ε are the control variables. A Pareto optimal control vector ω* = (ω*, ϖ*, ε*) for this problem is identified and reported for the first time. The utility index is shown to be remarkably periodically discontinuous in ε, even in the neighborhood of ε*. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.32239 https://doi.org/10.3897/nucet.5.32239 https://nucet.pensoft.net/article/32239/ https://nucet.pensoft.net/article/32239/download/pdf/ en

10.3897/nucet.5.34296 2019-04-11 nucet

Reliability of the RBMK-1000 coolant flow measurement system Pereguda,Arkady Parametric reliability coolant flow measurement system random variables time between failures random process mathematical time expectation distribution function exponential estimate Nuclear Energy and Technology 5(1): 81-87 An analysis of statistical data of diagnostic measurements of two parameters determining the performance of the RBMK-1000 SHADR-8A flowmeters – the minimum value of the negative amplitude half-wave at the transistor flow measuring unit (TIBR) input and the mean-square deviation over the flowmeter ball rotation period – made it possible to develop a mathematical model of the flowmeter parametric reliability. This mathematical model is a random process, which is a superposition of two delayed renewal processes. Studying the flowmeter operational reliability model provides an exponential estimate of the probability that the parameters determining the flowmeter performance will not exceed the specified levels. Using the Bernoulli scheme and the probability-estimating relationship for the flowmeter performance parameters, it is possible to calculate the probability of failure-free operation of both a single reactor quadrant and the coolant flow measurement system. In addition, it becomes possible to estimate the quadrant failure rate. Important for practice is the possibility of predicting the number of failed flowmeters depending on the system operation time. An indicator of the system reliability can be the average number of failed flowmeters, the relation for which is given in the paper. All the research results were obtained without any additional assumptions about the random values distribution laws. The obtained results can be easily generalized for the cases when the vector dimension of the determining parameters is greater than two. The use of the results of this study is illustrated by calculated quantitative values of the flowmeter parametric reliability indicators and the coolant flow measurement system. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.34296 https://doi.org/10.3897/nucet.5.34296 https://nucet.pensoft.net/article/34296/ https://nucet.pensoft.net/article/34296/download/pdf/ en

10.3897/nucet.5.34294 2019-04-11 nucet

Gas-cooled nuclear reactor core shaping using heat exchange intensifiers Kuzevanov,Vyacheslav Podgorny,Sergey Core shaping heat exchange intensification mass flow distribution maximum channel wall temperature Nuclear Energy and Technology 5(1): 75-80 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). info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.34294 https://doi.org/10.3897/nucet.5.34294 https://nucet.pensoft.net/article/34294/ https://nucet.pensoft.net/article/34294/download/pdf/ en

10.3897/nucet.5.34293 2019-04-11 nucet

Generation IV supercritical water-cooled nuclear reactors: Realistic prospects and research program Kirillov,Pavel Bogoslovskaya,Galina GIF-IV program supercritical water reactor (SCWR) prospects for nuclear power development advantages of the VVER-SKP concept problems of developing the VVER-SKP proposals for cooperation number of publications. Nuclear Energy and Technology 5(1): 67-74 Existing conditions make possible obtaining information that being discussed openly by wide scientific community could help outlining or even establishing the expediency of a particular area of present and future research. Use link http://www.sciencedirect.com to learn about the topics or areas that most attract researchers from different countries. The Generation IV International Forum (GIF-IV) established in January 2000 has set a goal to improve the new generation of nuclear technologies in the following areas: stability, safety and reliability, economic competitiveness, proliferation resistance and physical protection. The purpose of the present publication is to prepare a discussion of one of the directions of development of fourth-generation NPPs, the groundwork for which has already been laid in thermal power engineering in various countries. The number of papers published annually on this topic is the largest among other similar topics dedicated to nuclear power plants of the fourth generation. Judging from the operating experience of existing nuclear power plants using water as a coolant, it can be ascertained that the tendency of building water-cooled nuclear power plants will remain during the next 30 to 50 years. During the present stage the task in the development of alternative types of reactors will be limited to demonstration of their performance and acceptability for future power engineering and the society. The project of supercritical water-cooled reactor is based on the operating experience of VVER, PWR, BWR reactors (more than 14,000 reactor-years); many years of experience accumulated in operating fossil thermal power plants (more than 400 power units; 20,000 years of operation of power units) using supercritical (25 MPa, 540°C) and super-supercritical (35–37 MPa, 620–700°C) water steam. In Russia more than 140 supercritical pressure units are currently in operation. Numerical calculation and design of supercritical water-cooled reactor (similarly to BR-10 reactor) will allow not only training personnel for future development of this technology, but will also help revealing the most difficult points requiring experimental confirmation with application of independent test facilities, as well as formulating the plan of first priority experimental studies. Knowledge accumulated over the last 10 years in the world allows the following: further specifying the already developed concept; developing a plan of specific priority studies; compiling task order for designing small-power pilot VVER SKP-30 reactor (30 MW-th). The scope of problems that are to be solved to substantiate VVER-SCP reactor and commence designing an experimental reactor with thermal capacity of 30 MW is the same as that in developing any type of nuclear reactor: physics of the reactor core; material related matters (primarily concerned with the reactor pressure vessel, fuel, and fuel rod cladding); thermal hydraulics of rod bundles in the near- and supercritical areas; water chemistry at supercritical pressure; corrosion of materials, development of safety systems. Research must be carried out both in static conditions and under irradiation. The absence in Russia during the extended time period of approved program with allocation of appropriate funding and preservation of the existing status during the coming two or three years will lead to the situation when Russia will be hopelessly lagging behind in the development of SCWR technology. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.34293 https://doi.org/10.3897/nucet.5.34293 https://nucet.pensoft.net/article/34293/ https://nucet.pensoft.net/article/34293/download/pdf/ en

10.3897/nucet.5.35373 2019-04-12 nucet

Corrigendum: Comparison of spallation reaction models based on multiple-criteria decision analysis. https://doi.org/10.3897/nucet.4.31869 Andrianov,Andrey Korovin,Yury Kuptsov,Ilya Konobeyev,Aleksandr Andrianova,Olga Corrigenda: Comparison of spallation reaction models based on multiple-criteria decision analysis. https://doi.org/10.3897/nucet.4.31869 Nuclear Energy and Technology 5(1): 89-89 Corrigenda: Comparison of spallation reaction models based on multiple-criteria decision analysis. https://doi.org/10.3897/nucet.4.31869 info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Corrigendum text/html info:doi:10.3897/nucet.5.35373 https://doi.org/10.3897/nucet.5.35373 https://nucet.pensoft.net/article/35373/ https://nucet.pensoft.net/article/35373/download/pdf/ en

10.3897/nucet.5.35579 2019-05-17 nucet

A study into the dependence of the cladding-fuel pellet gap conductance on burn-up and the effects on the reactor core neutronic performance Vygovsky,Sergey Gruzdov,Fedor Al Malkawi,Rashdan VVER-1000 gap conductance coefficient burn-up xenon oscillations reactivity Doppler effect Nuclear Energy and Technology 5(2): 97-102 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.35579 https://doi.org/10.3897/nucet.5.35579 https://nucet.pensoft.net/article/35579/ https://nucet.pensoft.net/article/35579/download/pdf/ en

10.3897/nucet.5.35577 2019-05-17 nucet

Verification of the ROSFOND/ABBN nuclear data based on the OECD/NEA benchmark on criticality safety of mox-fueled systems Andrianova,Olga Golovko,Yury Manturov,Gennady MOX fuel integral experiments BFS critical facility accuracy estimation effective multiplication factor neutron data uncertainty OECD/NEA benchmark maximum likelihood method Nuclear Energy and Technology 5(2): 91-96 The paper presents the results of a computational analysis of the OECD/NEA benchmark conducted to estimate the accuracy of the critical safety parameters of multiplying MOX-fueled systems. The computational test is a set of 15 spherical multiplying systems that differ in their compositions and geometries. According to the test conditions, the keff values of the analyzed systems are unknown in advance. As part of the computational analysis of the test involving national codes and nuclear data libraries, along with the keff calculations, it is also necessary to estimate the a priori (due to the accuracy of the nuclear data used) and a posteriori (based on the accumulated experimental information) errors in the calculated keff values. Based on the benchmark, an updated version of the ROSFOND/ABBN-RF nuclear data was tested. The results of estimating the a priori and a posteriori errors in keff using the INDECS system for the proposed test models are presented. The analysis of the calculated data shows that (1) the observed spread in the keff values obtained from the Russian ROSFOND library and foreign evaluated nuclear data libraries (ENDF/B-VII.0, JEFF-3.2, JENDL-4.0) varies from –0.3 up to 0.8%; and (2) the deviation of the calculation results in the keff values obtained from the ROSFOND library and its group version, ABBN-RF, does not exceed 0.1%. The average a priori error in keff for all the tested options of multiplying systems is about 1% and, taking into account the selected set of experimental criticality data for MOX-fueled systems, including experiments at the BFS facilities, the average a posteriori error in keff can be reduced to 0.3%. The performed estimations confirm the high accuracy of the ROSFOND/ABBN-RF nuclear data for calculating the critical safety parameters of multiplying MOX-fueled systems. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.35577 https://doi.org/10.3897/nucet.5.35577 https://nucet.pensoft.net/article/35577/ https://nucet.pensoft.net/article/35577/download/pdf/ en

10.3897/nucet.5.35797 2019-05-17 nucet

A study into the scale effect on the strength properties of polymer composite materials Kiryushina,Valentina Kovaleva,Yuliya Stepanov,Pуtr Kovalenko,Pavel Polymer composite materials glass reinforced plastics scale effect strength Weibull statistical model size effect Nuclear Energy and Technology 5(2): 103-108 Polymer composite materials (PCM) are used extensively and are viewed as candidates for application in various industries, including nuclear power. Despite a variety of methods and procedures employed to investigate the mechanical characteristics of PCMs, the use of the laboratory sample mechanical test results to design and model large-sized structures is not always fully correct and reasonable. In particular, one of the problems is concerned with taking into account the scale parameter effects on the PCM strength and elastic characteristics immediately in the product. The purpose of the study is to investigate the scale effects on the mechanical characteristics of glass reinforced plastics using phenolformaldehyde and silicon-organic binders and a fabric quartz filler. Samples of four different standard sizes under GOST 25604-82 and GOST 4648-2014 were tested for three-point bending using an LFM-100 test machine to estimate the scale effect. The thicknesses of the model samples were chosen with regard for the wall thicknesses of full-scale products under development or manufactured commercially and the test machine features, and varied in the limits of 1.6 to 7.5 mm. The tests showed that strength decreased as the sample thickness was increased to 3 mm and more both at room and elevated (200 to 500 °C) temperatures, which can be described by an exponential function based on the Weibull statistical model. The values of the Weibull modulus that characterizes the extent of the scale effect on the strength of the tested materials were 4.6 to 6.7. The average bend strength in the sample thickness range of 3 mm and less does not vary notably or tends to increase slightly as the thickness is increased. This fact makes it possible to conclude that estimation of allowable stresses in a thin-wall product requires the use of test results for samples with a thickness that is equal to the product wall thickness since standard samples may yield overestimated allowable stress values and lead, accordingly, to incorrect calculations of the strength factor. The results obtained shall be taken into account when defining the allowable levels of operation for full-scale products and structures of polymer composites based on the laboratory sample strength data as well as when estimating their robustness as a characteristic of the product’s fail-safe operation. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.35797 https://doi.org/10.3897/nucet.5.35797 https://nucet.pensoft.net/article/35797/ https://nucet.pensoft.net/article/35797/download/pdf/ en

10.3897/nucet.5.35798 2019-05-17 nucet

Impact of secondary particles on microdistribution of deposited dose in biological tissue in the presence of gold and gadolinium nanoparticles under photon beam irradiation Konobeev,Ivan Kurachenko,Yury Sheino,Igor Nanoparticles of gold and gadolinium deposited dose photon beam therapy Geant4 Monte Carlo simulation Nuclear Energy and Technology 5(2): 109-116 It is experimentally proven that nanoparticles of high-Z materials can be used as radiosensitizers for photon beam therapy. In the authors’ opinion, data available as of today on the impact of secondary particles (electrons, photons and positrons generated in biological tissue by penetrating beam of primary photons) on the distribution of deposited dose during photon beam therapy in the presence of nanoparticles, are insufficient. Investigation of this impact constituted the main goal of this work. Two-stage simulation was performed using Geant4 platform. During the first stage a layer of biological tissue (water) was irradiated by monoenergetic photon sources with energies ranging from 10 keV to 6 MeV. As the result of this modeling spectra of electrons, photons and positrons were obtained at the depth of 5 cm. During the second stage the obtained photon spectra were used to irradiate gold, gadolinium and water nanoparticles. Radial distributions of energy deposited around nanoparticles were obtained as the result of this modeling. Radial DEF (Dose Enhancement Factor) values around nanoparticles of gold and gadolinium positioned in water at the depth of 5 cm were obtained after processing the collected data. Contributions from primary photons and secondary particles (electrons, photons and positrons generated in the layer of water with 5-cm thickness by the penetrating beam of primary photons) in the additional dose deposited around the nanoparticles were calculated as well. It was demonstrated that layer of biological tissue placed between the source of photons and nanoparticles considerably changes the initial spectrum of photons and this change is significant in the analysis of mechanism of radiosensitization of biological tissues by nanoparticles for all energies of photon sources (up to 6 MeV). It was established that interaction of electrons and positrons with nanoparticles does not lead to significant increase of additional dose in the vicinity of their surfaces and can be most likely excluded from consideration in the analysis of radiosensitization mechanism of nanoparticles. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.35798 https://doi.org/10.3897/nucet.5.35798 https://nucet.pensoft.net/article/35798/ https://nucet.pensoft.net/article/35798/download/pdf/ en

10.3897/nucet.5.35800 2019-05-17 nucet

Dismantling and decontamination of large-sized radiation-contaminated equipment during Research Building B decommissioning at the Bochvar Institute site Kuznetsov,Andrey Azovskov,Mikhail Belousov,Sergey Vereshchagin,Ilya Efremov,Alexey Khlebnikov,Sergey Decommissioning radiation-contaminated equipment tank extraction column dismantling uranium plutonium foam decontamination radioactive waste Nuclear Energy and Technology 5(2): 117-122 The article presents the results of work on dismantling the large installation equipment of Research Building B at the Bochvar High-technology Research Institute of Inorganic Materials (Bochvar Institute). The works were carried out as part of Building B preparation for decommissioning. The purpose of dismantling the large-sized capacitive equipment was to reconstruct the large installation site for managing radioactive waste generated during Building B decommissioning. The works on decommissioning a radioactively contaminated building within a densely populated district of megalopolis were carried out for the first time. The characteristics of the large-sized capacitive equipment are presented. Radioactive contamination of the capacitive equipment is determined by long-lived a-emitting isotopes: 235U, 238U, 239Pu. The sequence of works on dismantling the radiation-contaminated capacitive equipment includes preparatory work, dismantling the tank piping, localizing radioactive contamination of the external surface of the equipment as well as dismantling and moving it into a transport container. Dismantling and decontamination of the large-sized capacitive equipment was carried out by the Bochvar Institute Decommissioning Department. The following tools were used during the works: (1) a mobile foam decontamination facility to perform decontamination works and (2) a mobile high pressure facility to apply localizing and decontaminating film coatings. The tanks were dismantled by means of low-spark tools, i.e., reciprocating saws. Crane runways were made in order to move the dismantled equipment into transport containers: the movement was carried out with the help of a winch. The main results of dismantling and decontaminating the radioactively contaminated tanks are the dismantling of four units of long-length column-type equipment with heights from 4.2 to 6.4 m and 26 units of capacitive equipment (maximum capacity = 8 m3) as well as decontamination of the internal surfaces of radiation-contaminated equipment (decontamination factor = 25–70). As a result, the activity of the accumulated radioactive waste was reduced (the RW class was changed from 3 to 4). The main conclusion regarding the managment of large-sized radiation-contaminated tanks during Building B decommissioning is as follows: the works were organized and carried out at a high technical level, using modern decontamination and dismantling equipment and modern methods to ensure work safety at the Bochvar Institute site in the city of Moscow. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.35800 https://doi.org/10.3897/nucet.5.35800 https://nucet.pensoft.net/article/35800/ https://nucet.pensoft.net/article/35800/download/pdf/ en

10.3897/nucet.5.35801 2019-05-17 nucet

Nanofiltration fractionation of radioactive solution components as a method for reducing the volume of wastes intended for permanent disposal Chugunov,Aleksandr Vinnitskii,Vadim Radioactive waste nanofiltration complexation osmotic pressure reverse osmosis membrane permeability Nuclear Energy and Technology 5(2): 123-128 Baromembrane purification methods as part of liquid radioactive media processing complexes are increasingly included in the practice of radioactive waste management. The paper presents the results of a comparative study of the performance of commercially available hyper- and nanofiltration elements when a simulated solution is continuously phosphatized. The study revealed the influence of changes in the feed solution salinity on the permeability, working pressure in the brine chamber of the hyper- and nanofiltration apparatus and the permeate salinity. It is shown that, in a closed loop of liquid radioactive waste, the introduction of polyphosphates to stabilize the truly dissolved forms of multivalent metals on the ULP reverse-osmotic membrane leads, as expected, to a systematic performance degradation, first of all, in the membrane permeability at a fixed pressure in the apparatus. The permeate of the system with a nanofiltration membrane, VNF (Vontron NanoFiltration), contains a sufficiently high salt concentration indicating that sodium salts formed during complexation are removed from the circuit, thereby reducing the solution osmotic pressure which critically affects the yield of the purified solution. Thus, nanofiltration in combination with chelating agents can be an effective tool for fractionating components of radioactive solutions, ensuring the achievement of standard indicators for wastewater and biologically hazardous substances that are subject to permanent disposal. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.35801 https://doi.org/10.3897/nucet.5.35801 https://nucet.pensoft.net/article/35801/ https://nucet.pensoft.net/article/35801/download/pdf/ en

10.3897/nucet.5.36474 2019-06-21 nucet

Application of spiking neural networks for modelling the process of high-temperature hydrogen production in systems with gas-cooled reactors Starkov,Sergey Lavrenkov,Yury Spiking neural networks high-temperature gas-cooled reactors electro-optical neural commuting system hydrogen production forecasting centralized global parallel search circuit Nuclear Energy and Technology 5(2): 129-137 Hydrogen energy is able to solve the problem of the dependence of modern industries on fossil fuels and significantly reduce the amount of harmful emissions. One of the ways to produce hydrogen is high-temperature water-steam electrolysis. Increasing the temperature of the steam involved in electrolysis makes the process more efficient. The key problem is the use of a reliable heat energy source capable of reaching high temperatures. High-temperature gas-cooled reactors with a gaseous coolant and a graphite moderator provide a solution to the problem of heating the electrolyte. Part of the heat energy is used for producing electrical energy required for electrolysis. Modern electrolyzers built as arrays of tubular or planar electrolytic cells with a nuclear energy source make it possible to produce hydrogen by decomposing water molecules, and the working temperature control leads to a decrease in the Nernst potential. The operation of such facilities is complicated by the need to determine the optimal parameters of the electrolysis cell, the steam flow rate, and the operating current density. To reduce the costs associated with the process optimization, it is proposed to use a low-temperature electrolysis system controlled by a spiking neural network. The results confirm the effectiveness of intelligent technologies that implement adaptive control of hybrid modeling processes in order to organize the most feasible hydrogen production in a specific process, the parameters of which can be modified depending on the specific use of the reactor thermal energy. In addition, the results of the study confirm the feasibility of using a combined functional structure made on the basis of spiking neurons to correct the parameters of the developed electrolytic system. The proposed simulation strategy can significantly reduce the consumption of computational resources in comparison with models based only on neural network prediction methods. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.36474 https://doi.org/10.3897/nucet.5.36474 https://nucet.pensoft.net/article/36474/ https://nucet.pensoft.net/article/36474/download/pdf/ en

10.3897/nucet.5.36475 2019-06-21 nucet

Diagnostics of the critical heat flux state of a VVER reactor based on a channel steaming model Kachur,Svetlana Nuclear reactor critical heat flux power density thermophysical model identification neural networks Nuclear Energy and Technology 5(2): 139-144 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.36475 https://doi.org/10.3897/nucet.5.36475 https://nucet.pensoft.net/article/36475/ https://nucet.pensoft.net/article/36475/download/pdf/ en

10.3897/nucet.5.36476 2019-06-21 nucet

Ontologies and databases on thermophysical properties of nuclear reactor materials Chusov,Igor Kirillov,Pavel Pronyaev,Vladimir Obysov,Nikolay Novikov,Grigoriy Thermophysical properties reactor materials nuclear fuel ontology database data integration JSON-format Nuclear Energy and Technology 5(2): 145-153 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.36476 https://doi.org/10.3897/nucet.5.36476 https://nucet.pensoft.net/article/36476/ https://nucet.pensoft.net/article/36476/download/pdf/ en

10.3897/nucet.5.36477 2019-06-21 nucet

Improving the efficiency of fixed radionuclides’ removal by chemical decontamination of surfaces in situ Koryakovskiy,Yuriy Doilnitsyn,Valeriy Akatov,Andrey decontamination decommissioning fixed radioactive contamination film-forming compositions composite decontamination material Nuclear Energy and Technology 5(2): 155-161 The article presents the results of work aimed at improving of chemical decontamination methods. A brief description of existing chemical decontamination technologies used to remove radioactive contamination (RC) from walls, floors and external surfaces of equipment without dismantling, i.e., in situ, is given. A vector of research aimed at improving the efficiency of fixed RC’s removal is also determined. The first aim of this work is to improve the decontaminating properties of removable polymer coatings used in practice. The following domestic products were chosen as study objects: compositions presented under trademarks VA, VL; and also special formulation ZPS-1M. Modifications of these compounds performed in SPSIT in some cases made it possible to significantly increase the decontamination factors (DF). The best results were obtained for VL compositions: it was found that due to certain additives it is possible to increase the DF for metal surfaces by a factor of 5–35 over the base product. Along with film-forming decontaminating compositions, an alternative patented technique has been developed in SPSIT. The main feature of this technique is usage of sorbent-based composite covering material previously saturated by decontaminating solution. New technique allows to achieve far higher decontamination factors (150–500) when fixed RC is removed from metal surfaces. In addition, it can be applied to polymer and other non-metal materials. One of the main advantages of given technique is a drastic (11–16 times) reduction of time required for carrying out decontamination operations. The obtained results may be useful for further research in this area. Thus, research performed allows to come up with general conclusion: there are possibilities to efficiently remove fixed RC from surfaces using rather simple chemical means. That, in its turn, could be a rational alternative to high-priced robotic decontaminating systems. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.36477 https://doi.org/10.3897/nucet.5.36477 https://nucet.pensoft.net/article/36477/ https://nucet.pensoft.net/article/36477/download/pdf/ en

10.3897/nucet.5.36478 2019-06-21 nucet

Use of mathematical modeling to extend the scope of application for the procedure of measuring the mass of 235U in solid radioactive waste Rykov,Nikita Bezhunov,Gennady Gorbachev,Vyacheslav Nondestructive analysis of nuclear materials solid radioactive waste uranium mass gamma-spectrometry ISOCS system absolute efficiency curve Monte Carlo method MCNP code measurement procedure measurement procedure range Nuclear Energy and Technology 5(2): 163-169 The known dependence of absolute efficiency on energy and space for particular measurement conditions is used to determine the mass (activity) of 235U in solid radioactive waste by gamma-spectrometric method. The ISOCS system makes it possible to avoid laborious and time-consuming calibration measurements using standard samples to obtain the absolute efficiency curve due to using the so-called characterized detector having a file with a set of efficiencies for various measurement geometries. In many cases, the establishment of standard samples with parameters covering the 235U mass measurement range in the variation intervals of influencing factors, including density, non-uniformity, isotopic composition, geometry, etc., is very expensive and, most often, not feasible. With regard for this, a computational and experimental approach is used based on results obtained by Monte Carlo method using the MCNP code with variation of the key influencing parameters in a broad range. Calculations were performed for detector-recorded spectra of gamma quanta from casks containing waste differing in the density of the cask content (the density was calculated with regard for the uranium contained in waste) – from 0.016 to 1 g/cm3, in the mass of uranium in waste – from 0.64 g to 2 kg, and in the matrix material – graphite, cellulose, quartz, cellulose with 20 % of iron dust. Applicability boundaries have been defined for the developed procedure to measure uranium-containing waste in terms of the material matrix (~ 2.2 %) and its density (~ 10 %) and the contribution of the uranium mass uncertainty in the cask (5 % for nonporous matrices, 10 % for porous matrices) to the obtained result has been estimated. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.36478 https://doi.org/10.3897/nucet.5.36478 https://nucet.pensoft.net/article/36478/ https://nucet.pensoft.net/article/36478/download/pdf/ en

10.3897/nucet.5.36479 2019-06-21 nucet

Test bench for gas-dynamic studies in the furnace channel for nuclear fuel pellet sintering Kuzmin,Ilya Leshchenko,Anton Pavlov,Sergey Shamsutdinov,Rinat Mochalov,Yuriy gas-dynamic studies sintering furnace furnace channel furnace zone MNUP fuel gas distribution modeling gas concentration measurement gas sampling barrier Nuclear Energy and Technology 5(2): 81-85 Nuclear fuel pellets are sintered in high-temperature furnaces in an atmosphere with strictly defined requirements for the composition of the gas environments in the furnace’s different temperature zones. The preset process conditions in the mixed nitride uranium-plutonium (MNUP) fuel pellet sintering furnace is achieved through the respective gas supply arrangement and by the design of the barriers between the temperature zones and that of the gas supply and discharge units. A CFD model was created in the Ansys Fluent package and validated for testing the functionality of the design concepts used to develop the MNUP fuel sintering furnace channel. A mockup of the sintering furnace channel, which makes a part of the gas-dynamic test bench, was developed and fabricated for the analytical model validation. The paper presents a description of the test bench design and performance for measuring the concentration of gases in the channel simulating the nitride nuclear fuel sintering furnace channel. The results of the test bench gas-dynamic studies were used for the computational and experimental justification of the approaches used to develop the sintering furnace channel. The functionality of the barriers for the sintering furnace channel division into zones with the preset composition of the gas environments and the gas supply and discharge units has been tested experimentally. The obtained experimental data on the distribution of the process gas concentration makes it possible to validate computational thermophysical and gas-dynamic CFD models of the MNUP fuel sintering furnace channel. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.36479 https://doi.org/10.3897/nucet.5.36479 https://nucet.pensoft.net/article/36479/ https://nucet.pensoft.net/article/36479/download/pdf/ en

10.3897/nucet.5.34590 2019-06-21 nucet

Application of soft computing techniques in WWER nuclear power plant safety Abbasi,Akbar Sadikoglu,Fahreddin Nuclear Power Plant (NPP) Soft Computing Neuro Fuzzy Diagnosis System (NFDs) Genetic Algorithms (GAs) Safety Nuclear Energy and Technology 5(2): 177-182 Nowadays, Nuclear Power Plant (NPP) is one of the intended energy resources for the world requirement energy in future, and nuclear power plants provided 11 percent of the world’s electricity production in 2014. Meanwhile, nuclear power plant safety has always been one of the most critical issues in the world. In this paper, the nuclear power plant safety improvement using Soft Computing Techniques were analyzed. For this purpose, the support system based on Neuro-Fuzzy Diagnosis System (NFDs) method and Genetic Algorithms (GAs) approach were used. The obtained result showed that the first symptom is P3 (pressurizer pressure) and second order symptom is P2 (core coolant average temperature) in both approaches. The comparison between the NFDs method and the GAs approaches indicated that the GAs in data test results was faster than the NFDs results. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.34590 https://doi.org/10.3897/nucet.5.34590 https://nucet.pensoft.net/article/34590/ https://nucet.pensoft.net/article/34590/download/pdf/ en

10.3897/nucet.5.38117 2019-09-25 nucet

The technological challenge for current generation nuclear reactors D’Auria,Francesco Debrecin,Nenad Glaeser,Horst Nuclear reactor safety BEPU licensing extended safety margins additional safety barrier Nuclear Energy and Technology 5(3): 183-199 The present paper deals with the proposal of an additional safety barrier for the class of large (1000 MWe or more) Light Water Reactors (LWR) now in operation, in construction, or under design. Emphasis is given to the motivations or the needs for the barrier. Two main parts of the paper can be distinguished. The following topics are discussed in the former part (section 2): (a) the weakness of the barrier constituted by the current design of nuclear fuel; (b) the continuously increasing complexity of the system, with main reference to the Instrumentation and Control (I&C); (c) the role that the Large Break Loss of Coolant Accident (LBLOCA) had for arriving at the current layout of the Reactor Coolant System (RCS). Furthermore avoiding the severe accidents in 1979, 1987 and 2011, is at the basis of the proposal. In the latter part (sections 3 and 4), the elements of the proposed technological safety barrier are discussed: the As-Low-As-Reasonably-Achievable (ALARA) principle, the Best Estimate Plus Uncertainty (BEPU) approach, the Extended Safety Margin Detection (E-SMD) hardware, the Emergency Rescue Team (ERT) strategy (or a virtual entity for the reactor) and the Independent Assessment (IA) concept. The additional safety barrier, although not demonstrated in the paper, is expected to reduce for a factor in the range 10–1000 the probability of core melt and to have a cost in the order of 1% the cost of a nuclear reactor unit. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.38117 https://doi.org/10.3897/nucet.5.38117 https://nucet.pensoft.net/article/38117/ https://nucet.pensoft.net/article/38117/download/pdf/ en

10.3897/nucet.5.39239 2019-09-25 nucet

Personnel training for the developing nuclear power industry Putilov,Alexandr Strikhanov,Mikhail Tikhomirov,Georgy Nuclear power educational technology digital economy in the nuclear industry personnel development export of education Nuclear Energy and Technology 5(3): 201-206 The article briefly describes the history of training personnel for the nuclear industry and sets tasks for its improvement and development to ensure the future growth of this industry. Within the framework of the emerging digital economy, such a phenomenon as digital platforms erases the boundaries between industries, forming new unexpected industrial alliances, even new industries. Innovative activities in the power industry, including nuclear power, should provide the possibility of forming digital economic platforms in various energy segments as well as training personnel in using this new toolkit. Today, the Rosatom State Nuclear Energy Corporation is developing more than 30 projects of new nuclear power plants (NPP) in Russia and 12 other countries. This requires educational support, and for this purpose a Consortium of supporting universities of the Rosatom State Corporation was established, which includes 18 specialized higher educational institutions. More than half of them train personnel directly for designing, constructing and operating NPPs. The scale of the necessary personnel training in the near future indicates that we need a new “educational paradigm”, which can be described as “front-line education”, i.e., training personnel for developing digital economy technologies simultaneously along the entire “front”. This “front” stretches from schoolchildren preparing to enter universities to production personnel whose professional development should be carried out taking into account the specifics of the digital transformation of production. Partnership is one of the leading values of the modern young generation. To withstand high competition for the best personnel, organizations must not only be saturated with the culture of partnership from the inside but also act as reliable partners for one another in involving and training young employees. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.39239 https://doi.org/10.3897/nucet.5.39239 https://nucet.pensoft.net/article/39239/ https://nucet.pensoft.net/article/39239/download/pdf/ en

10.3897/nucet.5.39288 2019-09-25 nucet

Outcomes of the “steady-state crisis” experiment in the MIR reactor channel Alekseev,Aleksandr Dreganov,Oleg Izhutov,Aleksey Kiselyova,Irina Shulimov,Vitalij MIR reactor fuel element experimental fuel assembly (EFA) critical heat flux (CHF) RIA (reactivity-initiated accident) thermocouple (TC) temperature coolant flow rate Nuclear Energy and Technology 5(3): 207-212 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.39288 https://doi.org/10.3897/nucet.5.39288 https://nucet.pensoft.net/article/39288/ https://nucet.pensoft.net/article/39288/download/pdf/ en

10.3897/nucet.5.39290 2019-09-25 nucet

The study of the brittle fracture resistance in fusion areas between RPV steel 15H2NMFA grade 1 and austenitic padding Orlov,Viktor Anosov,Nikolay Skorobogatykh,Vladimir Gordyuk,Lyubov Yurgina,Zhanna Koshcheev,Konstantin Barsanova,Svetlana Shamardin,Valentin Assessment of brittle fracture resistance critical brittle point TK ductile-to-brittle-transition temperature TP conservative vessel life estimations padding and welded joints Nuclear Energy and Technology 5(3): 213-218 Brittle fracture resistance of RPV 15H2NMFA grade 1 steel is investigated. Sets of small-sized testing samples and a set of standard-sizes samples were used in the study. It was demonstrated that application of sets of small-sized specimens in mechanical tests for determining the brittle fracture resistance of RPV 15H2NMFA grade 1 steel makes possible the following: increasing the volume of tests in each batch by 8 times without significant changes in the design of irradiation facility thus ensuring maintaining the initial irradiation parameters during testing; substantially expanding the database of test results for statistical processing. The need for large-scale modeling of process conditions arising in weld joint zones inaccessible for direct testing, such as: (1) the welding zone between the base metal and the corrosion-resistant coating metal, (2) the welding area between the weld metal and the corrosion-resistant coating metal, and (3) the fusion area between the base metal, the weld metal, and the anticorrosive cladding metal, is demonstrated in the paper. Process modeling of the metal in fusion areas up to 0.5 mm wide (each is 100 μm in size) with an experimental electroslag refined (ESR) ingot of up to 300 mm long with similar microstructure and variable chemical composition allows the following: (1) examining not less than 1000 small-sized impact testing samples with continuous distribution of concentrations of chemical elements in accordance with a certain law; and (2) testing these samples and identifying brittle fracture dangerous zones across fusion areas between the base metal and the anti-corrosive padding metal in the initial state or after subsequent irradiation at a given fluence rate and temperature. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.39290 https://doi.org/10.3897/nucet.5.39290 https://nucet.pensoft.net/article/39290/ https://nucet.pensoft.net/article/39290/download/pdf/ en

10.3897/nucet.5.39318 2019-09-25 nucet

Development of the model to determine the fuel temperature field in a two-dimensional problem statement Gorbunov,Vladimir Ivanova,Natalya Lonshakov,Nikita Belov,Yaroslav Fuel element fuel temperature field analytical solution mathematical model to determine fuel temperature fields verification of numerical calculations influence of tuning coefficients safety of fuel element heating Nuclear Energy and Technology 5(3): 219-224 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.39318 https://doi.org/10.3897/nucet.5.39318 https://nucet.pensoft.net/article/39318/ https://nucet.pensoft.net/article/39318/download/pdf/ en

10.3897/nucet.5.39319 2019-09-25 nucet

Experimental studies of temperature pulsations during the process of mixing non-isothermal coolant flows in nuclear reactor equipment components Dmitriev,Sergey Mamaev,Alexandr Rayzapov,Renat Sobornov,Aleksey Kotin,Andrey Bescherov,Dmitry Bolshuhin,Mikhail Equipment life temperature pulsations coolant temperature field stress-strain state Nuclear Energy and Technology 5(3): 225-229 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.39319 https://doi.org/10.3897/nucet.5.39319 https://nucet.pensoft.net/article/39319/ https://nucet.pensoft.net/article/39319/download/pdf/ en

10.3897/nucet.5.39320 2019-09-25 nucet

Analytical version of the resonance coupled-channel model for D + T → 5He** → α + n reaction and its application for the description of low-energy D-T and D- 3He scattering Godes,Alexander Kudriavtseva,Anna Shablov,Vladimir Thermonuclear reactions resonance coupled-channel model effective range approximation S-matrix poles resonance and shadow poles Jπ = (3/2)+ state of 5He and 5Li nuclei Nuclear Energy and Technology 5(3): 231-235 The purpose of the present paper is the formulation of the analytical version of the resonance coupled-channel model (RCCM) originally developed for D + T → 5He** → α + n nuclear fusion reaction. The integral in the denominator of the Breit-Wigner type is examined in the expression for S-matrix elements of binary processes in this model. Imaginary part of this integral determines the energy-dependent decay width for the near-threshold channel. It is demonstrated that this integral can be calculated explicitly with the Binet representation for the ψ-function (the logarithmic derivation of the gamma function). As the result the explicit expression for the S-matrix elements in the form of analytical functions of the channel momenta are obtained and the equivalence of the RCCM and the effective range approximation (Landau – Smorodinsky – Bethe approximation) is established on this basis. This allows expressing the parameters of the RCCM through the model independent system characteristics: the complex scattering length and the complex effective range. Several sets of model parameters of both approaches that provide a good description of the measured data on D + T → α + n reaction and D-T elastic scattering are derived. By this means we find the location of the S – matrix poles on different Riemann sheets which corresponds to Jπ = (3/2)+ state of 5He and 5Li nuclei. In particular, the location of the resonance (R) and shadow (S) poles is determined: 5He**: ZR = 46.9 – i37.2 (keV) ZS = 81.7 – i3.5 (keV) 5Li**: ZR = 205.7 – i146.8 (keV) ZS = 264.4 + i112.0 (keV). Our results agree well with previous findings. The possible generalizations of the results obtained are discussed. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.39320 https://doi.org/10.3897/nucet.5.39320 https://nucet.pensoft.net/article/39320/ https://nucet.pensoft.net/article/39320/download/pdf/ en

10.3897/nucet.5.39340 2019-09-25 nucet

Experimental studies into the dependences of the axial lead coolant circulation pump performance on the pump straightening device parameters Beznosov,Aleksander Lvov,Aleksander Bokov,Pavel Bokova,Tatyana Lukichev,Nikita Heavy liquid metal coolant fast neutron reactor lead coolant axial pump pump head pump delivery rate pump impeller Nuclear Energy and Technology 5(3): 237-240 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.39340 https://doi.org/10.3897/nucet.5.39340 https://nucet.pensoft.net/article/39340/ https://nucet.pensoft.net/article/39340/download/pdf/ en

10.3897/nucet.5.46379 2019-09-25 nucet

Critical notes: history, state, problems and prospects of nuclear science and technology Murogov,Victor Nuclear power nuclear technologies system analysis of nuclear power Nuclear Energy and Technology 5(3): 241-248 This paper does not contain new computational and experimental scientific results. It attempts to analyze, based on a simplified phenomenological approach, the methodology of the evolution histories of nuclear science and technology, as well as the contradictions and issues which, if not resolved, make senseless any discussions of scenarios for the full-scale evolution of nuclear power. The paper analyzes in brief the evolution history of nuclear technologies in the USA and in the USSR. It also considers the present-day state of nuclear power. Two international projects, INPRO and GIF IV, were initiated in 2000. The INPRO objective is to define the evolution strategy for and the requirements to the nuclear power of tomorrow. The GIF IV project aiming to develop Generation IV reactors for future NPPs focuses on building innovative reactors capable to cope with the challenges involved in further evolution of nuclear power. The following issues were considered as the result of the system analysis – further evolution of nuclear power worldwide; – nuclear non-proliferation; – NPP safety; – nuclear waste; – climate and oxygen burning in the NPP operation; – education and training of younger generations of nuclear workers. A critical analysis into the history, status and future evolution of nuclear technologies at the present-day stage shows that the nuclear energy market has monopolized the design, development and construction of practically only one type of nuclear reactors for NPPs (95% of the NPPs under construction have water-cooled water-moderated reactors) which explains the fact that single-skilled personnel are largely trained for the construction and operation of this plant type. Achieving the full-scale evolution level of nuclear power capable to cope with the socio-economic and ecological issues faced by humankind requires a basically new evolution concept for all fields of nuclear industry. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Review Article text/html info:doi:10.3897/nucet.5.46379 https://doi.org/10.3897/nucet.5.46379 https://nucet.pensoft.net/article/46379/ https://nucet.pensoft.net/article/46379/download/pdf/ en

10.3897/nucet.5.46380 2019-09-25 nucet

Some aspects of the VVER-440 reactor plant life re-extension: a case study of the Novovoronezh NPP Unit 4 Povarov,Vladimir Fedorov,Anatoly Vitkovsky,Sergey Life re-extension VVER-440 V-179 fluence radiation embrittlement safety systems reactor vessel annealing Nuclear Energy and Technology 5(3): 249-256 The re-modernization of Unit 4 at the Novovoronezh NPP (Novovoronezh-4) made it possible to take a new approach to the problem of extending the VVER-440 reactor plant life and operation. The authors analyze the existing problems of the VVER-440/179 power unit, showing possible solutions to the identified shortcomings and the final state of the updated power unit. Modernization works significantly expanded the range of design-basis accidents from the primary coolant leak from an opening (DN = 100 mm) to the maximum possible, associated with a rupture of the main circulation pipelines (MCP) (DN = 500 mm). A unique experience was gained in using the safety systems of Unit 3, which was finally shutdown for decommissioning, to increase reliability and provide additional redundancy for the safety systems of Unit 4. The results of the performed works showed the correctness of the adopted concept of re-extending the service life of Unit 4 and ensured its compliance with the modern safety requirements in nuclear power engineering, including as it relates to the safety impact of the first-level probabilistic safety analysis model (PSA-1) for internal initiating events. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.46380 https://doi.org/10.3897/nucet.5.46380 https://nucet.pensoft.net/article/46380/ https://nucet.pensoft.net/article/46380/download/pdf/ en

10.3897/nucet.5.46381 2019-09-25 nucet

Problem of nuclear-laser power engineering and methods of their solution Dyachenko,Petr Suvorov,Alexey Kukharchuk,Oleg Zrodnikov,Anatoly Multi-core nuclear reactor neutrons fission fragments nuclear pumping laser optical quantum amplifier with nuclear pumping Nuclear Energy and Technology 5(3): 257-263 The concept of a high power reactor-laser system based on a nuclear pumped optical quantum amplifier (OKUYaN) was formulated at IPPE in the mid-1980-ies. The idea amounted to the use of wide-aperture OKUYaN as an amplifier within the already well-known “master laser – two-pass amplifier with phase conjugation” scheme. The structure of such an amplifier includes a system of two neutron-coupled units – an ignition reactor (RB) and a nuclear pumped laser amplifier (LB). The ignition unit is a compact multi-core pulsed fast neutron reactor. The laser amplifier unit operates on thermal neutrons and, with regard to the neutronics, it is a subcritical booster zone of the ignition reactor unit. Unique reactor-laser complex incorporating demonstration sample of a pulsed reactor-laser system based on OKUYaN (test facility “Stand B”) having no analogues anywhere in the world, was developed and put into operation at IPPE in 1999 for the purpose of substantiation of basic principles of the OKUYaN concept and demonstration of the possibility of its practical implementation, as well as verification of calculation codes and development of relevant equipment elements. Problems overcome in the development and construction of “Stand B” test facility, the choice and justification of the neutronics and laser characteristics of the OKUYaN demonstration sample are discussed in the present paper. Provided are the results of a detailed computational-experimental study of the demonstration sample characteristics, the data from systems studies of direct conversion of nuclear fission energy into laser radiation energy in complex reactor-laser devices and the results of examination of prospects for the development of nuclear-laser power engineering. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.46381 https://doi.org/10.3897/nucet.5.46381 https://nucet.pensoft.net/article/46381/ https://nucet.pensoft.net/article/46381/download/pdf/ en

10.3897/nucet.5.46382 2019-09-25 nucet

Prospects of creation of the two-component nuclear energy system Petrov,Andrey Shutikov,Alexander Ponomarev-Stepnoy,Nikolay Bezzubtsev,Valery Bakanov,Mikhail Troyanov,Vladimir Two-component nuclear energy system centralized nuclear fuel cycle commercial power supply complex of the Supplier (Rosatom) Nuclear Energy and Technology 5(3): 265-271 Possible options of organization of two-component energy system with closed nuclear fuel cycle (CNFC) and new business potential associated with provision of CFC services to foreign customers are examined. Dominating role in the development of nuclear power generation is assigned to VVER reactors with gradually increasing fraction of sodium-cooled fast breeder reactors (FR) incorporated within the joint nuclear fuel cycle operated on MOX-fuel. Components of such energy system perform the following functions: 1. Fast reactors: Generate electric power in base-load mode (possibility of fine tuning of reactor power within limited range (100 – 75 – 100%) is assumed); Utilize waste and/or regenerated uranium for re-fueling power reactors, produce plutonium applicable to the maximum extent for manufacturing MOX-fuel for VVER reactors; Incinerate long-lived highly radioactive wastes – minor actinides separated during reprocessing spent nuclear fuel of FR and VVER reactors. 2. VVER reactors: Generate electricity in compliance with manoeuvrability requirements imposed by the utility company operating the energy system; Utilize MOX-fuel instead of UO2 fuel; Are offered for export with the option of returning SNF back to Russia; Plutonium extracted from VVER spent fuel is used for manufacturing MOX-fuel for SFR. 3. Nuclear fuel cycle facilities: Provide reprocessing of SNF from VVER and SFR with extraction of nuclear materials for recycling; Use depleted or reprocessed uranium and plutonium extracted from spent nuclear fuel for manufacturing MOX-fuel; Provide partitioning of RAW for subsequent utilization of minor actinides and reduction of risks of proliferation of nuclear materials, conditioning and disposal of RAW. Russia possesses capacities for establishing the two-component system with CNFC, as well as the new business approach to rendering CNFC services to foreign customers. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.46382 https://doi.org/10.3897/nucet.5.46382 https://nucet.pensoft.net/article/46382/ https://nucet.pensoft.net/article/46382/download/pdf/ en

10.3897/nucet.5.39226 2019-09-25 nucet

Semantic web and knowledge graphs as an educational technology of personnel training for nuclear power engineering Telnov,Victor Korovin,Yury Nuclear education semantic web knowledge graph cloud computing Nuclear Energy and Technology 5(3): 273-280 The technologies of knowledge representation and inference in an artificial intelligence system focused on the domain of nuclear physics and nuclear power engineering are considered. The possibilities of description logics and graph databases of nuclear knowledge for the generation of cognitive hypotheses, using in addition to deduction and other ways of reasoning, such as inductive inference and reasoning based on analogies, are discussed. The use of adequate description logic and measures of semantic similarity is substantiated. Interactive visual navigation and reasoning on the knowledge graphs are performed by means of special retrieval widgets and the smart RDF browser. Operations with semantic repositories are implemented on cloud platforms using SPARQL queries and RESTful services. The proposed software solutions are based on cloud computing using DBaaS and PaaS service models to ensure scalability of data warehouses and network services. Example of use of the offered technologies and software has been given. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.39226 https://doi.org/10.3897/nucet.5.39226 https://nucet.pensoft.net/article/39226/ https://nucet.pensoft.net/article/39226/download/pdf/ en

10.3897/nucet.5.47972 2019-12-10 nucet

Heat transfer intensification in emergency cooling heat exchanger and dry cooling towers on nuclear power plant using air-water mist flow Abed,Akram Hamzah Shcheklein,Sergey Pakhaluev,Valery Nuclear power plants aerosol cooling air-water mist water concentration heat exchange intensification. Nuclear Energy and Technology 5(4): 281-287 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.47972 https://doi.org/10.3897/nucet.5.47972 https://nucet.pensoft.net/article/47972/ https://nucet.pensoft.net/article/47972/download/pdf/ en

10.3897/nucet.5.48391 2019-12-10 nucet

Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up Bulakh,Olga Kostylev,Oleg Nesterov,Vladimir Cherdizov,Eldar Critical fluence reactor-grade graphite damaging neutrons graphite lifespan nuclear fuel burnup GT-MHR HTGR Nuclear Energy and Technology 5(4): 289-295 High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.48391 https://doi.org/10.3897/nucet.5.48391 https://nucet.pensoft.net/article/48391/ https://nucet.pensoft.net/article/48391/download/pdf/ en

10.3897/nucet.5.48392 2019-12-10 nucet

Temperature field in gas-cooled reactor core in transient conditions under different approaches to mass flow profiling Kuzevanov,Vyacheslav Podgorny,Sergey Reactor core profiling transient processes temperature field porous body thermal conductivity equation heat sink Nuclear Energy and Technology 5(4): 297-303 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.48392 https://doi.org/10.3897/nucet.5.48392 https://nucet.pensoft.net/article/48392/ https://nucet.pensoft.net/article/48392/download/pdf/ en

10.3897/nucet.5.48393 2019-12-10 nucet

A study into the modes of the VVER-1000 RCP starting in an earlier inoperative loop Bragin,Iliya Belozerov,Vladimir Mode reactor coolant pump (RCP) circulation loop reactor plant scram reactivity coefficient safety margin reactor plant power Nuclear Energy and Technology 5(4): 305-311 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.). info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.48393 https://doi.org/10.3897/nucet.5.48393 https://nucet.pensoft.net/article/48393/ https://nucet.pensoft.net/article/48393/download/pdf/ en

10.3897/nucet.5.48394 2019-12-10 nucet

Roughness of the nuclear reactor pipe inner surface depending on the reactor operating time Trofimov,Maksim Globa,Ruslan Life extension surface roughness metal corrosion operating conditions Elcometer 7061 Marsurf PS1 roughness meter parent metal preventive maintenance Nuclear Energy and Technology 5(4): 313-316 During operation of nuclear reactors, there are various factors that affect the nuclear plant piping leading to erosion of the pipe inner surface and an increase of its micro-relief (roughness). Metal corrosion occurs and spreads faster on a surface having a higher value of the roughness parameter. Failure through erosive wear of the parent metal takes place predominantly in the pipe bend area. The roughness of the pipe inner surface has a sizeable effect on the signal attenuation in the process of the pipe wall ultrasonic testing. Defective main pipeline segments were cut out during preventive repairs from which samples with different operating times were taken. Five defective pipe segments of the austenitic 12Kh18N10T grade steel cut out of a high-pressure reheater’s piping and five defective pipe segments of the perlite-class steel of grade 20, after different operating times, were used to determine experimentally the actual value of the pipe inner surface roughness. Besides, a piece of a new Ø273 × 12 pipe of the 12Kh18N10T steel and a piece of a Ø159 × 6 pipe of grade 20 steel were cut out. The inner surface roughness was measured for different segments. Dependences of the roughness value on the operating time and the pipe segment type have been obtained. Company specimens were fabricated with the inner surfaces having the roughness corresponding to various pipe operating times. This made it possible to take into account the influence of the inner surface roughness on the signal attenuation in the process of the weld integrity ultrasonic testing and during ultrasonic measurements of the weld adjacent zone grain size value following the weld repair. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.48394 https://doi.org/10.3897/nucet.5.48394 https://nucet.pensoft.net/article/48394/ https://nucet.pensoft.net/article/48394/download/pdf/ en

10.3897/nucet.5.48397 2019-12-10 nucet

Calculations of research reactor thermal hydraulics based on VVER-440 fuel assamblies Doan,Thi Zieu Chang Lazarenko,Georgy Lazarenko,Denis Research reactor low-enriched fuel natural circulation long-term campaign export potential VVR IRT VVER-440 Nuclear Energy and Technology 5(4): 317-321 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.48397 https://doi.org/10.3897/nucet.5.48397 https://nucet.pensoft.net/article/48397/ https://nucet.pensoft.net/article/48397/download/pdf/ en

10.3897/nucet.5.48423 2019-12-10 nucet

Analyzing the causes for the dispersion of the fast reactor spent fuel rod cladding properties Barsanova,Svetlana Kinev,Evgeniy Kozlov,Aleksander Portnykh,Irina Panchenko,Valeriy Evseev,Mikhail Fuel rod cladding swelling rate corrosion grain size high-temperature embrittlement manufacturing technology. Nuclear Energy and Technology 5(4): 323-329 The swelling, corrosion and high-temperature embrittlement behavior of the fast-neutron sodium-cooled reactor standard and test fuel rod claddings was studied following the operation up to a damaging dose of 55 to 69 dpa. The tested characteristics were found to differ sensitively in conditions similar to irradiation for the claddings of the experimental tube conversion technology. Unlike the standard fuel rod claddings, the test rod claddings were additionally heated in the process of fabrication to homogenize the solid solution at different temperatures and austenitization times. On the whole, this led to an increased cladding resistance contrary the damaging factor of the reactor environment. The positive effect is explained by the influence of carbon and the morphology of swelling-reducing alloying elements, as well as by the nature of the carbide and intermetallide phase precipitation. However, the dispersion of the post-irradiation properties which remained significant and was also earlier observed in the standard rods is explained by potential differences in the heat treatment technology and the irradiation temperature in conditions of a hard-to-control coolant flow velocity. The swelling rate and the in-fuel corrosion depth for the test technology tubes were respectively 0.04 to 0.058%/dpa and 20 to 47 μm; similar values for the test material are 0.036 to 0.056%/dpa and 15 to 35 μm respectively. The short-term mechanical properties of the test fuel rods at a temperature of 600 °C showed a smaller tendency towards high-temperature embrittlement. The dispersion of the properties was caused by the chemical and structural heterogeneity as the result of the tube fabrication. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.48423 https://doi.org/10.3897/nucet.5.48423 https://nucet.pensoft.net/article/48423/ https://nucet.pensoft.net/article/48423/download/pdf/ en

10.3897/nucet.5.48425 2019-12-10 nucet

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 Ivanov,Igor Shelemetyev,Vasily Askhadullin,Radomir Heavy liquid metal coolants (HLMC) nuclear safety hydrogen tritium hydrogen purification of HLMC hydrogen afterburner bismuth oxide reaction kinetics thermal-conductivity analysis affine time transformation method Avrami-Erofeev equation. Nuclear Energy and Technology 5(4): 331-336 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.48425 https://doi.org/10.3897/nucet.5.48425 https://nucet.pensoft.net/article/48425/ https://nucet.pensoft.net/article/48425/download/pdf/ en

10.3897/nucet.5.48427 2019-12-10 nucet

Corrosion tests in water of fuel elements irradiated in the world’s first NPP reactor Ivanov,Sergey Porollo,Sergey Baranaev,Yury Timofeev,Vladimir Kharizomenov,Yury Fuel pin fuel assembly (FA) spent fuel assembly (SFA) corrosion defect aquatic environment artificial defect fuel composition metallographic studies γ-dose rate material microstructure fuel element (FE) claddings Bilibino NPP Nuclear Energy and Technology 5(4): 337-343 Spent nuclear fuel (SNF) storage in reactor spent fuel pools (SFP) is one of the crucial stages of SNF management technology: it requires special measures to ensure nuclear and radiation safety. During long-term storage in water-filled SFPs, leak-tight canisters in which SFAs are usually placed can become unsealed, which will result in the development of corrosion processes in the fuel element (FE) claddings. We studied fragments of spent fuel elements of the AM reactor of the World’s First NPP during their long exposure in the aqueous medium. The aim of the study was to obtain experimental data on the corrosion changes in the FE claddings and fuel composition during storage as well as on the release of radioactive fission products from them. For the study, a laboratory facility for exposing fuel elements in the water was developed and experimental fragments of fuel elements were made. The study was carried out in the hot chamber of the SSC RF-IPPE. The change in the activity of the water was estimated by the γ-dose rate from the selected water sample. The diameter measurements and metallographic studies were carried out in various sections of FE fragments. Corrosion tests were carried out on fragments of spent fuel elements of the AM reactor of the World’s First NPP that were stored for a long time (more than 50 years – FEs with U-Mo fuel and ~ 20 years – FEs with UO2 fuel) using standard technology – first in SFP canisters filled with water and then in dry canisters in the air. Placing the fuel elements in the water did not lead to through damage to the FE claddings and a significant change in the size (diameter) of the outer cladding. Metallographic studies of the FE fragments after the corrosion tests showed the presence of intergranular and local frontal corrosion on the surface of the claddings, the depth of which exceeded the depth of the cladding corrosion defects before testing. The rate of radionuclide release from the fuel composition was estimated by the γ-dose rate of water samples taken from the glasses with FE fragments. Throughout the test period, the dose rate of water samples from the glasses with defect-free FEs remained constant. The dose rate from water samples taken from the glasses with the FE fragments with an artificial defect grew during storage. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.48427 https://doi.org/10.3897/nucet.5.48427 https://nucet.pensoft.net/article/48427/ https://nucet.pensoft.net/article/48427/download/pdf/ en

10.3897/nucet.5.48426 2019-12-10 nucet

Heterogeneous effects in simulating a fast nuclear reactor on the BFS test facility Kazansky,Yury Karpovich,Gleb Heterogeneous effect simulation of nuclear reactors critical test benches fast reactor BN sodium void reactivity effect neutronics calculations Monte-Carlo Nuclear Energy and Technology 5(4): 345-351 Simulating fast neutron reactor cores for comparing experimental and calculated data on the reactor neutronics characteristics is performed using zero power test stands. The BFS test facilities in operation in Russia (Obninsk) are discussed in the present paper. The geometrical arrangement of materials in the cores of the simulated reactors (fuel pins, fuel assemblies, coolant geometry) differs from the simulation assembly on the BFS. This can cause differences between the experimental results obtained at the BFS and theoretical calculations even in the case when homogenized concentrations of all materials of the reactor are thoroughly observed. The resulting differences in neutronics parameters due to the geometry of arrangement of materials with the same homogeneous concentrations are referred to as the heterogeneous effect. Heterogeneous effects tend to increase with increasing reactor power and its size, mainly due to changes in the neutron spectra. Calculations of a number of functional values were carried out for assessing the heterogeneous effects for different spatial arrangements of the reactor materials. The calculations were performed for the following cases: a) heterogeneous distribution of materials in accordance with the design of a fast reactor; b) heterogeneous arrangement of materials in accordance with the capabilities and design features of the BFS test facility; c) homogeneous representation of materials in the reactor core and breeding blankets. The configuration of materials in accordance with the design data for fast reactors of the BN-1200 type was accepted as the basic calculation option, relative to which the effect called the heterogeneous shift of the functional value (HSF) was calculated. The effect of neutron leakage on the HSF obtained as the result of calculations using different boundary conditions was estimated. All calculations were carried out for the same homogeneous concentrations of all materials for all the above three configurations. Calculations were carried out as well for the case when plutonium metal fuel was used in the BFS. The values of the following functionals were calculated for different cases of arrangement of materials: the effective multiplication factor (reactivity), the sodium void reactivity effect, the average energy of fission-inducing neutrons, and the ratios of radioactive capture cross-sections to fission cross-sections for 239Pu. The calculations were performed using the Serpent 2.1.30 (VTT, Finland) Monte Carlo software package for neutronics simulations and ENDF/B-VII.0 and JEFF-3.1.1 evaluated nuclear data libraries. The effects of various options of material arrangement on the values of keff were found to be the greatest (about 1.6%) for the case when fissile material in the form of dioxide is replaced with metal fissile material. Homogenization of the composition reduces the keff value by about 0.4%. The average energy of fission-inducing neutrons depends to a significant extent on the leakage of neutrons and the presence of sodium (the average energy of neutrons increases and reaches in the presence of sodium about 100 keV, that is, it increases by about 11–13%). Replacing fissile material metal with its dioxide in the BFS test facility (while maintaining homogeneous concentrations, including that of oxygen) allows reducing the average energy of fission-inducing neutrons by about 60 keV. The highest values of HSF, reaching 65%, are observed when calculation of sodium void reactivity effect is performed with materials distributed homogeneously; however, HSF is equal to 1.5% when calculation of the reactor mock-up assembled on the BFS is performed. In the absence of neutron leakage (infinitely extended medium), the sodium void reactivity effect becomes positive and the HSF is equal to 4–7%. The heterogeneous effect of α for 239Pu noticeably (6–8%) depends only on the replacement of metallic plutonium with its dioxide (maintaining, of course, the homogeneous concentrations). info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.48426 https://doi.org/10.3897/nucet.5.48426 https://nucet.pensoft.net/article/48426/ https://nucet.pensoft.net/article/48426/download/pdf/ en

10.3897/nucet.5.46517 2019-12-10 nucet

Minor actinides transmutation in equilibrium cores of next generation FRs Egorov,Alexander Khomyakov,Yurii Rachkov,Valerii Rodina,Elena Suslov,Igor Fast reactor nitride fuel spent nuclear fuel minor actinides reactivity margin equilibrium cycle Nuclear Energy and Technology 5(4): 353-359 The Russian Federation is developing a number of technologies within the «Proryv» project for closing the nuclear fuel cycle utilizing mixed (U-Pu-MA) nitride fuel. Key objectives of the project include improving fast reactor nuclear safety by minimizing reactivity changes during fuel operating period and improving radiological and environmental fuel cycle safety through Pu multi-recycling and МА transmutation. This advanced technology is expected to allow operating the reactor in an equilibrium cycle with a breeding ratio equaling approximately 1 with stable reactivity and fuel isotopic composition. Nevertheless, to reach this state the reactor must still operate in an initial transient state for a lengthy period (over 10 years) of time, which requires implementing special measures concerning reactivity control. The results obtained from calculations show the possibility of achieving a synergetic effect from combining two objectives. Using МА reprocessed from thermal reactor spent fuel in initial fuel loads in FR ensures a minimal reactivity margin during the entire fast reactor fuel operating period, comparable to the levels achieved in equilibrium state with any kind of relevant Pu isotopic composition. This should be combined with using reactivity compensators in the first fuel micro-campaigns. In the paper presented are the results of simulation of the overall life cycle of a 1200 MWe fast reactor, reaching equilibrium fuel composition, and respective changes in spent fuel nuclide and isotopic composition. It is shown that МА from thermal and fast reactors spent fuel can be completely utilized in the new generation FRs without using special actinide burners. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.46517 https://doi.org/10.3897/nucet.5.46517 https://nucet.pensoft.net/article/46517/ https://nucet.pensoft.net/article/46517/download/pdf/ en

10.3897/nucet.5.39096 2019-12-10 nucet

Different distributions of gold nanoparticles on the tumor and calculation of dose enhancement factor by Monte Carlo simulation Keshavarz,Sajad Sardari,Dariush dose enhancement gold nanoparticles distribution nanoparticle size Monte Carlo Nuclear Energy and Technology 5(4): 361-371 Gold nanoparticles can be used to increase the dose of the tumor due to its high atomic number as well as being free from apparent toxicity. The aim of this study is to evaluate the effect of distribution of gold nanoparticles models, as well as changes in nanoparticle sizes and spectrum of radiation energy along with the effects of nanoparticle penetration into surrounding tissues in dose enhancement factor DEF. Three mathematical models were considered for distribution of gold nanoparticles in the tumor, such as 1-uniform, 2- non-uniform distribution with no penetration margin and 3- non-uniform distribution with penetration margin of 2.7 mm of gold nanoparticles. For this purpose, a cube-shaped water phantom of 50 cm size in each side and a cube with 1 cm side placed at depth of 2 cm below the upper surface of the cubic phantom as the tumor was defined, and then 3 models of nanoparticle distribution were modeled. MCNPX code was used to simulate 3 distribution models. DEF was evaluated for sizes of 20, 25, 30, 50, 70, 90 and 100 nm of gold nanoparticles, and 50, 95, 250 keV and 4 MeV photon energies. In uniform distribution model the maximum DEF was observed at 100 nm and 50 keV being equal to 2.90, in non-uniform distribution with no penetration margin, the maximum DEF was measured at 100 nm and 50 keV being 1.69, and in non-uniform distribution with penetration margin of 2.7 mm, the maximum DEF was measured at 100 nm and 50 keV as 1.38, and the results have been showed that the dose was increased by injecting nanoparticles into the tumor. It is concluded that the highest DEF could be achieved in low energy photons and larger sizes of nanoparticles. Non-uniform distribution of gold nanoparticles can increase the dose and also decrease the DEF in comparison with the uniform distribution. The non-uniform distribution of nanoparticles with penetration margin showed a lower DEF than the non-uniform distribution without any margin and uniform distribution. Meanwhile, utilization of the real X-ray spectrum brought about a smaller DEF in comparison to mono-energetic X-ray photons. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2019 Research Article text/html info:doi:10.3897/nucet.5.39096 https://doi.org/10.3897/nucet.5.39096 https://nucet.pensoft.net/article/39096/ https://nucet.pensoft.net/article/39096/download/pdf/ en

10.3897/nucet.6.50863 2020-03-11 nucet

A study into the structure and physical properties of the Cr18Ni9-grade steel following long-term irradiation as part of the BN-600 reactor internals Portnykh,Irina Kozlov,Aleksander Panchenko,Valeriy Barsanova,Svetlana Internals austenitic steel neutron irradiation porosity characteristics irradiation-induced swelling elasticity characteristics Nuclear Energy and Technology 6(1): 1-6 The microstructures and physical properties of the austenitic Cr18Ni9-grade steel after 22 and 33 years of operation as part of the reactor internals were tested for assessing the conditions of the BN-600 reactor non-replaceable components (internals) and the potential of their subsequent use in predicting the reactor ultimate life. The paper presents histograms of the porosity distribution depending on the void size, in samples taken from portions that were subjected to neutron irradiation with displacement rates ranging from 1.0×10–9 to 4.3×10–8 dpa/s at temperatures from 370 to 440 °C. The elasticity characteristics were measured by resonance-type ultrasonic technique for the samples taken from the same portions of material. It was demonstrated that swelling calculated using the histograms of the porosity distribution depending on the void size has the maximum value at ~415 °C and after 33 years of irradiation reaches values of ~3%. Long-term variations of Young’s modulus demonstrate non-monotonous dependence on the damage dose. The maximum relative variation of Young’s modulus after 22 and 33 years of operation does not exceed 2% and 6%, respectively, of the values corresponding to the initial state. It was shown that along with the irradiation-induced swelling the changes in the physical properties are also affected in the process of irradiation by other structural changes and, in particular, by the formation of secondary phases. As shown by the results of the studies, operation of the BN-600 reactor internals made of Cr18Ni9-grade steel can be extended beyond 33 years of service. The comparison of the results obtained for the material after 22 and 33 years of operation contains information required for describing subsequent changes of the structure and properties of the Cr18Ni9 internals. The obtained results can be used for forecasting the reactor ultimate life within the framework of existing and developed models. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2020 Research Article text/html info:doi:10.3897/nucet.6.50863 https://doi.org/10.3897/nucet.6.50863 https://nucet.pensoft.net/article/50863/ https://nucet.pensoft.net/article/50863/download/pdf/ en

10.3897/nucet.6.50865 2020-03-11 nucet

Circuit design solutions for the reactimeters Yuferov,Anatoliy Reactimeter instrumental function circuit design solution variant analysis Nuclear Energy and Technology 6(1): 7-14 A number of issues pertaining to comparative analysis of possible options for algorithmic and circuit embodiment of reactimeters were examined from the standpoint of the general theory of measuring instruments and the theory of digital filters. Structural diagrams of the linear part of the reactimeter, as well as the functional algorithms and their numerical implementation are described in terms of transient characteristics and transfer functions. Parallel, straight, canonical, symmetrized, lattice and ladder block structural diagrams are examined. The corresponding difference equations are given. The obtained results allow comparing possible circuit design solutions from the viewpoint of a number of criteria: the complexity of the elemental composition (the number of integrators, summation units, multipliers, delay elements), the number of necessary computing operations, the identifiability of the hardware function of the reactimeter, the coherence between the calculated and the measured values, the sensitivity to parameter uncertainties, etc. The possibility of considering the equations of the reactimeter as autoregressive is demonstrated, which ensures adaptability of the reactimeter under operating conditions. Certain algorithms for identification of the transient response characteristic and transfer function of the reactimeter are indicated. The possibility is shown of using identical algorithms in the main computing unit for solving the direct and inverse problems of nuclear reactor kinetics for ensuring consistency between the calculated and the measured reactivity values. Upper and lower estimated reactivity values are suggested for the moment of switching on the reactimeter. Implementation of such estimations in the reactimeter design allows minimizing the time needed for reaching by the reactimeter of the operating mode. Certain methodological simplifications were used in the development of ladder and lattice circuit design solutions. The database containing parameters of the instrumental functions of the circuit design solutions of the reactimeters is available on a public website. A number of tasks and directions for further research are identified. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2020 Research Article text/html info:doi:10.3897/nucet.6.50865 https://doi.org/10.3897/nucet.6.50865 https://nucet.pensoft.net/article/50865/ https://nucet.pensoft.net/article/50865/download/pdf/ en

10.3897/nucet.6.50867 2020-03-11 nucet

Safety features of fast reactor with heavy atomic weight weakly neutron absorbing reflector Kulikov,Gennady Shmelev,Anatoly Apse,Vladimir Kulikov,Yevgeny Nuclear safety reactivity accident delayed neutrons fast reactor radiogenic lead Nuclear Energy and Technology 6(1): 15-21 The purpose of the present study is the justification of the possibility of improving fast reactor safety by surrounding reactor cores with reflectors made of material with special neutron physics properties. Such properties of 208Pb lead isotope as heavy atomic weight, small neutron absorption cross section, and high inelastic scattering threshold result in certain peculiarities in neutron kinetics of the fast reactor equipped with 208Pb reflector, which can significantly enhance reactor safety. The reflector will also make possible generation of additional delayed neutrons characterized by the “dead” time. This will improve the resistibility of the fission chain reaction to stepwise reactivity excursions and exclude prompt supercriticality. Let us note that generation of additional delayed neutrons can be shaped by reactor designers. The relevance of the study amounts to the fact that generation of additional delayed neutrons in the reflector will make it possible mitigating the consequences of a reactivity accident even if the introduced reactivity exceeds the effective fraction of delayed neutrons. At the same time, the role of the fraction of delayed neutrons as the maximum permissible reactivity for reactor safety is depreciated. Scientific originality of the study pertains to the fact that the problem of yield of additional neutrons with properties close to normal delayed neutrons, has not been posed before. The authors suggest a new method for enhancing safety of fast reactors by increasing the fraction of delayed neutrons due to the time delay of prompt neutrons during their transfer in the reflector. In order to benefit from the expected advantages, the following combination is acceptable: lead enriched by 208Pb is used as a neutron reflector while natural lead or other material (sodium, etc.) is used as a coolant in the reactor core. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2020 Research Article text/html info:doi:10.3897/nucet.6.50867 https://doi.org/10.3897/nucet.6.50867 https://nucet.pensoft.net/article/50867/ https://nucet.pensoft.net/article/50867/download/pdf/ en

10.3897/nucet.6.50868 2020-03-11 nucet

Reactor with metallic fuel and lead-208 coolant Khorasanov,Georgiy Blokhin,Anatoliy Small reactor plutonium-americium-neptunium fuel lead-208 coolant incineration of minor actinides Nuclear Energy and Technology 6(1): 23-27 The paper considers the concept of a fast lead cooled 25MW reactor for a variety of applications, including incineration of minor actinides, production of medical radioisotopes, testing of radiation-damaged nuclear technology materials, etc. A specific feature of the proposed reactor is rather a high neutron flux of 2.6·1015 n/(cm2·s) at the core center, high average neutron energy of 0.95 MeV at the core center, and a large fraction (40%) of hard neutrons (En > 0.8 MeV). The extremely high estimated reactor parameters are achieved thanks to the small core dimensions (DxH ≈ 0.50×0.42 m2), innovative metallic fuel of the Pu-Am-Np-Zr alloy, and the 208Pb enriched lead coolant. A relatively high probability of 241Am fission (about 50%) is achieved in the reactor core’s hard spectrum, this making it possible to incinerate up to 4 kg of 241Am during one reactor campaign of 1000 effective days. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2020 Research Article text/html info:doi:10.3897/nucet.6.50868 https://doi.org/10.3897/nucet.6.50868 https://nucet.pensoft.net/article/50868/ https://nucet.pensoft.net/article/50868/download/pdf/ en

10.3897/nucet.6.51252 2020-03-27 nucet

A localization method for loose parts monitoring system of VVER reactor plants Maksimov,Ivan Vladimir,Vladimir Loose Parts Monitoring System (LPMS) diagnostics of NPPs impact localization acoustic wave main circulation circuit VVER reactor plant Nuclear Energy and Technology 6(1): 29-35 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. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2020 Research Article text/html info:doi:10.3897/nucet.6.51252 https://doi.org/10.3897/nucet.6.51252 https://nucet.pensoft.net/article/51252/ https://nucet.pensoft.net/article/51252/download/pdf/ en

10.3897/nucet.6.51253 2020-03-27 nucet

Development and verification of CFD model of carbothermal synthesis furnace for production of mixed nitride uranium-plutonium fuel Shamsutdinov,Rinat Pavlov,Sergey Leshchenko,Anton CFD model MNUP fuel SolidWorks Flow Simulationv furnace carbothermal synthesis mathematical model grid model Nuclear Energy and Technology 6(1): 37-42 The retort batch furnace for the carbothermal synthesis of uranium and plutonium nitrides is a component of the mixed nitride uranium-plutonium (MNUP) fuel Fabrication/Refabrication Module (FRM), a part of the Pilot Demonstration Energy Complex (PDEC). A CFD model of the furnace was built in the SolidWorks Flow Simulation code to check the feasibility of its thermophysical operating modes (heating and cooling rates, temperature of the product loaded into the furnace). The experimental data obtained from the performance tests was used to verify the developed CFD model and to confirm its adequacy. The relative deviation of the calculated temperature of the loaded product from the experimental data in the process of isothermal annealing does not exceed 0.7%. The temperatures of the loaded product predicted by the CFD model were used to justify engineering solutions for the uranium and plutonium nitrides carbothermal synthesis furnace. The CFD model can be used to define the furnace operation mode by selecting the gas flow rates inside and outside the retort, the heater temperature, and heating and cooling rates for the product loaded in the furnace. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2020 Research Article text/html info:doi:10.3897/nucet.6.51253 https://doi.org/10.3897/nucet.6.51253 https://nucet.pensoft.net/article/51253/ https://nucet.pensoft.net/article/51253/download/pdf/ en

10.3897/nucet.6.51778 2020-03-27 nucet

Problems of radiation safety calculations related to spent nuclear fuel transport casks Sobolev,Artem Danilov,Pavel SNF transport cask radiation safety neutron source gamma source radiation protection calculation Nuclear Energy and Technology 6(1): 43-47 The paper discusses the stages of calculating the radiation safety of spent nuclear fuel (SNF) transport packages, in particular, transport casks and some related problems. The problem of describing the source of neutrons and gamma radiation of spent nuclear fuel is shown. For individual designs of fuel assemblies, data are given on isotopes that make the main contribution to the neutron source as well as on gamma rays in nuclear fuel material and structural materials. The authors emphasize the necessity of analyzing the influence of the initial spent fuel parameters on the formation of the radiation spectrum and, therefore, on the radiation situation around the transport casks. Consideration is given to the problem of assessing the attenuation of gamma radiation in calculating protection analytically and using software. Due to the ambiguity of the position of the zone with the highest effective dose value on the SNF transport cask surface, it is indicated that preliminary estimates are required to take into account all radiation sources and their nonuniformities. All the problems presented in the paper are currently being solved by means of rather complex and voluminous calculations that take a long time. In order to be able to conduct a preliminary assessment of the radiation situation around the transport casks, the authors propose to create a methodology that will determine the type of interrelations between the maximum effective dose and input parameters, such as fuel burnup, decay, fuel composition, protection material in the SNF transport cask, etc. This methodology will make it possible to improve the efficiency of the process of designing the SNF transport casks, avoid possible design errors and, in particular, when used as intended, resolve the issue of the SNF cask loading configuration. info:eu-repo/semantics/altIdentifier/eissn/2452-3038 info:eu-repo/semantics/openAccess CC BY 4.0 Nuclear Energy and Technology 2020 Research Article text/html info:doi:10.3897/nucet.6.51778 https://doi.org/10.3897/nucet.6.51778 https://nucet.pensoft.net/article/51778/ https://nucet.pensoft.net/article/51778/download/pdf/ en cGFnZT0xJnNldD1udWNldCZmcm9tPSZ1bnRpbD0mbWV0YWRhdGFfcHJlZml4PW9haV9kYw==