Latest Articles from Nuclear Energy and Technology Latest 64 Articles from Nuclear Energy and Technology https://nucet.pensoft.net/ Thu, 28 Mar 2024 12:33:39 +0200 Pensoft FeedCreator https://nucet.pensoft.net/i/logo.jpg Latest Articles from Nuclear Energy and Technology https://nucet.pensoft.net/ Assessment of the possibility for large-scale 238Pu production in a VVER-1000 power reactor https://nucet.pensoft.net/article/117199/ Nuclear Energy and Technology 9(4): 297-301

DOI: 10.3897/nucet.9.117199

Authors: Anatoly N. Shmelev, Nikolay I. Geraskin, Vladimir A. Apse, Gennady G. Kulikov, Evgeny G. Kulikov, Vasily B. Glebov

Abstract: The paper presents the estimates for the possibility for large-scale production of 238Pu in the core of a VVER-1000 power reactor. The Np-fraction of minor actinides extracted from transuranic radioactive waste is proposed to be used as the starting material. The irradiation device with NpO2 fuel elements is installed at the reactor core center. The NpO2 fuel lattice pitch is varied and the irradiation device is surrounded by a heavy moderator layer to create the best possible spectral conditions for large-scale production (~ 3 kg/year) of conditioned plutonium with the required isotopic composition (not less than 85% of 238Pu and not more than 2 ppm of 236Pu). Plutonium with such isotopic composition can be used as the thermal source in thermoelectric radioisotope generators and in cardiac pacemakers. It has been demonstrated that the estimated scale of the 238Pu production in a VVER-type power reactor exceeds considerably the existing scale of its production in research reactors.

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Research Article Thu, 21 Dec 2023 13:40:09 +0200
A computer code for optimizing the neutronics model parameters based on results of reactor physics experiments https://nucet.pensoft.net/article/117198/ Nuclear Energy and Technology 9(4): 289-296

DOI: 10.3897/nucet.9.117198

Authors: Andrey A. Andrianov, Olga N. Andrianova, Yury A. Korovin, Iliya S. Kuptsov, Anastasiya A. Spiridonova

Abstract: The paper describes in brief the functional capabilities of a computer code for optimizing the neutronics model parameters (neutron data, technological parameters, and their covariance matrices) based on results of reactor physics experiments using conditional nonlinear multi-parameter optimization algorithms. The code’s application scope includes adjustment of neutron constants, technological parameters and their covariance matrices based on integral measurement results, formulation of requiremen117198ts with respect to the neutron data uncertainties for achieving the target accuracies in calculation of the reactor functionals, and estimation of the reactor performance prediction accuracy, as well as the informativity and similarity metrics of reactor physics experiments with respect to each other and in relation to the target reactor system. The paper also considers some examples of using the code to refine the neutronics models of nuclear reactor and fuel cycle systems based on results of reactor physics experiments.

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Research Article Tue, 19 Dec 2023 17:57:19 +0200
Enhancing the efficiency of the MOX fuel cycle for VVER-1200 using burnable absorbers https://nucet.pensoft.net/article/98689/ Nuclear Energy and Technology 9(4): 227-232

DOI: 10.3897/nucet.9.98689

Authors: Joy Ozoani, Yuri Volkov

Abstract: Margin adoption in a nuclear power plant (NPP) design is a frequent approach to strengthen the design’s robustness and provide an efficient way to handle uncertainties. However, the current trend of increasing fuel enrichment, including the use of MOX fuel to achieve a higher burnup, leads to non-uniformity in the energy release (power peaking factor) at the level of the fuel rod lattice, thereby causing a great effect on the reactor margins. One of the ways to reduce the power peaking factor is the use of burnable absorbers (BAs) which helps to minimize the power peaking factor. This work aims at enhancing the efficiency of the MOX fuel cycle for VVER-1200 reactor by replacing the Gadolinium burnable absorber to Erbia burnable absorber.

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Research Article Thu, 16 Nov 2023 17:53:12 +0200
Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code https://nucet.pensoft.net/article/91090/ Nuclear Energy and Technology 9(4): 215-225

DOI: 10.3897/nucet.9.91090

Authors: Md Imtiaj Hossain, Abdus Sattar Mollah, Yasmin Akter, Mehraz Zaman Fardin

Abstract: The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for the VVER-1000 MOX core, containing 30% mixed oxide fuel with low enriched uranium fuel, are presented in this study. As per the benchmark report, all six states are considered in the present study. The keff values, assembly average fission reaction rates, and pin-by-pin fission rates were calculated as per benchmark criteria. In addition, 2D thermal and fast neutron-flux distribution were also generated. The reactivity results and neutron flux distribution were compared with other results in which benchmark analysis was performed using the same core geometry and it showed great similarity with slight deviation. This shows that the modeling of the VVER-1000 MOX core was done successfully using OpenMC. Because OpenMC was successfully used for neutronics calculation of the VVER-1000 whole core, it may be mentioned here that OpenMC code can also be utilized for neutronics and other reactor core physics analyses of the VVER-1200 reactor which is to be commissioned in Bangladesh in the upcoming year.

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Research Article Fri, 10 Nov 2023 18:51:32 +0200
Lawson criterion for different scenarios of using D-3He fuel in fusion reactors https://nucet.pensoft.net/article/114267/ Nuclear Energy and Technology 9(4): 207-214

DOI: 10.3897/nucet.9.114267

Authors: Alexandr I. Godes, Vladimir L. Shablov

Abstract: The paper is devoted to refining the Lawson criterion for three scenarios of using D-3He fuel in fusion reactors (fully catalyzed and non-catalysed D-D cycles and a D-3He cycle with 3He self-supply). To this end, a new parameterization of the D + 3He → p + 4He fusion reaction cross-section and astrophysical factor has been developed based on the effective radius approximation (Landau-Smorodinsky-Bethe approximation), which is a model-free theoretical approach to investigating near-threshold nuclear reactions, including resonant reactions. In the framework of this approximation, experimental data from studies in the NACRE II and EXFOR libraries, believed to provide the most reliable results to date, have been described within the accuracy declared in the studies in question in the energy range of 0 to 1000 keV, and the fusion reactivity averaged over the Maxwell distribution has been calculated. The results obtained are in good agreement with the calculations based on the R-matrix theory and the NACRE II fusion reactivity data. For the fully catalyzed D-D cycle and the cycle with 3He self-supply, the Lawson criterion and the triple Lawson criterion have been calculated based on solving the equations of the stationary process kinetics in a fusion reactor for three fuel ions (D, 3He, and T) taking into account the potential for external supply of 3He and p and 4He impurity ions removed from the reaction zone. The parameters of the triple Lawson criterion found are as follows: nτT = 6.42∙1016 cm-3∙s∙keV (T = 54 keV) for the fully catalyzed D-D cycle, nτT = 1.03∙1017 cm-3∙s∙keV (T = 45 keV) for the cycle with 3He self-supply, and nτT = 4.89∙1016 cm-3∙s∙keV (T = 67 keV) for the non-catalyzed D-D cycle with equimolar D-3He fuel.

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Research Article Fri, 10 Nov 2023 18:48:28 +0200
Evaluation of neutronic performance for the VVER-1000 reactor core with regenerated uranium-plutonium fuel https://nucet.pensoft.net/article/112325/ Nuclear Energy and Technology 9(3): 177-182

DOI: 10.3897/nucet.9.112325

Authors: Viktor V. Semishin

Abstract: The possibility has been considered for the VVER-1000 reactor fuel loading to be formed based on regenerated fuel with the use of the spent fuel accumulated in reactors of the same type. A study was undertaken to investigate the change in the isotopic composition of the plutonium discharged from a thermal reactor in the course of its multiple reprocessing and recycle in a thermal neutron reactor. To obtain an equilibrium isotopic composition of the reactor-grade plutonium, 3D neutronic calculations were performed for the stationary fuel cycles of a VVER-1000 serial reactor with conventional oxide fuel and oxide fuel based on regenerated uranium and based on an undivided mixture of uranium and plutonium oxides from SNF. The neutronic performance of reactor cores was compared for the above mentioned fuel types in the course of the fuel company, including the following: in-core radial power density shaping, values of reactivity coefficients for various thermal parameters, reactivity control system efficiency, etc.

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Research Article Fri, 20 Oct 2023 19:00:00 +0300
The effect of the importance function resolution on the accuracy of calculating the functionals of the neutron kinetics in water critical assemblies by Monte Carlo method https://nucet.pensoft.net/article/112165/ Nuclear Energy and Technology 9(3): 171-175

DOI: 10.3897/nucet.9.112165

Authors: Daniil M. Arkhangelsky, Yuliya S. Daichenkova, Mikhail A. Kalugin, Dmitry S. Oleynik, Denis A. Shkarovsky

Abstract: The paper considers a computational study of the importance function effect on the accuracy of calculating the effective fraction of delayed neutrons, βeff, and generation time of instantaneous neutrons using the MCU Monte Carlo code based on the example of three criticality experiments from the ICSBEP handbook. In the MCU code, the importance function has a piecewise constant form: the computational model is broken down into a finite number of registration objects, and the neutron importance is calculated in each. The obtained importance values are used then to calculate the kinetic functionals due to which the calculation accuracy for the latter depends on the resolution. Three types of the importance function spatial partition (axial, radial, combined) have been studied. The numerical simulation results have shown that the axial component of the neutron importance function in all experiments has practically no effect on the calculation accuracy for βeff and Λ: the difference between the obtained values is less than 1%. The radial component has a notable effect (of up to 15.9%) on the Λ calculation accuracy while having almost no effect on the βeff estimate. Using combined partition, as compared with radial partition, improves the calculation accuracy insignificantly (< 1%).

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Research Article Fri, 20 Oct 2023 19:00:00 +0300
Effects of evaluated nuclear data libraries on the calculation results for fuel burnup with minor actinides in a VVER reactor https://nucet.pensoft.net/article/112327/ Nuclear Energy and Technology 9(3): 163-169

DOI: 10.3897/nucet.9.112327

Authors: Gleb W. Karpovich, Yury A. Kazansky, Kirill A. Bakhantsov, Kirill A. Isanov, Nikita O. Kushnir

Abstract: The paper deals with assessing the effects of the ENDF/B-VI.8, ENDF/B-VII.0, JEFF 3.1 and JEFF 3.1.1 nuclear data libraries on the results of calculating a number of functionals for a system based on a VVER reactor with fuel with a large fraction of minor actinides (up to 10%). Key estimates have been obtained for the errors introduced by libraries in calculations of systems with minor actinides (MA) based on a VVER-1200 reactor: – for reactivity, σρ = 0.3 βeff; – for isotopic compositions with minor actinides, ≤ 5% (the error for each particular isotope is different); – for the total mass of accumulated MAs, εm = 0.8%. Conclusions have been made with respect to the need for the further refinement of the library MA data proceeding from the nature of the calculation tasks that dictate the requirements for the accuracy of nuclear constants. It has been shown that systems based on VVER-1000/1200/1300 reactors with MAs need to be calculated using several libraries of evaluated nuclear data created at different organizations and based on the largest possible number of non-recurrent sets of experimental data.

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Research Article Fri, 20 Oct 2023 19:00:00 +0300
Nuclear data uncertainty on generation IV fast reactors criticality calculations analysis comparison https://nucet.pensoft.net/article/111919/ Nuclear Energy and Technology 9(3): 157-162

DOI: 10.3897/nucet.9.111919

Authors: Dimitar G. Chereshkov, Mikhail Yu. Ternovykh, Georgiy V. Tikhomirov, Alexander A. Ryzhkov

Abstract: The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO2, MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-VII.1 library via the SCALE 6.2.4 code system. The main nuclear data uncertainty contributors in the multiplication factor are determined. Recommendations are formulated for improving the cross sections accuracy for several nuclides in order to provide more reliable results of fast reactor criticality calculations. Lead-cooled reactors have no operational history compared to light-water and sodium-cooled reactors. The experimental data insufficiency calls in the question about reliability of the simulation results and requires a comprehensive initial data uncertainty analysis for the neutron transport simulation. The obtained results support the idea that lead- and sodium-cooled reactors have close nuclear data sensitivity using one and the same computation tools, nuclear data libraries and fuel compositions. This makes it possible to use the accumulated data of benchmarks for sodium-cooled reactors in the safety determination of lead-cooled reactors.

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Research Article Fri, 20 Oct 2023 19:00:00 +0300
Application of small perturbation theory for assessing variations of prompt neutron lifetime in a lead-cooled fast reactor https://nucet.pensoft.net/article/107977/ Nuclear Energy and Technology 9(2): 137-142

DOI: 10.3897/nucet.9.107977

Authors: Vladimir A. Apse, Anatoly N. Shmelev, Gennady G. Kulikov, Evgeny G. Kulikov

Abstract: The paper considers the applicability of small perturbation theory to assessing the variations of the prompt neutron lifetime caused by variations in the isotope composition of a lead-cooled fast reactor. The generalized small perturbation theory formulas have been developed to calculate derivatives of the prompt neutron lifetime regarded as a bilinear neutron flux and neutron worth ratio. A numerical algorithm has been proposed for the step-by-step application of the small perturbation theory formulas to assess the prompt neutron lifetime variations caused by a major perturbation in the reactor isotope composition, e.g. by the complete change of the material used earlier as the neutron reflector. The advantage of the proposed approach has been shown which consists in that it is basically possible to determine the role of different neutron reactions, isotopes and energy groups in and their contributions to the total prompt neutron lifetime variation caused by major changes in the reactor isotope composition.

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Research Article Wed, 19 Jul 2023 16:52:32 +0300
Use of remix spent mixed fuel plutonium in the BN-1200 reactor https://nucet.pensoft.net/article/107762/ Nuclear Energy and Technology 9(2): 131-136

DOI: 10.3897/nucet.9.107762

Authors: Nikita V. Kovalev, Alexander M. Prokoshin, Alexander S. Kudinov, Vladimir A. Nevinitsa

Abstract: The VVER-1000 thermal neutron reactor can operate on mixed uranium-plutonium fuel with a content of reactor-grade plutonium up to 5% with a 100% loaded core. In this case, plutonium burns up to 56% of odd isotopes. The energy potential of such plutonium is very low, and its further use in thermal reactors is impractical. However, such plutonium can be used in fast neutron reactors. The paper presents the results of investigating the possibility for such isotopic plutonium composition to be used in the BN-1200 thermal neutron reactor and its value be increased for the plutonium recycle in the reactor. For this purpose, a precision model of the BN-1200 reactor has been developed using the Serpent Monte Carlo code. The model has been verified against the reference values of the nuclear fuel burnup and breeding ratios. The study has shown that such plutonium can be used in the BN-1200 reactor MOX fuel. Maintaining the operating cycle length requires the plutonium fraction in the MOX fuel to be increased up to 2%. In the BN-1200 reactor, the isotopic composition has been found to improve for the further recycle of plutonium in the thermal reactor, i.e. odd plutonium isotopes increase. The fewer odd plutonium isotopes at the beginning of the BN-1200 operating cycle, the greater their increase. It can be seen as the result of the calculation that plutonium from VVER-1000 spent mixed fuel must be loaded into the BN-1200 reactor at least twice to increase the fraction of odd isotopes to the level of reactor-grade plutonium.

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Research Article Tue, 20 Jun 2023 15:31:40 +0300
Isotope kinetics modeling in a circulating fuel system: a case study of the MBIR reactor loop https://nucet.pensoft.net/article/107761/ Nuclear Energy and Technology 9(2): 127-130

DOI: 10.3897/nucet.9.107761

Authors: Dariya S. Kuzenkova, Victor Yu. Blandinskiy

Abstract: The paper presents the results of modeling of changes in the isotopic composition of the liquid-salt fuel circulating in the experimental channel of the MBIR reactor facility. The authors tested the ISTAR software environment adapted for solving burnup equations in problems with variable power levels. The loop channel parameters, including two heat exchanger options, were estimated to obtain the appropriate salt transit time through the loop channel zones. Two problems of a circulating fuel system (loop) modeling are considered, namely: (1) modeling the equilibrium salt isotope composition in such a system; and (2) developing a technique for modeling nonstationary isotope kinetics in the MBIR reactor loop. Non-stationary isotope kinetics can be modeled as sequential burnup of nuclides in the neutron field and decay during movement in the external circuit. The authors also developed an algorithm for modeling changes in the isotopic composition of fuel salt during its circulation, taking into account the sequential transfer of a given salt volume from the burnup zone to the zone outside the reactor core. Based on this algorithm, a software package was created using the Python 3.9 programming language and ISTAR modules. In addition, a description of the calculation methodology was given and some calculation results obtained using the software were presented. In the process of working with the program, it was found that, for the given times of the fuel being in each of the zones (2 and 200 seconds, respectively), modeling the change in the isotopic composition during the fuel campaign (500 days) will require the calculation of more than 500 thousand steps. In order to save time, it is necessary to find out whether it will be possible to reduce the number of calls to the neutronic calculation code due to a slight change in the isotopic composition of the fuel in the loop per one burnup step. Work is currently underway to optimize this process.

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Research Article Tue, 20 Jun 2023 15:31:19 +0300
Influence of the spatial grid type on the result of calculating the neutron fields in the nuclear power plant shielding https://nucet.pensoft.net/article/102507/ Nuclear Energy and Technology 9(2): 99-105

DOI: 10.3897/nucet.9.102507

Authors: Olga V. Nikolaeva, Sergey A. Gaifulin, Leonid P. Bass, Denis V. Dmitriev, Alexandr A. Nikolaev

Abstract: The paper considers the influence of the spatial grid type on the result of solving the equation of neutron transport in the nuclear power plant (NPP) shielding. Neutron fields were calculated in a realistic model of a liquid metal cooled fast neutron tank reactor with an integral equipment layout. Structured cubic and unstructured hexahedral grids (pmsnsys and FRIGATE codes) and unstructured tetrahedral and prismatic grids (RADUGA T code) are used. Limiting values of the group fluxes averaged over the material zones for refined grids have been obtained. It has been shown that the calculation results depend on the type of approximation for the curvilinear inner boundaries between the material zones rather than on the grid cell type (cube, hexahedron, tetrahedron, prism). Using “toothed” approximations for curvilinear boundaries leads to an increase in the area of the boundaries, as well as to the neutron flux refraction condition arising on them. These effects lead to an upward bias in the transport equation solution, and for all energy groups. Conclusion. When solving an equation of neutron transport in the NPP shielding by a grid technique, it is necessary to use grids other than leading to “toothed” approximations of the inner boundaries. Tetrahedral or prismatic grids, or grids of arbitrary hexahedrons can be recommended, as well as composite grids in which cubic cells are located inside the material zone, and hexahedron cells are located near the zone boundary.

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Research Article Tue, 20 Jun 2023 10:43:17 +0300
Evaluation of DNBR with neutronics calculation in LWR systems https://nucet.pensoft.net/article/98452/ Nuclear Energy and Technology 9(2): 77-83

DOI: 10.3897/nucet.9.98452

Authors: Suhail Ahmad Khan, Umasankari Kannan

Abstract: The heat flux in a Light Water Reactor (LWR) system is used to estimate the Departure from Nucleate Boiling Ratio (DNBR) of the system which is an important thermal hydraulic parameter for nuclear reactors from heat removal point of view. The DNBR signifies an operational safety limit i.e. the nuclear power plant has to be operated with sufficient margin from the specified DNBR limit for assuring its safety. The DNBR is evaluated using a thermal hydraulic analysis code using inputs from neutronics calculation. The present paper presents the evaluation approach of minimum DNBR (MDNBR) during standard neutronics calculation. The DNBR calculation is performed using a core physics analysis code and burnup variation of MDNBR is studied for the full cycle length. The results of calculation are presented using the equilibrium core of 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR). The calculations are performed using VISWAM-TRIHEXFA code system. The few group lattice parametric library for IPWR is generated by lattice analysis code VISWAM. The core follow up calculation for the equilibrium core configuration has been performed using core analysis code TRIHEXFA. A first order thermal hydraulic feedback model has been introduced into the 3D finite difference core simulation tool TRIHEXFA. The critical heat flux calculation, required for estimation of DNBR, has been performed using W-3 Tong and OKB-Gidropress correlations implemented in TRIHEXFA.

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Research Article Tue, 20 Jun 2023 10:35:33 +0300
A spatial dynamic model of the SHELF-M reactor facility with fuel and coolant temperature feedbacks https://nucet.pensoft.net/article/102912/ Nuclear Energy and Technology 9(1): 71-76

DOI: 10.3897/nucet.9.102912

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

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

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Research Article Fri, 17 Mar 2023 18:37:40 +0200
Fusion-fission hybrid reactor facility: power profiling https://nucet.pensoft.net/article/102781/ Nuclear Energy and Technology 9(1): 65-70

DOI: 10.3897/nucet.9.102781

Authors: Sergey V. Bedenko, Igor O. Lutsik, Anton A. Matyushin, Sergey D. Polozkov, Vladimir M. Shmakov, Dmitry G. Modestov, Vadim V. Prikhodko, Andrey V. Arzhannikov

Abstract: The current state of research in the field of nuclear and thermonuclear power aimed at creating power generation plants makes it possible to predict the further development of modern power industry in the direction hybrid reactor power plants. Such hybrid systems include a tokamak with reactor technologies, worked out in detail in Russia, and systems with an additional source of neutrons. Power generation plants using tokamaks and accelerators with the required level of proton energy will be of exceptionally large size and power, which will postpone their construction on an industrial scale to the distant future. The ongoing research is aimed at the development of small generation and has the prospect of entering the field of energy use in a shorter period. The hybrid reactor facility under study consists of an axisymmetric assembly of fuel blocks of a high-temperature gas-cooled reactor and a linear plasma source of additional neutrons. The paper demonstrates the results of optimization plasma-physical, thermophysical and gas-dynamic studies, the purpose of which is to level the distortions of the power density field, which are formed in the volume of the multiplicating part of the facility due to the pulsed operation of the plasma source of D-T-neutrons. The studies on increasing the “brightness” of the source and modeling its operating modes were carried out using the DOL and PRIZMA programs. The thermophysical optimization and gas-dynamic calculations were performed using the verified SERPENT and FloEFD software codes. The calculations were made on a high-performance cluster of the Tomsk Polytechnic University.

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Research Article Fri, 17 Mar 2023 18:37:28 +0200
Computational simulation of minor actinide burning in a BN-600 reactor with fuel without uranium and plutonium https://nucet.pensoft.net/article/102776/ Nuclear Energy and Technology 9(1): 59-64

DOI: 10.3897/nucet.9.102776

Authors: Valery V. Korobeynikov, Valery V. Kolesov, Igor A. Ignatiev

Abstract: The paper presents the results of studies on the burning of minor actinides (MA) extracted from SNF of thermal reactors in a BN-600 reactor, which uses the complete set of MAs instead of traditional nuclear fuel types: uranium and/or plutonium. The advantages of such approach to MA burning are that long-lived waste is recycled and energy is produced that can be used, e.g., to generate electricity. Besides, where, e.g., a reactor with uranium or MOX fuel is used for transmutation, apart from burning “foreign” minor actinides, it will additionally generate “its own” MAs. The studies have shown that such reactor can be efficient only if based on fast neutrons, which is due to the specific properties of the minor actinide neutron capture and fission cross-sections as compared with traditional fuel nuclides. The calculation results have shown rather a high rate of MA transmutation and burning in a reactor fueled with minor actinides.

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Research Article Fri, 17 Mar 2023 18:37:17 +0200
Heavy liquid metal cooled fast reactors: peculiarities and development status of the major projects https://nucet.pensoft.net/article/90993/ Nuclear Energy and Technology 9(1): 1-18

DOI: 10.3897/nucet.9.90993

Authors: Alexander I. Orlov, Boris A. Gabaraev

Abstract: Fast reactors with heavy liquid metal coolant (lead or eutectic bismuth-lead alloy) are one of the most advanced technologies capable to address the accumulated world nuclear energy issues. This innovative power technology is being developed in Russia, the USA, China and the European Union. Russia is the leader since it has focused on this topic for a number of decades. First concrete started to be poured in June 2021 to form the foundation of the Russian BREST-OD-300 lead cooled reactor scheduled to be started up in 2026. Attention is also given to the development status of the Chinese CLEAR reactor series. A large scope of R&D has been undertaken, and large-scale nonnuclear experimental facilities are under construction. International Euro-US consortiums for the development of the ALFRED, PLFR and MYRRHA reactors do not expect any unsolvable technical issues either and are currently formulating requirements to the test facilities and candidate materials and technologies required for further activities.

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Review Article Wed, 8 Mar 2023 21:34:44 +0200
On the concept of “effective delayed neutron fraction” https://nucet.pensoft.net/article/96567/ Nuclear Energy and Technology 8(4): 275-279

DOI: 10.3897/nucet.8.96567

Authors: Anatoly G. Yuferov

Abstract: The article considers methodological issues related to the conceptual and terminological apparatus of the dynamics of nuclear reactors. Based on an elementary analysis of the standard point reactor kinetics equations, the author shows that it is necessary to clarify the physical meaning of the parameter β included in the equations, which is traditionally interpreted as the “effective delayed neutrons fraction” (EDNF). It follows directly from the kinetics equations that the parameter β, which appears in these equations as the EDNF, is, from the point of view of the neutron balance, the fraction of prompt neutrons consumed for the generation of delayed neutron precursors (DNPs), and, from the point of view of the DNP balance, the DNP yield per prompt neutron in a single fission event. With these interpretations taken into account, the role of the β parameter is considered in situations related with its adjustment by multiplying it by the “delayed neutron efficiency factor” and with the establishment of the actual fractions of prompt and delayed neutrons. In particular, it is shown that: the statement “if the delayed neutron fraction is β, then the prompt neutron fraction is equal to 1 – β”, used in the problems of analyzing the nuclear reactor dynamics as a starting position, cannot be considered applicable to any reactor conditions; an increase in the β parameter by multiplying it by the “delayed neutron efficiency factor” leads, contrary to traditional interpretations, not to an increase but to a decrease in neutron reproduction in a supercritical reactor. The proposed clarifications are appropriate both in terms of more adequate descriptions of processes in nuclear reactors and in relation to the formulations of nuclear safety requirements.

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Research Article Tue, 13 Dec 2022 15:14:31 +0200
Investigation of algorithms for suppressing xenon oscillations in a VVER-1200 reactor https://nucet.pensoft.net/article/96566/ Nuclear Energy and Technology 8(4): 267-273

DOI: 10.3897/nucet.8.96566

Authors: Denis A. Soloviev, Artsrun G. Khachatryan, Yevgeny V. Chernov, Rashdan T. Al Malkawi

Abstract: This paper presents the results of numerical studies of various algorithms for suppression of xenon offset and power distribution oscillations in the core of a VVER-1200 reactor. The purpose of the research is to select an algorithm that minimizes the amount of liquid radioactive wastes during water exchange in the primary circuit of a nuclear power plant. For this, several algorithms for xenon oscillations suppression were considered. The first algorithm considered was an algorithm for suppression of xenon oscillations, which uses regulation due to AWP only, without utilization of any additional regulation. The second algorithm considered was an algorithm based on the use both AWP and boron regulation. In this algorithm suppression of xenon oscillations was carried out with the help of accelerated initiation of the work of the AWP by changing the boric acid concentration with constant second circuit pressure of the NPP and by utilization of the second control rods group. Last algorithm considered was algorithm based on the use of temperature control for accelerated initiation of the work of the AWP. In this algorithm, xenon oscillations suppression was carried out by changing coolant temperature at the reactor inlet caused by pressure change in the secondary circuit in the normal operation margins, and by involving the second group of control rods. It was shown that the best way to suppress xenon offset and power distribution oscillations in terms of minimization of radioactive liquid wastes amount is the algorithm with accelerated initiation of the AWP due to temperature regulation, with elimination of temperature regulation after minimizing of current axial offset value deviation from the nominal one.

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Research Article Tue, 13 Dec 2022 15:14:19 +0200
Simulating a lead-cooled reactor campaign using the EUCLID/V1 code https://nucet.pensoft.net/article/96565/ Nuclear Energy and Technology 8(4): 261-265

DOI: 10.3897/nucet.8.96565

Authors: Aleksandr A. Belov, Valery P. Bereznev, Galina S. Blokhina, Dmitry P. Veprev, Dmitry A. Koltashev, Vladimir S. Potapov, Olga I. Chertovskikh, Aleksey V. Shershov

Abstract: The paper presents the results of the development of the EUCLID/V1 integrated dynamic code designed to analyze and justify the safety of fast neutron reactor facilities with a liquid-metal coolant, in terms of simulating the reactor campaign. The relevance of this study lies in the need to simulate the behavior of the core at any time during the campaign. It lets us to obtain a full dataset for subsequent simulations of the reactor dynamic conditions (including transient states or accidents). The authors have developed a fuel archive to store calculated data in HDF5 format, created a computational model editor to generate input data in the fuel archive format, and also provided an example of computing the campaign of a lead-cooled fast reactor for three core design models shown in this paper. The main array of fuel assemblies was simulated as a single unit in the first model, as three units in the second model, and in the third every single assembly was unique. In addition, the authors have shown changes in the total masses of actinides in the core, revealed that the different core models have an insignificant effect on the evolution of the total masses of actinides, and given the fuel assembly burnup values for the three core models. For the third model, the largest difference between the minimum and maximum burnup values was obtained with an almost identical average over the fuel assemblies. The reactivity margin over time for the three core models was presented. It was shown that the values and behavior of the reactivity margin during the three micro-campaigns are almost equal. From the fourth to the sixth cycle, the reactivity margin value for the third core model was lower than for the first and the second ones. Finally, the authors conclude that it is desirable to evaluate the behavior of the reactivity margin for lead-cooled fast reactor campaigns based on the detailed model of the core.

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Research Article Tue, 13 Dec 2022 15:14:03 +0200
Physical feasibility of minor actinides transmutation in a two-component nuclear energy system in Russia https://nucet.pensoft.net/article/93664/ Nuclear Energy and Technology 8(4): 225-230

DOI: 10.3897/nucet.8.93664

Authors: Andrey A. Kashirskii, Andrey Yu. Khomiakov, Elena A. Rodina

Abstract: A transition to a two-component nuclear power structure with a reactor fleet consisting of thermal and fast reactors as envisioned in the Russian nuclear power development strategy to 2050 and outlook to 2100 will require optimal spent nuclear fuel and radioactive waste management solutions. A core issue in this regard is managing the long-lived minor actinide (MA) inventory that affects overall nuclear power ecological safety. The study examines several options for homogenous MA (Am and Np) transmutation using modern calculation codes with MA transmutation rate and material balances taken into account. Results demonstrate that if fast reactor installed capacity reaches 92 GWe by 2100 there would not be any need for dedicated MA-burners as the MA issue would be gradually resolved within the two-component nuclear energy system by the end of the century.

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Research Article Tue, 13 Dec 2022 15:12:43 +0200
The concept of a thermionic reactor-converter with evaporative heat transfer https://nucet.pensoft.net/article/93907/ Nuclear Energy and Technology 8(3): 179-185

DOI: 10.3897/nucet.8.93907

Authors: Pavel A. Alekseev, Georgiy E. Lazarenko, Vladimir A. Linnik, Aleksandr P. Pyshko

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

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Research Article Tue, 27 Sep 2022 13:00:16 +0300
The BFS complex – a unique facility to justify the neutronic parameters of the new generation fast reactor cores https://nucet.pensoft.net/article/83655/ Nuclear Energy and Technology 8(2): 97-105

DOI: 10.3897/nucet.8.83655

Authors: Sergey M. Bednyakov, Andrey V. Gulevich, Vladimir G. Dvukhsherstnov, Dmitry A. Klinov, Igor P. Matveenko, Gennadiy M. Mikhailov, Mikhail Y. Semenov

Abstract: The BFS complex comprising two fast critical facilities – BFS-1 and BFS-2 – is a unique experimental base for research into fast reactor physics, reactor safety, core optimization, justification of the closed fuel cycle parameters. The critical facilities have the same pitch of the core lattice, they are loaded with the same materials for core simulations but they differ in size. Over 60 years of the BFS operation, IPPE specialists have gained considerable experience in operating the facilities and carrying out experiments. More than 150 critical assemblies have been studied in BFS.

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Review Article Mon, 27 Jun 2022 18:46:51 +0300
Fusion-fission hybrid reactor facility: neutronic research https://nucet.pensoft.net/article/82294/ Nuclear Energy and Technology 8(1): 25-30

DOI: 10.3897/nucet.8.82294

Authors: Sergey V. Bedenko, Igor O. Lutsik, Anton A. Matyushin, Sergey D. Polozkov, Vladimir M. Shmakov, Dmitry G. Modestov, Vadim V. Prikhodko, Andrey V. Arzhannikov

Abstract: The authors investigate the neutronic characteristics of the operating mode of a hybrid nuclear-thermonuclear reactor. The facility under study consists of a modified core of a high-temperature gas-cooled thorium reactor and an extended plasma neutron source penetrating the near-axial region of the core. The proposed facility has a generated power that is convenient for the regional level (60–100 MW), acceptable geometric dimensions and a low level of radioactive waste. The paper demonstrates optimization neutronic studies, the purpose of which is to level the resulting offsets of the radial energy release field, which are formed within the fuel part of the blanket during long-term operation and due to the pulsed operation of the plasma D-T neutron source. The calculations were performed using both previously developed models and the SERPENT 2.1.31 precision program code based on the Monte Carlo method. In the simulation, we used pointwise evaluated nuclear data converted from the ENDF-B/VII.1 library, as well as additional data for neutron scattering in graphite from ENDF-B/VII.0, based on the S (α, β) formalism.

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Research Article Thu, 17 Mar 2022 10:26:56 +0200
Neutronics and burnup analysis of VVER-1000 LEU and MOX assembly computational benchmark using OpenMC Code https://nucet.pensoft.net/article/78447/ Nuclear Energy and Technology 8(1): 1-11

DOI: 10.3897/nucet.8.78447

Authors: Md. Imtiaj Hossain, Yasmin Akter, Mehraz Zaman Fardin, Abdus Sattar Mollah

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

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Research Article Mon, 14 Mar 2022 10:25:45 +0200
Increasing the production of the Mo-99 isotope by modernizing the design of targets irradiated in the experimental channels of the VVR-c reactor https://nucet.pensoft.net/article/78338/ Nuclear Energy and Technology 7(4): 291-295

DOI: 10.3897/nucet.7.78338

Authors: Denis A. Pakholik, Oleg Yu. Kochnov, Valery V. Kolesov, Vladimir V. Fomichev

Abstract: There are various ways to obtain Mo-99. Some of them are widely used in industrial production, others are in the research stage with the aim of increasing the product yield. The main industrial method for obtaining Mo-99 using a nuclear reactor is the fragmentation method. This method provides for the presence of a uranium target and a nuclear reactor. The target is placed in the channel of the reactor core and irradiated with neutrons for the required time. After that, the target is removed from the channel to the “hot” chamber for the chemical separation of Mo-99. This is how Mo-99 is obtained practically all over the world. The paper considers the fragmentation method for producing Mo-99, which is implemented on the basis of the engineering and technological complex of the VVR-c research nuclear reactor. In order to increase the yield of Mo-99, a modernized model of the “tube-in-tube” target is proposed. The assessment of the production of Mo-99 and the cooling efficiency of the modernized target was carried out. The calculations were performed using the VisualBurnOut and Ansys CFX software packages. Computational studies have shown an increase in the energy release and the amount of the produced Mo-99 isotope in the target of the modernized design. In the most stressed zones, the target wall temperature exceeds the water saturation temperature. Surface boiling occurs in these zones. As a result, turbulization and mixing of the near-wall boundary water layer increases. This improves heat dissipation.

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Research Article Thu, 16 Dec 2021 15:58:57 +0200
On the determination of neutron multiplication by the Rossi-alpha method https://nucet.pensoft.net/article/74156/ Nuclear Energy and Technology 7(3): 253-257

DOI: 10.3897/nucet.7.74156

Authors: Vladimir A. Grabezhnoy, Viktor A. Dulin, Vitaliy V. Dulin, Gennady M. Mikhailov

Abstract: Introduction. This work contains the results of determining the prompt neutron multiplication factor in the subcritical state of a one-core BFS facility, obtained by the neutron coincidence method, for which the influence of the error in the βeff in determining the multiplication factor turned out to be insignificant. The core of the facility consisted of rods filled with pellets of metallic depleted uranium, 37% enriched uranium dioxide and 95% enriched plutonium, sodium, stainless steel and Al2O3. Stainless steel served as a reflector. Methods. In contrast to the inverse kinetics equation solving (IKES) method, which is convenient for determining reactor subcritical states, the neutron coincidence method practically does not depend on the error in the value of the effective fraction of delayed neutrons βeff. If in the IKES method the reactivity value is obtained in fractions of βeff, i.e., from the measurement of delayed neutrons, the neutron coincidence method is based on the direct measurement of the value (1 – kσp)2, where is the effective multiplication factor by prompt neutrons. The total multiplication factor is defined as keff = kσp + βeff. If, for example, keff ≈ 0.9 (which is typical for determining the fuel burnup campaign), then it is the error in determining kσp that is the main one in comparison with the error in βeff. Thus, a 10% error in βeff of 0.003–0.004 (typical for plutonium breeders) will make a contribution to the error 1 – keff equal to 1 – kσp + βeff ≈ 0.00035, i.e., approximately 0.35%, but not 10%, as in the IKES method. Rossi-alpha measurements were carried out using two 3He counters and a time analyzer. The measurement channel width Δt was 1.0 μs. From these measurements, the value of the prompt neutron multiplication factor was obtained. In this case, the space-isotope correlation factor for the medium with a source was calculated using the following values: Φ(x) – solutions of the inhomogeneous equation for the neutron flux and Φ+(x) – solutions of the ajoint inhomogeneous equation. Results. The authors also present a comparison of the results of the Rossi-alpha experiment and measurements of the BFS-73 subcritical facility by the standard IKES method in determining the multiplication factor value. The data of the IKES method differ insignificantly from the results of the Rossi-alpha method over the entire range of changes in the subcriticality with an increase in the subcriticality of the BFS-73 one-core facility. Conclusion. It was impossible to apply the neutron coincidence method to fast reactors; however, the method turned out to be quite workable on their models created at the BFS facility, which was successfully demonstrated in this study.

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Research Article Thu, 23 Sep 2021 15:14:51 +0300
Effect of the burnable absorber arrangement on the VVER-1200 fuel assembly neutronic performance https://nucet.pensoft.net/article/73490/ Nuclear Energy and Technology 7(3): 215-221

DOI: 10.3897/nucet.7.73490

Authors: Ruslan A. Vnukov, Valery V. Kolesov, Irina A. Zhavoronkova, Yaroslav A. Kotov, Md Masum Rana Pramanik

Abstract: Optimizing the use of fuel in a power reactor is a task of current concern. However, little attention has been given to investigating the dependences among the enrichment used, the content of gadolinium oxide in fuel elements, and the life time in combination with assessing the efficiency of using Gd fuel elements with different Gd2O3 contents. The paper considers fuel assembly (FA) versions for VVER-1200 reactors having different enrichments for fuel elements, including those with Gd, and different contents of gadolinium oxide in fuel. A comparative analysis is presented for assemblies with homogeneous Gd2O3 arrangements in each fuel element and with profiled Gd2O3 arrangements. In the latter case, profiling depends on the neutron flux density in the layer which includes Gd fuel elements. This suggests that the arrangement of gadolinium oxide proportionally to the neutron flux density will improve the FA neutronic performance. The results were obtained using SERPENT (a continuous-energy multi-purpose three-dimensional Monte Carlo particle transport code). The assemblies with the used parameters for a 12-month fuel cycle have shown the method under consideration to be inefficient for a period of over 300 eff. days. With increased enrichment and content of gadolinium oxide, the use of profiled versions has turned out to be more rational for longer periods (up to 900 eff. days). Therefore, this phenomenon is relevant for the reactor life, whereas it proves to be insignificant for the fuel life. A complex relationship is noted between the gadolinium and uranium content in an assembly and the effective multiplication factor for the profiled and standard assemblies. This relationship requires further detailed consideration.

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Research Article Thu, 23 Sep 2021 15:12:48 +0300
Simulating operation of power units 1 and 2 at Novovoronezh NPP II with two electrical feed pumps disabled and the backup pump not enabled https://nucet.pensoft.net/article/72394/ Nuclear Energy and Technology 7(3): 187-194

DOI: 10.3897/nucet.7.72394

Authors: Igor N. Gusev, Aleksandr P. Vorobyev, Mikhail N. Kozlovsky, Sergey P. Padun

Abstract: Introduction. The article analyzes the operation of Unit 1 and 2 of Novovoronezh Nuclear Power Plant II (equipped with VVER-1200 reactors) with two electrical feed pumps disabled and the backup pump not enabled. These operating conditions are subsequently simulated using the power unit model software-hardware package (PUM SHC) developed by LLC IF SNIIP ATOM. Research objectives. The objective of this work was to check the reliability of the forecasts of changes in the power unit parameters obtained using the PUM SHC, based on operational data. Methods. The simulated power unit parameter changes in transient conditions were in good agreement with the data collected in real tests. During the simulation, the power unit dynamic stability was preserved, i.e., the operational parameters were within the design limits and did not exceed the protection operation set points. Results. The results of the work suggest the possibility of using current NPP power unit simulations: for developing proposals for adjusting the operation control algorithms in case of malfunctions and emergency modes with the main equipment shutdown and power unit protection actuation; and for verifying design solutions for updating the power unit systems, which are associated with the use of new equipment or changes in flow diagrams. Conclusion. Current power unit models can be applied both for existing power units and for new ones that are being commissioned.

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Research Article Thu, 23 Sep 2021 15:10:15 +0300
Assessment of the neutron flux stability in a high power BN-type reactor in terms of modal spatial kinetics https://nucet.pensoft.net/article/73488/ Nuclear Energy and Technology 7(3): 201-206

DOI: 10.3897/nucet.7.73488

Authors: Ivan V. Tormyshev, Andrey V. Gulevich, Vladimir A. Yeliseev, Victor Yu. Stogov

Abstract: The article discusses the neutron flux stability in the core of a high-power sodium-cooled fast reactor (of the BN-type) without feedbacks. The importance of this problem for high-power BN-type reactors is associated with the specific features of the layout of their cores, including a large diameter and height/diameter ratio about 5. The technique used to substantiate the stability of neutron fields is based on the analysis of the spectrum of the matrix of the system of spatial kinetics equations describing the core of a high-power BN-type reactor without feedbacks. A computational model of the spatial kinetics of a high-power BN-type reactor has been developed in the modal approximation based on the representation of an unsteady flux as a sum of orthogonal functions multiplied by time-dependent amplitudes. The eigenfunctions of the conditionally critical problem are used in the diffusion approximation, which in the discrete case form a complete system. The spectrum of the matrix of the system of ordinary differential equations describing the spatial kinetics of the reactor has been calculated. It is shown that the neutron flux in the core of a high-power BN-type reactor without feedbacks is stable. Test calculations have illustrated the damping of perturbations of the power distribution for a reactor in a critical state.

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Research Article Thu, 23 Sep 2021 15:08:51 +0300
Calculation and experimental analysis of benchmark experiments with a fast neutron spectrum and models of sodium and lead cooled fast reactors using different evaluated nuclear data libraries https://nucet.pensoft.net/article/68951/ Nuclear Energy and Technology 7(2): 103-109

DOI: 10.3897/nucet.7.68951

Authors: Olga N. Andrianova, Yury Ye. Golovko, Gleb B. Lomakov, Yevgeniya S. Teplukhina, Gennady M. Zherdev

Abstract: The paper presents the results of a comparative analysis of criticality calculations using a Monte-Carlo code with the BNAB-93 and BNAB-RF neutron group constants, as well as with evaluated neutron data files from the Russian ROSFOND evaluated nuclear data library and other evaluated nuclear data libraries (ENDF, JEFF, JENDL) from different years. A set of integral experiments on BFS critical assemblies carried out in different years at the Institute of Physics and Power Engineering (60 different critical configurations) was analyzed. The considered integral experiments are included in the database of evaluated experimental neutronic data used to justify the neutronic performance of sodium and lead cooled fast reactors, to verify codes and nuclear data as well as to estimate uncertainties in neutronic parameters due to the nuclear data uncertainties. It has been shown that the ROSFOND evaluated nuclear data library is a library that minimizes the calculation and experimental discrepancies for the considered set of integral experiments. The paper also presents the results of criticality calculations for models of sodium and lead cooled fast reactors based on different evaluated neutron data libraries and provides estimates for the uncertainty in criticality associated with nuclear data.

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Research Article Mon, 21 Jun 2021 10:48:11 +0300
Subcriticality control elements in a reactor system with an extended plasma source of neutrons with regard for temperature https://nucet.pensoft.net/article/68949/ Nuclear Energy and Technology 7(2): 97-101

DOI: 10.3897/nucet.7.68949

Authors: Vladimir V. Knyshev, Aleksandr G. Karengin, Igor V. Shamanin

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

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Research Article Mon, 21 Jun 2021 10:47:45 +0300
Analytical model for determining the leakage albedo component for a direct cylindrical channel passing through the nuclear reactor protective layer https://nucet.pensoft.net/article/68941/ Nuclear Energy and Technology 7(2): 91-95

DOI: 10.3897/nucet.7.68941

Authors: Kirill S. Kupriyanov, Vladimir V. Pereverzentsev

Abstract: The task of determining the radiation situation, including neutron and gamma-quantum flux density, radiation spectrum, specific volumetric activity of radioactive gases in the air, etc. behind the protective composition having inhomogeneities, has always been important in matters of radiation safety. One of the ways to solve the problem of determining gamma radiation fluxes was to divide the total ionizing radiation flux into four components: line-of-sight (LOS), leakage, line-of-sight albedo, and leakage albedo, and obtain an analytical solution for each component. The first three components have been studied in detail in relation to simple geometries and there are analytical solutions for them, but there is no such a solution for the last component. The authors of this work have derived an analytical representation for the leakage albedo component, which, in contrast to numerical methods (such as Monte Carlo methods), makes it possible to analyze the effect of inhomogeneities in protective compositions on the radiation environment as well as to quickly obtain estimated values of fluxes and dose rates. Performing a component-by-component comparison, it becomes possible to single out the most significant mechanisms of the dose load formation behind the nuclear reactor protection, to draw conclusions about the effectiveness of design solutions in the protection design and to improve the protection at significantly lower computational costs. Finally, the authors present calculations for the four components of the total ionizing radiation flux for various parameters of the cylindrical inhomogeneity in the reactor protection. Based on the obtained values, conclusions are made about the importance of taking into account the leakage albedo component in the formation of the radiation situation behind the core vessel.

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Research Article Mon, 21 Jun 2021 10:47:23 +0300
Experimental investigation of the coolant flow in the VVER reactor core with TVSA fuel assemblies https://nucet.pensoft.net/article/65313/ Nuclear Energy and Technology 7(1): 49-54

DOI: 10.3897/nucet.7.65313

Authors: Sergey M. Dmitriyev, Anton V. Gerasimov, Aleksander A. Dobrov, Denis V. Doronkov, Aleksey N. Pronin, Anton V. Ryazanov, Dmitry N. Solntsev, Aleksander Ye. Khrobostov, Aleksey S. Noskov, Oleg B. Samoylov, Yury K. Shvetsov, Dmitry L. Shipov

Abstract: The paper presents the results of an experimental study to investigate the coolant interaction in adjoining fuel assemblies in the VVER reactor core composed of TVSA-T and upgraded TVSA FAs. The processes of the in-core coolant flow were simulated in a test wind tunnel. The experiments were conducted using models representing different portions of the VVER reactor core fuel bundle and consisted in measuring the radial and axial airflow velocities in representative areas within the FAs and in the interassembly space. The results of the experiments can be translated to the full-scale conditions of the coolant flow with the use of the fluid dynamics simulation theory. The measurements were performed using a five-channel pressure-tube probe. The coolant flow pattern in different portions of the fuel bundle is represented by distribution diagrams and distribution maps for the radial and axial velocity vector components in the representative areas of the models. An analysis for the spatial distribution of the radial and axial velocity vector components has made it possible to obtain a detailed pattern of the coolant flow about the FA spacer, mixing and combined spacer grids of different designs. The accumulated database for the coolant flow in FAs of different designs forms the basis for the engineering justification of the VVER reactor core reliability and serviceability. The investigation results for the coolant interaction in adjoining TVSA FAs of different designs have been adopted for the practical use at JSC Afrikantov OKBM to estimate the heat-engineering reliability of the VVER reactor cores and have been included in the database for verification of computational fluid dynamics (CFD) codes and detailed by-channel calculation codes.

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Research Article Tue, 30 Mar 2021 19:38:50 +0300
Rapid preliminary modeling of transport reactor cores https://nucet.pensoft.net/article/65310/ Nuclear Energy and Technology 7(1): 41-47

DOI: 10.3897/nucet.7.65310

Authors: Vladimir I. Korolev

Abstract: At the present time, JSC Baltiskiy zavod has built and transported to the deployment site at Pevek Akademik Lomonosov, a floating nuclear power unit (FNPU), project 20870. There are also three multi-purpose nuclear icebreakers of project 22220 (Arktika, Sibir, Ural) under construction at Baltiskiy being at different readiness stages. A decision has been made to build a nuclear icebreaker, Lider, of even a higher power. Integral reactors developed by JSC OKBM Afrikantov are installed in the nuclear icebreakers using new assembly-type cores which have not been used earlier in floating facilities. A great deal of preliminary calculation is required to give these cores as advantageous characteristics as possible. The paper proposes a procedure for rapid modeling of floating cores with varied operating and design characteristics. This procedure can be used as part of preliminary modeling. The procedure is based on using a combined dimensionless parameter proposed by the author in (Korolev 2009). A chart is presented to model the key performance of cores for floating objects with a nuclear reactor NPPs. Eight assembly-type core options, which can be installed in transport reactors of a modular or integral design, are analyzed.

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Research Article Tue, 30 Mar 2021 19:38:26 +0300
Speeding up the ODETTA code for solving particle transport problems https://nucet.pensoft.net/article/64365/ Nuclear Energy and Technology 7(1): 15-20

DOI: 10.3897/nucet.7.64365

Authors: Anastasiya V. Shoshina, Viktor I. Belousov

Abstract: Mathematical simulation of fast neutron reactors requires high-precision calculations of protection problems based on unstructured meshes. The paper considers and analyzes a parallel version of the ODETTA code (Belousov et al. 2019) with the use of the MPI (Message Passing Interface) library technology (Knyazeva et al. 2006). The code is designed for numerical simulation of neutronic processes in shielding compositions of fast neutron lead cooled reactor plants in normal operating modes, and can be used to calculate the radiation conditions of using structural components and equipment of nuclear power facilities which are assumed to be the sources of and/or exposed to ionizing radiation during their safety justification. The operation of the generated code is compared against the previous version. The MPI-based development of the ODETTA code’s algorithmic part is described. Peculiarities and specific features of the code parallelization are presented, the code modification is given, and respective algorithms are considered. The structure of the ODETTA code based on the MPI is described in brief. The results of using the ODETTA code’s serial and parallel versions in OS Linux (Kostromin 2012) for NRNU MEPhI’s HPC cluster are provided (Savchenko et al. 2020). A comparative analysis is presented for two code implementation options in terms of speed and accuracy of results when using two different clusters and different numbers of nodes for these. Peculiarities of cluster-based calculations are noted.

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Research Article Wed, 24 Mar 2021 19:37:03 +0200
On a significant slowing-down of the kinetics of fast transient processes in a fast reactor https://nucet.pensoft.net/article/60379/ Nuclear Energy and Technology 6(4): 295-298

DOI: 10.3897/nucet.6.60379

Authors: Gennady G. Kulikov, Anatoly N. Shmelev, Vladimir A. Apse, Evgeny G. Kulikov

Abstract: The kinetics of nuclear reactors is determined by the average neutron lifetime. When the inserted reactivity is more than the effective delayed neutron fraction, the reactor kinetics becomes very rapid. It is possible to slow down the fast reactor kinetics by increasing the neutron lifetime. The authors consider the possibility of using the lead isotope, 208Pb, as a neutron reflector with specific properties in a lead-cooled fast reactor. To analyze the emerging effects in a reactor of this type, a point kinetics model was selected, which takes into account neutrons returning from the 208Pb reflector to the reactor core. Such specific properties of 208Pb as the high atomic weight and weak neutron absorption allow neutrons from the reactor core to penetrate deeply into the 208Pb reflector, slow down in it, and have a noticeable probability to return to the reactor core and affect the chain fission reaction. The neutrons coming back from the 208Pb reflector have a long ‘dead-time’, i.e., the sum of times when neutrons leave the reactor core, entering the 208Pb reflector, and then diffuse back into the reactor core. During the ‘dead-time’, these neutrons cannot affect the chain fission reaction. In terms of the delay time, the neutrons returning from the deep layers of the 208Pb reflector are close to the delayed neutrons. Moreover, the number of the neutrons coming back from the 208Pb reflector considerably exceeds the number of the delayed neutrons. As a result, the neutron lifetime formed by the prompt neutron lifetime and the ‘dead-time’ of the neutrons from the 208Pb reflector can be substantially increased. This will lead to a longer reactor acceleration period, which will mitigate the effects of prompt supercriticality. Thus, the use of 208Pb as a neutron reflector can significantly improve the fast reactor nuclear safety.

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Research Article Fri, 20 Nov 2020 10:56:05 +0200
Investigation of the impact of steady-state VVER-1000 (1200) core characteristics on the reactor stability with respect to xenon oscillations https://nucet.pensoft.net/article/60464/ Nuclear Energy and Technology 6(4): 289-294

DOI: 10.3897/nucet.6.60464

Authors: Rashdan Talal Al Malkawi, Sergey B. Vygovsky, Osama Wasef Batayneh

Abstract: The article presents a method for obtaining an analytical expression for the criterion of stability of a VVER-1000 (1200) reactor with respect to xenon oscillations of the local power in the core, containing an explicit dependence of the criterion ratio coefficients on the arbitrary axial neutron field distribution in steady states of the core. Based on the data of numerical experiments using a full-scale model of the Kalinin NPP power units, the authors present the results of checking the validity of this expression for the reactor stability criterion with respect to xenon oscillations for different NPPs with VVER-1000 (1200) reactors.

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Research Article Fri, 20 Nov 2020 10:55:26 +0200
Use of erbium as a burnable absorber for the VVER reactor core life extension https://nucet.pensoft.net/article/60563/ Nuclear Energy and Technology 6(4): 275-279

DOI: 10.3897/nucet.6.60563

Authors: Saleh H. Alassaf, Vladimir I. Savander, Ahmed A. Hassan

Abstract: The paper presents the results of a computational and theoretical analysis concerned with the use of erbium as a burnable absorber in VVER-type reactors. Partial refueling options for the reactor life extension to 18 and 24 months is considered, the refueling ratio being equal to three for the 18-month life and to two for the 24-month life. Erbium is expected to be present in all fuel elements in the FA with the same weight content. The influence of the erbium weight content on such neutronic characteristics of the reactor and fuel as burn-up, reactivity coefficients, residual volume of “liquid" control, and amounts of the liquid radioactive waste (LRW) formed was assessed. The calculations were performed using a simplified model of refueling without FA reshuffling. An infinite array of polycells consisting of FAs with different in-core times was considered. The escape of neutrons from the core was taken into account by selecting the critical value K∞ at the end of life. Erbium does not burn up in full for the lifetime which affects the fuel burn-up as compared with the liquid excessive reactivity compensation system. The reduction is 0.7% per 0.1% of the erbium weight load in the fuel elements. This, however, also reduces the maximum content of the boron absorber in the coolant and the LRW accumulation in the ratio of 5% per 0.1% of the erbium weight load. Erbium influences the spectral component of the coolant temperature reactivity coefficient which turns out to be negative even with its minor weight fraction in fuel elements, and a reduction in the boron absorber fraction leads to a positive value of the density reactivity coefficient. As a result, the overall coolant temperature reactivity coefficient has a negative value throughout the lifetime.

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Research Article Fri, 20 Nov 2020 10:54:20 +0200
Codes of new generation for safety justification of power units with a closed nuclear fuel cycle developed for the “PRORYV” project https://nucet.pensoft.net/article/54710/ Nuclear Energy and Technology 6(3): 203-214

DOI: 10.3897/nucet.6.54710

Authors: Leonid A. Bolshov, Valery F. Strizhov, Nastasya A. Mosunova

Abstract: The article describes the status of development of codes of new generation for the “PRORYV” Project by the end of 2019: twenty-five commercial-grade software products to justify design solutions and safety of power units with fast neutron reactors and liquid metal coolant (sodium and lead) in a closed nuclear fuel cycle. The developed system of codes is multi-physical and multi-scale that allows performing both calculations of the whole installations and high precision calculations of their individual elements. The developed codes offer unique features. Twelve developed codes have already been certified by Rostechnadzor, and six more have been submitted for certification. In addition to creating the software products, a large-scale work is being carried out to conduct experimental studies for code validation that meet modern requirements imposed by the codes: unique measurement techniques have been created; experimental data on flow characteristics of heavy liquid metal coolant (HLMC) in a fuel assembly simulator have been obtained, as well as of “gas-HLMC” interphase interaction after inert gas injection in HLMC and characteristics of heat exchange between the inert gas and HLMC. The results are already used for validation of system and CFD codes used in the “PRORYV” Project.

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Research Article Fri, 6 Nov 2020 09:51:36 +0200
Power density dynamics in a nuclear reactor with an extended in-core pulse-periodic neutron source based on a magnetic trap https://nucet.pensoft.net/article/57976/ Nuclear Energy and Technology 6(3): 175-179

DOI: 10.3897/nucet.6.57976

Authors: Igor V. Shamanin, Sergey V. Bedenko, Vladimir M. Shmakov, Dmitry G. Modestov, Igor O. Lutsik

Abstract: The article examines the features of the spatial kinetics of an innovative hybrid nuclear power facility with an extended neutron source based on a magnetic trap. The fusion-fission facility under study includes a reactor plant, the core of which consists of an assembly of thorium-plutonium fuel blocks of the HGTRU reactor of a unified design and a long magnetic trap that penetrates the near-axial region of the core. The engineering solution for the neutron plasma generator is based on an operating gas-dynamic trap based on a fusion neutron source (GDT-FNS) developed at the Novosibirsk G.I. Budker Nuclear Physics Institute of the Siberian Branch of the Russian Academy of Sciences. The GDT-FNS high-temperature plasma pinch is formed in pulse-periodic mode in the investigated hybrid facility configuration, and, at a certain pulse rate, one should expect the formation of a fission wave that diverges from the axial part of the system and propagates throughout the fuel block assembly in a time correlation with the fast D-D neutron pulse source. In these conditions, it is essential to study the fission wave propagation process and, accordingly, the power density distribution formation within the facility blanket. The paper presents the results of a study on the steady-state and space-time performances of neutron fluxes and the power density dynamics in the facility under investigation. The steady-state neutronic performance and the space-time fission wave propagation were simulated using the PRIZMA software package developed at FSUE RFNC-VNIITF.

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Research Article Tue, 29 Sep 2020 17:41:29 +0300
A study into the propagation of the uncertainties in nuclear data to the nuclear concentrations of nuclides in burn-up calculations https://nucet.pensoft.net/article/57802/ Nuclear Energy and Technology 6(3): 161-166

DOI: 10.3897/nucet.6.57802

Authors: Alexander N. Pisarev, Valerii V. Kolesov

Abstract: The key papers on estimating the uncertainties in nuclear data deal with the influence of these uncertainties on the effective multiplication factor by introducing the so-called sensitivity factors and only some of these are concerned with the influence of such uncertainties on the life calculation results. On the other hand, the uncertainties in reaction rates, the neutron flux, and other quantities may lead to major distortions in findings, this making it important to be able to determine the influence of uncertainties on the nuclear concentrations of nuclides in their burn-up process. The possibility for the neutron flux and reaction rate uncertainties to propagate to the nuclear concentrations of nuclides obtained as part of burn-up calculations are considered using an example of a MOX-fuel PWR reactor cell. To this end, three burn-up calculation cycles were performed, and the propagation of uncertainties was analyzed. The advantages of the uncertainty estimation method implemented in the VisualBurnOut code consists in that all root-mean-square deviations are obtained as part of one calculation as the statistical method, e.g. GRS (Generation Random Sampled), requires multiple calculations. The VisualBurnOut calculation results for the root-mean-square deviations in nuclear concentrations were verified using a simple model problem. It is shown that there is a complex dependence of the propagation of the root-mean-square deviations in the nuclear concentrations of nuclides in the process of fuel burn-up, and, therefore, further studies need to aim at investigating the influence of uncertainties in nuclear data on the nuclear concentrations of nuclides.

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Research Article Wed, 16 Sep 2020 12:00:02 +0300
Justification of VVER-1000 safety when using fuel compositions doped by protactinium and neptunium https://nucet.pensoft.net/article/55218/ Nuclear Energy and Technology 6(2): 99-104

DOI: 10.3897/nucet.6.55218

Authors: Tuul Baatar, Evgeny G. Kulikov

Abstract: Increasing fuel burnup is one of the important areas of nuclear power development. Currently, the most common type of light-water reactors is characterized by burnup ratios of about 5%, i.e., only a small fraction of fuel is used to generate electricity. The paper considers the possibility of a significant increase in fuel burnup due by introducing protactinium and neptunium into the fuel composition. The chains of nuclide transformations starting with protactinium and neptunium are characterized by a gradual improvement in the multiplying properties, which ensures increased fuel burnup. At the same time, a situation may be observed when the multiplying properties of a fuel composition are improved during the campaign, which indicates that at a certain point in time the accumulation rate of fissile nuclides from protactinium and neptunium exceeds the accumulation rate of fission products. While protactinium is hardly accessible in sufficient quantities, neptunium is contained in spent nuclear fuel, a significant amount of which is stored in on-site facilities. Therefore, from a practical perspective, the introduction of neptunium into fuel compositions seems to be more preferable. The novelty of the work is the analysis of the effects of protactinium and neptunium on the reactivity coefficients during fuel campaigns. The calculations were carried out for a VVER-1000 type reactor using the SCALE-6.2 software package.

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Research Article Thu, 25 Jun 2020 15:42:31 +0300
Evaluation of the neutronic performance of a fast traveling wave reactor in the Th-U fuel cycle https://nucet.pensoft.net/article/54629/ Nuclear Energy and Technology 6(2): 77-82

DOI: 10.3897/nucet.6.54629

Authors: Alina Ye. Pomysukhina, Yury P. Sukharev, German N. Vlasichev

Abstract: The possibility for all of the uranium or thorium fuel to be used nearly in full is expected in traveling wave reactors. A traveling wave reactor core with a fast neutron spectrum in a thorium-uranium cycle has been numerically simulated. The reactor core is shaped as a rectangular prism with a seed region arranged at one of its ends for the neutron fission wave formation. High-enriched uranium metal is used as the seed region fuel. Calculated power density dependences and concentrations of the nuclides involved with the transformation chain along the core at a number of time points have been obtained. The results were graphically processed for the clear demonstration of the neutron fission wave occurrence and transmission in the reactor. The obtained power density dependence represents a soliton (solitary wave) featuring a distinct time repeatability. Neutron spectra and fission densities are shown at the initial time point, when no wave has yet formed, and at the time of its formation. The wave rate has been calculated based on which the reactor life was estimated. The fuel burn-up has been estimated the ultra-high value of which makes the proposed reactor concept hard to implement. The burn-up of most of both the raw material and the fissile material it produces indicates a high potential efficiency of the developed reactor concept in terms of fuel utilization and nuclear nonproliferation.

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Research Article Fri, 19 Jun 2020 18:20:33 +0300
Consideration of heterogeneous effects in preparing neutron multigroup constants in the CONSYST/BNAB-RF system https://nucet.pensoft.net/article/52112/ Nuclear Energy and Technology 6(1): 63-70

DOI: 10.3897/nucet.6.52112

Authors: Olga N. Andrianova, Gennady M. Zherdev, Gleb B. Lomakov, Yevgeniya S. Teplukhina

Abstract: The need for building mutually self-agreed computational models for high-precision and engineering neutronic codes is defined by requirements to certification and verification of software products and nuclear data in accordance with the Regulations for Verification and Expert Examination of Neutronic Calculation Software Tools (RB-061-11). The key requirement in RB-061-11 is that there shall be a methodically transparent and reproducible procedure to estimate the methodological and nuclear data component of the neutronic parameter uncertainties to be implemented only if there are mutually self-agreed computational neutronic models. Using an example of a series of measurements carried out on three critical BFS-61 configurations, factors are discussed which need to be taken into account when building such types of neutronic models and the peculiarities of their application for calculating the neutronic parameters of BFS-61 assemblies. Improved functional capabilities of updated software tools and nuclear data for the computational and experimental analysis of integral BFS experiments (ROSFOND/BNAB-RF, CONSYST and MMK-RF) have been demonstrated, allowing a much shorter time and the smallest risks of errors in preparing mutually self-agreed computational models for various neutronic codes, as well as correct estimation of the methodological and nuclear data components of the uncertainties in neutronic parameters in accordance with RB-061-11. The results of estimating the uncertainty in neutronic parameters with respect to the group approximation approach, are presented. It has been shown based on an analysis of the obtained results that the discrepancies in the calculations of the BFS-61 configurations in the transition from the ROSFOND evaluated neutron data library to its group version, BNAB-RF, does not exceed 0.3% in criticality (heterogeneous effects uncertainty of 0.2 to 0.8 %). The estimated spectral index data biases lie in the limits of the Monte Carlo statistical error. Based on the results of a computational and experimental analysis for the entire set of measurements performed on a series of BFS-61 assemblies, the ROSFOND library is the optimal nuclear data library.

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Research Article Fri, 27 Mar 2020 18:18:45 +0200
Reactor with metallic fuel and lead-208 coolant https://nucet.pensoft.net/article/50868/ Nuclear Energy and Technology 6(1): 23-27

DOI: 10.3897/nucet.6.50868

Authors: Georgiy L. Khorasanov, Anatoliy I. Blokhin

Abstract: 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.

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Research Article Wed, 11 Mar 2020 14:06:31 +0200
Safety features of fast reactor with heavy atomic weight weakly neutron absorbing reflector https://nucet.pensoft.net/article/50867/ Nuclear Energy and Technology 6(1): 15-21

DOI: 10.3897/nucet.6.50867

Authors: Gennady G. Kulikov, Anatoly N. Shmelev, Vladimir A. Apse, Evgeny G. Kulikov

Abstract: 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.

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Research Article Wed, 11 Mar 2020 14:06:08 +0200
Minor actinides transmutation in equilibrium cores of next generation FRs https://nucet.pensoft.net/article/46517/ Nuclear Energy and Technology 5(4): 353-359

DOI: 10.3897/nucet.5.46517

Authors: Alexander V. Egorov, Yurii S. Khomyakov, Valerii I. Rachkov, Elena A. Rodina, Igor R. Suslov

Abstract: 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.

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Research Article Tue, 10 Dec 2019 16:27:05 +0200
Heterogeneous effects in simulating a fast nuclear reactor on the BFS test facility https://nucet.pensoft.net/article/48426/ Nuclear Energy and Technology 5(4): 345-351

DOI: 10.3897/nucet.5.48426

Authors: Yury A. Kazansky, Gleb V. Karpovich

Abstract: 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).

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Research Article Tue, 10 Dec 2019 14:35:21 +0200
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 https://nucet.pensoft.net/article/48391/ Nuclear Energy and Technology 5(4): 289-295

DOI: 10.3897/nucet.5.48391

Authors: Olga I. Bulakh, Oleg K. Kostylev, Vladimir N. Nesterov, Eldar K. Cherdizov

Abstract: 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.

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Research Article Tue, 10 Dec 2019 14:31:15 +0200
Some aspects of the VVER-440 reactor plant life re-extension: a case study of the Novovoronezh NPP Unit 4 https://nucet.pensoft.net/article/46380/ Nuclear Energy and Technology 5(3): 249-256

DOI: 10.3897/nucet.5.46380

Authors: Vladimir P. Povarov, Anatoly I. Fedorov, Sergey L. Vitkovsky

Abstract: 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.

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Research Article Wed, 25 Sep 2019 12:25:29 +0300
Application of soft computing techniques in WWER nuclear power plant safety https://nucet.pensoft.net/article/34590/ Nuclear Energy and Technology 5(2): 177-182

DOI: 10.3897/nucet.5.34590

Authors: Akbar Abbasi, Fahreddin Sadikoglu

Abstract: 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.

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Research Article Fri, 21 Jun 2019 15:00:07 +0300
Application of spiking neural networks for modelling the process of high-temperature hydrogen production in systems with gas-cooled reactors https://nucet.pensoft.net/article/36474/ Nuclear Energy and Technology 5(2): 129-137

DOI: 10.3897/nucet.5.36474

Authors: Sergey O. Starkov, Yury N. Lavrenkov

Abstract: 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.

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Research Article Fri, 21 Jun 2019 15:00:01 +0300
Optimization of two opposing neutron beams parameters in dynamical (B/Gd) neutron cancer therapy https://nucet.pensoft.net/article/32239/ Nuclear Energy and Technology 5(1): 1-7

DOI: 10.3897/nucet.5.32239

Authors: Nassar H. S. Haidar

Abstract: 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*.

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Research Article Wed, 20 Mar 2019 16:41:31 +0200
ROCOCO system of combined neutron constants – current status and results of testing using geometrical module of the MMK code https://nucet.pensoft.net/article/31861/ Nuclear Energy and Technology 4(3): 217-222

DOI: 10.3897/nucet.4.31861

Authors: Gennady M. Zherdev, Tamara S. Kislitsyna, Mark N. Nikolayev

Abstract: 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.

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Research Article Fri, 7 Dec 2018 10:44:00 +0200
Application of diffusion approximation in the calculations of reactor with cavities https://nucet.pensoft.net/article/31863/ Nuclear Energy and Technology 4(3): 203-209

DOI: 10.3897/nucet.4.31863

Authors: Evgeny F. Seleznev, Valery P. Bereznev

Abstract: 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.

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Research Article Fri, 7 Dec 2018 10:42:00 +0200
Reactivity margin evaluation software for WWR-c reactor https://nucet.pensoft.net/article/31862/ Nuclear Energy and Technology 4(3): 197-201

DOI: 10.3897/nucet.4.31862

Authors: Ivan P. Belyavtsev, Sergey O. Starkov

Abstract: 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.

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Research Article Fri, 7 Dec 2018 10:41:00 +0200
Power coefficient of reactivity: definition, interconnection with other coefficients of reactivity, evaluation of results of transients in power nuclear reactors https://nucet.pensoft.net/article/30663/ Nuclear Energy and Technology 4(2): 111-118

DOI: 10.3897/nucet.4.30663

Authors: Yury Kazansky, Yanis Slekenichs

Abstract: 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.

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Research Article Mon, 26 Nov 2018 16:05:00 +0200
ROCOCO: A constants supply system for Monte-Carlo reactor calculation https://nucet.pensoft.net/article/30662/ Nuclear Energy and Technology 4(2): 103-109

DOI: 10.3897/nucet.4.30662

Authors: Gennady Zherdev, Tamara Kislitsyna, Mark Nikolayev

Abstract: 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.

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Research Article Mon, 26 Nov 2018 16:04:00 +0200
Lead reactor of small power with metallic fuel https://nucet.pensoft.net/article/30527/ Nuclear Energy and Technology 4(2): 99-102

DOI: 10.3897/nucet.4.30527

Authors: Georgiy Khorasanov, Dmitriy Samokhin, Aleksandr Zevyakin, Yevgeniy Zemskov, Anatoliy Blokhin

Abstract: 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.

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Research Article Mon, 26 Nov 2018 16:03:00 +0200
Solution of neutron-transport multigroup equations system in subcritical systems https://nucet.pensoft.net/article/29837/ Nuclear Energy and Technology 4(1): 79-85

DOI: 10.3897/nucet.4.29837

Authors: Igor V. Shamanin, Sergey V. Bedenko, Vladimir N. Nesterov, Igor O. Lutsik, Anatoly A. Prets

Abstract: 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.

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Research Article Thu, 18 Oct 2018 10:09:13 +0300
Converting ENDF libraries into relational format https://nucet.pensoft.net/article/29858/ Nuclear Energy and Technology 4(1): 57-63

DOI: 10.3897/nucet.4.29858

Authors: Anatoliy Yuferov

Abstract: 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.

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Research Article Wed, 17 Oct 2018 11:30:13 +0300
Minimize fission power peaking factor in radial direction of water-cooled and water-moderated thermionic conversion reactor core https://nucet.pensoft.net/article/29453/ Nuclear Energy and Technology 4(1): 7-11

DOI: 10.3897/nucet.4.29453

Authors: Pavel A. Alekseev, Aleksei D. Krotov, Mikhail K. Ovcharenko, Vladimir A. Linnik

Abstract: 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.

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Research Article Tue, 25 Sep 2018 09:55:03 +0300