bluuebunny/crossref_metadata_2025_split huggingface dataset:
The analysis of RBMK neutron flux profile distortions using operational data history
Authors: Zagrebayev A M , Trifonenkov A V 
Published: December 2020

Abstract
In this research the operational data history was analysed. The data set given represents RBMK-1000 parameters for one month of the reactor operation. Axial flux sensors data and channel power data taken from database were considered. The transients of axial and radial-azimuthal neutron flux distributions were visualized. For the flux distribution data from 72 core channels with axial flux sensors the axial offset transients were obtained. In the operational data considered xenon-induced power oscillations were revealed. Phase and amplitude transients of the oscillations were calculated. The dynamics of phase and amplitude were visualized over the reactor core cross-sections. The values of quantities were displayed with different color shades. The iodine-135 and xenon-135 concentrations recovery problem solution was considered. Formulas for concentrations recovery were derived. The recovery procedure obtained is based on Cauchy problem solution calculation process. The dependence of accuracy of the recovered transients on the amount of the available operational data was studied. The dependency graph of calculation error on the amount of operational data was plot. The recovery of xenon-135 concentration was carried out based on the operational data history of the power unit where the phenomenon of xenon-induced power oscillations was observed.
Neutron-physical characteristics modeling of light water reactors critical corium slurry in water
Authors: Ternovykh M Y , Abdumalikov H M , Ryzhov S N 
Published: December 2020

Abstract
In this paper, the idea of modifying the benchmark by increasing effective multiplication factor value in the considered system without increasing the number of particles under consideration and fuel enrichment is discussed. The technology of data transmission, processing and comparison of the results of neutron-physical calculations using three modern codes developed in different countries and implementing the Monte Carlo method has been worked out. For the corium and water mixture parameters search, the method conventionally called the gradient descent method is used. The search method for areas of critical state corium slurry in water is described. This method can be used to justify nuclear safety in the corium extraction and transportation processes. The methodology is based on the combined use of 1-D and 3-D criticality calculations capabilities of the SCALE 6.2 program pack. Fall of corium particles in water simulation benchmark version is formulated. The possibility of using the algorithm to find the parameters of a corium and water mixture is demonstrated. This benchmark includes the critical state of corium slurry in water and assumes the use of regular structures in the formation of geometric models. The proposed version of the critical benchmark for the corium particles in water state contains 55% fuel.
Experience of long-term U–Al fuel performance in research reactors with low energy generation
Authors: Alferov V P , Geraskin N I , Kozhin A F , Kruglikov A E , Ozherelev S A , Serebryakov A A , Shchurovskaya M V , Yakovlev L I , Aleeva T B , Andreev D V , Gaiduchenko A B 
Published: December 2020

Abstract
Life cycle extension practices for U–Al fuel in Al cladding used at the OR and the IRT MEPhI research reactors are discussed. The IRT MEPhI reactor uses tube type IRT-3M FA with dispersed UO2–Al fuel meat and SAV–1 aluminum alloy as cladding. The OR reactor uses pin type fuel elements with U–Al alloy fuel meat in cladding of Al alloy AMSN. The necessity of life cycle extension for the fuel elements is caused by a large calendar time of fuel assembly operation (10–30 years) and regulatory bodies concerns about this issue. The results of fuel cladding corrosion effects estimation using available empirical correlations for the corrosion rate and the rate of oxide film growth are presented. The results of visual inspection of the fuel elements are shown. Water chemistry monitoring practises in the reactor pool and in spent fuel storage are analysed. The successful experience of the IRT MEPhI (in 2003–2007) on repeated operation of the fuel assemblies previously discharged from the core (after the storage over 5–21 years) is presented.
Test task for transmutation of 241Am in molten salt reactors
Authors: Ashraf O , Tikhomirov G V 
Published: December 2020

Abstract
Molten Salt Reactor (MSR) is one of the six innovative systems, which has been selected by the international Generation IV forum as a candidate for the next-generation reactors. In this paper, the transmutation of241Am, the main contributor to the waste toxicity in the nuclear fuel cycle, in both thermal- and fast-spectrum molten salt reactors is compared. The Am-cylinder is irradiated in the center of the thermal (SD-TMSR) and fast (SMSFR) spectrum MSRs, respectively, and during a long period of time. The total neutron fluxes in SD-TMSR and SMSFR are 4.1×1014 and 1.8×1015 n.cm-2.s-1, respectively. The overall change in actinides and fission products (FPs) mass during the irradiation has been investigated using direct Serpent2 calculations. Under fast spectrum irradiation, transmutation leads to produce short-lived FPs (by fission) rather than the accumulation of heavier actinides. The ratio between the maximum and initial released energy from all isotopes in the Am-cylinder has been calculated in this paper.
Water exchange amount reduction algorithm for power control of VVER-1200 reactor operating in load-following mode
Authors: Djaroum B , Solovyov D A , Semyonov A A , Vigovskiy S B , Schukin N V , Al-Shamayleh A I , Tanash H A , Molev I A 
Published: December 2020

Abstract
In order to solve the problem of water exchange minimization in the primary loop of VVER-1200 reactor operating in load-following mode, we chose the dynamic programming method (DPM). However, this method is very expensive in terms of RAM and computing time. A heuristic is introduced for the DPM to decrease the computational costs. The introduction of the heuristics makes it possible to cut the inefficient branches of the DPM algorithm and to increase its speed without loss of precision. In the proposed algorithm, the use of temperature regulation control reduces the amount of coolant removed from the reactor primary loop, the movement of control rods is limited, which keeps the current axial offset (AO) in the recommended area, makes it possible to equalize the energy release in the reactor core volume. The simulation results obtained confirm the improvement of reactor power control in terms of load following: deviation of core power from the load schedule is reduced, water exchange is decreased by more than 50%, which is useful from the point of view of liquid waste management and treatment.
The problem of large-scale production of plutonium-238 for autonomous energy sources
Authors: Shmelev A N , Geraskin N I , Kulikov G G , Kulikov E G , Apse V A , Glebov V B 
Published: December 2020

Abstract
The article is devoted to the urgent problem of large-scale production of plutonium-238. Various starting isotopes and Pu-238 production schemes are analyzed. The isotope characteristics of the chain based on the starting Np-237 isotope are presented. It has been substantiated that the region of neutron resonant absorption for the isotope Np-237 is the preferred spectrum of its irradiation. It is proposed to form a region with a preferred neutron spectrum in the fast reactor core for irradiation of neptunium-237. The preferred spectrum is achieved by a heterogeneous target structure including neptunium-237 and a moderator with a high atomic weight (Pb, Pb-208, Bi, Pb-Bi eutectic). It is shown that when a moderator with a high atomic weight is used, it becomes possible to form an extensive region with the preferred spectrum. In conclusion, the main conditions for the large-scale production of isotopically pure Pu-238 in a fast reactor are formulated.
The main problems of transmutation of actinides and long-lived fission products from NPP spent fuel
Authors: Tikhomirov G V , Gerasimov A S 
Published: December 2020

Abstract
The paper formulates the main problems associated with research on transmutation, which should be paid attention to by today’s young researchers. The processes of production of hazardous nuclides during transmutation in reactor facilities are considered. The goals of transmutation and the choice of nuclides to be transmuted are discussed. The concept of radiotoxicity is explained as a measure of the radiological hazard of radioactive nuclides, based on the maximum permissible concentration of nuclides according to the IAEA standards. The problem of the formation of secondary radioactive nuclides in nuclear fuel during generation of neutrons for transmutation is discussed. The advantages and disadvantages of various methods of transmutation in nuclear installations are considered: inclusion of transmutable nuclides in nuclear fuel in fast reactors, transmutation in specialized thermal and fast transmutation reactor installations and ADS systems. The problem of accumulation of highly radioactive actinides in a transmutation facility during long-term transmutation and the problem of a potential hazard of the transmutation facility itself are discussed. The unacceptability of application of common-type power reactors for the transmutation of long-lived fission products is demonstrated.
Assessment of a possibility to use 232U in radioisotope thermoelectric generators
Authors: Kulikov G G , Shmelev A N , Apse V A , Kulikov E G 
Published: December 2020

Abstract
The paper analyzes advantages and drawbacks of the radioisotope thermoelectric generators (RTG) based on application of 238Pu and 232U. Currently, the RTG on 238Pu are widely used for long-term autonomous power supply of spaceships due to the following positive properties of 238Pu: high specific heat generation rate, long half-life, weak emission of neutrons and γ-rays. Isotope 238Pu may be produced by neutron irradiation of 237Np (main component of transuranium radioactive wastes) in nuclear reactors. The paper considers application of 232U for the same purpose because of the following positive properties of 232U: power elevation during initial 10 years of the RTG operation (effect of α and β-emitters in 232U decay chain), long half-life (comparable with 238Pu half-life). Unfortunately, 232U decay chain includes also intense emitters of high-energy γ-rays. As a compromise, a possibility is considered to create the RTG based on mixture of 238Pu with small (below 5%) additive of 232U. Such RTG will be able to keep long-term stable operation at high power level. The following two-step technology is proposed to produce 232U: generation of 231Pa in thorium blanket of hybrid thermonuclear reactors and subsequent neutron irradiation of 231Pa in nuclear reactors till accumulation of significant 232U quantities.
Possible ways to solve the problems of using MOX fuel in thermal reactors
Authors: Kosyakin D A , Stogov V Yu , Korobeinikov V V 
Published: December 2020

Abstract
Modern nuclear power has a number of crucial problems, including the accumulation of spent nuclear fuel (SNF) and the projected shortage of nuclear resources. The conversion to a closed fuel cycle is the most promising solution to these problems. Mixed uranium-plutonium fuel (MOX fuel) has been successfully tested and commissioned under this concept by several European countries. However, its operation in reactors designed for uranium fuel has a number of problems, most of which have been resolved. Nevertheless, the spent MOX fuel has neutronic characteristics that do not allow its reuse in thermal reactors and solve the problem of SNF accumulation. This is due to the accumulation of heavy (even) plutonium isotopes, which are not fissionable in the thermal spectrum. In fast reactors, it is possible to create conditions under which the isotopic composition of plutonium from the spent MOX fuel from thermal reactors changes in such a way that it becomes suitable for reuse in thermal reactors. This process is called improving. The concept for the development of two-component nuclear energy based on the joint operation of thermal and fast neutron reactors and the creation of a closed fuel cycle was adopted in Russia to implement this process and solve crucial problems of nuclear power.
Features of training specialists for responsibility centers of the project direction “Proryv”
Authors: Kutsenko K V , Lavrukhin A A , Pershukov V A , Ternovykh M Yu , Tikhomirov G V 
Published: December 2020

Abstract
National Research Nuclear University “MEPhI” (NRNU MEPhI) is the leading university of the consortium of reference universities of the State Atomic Energy Corporation Rosatom (ROSATOM), a regular supplier of young specialists of the highest level. Since 2014 NRNU MEPhI has been training graduates for responsibility centers of the project direction “Proryv”. University responds quickly to staffing challenges, and provides the project “Proryv” the personnel with relevant unique competencies. These competencies include the ability to apply the knowledge gained to develop a new technological platform for nuclear power with the involvement of uranium-238 and spent nuclear fuel reprocessing products in the fuel cycle. Obtaining such unique knowledge is impossible without the use of external resources, namely, the involvement of leading scientists and practitioners of ROSATOM for conducting classes with students, organizing practices in scientific centers and at nuclear fuel cycle enterprises. The article discusses the features of the organization of the educational process in this area.
Experimental comparison of the measurement results of two different silicon photomultipliers and organic scintillator to detect fast neutrons
Authors: Chergui A C , Popova E V , Stifutkin A A , Dendene O , Ilyin A L , Vinogradov S L , Bychkova O V 
Published: December 2020

Abstract
We report on the study of experimental comparison between the results of two types of silicon photomultiplier (SiPM), to discriminate the signals from fast neutrons and gamma-rays by the digital charge pulse shape discrimination (PSD) method. This uses a total to partial charge ratio analysis. The signals are digitized by a high-speed digital oscilloscope LeCroy WR 620zi with a sample rate of 10 GS/s. This oscilloscope was used to acquire data, which were later processed offline in a specially developed software package written in LABVIEW. The tested detector part consisted from a stilbene crystal (3x3x6 mm3 and 6x6x6 mm3) optically coupled to a SiPM (KETEK 3x3 mm2 and SensL 6x6 mm2 respectively) as a photodetector. Measurements with a Cf-252 source were performed and a figure of merit (FOM) for discrimination between neutron and gamma-ray was calculated for each assembly. The resulting value of FOM for the KETEK 3x3 mm2 SiPM and SensL 6x6 mm2 SiPM was 1.6±0.03 and 1.76±0.05, respectively, the SensL assembly having a slightly better discrimination factor than KETEK one. The obtained results prove a good fast neutron detection performance of the SiPMs, which makes it possible to use these types of neutron detectors in the several fields of radiation protection applications and safety of nuclear facilities.
A new approach of spectral regulation in the pressurized water reactors using Zr rods
Authors: Elazaka A I , Tikhomirov G V 
Published: December 2020

Abstract
The elongation of the fuel cycle period and the reactor reactivity control are crucial for all reactor suppliers. The traditional methods of boric acid, burnable absorbers, and control rods to control the reactor reactivity have their defects. A new method of the neutron spectral shift is studied to perform the reactor reactivity control and maintain the length of the fuel cycle. The neutron spectrum regulation can be achieved with different methods, including the Zr rods insertion in the fuel assemblies as water displacers. In this work, SERPENT2 version 2.1.30 was used to investigate a new fuel assembly design in the basics of Sierpinski carpet geometry to utilize the Zr rods as water displacers. The study of different fuel pitches of the Sierpinski fuel assembly is investigated to choose the reference fuel of the new design. The effect of boric acid concentration increasing and defining the maximum boric acid concentration in the Sierpinski fuel assembly are studied. The primary safety parameters of the Doppler effect and moderator temperature reactivity coefficients calculations approved the safety of the new Sierpinski fuel assembly design. The infinite multiplication factor at different steps of burnup showed the efficiency of the Sierpinski fuel assembly model to control the reactor reactivity. The average breeding factor to the burnup limit of 65 GWd/T is 0.68 in the Sierpinski fuel assembly with the presence of Zr rods.
Effect of gadolinium absorber radial profiling in fuel pins on VVER-1000 assembly neutron-physical characteristics
Authors: Vnukov R A , Zhavoronkova I A , Kolesov V V , Karpovich G V , Teplyakova A R 
Published: December 2020

Abstract
The paper raises the problem of non-optimal use of resources. The purpose is to extend the fuel campaign. An alternative task is to increase the duration of the reactor campaign to increase the flexibility of decision-making during the operation of a nuclear reactor. The paper presents calculations of the profiling of a burnable absorber in fuel assemblies for VVER-1000 reactors. A quantitative assessment of the effect of alternative placement of gadolinium in fuel elements is made. Changes in the burnup of fuel elements in fuel assemblies are noted, as well as an advantage expressed in a change in the kinf. The results showed a strong dependence of the isotopic composition of the peripheral layer on the neutron flux density and a weak dependence of the isotopic composition of the central fuel element with gadolinium layers on the neutron flux density. The burnup when using the iterative process turned out to be impossible to align along the radius presumably due to the production of plutonium. A significant advantage of profiling a fuel cell with gadolinium is expressed in the period before the first overload, with each subsequent overload the efficiency of the technology decreases. However, kinf of the profiled assembly throughout the entire fuel campaign exceeds the kinf of the standard assembly, which indicates the efficient use of a burnable absorber.
Analysis of the connection of an idle MCP with three operating without a preliminary reduction in the power of the VVER-1000 reactor
Authors: Muzafarov A , Nikonov S 
Published: December 2020

Abstract
The paper analyzes the transient process caused by connecting an idle loop (turning on an idle main circulation pump (MCP)) without preliminary reduction in the power of a VVER-1000 reactor plant (RP) operating at a rated power corresponding to three MCPs. The process under consideration is included in the “Reactive accidents” section of materials for justifying the safety of VVER-1000. Modeling is carried out on the example of the 3rd power unit of the Kalinin NPP [1]. For the calculations, we used the improved evaluation code ATHLET [2], which is included in the AC2 software package, officially obtained by the National Research Nuclear University MEPhI on the basis of a license agreement with Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS) gGmbH, Germany [3]. The ATHLET code has been certified in Russia for calculating stationary and transient conditions in water cooled reactors [4]. For the conservatism of the calculation, the study did not take into account the work of the PRL, APR, WP-1, WP-2 and AWP, the functioning of which is aimed at limiting the increase in the reactor power during the transient process. Blocking the signal to turn off the MCP on the fact of a decrease in the level in the SG and of the three ECCS systems, the operation of only one system is taken into account. For a number of reasons, this calculation was carried out for the end of the campaign. Implementation of design safety criteria for the adopted transition scenario is demonstrated.
Investigation the possibility of burning Cm a curium fuel reactor
Authors: Dikova T S , Karazhelevskaya Yu E , Terekhova A M 
Published: December 2020

Abstract
Radioactive waste management is one of the main tasks of the nuclear power industry. Spent nuclear fuel poses a threat to the environment if it leaks from the storage facility. A possible solution to reduce waste storage and reduce the economic pressure of natural uranium shortages is to use minor actinides as fuel for nuclear reactors. This article is dedicated to the possibility of burning Cm in a reactor with curium fuel, as a result of which It can be used as fuel for a nuclear reactor. The purpose of the research is to study the possibility of creating a model of a reactor with curium fuel. Calculations were performed for a simplified single-zone model of the RBEC reactor using the «Serpent» software package. Calculations have shown that the use of the actinides as an alternative fuel is possible. That can later play an important role in reducing emissions of nuclear waste in nuclear power.
Irregularity of plutonium isotopic composition of the BREST-OD-300 initial load
Authors: Karazhelevskaya Y E , Levon M A , Terekhova A M , Zlobin A S 
Published: December 2020

Abstract
This article is dedicated observation of irregularity of plutonium isotopic composition of the BREST-OD-300 initial load. The purpose of the research is to study the effect of the plutonium isotopic composition in the nitride fuel of the BREST–OD-300 reactor core starting load on the reactivity reserve. The calculations is creating by writing simplified three-dimensional geometry of the BREST-OD-300 reactor in software package “SERPENT”. There are five options in the fuel assemblies locations are considered with different cases of using the plutonium isotopic composition with deviations for some isotopes. Calculations have shown that the deviation of the plutonium isotope composition from the design values of the fuel loading of the BREST-OD-300 reactor does not affect the stability of reactivity during fuel burnout within the effective fraction of delayed neutrons. In the course of the research, changes in the effective multiplication coefficient are obtained over five years of reactor operation. The value of this indicator does not depend much on the plutonium deviation in the considered variations in the arrangement of fuel assemblies in the reactor core.
Characteristics of uranium mineralization in the Tin-Seririne sedimentary basin (Algeria)
Authors: Badahmaoui T , Bugrieva E P , Lavrukhin A A 
Published: December 2020

Abstract
This article presents the results of studying the uranium mineralization of various parts of the Tin-Seririne sedimentary basin located in the southern part of the Hoggar massif in south-eastern Algeria. Mineralization of uranium is localized at different stratigraphic levels: in altered rocks of the Archean basement and unconformity overlying basal conglomerates (Tahaggart region), and in Devonian sandstones (uranium ore occurrences Tamart-N-Iblis, Timouzaline, etc.). For laboratory studies, 07 samples were taken: As and A1 granites, Tedjart, Timouzaline and Tamart-N-Iblis sandstones, gneisses and conglomerates of the Tahaggart deposit. The samples were examined using gamma spectrometric, chemical and microprobe analyzes. Mineralization is represented by minerals of hexavalent uranium: autunite, torbernite, carnotite, vanuralite and uranophane. The hypergenesis zone of the deposit, it is noted that the geochemical features of the formation of uranium mineralization are typical for regions of arid climate. The arid climate corresponds to oxidative processes of chemical weathering, with weakly alkaline or neutral uranium-bearing solutions. When studying samples taken from different stratigraphic levels, a difference in mineral composition was noted, which indicates some differences in the conditions of formation of uranium mineralization.