bluuebunny/crossref_metadata_2025_split huggingface dataset:
Description of transient processes in multiplying medium by the physical birth-and-death model
Authors: Bobkova M V , Korbut T N , Rudak Ed A 
Published: December 2020

Abstract
The work presents an approach to describe neutron and neutron-breeding medium interaction process based on the physical birth-and-death model. This model is an extended form of the basic birth-and-death model and operates with breeding medium parameters. Analytical solutions and physical interpretation of the processes within reactor can be obtained by such method. Recently, thermal reactor theory with an internal neutron source in the form of delayed neutrons on the basis of birth-and-death model within linear growth approximation was proposed. Particles radioactive decay properties and equation analysis for describing breeding medium parameters using point thermal reactor framework were also studied within the model using Poisson and binomial distribution formalism. Experimental data of three nuclear subcritical facilities were used in this work, KUCA (Japan), MASURCA (France) and GUINEVERE (Belgium). Reactivity and the average number of particles in the system were estimated in this work within physical model framework for three subcritical systems.
Novel Bio-compatible Tracers for Positron Emission Particle Tracking
Authors: Chang Yu-Fen , Balakin Boris V. 
Published: December 2020

Abstract
To study the flow of objects in opaque flow processes, the PEPT (positron emission particle tracking) technique may be utilized. The PEPT technique is based on the same principle of positron emission tomography (PET). Both techniques detect the characteristic back-to-back gamma pairs caused by positron-electron annihilation. PEPT allows tracking of β+ - activated objects (tracers) with high spatial (microns) and temporal resolutions (milliseconds). This makes the technique promising in a number of medical applications that require observation of fast transient phenomena e.g. heart and brain hemodynamics, the aerodynamics of respiratory tract etc. However, a majority of PEPT-tracers used nowadays are tailored for industrial applications and are based on solid particles of a micrometric size, which are hardly suitable for in-vivo tests. In this contribution, we propose new bio-compatible tracers. The tracers are based on frozen radioactive solutions rather than activated solids. We produced a number of millimetre-sized ice tracers dispersing the droplets of radioactive solution in oil at negative temperatures. In this preliminary study, we demonstrate how the tracers behave in a settling column and in an agitated vessel. The PEPT-tracks are presented together with a supplementary numerical analysis that was conducted in order to estimate the freezing time of the droplets.
Mathematical model of the distribution of the coolant flow rate in the group distribution header and its use to determine the flow rate in the RBMK reactor channel
Authors: Zagrebayev A M , Ramazanov R N 
Published: December 2020

Abstract
The paper proposes an elementary mathematical model of the distribution of coolant flows through water communications in the group distributing header (GDH) of a power unit with an RBMK reactor. This model used to determine the flow rate in the channel with a “forbidden” flow meter. An equation is given based on the mass balance of the coolant in the allocated volume. The results of the solution are used to construct an algorithm for determining the flow rate of the coolant in the channel. The results of the application of the technique on the measurement data from the power unit are presented. It is shown that the relative discrepancy with the measured value of the flow rate at a constant position of the flow control valve is on the order of several percent.
Computational analysis of experiments to determine the kinetic parameters at the ASTRA HTGR type critical facility
Authors: Kruglikov A E , Volkov Y N , Zizin M N , Boyarinov V F , Nevinitsa V A , Fomichenko P A 
Published: December 2020

Abstract
The zero power ASTRA critical facility is intended for experimental investigations of neutronic features of modular High Temperature Gas Cooled Reactors. The computational analysis of non-stationary experiments carried out at the ASTRA critical stand to determine the kinetic parameters is carried out using SHIPR diffusion program. To measure the kinetic parameters, the impulse Simmons-King method was used. The computational analysis were carried out for 4 different configurations of ASTRA critical stand. Since in case of nonstationary calculation by implementation of the spatial kinetics approximation, it is necessary to set the neutron group velocities the full-scale calculations of neutron spectra were carry oud using the MCU program in order to take into account the spectrum that is established in the full-scale reactor model. On the basis of a comparison of the experimental and calculated results, a conclusion was made about the applicability of the applied effective calculation models for neutronic calculations of HTGR-type reactors.
Homogenised group constants generation using Monte Carlo codes for diffusion calculation of pebble bed reactor
Authors: Kruglikov A E , Shchurovskaya M V , Volkov Yu N , Nevinitsa V A , Fomichenko P A , Zizin M N 
Published: December 2020

Abstract
This study is focused on the development of homogenization technique for the ASTRA pebble bed critical facility. The methods of generating homogenized few-group cross sections using continuous energy Monte Carlo codes MCU-PTR and Serpent 2.1.29 for full core diffusion calculation of pebble bed reactor were studied. 13-group cross sections were used in the SHIPR diffusion code to calculate the simplified cores of the ASTRA critical facility with HTGR-type fuel and a graphite reflector. Cross-verification of Serpent and MCU-PTR applied to HTGR system was carried out. Different homogenization schemes were tested for the SHIPR model and compared against Monte Carlo solutions. Diffusion coefficients for the fuel calculated using out-scatter approximation and cumulative migration method (CMM) were compared. It was shown that diffusion coefficients for fuel of pebble bed reactor calculated by CMM may reduce the difference between the diffusion and Monte Carlo simulation. The abilities of Serpent and MCU-PTR for modelling pebble bed reactors were compared.
Investigation of the power peaking factor Krin the fuel assembly during the maneuvering mode of the VVER-1000 reactor
Authors: Anisur Rahman SK , Uvakin M A 
Published: December 2020

AbstractThe aim of this thesis paper is to analysis the power peaking factors Krof the VVER-1000 reactor. Power peaking factor is important for the safety operation of the VVER-1000 power reactor. The maintenance of the core power distribution in a nuclear power plant is a precondition for the safety operation of a reactor. In this investigation used boron carbide. Boron carbide is a compound that contains boron and carbon, especially B4C; an extremely hard, non-metallic, black crystalline compound or solid solution. Only diamond is harder. In addition to its hardness property, boron carbide has the physical properties of corrosion resistance, heat strength, low specific gravity and high elastic modulus. It is used as an abrasive, neutron absorber for nuclear reactors, and as an alloying agent in composite structural materials. For this reason, when need to change the power of a reactor then the boron carbide (B4C) control insert the reactor core. In this step investigated the power picking factors and its deviation using the Boron carbide (B4C). Power peaking factors are connecting between the thermal-hydraulics and the nutronics of the nuclear power plant core. This power peaking factors was calculated by the comprehensive nutronics calculation program GETERA-93.
Investigation of using U-233 in thorium base instead of conventional fuel in Russian PWR by SERPENT Code
Authors: Hassan A A , Alassaf S H , Savander V I , Afanasyev V V , Abu Sondos M A 
Published: December 2020

Abstract
In this work, the fuel assembly of type TVS-2006, which is used in Russian reactor VVER-1200, is considered. A proposed fuel of Thorium and Uranium-233 ((Th232+U233)O2 has been investigated to replace the fuel currently used (U238 as a fertile and U235 as fissile). This has been done in order to compare the; conversion ratio, reproduction factor (η), burnup, effective delayed neutrons fraction (βeff), Minor Actinides (MA) inventory, and Temperature coefficients of reactivity. The calculations performed using the Monte Carlo code (SERPENT-2.31) to analyze and compare the two fuels. Results showed that using U-233 in the Thorium base will considerably increase the reactivity at the Beginning Of Cycle (BOC), increase the conversion ratio, and burn up. Besides, the safety parameters for the proposed Th-based fuel, in general, lower than that for Low Enriched Uranium (LEU) but still acceptable.
Possibility assessment for production of non-traditional nuclear fuel in thorium blanket of hybrid thermonuclear reactor
Authors: Kulikov G G , Shmelev A N , Kruglikov A E , Apse V A , Kulikov E G 
Published: December 2020

Abstract
The paper aims at studying specific peculiarities in isotope composition of thorium blanket under irradiation by fusion neutron source (FNS) in hybrid thermonuclear reactor (HTR). High-energy (14 MeV) component of neutron spectrum in thorium HTR blanket results in production of non-traditional fissile mixture including not only233U, but also231Pa232U and 234U. Extraction of such non-traditional fuel from spent Th-blanket and its utilization in traditional nuclear power reactors could increase fuel burn-up and strengthen regime of nuclear non-proliferation. The detailed investigations of these positive effects require high-precision neutron-physical analyses of thorium HTR blanket. The results obtained in these investigations are presented in the paper. The following results were obtained: the chosen model of HTR allowed us to form high-energy neutron spectrum in Th-blanket with significant fraction of 14 MeV neutrons; it was evaluated that threshold (n,2n) and (n,3n)-reactions are able to produce significant amounts of non-traditional target isotopes231Pa and232U; it was shown that accumulation of non-traditional target isotopes weakened substantially in depth of Th-blanket. So, it seems reasonable to seek for optimal thickness of Th-blanket and, thus, to find optimal loading of natural thorium.
Diagnostics of coolant boiling onset based on the analysis of fluctuations of thermohydraulic parameters
Authors: Litvintsova Y E , Laouar S , Delov M I , Kuzmenkov D M , Lavrukhin A A , Kutsenko K V 
Published: December 2020

Abstract
When operating heat-exchange facilities, it is of interest to develop methods of diagnostic and prediction for coolant boiling onset. It is known that the surface temperature of a heating element always has stochastic oscillations around an average value, the analysis of such fluctuations will make it possible to predict the change in the heat transfer regimes.
This paper presents the results of an experimental study of temperature fluctuations for an inertial heater and using of method of diagnostic are based on statistical and frequency analysis of temperature fluctuations. It is shown that there is a correlation between the frequency characteristics of heater temperature fluctuations and oscillations in the thermohydraulic parameters of the coolant (temperature, flow, pressure). The obtained results will find application in the development of an automated system for diagnostics the coolant boiling onset in power facilities.
Simulation of 3D hydraulic effects in the IR-8 reactor’s IRT-3M FA using ATHLET code
Authors: Glyva K I , Nikonov S P 
Published: December 2020

Abstract
This article presents the results of the calculated determination of the 3D hydraulic effects in the gaps of the fuel assembly (FA) IRT-3M. Determination 3d hydraulic effects was carried out using the one-dimensional ATHLET code. Previous work has already been done to analyze the test results hydraulic eight- and six-tube FA IRT-3M reactor IR-8. However, it has not been paid attention to the partition of hydraulic structures in the FAs in the gaps between the fuel rods in the sector in accordance with the division of thermal structures. The new improved FA model was created according to model for neutron physics calculations to create coupling calculation further. The total consumption through the FAs of the improved model changed by less than 0.01% from the base model. The obtained differences in the velocities in the flat and angular sectors will make it possible to obtain a more accurate temperature field distribution for estimation of the margin before the onset of surface boiling. The results of this work can be used in determining the permissible power of the IRT type research reactors: IR-8 at NRC KI (Moscow), IRT-MEPHI at NRNU MEPHI (Moscow), IRT-T at TPU (Tomsk) and WWR-SM at INP (Tashkent).
Calculation of VVER MOX pin-cell benchmark with OpenMC code
Authors: Nikalaichyk A , Bahdanovich R 
Published: December 2020

Abstract
This article presents the results of the ‘Weapons-Grade MOX VVER-1000 Neutronics Benchmarks’ calculated with the OpenMC code. The purpose of this work was to master the OpenMC code and adapt it for VVER calculations in order to use it in the research work conducted at the department. Previously the benchmark was calculated with the SAS2H, HELIOS and MCU codes. Cases calculated with the OpenMC code consist of eight fuel variants with a specific sets of reactor conditions which represent normal and off-normal situations. The obtained infinite neutron multiplication factors are consistent with the presented in the benchmark at the level of 0.02% to 4.71%. The best consistency is observed with the HELIOS code, deviations are less than 1.6%. Despite the rather large deviations for some variants, it can be argued that the OpenMC code allows getting close to other codes results and can be used for similar calculations of VVER-1000.
CORIUMSITY program code for the consequences analysis of a severe core melt accident
Authors: Saldikov I S , Bogdanova E V , Pugachev P A , Ryzhov S N , Smirnov A D , Ternovykh M Y , Tikhomirov G V 
Published: December 2020

Abstract
As part of the tasks to improve the nuclear safety of nuclear power plants, a new program code was developed. The CORIUMSITY program code developed, considered in this work, is intended to analyze the scenario in which an accident at a nuclear power plant is simulated with the melting of the core and the formation of the so-called “corium” - a mixture of nuclear and structural materials of the nuclear reactor core, formed as a result of thermal and mechanical impact during an accident. The CORIUMSITY program code, is intended to analyze several scenarios of different accidents, include an accident with reactor core melting. The functions of this code can help in solving many urgent nuclear safety problems. One of the main methods of operation of the CORIUMSITY code algorithms is the matrix exponential method, which consists in using a matrix function of a square matrix, in which as values are used indicators corresponding to nuclides from the CORIUMSITY code database. The program implements an iterative Euler method for solving the system of levels of nuclear fuel burnup. The CORIUMSITY code was verified with benchmark data to assess the accuracy of the calculation.
SKETCH-N/ATHLET coupled calculations using the boundary conditions plugin
Authors: Romanenko V I , Nikonov S P , Zimin V G , Perin Y , Henry R , Velkov K 
Published: December 2020

Abstract
The paper describes recent developments in the multi-physics coupling scheme between the SKETCH-N nodal neutronics code and the best-estimate thermohydraulic code ATHLET v3.2. The boundary conditions plugin was implemented. The verification and validation were performed using the transient of the Kalinin-3 international Benchmark. The simulation results using the boundary conditions plugin show good agreement with experimental data and calculations performed by using SKETCH-N/ATHLET direct calculations. The calculations of the transient “Switch off of one MCP (Main Coolant Pump) at nominal power” is performed applying a simple core thermohydraulic model without taking into account inter-channel mass transfer.
Estimation of Accuracy of the Method of Gamma- Spectrometry for Non-Destructive Control of Gadolinium Content and Uranium Enrichment in Non-Irradiated VVER-type Fuel Pellets
Authors: Stogov Yu V 
Published: December 2020

AbstractThe gamma-spectra of the single (UO2-Gd2O3)-pellets were measured on the planar detector HPLeGe. The measurements of Gd content in (UO2-Gd2O3)-pellets were performed by X-ray fluorescence method. The total count rate in the doublet 42,3 and 43,0 keV Kα1,2-X-rays from Gd was used as a measure of Gd content. In measurements was used241Am as an external radiation source for exciting the doublet 42,3 and 43,0 keV Kα1,2-X-rays from Gd. In order to obtain the calibration dependence of count rate in doublet 42,3 and 43,0 keV from Gd content in (UO2-Gd2O3)-pellets there were used standard (UO2-Gd2O3)-pellets with the known Gd content, density and enrichment. The estimates of influence of deviations of the density of (UO2-Gd2O3)-pellets and Gd content on total count rate in doublet 42,3 and 43,0 keV were performed. The “pure” gamma-spectra (without employing of external radiation source) were used to obtain the data on uranium enrichment of fuel pellets using multigroup gamma-spectra analysis code (MGAU) and additional information about Gd content. The dependence ofS185,7keVS98,43keVfrom Gd content in (UO2-Gd2O3)-pellets is present (the ratio of count rates for two peaks: Eγ=185,7 keV (the measure of235U) and E=98,43 keV (the measure of235,238U).
Tolerant fuel for VVER reactors
Authors: Alhmoud A Z , Kruglov V B , Kruglov A B , Tanash H A 
Published: December 2020

Abstract
The main factor of destruction of fuel rods in accidents with loss of coolant is associated with the vapor-zirconium reaction occurring between the fuel rod shell and the coolant (water). Improving the reliability of fuel cells can be obtained by modifying or replacing the fuel shell, materials that do not interact with the coolant during normal operation and in emergencies. The loss of coolant accident is a design-basis accident in light water reactors. The postulated accident requires an analysis of the double guillotine break in a main primary coolant pipe, which allows the coolant to freely discharge out of the primary system into the containment of the building. In the present work, the heat stored in the active zone of the reactor, the heat capacity of the fuel and the shell temperatures are determined for ATF materials U-10Mo and U3Si2. The calculated values are compared with the usually used fuel UO2. The results showed beneficial effect of using ATF materials. The above quantities are decreased dramatically for ATF, promising to replace UO2 with ATF fuels. This reduces the probabilities of accidents and oxidation in the nuclear reactors. The calculations are done for nuclear power plant, type VVER-1200.
Membrane vesicles nanotheranostic systems: sources, engineering methods, and challenges
Authors: Nie Weidong , Wu Guanghao , Zhong Haizheng , Xie Hai-Yan 
Published: December 2020

Abstract
Extracellular vesicles (EVs) are cell secretory native components with long-circulation, good biocompatibility, and physiologic barriers cross ability. EVs derived from different donor cells inherit varying characteristics and functions from their original cells and are favorable to serve as vectors for diagnosing and treating various diseases. However, EVs nanotheranostics are still in their infancy because of their limited accumulation at lesion sites and compromised therapy efficiency. Hence, engineering modification of EVs is usually needed to further enhance their stability, biological activity, and lesion-targeting capacity. Herein, we overview the characteristics of EVs from different sources, as well as the latest developments of surface engineering and cargo loading methods. We also focus especially on advances in EVs-based disease theranostics. At the end of the review, we predict the obstacles and prospects of the future clinical application of EVs.
High-resolution forest carbon mapping for climate mitigation baselines over the RGGI region, USA
Authors: Tang Hao , Ma Lei , Lister Andrew , O’Neill-Dunne Jarlath , Lu Jiaming , Lamb Rachel L , Dubayah Ralph , Hurtt George 
Published: December 2020

Abstract
Large-scale airborne lidar data collections can be used to generate high-resolution forest aboveground biomass maps at the state level and beyond as demonstrated in early phases of NASA’s Carbon Monitoring System program. While products like aboveground biomass maps derived from these leaf-off lidar datasets each can meet state- or substate-level measurement requirements individually, combining them over multiple jurisdictions does not guarantee the consistency required in forest carbon planning, trading and reporting schemes. In this study, we refine a multi-state level forest carbon monitoring framework that addresses these spatial inconsistencies caused by variability in data quality and modeling techniques. This work is built upon our long term efforts to link airborne lidar, National Agricultural Imagery Program imagery and USDA Forest Service Forest Inventory and Analysis plot measurements for high-resolution forest aboveground biomass mapping. Compared with machine learning algorithms (r
2 = 0.38, bias = −2.3, RMSE = 45.2 Mg ha−1), the use of a linear model is not only able to maintain a good prediction accuracy of aboveground biomass density (r
2 = 0.32, bias = 4.0, RMSE = 49.4 Mg ha−1) but largely mitigates problems related to variability in data quality. Our latest effort has led to the generation of a consistent 30 m pixel forest aboveground carbon map covering 11 states in the Regional Greenhouse Gas Initiative region of the USA. Such an approach can directly contribute to the formation of a cohesive forest carbon accounting system at national and even international levels, especially via future integrations with NASA’s spaceborne lidar missions.
Relationship of surface soil moisture with solar-induced chlorophyll fluorescence and normalized difference vegetation index in different phenological stages: a case study of Northeast China
Authors: Shen Qiu , Liu Leizhen , Zhao Wenhui , Yang Jianhua , Han Xinyi , Tian Feng , Wu Jianjun 
Published: December 2020

Abstract
Solar-induced chlorophyll fluorescence (SIF) is closely related to vegetation photosynthesis, and is considered as a direct and non-invasive indicator of the functional status of photosynthetic machinery; the normalized difference vegetation index (NDVI) can also reflect photosynthetic characteristics. Both are related to surface soil moisture (SSM), which is susceptible to drought, and phenology plays an important role in these vegetation-moisture relationships. Based on a variety of gridded SIF, NDVI, and SSM datasets obtained from satellite sensors, we presented the detailed relationships of SSM with SIF and NDVI in different phenological stages. Results showed that SIF and NDVI were significantly correlated with SSM in different phenological stages, especially during the maturity and senescence stages. For different vegetation types, SIF was more sensitive than NDVI to SSM anomalies in grasslands and forests during the maturity stage and rainfed croplands during the senescence stage. These relationships were regulated by precipitation and mean temperature, and decreased precipitation had the most significant impact when accompanied by increased temperature during the maturity stage or decreased temperature during the senescence stage. Our findings shed light on the role that phenology has in the relationships of SSM with SIF and NDVI on a large scale, which can further improve the understanding of vegetation-moisture relationships.