bluuebunny/crossref_metadata_2025 huggingface dataset:
Analisis Kuantitatif Informed Consent pada Tindakan Sectio Caesarea di Rumah Sakit Patria IKKT Jakarta Barat
Authors: Nurmayantih  , Nanda Aula Rumana  , Daniel Happy Putra  , Puteri Fannya
Published: January 2022

Informed Consent is the consent given by the patient or his family on the basis of an explanation of the medical/surgical action to be performed on the patient and this informed consent must be complete. In performing sectio caesarea, the informed consent sheet is not filled in, so any action taken can be categorized as malpractice. Researchers found that there were still many incomplete informed consent forms, especially informed consent for sectio caesarea surgery. The purpose of the study was to determine the quantitative analysis of informed consent for sectio caesarea at the Patria IKKT Hospital for the period March - April 2021 based on 4 components, namely knowing the completeness of patient identification, author authentication, and completeness of important reports, good records. This type of research is quantitative with descriptive design and data collection techniques are observation, checklist. This research was conducted using systematic random sampling method. The results of the study of 90 informed consent sheets for sectio caesarea, the average completeness of the patient identification filling component was 100%, the important component of filling out the report was an average of 86%, the author's authentication component had an average of 97.9%, the component of filling out good notes had an average of 97.9%. the average completeness is 93.7%. The results of the recapitulation of quantitative analysis have an average completeness of 94.4%.
Tinjauan Kualitas Pelayanan Pasien JKN Non PBI di Instalasi Rawat Jalan Puskesmas Batua Kota Makassar
Authors: Leni Karunia Septiani  , Nanda Aula Rumana  , Deasy Rosmala Dewi  , Noor Yulia
Published: January 2022

Service quality comes from the comparison of what consumers feel (expectations) with consumer perceptions about the performance of companies that provide services. The purpose of this study was to describe the quality of service for JKN non-PBI patients at the Batua Public Health Center in Makassar. The method used is quantitative analysis with descriptive method. Calculation of sample size using the proportion estimation formula obtained 96 samples with the sampling method, namely Incidental/Convenience Sampling. Data collection techniques using interviews and self-filling by patients with interview guide instruments and questionnaires, and data were analyzed using SPSS. Service quality is assessed based on 5 dimensions, namely Tangible, Responsiveness, Reliability Assurance and Empathy. From the results of the study, it was found that the percentage of service quality at Batua Health Center was 57.3% of good service quality and 42.7% not good, meaning that the service quality was classified as not good because ≤ 90% based on outpatient SPM. Obstacles that often occur when serving patients such as problems with communication tools/speakers, lack of computers, patients do not want to queue and do not bring medical cards/BPJS cards/identity cards. It is recommended to the Puskesmas to pay more attention to the cleanliness and comfort of the registration place, the officers should improve their competence, advance working hours so that there are no delays in opening the counter, apply 3S 1T, pay attention to the problematic speakers and computers as well as socialize the files that must be brought during registration.
Coleoptera fauna in forest ecosystem of Shivalik hills region of India - A mini Review
Authors:   Zoological Survey of India, Northern Regional Centre, 218, Kaulagarh Road, Dehradun-248195, Uttarakhand, India, Paunikar Sanjay
Published: January 2022

A review is given on documentation of important and major insect order coleoptera in forest ecosystem of Shivalik Hill regions of India. Shivalik Hills are situated in north Indian region to foot hills of North-West Himalaya due to position make itsunique bio geographic location, diversified climatic conditions and enormous eco-diversity and geo-diversity of flora and fauna. Several renowned workers contributed and documented number of species of coleoptera of the Shivalik Hills region of India. After reviewing the literature on coleoptera fauna in the Shivalik Hills areas indicates some fragmentary work on different aspects of coleoptera such as taxonomic, bioecology, species diversity, seasonal abundance, distribution and pest status. Though, it was found that some documents published but, it is needed to explore more coleopteran fauna and prepare consolidate documents from forest ecosystem of Shivalik Hills and detailed study on taxonomy, bioecology, species diversity, abundance and distribution in forest ecosystem due to current changing climatic condition
'Meritocracy' to 'Parentocracy': (De) constructing the Theoretical Dilemma in Basic Education of Japan and Bangladesh
Authors:   University of Dhaka, Bangladesh, Alam Md. Jahangir
Published: January 2022

Globally, access to basic education is a universal fundamental human right. Conversely, governments of underdeveloped and developing countries experience several challenges in ensuring equity in basic education. Along with the government initiatives, equity in basic education also depends on parental socio-economic status. Japan has achieved significant success in students' learning outcomes regardless of parental SES and might serve as a role model for Bangladesh to ensure equity in education. This case study research comprehends the dynamics of parental SES in accessing quality basic education for their children in Japan and Bangladesh. This paper argues that historically, meritocracy played a fundamental role in education. However, owing to the cultural paradigm shift, Parentocracy appears to be the better option for improving children's academic outcomes. The central theoretical argument of this study is to cognise whether meritocracy or Parentocracy in the modern education system can bring equity among children. This paper finds that Bangladesh can implement the Japanese experience through curriculum modifications and play-based learning approaches to ensure equity in basic education. Finally, parental involvement is vital in ensuring equity in education to accommodate all children to receive quality basic education in Bangladesh.
INTERACTION OF MATERIALS OF THE SEPARATOR OF HYDROGEN WITH SODIUM AND IMPURITY IN IT
Authors: Solomatin A , Orlova E , Sorokin A , Zasorin I , Zagrebaev S , Orlov A , Belozerov V
Published: January 2022

Work is devoted to a research of materials (zirconium and nickel) for hydrogen separation by its production with use of nuclear reactor in relation to the high-temperature fast sodium reactor (BN-VT) developed in IPPE, for production of hydrogen from hydrocarbon raw materials. These materials can be used in relation to the thermal reactor of the operating nuclear power plants or the reactor of new generation of VTGR, the electrolysis of water offered for receiving hydrogen in Research Center KI. Hydrogen on the NPP is formed also at reaction of water with structural materials, at water leak emergencies in sodium and steam - Zr reaction, at nuclear reactions, etc. In the real work results of LAES of the analysis of zirconium and nickel after their interaction with sodium and impurity in it are presented at 1053 K during 100 h in the sodium purified from oxygen to the level which is not exceeding 2 mg/kg and containing 6 mg/kg of hydrogen. Interaction of zirconium and nickel with sodium and impurity in it is revealed (hydrogen, oxygen, carbon). The maximum quantity of carbon, oxygen, nitrogen in zirconium is concentrated in a small layer of a sample by thickness near 10 microns. Hydrogen is concentrated probably near borders of grains, oxygen content in zirconium significantly exceeded oxygen content in initial zirconium that will be coordinated with results of thermodynamic calculations. It is shown that for increase in purity of hydrogen and durability of service of the device it is necessary to have zirconium as well as a nickel membrane in a gas phase.
ON COMBINED BURNING RATE OF LIQUID SODIUM STREAM AND POOL
Authors: Bondarenko A , Savin M , Supotnitskaya O , Yantseva L
Published: January 2022

Sodium fires with heat and fume release often happen at coolant circuit rupture of sodium fast reactor. At nuclear power plant with the sodium fast reactor the more probable scenario of sodium fire is considered burning of falling from damaged pipeline liquid sodium stream and the pool formed under it. The simple models for sodium fire which contains minimum of phenomenological parameters and gives the opportunity to get the upper limits of heat generation and amount of aerosol generated at sodium fire are more preferable. The physical and mathematical model for estimation a combined burning rate of the freely falling liquid sodium stream and the pool formed under it is proposed. The model is based on the laws of energy and mass conservation, the empiric correlations for burning rate of stream and pool sodium and assumption that pool depth cannot be less than equilibrium. The equilibrium pool depth depends on surface tension of liquid sodium. The probable heat transfer techniques for steam-phase and heterogeneous burning of metal are discussed. The experimental data of the Monju's sodium fire imitation are used for verification of the model presented. Good agreement between experimental and calculated data is observed.
THE EFFECT OF GAS BUBBLES ON THE EROSION OF STEEL IN LEAD
Authors: Alekseev V , Orlova E
Published: January 2022

The erosion of the steel surface in lead-based coolants is a powerful factor that initiates corrosion of the structural material. In the conditions of cavitation effect of gas bubbles on the oxide coating, breaches of its integrity can be created, as a result of which an intensive influence of lead on steel occurs on the "bare" areas. This leads to a multiple increase in the rate of corrosion. The regularities of erosive action of lead or its alloys on structural materials have not been sufficiently studied. In the experiments of different authors, differing data of flow rates of lead or lead-bismuth, to cause the erosion of steel, were obtained. In general, this phenomenon is a combination of a number of simultane-ously occurring thermohydraulic and physicochemical processes. Erosion failure of structural materials in a heavy coolant is rapidly developing and occurs in places with increased coolant velocities, in the field of hydrodynamic disturbances, and so on. Protective coatings formed on steels based on oxide technology are vulnerable from the point of view of erosion wear. The erosion of protective coatings on steels can in turn cause peeling of fuel rod cladding with end-to-end sheath breaking, which can proceed more intensively than the front dissolution of unprotected shells.The most significant factor leading to the erosion of steel in lead is the hydrodynamic effect of the coolant flow. In the presence of gas bubbles in the coolant, erosion occurs at the lowest flow rates, and the erosive effect itself has a cavitation character. The present report presents the results of studies of erosion of steel in lead under gas cavitation. A methodology for appropriate erosion testing was developed. Experiments were carried out on the installation with a rotating cylinder at different cylinder speeds and at various temperatures. A numerical solution of the differential equation of motion of the bubble wall is performed for the different hydrodynamic parameters of the lead flow. The experimental and calculated data obtained make it possible to estimate the process parameters under which a cavitation damage of the oxide coating occurs on the steel surface in the flowing lead.
LEAD CONTAINING COOLANTS IN PERSPECTIVE TECHNOLOGIES OF SOLID, LIQUID AND GASEOUS RAW MATERIALS PROCESSING
Authors: Ulyanov V , Askhadullin R , Melnikov V , Gulevskiy V , Koshelev M , Storozhenko A
Published: January 2022

The possibilities of innovative use of fusions of fusible metals (Pb, Pb-Bi) in the solution of hi-tech tasks are considered. The main actions for control and maintenance of quality of heavy liquid metal coolant in reactor installations and research stands of the Russian Federation taking into account operating experience of the lead-bismuth coolant in transport reactors are proved. The directions of innovative use of fusions of lead and the eutectic alloy of lead with bismuth, including in problems of processing of solid, liquid and gaseous raw materials are presented. Theoretically both the possibility of realization is experimentally proved and the prospects of thermoelectrochemical decomposition of water in liquid metal with receiving gaseous hydrogen and oxygen, receiving synthesis gas (mix of hydrogen and monoxide of carbon) from organic raw materials in the liquid-phase metal-oxide cycles "lead-lead oxide", liquid metal pyrolysis of solid organic polymeric waste with receiving the whole range of commodity products are confirmed, to distillation of water solutions without preliminary water treatment, synthesis of nanomaterials in liquid metal matrix.
APPLICATION OF HLMC IN THE TECHNOLOGY OF METAL RADIOACTIVE WASTE PROCESSING
Authors: Voronin I , Melnikov V , Yagodkin I , Posazhennikov A , Legkikh A , Nikolaev A
Published: January 2022

For the processing of radioactive waste is possible to use a variety of methods. One of the promising methods is the processing of metal radioactive waste (MRW) using smelting units. The article shows the possibility of using a eutectic lead-bismuth alloy for cooling the melting unit. Estimates show that when the body of the melting unit is cooled by a lead-bismuth eutectic, it is possible in principle to maintain the temperatures of the steel surfaces, not exceeding the allowable values. To solve this problem, a schematic diagram of an experimental setup for testing the MRW decontamination processes was developed, using lead-bismuth eutectic for cooling the hull. Gas purification systems at industrial enterprises are considered and on the basis of their analysis, as well as taking into account the specifics of the work, a gas purification system has been developed, which includes several elements: a precipitator, dilution sites, filters of preliminary and highly efficient cleaning. Tested experimental setup. A decontamination factor was obtained, which is more than three times the design value. For preparing this paper we used the results that had been obtained the course of implementation of the project of the Russian Ministry of Education and Science - “Development of a technology of utilization of metal radioactive waste on the basis of melting units with liquid-metal heat removal”. The unique identifier of PNIER (Applied Research and Experimental Developments) is RFMEFI62614X0002.
BOILING WATER REACTOR WITH TIGHT LATTICE OF FUEL RODS - DIRECT-CIRCUIT WATER-COOLED FAST REACTOR
Authors: Pivovarov V
Published: January 2022

The original concept of the boiling water reactor core with the reduced moderation of neutrons is proposed, in which a negative void reactivity effect is provided not by increase leakage of neutrons in the axial direction, but by an another physical principle. Instead of the traditional core flattening, a special heterogeneous arrangement is proposed, in which, along with tight lattice fuel assemblies (fuel rod diameter is 13.5 mm, the distance between the fuel rods is 1.3 mm) containing uranium-plutonium (MOX) fuel, there are fuel assemblies with uranium-thorium fuel (UO2+ThO2) with a small (~1 %) initial content of 233U or 235U and an increased water-fuel ratio (fuel rod diameter is 12.6 mm, the distance between the fuel rods is 2.2 mm). Uranium-thorium assemblies provide a negative component of the reactivity effect during dehydration of the core. The results of the calculation of the reactor with a capacity of 3000 MW (t) showed the possibility of achieving a breeding ratio of fuel within 0.96-1.0 with a negative void reactivity effect (-0.2 %). The main advantages of the proposed concept are a directcircuit scheme, medium technological parameters close to traditional boiling reactors, allowing the use of available construction materials and equipment.
ANALYSIS OF VVER REACTIVITY IN ACCIDENT WITH UNCONTROLLED WITHDRAWAL GROUP OF CONTROL AND PROTECTION SYSTEM CONTROL RODS AT MINIMAL POSSIBLE INITIAL NEUTRON FLUX
Authors: Nikolaev A , Uvakin M
Published: January 2022

The numerical simulation of accident with uncontrolled withdrawal of the most efficient group of control and protection system control rods at the initial states of power levels less than minimum controlled power level were realized. Analysis of this accident were perfomed for different initial states of reactor: for critical state and for a few states with various values of subcriticality. Beginning and the end of cycle for first and stationary fuel loading were considered in calculations. All calculations were done for AES-2006 project reactor plant VVER. The most important results obtained from the multivariate calculations with various combinations of initial power level, value of subcriticality, group of control and protection system control rods, No. of fuel loading, moment of fuel cycle were analyzed in more details. Consequences on the calculation results were done in conclusions. Similarly practical significant recommendations for safety analyzes that may be used for complement traditional calculations for minimum controlled power level were proposed. The calculations were performed by KORSAR/GP code with three-dimensional model of neutron kinetics.
EXPORT POTENTIAL OF RUSSIAN FAST REACTORS AND TECHNOLOGIES OF THE CLOSED NUCLEAR FUEL CYCLE
Authors: Gulevich A , Decusar V , Klinov D , Chebeskov A
Published: January 2022

The key competitive advantage of the Rosatom State Corporation on the world nuclear energy market is an integrated offer of a range of services, including construction, operation and maintenance of NPPs abroad for the entire life cycle up to 60 years, adopted for new projects with VVER-1200 reactors. In the long term, a significant expansion of the scale of international business is planned, which is reflected in the target indicators of Rosatom State Corporation and its organizations. Russia is a recognized world leader in the development of fast neutron reactors. To date, the technology of sodium fast reactors BN successfully demonstrated and entered the stage of commercialization. Russia is the only country in the world that has in its nuclear power structure operating BN 600 and BN 800 fast reactors. Technologies of fast reactors and closed nuclear fuel cycle, subject to their commercialization, may become one of the main export products of the Russian nuclear energy industry. Therefore, the leadership of the Rosatom State Corporation has set the task of developing the conditions for entering Russian NPPs with fast reactors the world nuclear energy markets.
THE THERMOHYDRAULIC MODEL OF A REACTOR WITH A LIQUID METAL COOLANT
Authors: Chudinova V , Nikonov S
Published: January 2022

This paper presents the results of calculations on the heat-hydraulic model of a reactor with lead coolant, which is based on the design scheme for the ATHLET code, obtained on the basis of public information on the BREST-OD-300. In the model used in this work, firstly, the descending sections of each loop are interconnected by transverse hydraulic connections and, secondly, the in-reactor space from the lower pressure collector of the reactor to the upper distributing collector of the reactor is also divided into a system of parallel cross-linked channels. The model includes a reactor, four circuits with main circulation pumps, two sectional steam generators on each loop. The secondary loop for each steam generator has a feed pump. The article describes the method of obtaining the initial stationary state for the developed scheme and presents the results of the calculation of the transitional regime associated with the disconnection of one of the MCP with saving the reactor power. It is shown that in the emergency loop a reverse coolant current occurs when the MCP is disconnected. This leads to the fact that less coolant flow passes through the core.The cold coolant from the fourth loop is not fed to the entrance to the core, but is mixed in the upper distributing collector after the core.
TO EXPERIMENTAL EVALUATION OF SOLUTIONS OF PARTIAL NEUTRON TRANSPORT EQUATIONS IN FAST BREEDER REACTORS
Authors: Seleznev E , Bereznev V , Chernova I
Published: January 2022

Solutions of the partial neutron transport equations, being components of the neutron flux density, which is usually measured by ionization chambers, would seem to not be fixed experimentally. But ionization chambers can be created on the basis of various fissile materials, including those having a threshold fission cross section, which is used in experiments on active reactors to “filter” the signal from the background, when, for example, the signal in the chamber is used only on the basis of high-energy neutrons eliminate taking into account the ionization of the gas medium in the ionization chamber from gamma-quanta, i.e. exclude the background. Such ionization chambers include cameras based on the use of 232Th, 238U, 236U, 237Np, 234U with a fission threshold energy of 1.5, 1.3, 0.70, 0.32 and 0.26 MeV, respectively. Together with the use of a 235U non-threshold chamber, these chambers offer the possibility of obtaining some experimental estimates of the partial solutions of the neutron transport equations in a fast neutron reactor. At the same time, the use of a threshold ionization chamber that records only a fraction of the neutrons from their full spectrum during a non-stationary process in a reactor with a change in the fraction of delayed neutrons in it, i.e. amplification of one or another part of the neutron spectrum may not fully take into account these changes and, thus, increase the measurement error. The possibility of estimating the magnitude of such an error is given by partial solutions of the neutron transport equations.
MEASUREMENTS OF 14-MeV NEUTRON TRANSMISSION THROUGH LITHIUM HYDRIDE LAYERS BY NEUTRON ACTIVATION ANALYSIS
Authors: Andreev S , Gabbasov D , Kostenko I , Litvin V , Lukin A , Moseeva A , Namakonov V , Sokolov Yu , Khmelnitskii D , Yudov A
Published: January 2022

The paper presents the results of experiments in 14-MeV neutron transmission through lithium hydride layers 25 cm thick. Measurements were performed by neutron-activation analysis using NG-12I neutron generator. The neutron generator operated in static mode. In order to determine 14-MeV neutron transmission through sample layers and to evaluate neutron spectrum changes we have used neutron-activation detectors based on the following activation reactions: 19F(n,2n)18F, 27Al(n,α)24Na, 64Zn(n,p)64Cu, 115In(n,n')115mIn, 65Cu(n,2n)64Cu, 63Cu(n,γ)64Cu. The activity of irradiated neutron-activation detectors was measured by the gamma-spectrometer with a semiconductor HPGe-detector. Neutron yield from the generator target was determined by monitor readings with fluoroplastic neutron-activation detector. The average 14-MeV neutron flow from the target came up to ~1010 n/s. The experimental geometry and measurement results are described. Based on measurement data we determined the number of reactions for each neutron-activation detector, as well as coefficients of 14-MeV neutron transmission through lithium hydride layers of different thickness. The obtained results can be used to validate neutron-physical calculations and to update neutron constant libraries.
SURFACE HARMONICS METHOD FOR NEUTRONIC CALCULATION OF 3D HETEROGENEOUS REACTOR WITH NON-SYMMETRIC CELLS
Authors: Elshin A , Agalina P
Published: January 2022

The paper addresses the application of the surface harmonics method to building a system of finite difference equations for describing the neutron field of a 3D heterogeneous reactor. The reactor is comprised of cells shaped as right-angled parallelepipeds with non-symmetric contents. Neutrons distribution in the cells is presented as a linear combination of trial functions that satisfy the neutron transport equation and some non-uniform boundary conditions allowing to consider the influence of the cell environment on the neutron distribution within the cell. Even given non-symmetric cells, we can build a system of equations (by sewing the neutron distribution in adjacent cells on joint faces), which, if diffusion approximation is used on the cell faces, are similar in appearance to the finite difference approximation of a group-diffusion equation (using a different method to obtain coefficients of equations, with complete matrices of group neutron diffusion coefficients). Finite-difference equations can also be defined by considering the additional neutron distribution components in cells (trial functions), which are necessary for valid transition to particular cases: two-dimensional and one-dimensional geometry. Due to cell asymmetry, additional components are introduced in equations and formulas for equation coefficients are corrected (correction is similar to the introduction of "discontinuity coefficients" used in foreign codes). As it is shown, the system of equations allows us to avoid using the diffusion approximation in reactor calculations since we can now sew together not only the neutron current density and neutron flux but also higher angular momenta at boundaries of cells without changing the form of equations (increasing the dimension of coefficient matrices and dimension of required vectors). Numerical results are presented on an example of 1D single-velocity test problems.
ON THE STABILITY OF A NEUTRON MATTER AND THE POSSIBILITIES OF ITS PRODUCTION ON THE EARTH. PROBLEM OF “ZERO ELEMENTS” IN WORKS D.I. MENDELEEV
Authors: Ryazantsev G , Beckman I , Buntseva I
Published: January 2022

D.I. Mendeleev assumed the existence of elements X (“Newtonium”) and Y (“Coronium”) in front of hydrogen in the zero group of the periodic system of chemical elements (PS). It should be recalled that Mendeleev was not mistaken in his predictions of new elements. When he applied the periodic law (PL) to analogues of boron, aluminum and silicon, he had complete confidence in success, because everything was obvious to him there. The idea of elements before hydrogen came to him immediately after the opening of the PL, but he published this only before his death. This question haunted him almost his entire creative life. D.I. Mendeleev did not have time to solve this problem, and his students and followers tried to forget it as “erroneous”. It should be noted that after D.I. Mendeleev the question of “zero” elements was repeatedly raised by many authors both in the past and in the present centuries, however, for brevity, we only mention the very first and most famous of them: for example, Ernest Rutherford in 1920 and Andreas von Anthropoff in 1926. Anthropoff was the first to propose the term «neutronium» to refer to a hypothetical element with atomic number zero, which he placed at the beginning of the periodic table (PT). At present, neutron matter, like neutron stars, is a recognized reality in astro- and nuclear physics. From the standpoint of general chemistry, a neutron substance can be classified as chemically simple (that is, it cannot be decomposed into simpler ones by chemical means), then the question inevitably arises of the corresponding element and its place in the PS. Based on the logic of the PL - (ordinal number - electric charge) - the ordinal number of neutron matter will correspond to zero, which makes us remember and develop the ideas of Dmitry Ivanovich Mendeleev about the zero group and period. Based on the works of Tamm, Hund and Migdal, the possibility of the stable existence of neutron matter at the micro level, and not only at the macro level, as is now believed in astrophysics, is stated. Neutron matter is considered as the primary cosmological substance, a candidate for dark matter and its production in laboratory conditions on Earth.
VERIFICATION OF SN-MODELLING OF PROGRAM FRIGATE ON SPECIAL TEST PROBLEMS
Authors: Zalesova A , Nikolaev A
Published: January 2022

The main features of the SN-block of program FRIGATE aimed at optimizing the processes of calculated justification are considered in this paper. The original features of program FRIGATE are noted, in particular, the ability to perform calculations on grids composed of arbitrary hexahedrons (including those containing non-planar faces) with an arbitrary way of joining grid cells along their faces. The program implements a wide range of boundary conditions (reflection, periodicity, rotational symmetry, surface source), which allow to create economical grid approximations of reactors with a high level of spatial detail and high quality of the form of the vast majority of hexahedral grid cells. FRIGATE program allows to perform calculations with quadratures with local thickening of nodes to account for the contribution of the effect of radiation through channels, cracks and gaps in structures. Also in the program FRIGATE the ability is implemented to calculate the method of "bootstrapping" large models in case of lack of computer memory in the calculation of the problem. The results of verification possibilities SN-block of program FRIGATE on two special developed problems and benchmark-experiment of fast reactor JOYO are presented in this paper.
PROGRAM DECAR FOR UNSTABLE NUCLIDE TRANSFER MODELING IN ROOMS AND CALCULATION OF EFFECTIVE DOSE
Authors: Davidenko O , Tsibulsky S
Published: January 2022

At the present time, the improving the accuracy of the assessment of effective doses to personnel and the public and of fuel cycle room pollution is actual. Description of algorithms and software implementation of computer system DECAR are considered in the article. DECAR is intended for computational simulation of transfer of unstable nuclides in fuel cycle facilities and for computational simulation of the exit of fission products beyond the protective shells. Description of method of solving isotope kinetics equations is presented. This method is realized by computer complex DECAR. Results of verification this complex are presented. They show that the program can be used to calculate the residual energy release and to calculate the isotopic composition of fission products. It is shown that deviations in the concentrations of nuclides do not exceed the established norm. The results of computational modeling of the activity exit from the reactor box to the hall with various parameters of mass transfer are obtained. The results of calculating the effective dose from external radiation in the reactor box and the integral inhalation doses in the room for various parameters of activity output to the room are presented.
COMPUTATIONAL SUBSTANTIATION OF ″MASTER″ REACTOR SAFETY
Authors: Dorokhovich S , Zajtsev A , Kascheev M , Levchenko V
Published: January 2022

Safety of the ″Master″ reactor plant in conditions of blocking flow section of a technological channel in the form of a Field’s tube is considered. The possibility of cooling a separate technological channel in the passive mode, considering heat-conducting zirconium matrix, is evaluated. It is noted that the project of the power unit with the ″Master″ reactor with 30 MW power has all the prerequisites for a successful application. General description of the power unit is presented. The mathematical formulation of the problem is formulated. Temperatures in fuel assembly with averaged properties and zirconium matrix (E-110 alloy) are described by differential equations of thermal conductivity with appropriate initial and boundary conditions. Temperatures in external and internal channels of the Field’s tube are calculated using a lumped parameters model and depend only on time. The mathematical formulation also includes an equation for determination of the flow rate of the coolant in the natural circulation circuit. Temperatures in assembly and zirconium matrix are determined by the tridiagonal matrix algorithm. System of equations for determination of temperatures in channels is solved analytically in a quasi-stationary approximation using the Laplace transformation. The developed mathematical model is implemented in the form of TK computer code. Calculations showed that the fuel temperature remains relatively low, and there is no melting of fuel elements.
EVALUATION OF ACCUMULATION SPEED OF ISOTOPES EINSTEINIUM AND BERKLIUM AT REACTOR PIK
Authors: Onegin M
Published: January 2022

The production of einsteinium and berklium isotopes at reactor PIK is considered. Presently the design of the reactor core for reactor operation after full power startup is under consideration. The additional places for isotopes production in the reactor core can be organized besides the central experimental channel. The irradiation conditions in the central water trap and in the core are comparing. For this purpose we are comparing calculated here the neutron spectrums, neutron fluxes and cross sections of the main reactions which are influence on the isotope production. We maid the approximation of the averaged over reactor spectrums neutron reaction cross sections using thermal cross sections, resonance integral and Westcott factors. For the realistic isotope targets we have calculated the amound of produced einsteinium and berclium. Also we compare the accumulation speed of these two elements in the central water trap and in the reactor core. We have made the verification of the MCNP code calculations of the isotope production using analitic calculations and code SCALE. We have demonstrated that for targets with initial isotopes placed in the core the amount of produced einsteinium and berclium can be 4-5 times greater then for targets placed in the central experimental channel. However the time for isotope production in the core can be about two times longer. We are discussing also the main ambiguities in the reaction cross section which influences on the amount of the produced einsteinium in the target. The main ambiguity in the calculations of the produced amount of rather long-lived isotope Es-254 is introduced the neutron reaction cross sections of Es-253 nuclei espessially the isomeric ratio of the reaction of the neutron radiation capture.
V.I. SUBBOTIN - FOUNDER OF SCIENTIFIC SCHOOL ″HEAT AND MASS TRANSFER, PHYSICAL CHEMISTRY AND TECHNOLOGY OF COOLANS IN ENERY SYSTEMS″ (ON THE 100TH ANNIVERSARY OF BIRTH)
Authors: Arnoldov M , Orlov Yu , Sorokin A , Rachkov V
Published: January 2022

Biographical data on the life and activities of the academician of the USSR Academy of Sciences V.I. Subbotin, presented and analyzed are the key areas of activity of the scientific school ″Heat and Mass Transfer, Physical Chemistry and Technology of Heat Transfer Fluids in Energy Systems″, which he created, which he was the head of. During the years 1953-1975 V.I. Subbotin led the research in the field of thermal physics at the Physics and Energy Institute (Obninsk). During this period, as a result of a large complex of experimental and computational studies of heat and mass transfer, physical chemistry and technology of water and liquid metals (sodium, eutectic alloys of sodiumpotassium and lead-bismuth, lithium, lead), the scientific foundations were created in the nuclear power plant under his leadership the use of liquid metal coolants in nuclear energy. The fundamental physicochemical and thermohydraulic laws of the coolant - impurities - structural materials - protective gas system were studied. The results of these studies made it possible, together with the institutes and the OKBM of the country performing the development of nuclear power and power plants cooled by water and liquid metals, to scientifically substantiate the thermohydraulic parameters and highly efficient technological processes, to develop and practically implement apparatuses and systems, ensured the successful operation of fundamentally new nuclear power plants with original scientific and technical solutions that had no analogue in world practice. In 1975, V.I. Subbotin was transferred to work in Moscow as director of the newly created wide-profile organization NPO ″Energia″ of the USSR Ministry of Energy and Electrification. While in this position, he put a lot of effort into creating a fundamentally new institute in the nuclear energy industry - VNIIAES, and also was the initiator and achieved the creation of a new institute for the nuclear industry - VNIIAM. In 1977-1988, V.I. Subbotin worked as the head of the Department of Thermophysics at MEPhI. During his work in this position, the composition of the department grew from 39 to 120 people; three collaborators defended doctoral dissertations, more than 30 defended doctoral dissertations. In 1968, he was elected a corresponding member of the USSR Academy of Sciences, and in 1987 a full member of the USSR Academy of Sciences. In 1988, Valery Ivanovich transferred to work at the USSR Academy of Sciences. The traditions of the scientific school are currently preserved and developed by the students, associates and followers of V.I. Subbotin.
THEORETICAL AND EXPERIMENTAL ESTIMATE OF 16O(p, α)13N AND 12C(p, γ)13N, 13C(p, n)13N REACTION CROSS-SECTIONS AVERAGED OVER THE RECOIL-PROTON SPECTRUM UNDER THE ELASTIC (n, p)-SCATTERING OF 14 MeV NEUTRONS IN HYDROGEN-CONTAINING MEDIA
Authors: Chernukhin Yu , Kostenko I , Sokolov Yu
Published: January 2022

The paper describes the calculation-and-experimental method used to measure σO, σC cross-sections of 16O(p, α)13N and 12C(p, γ)13N, 13C(p, n)13N reactions when these cross-sections were averaged over the spectrum of recoil protons in hydrogen-containing media. Measurements were performed with the help of the ∼14 MeV neutron source. As a result, cross-sections were determined to be σO=17,2 Mbarn, σC =2,59 Mbarn and this fairly well agrees with the sufficiently limited data available on the energy dependence of cross-sections from the above reactions in the proton energy region E≤10 MeV. Cross-sections of 16O(p, α)13N and 12C(p, γ)13N, 13C(p, n)13N reactions in the region of small and medium proton energies E≤10 MeV are of interest for the fundamental science when studying the CNO - cycle of hydrogen burning in rather large stars and also for applied sciences when it is necessary to solve certain problems like (i) considering "false" nitrogen formation when protein content is determined in grain crops using the neutron activation analysis and the 14 MeV neutron source; (ii) studying applicability of 16O(p, α)13N and 12C(p, γ)13N, 13C(p, n)13N reactions to produce the shortlived β+-radioactive isotope 13N directly in water and alcohol solutions of preparations for positronemission tomography (PET) with high-energy sources of (D,T)-neutrons; and (iii) studying applicability of 16O(p, α)13N and 12C(p, γ)13N, 13C(p, n)13N reactions with the formation of the β+-radioactive isotope 13N to determine actual distribution of absorbed doses in the patient body or in the phantom thereof based on the annihilation radiation after a session of neutron therapy with the 14 MeV neutron source or after a proton therapy session.
ON THE EFFECT OF OXYGEN IMPURITIES ON HEAT TRANSFER AT TRANSVERSAL FLOW OF STEAM-GENERATING TUBES IN NORMAL HEAT TRANSFER MODES AND WITH LEAD FREEZE
Authors: Grabezhnaya V , Mikheyev A , Kryukov A
Published: January 2022

The BREST-OD-300 steam generator project being developed at NIKIET is pioneering both in terms of the heat carrier used (lead) and in design implementation (coils of helical heat transfer tubes). The advantages of the designs of steam generators made in the form of helical coiled tubes, in comparison with straight tube designs are obvious. Helical coiled tubes are used in heat exchange equipment not only to increase the heat transfer surface, to solve the problem of thermal expansion, but also to increase the coefficient of heat transfer to the fluid flowing inside the tubes. In 2011-2017 years the thermohydraulic tests of various models of lead-heated steam generator were carried out at the IPPE SPRUT facility (IPPE). The test program was aimed to study the heat transfer and the thermal-hydraulic stability of the steam generating tubes. Throughout the entire range of changes in operating parameters, no pulsating modes were detected with overturn of circulation in the water circuit. The design temperatures of superheated steam were obtained in nominal operation. The results provide extensive information on water heat transfer in different zones of the steam generating channel under various operating conditions (nominal and partial modes, starting modes). However, to verify the computer codes, experimental data on the heat transfer of lead coolant around the bundles of heat transfer tubes are necessary. Due to the small twist angle of the tubes in a full-scale steam generator, it can be said that heat transfer is close to heat transfer during transverse flow. A model with a transverse flow of lead coolant around steam-generating tubes was developed at the SSC RF - IPPE. The main goal of the research was to obtain data on the effect of oxygen concentration in lead on heat transfer in normal heat transfer modes and with lead freezing. Throughout the entire range of changes in the initial temperature values of lead and water, blocking of the annular space by frozen lead was not recorded.
IMPLEMENTATION OF COMPARATIVE BURNUP CALCULATIONS
Authors: Los V , Kuryndin A , Kirkin A , Sinegribov S , Makovskiy S
Published: January 2022

Features of calculation methods implemented in computer codes, evaluated nuclear data libraries and computer models have a significant impact on the results of burnup calculations. This article contains comparative calculations of VVER-1000 reactor nuclear fuel burnup and demonstrates an influence of different factors on the results of these calculations. The results of comparative analisys demonstrate the necessity to intensify creation of benchmarks with experimental data on nuclide concentarions in fuel including fuel for fast neutron reactors.
ASSESSMENT OF THE SYSTEM CHARACTERISTICS OF A REACTOR WITH SUPERCRITICAL COOLANT PARAMETERS FOR VARIOUS FUEL CYCLES
Authors: Lapin A , Bobryashov A , Blandinsky V , Bobrov E
Published: January 2022

Nowadays nuclear energy operates in an open fuel cycle. One of the most important directions in the development of nuclear energy is the closure of the nuclear fuel cycle. The solution to this problem is possible with the use of fast neutron reactors. To achieve this goal, the possibility of using a reactor with a fast-resonance neutron spectrum cooled by supercritical water (SCWR) was considered. The SCWR reactor can be effectively used in a closed nuclear fuel cycle, since it makes it possible to use spent fuel and dump uranium with a small amount of plutonium added. The layout options of the core with a change in the size of the core and reproduction zones are considered. The possibility of placing reproduction zones from various materials inside the active zone was evaluated. Based on the studies, an acceptable version of the core is selected in terms of system characteristics. For the considered arrangement of the reactor core, the possibility of shorting the uranium-plutonium and uranium-thorium fuel cycles has been investigated. The system characteristics of the reactor installation were studied for the following fuel load options: 1. Loading MOX fuel into the core, depleted uranium in the lateral zone of reproduction. 2. Loading of uranium-thorium fuel into the core and side screens. The results of the assessments of the system characteristics of the reactor are considered in the article.
FLUID DYNAMICS IN THE CORE OF A WWER REACTOR WITH TVSA OF VARIOUS DESIGNS
Authors: Dmitriev S , Dobrov A , Doronkov D , Lyskova V , Pronin A , Rubtsova E , Ryazanov A , Solntsev D , Khrobostov A
Published: January 2022

The results of experimental studies of the interassembly interaction of coolant in VVER reactor core which consists of TVSA-T and upgraded TVSA are presented. The modeling of coolant flow in the fuel assembly (FA) was carried out on an aerodynamic stand. The studies were carried out on a fragment model of VVER reactor core and consisted in measuring the velocity vector modulus in the characteristic zones of both TVSA and interassembly space of VVER reactor core. The measurements were carried out by a five-channel pneumometric probe. An analysis of the spatial distribution of projections of the absolute flow velocity allowed to detail the pattern of flow-round by the coolant flow of spacer, mixing and combined spacer grids TVSA. The results of study of interassembly interaction of coolant between neighboring TVSA-T and upgraded TVSA were adopted for practical use by JSC “Afrikantov OKBM” in assessing heat engineering reliability of VVER reactor cores and are included in the data-base for verification of computational fluid dynamics programs (CFD codes) and detailed cellular calculation of VVER reactor core.
SPECIFIC FEATURES OF NUMERICAL SIMULATION OF THERMAL OPERATION MODE THE SPENT FUEL POOLS OF BILIBINO NPP
Authors: Sergeev V
Published: January 2022

The paper considers methodological aspects during the development of thermo-hydro-dynamic of numerical calculation models for spent nuclear fuel pool (SNFP) on the example of Bilibino NPP by using international industry codes. The purpose of these models development is the fast simulation of thermal and humidity operational modes of spent fuel pool during periods of filling, "wet" and dry storage with an interval of up to 50 years, taking into account passive and forced convection heat removal systems. The following methodological aspects are considered in detail: 1. Use of isotope kinetics codes for calculating of the heat power dynamics for separate spent fuel assembly. 2. Calculation method of the heat removal power from the evaporation mirror of SNFP during “wet” storage, including evaporation power calculation and exhaust ventilation operation. Using the law of similarity of heat transfer and mass transfer processes (Lewis' law) for evaporation calculations. 3. Methods of accelerated computational forecasting of the dynamics of the thermal regimes of the SNFP during “wet” storage. 4. Condensation of atmospheric moisture at the bottom of the SNFP after “dry” storage and methods for its removal. It is shown that TRAC (TRACE) code with a 3D porous body model and complete evaporation-condensation models is the most suitable for solving the problems under consideration among the system thermo-hydraulic two-phase codes for nuclear energy.
MAINTAINING THE QUALITY OF LEAD CONTAINING COOLANT IN ADVANCED TECHNOLOGIES RECYCLING ORGANIC AND INORGANIC WASTE
Authors: Ulyanov V , Koshelev M , Kharchuk S
Published: January 2022

The unique properties of heavy liquid metal coolants (low chemical activity when interacting with air, organic compounds, water and water vapor, high boiling point, low vapor pressure) allow them to be used both in the first circuits of fast neutron reactor plants and in heat exchangers used for processing various organic and inorganic raw materials. Such devices can be either recuperative or mixing type. The latter additionally have the following advantages: lower cost and simplicity of construction; no heat transfer surfaces that can be subject to corrosion and contamination; reduced dimensions due to a larger specific heat transfer surface. The greatest degree of study relates to liquid metal pyrolysis of solid substances, direct contact distillation of aqueous solutions without preliminary preparation, improving the efficiency of continuous casting machines due to the replacement of water with lead-bismuth eutectic. The results of research in support of the above processes are presented. The methods of maintaining the quality of lead-containing heat carriers do not have fundamental differences for the conditions of reactor installations and in relation to promising technologies for processing organic and inorganic raw materials. At the same time, the requirements for devices for implementing these methods in promising technologies for processing various raw materials are significantly lower: in general, a sensor for measuring the thermodynamic activity of oxygen and a device for entering gas (hydrogen-containing or oxygen-containing) mixtures in the coolants.