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The article presents the results obtained from verification of the EUCLID/V2 coupled code developed at the Nuclear Safety Institute of the Russian Academy of Sciences, which is intended for analysis of accident conditions in liquid metal cooled fast reactors. The additional capabilities available in the code in comparison with its first version include, in particular, analysis of individual equipment components in the 3D approximation, consideration of the transport of fission products and corrosion products in the coolant and in the nuclear power plant buildings, and also analysis of severe accidents in a fast reactor. The article presents the code verification results and assessment of its applicability for analysis of accidents involving destruction of fuel pins and the reactor core. The verification was carried out against the data obtained at experimental facilities and from analytic tests. Information about the key experiments used to validate the code is briefly outlined. In particular, data of experiments carried at the Oak Ridge, Argonne, and Sandia National Laboratory, the United States; at the National Nuclear Center of the Kazakhstan Republic; and on the test bench at the Nizhny Novgorod State Technical University (NSTU) in Russia are used. Modules of the coupled code EUCLID/V2 integrated code verification matrix are given. The errors of calculating the parameters most important for analysis of an accident's consequences evaluated using the EUCLID/V2 code are proven with due regard to the computation and experimental results. The ranges of parameters within which the code has been verified are determined. The uncertainty and sensitivity of computation results are also analyzed based on the results from simulating certain experiments. Factors having the main influence on the computation results are determined. It is shown that the computation results are consistent with the experimental results subject to the input data uncertainty. [ABSTRACT FROM AUTHOR]

The article presents and analyses the results from first-priority thermophysical investigations aimed at assessing the characteristics and safety of water-cooled reactors, including enhancement of the nuclear fuel technical and economic characteristics, design solutions for passive safety features, and stable operation of the passive heat removal system (PHRS) air-cooled heat exchangers under extreme meteorological conditions. It is shown that the fuel assembly (FA) critical power level depends both on the type and number of spacer and mixing grids and on the burnout occurrence spot. The critical heat fluxes that occur when the coolant flowrate drops by 4% are higher in unsteady modes than they are in steady modes of operation at the same flow parameters. Information about the upgraded base of estimated experimental data on the critical heat fluxes in the new FA designs is given. The article also presents data on pressure variations in the main components of the RU-ZO experimental bench (containing the reactor plant and containment models) in the experiment modeling an accident involving reactor coolant pipeline (RCPL) rupture at the reactor inlet (scenario 1) and decrease of the steam generator (SG) model condensing power in the bench (scenario 2). It is found that the drop of SG condensing power is approximately 30% of its level at the onset of the SG condensing operation mode during accidents according to both the scenarios. However, the remaining condensing power level is sufficient for reliably cooling the VVER-1200 reactor core for no less than 24 h with joint operation of the second-stage hydro accumulators (HA-2) and PHRS. Experimental investigations of the external aerodynamic characteristics of the Novovoronezh-2 NPP (NVNPP-2) main building aimed at checking if proper operation of the PHRS is secured independently of the wind load have shown that extreme meteorological conditions, including hurricane winds and dust storms, may manifest themselves in a decreased capacity of part of the PHRS air-cooled heat exchangers. Studies carried out for assessing the design of a water-cooled nuclear power facility (NPF) for an electrical power capacity of 1700 MW operating at supercritical parameters and the experimental VVER-SKD-30 reactor with a thermal power capacity of 30 MW are mentioned. The problems that have to be solved in carrying out further thermophysical investigations of water-cooled reactors for supercritical parameters are listed. [ABSTRACT FROM AUTHOR]

The problem of an increase in the Al concentration in the primary coolant of a research nuclear reactor is considered. The dispersion of aluminum hydroxide by fission products of uranium sorbed on the reactor core surfaces is suggested as a possible cause of this phenomenon. The correlation between the Al concentration and the reactor core contamination with uranium was considered with the IVV-2M research nuclear reactor as an example. Contamination of the reactor core with uranium was accompanied by a considerable increase in the activity concentrations of 24Na, 27Mg, and 239Np washed out from oxide deposits on the reactor core surfaces, which confirms the hypothesis put forward. [ABSTRACT FROM AUTHOR]

Depletion calculations have been performed for the McClellan reactor history from January 1990 through August 1996. A database has been generated for continuing use by operations personnel which contains the isotopic inventory for all fuel elements and fuel-followed control rods maintained at McClellan. The calculations are based on the three-dimensional diffusion “theory code REBUS-3 which is available through the Radiation Safety Information Computational Center (RSICC). Burnup-dependent cross-sections were developed at zero power temperatures and full power temperatures using the WIMS “code (also available through RSICC). WIMS is based on discretized transport theory to calculate the neutron flux as a function of energy and position in a one-dimensional cell. Based on the initial depletion calculations, a method was developed to allow operations personnel to perform depletion calculations and update the database with a minimal amount of effort. Depletion estimates and calculations can be performed by simply entering the core loading configuration, the position of the control rods at the start and end of cycle, the reactor power level, the duration of the reactor cycle, and the time since the last reactor cycle. The depletion and buildup of isotopes of interest (heavy metal isotopes, erbium isotopes, and fission product poisons) are calculated for all fuel elements and fuel-followed control rods in the MNRC inventory. The reactivity loss from burnup and buildup of fission product poisons and the peak xenon buildup after shutdown are also calculated. The reactivity loss from going from cold zero power to hot full power can also be calculated by using the temperature-dependent, burnup dependent cross-sections. By calculating all of these reactivity effects, operations personnel are able to estimate the total excess reactivity necessary to run the reactor for the given cycle. This method has also been used to estimate the worth of individual control rods. Using this approach, fuel management and core loading can be optimized such that each individual fuel element and fuel-followed control rod is used to its full potential before being replaced with fresh fuel. This fuel management strategy allows a significant cost saving to MNRC by reducing fuel replacement costs and maximizing the usefulness of each element in the inventory.

• Incore detectors results better than ex-core detectors, blending them leads to highest accuracy. • ROM is applicable when direct measurement information is unavailable. • Bayesian principles play a crucial role for core digital twins under uncertain conditions. In constructing a digital twin for a nuclear reactor core, it is important to consider the influence of randomness from various sources. Data assimilation (DA) can combine time distribution observations with dynamic models to approximate the real state of a physical system. Machine learning (ML) and DA share similarities under the Bayesian framework, and using probabilistic ML may provide a way to improve or replace current DA techniques. This paper proposes using a probabilistic ML as Bayesian neural network (BNN) to solve an inverse problem of core monitoring and demonstrates its feasibility through a pressurized water reactor core simulation analysis. Model order reduction technology is also analyzed, and the feasibility and benefit of using it to achieve core monitoring under steady-state conditions is preliminarily verified and discussed. Future work will focus on improving estimation and prediction models under transient operating conditions by unifying DA and ML under the Bayesian framework. [ABSTRACT FROM AUTHOR]

Measuring the neutron spectrum in the reactor core is quite difficult to do. The commonly used technique is using a multi-foil activation-based neutron spectrometer. This technique provides precise positional measurement but low accuracy and is difficult and expensive to manufacture. A technique for measuring the neutron spectrum in the reactor core has been developed using a single moderator-based neutron spectrometer. This technique is simple, easy to make, and inexpensive. The measurement time is relatively fast, and the results are quite accurate. This technique uses a single cylindrical moderator and places seven passive Au detectors with varying depths. The response of each detector is calculated using the Monte Carlo program MCNPX. The neutron spectrum measurement in the reactor core was carried out by placing the spectrometer above the core for a few minutes. The foil activity was measured using a gamma counter, and the neutron spectrum was calculated using the UMG unfolding program. After measuring the neutron spectrum with a single moderator-based neutron spectrometer in the Triga Mark II reactor core, Kartini, compared with the simulation results, the comparison results show values that are close to the same, where the difference is around 2.2%. [ABSTRACT FROM AUTHOR]

The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP) VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD) codes and for detailed cellwise analyses of the VVER-100 reactor cores.

The paper considers the flow and heat transfer in coolant flows cooling the fuel rod bundles of nuclear reactors with modified cell-type spacer grids. The spacer grids were developed by JSC Mashinostroitelny Zavod, Elektrostal, and differ from standard grids in the inclination of the corrugations that contact the fuel cladding outer surface. The coolant flow inside such cells leads to tangential components of the velocity vector formed inside the flow with an orientation in accordance with the corrugation inclination direction. By arranging geometrically different or identical cells within the grid array, it is possible to generate different secondary flows in the rod bundles downstream of the grid. The paper considers spacer grids that generate two types of secondary coolant flows: around the fuel rods and between the rod rows. The investigation was based on computational fluid dynamics methods with the calculation results validated against aerodynamic test data. Mechanisms of the secondary flow formation downstream of the grids have been described, and the intensities thereof have been quantitatively estimated. Data is presented on the grid flow resistance coefficients at different Reynolds numbers both in conditions of a weakly compressed isothermal air flow and with parameters representative of the VVER-1000 reactor primary coolants. Intensification of the coolant mixing downstream of the spacer grids was analyzed by a thermal wake method for different boundary conditions. A conclusion has been made that the modified cell-type spacer grids can be used for and are efficient in flattening temperature non-uniformities in the coolant flow due to creating a directed convective transport.

Description is given of the methodology for experimental substantiation of hydrodynamic characteristics of imitators of fuel channels in the model of reactor unit with two-loop configuration. It is known from the experience of development of reactor units of different types that it is practically not feasible to select the channel to serve as the imitator with simple geometry for which the dependence of pressure drop due to hydraulic losses would be the same as the dependence for full-scale operating fuel channel containing fuel assemblies. Therefore, only approximate modeling of hydraulic losses in operational fuel channels with fuel assemblies and imitating channels within the limited range of Reynolds numbers can be discussed as well as inclusion of all coolant flow regimes within the reactor vessel predicted by design calculations. Selection of final geometry of imitating channels was made based on the following several basic assumptions of approximate modeling of flow path in the full-scale fuel channel with fuel assembly under nominal operational mode of the nuclear reactor using the channel with simplified geometry: (1) range of variation of coolant flow rates in the operational fuel channel with fuel assembly in nominal operational mode of the reactor is known from physical and from preliminary thermal physics calculation studies; (2) equality of Euler numbers for nominal coolant flow regimes in the reactor and for coolant flow in imitating channels. Experiments were conducted with three types of imitators and the most suitable among them was chosen as the result. The finally selected design of imitating channel included a pipe with high-quality of manufacturing, two diaphragms and a nozzle with configured using the method suggested by Vitoshinsky.

Abstract: The authors analyzed events related to the largest nuclear accident in the history of the Soviet Navy, which occurred on the nuclear submarine K-431 on August 10, 1985, during scheduled operations at a dockyard in Chazhma Bay near Vladivostok. For reasons of security, the aftermath of these events remained unknown to a wide circle of specialists and public for over 30 years, prompting ambiguous interpretations and conjectures. This article describes the findings, including the collection, generalization, and systematization of surviving incomplete and limited data; numerical and model calculations; and the analysis of the accident’s causes and development. The radioactive and radioecological consequences, rehabilitation measures, and the disposal of the formed radioactive waste and the damaged nuclear submarine are considered. The consequences of the radioactive release into the Sea of Japan, the cloud traveling to the north along the Russian-Chinese border, and the transborder transfer of radioactive substances for the populations of the neighboring countries are evaluated for the first time. [ABSTRACT FROM AUTHOR]

High reliability and security of power unit are basic requirements when the power unit maneuvering mode operation. The reactor stability under disturbances both at steady load and maneuvering load embodies the guarantees of power unit safe and reliable operation. A quantitative measure of the reactor stability is assessed by the axial offset representing the technological characteristics of energy release uniformity, therefore the axial offset minimum deviation is WWER-1000 operation efficiency measure. The power unit capacity automated control systems’ influence on axial offset under maneuvering mode is investigated. Considered is the power unit compromise-combined control program, which maintains a constant axial offset value when power unit switching from one power level to another.