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The focus of this publication is on collecting current practices in Member States related to design extension conditions (DECs) with core melting. The information provided is based on the feedback from technical experts from Canada, France, Finland, India, the Islamic Republic of Iran, the Russian Federation, and the United States of America. There is, however, still no common understanding of DECs due to the complexity of phenomena and insufficient experimental data. This publication identifies current approaches of IAEA Member States with active nuclear power programmes and discusses the regulatory perspective and technical rationale. It attempts to find common practices and possible areas for harmonization of the main rules related to the analysis of DECs with core melting for new water cooled reactors, including their selection for the safety demonstration.

This article presents the results of an experimental study of the coolant flow in a fuel rod bundle of a nuclear reactor fuel assembly of a small modular reactor for a small ground-based nuclear power plant. The aim of the work is to experimentally determine the hydrodynamic characteristics of the coolant flow in a fuel rod bundle of a fuel assembly. For this purpose, experimental studies were conducted in an aerodynamic model that included simulators of fuel elements, burnable absorber rods, spacer grids, a central displacer, and stiffening corners. During the experiments, the water coolant flow was modeled using airflow based on the theory of hydrodynamic similarity. The studies were conducted using the pneumometric method and the contrast agent injection method. The flow structure was visualized by contour plots of axial and tangential velocity, as well as the distribution of the contrast agent. During the experiments, the features of the axial flow were identified, and the structure of the cross-flows of the coolant was determined. The database obtained during the experiments can be used to validate CFD programs, refine the methods of thermal-hydraulic calculation of nuclear reactor cores, and also to justify the design of fuel assemblies.

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

Described shortly here is a procedure of demounting, removal, transport and long-term storage of the SM-3 core, based on the previous experience of reactor refurbishment undertaken in 1991. Prior to performing refurbishment, computations and calculated data analysis were performed to prove radiation safety of this work, which included estimation of the activity level for activation products in the structural materials of the nuclear research reactor core and the radiation conditions at different stages of its handling. As evidenced by the calculated data, the activity of the main dose-forming radionuclide 60 Co attains equilibrium in about 12 years of radiation exposure. Taking into account the fact that the time period between two refurbishments was longer than 12 years, the calculated values of the equivalent dose rate were normalized to the radiation monitoring data obtained during the previous refurbishment, taking into account the calculated activity of 60 Co radionuclide. The normalization made it possible to confirm reliability of estimates. The obtained activity data of activation products and taking into account the time spent during the SM-3 refurbishment in 1991, the radiation impact on personnel was estimated. Calculated values of the anticipated effective radiation exposure doses to the personnel engaged in the refurbishment revealed that the main limits of the personnel radiation exposure established in accordance with NRB-99/2009 were not exceeded. Comparison of the results of calculating the equivalent dose rate with the results of radiation monitoring at various points allowed us to establish that during the calculation and analytical justification of the radiation safety of work, the assessment of reflected radiation was significantly underestimated. But the radiation monitoring data, personal radiation monitoring, as well as recorded data of automatic radiation monitoring system show that all work was performed in compliance with the requirements of regulatory documents in the field of radiation safety.

Neutron flux measurement for research reactor in-core is typically performed using the neutron activation method. In this study, an 197 Au sample The sample in this study used an 197 Au was irradiated to measure neutron.The neutron was calculated by determining the saturation activity of the irradiated sample, with correction factors applied for counting efficiency and uncertainty. This study analytically evaluated the influence of each correction factor on the neutron flux measurements in the TRIGA 2000 reactor core. The results indicate significant difference in flux calculations when using ϵ Overall Counting and ϵ Calibration. The averages value of neutron flux using ϵ Overall Counting is only 0.5% of the flux value using ϵ Calibration . The difference is due to the lack of consideration of intrinsic efficiency, which affects the full peak energy efficiency of the detector. As a result, ϵ Overall Counting is also affected by intrinsic efficiency.

A methodological approach has been developed for the computational investigation of the thermal-hydraulic processes taking place in a sodium cooled fast neutron reactor based on a Russian computational fluid dynamics code, FlowVision. The approach takes into account the integral layout of the reactor primary circuit equipment and the peculiarities of heat exchange in the liquid metal coolant, and makes it possible to model, using well-defined simplifications, the heat and mass exchange in the process of the coolant flowing through the reactor core, and the reactor heat-exchange equipment. Specifically, the methodological approach can be used for justification of safety during the reactor cooldown, as well as for other computational studies which require simulation of the integral reactor core and heat-exchange equipment. The paper presents a brief overview of the methodological approaches developed earlier to study the liquid metal cooled reactor cooldown processes. General principles of these approaches, as well as their advantages and drawbacks have been identified. A three-dimensional computational model of an advanced reactor has been developed, including one heat-exchange loop (a fourth part of the reactor). It has been demonstrated that the FlowVision gap model can be applied to model the space between the reactor core fuel assemblies (interwrapper space), and a porous skeleton model can be used to model the reactor’s heat-exchange equipment. It has been shown that the developed methodological approach is applicable to solving problems of the coolant flow in different operating modes of liquid metal cooled reactor facilities.

Advanced pressurized water reactors are the main part of a new generation of nuclear power plant projects under development that provide cost-effective power production for various needs (Yemelyanov et al. 1982, Klimov 2002, Boyko et al. 2005, Baklushin 2011, Bays et al. 2019, Nuclear Technology Review 2019). The innovative technologies are aimed at improving the safety and reliability as well as at reducing the cost of NPPs. At the same time, improvements in design, technological and layout solutions are focused primarily on the reactor core. Assessments of the efficiency of these improvements are preceded by numerical simulations of the processes in the core, in particular heat generation and sink, with account for the difference between the study object and the standard version tested in operational practice. The authors of the article propose a method for calculating the temperature field in the core of a heterogeneous reactor (using the example of a pressurized water reactor), which makes it possible to quickly assess the level of temperature safety of various changes in the core and has the necessary speed for analyzing transients in real time. This method is based on the energy equation for an equivalent homogeneous core in the form of a heat equation that takes into account the main features of the simulated heterogeneous structure. The procedure for recovering the temperature field of a heterogeneous reactor uses the analytical relation obtained in this work for the heat sink function, taking into account inter-fuel element heat leakage losses. Calculations of temperature fields in the model of the PWR type reactor (The Westinghouse Pressurized Water Reactor Nuclear Plant 1984) were carried out in stationary and transient operating modes. The calculation results were compared with the results of CFD simulation. The area of competing use of the temperature field recovery method was indicated.

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

Anomaly detection for core temperature has great significance in maintaining the safety of nuclear power plants. However, traditional auto-encoder-based anomaly detection methods might extract the latent space features with redundancy, which may lead to missing and false alarms. To address this problem, the idea of feature disentangling is introduced under the auto-encoder framework in this paper. First, a feature disentangling auto-encoder (DAE) is proposed where a latent space disentangling loss is designed to disentangle the features. We further propose an incrementally feature disentangling auto-encoder (IDAE), which is the improved version of DAE. In the IDAE model, an incremental feature generation strategy is developed, which enables the model to evaluate the disentangling degree to adaptively determine the feature dimension. Furthermore, an iterative training framework is designed, which focuses on the parameter training of the newly incremented feature, overcoming the difficulty of model training. Finally, we illustrate the effectiveness and superiority of the proposed method on a real nuclear reactor core temperature dataset. IDAE achieves average false alarm rates of 4.745% and 6.315%, respectively, using two monitoring statistics, and achieves average missing alarm rates of 6.4% and 2.9%, respectively, using two monitoring statistics, outperforming the other methods.

Small modular reactors (SMRs; i.e., nuclear reactors that produce < 300 MWelec each) have garnered attention because of claims of inherent safety features and reduced cost. However, remarkably few studies have analyzed the management and disposal of their nuclear waste streams. Here, we compare three distinct SMR designs to an 1,100-MWelec pressurized water reactor in terms of the energy-equivalent volume, (radio-)chemistry, decay heat, and fissile isotope composition of (notional) high-, intermediate-, and low-level waste streams. Results reveal that water-, molten salt–, and sodium-cooled SMR designs will increase the volume of nuclear waste in need of management and disposal by factors of 2 to 30. The excess waste volume is attributed to the use of neutron reflectors and/or of chemically reactive fuels and coolants in SMR designs. That said, volume is not the most important evaluation metric; rather, geologic repository performance is driven by the decay heat power and the (radio-)chemistry of spent nuclear fuel, for which SMRs provide no benefit. SMRs will not reduce the generation of geochemically mobile 129I, 99Tc, and 79Se fission products, which are important dose contributors for most repository designs. In addition, SMR spent fuel will contain relatively high concentrations of fissile nuclides, which will demand novel approaches to evaluating criticality during storage and disposal. Since waste stream properties are influenced by neutron leakage, a basic physical process that is enhanced in small reactor cores, SMRs will exacerbate the challenges of nuclear waste management and disposal.