Atomic Energy最新文献

The study presents an analytical expression based on the generalized diffusion equation derived using the radial averaging method to determine the rotor radius of a separation gas centrifuge, which is optimal in terms of maximum separation power per unit volume of the centrifuge. The derived generalized equation divides the gas flow inside the rotor into circulation and transit streams, using the vortex potential instead of the current function. This approach assumes the optimal radius to be proportional to the feeding stream of the centrifuge.

The article describes a method for calculating surface tension in the TwoPhase code for direct numerical simulation of thermal fluid dynamics for a two-phase compressible medium, including two-component mixtures, taking into account interphase heat and mass transfer and condensed gas or Noble–Abel equation of state for a weakly compressible medium. A six-equation model is used to calculate the two-phase medium. The surface tension is taken into account in the form of additional terms of equations for the components of momentum and total energy, which are calculated using an additional vector parameter equal to the volume fraction gradient of the liquid phase. The results of method verification on the classical problem of a gas bubble at rest in a fluid are presented.

Background

The main conditions for the safe operation of fuel assemblies (FAs) in the core of the research IVV-2M heterogeneous water-water reactor of a pool type involve the absence of surface boiling on the fuel element cladding, including under the layer of deposits. The values of the coolant flow rate through the inter-fuel gaps can be used to predict the coolant temperature at the outlet of the FA over the campaign and justify the operating limit settings.

Aim

To determine the functional dependence of the coolant flow rate through FAs on their burnup depth and residence time in the core based on the results of measuring the coolant flow rate through individual FAs.

Materials and methods

A FLUXUS ADM 7407 ultrasonic device was used for mobile monitoring of the coolant flow rate through the inter-fuel gaps of FAs in the IVV-2M nuclear research facility; the device was developed and manufactured at the N.A. Dollezhal Research and Design Institute of Power Engineering JSC. The measurements were carried out under a layer of water; the flowmeter was installed directly on the FA. The measurement methodology and derived analytical dependence of the flow rate on the pressure drop in the core are presented in the article “Study of the coolant flow rate through the fuel assemblies of the IVV-2M reactor core. Part 1”.

Results

Analytical dependencies of the flow rate on the burnup depth and residence time of the FA in the core are derived using the data from measurements of the coolant flow rate in the IVV-2M FAs during several years of reactor operation.

Conclusion

The obtained dependencies can be used to estimate the coolant flow rate through the inter-fuel gaps of the FA without conducting direct measurements, as well as to more accurately determine the temperature on the fuel element cladding, thereby increasing the operational safety of the nuclear research facility. The measurements should be continued on operating reactor for accumulating statistical data to clarify the dependencies and reduce errors.

Background

To close the nuclear fuel cycle, the state program for the development of nuclear power industry in the Russian Federation assumes the development of new radiochemical production facilities and technologies. The most important task is to justify fire, explosion, nuclear, and radiation safety, as well as to minimize the risk of accidents at newly developed radiochemical facilities for nuclear fuel production and reprocessing of spent nuclear fuel (SNF). However, verified and certified codes for predicting processes in the non-reactor part of the fuel cycle are actually unavailable.

Aim

To develop a system of mathematical models and codes substantiating safety of radiochemical technologies in the production of nuclear fuel and reprocessing of SNF; to test the operability of the codes, their system, and proposed approach.

Materials and methods

A calculated assessment of fire, explosion, nuclear, and radiation safety of radiochemical production processes and equipment was carried out using a system of mathematical models and codes providing the necessary overall operability. The assessment method includes calculation of chemical and isotopic composition of fuel, determination of process characteristics, simulation of fluid dynamics processes, as well as calculation of critical characteristics and intensity of radiation fields, explosion and fire hazard indicators.

Results

Requirements for mathematical models are defined. An approach to the calculation safety assessment is proposed based on end-to-end calculations using a system of mathematical models and codes. The effectiveness of the developed system for analyzing emergency scenarios and assessing safety criteria for technological solutions is demonstrated. The dose load of the extractant and release of radiolytic hydrogen in dissolution and fractionation operations are assessed.

Conclusion

The operability of both individual codes and proposed approach to the calculated safety assessment as a whole is confirmed. A system of mathematical models and codes should be developed and verified for subsequent application in safety substantiation of radiochemical production facilities. The proposed system of models and codes will significantly reduce the time and costs of experimental substantiation.

Background

Transmutation of Np, Am, and Cm minor actinides appears to be a key direction for reducing the long-term radiotoxicity of spent nuclear fuel from thermal and fast reactors. High specific heat emission of ~2.5 kW/kg makes Cm isotopes difficult to include in fast reactor fuel due to the features of fabrication. The solution may be the transmutation of Cm in specialized units.

Aim

To perform a computational study of Cm transmutation in the MSR‑B molten salt nuclear reactor for reducing the radiotoxicity of spent nuclear fuel.

Materials and methods

The object of the study is the fuel campaign of the MSR‑B reactor: 2400 MW(t), 73% LiF—27% BeF₂ carrier salt, and CmF₃ fuel component. The research method is numerical simulation in the MCU-MSR software developed by the National Research Center “Kurchatov Institute”. Calculation methods include simulation of radiation transfer in three-dimensional systems using the method of Monte Carlo and nuclide kinetics with quasi-continuous correction of material compositions.

Results and discussion

The concept of the reactor, its main technical parameters, and design features are presented. Calculations confirm the possibility of effective Cm transmutation. In contrast to Np and Am, high fission cross-sections of ²⁴³Cm, ²⁴⁵Cm, and ²⁴⁷Cm ensure a high rate of their transmutation without the accumulation of secondary long-lived nuclei. The reactor requires no additional Pu or other fissile isotopes introduced to the Cm transmutation cycle. The presence of light nuclei shifts the spectrum to the thermal region optimal for Cm transmutation. Homogeneous salt simplifies fuel composition control and fission product removal. The equilibrium content of key and all fuel isotopes in the fuel circuit is demonstrated to be reached within the first 3 and 15 years of reactor operation, respectively. Over 50 years of operation, ~39 t of Cm are loaded into the MSR‑B circuit: 95% is transmuted with < 1% (~1.9 t) remainder of secondary long-lived actinides.

Conclusion

A molten salt reactor based on LiF–BeF₂ ensures highly efficient transmutation of Cm with minimal accumulation of secondary actinides. A qualitative analysis shows that one ~600 MW(t) MSR-Cm specialized reactor is sufficient to transmute the forecast amount of Cm accumulated by the Russian nuclear power industry until 2100.

Background

Dry cooling of the solid blanket designed by the NIKIET JSC for a tokamak fusion-fission hybrid reactor can significantly facilitate design and technological implementation of the shutdown mode. However, this requires a calculated confirmation of (non)exceedance for the safe operation limits of the blanket with dry cooling in the shutdown mode.

Aim

To determine the safe conditions for reloading the source material blanket of a hybrid fusion-fission reactor in a shutdown mode.

Materials and methods

The considered solid blanket options include uranium and thorium source materials, as well as heavy and light water coolant under the most conservative conditions. The blanket is simulated as a subcritical system irradiated with 14.1 MeV fusion neutrons. The calculations are performed using the MCU-BR software with the MDBBR50 nuclear database. Studies of dry cooling conditions in the shutdown mode include the calculation of decay heat and analysis of its behavior, as well as thermophysical simulation of dry cooling for the blanket module based on calculated neutron-physical characteristics.

Results

The power of a thorium blanket during the irradiation period is 4–5 times less than of uranium one. However, the difference in decay heat between the two blankets is insignificant due to the considerable contribution of ²³⁹Np and ²³³Pa nuclides, taking into account the rapid decline in the contribution of ²³⁹Np and long-term influence of ²³³Pa contribution. Neutron-physical calculations show the decay heat of all considered blanket options comparable during 10 to 100 days of cooling. During dry cooling in the shutdown mode, the maximum temperature values of the source material in the solid blanket module and its most important structures quickly (in ~2 h) reach and then significantly exceed the operating limits requiring forced coolant circulation.

Conclusion

Dry cooling during reloading of the solid blanket in a hybrid fusion-fission reactor with both uranium and thorium source materials is demonstrated unacceptable due to the rapid increase in the temperature of the source material and structure to a maximum level significantly exceeding the limits of safe operation. Taking into account the obtained results, systems for reloading operations will be developed at the next design stages.

Background

One of the challenges of fuel integrity analysis during reactor operation is significant uncertainties when leaking fuel burnup is estimated by ¹³⁴Cs and ¹³⁷Cs activities during spiking events. For better radiation safety and lower financial losses, advanced methods of fuel integrity analysis are required.

Aim

To develop an advanced technique for detecting leaking fuel assemblies (FAs) in the core of VVER reactors using the features of ¹³⁴Cs and ¹³⁷Cs accumulation depending on the fuel type and irradiation history of each fuel rod.

Materials and methods

Axial distributions of ¹³⁴Cs and ¹³⁷Cs in each fuel rod are calculated for the entire history of fuel assembly operation, taking into account the dependence of the ¹³⁴Cs production on the neutron spectrum. The spectrum is sensitive to the fuel enrichment and burnup, gadolinia content in the Gd-fuel rods, position of the fuel rod in the fuel assembly, and characteristics of the nearest fuel assemblies. In this regard, ¹³⁴Cs/¹³⁷Cs activity ratio as a function of fuel burnup for different fuel rods differs and changes every time the fuel assembly layout in the reactor core changes from campaign to campaign. The initial data for calculations are standard output files of the KASKAD software package for each campaign. The calculated concentration of ¹³⁴Cs and ¹³⁷Cs in each fuel rod is compared with the activity measured during the spike effects. The developed software automatically selects fuel rods with coincinding calculated and measured values.

Results

A new method for identifying leaking fuel assemblies of the VVER reactor is developed. The CAESAR software is developed for automated calculation of cesium concentration in every fuel rod in the core. Validation on NPP data shows the proposed technique significantly more effective than the standard method for assessment of leaking fuel burnup.

Conclusion

The developed technique eliminates most of the uncertainties that lead to significant errors in assessing the fuel burnup of leaking fuel assemblies using the standard method specified for VVER reactors. The technique also helps to optimize the sequence of fuel testing in the casks of spent fuel pool in order to find quickly the leaking fuel assembly and reduce the total duration of the reactor outage.

Background

The Proryv project direction assumes the simulation modeling (SM) of process boxes with an inert atmosphere particularly important for the development of a digital twin for the module of fabrication-refabrication of mixed uranium-plutonium nitride (MUPN) fuel due to the pyrophoricity of the latter and high requirements for the accuracy of gas environment control.

Aim

To develop a mathematical model that reproduces the dynamics of parameters of the gas environment in the process box of the fabrication-refabrication module, as well as the operation of control and emergency protection systems.

Materials and methods

Key requirements for the model include: SM of emergency operating modes in case of box depressurization, valve failure, reduction of filter efficiency; SM of automatic control and emergency protection systems; SM of dynamic changes in the parameters of the gas environment inside the process box; calculation of fluid dynamics parameters in pipeline systems. The mathematical model is based on the equations of material and heat balance, state of ideal gas, as well as on fluid dynamics calculations, including submodels of in-box gas environment, hydraulic network, valves, sensors, and controllers.

Results

The developed mathematical model was tested in the KT-Nimfa software package for the following scenarios: normal box operation (Ar flow rate of 3.2–6 m³/h); emergency mode (500 ppm О₂ concentration of the inlet gas). The relative error in the time to reach the operating mode for chambers with a volume of 38.5 and 102 m³ is 8.3 and 6.5%, respectively. The protections in the emergency mode of more than 50 ppm O₂ in the box were triggered correctly. The model was used in digitalization tasks under the Prioritet-2030 program of the National Research Tomsk Polytechnic University.

Conclusion

The simulation mathematical model is supposed to be used for optimizing the operating parameters of the MUPN fuel pellet pressing plant, debugging and modernizing control and protection algorithms, as well as training personnel. The model needs to be verified and validated on experimental data and adapted for other plants.

Background

Keeping radioactive iodine within the containment in the event of accidents at NPPs with VVER reactors is one of the important tasks that needs to be solved. The key factor affecting the volatility of iodine is the pH of the aqueous medium inside the containment. According to EUR safety requirements, the pH should exceed 7.0 to suppress volatile forms of iodine. The lack of domestic calculation models for assessing pH dynamics during accidents significantly complicates the design process.

Aim

To develop and validate a model for calculating the pH values of the aqueous medium in the containment during design basis and beyond design basis accidents at NPPs with VVER-1200 reactors.

Materials and methods

The paper considers two scenarios with an initial event of a double-ended guillotine rupture of the main circulation circuit: a design basis accident without damage to the core and beyond design basis accident with significant damage to the core and failure of the active part of the emergency cooling system. The studied object is the pH of the aqueous medium in the containment. The research method is numerical simulation based on the developed mathematical model and software code for solving the equations of chemical equilibrium between components in aqueous solutions. Experimental methods for validation of the mathematical model involve potentiometric pH measurements in aqueous solutions with known concentrations of components at 25 °C.

Results

A mathematical model for calculating pH values of the aqueous medium in the containment is developed and validated. Calculations show that during a design basis accident, pH ranges from 4.2 to 8.0; the maximum pH values are reached by the 75th minute. Ten hours after the onset of the beyond design basis accident, the pH drops to 3.2 due to the formation of nitric and hydrochloric acids. After the sprinkler system is turned on, the pH rises to 7.8 and remains at this level for up to 30 days after the accident. Due to design solutions of introducing additional alkali starting from the 40th minute, the pH becomes higher than 7.0 and remains in the range from 8.0 to 8.4 until the sprinkler system is turned on.

Conclusion

The adequacy of the proposed model is confirmed by experimental data. The performed verification of the model shows the engineering solutions adopted at NPPs with VVER-1200 ensuring compliance with the requirements for binding radioactive iodine in design basis and beyond design basis accidents.

Background

Transient processes occurring at operating VVER reactors contribute to the overall damage to equipment and pipelines. The damage is determined by stresses calculated using indirect sensor data at control points, yet with assumptions and a large margin. Thus, the development of more advanced methods for determining stresses is of great interest.

Aim

To use a neural network for determining stresses at arbitrary points of equipment and pipelines based on readings from external thermocouples and pressure in the coolant circuit.

Materials and methods

The present study applies a neural network approach to solve the inverse problem of thermoelasticity. The developed neural network establishes a relationship between measured and predicted values. The stress range is selected as the criterion for calculation accuracy.

Results

We determined the stress values based on the readings of external surface thermocouples at control points for a number of pipelines in the primary circuit of the VVER reactor plant and branch pipe connection to the main circulation pipeline. The difference between predicted values of stress ranges and those obtained in solving the direct problem falls within 10%. The main approaches to increasing the stability of the resulting solution with insufficient quality of input data are considered.

Conclusion

Neural networks of simple configuration can be used in the monitoring system, since they quickly and quite accurately calculate the stresses in the pipelines of the VVER reactor plant. The proposed approach has great potential for further development and application at NPPs. The work for determining the optimal hyperparameters of the neural network should be carried out to improve its predictive ability.