Atomic Energy最新文献

The article presents the results of numerical simulation and experimental study of conditions for igniting the gas discharge on the outer side of the cathode in a Penning ion source. The deposited plasma products of this discharge metallize the surface and decrease the electrical strength of the feedthrough insulator. We propose the options of magnetic systems preventing the ignition of a side discharge to increase the service life of the ion source. The obtained experimental data demonstrate the proposed magnetic systems maintaining the neutron flux generation modes during the operation of a miniature linear accelerator as part of a neutron generator.

The paper presents the results of a numerical study of mercury flowing downward in a vertical one-side heated round pipe under a transverse magnetic field. The problem was solved using the LES method in a wall-conjugate formulation by accounting for the physical properties of the pipe wall material and simulating a thin layer of possible contaminants on the pipe inner surface. The performed study revealed the electrical conductivity of the contaminant layer significantly effecting hydraulic resistance, heat exchange intensity, and structure of the flow, as well as the occurrence of quasi-periodic, abnormally high-amplitude temperature fluctuations of liquid and pipe wall. The fields of longitudinal velocity, temperature, and temperature fluctuation oscillograms are provided for various values of dimensionless contact resistance.

In the present paper, we use computational simulation to investigate the response of self-powered neutron detectors (SPNDs) to changes in the coolant flow rate of the measured fuel assembly (FA). A model problem of reducing the coolant flow rate in several measured FAs is considered. Their response is assessed, including the difference in the SPND current before and after introducing a disturbance caused by a decrease in the flow rate through the FA. In some cases, readings of in-core monitoring system SPNDs can be used to diagnose a decrease in the flow rate through the FA.

The paper presents the results of testing the operation of a reactor at the 1st power unit of the Leningrad NPP‑2 in the daily load schedule mode. The tests aimed to experimentally substantiate the operation of the VVER-1200 power unit in a daily load schedule, as well as to develop control methods and check the operation of standard equipment. Two alternative methods for controlling the reactor power density were tested in combination with soft temperature control of reactivity by changing the steam pressure of the secondary circuit. The operation of the boron control system modernized to automatically compensate for reactor xenon poisoning was tested and adjusted. The performed tests confirmed the ability of the VVER-1200 power unit to operate in a daily load schedule both at the beginning and end of the fuel campaign.

Background

As part of the Proryv project, the final stage of reprocessing spent nitride fuel involves the separation of americium and curium on a cation-exchange sorbent using displacement complexation chromatography. For radioecological safety of the environment, spent radioactive sorbent should be converted into a stable compound.

Aim

To test magnesium potassium phosphate (MPP) matrix for immobilization of spent radioactive sorbent formed during the reprocessing of spent nitride fuel, as well as to determine the quality indicators of the resulting MPP compound.

Materials and methods

The phase composition of obtained samples was determined by X‑ray diffractometry. Water resistance, mechanical compressive strength, resistance to thermal freeze-thaw cycles, as well as leaching of matrix-forming elements and ^{241, 243}Am and ²⁴⁴Cm isotopes was determined in accordance with standard tests.

Results and discussion

The obtained samples of MPP compound contain up to 20 wt % of the spent sorbent simulant and 10 wt % of wollastonite; their main crystalline phase represents an analogue of the K‑struvite natural mineral. The compressive strength of MPP compound is ~6 MPa, including upon the completion of freeze-thaw and water resistance tests after 90 days of immersion in water. The compound is demonstrated highly resistant to leaching of matrix-forming elements and isotopes: the leaching rate of ^{241, 243}Am and ²⁴⁴Cm is approximately 5.8∙10⁻⁷ and 2.5∙10⁻⁷ g/(cm²∙day), respectively.

Conclusion

The prospects of using MPP compound for immobilization of spent radioactive sorbent formed during reprocessing of spent nitride fuel are shown. Quality indicators of obtained samples meet the requirements for solidified radioactive waste.

Background

Dense nitride fuel is promising for reactors on fast neutrons (fast reactors) with liquid-metal coolant to solve the problems of closing the nuclear fuel cycle. An important aspect of research involves the study of the nitride fuel dissociation rate under various conditions, including that characteristic of severe accidents.

Aim

To numerically study the time of uranium nitride fuel dissociation under various external conditions characteristic of severe accidents with core destruction.

Materials and methods

The object of the study is uranium nitride. The research method is numerical simulation using the validated SAFR severe accident module of the EVKLID/V2 integral code developed by the IBRAE RAS, Afrikantov OKBM JSC, NIKIET JSC, and NRC Kurchatov Institute (Russian Federation). The SAFR module uses a model for calculating the behavior of nitride fuel elements in severe accidents. Computational methods of the SAFR module include enthalpy formulation to solve the thermal conductivity equation; the melt flow is simulated by solving one-dimensional equations of mass, energy, and momentum conservation.

Results

Depending on the properties of the contact environment, time of fuel dissociation may vary by several orders of magnitude. In a leaking fuel element with a helium atmosphere without nitrogen at a relatively high temperature of 1800 ℃, the dissociation time for 50% of uranium nitride ranges from 40 days to ~3 years, depending on the conditions of wetting the fuel column by liquid uranium. At high values, close to the nitride melting point of 2600 ℃, the dissociation time is from hundreds of seconds to 1 h. The presence of nitrogen with a partial pressure of ~2.5 atm completely suppresses dissociation up to melting.

Conclusion

The longest time of uranium nitride dissociation, and therefore its greatest thermochemical stability, is observed during dissociation into a gas atmosphere in the presence of a residual melt film on the fuel surface.

Background

Development of a two-component nuclear power system with a closed nuclear fuel cycle based on new-generation fast reactors represents an important task put forward in the Strategy for the Development of Nuclear Power in the Russian Federation until 2050.

Aim

To generalize the results of developments for the project of a new generation commercial power unit with a BN-1200M reactor and to substantiate innovative engineering solutions for its equipment and systems.

Materials and methods

The BN-1200M project uses both reference engineering solutions that have proven themselves well during the operation of BN-600 and BN-800 power units and advanced solutions for the core, main circulation pumps, steam generator, cold traps, reloading system, and layout of the primary and secondary circuit equipment. The current work is focused on the justification of mixed uranium-plutonium fuel, design of the emergency heat removal system, operational safety, manufacturing technology for structural elements of a two-module vessel steam generator, new structural materials, etc.

Results

An engineering design for a sodium-cooled BN-1200M fast neutron reactor has been developed; the competitiveness of the project has been substantiated; design materials for the power unit have been developed. The results of R&D on selecting emergency core protection using passive equipment in the emergency heat removal system for natural circulation along the circuits and justifying new structural materials in heat exchange equipment have increased the level of safety, reliability, and cost-effectiveness of the BN-1200M reactor as compared to previous projects and promising energy sources.

Conclusion

The completed work became a transition to the commercial development of the designed power unit with a high-power BN-1200M reactor plant in a two-component nuclear power system. This stage of commercial development solves the urgent task of completing the experimental validation of engineering solutions for equipment, including the development and testing of prototypes on sodium benches.

Background

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

Aim

To determine the analytical dependence of the relative quantitative change in the coolant flow rate through the FA on the pressure drop in the core based on the obtained primary empirical data from hydraulic tests on an FA model.

Materials and methods

A methodology for measuring the coolant flow rate through FAs is developed. Hydraulic tests of an FA model were performed to determine a formula for calculating the coolant flow rate through the FA in the core depending on its pressure drop.

Results

The developed methodology is appropriate for determining the coolant flow rate through the FA of the IVV-2M reactor. Measurements carried out on an FA model were used to obtain an empirical dependence of the change in the flow rate on the pressure drop in the core.

Conclusion

The considered methodology increases safety during reactor operation, as it allows the temperature at the outlet of the FA, and therefore, on the fuel element cladding to be predicted. The analytical formula of obtained dependencies can be used to calculate the coolant flow through the FA in the reactor core and to analyze the measurement results for different core configurations and coolant temperatures. The next article “Study of the coolant flow rate through the fuel assemblies of the IVV-2M reactor core. Part 2” will represent a description of the flow meter design and an analysis of flow rate measurements over several years.

Background

Process monitoring for the composition of spent nuclear fuel (SNF) is closely related to the risks of radiation exposure to personnel and formation of secondary radioactive waste. In addition, it requires labor-intensive sample preparation, as well as costly, long-term, and numerous measurements, making the transition to remote measurement methods, such as laser-induced breakdown spectrometry (LIBS), extremely important.

Aim

To determine the possibility of introducing the LIBS method into the system for chemical composition monitoring of hydrometallurgical reprocessing of SNF from the BREST-OD-300 reactor plant for an alternative replacement of existing physical and chemical methods including atomic emission spectrometry, spectrophotometry, and X-ray fluorescence analysis for measuring the content of U, Pu, and rare earth elements (REE).

Materials and methods

A review of literature sources was conducted. A model of the experimental laboratory LIBS unit was developed; test measurements of the element composition were made using the spectral characteristics of solid and liquid samples simulating oxidized SNF.

Results

The performed literature review showed the potential of using LIBS. The developed LIBS unit was effectively used to determine actinide atomic emission lines using U as an example and identify the presence of nuclear fuel fission products. The measurement error for the concentration of La, Nd, Ce, Fe was 10, 12, 4.5, and 13.5%, respectively, in solid samples of REE oxides with a concentration from 10³ to 10⁵ ppm. For Zr, the error is 3.3% in the samples of nitric acid solutions with a concentration of 0.9–15 g/dm³. Economic efficiency of using the LIBS method involves the reduction in the daily number of measurements and equipment costs by ~29 and 40%, respectively, at the duration of measurements decreased to 24 h.

Conclusion

The experimental results confirm the operability of the developed LIBS unit and possibility of determining uranium, plutonium, and REE, as well as measuring their content. Due to reduced risks of personnel exposure, the LIBS method can be used to replace a number of quantitative chemical analysis methods for conducting remote online radiation monitoring and decelerating the formation of secondary radioactive waste. In the future, the work on reducing detection limits and measurement errors is required.

Background

Technologies based on the principles of “green chemistry” are actively used around the world to treat contaminated areas. Field studies at radiochemical plants in the USA and China have demonstrated bioremediation potential for complex treatment of groundwater. The article presents the experience of applying this approach in the Russian Federation.

Aim

To study the mechanisms of immobilizing redox-sensitive radionuclides in the presence of an in situ biogeochemical barrier for treatment of groundwater with polyelement radioactive contamination at the territories of the Siberian Chemical Plant JSC (SCP) and Chepetsk Mechanical Plant JSC (CMP), as well as to justify optimal compositions for the activation of indigenous microflora.

Materials and methods

The elemental composition of samples was determined using inductively coupled plasma mass spectrometry (ICP-MS); the CGE method of capillary gel electrophoresis was used to determine the ion concentration. Various organic substrates and food production waste were used. Field experiments were conducted on the territories of the CMP (2020), as well as near the SCP B2 pool (2015) and Sublimate Plant (2018).

Results

Whey, molasses, and a mixture of sucrose and phosphates yielded less than 10% of desorbed actinides. Comparatively high desorption rates are characteristic of stillage, glucose, sucrose, and a mixture of microbial metabolites. The maximum effect is observed in distilled water with 3 g/L of nitrate ions: 33, 55, and 45% desorption for U, Pu, and Np, respectively.

Conclusion

Field tests have proven high efficiency of using food production waste for the biological treatment of radioactive areas. The biogeochemical barrier zone reduces colloidal and pseudocolloidal transfer of radionuclides due to the coagulation of iron and clay particles. Oxidized iron phases ensure long-term reliable immobilization of redox-sensitive radionuclides under oxidative conditions. The total cost of materials for treating a formation of more than 600 m³ with a single injection of organic mixtures amounted to about 25,000 RUB. The obtained results allow this approach to be recommended for large-scale use to treat the territories near radioactive waste storage reservoirs during operation and post-mothballing periods.