Atomic Energy最新文献

Background

In nuclear medicine, radiopharmaceutical drugs (RPDs) containing the technetium isotope Tc-99m are used to treat oncological diseases, such as brain, thyroid, and salivary gland cancer, as well as for diagnostic studies of the cardiovascular system. These RPDs are created using a molybdenum-technetium Tc-99m generator.

Aim

To determine the main parameters of complex protection for a designed transport container of a technetium generator for the production of RPDs in accordance with the GMP standard.

Materials and methods

Eighteen container material options were considered. To assess the protective characteristics of the container, we used the MicroShield v. 8.01 software. Transport indices and categories of all container options were determined taking into account the calculated rate of equivalent doses for gamma radiation from the Mo-99/Tc-99m sorption column and the nominal source activity.

Results

We have simulated an advanced protective container for the Tc-99m generator. The best protective characteristics are noted for the option combining tungsten and lead 2.5 and 3.5 cm thick, respectively.

Conclusion

This complex design of the protective container ensures the safe operation and transportation of the Tc-99m generator.

Background

Issues of hydrogen explosion safety are extremely relevant. Search for solutions requires both experimental and computational methods.

Aim

To experimentally study the turbulent combustion of hydrogen-air mixtures and to test the CABARET-COMBUSTION CFD calculation code using the obtained data.

Materials and methods

The study includes experiments conducted in a facility representing a large diameter pipe, as well as numerical simulation of deflagration combustion.

Results

The calculated data on the flame front propagation velocity and pressure dynamics at the shock wave front are consistent with experimental results.

Conclusion

The obtained results indicate the potential of using the CABARET-COMBUSTION CFD code in the numerical solution of hydrogen-air mixture combustion problems. This code and its testing on data obtained with high-quality diagnostics will increase the predictive capabilities of supercomputer simulation for the analysis of hypothetical accidents at a qualitatively new level.

Background

The total pollution of the ecosphere with anthropogenic tritium, which is many times higher than its natural level, is an urgent strategic task.

Aim

To assess the contribution of nuclear power plants (NPPs) to the global tritium release into the ecosphere depending on the type of the NPP reactor.

Materials and methods

We have collected and analyzed more than 6 thousand records on specific and absolute indicators for annual emissions and discharges of tritium entering the environment from NPPs of all reactor types and from other sources worldwide. The statistical analysis of the data was carried out using Microsoft Excel and Statistica software packages.

Results

We have determined the statistical characteristics of specific tritium emissions and discharges for all types of operating reactors. The highest specific emission and discharge is noted for NPPs with PHWR and AGR reactors, respectively; the lowest values are characteristic of RBMK NPPs, including EGP ones. During normal operation of NPPs, discharges cause 61% of the tritium entering the ecosphere. PHWR NPPs make the largest contribution of more than 40 and 99% to global ecosphere discharges and emissions, respectively. As compared to other reactor types, BN breeder reactors insignificantly affect the tritium pollution of the ecosphere.

Conclusion

The accumulated activity of tritium from emissions and discharges during normal operation of the considered NPPs for the period of 1970–2022 is less than 10% of the natural tritium level and less than 2% of the level caused by fission and fusion nuclear weapons tests.

Background

Nuclear power waste can be reduced by replacing depleted uranium with spent nuclear fuel (SNF) from high-power channel (RBMK) reactors in the fuel compositions of fast neutron reactors.

Aim

To determine the possibility of using RBMK SNF, processed through a simplified reprocessing technique, as a feedstock for fast reactors.

Materials and methods

We have simulated a fuel campaign for a modernized 1200 MW BN-1200M sodium-cooled fast neutron reactor fueled with U‑Pu nitride fuel. Physical calculations were carried out for scenarios with both homogeneous and heterogeneous layouts of the fast reactor core; the layouts are planned to be used at the stage 1 and stages 2, 3 of operation, respectively.

Results

The results of physical calculations are presented for a scenario with a simplified technology of RBMK SNF reprocessing for the first active loading and makeup fuel: the maximum reactivity margin for the burnup of the core fuel has increased by 0.3 and 0.4% for a homogeneous and heterogeneous layout, respectively. The equivalent dose rate (EDR) of photon radiation from the fresh fuel assembly similarly increases by ~250 and ~40 times for a 30-day and 3‑year storage, respectively. The EDR of fission products for a fresh assembly decreases to 70 and 50 μSv/h at a purification factor of 1·10⁴ and its increased value, respectively.

Conclusion

Reprocessed RBMK SNF used instead of depleted uranium fully realizes the energy potential of natural uranium, as well as reduces nuclear power waste and load on SNF storage facilities.

Background: As a key task of nuclear power industry, transmutation raises the problem of radiation equivalence and utilization of minor actinides (MAs). The conversion of MAs into fission products appears to be the best way to solve it.

Aim: To develop criteria for comparing the efficiency of transmutation in different reactor plants (RPs) without the deep analysis of nuclear power development scenarios and numerous assumptions regarding external parameters.

Materials and methods: The selected criteria included the absolute mass balance and number of fissions, as well as those taking into account the operation of the RP as part of nuclear power industry. Lists of unfavorable MAs such as long-lived high-level waste (LLHW) have been compiled for the developed transmutation efficiency methodology. The selected criteria and LLHW lists were validated by simulating the first campaign of a lead-cooled fast neutron reactor (FR) and a VVER-1200 water-water power reactor with the addition of MAs.

Results and discussion: We propose to normalize the amount of transmuted material to the thermal power in the core in order to adequately compare the transmutation efficiency of different reactors. This approach has revealed and confirmed the efficiency of transmutation in FRs, both homo- and heterogeneously loaded, to be significantly higher than that for VVER-1200 reactors. We recommend to determine the mass balance using the LLHW list including all MAs and ²³⁸ Pu.

Conclusion: To evaluate the transmutation efficiency by the selected RPs, we propose to use the criterion of the difference between the masses of loaded and unloaded actinides, as well as to normalize the amount of the transmuted material to the thermal power of the core.

Background: Increasing the service life of pumping collector equipment at uranium isotope separation plants and the degree of exhaust gas neutralization, as well as reducing the mass of radioactive waste represent urgent tasks for the isotopes separation industry.

Aim: To perform a comparative study on the efficiency of neutralizing gaseous fluoride ion under low pressure conditions for a number of chemical absorbents used in pumping collectors of uranium isotope separation plants.

Materials and methods: The studied sources of gaseous F‑ion include HF and UF₆ as the main and impurity one (< 0.3 vol.%), respectively. The tested chemical absorbents of the K‑09-01 pump collector are Ca(OH)₂, Ca₂(OH)₂CO₃, CaCO₃, Al₂O₃, Na₂CO₃, Mg(HCO₃)₂, as well as vinylpyridine cationite, mercerized wood with NaOH, and modified mercerized wood admixed with Na₂SO₃.

Results: Under pumping collector conditions of t = 25 ± 3 °C and 0.4 < P < 1.3 kPa, the degree of absorption for F‑ and U‑containing gas components by lime and mercerized wood absorbents has reached 99.7 and almost 100%, respectively; other absorbents have proved themselves ineffective. Acting as a neutron moderator due to carbon, spent mercerized wood absorbents have been established to contain a small amount of uranium.

Conclusion: Mercerized wood and lime chemical absorbents demonstrate the highest absorption efficiency for UF₆ and HF fluorine-containing gaseous compounds. Thus, they can be used in pumping collectors of uranium isotope separation plants. Spent mercerized wood absorbents cannot be attributed to nuclear hazardous objects and should be classified as radioactive waste.

Background: Existing schemes for the output and transmission of large electric energy flows from sources to consumers over long distances entail significant energy losses during transmission, economic costs during the construction of step-up and step-down substations, as well as the alienation of large territories.

Aim: To develop alternative highly efficient technologies for the output of large energy flows and their transmission from remote generation facilities using superconducting cable lines increasing their efficiency, reliability, and environmental friendliness.

Materials and methods: The paper reviews completed, ongoing, and planned Russian and foreign long-distance power transmission projects using high-temperature superconductors (HTSs). A prototype of a generator current lead phase with HTSs has been developed. Experimental studies conducted at the D.V. Efremov Institute of Electrophysical Equipment JSC jointly with the ROSSETI Scientific and Technical Center JSC have confirmed the possibility and feasibility of developing a full-scale HTS project for the Leningrad NPP.

Results: The performed theoretical assessment of general approaches to the development of HTS cable lines and the analysis of the experiment have demonstrated the superconductivity effect appropriate to achieve the transmitted power level of several gigawatts per circuit and simultaneously reduce energy losses to 3%. Calculations have confirmed that the cryogenic cable line system can be indefinitely extended under high-quality cooling ensuring a virtually unlimited length of HTS lines.

Conclusion: The accumulated experience is sufficient to start the implemention of pilot projects using HTS cable lines. However, further research is needed for taking into account technical, infrastructural, and economic factors.

Background

The assessment of the external radiation dose to reference herbage based on dose conversion factors assumes uniform distribution of radionuclides along the vertical soil profile. In fact, this often conflicts with reality and leads to incorrect assessments of radiation exposure.

Aim

To evaluate the external radiation doses to reference herbage in four options of radionuclide distribution along the vertical soil profile using different calculation methods; to select the most relevant calculation method.

Materials and methods

The study assesses external radiation doses to reference herbage due to γ‑radiation of ⁵⁴Mn, ⁶⁰Co, ¹⁰⁶Ru¹⁰⁶Rh, ¹³⁴Cs, ¹³⁷Cs^137mBa, ¹⁴⁴Ce¹⁴⁴Pr, and ¹⁵²Eu in four options of soil radionuclide distribution. The assessment methods are based on dose conversion factors, engineering calculations, and solution to radiation transfer equations.

Results

The results of calculating external radiation doses to reference herbage using different methods for the same geometries of the source and receiver of γ‑radiation are comparable. The assumption of a uniform radionuclide distribution along the soil depth with their actual predominance in the upper part of the vertical profile leads to an almost twofold underestimation of the calculated doses.

Conclusions

We recommend assessing the γ‑radiation impact of soil radionuclides on reference herbage using engineering methods taking into account the actual distribution of radionuclides by the depth of the soil profile.

Background: The operation of radioactive waste (RAW) pools is associated with potential environmental risks. Remediation of adjacent territories represents a priority in the Strategy for Environmental Safety of the Russian Federation.

Aim: To determine the characteristics of a biogeochemical anti-migration barrier created for the remediation of aquifers with complex contamination, as well as to assess the effectiveness of this barrier.

Materials and methods: The elemental composition of samples was determined by inductively coupled plasma mass spectrometry (ICP-MS); the method of capillary gel electrophoresis (CGE) was used to determine the ion concentration.

Results and discussion: Maximum concentrations of ammonium, sulfates, uranium, and nitrates in RAW filtration zones are 448, 1800, 4.9, and 10,000 mg/L, respectively. As a result of bioremediation, iron partially passes into sulfide phases, while the remained iron re-precipitates into hydroxide phases during bio- or reoxidation. Purification yielded a chemically active mineral sediment preventing the spread of U, Np, Pu, and Tc redox-sensitive radionuclides, as well as Sr and Am.

Conclusion: An effective and economically feasible approach to the purification of groundwater near nuclear fuel cycle facilities during their operation and post-mothballing periods has been tested. The tested method involves the in situ intensification of microbial processes by introducing soluble sources of organic carbon and phosphorus.

Background: A modernized floating power unit with the RITM-200S reactor was developed to provide electricity to the areas of decentralized energy supply. An advanced design of the cassette core with increased energy resource, reliability, and safety indicators requires final comprehensive justification.

Aim: To experimentally study the coolant hydrodynamics in a fuel rod bundle of a fuel assembly for further justification of characteristics for an advanced cassette core of the RITM-200S reactor.

Materials and methods: Developed by order of the Rosatom State Corporation, a scale model of a fuel rod bundle fragment was tested on an aerodynamic facility of the Nizhniy Novgorod State Technical University named after R.E. Alekseev. The flow structure was studied according to a pneumometric method using a five-channel sensor.

Results: In regular cells behind the spacer grid, the formation of transverse flows is recorded at a distance of L/d_h ≈ 1 from the plates. Regular cells align the structure of the axial flow at a distance of L/d_h ≈ 10 from the grid; the dimensionless velocity is 1.0–1.1. The lowest axial flow velocity is observed in corner cells with a section covered by plates; the dimensionless value axial of the flow velocity ranges 0.3–0.6 at a distance of L/d_h ≈ 1 from the plates. The velocity of the axial flow at the central displacer is higher than at the periphery; its dimensionless value at a distance of L/d_h ≈ 10 from the grid is 0.75–0.9.

Discussion: Three structural zones of the flow are identified in the area of regular cells, central displacer, and on the periphery near the fuel assembly jacket; the flow velocity in them differs by 25–30%. The arrangement of the grid plates has a significant effect on the flow structure at a distance exceeding L/d_h ≈ 10.

Conclusion: The identified flow characteristics should be taken into account for justifying the thermal reliability of newly developed cassette cores using the KANAL thermal-hydraulic code. The calculation methodology of the code should be changed by increasing the number of calculation cell types and taking into account the non-uniformity of the flow rate over the areas of the fuel rode bundle and cell types.