Atomic Energy最新文献

The PIK research reactor includes circuits of heavy-water reflector and liquid regulation representing systems important for nuclear safety. Therefore, the issue of organizing the optimal water-chemical mode and chemical control of coolants for these two systems is relevant. The volume of chemical control for coolants of heavy water circuits was developed taking into account existing experience and the physicochemical characteristics of heavy water. In order to return heavy water to the heavy-water reflector circuit and reduce the dose load on laboratory personnel, a unique automated system of sampling and analysis was created for online monitoring of heavy water in the heavy-water reflector of the PIK reactor.

An assessment of the radioecological state of water bodies on the territory of the uranium mining enterprise represented by the decommissioned Almaz Scientific and Production Association (LPO Almaz) uranium mining plant located in the town of Lermontov and in the adjacent residential area was carried out. Water bodies represent a critical factor for characterization of the radiation factor in terms of its impact on the population. After measuring the content of natural radionuclides in water and bottom sediments, the current radioecological state of water bodies as a whole was assessed. The content of natural radionuclides in water was estimated by multiplying the excess of the intervention level for each radionuclide present in a significant amount; for several radionuclides, the summarized ratios of specific activity to the corresponding level of intervention were used. The content of natural radionuclides in bottom sediments was estimated according to the effective specific activity of solid materials. A conservative average annual effective dose potential for the population in the residential area due to water bodies was calculated. Radium-226 was identified as the main radionuclide forming a dose load of internal exposure due to water bodies of the residential area in the immediate vicinity of the decommissioned LPO Almaz plant. Based on this radionuclide, an individual potential radiation risk to public health was calculated for recreational and agricultural modes of water use.

Results of experiments to test the modes of plutonium and neptunium displacement re-extraction with a nitrate solution of uranium (VI) are presented. The developed method of plutonium displacement re-extraction was tested at the refining extraction and crystallization facility of JSC SChC as part of the comprehensive program “Development of equipment, technologies, and scientific research in the field of nuclear energy use in the Russian Federation”. Due to a high accumulation of plutonium in the fuel of fast reactors, the main task of plutonium re-extraction consists in ensuring the re-extract ratio of Pu / (Pu + U) specified by the manufacturers of oxide fuel. According to the results of the performed tests, the developed method of plutonium displacement re-extraction with a solution of uranyl (VI) nitrate allows uranium, plutonium, and neptunium to be re-extracted in the ratio specified by fuel manufacturers. In this case, the completeness of plutonium extraction into the re-extract and its purification according to the proposed method are comparable to the results obtained in the process of reductive plutonium re-extraction.

An algorithm for estimating tritium emission during the evaporation of water from spray cooling ponds of nuclear power plants is considered. The factors determining the intensity of evaporation and intake of tritium into the atmospheric air are analyzed. For the considered meteorological and technological factors of area source operation, the intensity of tritium emission averages 6.2 Bq/h from each 1·m² of the spray cooling pond area per each 1 Bq/L of specific water activity. The possibility of regulating the emission activity is demonstrated taking into account the data of controlled technological parameters for the water-cooling system of spray cooling ponds and the meteorological parameters of the atmosphere in the NPP location area.

The paper presents the results of validating the severe accident block of the EVKLID/V2 integral code used to calculate the processes of fission product release from the oxide fuel melt and dissociation of nitride fuel observed during the destruction of the core in a fast neutron reactor with liquid metal cooling. Based on the obtained results, the uncertainty of calculating individual parameters, including the fraction of released fission products and the loss of the fuel mass during dissociation, is presented.

The article presents a system for measuring the level and density of liquids in nuclear-safe devices, which is based on intelligent sensors and differential pressure measurements at several points of the process medium using capillary impulse lines. The presented system, which consists of a probe, differential pressure sensors, and a separating medium supply system, is operable when both air and liquid are used as a separating medium. The results of testing the system using a compressed air separating medium are presented.

The article describes a set of studies for estimating and substantiating the limit mass of graphite removed during the resource characteristic management of RBMK reactors. The limit mass is understood as the maximum mass of graphite that can be removed from the reactor core at the stages of resource characteristic management without violating the operational limits of neutron-physical characteristics during the operation of power units over a period of 45 years. The substantiation of the limit mass was made based on the detailed computational simulation of reactor operation for realistic operational scenarios of the third and fourth power units of the Leningrad NPP. Particular attention is paid to the change in the void coefficient for the predicted states of reactors. The uncertainty of the limit mass of removed graphite is analyzed. The trueness of the calculated predictive estimates of neutron-physical characteristics and the limit mass of removed graphite is confirmed by the measurement results. It is shown that, by taking into account the considered operational scenarios for the reactors of the third and fourth power units of the Leningrad NPP, nuclear safety is ensured.

The article presents an economic and analytical model that evaluates various criteria for the effectiveness of investments in spent nuclear fuel reprocessing plants at given capital, operating, and decommissioning costs, as well as expected project implementation deadlines. The model also solves the inverse problem of determining the requirements for plant parameters based on the criteria of investment effectiveness. The results of the calculation of the most important of these parameters, including the internal rate of return, the discounted payback period, and the levelized cost of reprocessing spent nuclear fuel, are presented. The mathematical definition of the net present value, representing the main criterion, depends on the plant capacity, which varies from 200 to 3000 t of fuel per year. The estimated conditions under which the net present value of reprocessing spent nuclear fuel does not exceed $1000/kg are presented.

The paper describes a methodology for calculating break probability and the applicability of a leak-before-break (LBB) concept for the secondary circuit pipelines at nuclear power plants with VVER reactors. As well as a calculation based on the LBB concept, an example of break elimination using probabilistic approaches is presented. The study was carried out on the feedwater and main steam pipelines of an AES-2006 project. As well as taking into account the known defectiveness, data setting used a calculation method in the absence of such information. An example of calculating secondary circuit pipelines using experimentally obtained defectiveness data is presented. The leak probability for various nominal crack diameters is provided taking into account the detection of cracks during non-destructive testing, as well as cases of non-detection.

The article proposes the use of permanent additional low-power steam turbines to back up the auxiliary units of nuclear power plants for removing decay heat from the reactor core. A study was carried out to develop a method of plant-wide multi-channel backup of nuclear power plant auxiliaries based on low-power steam turbines for a multi-unit station. In order to improve the efficiency of plant-wide auxiliary backup for a nuclear power plant having more than two power units, a method for controlled maintenance of the required reactor power at a lack of decay heat was developed based on the use of permanently operating low-power turbines for each power unit. Conditions for the removal of decay heat from one, two, and four reactors, which are demonstrated in the event of the nuclear power plant being disconnected from the power system, are based on the decay heat of one reactor and one low-power steam turbine. The study proved the possibility of creating a multi-channel plant-wide backup of nuclear power plant auxiliaries based on the equipping of each power unit with a low-power steam turbine.