Kola NPP Research Articles

Comprehensive assessment of system efficiency and competitiveness of nuclear power plants in combination with hydrogen complex

The strategy provides construction and commissioning of a number of new nuclear power units for the development of nuclear energy in Russia. The share of nuclear power plants increase in the energy systems of Russia is predicted from 19 to 22% in the future, up to 2050. Nuclear power plants planned to involve in the primary frequency control at the same time. All these circumstances exacerbate the problem of providing nuclear power plants with a basic electrical load in the night period, including during the daily period. The energy strategy of Russia provides for the production of hydrogen by low-carbon methods, one of which is water electrolysis using nuclear power. Hydrogen production is included in the development strategy of the at operating Russian NPPs. Hydrogen production planned at the Kola NPP by water electrolysis. Thus, the article provides a rationale for the effectiveness of combining nuclear power plants with a hydrogen complex based on the production of hydrogen by electrolysis of water. The effectiveness substantiated of the new principle of combination with overheating of the working fluid steam turbine cycle of the NPP taking into account the safety of handling hydrogen. A new system proposed for the combustion of hydrogen in oxygen, which makes it possible to overheat the working fluid of the NPP steam turbine cycle with undissociated steam, which significantly reduces the content of unreacted hydrogen in the working fluid flow. In addition, a system was developed and proposed for removing unreacted hydrogen and oxygen from the steam phase of the working fluid of the NPP steam turbine cycle. Thermodynamic and technical-economic new estimates are presented and analyzed of the efficiency of combining NPP with a hydrogen complex.

Prospects for the development of hydrogen technologies in the domestic metallurgy

The inevitability of the development of carbon-free steel production technologies in the domestic metallurgy is due not only under the pressure of environmental requirements for decarbonization, but also by the need to bring steelmaking technologies in line with the level of modern science. The electronic theory of oxidation/reduction can serve as a scientific basis for new technologies for the reduction of metals. From the standpoint of this theory, the final stage of the electrochemical process of reducing metals with any reducing agent, including hydrogen, is the electrolysis of oxides with the release of metal and gaseous oxygen under the action of an electron flow generated in the fuel cell during the combustion of the reducing agent due to the oxygen of the oxides. Comparison of two technology options ‒ iron reduction by “green” hydrogen or direct electrolysis of ore oxides using electricity from the electrical network shows a multiple advantage of electrolysis in terms of energy costs. It is concluded that the priority in the development strategy of the industry should not be given to the production and use of “green” hydrogen, but to the electrolysis of ore. The use of hydrogen as a reducing agent can be justified in the selective extraction of iron from complex ores in units such as plasma shaft furnaces, plasma reactors or fluidized bed furnace, in which, along with reduction, nitriding would also take place to convert soft iron into steel. In Russia, there is already an opportunity to master the production of steel using “green” or “turquoise” hydrogen. It is most rational to start the introduction of hydrogen reduction of iron on the basis of the deposits of titanomagnetites of the Kola Peninsula using the excess capacity of the Kola NPP to produce hydrogen

Biogeochemical Migration and Accumulation of Chemical Elements by Plants in the Aquatic System of the Discharge Channel of the Kola NPP

Biogeochemical Migration and Accumulation of Chemical Elements by Plants in the Aquatic System of the Discharge Channel of the Kola NPP

Роль Кольской АЭС в энергетике Мурманской области.

The determining value of the Kola NPP in the structure of the Kola electric power system is shown. The mutually complementary operation of the 1,760 MW KAES with seventeen hydroelectric power stations with a capacity of about 1,600 MW and two 240 MW thermal power plants makes the region's power supply system unique, producing the cheapest electricity in the North-West of Russia. The issues related to the further extension of the operating life of the existing nuclear power plant and the construction of theKola NPP-2 in the future are considered.

Simulating the Operation of the Jet-Vortex Condenser Used in the VVER-440 Reactor Plant’s Confinement System

A jet-vortex condenser (JVC) used as part of the confinement system (CS) serves for limiting the pressure in sealed rooms to the maximum admissible value for which the civil structures of NPPs equipped with VVER-440 reactors made according to the V-230 and V-179 projects are designed under the loss of coolant accidents' conditions. Fitting the Kola NPP’s first stage equipped with the VVER-440 reactor (made according to the V-230 project) in combination with taking certain other measures opened the possibility to extend the NPP service life. To substantiate the possibility of extending the service life of NPP power units, their safety will be assessed using an analysis model. The article presents the results from such modeling aimed at analyzing the performance of the jet-vortex condenser that is used as part of the confinement system for limiting the pressure in the sealed rooms down to the maximum admissible value for which the VVER-440 reactor-based NPP civil structures are designed with due regard to loss of coolant accidents (LOCA). The jet-vortex condenser operation model incorporated into the KUPOL-M computer code was verified against the experimental data obtained on the test facility installed at the All-Russian Research Institute for Nuclear Power Plants Operation (VNIIAES). A comparative analysis has shown that the calculated and experimental data are in fairly good agreement with each other. The KUPOL-M code with the jet-vortex condenser model integrated into it was used for calculating the parameters of medium in the system of sealed rooms in modeling the large break LOCA and small break LOCA conditions for the Kola NPP’s first stage, the distinctive feature of which is that it contains a JVC in the confinement area. The calculation results have confirmed the JVC efficiency under the conditions of a wide range of loss of coolant accidents. At present, the JVC has been put into use in the Kola NPP units 1 and 2, in the Novovoronezh NPP units 3 and 4, and in the Kozloduy NPP units 3 and 4 in Bulgaria.

Effect of Microadding Amine Surfactant on Erosion-Corrosion Wear of Structural Steel in NPP

Corrosion processes, specifically, erosion-corrosion, are one of the causes of damage occurring in NPP equipment during operation. Measures are taken to prevent erosion-corrosion wear of thermal-power equipment and pipelines. One such measure is the development and adoption of water chemistry regimes that are capable of forming a protective fi lm on a metal surface. Octadecylamine is one such reagent. The results of laboratory tests and industrial experiments studying the effect of octadecylamine on the development of erosion-corrosion wear of NPP equipment are presented. The octadecylamine-hydrazine water chemistry regime at octadecylamine concentration 0.5 mg/dm3 reduces the rate of erosion-corrosion wear by about 15-fold. The positive effect of micro-additions of octadecylamine is confirmed by direct erosion-corrosion studies performed at the Kola NPP.

Оценка состояния лесных фитоценозов в 30-км зоне Кольской АЭС по наземным и спутниковым данным

Оценка состояния лесных фитоценозов в 30-км зоне Кольской АЭС по наземным и спутниковым данным

Uncertainties in the Fluence Determination in the Surveillance Samples of VVER-440

The reactor pressure vessel (RPV) represents one of the most important safety components in a nuclear power plant. Therefore, surveillance specimen (SS) programs for the RPV material exist to deliver a reliable assessment of RPV residual lifetime. This report will present neutron fluence calculations for SS. These calculations were carried out by the codes TRAMO [1] and DORT [2]. This study was accompanied by ex-vessel neutron dosimetry experiments at Kola NPP. The main neutron activation monitoring reactions were 54 Fe(n,p)54 Mn and 58 Ni(n,p)58 Co. Good agreement was found between the deterministic and stochastic calculation results and between the calculations and the ex-vessel measurements. The different influences on the monitors were studied. In order to exclude the possible healing effects of the samples due to excessive temperatures, the heat release in the surveillance specimens was determined based on the calculated gamma fluences. Under comparatively realistic conditions, the heat increased by 6 K.

DIATOM COMPLEXES OF THE IMANDRA LAKE SEDIMENTS IN THE AREA AFFECTED BY HEATED WATER FROM THE KOLA NUCLEAR POWER PLANT

Diatom assemblages in Lake Imandra sediments (Kola Peninsula) were investigated in the area affected by heated discharges from the Kola NPP. Diatoms responses to the environmental changes were analyzed; the main stages in the lake ecosystem development were revealed; the water characteristics (pH, saprobity) were reconstructed by diatoms models. It is shown that the influence of thermal pollution has been the most conspicuous at the present stage of the lake development. The consequences of climatic changes since the Little Ice Age were revealed.

VVER-440 fuel cycle improvement: status and prospects

The results of development work on the base five-year fuel cycles with second-generation cassettes and average enrichment 4.87% for NPP with VVER-440 are presented. Such power units make it possible to increase thermal power while preserving an acceptable run time and RK-2+ cassettes with an enlarged uranium load for reactor operation at elevated thermal power 107, 109, and 112%. The results of trial commercial operation of third-generation fuel in the No. 4 unit at the Kola NPP in the 23rd load are presented. Increasing fuel enrichment above 5% is a promising direction for improving the VVER-440 fuel cycle. The results of development work on a full six-year fuel cycle with some cassettes kept for the seventh year of operation, which satisfies the operating conditions at elevated thermal power using RK-2+ fuel assemblies with >5% enrichment, are presented.

Experimental Studies for the VVER-440/213 Bubble Condenser System for Kola NPP at the Integral Test Facility BC V-213

In the frame of Tacis Project R2.01/99, which was running from 2003 to 2005, the bubble condenser system of Kola NPP (unit 3) was qualified at the integral test facility BC V-213. Three LB LOCA tests, two MSLB tests, and one SB LOCA test were performed. The appropriate test scenarios for BC V-213 test facility, modeling accidents in the Kola NPP unit 3, were determined with pretest calculations. Analysis of test results has shown that calculated initial conditions and test scenarios were properly reproduced in the tests. The detailed posttest analysis of the tests performed at BC V-213 test facility was aimed to validate the COCOSYS code for the calculation of thermohydraulic processes in the hermetic compartments and bubble condenser. After that the validated COCOSYS code was applied to NPP calculations for Kola NPP (unit 3). Results of Tacis R2.01/99 Project confirmed the bubble condenser functionality during large and small break LOCAs and MSLB accidents. Maximum loads were reached in the LB LOCA case. No condensation oscillations were observed.

Development and trials of a technology for reprocessing of NPP liquid radioactive wastes

Laboratory trials of a process for sorption decontamination of real NPP bottom residues from Cs radionuclides with Termoksid-35 ferrocyanide sorbent and rig and pilot trials of the technology for reprocessing of bottom residues from a number of Russian nuclear power plants were performed. The efficiency of the Cs sorption drastically decreases in the presence of organic complexing agents (EDTA, oxalate ions). Preliminary ozonation of the bottom residue allows the working life of Termoksid-35 sorbent in the decontamination process to be increased by a factor of several tens. Rig and pilot trials showed that ozonation of bottom residues, followed by microfiltration separation of the precipitate formed, ensures efficient decontamination of the solutions from Co, Mn, and Sr radionuclides and from α-emitters. A technology was suggested for decontamination of NPP bottom residues from radionuclides, involving ozonation, microfiltration, and selective sorption of Cs radionuclides on Termoksid-35 ferrocyanide sorbent. Process and technical solutions obtained in the course of trials were implemented in an industrial installation for reprocessing of bottom residues from the Kola NPP.

The effect of thermal flow of large power facilities on zooplankton community under subarctic conditions

Subarctic water bodies are extremely vulnerable with respect to changes in the environmental characteristics, including an increase in temperature. The thermal effect of spent waters of power plants in water bodies of the Kola Peninsula was found to cause changes in communities comparable, within certain limits, with the effect of eutrophication. The most informative indices, characterizing the response of zooplankton community to the thermal impact of spent waters of the Kola NPP are determined for the case of Lake Imandra, the Kola Peninsula. The qualitative and quantitative characteristics of zooplankton are shown to have some specific features depending on the extent of the thermal pollution of the water body and to adequately reflect the difference between the conditions of zooplankton existence in its individual parts. The indicator role of zooplankton is evaluated, and the possibility to incorporate it in the monitoring system of water bodies is substantiated.

Analysis of phase composition of corrosion-products deposits on the surfaces of a PG-440 steam-generator tube bundle by Mossbauer spectroscopy method

Possibilities of using Mossbauer spectroscopy for the analysis of the phase composition of corrosion deposits at the outer and inner surfaces of the tube bundle of a PG-440 steam generator of one of the Kola NPP power units are shown.

Assessment of the WWER-440/V-213 reactor condition

The current state of materials (steel, welds, anticorrosive cladding metal) of WWER-440/V-213 RPVs has been evaluated with respect to the peculiarities of their manufacturing process and operation for 30 years. The possibility of operating these reactors after the design service life has been assessed considering the degradation of properties due to thermal and radiation embrittlement, as well as damage from cyclic loads. Taking into account that the main results of operating effects were obtained in research laboratories it is necessary first of all to continue the investigations in the commercial reactors. In future, it will be possible to extend the service life of this type of reactors. For example, the positive results have been already obtained for unit No. 3 Kola NPP reactor.

Consolidation of scientific and technological expertise to assess the reliability of reactor pressure vessel embrittlement prediction in particular for the arctic area plant (COBRA)

The evaluation and prognosis of reactor pressure vessel (RPV) material embrittlement in WWERs and the allowable period of their safe operation are performed on the basis of impact test results of irradiated surveillance specimens. The main problem concerns the irradiation conditions (irradiation temperature, neutron flux and neutron spectrum) of the surveillance specimens that have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV wall condition. In particular, the key issue is the possible difference between the irradiation temperature of the surveillance specimens and the actual RPV wall temperature. It is recognized that the direct measurement of the irradiation temperature by thermocouples during reactor operation is the only way to obtain reliable information. In addition, the neutron field's parameters in the surveillance specimens location have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level. The COBRA project, which started in August 2000 and had a duration of 3 years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contained state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during reactor operation. The selected reactor for the experiment was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings, the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5 ± 4 °C. Uncertainties were evaluated also with experimental work carried out in the WWRSZ research reactor and by finite element modelling of surveillance capsules. The results obtained show that there is not need to perform temperature correction when surveillance data of irradiated specimens are used for embrittlement assessment of WWER-440(213) reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals ∼2.7 × 10 12 cm −2 s −1 with E > 0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%.

Threat to Norway from potential accidents at the Kola nuclear power plant. Climatological trajectory analysis and episode studies

Following the experiences after the Chernobyl accident in 1986, Norwegian Authorities regard the effects from accidental releases at nuclear installations in neighboring countries to be among the greatest environmental threats in the coming years. One of these nuclear installations is the Kola Nuclear Power Plant (Kola NPP). The unsatisfactory safety at the Kola NPP has been of major concern and a `Norwegian Plan of Action for Nuclear Safety’ has been worked out ( Ministry of Foreign Affairs, 1995. Plan of action for follow-up activities to Report no. 34 to Norwegian parliament (1993–1994)). As a response to this plan, DNMI has been involved in a project called: `Consequence Analysis of Potential Accidents at the Kola Nuclear Power Plant'. DNMI's part of the project consisted of analyzing the atmospheric transport and deposition pattern resulting from potential accidents at the Kola NPP. Results based on two different methodologies are presented in this paper. (1) Trajectory analysis as a tool for describing the air pollution transport pattern and screening of a large set of meteorological data for the selection of weather situations suitable for episode studies. (2) Episode studies using DNMI's dispersion model `Severe Nuclear Accident Program’ (SNAP) for the selected episodes.

The Kolanet International Project: Quick-Response System on Radiation Accidents for the Kola Peninsula

According to the international project Kolanet dealing with information defining the ecological problems in the Barents Euro-Arctic Region, one of the main tasks is the elaboration of the information system for quick response to radiation accidents on the Kola Peninsula and in the counties of the Northern Fennoscandia, The concept of the system includes: a radioactive monitoring net; telecommunications system; database of nuclear risk objects; and real-time prediction of the radiation situation and possible consequences using mathematical models with GIS technology. The first step in the development of this project is discussed, using the Kola NPP as an example.

The experimental definition of the acoustic standing waves series shapes, formed in the coolant of the primary circuit of VVER-440 type reactor

On the basis of pressure fluctuation measurements in some primary circuit loops at 2 nd Unit of Kola NPP with VVER-440 type reactors, the shapes of acoustic standing waves (ASW) were determined at frequencies corresponding to four minimal oscillation eigenfrequencies in the primary circuit coolant. On identification of the ASW modes and properties, experimental results based on six circulating loops in symmetric arrangement allowed determination of the three-dimensional space structure of the wave nodes and antinodes inside and outside of the reactor vessel (RV). As part of this analysis, the geometric features of the primary circuit that caused the formation of these standing waves were identified. Differences in each ASW shape were shown to cause different individual effects on the neutron field in the reactor core and on fuel assembly vibration. This has been partially confirmed by ex-core neutron ionization chamber noise analysis. One type of ASW, possessing an antinode inside the RV, can be used for measurement of the pressure coefficient of reactivity. However, this must be done with care to avoid the potential for incorrect results in some cases. The results presented in this paper can be readily extended to other VVER type reactors with both odd and even number of loops.