VVER-1000 Reactor Research Articles

The long-lived radionuclides radiotoxicity ranking the nuclear fuel cycle of the Belarusian NPP

Using a methodical approach to the potential biological hazard (radiotoxicity) assessment of radionuclides, computational studies and a comparative analysis of the contributions to the total radiotoxicity of long-lived fission products and actinides accumulated in the nuclear fuel cycle of the Belarusian NPP were carried out. In order to estimate correctly the 237Np isotope mass at the long-term stage of SNF management, the mathematical model of the nuclear fuel cycle of the Belarusian NPP of the code CUB was modified to take into account the transmutation chain of 241Pu, 241Am and 237Np isotopes. The initial concentrations for the main radiation-hazardous radionuclides at the time of fuel unloading from the VVER1200 reactor core are presented in the database “Radiation Characteristics of Spent Nuclear Fuel of the Belarusian NPP” developed by the authors. The database development is carried out in accordance with the recommendations of the RB-093-20 “Radiation and thermophysical characteristics of spent nuclear fuel of pressurized water power reactors and high-power channel reactors”. The main characteristics used in the nuclear fuel cycle simulation in the CUB code, such as the total mass of spent nuclear fuel (2492 tons), the number of spent fuel assemblies (5294) and the average burnup, are taken in accordance with the Spent Fuel Management Strategy of the Belarusian NPP data. It is shown that at the stage up to 100 years, the main contributors to the total radiotoxicity are the isotopes of 238Pu, 239Pu, 240Pu, 243Am, 241Pu, 241Am, 237Np, where contributions of three last of them are 0.07, 49.0 and 0.007%, respectively. But 5000 years later, the contribution of 241Am will be 0.24%, and the isotope 237Np – up to 0.1% of the total actinides radiotoxicity. The results obtained can be used to substantiate for the recovery degree requirements of 237Np, 241Pu, 241Am during the SNF of the Belarusian NPP reprocessing.

Validation and application of the Dragon5 lattice code for neutronics and burnup analysis of VVER-1000 pin cell and assembly model

This work deals with the validation and assessment of neutronics and burnup analysis capacity of the deterministic lattice code Dragon5 for VVER-1000 hexagonal pin and assembly model configuration. Various VVER-1000 hexagonal pin-cell benchmark cases including the fresh uranium fuel and mixed oxide fuel of ORNL are simulated and compared with the HELIOS and OpenMC code. The data library differences of certain nuclides will lead to large deviations of the results. Accurate results can be presented by the Dragon5 lattice code when compared with OpenMC code. The OECD LEU and MOX assembly benchmarks are computed to further validate the prediction ability for VVER-1000 assembly, reasonable agreement between Dragon5 solutions and published values of other codes is observed for burnup parameters including infinite multiplication factors, concentrations of important actinides and fission product nuclides. Moreover, the post-irradiation of UO2 sample experiment of the Russian Kalinin-1 VVER-1000 reactor is simulated and compared with laboratory measured values to provide the experiment validation of the Dragon5 lattice code. The predicted concentrations of actinides up to Cm244 are compared with HELIOS and Serpent Monte Carlo code. The results show that the Dragon5 lattice code presents better prediction accuracy for Calculated-to-Experimental ratios (C/E's) of most actinide concentrations than the HELIOS and Serpent code. Therefore, the École Polytechnique de Montréal Dragon5 lattice physics code can be applied to the neutronics and burnup analysis of VVER-1000 hexagonal pin-cell and assembly model.

A study of VVER-1000 fuel rod integrity during LOFA

The safety of the fuel rod is of great importance in all operational states, especially during accidents. This paper is the third part of a three-part study aiming at developing the coupled neutronic/thermal-hydraulic codes to simulate a full 3-D model for VVER-1000 assembly and predict the essential safety parameters. In our previous works, three scenarios for loss of flow accident (LOFA) were studied by developing a simulation model for the thermal analysis of a fuel rod with the ANSYS-FLUENT17.2 code. The present research is aimed at studying the mechanical behaviour of the VVER-1000 reactor fuel rod under the worst LOFA conditions. To perform this analysis, the thermal-hydraulic parameters with and without coupling transfer from the ANSYS-FLUENT17.2 thermal-fluid module are used in the Static Structure (Mechanical) module. The value and the location of the maximum mechanical stress and elongation of the cladding during the accident are demonstrated, as well as the time when the clad critical temperature occurs. Through the mechanical module, the mechanical behaviour of the fuel clad is studied at several axial locations in the fuel, to account for power oscillations. As the power of the reactor increases, the fuel rod suffers from stress and elongation caused by thermal expansion. Also, the melting of the fuel rod starts after 80 s.

Thermal hydraulic analysis of VVER spent fuels stored in vault dry system under different operating and design conditions

Abstract The spent nuclear fuel discharged from power reactors is a very important problem facing the future of using power reactors in electricity production. This paper focuses on the thermal-hydraulic behaviour of the VVER spent fuel in the vault dry storage system under forced convection mode, which is experimentally and numerically investigated. For this purpose, a test rig is designed and constrained to simulate the cooling loop vault system that contains four spent fuel assemblies discharged from the VVER reactor, which are represented by four electric heaters. A numerical simulation is performed by the ANSYS-CFX fluid dynamics code. The effects of decay heat generation and inlet air velocity are investigated as an operating condition. Also, the effect of the type of the Vault System tube material is being studied. The results show that the increase in the inlet air velocity improves the coolability of the fuel, while the increase in decay heat leads to a decrease in the coolability of the fuel. The used velocity range is (0.1 < V < 0.5 m/s) for inlet coolant air and heater power (20 < P < 100 W). Three tube materials (aluminum, copper, and stainless steel) were evaluated for mechanical properties, including thermal conductivity, to assess the feasibility of their use as tubes in the spent fuel storage.

Vitrification of a Simulator of Vat Residues from Liquid Radioactive Waste

The study on the posibility of the use of the optimal glass compositions for vitrification of an imitator of vat residues of liquid radioactive waste from nuclear power plants with VVER-1000 reactors was carried out. The main process parameters such as vitrification temperature, strength, corrosion resistance, absence of crystalline phases, minimization of glass-forming additives and inclusion the maximum amount of waste were analyzed. It has been established that the melting temperature of lead-borosilicate glass matrices was 1150 °C, which satisfies the requirements for vitrification of low- and medium-level waste. The ultimate compressive strength of the obtained samples of glass matrices was 136.0 MPa. In addition, it has been shown that lead-borosilicate glass matrices are the most resistant to leaching. The cesium leaching rate was 1.5·10-5 g/cm2·day.

Effects of Evaluated Nuclear Data Libraries on the Calculation Results for Fuel Burnup with Minor Actinides in a VVER Reactor

Effects of Evaluated Nuclear Data Libraries on the Calculation Results for Fuel Burnup with Minor Actinides in a VVER Reactor

Evaluation of Neutronic Performance for the VVER-1000 Reactor Core with Regenerated Uranium-Plutonium Fuel

Evaluation of Neutronic Performance for the VVER-1000 Reactor Core with Regenerated Uranium-Plutonium Fuel

DISTRIBUTION OF RADIOACTIVE ELEMENTS DURING SLAG REMELTING OF STRUCTURAL MATERIALS OF VVER FUEL ELEMENT SHELLS

At the radiochemical plant of FSUE “PO “Mayak” in the process of processing spent nuclear fuel (SNF), about 170 tons/year of metal radioactive waste (MRW) are formed, mainly represented by fuel element shells structural materials (SM) containing residual amounts of radionuclides after fuel dissolution, and spent fuel assemblies (SFAS) sent to storage. It is possible to achieve compaction and decontamination of MRW by a method based on slag remelting in an induction furnace with a cold crucible. In order to establish the distribution of actinides and fission products (FP), thermodynamic slag remelting process modeling of the VVER-1000 reactor plant SM fuel rods shells and SFAS, experimental data obtained verification were carried out. The most probable distribution of actinides and FP by melting products is shown. Most of the curium and americium are concentrated in the metal – 99 and 94 wt %, respectively. The maximum extraction of uranium into the slag phase in the form of UO2 dioxide is up to 40 wt %. The distribution of plutonium in the temperature range of 1500–2000°C over the slag (in the form of PuO and PuO1.61 oxides) and metal phases occurs in almost equal proportions. It has been established that up to 99.78 wt % cesium, the content of europium and americium is 0.05 and 0.17 wt %, respectively.

Searching for optimal accident tolerant fuel for the VVER-1200 reactor from the neutronic point of view

ABSTRACT In an attempt to find an optimum fuel, which can withstand the harsh operation condition in the VVER-1200 reactor, different fuel types were investigated in this work. Uranium carbide and uranium nitride are proposed as alternative fuels to UO2. MCNPX 2.7 with cross-section library ENDF/B-VII was used to design 3D model of VVER-1200 to investigate the neutronic properties of the suggested fuel types. The output results of the burnup of the proposed fuels have been analyzed to determine which of them achieves the longest fuel cycle. It was found that UO2 and U12C achieved criticality till 35 GWD/MTHM, while U14N and U15N achieved criticality at 30 GWD/MTHM and 24 GWD/MTHM, respectively. The traditional method of reactivity management has been tested on the proposed fuel to determine to what extent it is valid for them. To ensure the operation safety of the proposed fuel, some safety factors such as moderator temperature coefficient (MTC‎), fuel temperature coefficient (FTC‎), and effective delayed neutron fraction (βeff) have been calculated. The distribution of the thermal neutron flux and normalized power has been displayed to give us a good insight view of the VVER-1200 assembly to determine which of the proposed fuels is optimum compared to UO2.

Neutronic performance of VVER-1200 reactor with advanced fuel and cladding materials

Nuclear reactor operation and safety are connected with a nuclear fuel design and its material and geometry properties. An increase in nuclear safety is therefore based on a neutronic and thermal analysis of the materials used in a core. This paper analyzes neutronic properties of a VVER-1200 fuel assembly with alternate fuel and cladding materials that consider Accident Tolerant or Advanced Technology Fuels (ATF). The ATF nuclear fuel development started with the post-Fukushima efforts and its main purpose is a reduction of the high-temperature hydrogen production from the zirconium water reaction. Five ATF cladding candidates with reference Zr-based alloy are studied in this paper as well as six ATF fuel candidates together with an additional standard UO2 reference case. The results are analysed based on the multiplication characteristics, reactivity feedbacks, and spent nuclear fuel composition.

Radiation-induced structure of austenitic steels with different nickel content under neutron irradiation in SM-3 and BOR-60 reactors

Comparative studies of the radiation-induced structure of austenitic steels with a nickel content of 10, 20 and 25 wt.%, irradiated sequentially in the SM-3 and BOR-60 reactors, as well as to higher damaging doses in the BOR-60 reactor, have been carried out. The phase composition, dislocation structure, pores, and radiation-induced segregations at grain boundaries were studied by high-resolution analytical methods of transmission electron microscopy, scanning electron microscopy, and atomic probe tomography. The formation of radiation-induced phase precipitates based on nickel has been established, and its volume fraction correlates with the level of radiation-induced segregations, and increases, the higher the nickel content in the steel. The values of barrier strength factors for radiation-induced structural elements in the studied steels are adjusted by calculation and experiment, which makes it possible to determine their contribution to radiation hardening.It is shown that the largest contribution to radiation hardening as a result of neutron irradiation in BOR-60 at high irradiation temperature up to 29 dpa is made by large radiation-induced precipitates of (G + γ') phases. It is shown that with an increase in the damaging dose, the main factor limiting the performance of internal devices will be radiation swelling, since the contribution to the change in properties from radiation-induced phases and radiation defects will not increase due to their density reaching saturation. Steel with 25 wt.% Ni exhibits the lowest level of swelling at high radiation doses, which makes it possible to consider it as a material-candidate for internals for promising VVER reactors with higher temperatures and longer service life.

Thermal-hydraulic analysis of the noise effects in super critical water reactor by using the porous media approach in the frequency domain

Noise analysis in a supercritical water reactor (SCWR) is considered. This type of reactor is one of the generation IV reactors in which supercritical water is used as a coolant. Since the thermodynamic properties of water at supercritical pressure near the pseudo-critical point is quite sensitive to any perturbation in temperature, one would be interested in having a precise stability analysis in this region.In this research, the effects of coolant temperature fluctuations at the inlet of a fuel assembly were investigated by using the porous media approach (PMA) for a typical SCWR reactor.This analysis is performed using the Fourier transform of the mass, momentum, and energy conservation equations in the frequency domain. The transformed equations were discretized with the finite volume method and numerically solved. Noise calculation was performed for two different frequencies of 0.1 and 1 Hz.The comparison between the achieved results of the applied model and previous studies performed with sub-channel analysis in a typical SCWR (in case of steady-state condition) as well as a VVER-1000 (in case of noise condition) reactors, showed a good agreement.The porous media approach noise calculation code showed that any increment in coolant temperature would result in severe changes in thermal–hydraulics parameters.

Comparative analysis of the results of simulation of the consequences of a serious accident at a Belarusian NPP with the use of J-RODOS and RECASSEXPRESS software complexes

Using the J-RODOS and RECASS-EXPRESS software packages, modeling of the radiological and radiation-ecological consequences of a severe beyond design basis accident at the power unit of the Belarusian NPP with a VVER-1200 reactor was carried out. The simulation results showed that if radionuclides were released from the reactor containment at the level of 2.15·1015 Bq, the maximum values of the fallout from the atmosphere to the surface 1 day after the accident will be 2.98∙105 Bq/m2 for I-131 (J-RODOS) and 7∙105 Bq/m2 (RECASS-EXPRESS), and for Cs-137 − 1.11∙104 Bq/m2 and 1.3∙105 Bq/m2, respectively. The difference in the values of the radionuclide fallout, modeled by these software packages, increases with the distance from the release source and at a distance of more than 10 km reaches 1 order of magnitude. The predicted values of the equivalent thyroid dose for adults and children for the first 7 days after the accident, obtained using J-RODOS calculations, are 3–4 times higher than using RECASS-EXPRESS software product. At a distance of 3–5 km from the NPP site, doses ranged from 3.7 to 14.8 mSv for adults and 16.2 to 32.5 mSv for children. The values of the total effective dose with the J-RODOS simulation are 2 times higher than those with RECASS-EXPRESS at a distance of 3 km from the site (0.60 and 0.32 mSv, respectively) and 1.5 times higher at a distance of 5 km from the site (0.30 and 0.21 mSv respectively). As the distance from the NPP increases, the convergence in the obtained evaluation results of total effective dose increases. None of the obtained prediction results exceeds the established emergency response criteria and do not indicate the need for protective measures for the public.

Study of burnup process in the VVER-1200 reactor with plutonium refueling options

Study of burnup process in the VVER-1200 reactor with plutonium refueling options

Radiological characteristics of spent nuclear fuel from small modular reactors under consideration for deployment in Ukraine

Small modular reactors represent a promising technology for power generation, offering solutions to the energy crisis and mitigating greenhouse gas emissions. As Ukraine considers the deployment of NuScale, UK SMR, and SMR-160, it is crucial to address the safe management of spent nuclear fuel. This study focuses on evaluating the radiological characteristics of spent nuclear fuel from the selected small modular reactors and for comparison, from the VVER-1000 reactor. Using the Monte Carlo code Serpent, depletion calculations were performed for an assembly in an infinite 2-D geometry, and the activity, decay heat, and inhalation toxicity of the spent nuclear fuel were assessed. We determined the main nuclides contributing to the radiological characteristics and quantified the mass content of these nuclides. The total number of spent nuclear fuel assemblies produced during the entire life of each small modular reactor type was estimated. The radiological characteristics assessed for the three small modular reactors do not exceed those observed for VVER-1000 reactors currently operating in Ukraine. So, spent nuclear fuel generated by the selected small modular reactors will introduce no new challenges to Ukraine's radioactive waste management system. The results of this work provide valuable insights for identifying the optimal small modular reactor technologies for Ukraine concerning safe spent nuclear fuel management.

Development of a Risk Monitoring Model for NPPs with a VVER-440 Reactor

Development of a Risk Monitoring Model for NPPs with a VVER-440 Reactor

Loading Pattern Optimization of Vver-1000 Reactor Core Based on the Discrete Golden Eagle Optimization Algorithm

Loading Pattern Optimization of Vver-1000 Reactor Core Based on the Discrete Golden Eagle Optimization Algorithm

The concept of a fast reactor with liquid metal fuel in tungsten capsules

The concept of a dual-purpose high-power nuclear reactor is proposed. One of the goals is the production of electricity, the other is the production of high-potential thermal energy. It is proposed to use liquid fuel based on waste uranium and plutonium extracted from the spent fuel of VVER reactors (purified from the 238Pu isotope). The fuel is in sealed tungsten capsules. Lead extracted from thorium ores is used to cool the reactor. The electrical power of the reactor is 3.3 GW. The layout of the reactor is identical to the BREST-OD-300 reactor under construction. The analysis of emergency modes from among ATWS (anticipated transients without scram) is carried out. The reactor is reliable and safe. The maximum temperature of a high-temperature reactor coolant is close to the boiling point of lead. By the nature of the change in the maximum temperatures of the core components, the reactor occupies an intermediate position between a reactor with solid metallic fuel and a reactor with cermets based on UN-PuN and metal uranium nanopowder.

Preliminary design of soluble boron free small modular reactor core based on hexagonal lattice

This study made an effort to design a novel high-power density SMR core based on a hexagonal lattice for the soluble boron free environment. The VVER-1000 reactor core was converted to an SMR core with a “halved scale”. The core design process is divided into two parts; Part I: assembly-level analysis and Part II: whole-core analysis. Important design parameters such as core dimensions, number, and arrangement of fuel assemblies, fuel enrichment, appropriate burnable poison, and placement of burnable poison pins have been studied in Part I. Also, by designing an optimized burnable poison pattern, the excess reactivity and acceptable radial pin power distribution were treated without soluble boron. WIMS, and PARCS codes have been investigated for fuel assembly and whole core analysis. Finding the most appropriate fuel-loading pattern involved an iterative manual optimization between the core loading pattern and FA design to assess the acceptable neutronics and thermal-hydraulics core features and satisfy the safety margins.The design of high-power density core requires accurate prediction of the thermal–hydraulic parameters to ensure safe operation. In Part II coupled 3-D neutronic/thermal–hydraulic analysis has been done to identify the main safety challenges of the developed core. To access the proper discharge burn-up, an initial enrichment of 6.6% U235 was determined for the developed core. The developed core maximum discharge burn-up is equal to ∼49 Mwd/KgU at full power operation. Radial-zoning burnable poison is investigated in the seek of a proper pattern to limit power peaking.

Оценка доз облучения населения от жидких сбросов Белорусской АЭС в режиме нормальной эксплуатации

Radiation safety of the population from liquid discharges of the Belarusian NPP (hereinafter - BelNPP) into surface water bodies is considered to be ensured if the exposure dose to the population due to the discharges does not exceed the value of the Dose Constraint (hereinafter - DC) established by the sanitary norms and rules for ensuring radiation safety: 50 gSv/year. If there are prerequisites for the possibility of reducing the specified DC standard, it can be revised by decision of the state sanitary supervision institutions based on estimates of annual effective doses to the population (hereinafter -referred to as YED) for a certain period of NPP operation. The purpose of the study was to consider the rightfulness of the established value of the DC standard by comparing it with the calculated values of the YED of the population from the design and predictive actual liquid discharges of the BelNPP after the commissioning of the plant. As model parameters for predictive estimates, data on actual liquid discharges from NPP - analogues of the BelNPP with a VVER-1200 reactor were used. The article presents the results of the assessment and analysis of the YED from the design and predictive actual liquid discharges of the BelNPP through all exposure pathways. It is shown that the predictive estimates of the actual YED of the population are comparable with the design YED from BelNPP discharges into the Viliya River and do not exceed 22 % of the corresponding DC standard. The remaining reserve of 78 % is sufficient to reliably ensure compliance with the DC standard for public exposure from liquid radioactive discharges of the BelNPP, including after the commissioning of the second power unit. At least 80 % of the YED of the population from liquid discharges of the BelNPP into the Viliya River will be formed due to internal irradiation, to a large extent from the consumption of fish caught in the river. External exposure of the population from liquid discharges of the BelNPP occurs mainly when staying on the beach, its contribution to the total YED from discharges does not exceed 20 %. The list of the main dose-forming radionuclides was determined and the necessity of mandatory control of the volumetric concentration of radionuclides 131I, 60Co, 58Co, 54Mn, 90Sr in waste waters of the BelNPP was substantiated.