Water-Water Energetic Reactor Research Articles

ANALYSIS OF STRESS CORROSION CRACKING OF HETEROGENEOUS WELDED JOINTS IN SIMULATED PRIMARY WATER ENVIRONMENT

The paper deals with the stress corrosion cracking of heterogeneous welded joints. The welded joints were made of austenitic base metal (AISI 321) and ferritic weld metal by arc welding. Two methods were used to analyze the stress corrosion cracking of samples of this type of welded joint. The first method was a slow strain rate test (SSRT) using a tabletop device at a temperature of 60°C with a graded strain rate for three specimens (10−5, 10−6, and 10−7 1/s). The second method was exposure in a corrosion autoclave at a temperature of 270°C and pressure of 12.26 N/mm2, while prestressing devices were used to achieve the prestressing at the level of yield strength and 3% plastic deformation. One specimen was bent to 60°. Boric acid solutions were used as the medium, which was supposed to simulate the environment of the primary circuit of nuclear power plants, a type of water-water energetic reactor. The surface and the presence of corrosion products, cracks, fractures, or pits on specimens were monitored using scanning electron microscopy, stereomicroscopy, confocal microscopy, and light microscopy. A gravimetric analysis was performed, as well, to determine the corrosion rate after exposure to the autoclave.

Economic evaluation of hydrogen production in second and third units of Bushehr nuclear power plant regarding future need of nuclear fission technology

Today, the most important challenge for hydrogen production is the use of suitable energy source for sustainable production. The dual use of nuclear power plants allows them to expand into other markets by eliminating greenhouse gas emissions from processes such as hydrogen production. The significance of this matter especially in countries with ongoing new nuclear power plants is high and shows the future need for nuclear fission technology. In this paper, an economic assessment of hydrogen production in the second and third units of the Bushehr nuclear power plant (BNPP) with the Hydrogen Economy Evaluation Program (HEEP) is carried out and a comparison of CO2 emission in nuclear and non-nuclear methods of hydrogen production is presented. Most nuclear reactors, including Advanced Pressurized WaterReactor (APWR), High Temperature Gas Reactor (HTGR), and Water-Water EnergeticReactor (WWER), can be used as a source of thermal and electrical energy for the process of hydrogen production. In this research, hydrogen production processes based on the Conventional Electrolysis (CE), High Temperature Steam Electrolysis (HTSE), and Sulfur-Iodine (SI) Process using different storage and transportation options are compared to choose the cost-effective method for coupling with the WWER1000 reactor. Hydrogen production, thermal energy, and electricity costs have been calculated and explained in detail. The results showed that the cost-effective option for hydrogen production is using the High Temperature Steam Electrolysis (HTSE) method, Compressed Gas (CG) storage, and pipeline transportation, by 2.88 $/kg. Also, applying this method to produce hydrogen, the nuclear power plant thermal efficiency, and the hydrogen plant efficiency are kept in the best possible conditions by 51.98% and 58.21%, respectively.

Development and benchmarking of SIMULATE5-K-VVER

SIMULATE5-K-VVER (S5K-VVER) is the newest addition to the CMS5-VVER modeling suite developed by Studsvik Scandpower, Inc. (SSP). In S5K-VVER, the coupled time-dependent multigroup neutron diffusion equation, delayed neutron precursor concentrations, and thermal hydraulic models are solved in triangular nodal geometry to model transients in Water-Water Energetic Reactors (VVERs).The neutron kinetics equations in S5K-VVER are solved in a two-step process: temporal integration and spatial integration. In S5K-VVER there are two options for temporal integration: the Frequency Transform Method (FTM) and a Backward-Euler Method (BEM).S5K-VVER is then used to solve a series of benchmark problems including the AER-DYN-001 and AER-DYN-002 numerical transient benchmarks. Finally, the Kalinin-3 benchmark is modeled which describes the shutting-off of a single Main Coolant Pump (MCP) in the Kalinin-3 VVER-1000/320 nuclear power plant. Compared to plant data, the S5K-VVER solution agrees within 1.5% of core total power during this transient.

Advanced fuel and burnable absorbers designed for long-cycle operation of BNPP

The present study is based on the facts that are happening for the Bushehr Nuclear Power Plant (BNPP) and coincides with the operational changes and step-by-step modifications of the BNPP in the near future. Bushehr reactor core will change from BNPP core to TVS-2M core. In other words, a new fuel, called the TVS-2M, is being used by the Russians to upgrade WWER-1000 reactors to reduce power generation costs, increase cycle length and fuel consumption. The neutronic analysis performed in this work is in line with this change. It provides a feasibility study of BA change and increase of cycle length in BNPP reactor without changing the initial structure of the reactor core. This research provides a new core configuration that can achieve the stated goals to a large extent without changing the BNPP core to the TVS-2M core. This research provides a feasibility study of Integrated burnable absorber (IBA) change by maintaining the initial neutronic conditions and a new high-performance configuration aiming for a long-cycle operation in Iran's nuclear power reactor. In Water-Water Energetic Reactors (WWER) reactors, two alternatives to CrB2Al (current burnable absorber in BNPP) are gadolinia (Gd2O3-UO2), and erbia (Er2O3-UO2). The benefits of the combinational IBA concept are expected by compensating weak points of each other. The main contribution of this research is the use of two cylindrical coaxial IBA layers with the same volume of gadolinium pins plus erbium (separated by a thin layer of zirconium alloy (Zr + 1% Nb)) and gadolinium pins. A neutronic evaluation has been conducted to investigate the variations of effective multiplication factor versus burnup, reactivity swing, power peaking distribution, moderator temperature coefficient (MTC), critical boron concentration (CBC), and cycle length. The result shows that the cycle length can be extended from 296 days in the BNPP core to 366 days in the proposed core, which is comparable to the cycle length of the TVS-2M core. The average BA concentration in the proposed core is 5%, which is lower than the average BA concentration of 5%–8% in the TVS-2M core. Regarding the gain of 73.90 days in cycle length, the revenue is estimated to be around 19,503,000 USD, making it worth the implementation considering the number of gadolinium and erbium pins in the proposed core.

Analysis of the effect of pollution of component surfaces as a result of possible contact of fresh nuclear fuel with sea atmosphere on corrosion

Acquisition and analysis of the information was performed on possible effect of components of sea atmosphere in case of their precipitation on the assembly surfaces on the corrosion of structural materials of items of fuel assemblies (FAs) of water-water energetic reactor (VVER-1000) in their subsequent operation. The data is presented on static and dynamic tests of alloy E-110 at temperatures of 300–350 С in water and standard coolant of the reactor polluted with chlorides and iodine, as well as of alloy E-110 under special pollution of its surface with chlorides. It is demonstrated that in keeping the conditions of storage of fresh fuel, as well as the recommendations of the present work, its high corrosion resistance and operability in further operation is ensured.

Increasing the cavitation coefficient of injection of a jet apparatus for the implementation of environmentally friendly technologies

Ejector systems have proven themselves in various industries for their simplicity, reliability and environmental friendliness. To increase the efficiency of the emergency core cooling systems of innovative projects of nuclear power plants with water-water energetic reactor, the pump-ejector system can be used, but to achieve design indicators for maximum water flow and pressure, it is necessary to operate jet devices in cavitation mode. The article presents the results of bench tests of ejectors with single-barrel and multi-barrel nozzles, as well as their pressure characteristics. A recommendation is given on the use of ejectors with multi-barrel nozzles and an increased cavitation injection coefficient in emergency core cooling systems for nuclear reactors. Keywords: ejector; jet pump; emergency core cooling systems; nuclear power; environmentally friendly technologies.

Correlation between grain boundary strength determined by impact test of miniature specimen and stress corrosion cracking resistance of irradiated austenitic steels used for the internals of WWER-type and PWR-type nuclear reactors

Austenitic chromium-nickel 321, 316 and 304 stainless steels used for internals of PWR (pressurized water reactor) and WWER (water-water energetic reactor) reactors were irradiated to 4–125 dpa in different neutron energy spectra and investigated by impact bending tests of miniature specimens. Relying on these tests, the effects of neutron dose and neutron energy spectrum on the intergranular fracture energy determining grain boundary strength have been analyzed.Two neutron spectra were considered. One spectrum is typical of the core of light water moderated reactors PWR and WWER, another spectrum is typical of the core of fast neutron reactors (FR).A correlation between resistance to stress corrosion cracking (SCC) and grain boundary strength in irradiated chromium-nickel 321, 316 and 304 stainless steels has been found.

The method for calculating the probability of brittle destruction of NPP equipment in different operating modes with postulated defectiveness

The destruction of equipment metal by a brittle fracture mechanism is a probabilistic event at nuclear power plants (NPP). The calculation for resistance to brittle destruction is performed for NPP equipment exposed to neutron irradiation; for example, for a reactor plant such as a water-water energetic reactor (WWER), this is a reactor pressure vessel. The destruction of the reactor pressure vessel leads to a beyond design-basis accident, therefore, the determination of the probability of brittle destruction is an important task. The research method is probabilistic analysis of brittle destruction, which takes into account statistical data on residual defectiveness of equipment, experimental results of equipment fracture toughness and load for the main operating modes of NPP equipment. Residual defectiveness (a set of remaining defects in the equipment material that were not detected by non-destructive testing methods after manufacturing (operation), control and repair of the detected defects) is the most important characteristic of the equipment material that affects its strength and service life. A missed defect of a considerable size admitted into operation can reduce the bearing capacity and reduce the time of safe operation from the nominal design value down to zero; therefore, any forecast of the structure reliability without taking into account residual defectiveness will be incorrect. The application of the developed method is demonstrated on the example of an NPP reactor pressure vessel with a WWER-1000 reactor unit when using the maximum allowable operating loads, in the absence of load dispersion in different operating modes, and taking into account the actual values of the distributions of fracture toughness and residual defectiveness. The practical significance of the developed method lies in the possibility of obtaining values of the actual probability of destruction of NPP equipment in order to determine the reliability of equipment operation, as well as possible reliability margins for their subsequent optimization.

Аnalysis Catalytic Hydrogen Recombiner Capacity Calculation Taking into Account Conditions Inside Sealed Enclosure of Containment Safety System of Nuclear Power Plants with Water-Water Energetic Reactor

Localizing safety systems are provided to contain radioactive substances in an accident and attenuate ionizing radiation at a modern nuclear power plant. Together with radioactive substances, hydrogen is also retained, which is formed during the decomposition of the primary coolant. The accumulation of hydrogen in the presence of oxygen from the atmosphere in the accident localization zone carries the danger of the formation of flammable and explosive concentrations of these components. Nuclear power plant (NPP) deigns with water-water energetic reactor (WWER) provides for a hydrogen removal system including passive catalytic hydrogen recombiners. The device capacity is confirmed experimentally under reference conditions (lean air-hydrogen mixture, pressure and temperature close to normal, no interference with gas exchange). Capacity is an important safety parameter. In the event of an accident, conditions inside the ealed enclosure of the localizing system of NPP with WWER can differ from the reference ones and affect the capacity. On the basis of calculations, the operation of recombiners with lack of oxygen and with hindered gas exchange has been investigated in the paper. The decrease in capacity with lack of oxygen reaches 50 %, which is mainly caused by an increase in underburning. Compared to the reference conditions, the effect is more pronounced in the event of an accident – 60–70 %. The hindered gas exchange is modeled by a decrease in the height of recombiner traction channel. This case can be reduced to the placement of the device in cramped conditions and the effect of the atmosphere speed inside the enclosure. Regardless of the hydrogen concentration, the operating characteristic of the device remains linear, with a two-fold decrease in height leads to a decrease in capacity by 20 %. The results can be used to substantiate the safety of NPPs with WWER and to review on the safety subtantiation of power units.

Determining the critical concentration of boric acid and the time of its onset when reaching to minimum controllable power for the water-water energetic reactor

Start-up to minimum controllable power level (criticality approach) is one of the most hazardous nuclear operations during operation. In particular, the spontaneous and unauthorised start-up is very dangerous, and it occurs as a result of some technological operations or changes in technological regimes. Currently, there are codes for neutron-physical calculations at NPPs with VVER, such as reactor simulator (IR) and BIPR-7A. These codes calculate the boric acid critical concentration without relying on ex-core ionisation detectors data, which may result in inaccuracies in determining the critical concentration. In addition, feeding the primary circuit with clean condensate must be stopped at least 15 minutes before is reached, and these codes do not calculate the time to reach the critical state. As a result, the idea arose to develop a code that would predict the time to reach the critical state and the critical concentration of boric acid only using the measuring equipment readings without reliance on additional calculations.

Determining the critical concentration of boric acid and the time of its onset when reaching to minimum controllable power for the water-water energetic reactor

Determining the critical concentration of boric acid and the time of its onset when reaching to minimum controllable power for the water-water energetic reactor

Iterative reconstruction algorithm comparison using Poisson noise distributed sinogram data in passive gamma emission tomography

ABSTRACT Gamma emission tomography (GET) detects fuel rods that emit gamma rays for potential use as verification tools in nuclear safeguards. In GET, iterative reconstruction algorithms are often used to reconstruct passive gamma-ray emitter source distributions. Generally, using noisy sinogram data, the iterative reconstruction algorithms increase the noise level on a reconstruction image with iterations. Thus, it is important to evaluate how performances of representative iterative reconstruction algorithms change with noisy sinogram data. The passive gamma-ray source distributions inside the mock-up of the water-water energetic reactor (WWER) fuel assembly having missing fuel rods were reconstructed by using the following algorithms: gradient method, steepest descent method, conjugate gradient reconstruction method, algebraic reconstruction technique, simultaneous iterative reconstruction technique, and maximum likelihood expectation maximization (MLEM). Consequently, MLEM algorithm yielded a higher contrast reconstruction image and was regarded as higher reliable algorithm to discriminate the fuel rods from the passive gamma-ray emitter source distribution.

Investigation of the probability of existence of defects with a size exceeding the allowed value

Relevance. Non-destructive testing of metal determines the actual state of the metal, the presence of discontinuities and their sizes, and also allows to determine what mechanisms of metal degradation were subjected to. One of the main characteristics of the quality of non-destructive testing is the detectability of discontinuities and defects. If no defects were missed, then its possible to guarantee the reliable operation of the facility until the next scheduled inspection. The article is devoted to the study of the probability function of detecting defects and determining the probability of the existence of a residual defect with a size exceeding the permissible value. The aim of the work - to develop a method to determine the probability of the existence of a residual defect with a size exceeding the permissible value after non-destructive testing and repairs of equipment and pipelines of a nuclear power plant. Methods. During the work formulas for the probability of detecting a defect and initial defectiveness, regulatory requirements in the field of certification of flaw detectors, and the results of research on non-destructive testing were used. Results. A method for determining the probability of defects with a size exceeding the allowed value, using the example of a reactor vessel, is presented. The method is based on residual defects, which takes into account the detectability of defects. The value of the coefficient that takes into account the influence of the human factor, instrument and methodological shortcomings or complexity of access to the control point is determined, which reduces the degree of uncertainty in determining the residual defect. The results of this work permit to evaluate the probability of the existence of a defect with a size exceeding the allowed value. The development of a residual defect to critical values characterizes the initial event for the destruction of the integrity of the structure. Thus, the probability of a residual defect can be used when performing a safety analysis of the water-water energetic reactor vessel.

Statistical investigation of the weld joint efficiencies in the repaired WWER pressure vessel

From a fatigue point of view, the second most important WWER (Water-Water Energetic Reactor) pressure vessel component (after closure studs) are the primary nozzles. These parts are included in the in-service inspection performed every 4 years when non-destructive testing techniques like ultrasonic, eddy-current and dye-penetration methods are applied. After the determination of the severity of fatigue crack, the progress of initial defects over time can be prevented by repair welding. Purposedetermination of values of the eight main mechanical properties of the welded nozzle after major repair, evaluation of statistical data for reliability and recommendations for the maintenance. Study designthe examination of the nozzle Du-500 revealed the presence of perpendicular crack on the outside surface in the base metal. The crack in the base metal has an angle of 72° with the horizontal plane which was repaired at normal (T = 20 °C) and elevated (T = 350 °C) temperatures by (i) automatic arc welding under the flux layer AN-42, and (ii) ceramic flux. The analysis was performed with 15 different batches of welded steel specimens and every single parameter each time was established by using 45 repetitions of the test procedure for the same type of specimen. Findingsthe study describes the specific case of defected nozzle's housing that was made of steel 15Cr2MoVA and after the crack repair along the length of nozzle diameter could potentially be further exploited. Value – the analysis of the welded specimens showed good agreement between the empirical and theoretical curves and can be used in lifetime and fatigue life predictions for WWER pressure vessels still applicable in power generation industry.

Analysis of experimental data on the hydrodynamics in a horizontal steam generator using the STEG code

The new model of interfacial drag is proposed and implemented in the STEG code based on the 3D two-fluid model is presented. Experimental data obtained at the PGV test facility is used for validation of the modified STEG code. The PGV test facility comprises a model of the horizontal steam generator used for Russian type water-water energetic reactors. A quantitative comparison with experimental values of void fractions has been performed. It can be argued that the results generally indicate the relatively good predictive ability of the modified STEG code.

Fuzzy logic algorithms for classification and parameterization of eddy-current signals

Safety of nuclear power plants operation depends on steam generator safe operation. Russian nuclear power plants with water-water energetic reactors WWER-440 and WWER-1000 use horizontal type steam generators which are cylindrical objects of 3 m diameter and 12 - 15 m length. Steam goes through steam generator tube sheets in the bottom from hot end and passes its energy through heat-exchanger tube system and become colder on the other tube sheet. Reactors WWER-440 have more than 5,000 tubes in one steam generator and reactors WWER-1000 have already 11,000 tubes in one steam generator. Each block of WWER-440 reactor has 6 steam generators, and each block of WWER-1000 reactor has 4 steam generators. A thin-walled steam generator tube can be a threat of nuclear fallout in case of high pressure in primary coolant loop. To be an effective barrier these tubes should be free from corrosion, thinning and cracks.

Computational modeling of boric acid droplet entrainment during water-water energetic reactor operation in emergency boiling mode

The results of a computational modeling of the processes of droplet entrainment of boric acid by a vapor stream in the event of an emergency at a water-water energetic reactor (WWER) nuclear power plant associated with a guillotine rupture of the main circulation pipeline and the loss of all alternating current sources are presented in the article. The dependence received in the course of computational modeling makes it possible to estimate the maximum size of the droplets carried away by the steam stream during the emergency process in the WWER reactor unit during the operation of the complex of passive safety systems in the pressure range 0.2-0.5 MPa. It was established that to the end of 72 hours of an accident the maximum size of droplets carried away from reactor is 0.11 mm.

Investigation of the equalization capability of submerged perforated sheets under thermal–hydraulic conditions of a horizontal steam generator

The results of a study on the equalization of a non-uniform steam load using submerged perforated sheets (SPSs) are presented. The experimental data obtained at the PGV test facility and calculations performed using the STEG code are used. The PGV test facility comprises a model of the horizontal steam generator used for Russian type water-water energetic reactors. A short description of the 3D thermal–hydraulic code STEG is presented. Tests for determining the equalization capability of the SPSs are performed using a non-uniform steam supply on the hot and cold sides of the test facility. The equalization of the steam flow rate is estimated using experimental pressure drops through the SPSs. Furthermore, a detailed analysis of the tests conducted with the STEG code has been performed. The SPS perforation degrees, which improve the equalization of the non-uniform steam load, are found using the STEG code.

Non-natural ruthenium isotope ratios of the undeclared 2017 atmospheric release consistent with civilian nuclear activities

Understanding the circumstances of the undeclared 2017 nuclear release of ruthenium that led to widespread detections of the radioisotope 106Ru in the Eurasian region, and whether it derives from a civilian or military source, is of major importance for society and future improvements in nuclear safety. Until now, the released nuclear material has merely been studied by analyzing short-lived radioisotopes. Here, we report precise measurements of the stable isotopic composition of ruthenium captured in air filters before, during, and after the nuclear release, and find that the ruthenium collected during the period of the 2017 nuclear release has a non-natural isotopic composition. By comparing our results with ruthenium isotopic compositions of spent nuclear fuels, we show that the release is consistent with the isotopic fingerprints of a civilian Russian water-water energetic reactor (VVER) fuel at the end of its lifetime, and is not related to the production of plutonium for nuclear weapons.

Decision support system for WWER-1000 reactivity management

The article describes the developed software for automating operational calculations of reactivity changes of the water-water energetic reactor (WWER-1000). The main calculation dependencies, screen forms and test results on the Rostov nuclear power plant (Rostov NPP) are given. The calculation accuracy increase is shown.