New Generation Of Nuclear Reactors Research Articles

Wetting of laser-textured tungsten substrate by molten tin

In the new generation of nuclear reactors, plasma facing components tend to be designed as Sn/W composite components. Considering the combination of tin and tungsten, the wettability evaluation of Sn/W system and its influencing factors are urgent work. Wetting experiments of Sn/W system were carried out in high vacuum by using modified sessile drop method. The results show that after vacuum annealing at 1100℃, good wettability (∼ 48° intrinsic contact angle) can be obtained quickly at 900℃ below the critical temperature of 1050℃. After laser surface texturing, the greater the roughness of the tungsten substrate surface, the better the wettability can be obtained. In essence, this correspondence is closely related to the geometric parameters of the grooves formed by laser etching, which corresponds to the predicted results in Wenzel model. The phenomenon of pinning and depinning of triple lines in the later stage of spreading is related to the spreading of liquid in the groove, and the spreading rate of the latter is faster than that of macro triple line, which leads to this phenomenon.

Simulation and design of passive residual heat removal system for Fluoride-Salt-cooled high-Temperature Advanced reactor (FuSTAR)

The new generation of reactor technology: Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR), mainly applied to the steady supply of stable electricity and high-temperature process heat for remote, underground, and arid areas. So far, the design of neutron physics and thermal hydraulics of the core have been finished, but the heat transport system and Passive Residual Heat Removal System of FuSTAR still need to be designed and optimized. Given the complexity and consistency of the design, a unified design and optimization method is required, which can directly obtain detailed configurations and dimensions. In this paper, a design and optimization method - Multi-Layer Nonlinear Programming was proposed to obtain the optimal parameters, including vital thermodynamic values, the dimensions of heat exchangers, the configurations of pipelines, and the optimal equivalent parameters in the transient simulation. Finally, based on Multi-Layer Nonlinear Programming, the effectiveness of the Passive Residual Heat Removal System was demonstrated by the deterministic safety analysis. The detailed parameters of the Passive Residual Heat Removal System in this paper can provide references for subsequent experimental verification. Moreover, for any reactor relying on a heat transport system, researchers can directly obtain its excellent preliminary configurations by this method avoiding a lot of iteration and conflict design, which is conducive to the integrated modeling and simulation of a new generation of reactors.

Optimization of the Pressure Resistance Welding Process for Nuclear Fuel Cladding Coupling Experimental and Numerical Approaches

An approach coupling experimental tests and numerical simulation of the pressure resistance welding (PRW) process is proposed for optimizing fuel cladding welds for the new generation of nuclear reactors. Several experimental welds were prepared by varying the dissipated energy, which accounts for the effect of electric current and welding time applied during the PRW process. A working zone, a function of both applied dissipated weld energy and plug-displacement, was then identified on the basis of the microscopy observations of the weld defects. In addition, the numerical approach, based on a 2D axisymmetric multi-physics finite element model, was developed to simulate the PRW process in a plug-tube configuration. The proposed model accounted for interactions between the electrical, thermal and mechanical phenomena and the electro-thermo-mechanical contact between the pieces and electrodes. Numerical simulations were first validated by comparison to experimental measurements, notably by comparing the plug-displacement and the size and position of the heat-affected zone (HAZ). They were then used to assess the effect of the applied parameters on the maximum temperature and cumulated plastic strain reached during welding and the effect of the welding force on the quality of the weld. According to the numerical computations, the maximum temperature reached in the weld remains well below the melting temperature. Changing the welding force implies also modifying the applied energy in order to maintain the quality of the welds. Applied to different plug and clad geometries, the proposed model was shown to be useful for optimizing the joint plane for such a welding configuration.

High-Fluence Multi-Energy Ion Irradiation for Testing of Materials.

Structural materials of the new generation of nuclear reactors, fission as well as fusion, must often cope with high production rates of transmutation helium. Their testing hence requires either a powerful source of fast neutrons or a high-fluence ion-irradiation facility providing sufficient amounts of high-energy helium to investigate its effect on the material. Most ion irradiation studies, however, concentrate on basic effects such as defect evolution or bubble swelling in narrow near-surface regions modified by ion bombardment. Studies on bulk samples with a relatively thick implanted region, which would enable, for instance, micromechanical testing, are underrepresented. This gap might be filled by high-fluence multi-energy ion irradiations modifying several tens of micrometres of the investigated substrate. High-energy ion accelerators providing reasonable currents with energies of tens of MeV are rarely employed in such studies due to their scarcity or considerable beamtime costs. To contribute to this field, this article reports a unique single-beam He implantation experiment aimed at obtaining quasi-uniform displacement damage across >60 μm with the He/dpa ratio roughly one order of magnitude above the typical spallation neutron target irradiation conditions. Some technical aspects of this irradiation experiment, along with recent developments and upgrades at the 6 MV Tandetron accelerator of the Slovak university of technology in Bratislava, are presented.

A New View of the Dynamics of Particle Structure

The new view denies the existence of fundamental particles in matter, and instead states that any particle and matter is a dynamic superposition of three natural elements, motion, force, and emptiness. This paper proposes a new theory of the structure of matter and the unification of the four forces. A correct understanding of the logic of material structure will enable atomic energy and nuclear physics to generate new manufacturing ideas. The dissertation expounds the natural distribution of dynamics and forces in space and the new deterministic theory of dynamic relationship. This theory solves problems such as the unity of the four forces and the dynamic laws of the material structure. In practical applications, it can provide new theoretical guidance for industrial manufacturing such as nuclear energy, fusion engineering, new energy, new materials, battery energy storage, new-generation reactors, etc., and then adopt effective means in line with the laws of natural power to make the substances needed by human beings. The supply of resources and energy is perfectly satisfied. Applying this theoretical model of material structure to all aspects of the manufacturing industry will surely open up a new era of material civilization for human beings.

Synergistic effect of residual stress and phosphorus grain-boundary segregation on embrittlement of coarse-grained heat-affected zone in a novel Ni-Cr-Mo reactor pressure vessel steel

SA508Gr.4N steel is a novel Ni-Cr-Mo low-alloy steel for the new-generation reactor pressure vessel. Herein, a coarse-grained heat-affected zone (CGHAZ) in a P-doped SA508Gr.4N steel is simulated using Gleeble thermal cycling simulation with a peak temperature of 1320 ℃ under a heat input of 100 kJ cm−1. For the as-simulated specimen, due to the synergistic effect of high residual stress and P grain-boundary segregation, the fracture appearance changes in the order of ductile fracture to intergranular fracture, and then to cleavage fracture with decreasing impact temperature. With the residual stress relieved by annealing at 620 ℃, the specimen toughens to a certain extent even after ageing at 560 ℃, causing a decrease in ductile-to-brittle transition temperature (DBTT), although the boundary cohesion declines to a certain degree owing to the increase of P segregation. Simultaneously, the intergranular-to-cleavage fracture transition temperature (ICFTT) shifts to a lower temperature below –190 ℃, enabling the specimen to exhibit intergranular fracture even at −190 ℃. With the ageing temperature lowered to 500 ℃, the boundary segregation of P is raised to a high level, leading to a large increase in DBTT and a further decrease in ICFTT. Accordingly, the synergistic effect of residual stress and P boundary segregation affects the fracture nature of CGHAZ, thereby making fracture appearance have different characteristics. A mechanism diagram for the above effect is proposed based on the experimental results.

Study of the Effect of Doping ZrO2 Ceramics with MgO to Increase the Resistance to Polymorphic Transformations under the Action of Irradiation.

The purpose of this study is to assess the effect of doping ZrO2 ceramics with MgO on radiation swelling and polymorphic transformations, as a result of irradiation with heavy ions. Interest in these types of materials is due to the great prospects for their use as structural materials for new-generation reactors. The study established the dependences of the phase composition formation and changes in the structural parameters following a change in the concentration of MgO. It has been established that the main mechanism for changing the structural properties of ceramics is the displacement of the cubic c-ZrO2 phase by the Zr0.9Mg0.1O2 substitution phase, which leads to an increase in the stability of ceramic properties to irradiation. It has been determined that an increase in MgO concentration leads to the formation of an impurity phase Zr0.9Mg0.1O2 due to the type of substitution, resulting in changes to the structural parameters of ceramics. During studies of changes in the strength properties of irradiated ceramics, it was found that the formation of a phase in the Zr0.9Mg0.1O2 structure leads to an increase in the resistance to cracking and embrittlement of the surface layers of ceramics.

Radiation swelling and hardness of high-entropy alloys based on the TiTaNbV system irradiated with krypton ions

The aim of this work is to study the effect of irradiation with low-energy Kr14+ ions with an energy of 280 keV on radiation swelling and a decrease in the strength characteristics of high-entropy alloys based on the TiTaNbV system. The prospects of these studies are due to the possibility of obtaining new data on the radiation resistance of high-entropy alloys, which have great potential for use as structural materials for new-generation reactors. The choice of this type of irradiation and a dose of 5 × 1015 Kr/cm2 made it possible to simulate the effect of radiation swelling arising at a displacement per atom of 30–50 dpa in a small surface layer no more than 100–150 nm thick. During the studies carried out, it was found that the formed medium- and high-entropy alloys TaNbV and TiTaNbV have increased resistance to swelling and deformation of the crystal structure. At the same time, alloys of the NbV and TaNbV type have the highest degree of resistance to softening, while the TiTaNbV alloy has a lower resistance to swelling, since the addition of titanium leads to a decrease in the density and hardness of the alloy.

Analysis of system characteristics of a reactor with supercritical coolant parameters

For 60 years of its existence, nuclear energy has passed the first stage of its development and has proven that it can become a powerful industry, going beyond the 10% level in the global balance of energy production. Despite this, modern nuclear industry is capable of producing economically acceptable energy only from uranium-235 or plutonium, obtained as a by-product of the use of low enriched uranium for energy production or surplus weapons-grade plutonium. In this case, nuclear energy cannot claim to be a technology that can solve the problems of energy security and sustainable development, since it meets the same economic and ‘geological’ problems as other technologies do, based on the use of exhaustible organic resources. The solution to this problem will require a new generation of reactors to drastically improve fuel-use characteristics. In particular, reactors based on the use of water cooling technology should significantly increase the efficiency of using U-238 in order to reduce the need for natural uranium in a nuclear energy system. To achieve this goal, it will be necessary to transit to a closed nuclear fuel cycle and, therefore, to improve the performance of a light-water reactor system. The paper considers the possibility of using a reactor with a fast-resonance neutron spectrum cooled by supercritical water (SCWR). The SCWR can be effectively used in a closed nuclear fuel cycle, since it makes it possible to use spent fuel and discharge uranium with a small amount of plutonium added. The authors discuss the selected layout of the core with a change in its size as well as the size of the breeding regions (blankets). MOX fuel with an isotopic plutonium content corresponding to that discharged from the VVER-1000 reactor is considered as fuel. For the selected layout, a study was made of the reactor system features. Compared with existing light-water reactors, this reactor type has increased fuel consumption due to its improved efficiency and nuclear fuel breeding rate up to 1 and above.

Smart reactors – Combining stimuli-responsive hydrogels and 3D printing

Reactor design and development are key disciplines in process engineering. In order to keep up with modern-day production demands, new reactor concepts have to be conceived, which transcend the mere purpose of a reaction vessel. With the availability of improved fabrication methods and insight into new materials, novel process integration measures and process-adapted setups can be developed to provide a new generation of reactors. The application of smart materials allows for response-triggered actuation, which can be used in setups for self-contained process control without external input. In combination with available 3D fabrication technologies, smart materials can be integrated and adapted to the reaction system demands, to develop smart reactor concepts. Herein, we demonstrate an exemplary application of a smart reactor module for an exothermic emulsion copolymerization of vinyl acetate (VAc) and Versa® 10, a vinyl neodecanoate. The flow manipulation of the reaction medium on a preparative scale is presented by using poly(N-iso-propylacrylamide) (pNIPAM) hydrogels as temperature-triggered actuators, which are able to distinguish the course of the reaction by separating the flow of reactant from product stream during continuous operation.

UK nuclear energy system study from 1956 to 2035

The first nuclear reactor built in 1947 had a research purpose, but the first commercial reactor power station was built nine years later at Windscale, England. After that, 45 reactors were built until 1995 for electricity production. Currently, 30 of them are permanently shutdown and 15 are still operational which comprises around 19% of the United Kingdom (UK) energy matrix. Despite the last reactor built in 1995 was a PWR - SIZEWELL-B connected to the grid 23 years ago, the UK government plan to build new generation plants to supply 19 GWe until 2025 and aims to have an additional 16 GWe until 2023. Nevertheless, half of the current capacity should be retired by 2025 and the rest of the AGR generation until 2030. Therefore, the UK nuclear energy system is modeled taking into consideration the retirement and construction of different reactors. The results show the UK nuclear energy system and the reactor transition from old AGRs to a new generation of nuclear reactors. Also, economic features and spent fuel produced due to the nuclear activity up to 2035 are presented.

Estimation influence of boric acid drop entrainment to its accumulation in the VVER reactor in the case of accident

Process of boric acid mass transfer during accidents accompanied with rupture of circulation pipelines in VVER reactors of new generation equipped with passive safety systems are examined. Results of calculation of variation of boric acid concentration in VVER-TOI reactor in case of accident development process are presented. Positive effects of boric acid droplet entrainment on the processes of acid accumulation and crystallization in the reactor core are demonstrated. The obtained results allow formulating the conclusion on the possibility of these processes in the reactor core which may lead to the disruption of heat removal from fuel pins. Review of available published reference data on physical properties of boric acid solutions (density, viscosity, thermal conductivity) is given. It is established that available information is of too general nature and fails to cover the whole range of parameters (acid temperature, pressure and concentration) typical for potential emergency situation on NPP equipped with VVER reactor. Necessity of experimental study of processes of droplet entrainment under parameters typical for VVER emergency operation conditions, as well as investigation of thermal physics properties of boric acid within wide range of acid concentration values is required.

Neutronic Analysis of Heavy Water Application in HPLWR Emergency Core Cooling System

Since 2002 a generation IV nuclear energy system development project has been initiated to enhance the future role of nuclear energy systems. Challenging technology goals for generation IV nuclear reactor systems were defined in 4 areas: sustainability, economics, safety and reliability and proliferation resistance. Light water reactors at supercritical pressure, being currently under design, are the new generation of nuclear reactors. The high-performance light water reactor (HPLWR) is the European version of the supercritical-pressure water cooled reactor (SCWR). The aim of this study is to reliability analysis of heavy water emergency core cooling system (ECCS) (instead of light water ECCS) in HPLWRs during LOCA (loss of coolant accident) through neutronics calculations and applied feedback.. Reliability analysis (neutronic calculations) are performed using WIMS and CITATION codes. Changes in neutronic parameters, such as neutron cross-sections, effective multiplication factor and power peaking factor during LOCA, are discussed by injecting different concentrations of the heavy water coolant by ECCS.

Development of group-constants production system for the new generation of reactors

Development of group-constants production system for the new generation of reactors

NEUTRONIC INVESTIGATION OF SEMI-HEAVY WATER APPLICATION IN HPLWR NEW FLOW PATTERN

The high-performance light-water reactor (HPLWR) is the European version of the supercritical-pressure water-cooled reactor. Light water reactors at supercritical pressure, being currently under design, are the new generation of nuclear reactors. The variable moderation high performance light-water reactor (VMHWR) is the new version of HPLWR conceptual design. In this design, the fluid passing through the moderator channels and the assembly gaps (known as a moderator) flows in channels detached from coolants. This paper presents a neutronic analysis of semi-heavy water in VMHWR. Neutronic calculations were made through WIMSD5 and CITATION codes. Neutronic parameters such as effective multiplication factor, power peaking factor, and the effect of semi-heavy water on these parameters were investigated.

Radiation chemistry for modern nuclear energy development

Radiation chemistry plays a significant role in modern nuclear energy development. Pioneering research in nuclear science, for example the development of generation IV nuclear reactors, cannot be pursued without chemical solutions. Present issues related to light water reactors concern radiolysis of water in the primary circuit; long-term storage of spent nuclear fuel; radiation effects on cables and wire insulation, and on ion exchangers used for water purification; as well as the procedures of radioactive waste reprocessing and storage. Radiation effects on materials and enhanced corrosion are crucial in current (II/III/III+) and future (IV) generation reactors, and in waste management, deep geological disposal and spent fuel reprocessing. The new generation of reactors (III+ and IV) impose new challenges for radiation chemists due to their new conditions of operation and the usage of new types of coolant. In the case of the supercritical water-cooled reactor (SCWR), water chemistry control may be the key factor in preventing corrosion of reactor structural materials. This paper mainly focuses on radiation effects on long-term performance and safety in the development of nuclear power plants.

Neutron Induced Capture and Fission Processes on238U

Nuclear data on Uranium isotopes are of crucial interest for new generation of nuclear reactors. Processes of interest are the nuclear reactions induced by neutrons and in this work mainly the capture and the fission process on 238 U will be analyzed in a wide energy interval. For slow and resonant neutrons the many levels Breit – Wigner formalism is necessary. In the case of fast and very fast neutrons up to 200 MeV the nuclear reaction mechanism implemented in Talys will be used. The present evaluations are necessary in order to obtain the field of neutrons in the design of nuclear reactors and they are compared with experimental data from literature obtained from capture and (n,xn) processes.

Axial enrichment profile in advance nuclear energy power plant at supercritical-pressures

Abstract The High-Performance Light Water Reactor (HPLWR) is the European version of the advance nuclear energy power plant at Supercritical-pressure. A light water reactor at supercritical pressure, being currently under design, is the new generation of nuclear reactors. The aim of this study is to predict the HPLWR neutronic behavior of the axial enrichment profile with an average enrichment of 5 w/o U-235. Neutronic calculations are performed using WIMS and CITATION codes. Changes in neutronic parameter, such as Power Peaking Factor (PPF) are discussed in this paper.

Subchannel analysis of Al2O3 nanofluid as a coolant in VMHWR

AbstractThe main objective of this study is to predict the thermal hydraulic behavior of nanofluids as the coolant in the fuel assembly of variable moderation high performance light water reactor (VMHWR). VMHWR is the new version of high performance light water reactor (HPLWR) conceptual design. Light water reactors at supercritical pressure (VMHWR, HPLWR), being currently under design, are the new generation of nuclear reactors. Water-based nanofluids containing various volume fractions of Al2O3 nanoparticles are analyzed. The conservation equations and conduction heat transfer equation for fuel and clad have been derived and discretized by the finite volume method. The transfer of mass, momentum and energy between adjacent subchannels are split into diversion crossflow and turbulent mixing components. The governed non linear algebraic equations are solved by using analytical iteration methods. Finally the nanofluid analysis results are compared with the pure water results.

Simulation of an Accelerator Driven Subcritical Core with Mixed Uranium-Thorium Fuel

During recent years, a new generation of nuclear reactors, known as “Accelerator Driven Subcritical Reactors”, has been developed. One of the new application aspects for such reactors (besides transmutation of High Level Waste and burning Minor Actinides) is usage of thorium as nuclear fuel. In this work a subcritical core in experimental scale is simulated by MCNPX code. The core contains two types of fuel assemblies: (85% ThO2 + 15% UO2) and MOX (U-Pu). In the first step, only the thorium-contained fuel assemblies are loaded into the core. Criticality calculations using MCNPX show that the keff is so low that the fuel assemblies cannot run the subcritical core. This implies that MOX (U-Pu) assemblies must be loaded as well. Neutronic parameters of the thorium- fueled Accelerator Driven Subcritical core are then calculated as well as some other parameters related to accelerator coupled with the core. The main objective of this simulation is to study the behavior of Accelerator Driven Subcritical core with thorium assemblies.