Next Generation Nuclear Research Articles

Miniature Dc Electromagnetic Pumps of Molten Lead and Sodium to Support Development of Gen-Iv Nuclear Reactors

Miniature Dc Electromagnetic Pumps of Molten Lead and Sodium to Support Development of Gen-Iv Nuclear Reactors

Lattice Boltzmann Solidification Modeling of Forced Convection Internal Flows Applied to Gen-IV Nuclear Reactors

Lattice Boltzmann Solidification Modeling of Forced Convection Internal Flows Applied to Gen-IV Nuclear Reactors

Towards a single European strategic research and innovation agenda on materials for all reactor generations through dedicated projects

The goal of the ORIENT-NM action is to produce a single European strategic vision on research and innovation concerning nuclear materials in the EU, serving all reactor generations and nuclear systems. The key in this endeavour is to focus on advanced materials science practices that, combined with digital techniques, will enable acceleration in materials development, manufacturing, supply, qualification, and monitoring, in support of nuclear energy safety, efficiency, economy and sustainability. The research agenda will be rooted in existing virtuous examples of nuclear materials science projects. Here the results of three of them are summarised, thereby covering different reactor applications and families of materials, as well as a range of advanced material research approaches. GEMMA addressed a number of key areas concerning the development and qualification of metallic structural materials for GenIV reactor conditions, focusing on austenitic steels and their compatibility with several non-aqueous coolants, their welds and the modelling of their stability under irradiation. INSPYRE was an integrated project applying a basic science approach to (U,Pu)O2fuels, to develop physics-based models for the behaviour of nuclear fuels under irradiation and improve fuel performance codes. Modelling was also the focus of the M4F project, which brought together the fission and fusion materials communities to study the effects of localised deformation under irradiation in ferritic/martensitic steels and to develop good practices to use ion irradiation as a tool to evaluate radiation effects on materials.

Innovation to zero emissions: nuclear energy trough nonproliferation and environment protection- Romanian experience for GEN 4

The UN Intergovernmental Panel on Climate Change (NY 2019) has concluded that nations must move more swiftly to reduce emissions of carbon dioxide, methane, and other greenhouse gases to avoid the most devastating effects of global warming. The paper presents nuclear energy as part of the solution. Due to the fact that the population is concerned about nuclear proliferation, plant safety and radiation protection, the paper presents the Romanian experience regarding the reduction of the risk of proliferation as well as the project of the 4th generation reactor ALFRED. One of the most important steps in assessing the candidate materials for Generation IV reactors is the material performance under neutron irradiation. In this respect, the paper also presents the results of the evaluations on some potential materials to be used in fast lead cooled reactors.

Performance enhancement of methanol reforming reactor through finned surfaces and diffused entry for on-board hydrogen generation

Hydrogen-rich combustion in engines helps in reducing pollutants significantly. But hydrogen usage on a moving vehicle is not getting large-scale user acceptance mainly due to its poor energy storage density resulting in shorter driving ranges. This storage issue led to the hunt for mediums that can efficiently produce on-board hydrogen. Methanol proves to be an efficient alcohol fuel for producing hydrogen through steam reforming reaction. The heat energy required for such endothermic reaction is obtained through exhaust engine waste energy and this process is collectively known as thermochemical recuperation. However, the conventional reactor used for this process faces a lot of problems in terms of efficiency and methanol conversion. In this study, an attempt has been made to improve the design of the reactor for on-board hydrogen generation using engine exhaust heat for addressing the challenges related to performance and hydrogen yield. For enhancing the heat transfer, a finned surface (straight & wavy) was introduced in the reactor which resulted in an increment in methanol conversion significantly. It was found that wavy fin improved the methanol conversion up to 96.8% at an exhaust inlet temperature of 673 K. Also, a diffusing inlet section was introduced to increase the residence time of reactant gases while passing through the catalyst zone. Under given inlet conditions, the methanol conversion for 6° diffuse inlet reactor goes up to 87.9% as compared to 75.4% for the conventional reactor.

Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation

Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe2+ up to 600 peak displacement-per-atom (dpa) at 450 ℃. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. The comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.

Corrosion behavior of refractory metals in liquid lead at 1000 °C for 1000 h

Lead-based fast reactor (LFR) has become one of the most promising reactors for Generation IV nuclear systems. A developing trend of LFR is high efficiency, along with operation temperatures up to 800 °C or even higher. One of key issues in the high-efficiency LFR is corrosion of cladding materials with lead at high temperatures. In this study, corrosion behavior of some refractory metals (Nb, Nb521, and Mo-0.5La) was investigated in static lead at 1000 °C for 1000 h. The results showed that Nb and Nb521 exhibited an intense dissolution corrosion with obvious lead penetration after corrosion, and lead penetration extended along the grain boundaries of the specimens. Furthermore, Nb521 showed a better corrosion resistance than that of Nb as a result of the elements of W and Mo included in Nb521. Mo-0.5La showed much better corrosion resistance than that of Nb and Nb521, and no lead penetration could be observed. However, an etched morphology appeared on the surface of Mo-0.5La, indicating the occurrence of corrosion to a certain degree. The results indicate that Mo-0.5La is compatible with lead up to 1000 °C. While Nb and Nb alloys might be not compatible with lead for high-efficiency LFR at such high temperatures.

A New Stress Intensity Factor Solution Based on the Response Surface Method for Nozzle Corner Cracks in Nuclear Reactor for Thermal Energy Generation

As considered carbon-free, the use of nuclear energy for thermal energy generation may expand in the future, with the guarantee of safe operation of the nuclear reactor. In a nuclear reactor pressure vessel (RPV), the nozzle area is an important part of the safe operation. It bears internal pressure, axial force, and overall moment, and at the same time bears higher stress than the rest of the vessel. To assess the integrity of the nozzle structure with a crack under combined load, an accurate solution of stress intensity factors (SIF) along the crack front is necessary. To obtain the SIF, this paper proposes a solution method that uses the stress on the crack surface and the response surface method to fit the stress under the framework of the linear superposition technique. This method is the first choice to determine a series of influence coefficients under unit pressure load. Then, one can estimate the SIFs by superposition method for an arbitrary stress distribution resulted from combined loads. The proposed solution is verified for a typical RPV with cracks under internal pressure, axial force, and global bending moment. The results reveal that the proposed solution is in good agreement with the existing solutions under internal pressure. Therefore, it can be obtained that this solution can be effectively used for the structural integrity assessment of RPV with nozzle corner cracks.

Numerical analysis of wave effect on steam–air condensation on a vertical surface

Steam condensation with non-condensable gas is an unavoidable problem for passive containment cooling system (PCCS) in many third generation reactors. The effects of the air mass fraction in steam–air flow and condensing wall subcooling on the steam condensation at the different gas velocities under the different types and amplitudes of solitary or sinusoidal wave walls are numerical analyzed, in the cases that the wavy film is regarded as wavy solid wall. The 41% rise of average condensation flux (ACF) of steam–air flow can be obtained in the channel with large-amplitude solitary wave wall than that of flat wall while the less variation of ACF happens in the mode with sinusoidal wave wall. The sudden expansion and contraction of flow section along the wavy wall as well as vortex generation in the channel influence local condensation flux greatly. All these results are helpful for thermal analysis of condensation in the PCCS.

Atomistic simulations to study the effect of helium nanobubble on the shear deformation of nickel crystal

So far, many theoretical and experimental studies have investigated the detrimental effect of helium nanobubble on the mechanical properties of nuclear structural materials. In this article, efforts have been made to study the effect of helium nanobubble on the onset of plastic deformation in a single crystal Ni. The onset of plastic deformation has been quantified as a function of the orientation of slip planes with shear force and configuration of helium nanobubbles. Two configurations were generated for helium nanobubble, one with radius fixed where the He to Ni vacancy ratio was varied. In contrast, the ratio was kept constant in another configuration, and the nanobubble radius was varied. From the simulation, it was predicted that irrespective of the orientation of Ni crystal with respect to shear forces, helium nanobubble has a detrimental effect on the mechanical properties of the single-crystal Ni. Compared to the size of helium nanobubble, the higher concentration of helium atoms in the nanobubble has a more detrimental effect on mechanical strength. The focus of nuclear scientists is to develop future generation nuclear reactors and predict the safe life cycle of existing nuclear structures, which are in use for a long time. Helium nanobubbles change the mechanical behavior of underlying nuclear material in an effective manner. The results presented in this article will help in developing a better understanding of the deformation in irradiated Ni crystal.

The electron accelerator driven sub-critical system

We reference a typical pressurized water reactor and design an electron accelerator driven subcritical system (eADS), the reactor is a thermal reactor. The physics to create external neutrons in eADS is different to the spallation reactions which are used in the proton accelerator driven ADS system.The neutron energy spectra in eADS system are the combination of neutron energy spectra from photonuclear reactions and fission reactions. Simulation calculation shows that the neutron energy spectra from photonuclear reactions and from fission reactions have a remarkably difference. A large portion of neutrons emitted from the photonuclear reaction target is peaked in the energy region 10-8 ∼ 10-7MeV, the neutrons within this energy region are well thermalized neutrons which are readily absorbed by the fissile isotopes to split and release energy.The power peaking factors are greater than 2.0 in the eADS system driven by one and five electron accelerator(s). After rearranging the burnable poison rods in 1.6 % and 2.4 % fuel enrichment assemblies and inserting some burnable poison rods in the fuel assemblies with 3.1 % enrichment in the periphery area of reactor core, the power peaking factors can be reduced to about 1.2. The multiple electron accelerators in the subcritical system flatten the neutron flux distribution and improve the power peaking factors. Our calculation results indicate that with the symmetrical distribution of the four electron targets about the electron target on the center of the reactor core, the larger the radius of other four electron targets to the center of reactor core, the higher the keff and the lower the power peaking factors. This implies that in designing an eADS system all the electron targets should be in the periphery area of the reactor core except the electron target on the center.The electron accelerator driven system may be the real next generation reactor to burn depleted nuclear fuel, generate electric power and transmute nuclear wastes, is promising and deserves to explore with.

Radiogenic neutron background in reactor neutrino experiments

We report a novel correlated background in the antineutrino detection using the inverse beta decay reaction. Spontaneous fissions and $(\alpha,n)$ reactions in peripheral materials of the antineutrino detector, such as borosilicate glass of photomultipliers, produce fast neutrons and prompt gamma rays. If the shielding from the material to the detector target were not thick enough, neutrons and gammas could enter the target volume and mimic antineutrino signals. This paper revisits the yields and energy spectra of neutrons produced in B$(\alpha,n)$N and F$(\alpha,n)$Na reactions. A Geant4 based simulation has been carried out using a simplified detector geometry for the present generation reactor neutrino experiments. The background rates in these experiments are estimated. If this background was not taken into account, the value of the neutrino mixing angle $\sin^22\theta_{13}$ would be underestimated. We recommend that Daya Bay, RENO, Double Chooz, and JUNO, carefully examine the masses and radiopurity levels of detector materials that are close to the target and rich in boron and fluorine.

Evolving from a hydrocarbon-based to a sustainable economy: Starting with a case study for Iran

Environmental and socioeconomic challenges facing many hydrocarbon-rich nations could help them evolve from hydrocarbon-based (black) into renewables-based (green) economies. In this paper, we present a case study for application of this Black Into Green (BIG) method for Iran. We show that with international cooperation Iran could indefinitely meet its energy and freshwater needs by leveraging large-scale renewable energy harvesting systems integrated with seawater pumped storage hydropower, reverse osmosis desalination, and enhanced distribution systems. This BIG transition could be facilitated by allocating a significant portion of Iran’s potential hydrocarbon-based income to acquire, develop, and deploy renewable energy technology that utilizes Iran’s immense solar and wind energy potential. Our analysis shows that developing just 1% of identified sites would provide over 30 TWh of annual renewable electricity generation and storage, and 221 million m3/day of renewably-powered desalination capacity to meet the freshwater needs of over 85% of its population. In addition to providing renewable energy design and manufacturing technology, the international community could also work with Iran to help evolve its nuclear power program toward small modular reactors for baseload power generation. Such an arrangement would contribute to regional and global security, and greatly reduce the Middle East’s carbon emissions of which Iran is the largest emitter. Results from this holistic approach could then serve as a model for the world economy to work with other resource-rich nations to reduce their environmental impact while modernizing their economies and helping them achieve the socioeconomic goals outlined in the UN’s Brundtland report.

Severe accident analysis of the Qinshan Nuclear Power Plant and evaluation of boundary conditions for ex-vessel heat transfer

Severe accident analysis is critical for the operation and licensing of new generation reactor designs as well as for the prevention and management of severe accidents. The Severe Accident Management Guidelines (SAMGs) are developed and implemented for prevention and mitigation measures of a severe accident. Retention of corium in the reactor vessel via ex-vessel cooling is one of the most promising mitigation strategies. Its success requires a thorough analysis of the reactor core following a severe accident to determine the boundary conditions for external cavity flooding. In this study, two loss of coolant accident cases at the Qinshan nuclear power plant are analyzed using RELAP5/SCDAP simulations. Passive accumulators are credited in the first case (Case-1), but not in the second (Case-2). The important severe accident events are studied for the plant and the sequence of events and melt conditions are captured. Hydrogen generation has been estimated to be 120 kg in Case-1 and 108 kg in Case-2. In both instances, a molten pool is formed. The corium reached a maximum temperature of 3016 K and 3006 K, respectively in Case-1 and Case-2. The metallic/oxidic layer properties and heat transfer coefficient of the corium pool as well as the ex-vessel heat transfer boundary conditions are evaluated analytically for the reference plant. The analytically determined temperatures of oxidic and metallic phases in the corium pool are 3006 K and 1919 K respectively. The upward and sideward heat fluxes (vessel wall) are determined to be 298.3 kW/m2 and 730.75 kW/m2 respectively, while the heat transfer coefficients for oxidic and metallic phases are 2.80 W/m2-K and 4.42 W/m2-K respectively.

Fiber/matrix debonding evaluation of SiCf/SiC composites using micropillar compression technique

Third-generation silicon carbide (SiC) composites reinforced by SiC fibers (Hi-Nicalon S [HNS] and Tyranno SA3 [SA3]) are attractive for use in next generation reactors owing to their high strength and chemical inertness at high temperatures, as well as enhanced radiation tolerance under neutron irradiation environments. To optimize composite performance, the interfacial mechanical properties of chemical vapor–infiltrated (CVI) SiCf/SiC composites are investigated in this effort by using a slant interface micropillar compression testing procedure. The micropillar test specimens, containing an inclined pyrolytic carbon (PyC) interphase, are prepared using a focused ion beam. The novel microcompression testing successfully quantifies the debond shear strength and internal friction coefficient of micropillar test samples by using the Mohr-Coulomb formulation. According to four types of SiCf/SiC composite microcompression test results, interfacial properties and debond mechanisms are significantly affected by the PyC layer thickness, the local bonding mechanism of the PyC interphase on the SiC fiber surface, and the surface roughness of fibers. Regardless of PyC thicknesses, SA3-reinforced CVI SiCf/SiC composites are found to have much higher debond shear strengths than HNS-reinforced SiCf/SiC CVI composites. By using this micropillar compression technique alongside analytical methods, we uncover new understandings of PyC interface properties. Additionally, the micropillar test results obtained are correlated with macroscopic mechanical properties of neutron-irradiated CVI SiCf/SiC composites.

Research on Response Characteristics and Control Strategy of the Supercritical Carbon Dioxide Power Cycle

With the development of GEN-IV nuclear reactor technology, the supercritical carbon dioxide (SCO2) Brayton cycle has attracted wide attention for its simple structure and high efficiency. Correspondingly, a series of research has been carried out to study the characteristics of the cycle. The control flexibility of the power generation system has rarely been studied. This paper carried out a dynamic performance of the 20 MW-SCO2 recompression cycle based on the Simulink software. In the simulation, the response characteristics of the system main parameters under the disturbances of cooling water temperature, split ratio, main compressor inlet temperature and pressure were analyzed. The results show that the turbine inlet temperature is most affected by the disturbances, with a re-stabilization time of 2500–3000 s. According to the response characteristics of the system after being disturbed, this study proposed a stable operation control scheme. The scheme is coordinated with the main compressor inlet temperature and pressure control, recompressor outlet pressure control, turbine inlet temperature control and turbine load control. Finally, the control strategy is verified with the disturbance of reduced split ratio, and the results show that the control effect is good.

Techno-Economic Assessment of Fuel Cycle Facility of System Integrated Modular Advanced Reactor (SMART)

The economic assessment of advanced nuclear power reactors is very important, specifically during the early stages of concept design. Therefore, a study was performed to calculate the total cost estimation of fuel cycle supply for a system modular advanced reactor (SMART) by using the Generation-IV economic program called G4-ECONS (Generation 4 Excel-based Calculation of Nuclear Systems). In this study, the detailed description of each model and methodology are presented including facility, operations, construction matrix, post-production model, and fuel cycle cost estimation model. Based on these models, six Generation-III+ and Generation-IV nuclear reactors were simulated, namely System 80+ with benchmark data, System 80+ with uranium oxide (UOx) and mixed oxide (MOx) fuel assemblies, fast reactor, PBMR (Pebble Bed Modular Reactor), and PWR (Pressurized Water Reactor), with partially closed and benchmarked cases. The total levelized costs of these reactors were obtained, and it was observed that PBMR showed the lowest cost. The research was extended to work on the SMART reactor to calculate the total levelized fuel cycle cost, capital cost, capital component cost, fraction of capital spent, and sine curve spent pattern. To date, no work is being reported to calculate these parameters for the SMART reactor. It was observed that SMART is the most cost-effective reactor system among other proven and advanced pressurized water-based reactor systems. The main objective of the research is to verify and validate the G4-ECONS model to be used for other innovative nuclear reactors.

Pulse/Pulse Reverse Deposition of NiMo Alloys for Improved Molten Salt Reactor Performance

The use of molten fluoride salts as a primary coolant within next generation nuclear reactors has the potential to improve efficiency and nuclear reactor safety by operating at low pressures and high temperatures without boiling. However, these coolants will require the development of new corrosion resistant material systems that will have to meet or supersede existing standard codes for these systems. Therefore, state of the art reactors require validation and testing of new material systems that can produce robust component structures and enable them to withstand these corrosive environments. Furthermore, any next generation process to apply protective coatings must allow for a conformal, thick, dense, and bonded materials to be formed onto any low cost ASME material (like 316H SS) that comprises the structure of heat exchanger and containment vessel components.Within this context, Faraday Technology Inc. is working on developing low cost and high value corrosion-resistant alloy coatings for next generation molten salt reactors with the goal of increasing their functional lifetime, while reducing the component cost. The manufacturing process involves electrodeposition of functionally graded NiMo alloys onto ASME certified substrates and their subsequent high temperature corrosion evaluation in molten fluoride salt electrolytes. A wide array of electrolytes and processing parameters were evaluated in order to understand these effects on the deposit composition, structure, and corrosion resistance properties towards the goal of developing an ideal alloy coating. Specifically, we demonstrated the potential to apply graded NiMo coating to the 316H SS surface (Figure 1) and found that a coating composition of ~25% Mo has the potential to improve the components lifetime. Acknowledgements: The financial support of DOE Contract No. DE-SC0019602 is acknowledged. Figure 1: Top) show a cross-sectional image of a NiMo coated 316HSS component. Bottom) shows composition of Ni (Left/Dashed) and Mo (Right/Solid) as measured by EDX trace from the SEM along the red line (Top image). Figure 1

CRITICAL FLOW PREDICTION MODELS FOR THE COOLANT AT SUPERCRITICAL PARAMETERS


 
 
 An interest of the problems of various thermophysical and hydrodynamic phenomena in the nuclear industry, determined by the real application in the field of analysis of the accident scenarios related to the loss of coolant accident. For the generic super critical water reactor the meaning of the problem at the initial stage of the critical flow process, is the existing of the uncertainty in the accepting boundary conditions to predict the flow characteristics. The article provides an analytical review of existing approaches for describing the critical flow phenomenon of the medium and to focus on the current predictive models. A description of the physical nature of such a phenomenon is provided. The scope of consideration includes information from the literature for single and two-phase flow, taking into account their physical basis and the assumptions made. The task of the work was to analyze the information found and to evaluate and update the data on the application of the models to obtain the critical characteristic. It was supposed to highlight the physical aspects and peculiarities of this phenomenon, as applied to the coolant at supercritical parameters. To formulate important requirements to the representative critical flow model for the possibility of its use in the system codes for evaluation of the nuclear safety problems of promising fourth generation nuclear reactors.
 
 

On-site H2O2 electro-generation process combined with ultraviolet: A promising approach for odorous compounds purification in drinking water system

The on-site generation of hydrogen peroxide (H2O2) attracts much attention for advanced oxidation processes (AOPs) in water treatment. In this work, we reported a novel on-site H2O2 generation reactor combined with ultraviolet (UV) for taste and odor (T&O) compounds removal in drinking water purification systems. The reactor is equipped with a gas diffusion electrode (GDE) accompanied by a proton-conducting solid electrolyte. In this reactor, GDE is used to improve the oxygen transfer efficiency. The porous solid electrolyte is selected for the proton and electron conduction instead of an electrolyte solution, which is crucial for producing pure H2O2 solution. The production rate and concentration of H2O2 can be controlled through adjusting the applied current efficiency and the flow rate of deionized water in a solid chamber. Meanwhile, this system exhibits long-term stability during a continuous operation for 96 h. For the application of the system, the single-cell was assembled into cell stacks, which improved the H2O2 concentration from 780.9 to 2301.3 mg·L−1, along with the energy consumption reduced from 23.81 to 6.67 kWh·kg−1 under the applied current density of 30 mA·cm−2. To evaluate the water purification performance of this system, the oxidation degradation of 2,4,6-trichloroanisoles (TCA) was conducted using the on-site H2O2 generation process under both static and dynamic flow conditions combined with UV light. The degradation rate of TCA was >95% with its initial concentration of 1 μg·L−1 under the hydraulic retention time from 10 to 40 min−1 and H2O2 dosage from 10 to 100 mg·L−1. Meanwhile, almost no residual H2O2 was detected after the degradation process. This study proposed a promising device for drinking water purification.