Loop Of Reactor Research Articles

Understanding the high-temperature corrosion behavior of zirconium alloy as cladding tubes: a review

Operated under extreme conditions, corrosion occurs between zirconium alloy cladding tubes and the coolant in the primary loop of pressurized water reactors (PWRs), contributing to a reduction in the effective metallic material thickness. Therefore, understanding the corrosion behavior of zirconium alloy is vital to both raising the burnup of PWR and the improvement of safety properties of these reactors. During the past decades, extensive investigation was conducted with various conditions, such as changing corrosion temperatures and alloying elements, but contradiction persists and universal conclusion remain elusive. In the present work, a variety of research results that focused on corrosion kinetics, microstructural evolution, and the influence of alloying elements were integrated and summarized, so that a valuable reference can be provided to further research.

Design of a novel gamma camera with large field of view for 16N diagnosis in the primary loop of nuclear reactor.

The assessment of the concentration and distribution of l6N, derived from 16O in the cooling water exposed to neutron irradiation, is essential for ensuring radiation safety during nuclear reactor operation. The imaging method allows for the visualization of the intensity distribution of these l6N by capturing gamma-rays emitted during their decay process. However, the existing gamma camera is exclusively compatible with gamma-rays below 2MeV. In this paper, a novel gamma camera featuring a thick double-conical penumbra aperture, a pixelated Lu1.8Y0.2SiO5:Ce scintillator array, and a position-sensitive photomultiplier tube is proposed to address this limitation. This innovative design offers a large field of view (FOV) and is suitable for high energy extended gamma source imaging. The optimization of key parameters of the camera was conducted, and a FOV of 60° and an angular resolution of up to 4.57° were achieved. Imaging simulations, including a simplified model of the primary loop of the pressurized-water reactor by GEANT4 code and image reconstruction using the expectation maximum algorithm, demonstrated that the proposed gamma camera could obtain a satisfactory spatial resolution for diagnosing the distribution of 16N in the primary loop of a nuclear reactor.

Prediction of sodium leakage droplet size distribution in sodium-cooled fast reactor loop

Sodium spray fire is a significant containment-related safety concern in sodium-cooled fast reactors (SFRs). The droplet size distribution of sodium spray is an important factor affecting the combustion intensity of sodium spray fire. Due to the unstable chemical properties of sodium, water is usually used to simulate sodium to carry out spray experiments, and the droplet size relationship between them needs to be obtained. Based on the unstable wave theory, the droplet size similarity model characterized by liquid physical property parameters and atomization pressure difference is developed and verified from the two aspects of pressure and physical property parameters, and compared with existing models. Furthermore, an experimental facility of water atomization using a nozzle to simulate a break is built and the droplet size of water atomization under the differential pressure range of 1.0 bar–6.0 bar is obtained. The droplet size similarity model is used to predict the sodium atomization droplet size at 300–600 °C under different pressures based on these experimental data.

Corrosion products deposition and its effects on conjugate heat transfer for helical cruciform fuel assembly

The corrosion products in the primary loop of the nuclear reactor can be deposited on the surface of the fuel rod, forming a deposition layer called CRUD layer, introducing the heat transfer resistance. The CRUD layer will result in an increase in the temperature of the cladding and accelerating the corrosion of the cladding. In this work, the corrosion product deposition and thermal resistance models of the deposition layer were established. The deposition behavior of corrosion products on the surface of a novel Helical-Cruciform Fuel (HCF) assembly and its influence on fluid-solid conjugate heat transfer were studied by coupling the deposition models with the flow and heat transfer models for fuel rods and coolant. The results show that the deposition rate of particles is affected by the wall shear force; the deposition rate of impurity ions is affected by the wall shear force and the temperature of the coolant in the viscous layer. The erosion rate of the deposited layer is related to the wall shear force and the sum of the particle and ion deposition rates. The thickness of CRUD layer on the surface of HCF assembly increases with the decrease of coolant flow rate and the increase of fuel power. Furthermore, the CRUD layer thickness varies in a period of 90° along the circumference of the fuel and reaches a peak at the root of the fuel ribs.

A Method for Backward Failure Propagation in Conceptual System Design

The increased complexity of modern system designs and demands for quicker time to market have made safety-related verification and validation of such systems more challenging. Incorporating safety and risk considerations at the early stages of design is one way to acquire a more robust initial design for novel systems. Inductive fault analysis has its significance at final stages of design, e.g., verification and validation. However, to preclude certain system failure states—especially for the systems with high failure consequences, a designer would innately prefer to trace back and remedy the causes of failure, as compared to a more cumbersome activity of identifying the faults individually and sifting the combinations that lead to the failure of interest. The work presented in this paper is aimed at the development of a backward failure propagation methodology for analyzing the origins of functional failures in a conceptual design of systems including but not limited to nuclear, mechanical, aerospace, process, electrical/electronics, telecommunication, automotive, etc. This method allows the designer to achieve a robust early design based on the analyses of the system’s functional dependencies before proceeding to the detailed design and testing stages. The insights provided by the analysis at the conceptual design stage also reduce redesign efforts, testing costs, and project delays. The proposed method is a functional analysis approach that extends the Integrated System Failure Analysis for backward failure propagation. When provided with an abstract system configuration, a system’s functional model, and a system’s behavioral model, it utilizes a known functional state (typically a failure) to acquire system component modes and the states of other functions. The method includes inversion of the functional failure logic and component behavioral rules using propositional logic and deductive analysis to assess valid states of a system that satisfy the given initial conditions. To test the method’s scalability, we applied the proposed method to a simplified representation of the secondary loop of a typical pressurized water reactor.

Multi-physics core analysis and verification of NuScale reactor with coupling PARCS/RELAP

The attractiveness of small modular reactors (SMR) lies in high safety, simplicity of design, reduced siting costs, etc.The NuScale is considered one of the unique SMR reactors due to design advantages such as the immersion of the containment vessel in the reactor pool, natural circulation cooling mechanism, integrated loop, and flexibility of its safety functions for up to 72 h after severe accidents. Investigating the neutronic and thermal–hydraulic parameters of a reactor is one of the concerns of nuclear-reactor designers because such parameters play an essential role in the design process and safety assessment. Therefore, in the present study, the internal coupling capability of the PARCS and RELAP codes has been used to analyze the multi-physics parameters of the core and loop of the NuScale reactor. The DRAGON5 code was used to perform lattice calculations and prepare the PMAXS library for PARCS. The code-by-code verification of lattice and core calculations was carried out by the MCNP code. Multiplication factor, power distribution, fuel and moderator feedback coefficients, and other thermal–hydraulic parameters were calculated under a steady-state condition. Burnup calculations have also been performed to investigate the fuel cycle parameters. The results show an acceptable accuracy in comparison with the NuScale reported data.

Isotope kinetics modeling in a circulating fuel system: a case study of the MBIR reactor loop

The paper presents the results of modeling of changes in the isotopic composition of the liquid-salt fuel circulating in the experimental channel of the MBIR reactor facility. The authors tested the ISTAR software environment adapted for solving burnup equations in problems with variable power levels. The loop channel parameters, including two heat exchanger options, were estimated to obtain the appropriate salt transit time through the loop channel zones. Two problems of a circulating fuel system (loop) modeling are considered, namely: (1) modeling the equilibrium salt isotope composition in such a system; and (2) developing a technique for modeling nonstationary isotope kinetics in the MBIR reactor loop. Non-stationary isotope kinetics can be modeled as sequential burnup of nuclides in the neutron field and decay during movement in the external circuit. The authors also developed an algorithm for modeling changes in the isotopic composition of fuel salt during its circulation, taking into account the sequential transfer of a given salt volume from the burnup zone to the zone outside the reactor core. Based on this algorithm, a software package was created using the Python 3.9 programming language and ISTAR modules. In addition, a description of the calculation methodology was given and some calculation results obtained using the software were presented. In the process of working with the program, it was found that, for the given times of the fuel being in each of the zones (2 and 200 seconds, respectively), modeling the change in the isotopic composition during the fuel campaign (500 days) will require the calculation of more than 500 thousand steps. In order to save time, it is necessary to find out whether it will be possible to reduce the number of calls to the neutronic calculation code due to a slight change in the isotopic composition of the fuel in the loop per one burnup step. Work is currently underway to optimize this process.

Simulation of radioactive fission gas release from defective PWR fuel rod using the MFPR/R mechanistic code

A software package was elaborated by coupling the mechanistic code MFPR/R with a recently developed advanced model of mass exchange in defective fuel rods, to simulate the release of radioactive gaseous fission products into the coolant during operation of a thermal reactor. To simulate the oxidation of UO2 fuel, the processes of steam dissociation and radiolysis in the rod free volume were additionally considered. The package was validated by calculating the release of krypton, iodine and xenon isotopes into the coolant from defective rods under irradiation in the pressurized light water loop of the NRX reactor at Chalk River Nuclear Laboratories. It is shown that a reasonable overall agreement between calculations and experiment for noble gas isotopes can be obtained by introducing a single fitting parameter that takes into account uncertainties in the rod geometry and the rate of hydrogen production in the rod free volume.

A microstructure-based three-scale homogenization model for predicting the elasto-viscoplastic behavior of duplex stainless steels

A new microstructure-informed three-scale homogenization scheme for elasto-viscoplastic heterogeneous materials is developed. It is applied to predict the mechanical behavior of cast duplex austenitic-ferritic (AF) steels, which are widely used in primary loop of pressurized water reactors (PWR). At the first scale (microscale), the elasto-viscoplastic behavior of single crystals for both phases is modeled using linear elasticity and a viscoplastic crystal plasticity model. An “affine” type formulation based on the first moments of stresses is applied to the inelastic non-linear part of the deformation. Then, at the second scale (mesoscale), an EBSD-informed two-phase austenite/ferrite laminate structure (LS) model is developed with {110}-type habit planes (HP) in ferrite and a Kurdjumov-Sachs orientation relationship (KS-OR) between both phases. At the third scale (macroscale), the model considers a single ferritic primary grain as an aggregate of spherical two-phase laminate structure domains corresponding to the 24 KS-OR variants. The elasto-viscoplastic self-consistent scheme (EVPSC) is used through the Translated Fields (TF) method to obtain the effective behavior at this scale. The TF-EVPSC scheme is also used for an ensemble of primary ferritic grains to identify materials parameters from experimental tensile curves corresponding to as received and aged specimen. From EBSD measurements, the crystallographic data are thoroughly analyzed to physically feed the three-scale model. The results are discussed in terms of stress/strain responses in ferrite and austenite, which are correlated to the different KS-OR variants and HP orientations. The effect of primary ferritic grain crystallographic orientation and aging is also studied regarding monotonic and cyclic tests to study the possible origins of backstress in this material.

Experimental and numerical investigation on the pressure pulsation in reactor coolant pumps under different inflow conditions

A reactor coolant pump (RCP) is essential for transporting coolant in the primary loop of pressurized water reactors. In the advanced passive reactor, the absence of a long pipeline between the steam generator and RCP serves as a transition section, resulting in a non-uniform flow field at the pump inlet. Therefore, the characteristics of the pump should be investigated under non-uniform flow to determine its influence on the pump. In this study, the pressure pulsation characteristics were examined in the time and frequency domains, and the sources of low-frequency and high-amplitude signals were analyzed using wavelet coherence analysis and numerical simulation. From computational fluid dynamics (CFD) results, non-uniform inflow has a great effect on the flow structures in the pump's inlet. The pressure pulsation in the pump at the rated flow increased by 78–128.7% under the non-uniform inflow condition in comparison with that observed under the uniform inflow condition. Furthermore, a low-frequency signal with a high amplitude was observed, whose energy increased significantly under non-uniform flow. The wavelet coherence and CFD analysis verified that the source of this signal was the low-frequency pulsating vortex under the steam generator.

Solubility of Bi in Li–Pb Eutectic Alloy Between 508 and 623 K

Liquid lithium–lead (Li–Pb) eutectic alloy is one of the candidate liquid tritium breeding materials of fusion blanket and it contains bismuth (Bi) as an initial impurity or transmutation product from lead. The saturated solubility of Bi, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$S$ </tex-math></inline-formula> (wt.ppm), in Li–Pb was measured in the temperature range from 508 to 623 K. The measured results were plotted on the graph with <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$S$ </tex-math></inline-formula> (wt.ppm) versus 1/ <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$T$ </tex-math></inline-formula> . From these plots, a positive temperature dependence, where the solubility decreases as the temperature is lowered, is confirmed. The equation of the solubility of Bi in Li–Pb was obtained, too. In addition, the dissolution enthalpy of Bi in Li–Pb, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\Delta H_{\mathrm {sol}}$ </tex-math></inline-formula> , was quantified from this equation. The precipitation morphology of Bi in Li–Pb was also investigated. X-ray diffraction was used to analyze the precipitates collected from the surface of the liquid phase. As a result, the peak pattern of Li3Bi was found from the precipitate. Moreover, the amount of polonium produced when the cold trap is introduced to a Li–Pb loop of a fusion reactor was predicted from the solubility obtained in this study.

3D field data of flow structures in dead-ended, coolant loop reactor branch lines for CFD validation of thermal fatigue onset modeling

Thermal fatigue phenomena in normally stagnant branch lines connected to the primary loop of pressurized water reactors has been identified as a primary cause for at least ten pipe cracking events in the past decade. Current prediction and inspection of at-risk piping is based on semi-empirical models, derived on the basis of low-resolution experimental data. This data does not include sufficient information on contributing factors that lead to thermal boundary development and fluctuation in dead-ended branch lines. Without comprehensive, high-resolution data of the phenomena involved, CFD models attempting to account for the re-laminarization or quasi-laminar flow that persists in the branch line cannot be validated.Validated CFD models for dead-ended branch line flow conditions will greatly expand the predictive capabilities the method currently offers. To build the database necessary for related CFD validation campaigns, 3D high-resolution measurements were made in the branch line of an experimental facility built at the University of Michigan. The facility is a scaled coolant flow loop with a reduced main line diameter. Nuclear power plant coolant loop flow rates near 10 m/s were achieved for study. High-resolution, 3D velocity field data in the dead-ended branch line was obtained using Tomographic Particle Image Velocimetry (T-PIV) techniques. With this data as a basis for validation, CFD models such as Low-Re will be able to accurately predict flow phenomena in dead-ended branch line simulations where the branch line diameter, opening geometry, or piping component (elbow) locations are altered.

Optical Flowmeter for Measuring the Coolant Flow Rate in the Primary Loop of a Nuclear Reactor

Optical Flowmeter for Measuring the Coolant Flow Rate in the Primary Loop of a Nuclear Reactor

Analysis of activated colloidal crud in advanced and modular reactor under pump coastdown with kinetic corrosion

The analysis of rapid flow transients in Reactor Coolant Pumps (RCP) is essential for a reactor safety study. An accurate and precise analysis of the RCP coastdown is necessary for the reactor design. The coastdown of RCP affects the coolant temperature and the colloidal crud in the primary coolant. A realistic and kinetic model has been used to investigate the behavior of activated colloidal crud in the primary coolant and steam generator that solves the pump speed analytically. The analytic solution of the non-dimensional flow rate has been determined by the energy ratio β. The kinetic energy of the coolant fluid and the kinetic energy stored in the rotating parts of a pump are two essential parameters in the form of β. Under normal operation, the pump's speed and moment of inertia are constant. However, in a coastdown situation, kinetic damping in the interval has been implemented. A dynamic model ACCP-SMART has been developed for System Integrated Modular and Advanced Reactor (SMART) to investigate the corrosion due to activated colloidal crud. The Fickian diffusion model has been implemented as the reference corrosion model for the constituent component of the primary loop of the SMART reactor. The activated colloidal crud activity in the primary coolant and steam generator of the SMART reactor has been studied for different equilibrium corrosion rates, linear increase in corrosion rate, and dynamic RCP coastdown situation energy ratio β. The coolant specific activity of SMART reactor equilibrium corrosion (4.0 mg s−1) has been found 9.63 × 10−3 μCi cm−3, 3.53 × 10−3 μC cm−3, 2.39 × 10−2 μC cm−3, 8.10 × 10−3 μC cm−3, 6.77 × 10−3 μC cm−3, 4.95 × 10−4 μC cm−3, 1.19 × 10−3 μC cm−3, and 7.87 × 10−4 μC cm−3 for 24Na, 54Mn, 56Mn, 59Fe, 58Co, 60Co, 99Mo, and 51Cr which are 14.95%, 5.48%, 37.08%, 12.57%, 10.51%, 0.77%, 18.50%, and 0.12% respectively. For linear and exponential coastdown with a constant corrosion rate, the total coolant and steam generator activity approaches a higher saturation value than the normal values. The coolant and steam generator activity changes considerably with kinetic corrosion rate, equilibrium corrosion, growth of corrosion rate (ΔC/Δt), and RCP coastdown situations. The effect of the RCP coastdown on the specific activity of the steam generators is smeared by linearly rising corrosion rates, equilibrium corrosion, and rapid coasting down of the RCP. However, the time taken to reach the saturation activity is also influenced by the slope of corrosion rate, coastdown situation, equilibrium corrosion rate, and energy ratio β.

An ab initio molecular dynamics investigation of the thermophysical properties of molten NaCl-MgCl[formula omitted

Molten salts have many applications in the nuclear and solar energy industries for thermal storage and heat transfer applications. However, there is a knowledge gap in molten salt thermophysical properties which hinders the technical readiness level of molten salt applications, especially in the nuclear industry. A common method of investigating new materials is through ab initio Molecular Dynamics (AIMD) simulations which is an effective tool to investigate structural and thermophysical properties at realistic temperatures. NaCl-MgCl2 is an inexpensive salt that is a good candidate for use as a heat transfer medium in solar power applications or in the secondary loop of a nuclear reactor. In this article, the thermophysical properties of NaCl-MgCl2 are computed via AIMD calculations to supplement the limited experimental studies in the literature. A wide range of compositions and temperatures for the pseudo-binary NaCl-MgCl2 were used to calculate the density, heat capacity, compressibility, enthalpy of mixing, and volumetric thermal expansion coefficient. AIMD is shown to accurately model the densities of molten NaCl-MgCl2 as there is good agreement with the available literature. This work observed a transition to a monotonic increase of the density with respect to MgCl2 composition occurring above 1100 K. The heat capacity values increase uniformly with respect to concentration of MgCl2 at a rate of 2.85 J/mol-K per 10 mol% of MgCl2. Select thermophysical properties are fit to a Redlich-Kister expansion for utilization in multiphysics simulations.

Nuclear-Magnetic Flowmeter-Relaxometer for Monitoring the Flow Rate and State of the Coolant in the First Loop of the Nuclear Reactor of a Moving Object

Nuclear-Magnetic Flowmeter-Relaxometer for Monitoring the Flow Rate and State of the Coolant in the First Loop of the Nuclear Reactor of a Moving Object

Flowing Refractometer for Feed Water State Control in the Second Loop of Nuclear Reactor

The necessity to control the feed water state in the second loop of a nuclear power plant nuclear reactor is justified. The different methods of the state control of flowing water in the pipeline are reviewed. It has been established that controlling the feed water state should not result in irreversible changes in its chemical composition and physical structure. A change in the composition or structure of feed water leads to a change in its heat capacity. The heat transfer deteriorates, the production of electrical energy in the installation decreases, and the additional release of heat into the atmosphere increases. This process also occurs during the heat capacity changes for other reasons. The method for controlling the feed water heat capacity by measuring the value of the refractive index n is developed. The design of a flow-through refractometer based on the total internal reflection for control of the feed water state in the stream is made. The dependence of the heat capacity change of feed water from the refractive index is established. The results of research on different types of water are presented.

Computational fluid dynamics simulation of internally circulating fluidized bed reactor for dry reforming of methane

Internally circulating fluidized bed reactor (ICFB) is the reactor that combines the conventional circulating fluidized bed (CFB) reactor system, including riser, downer, loop seal and cyclones, into a single compact reactor column. The ICFB reactor column is separated into two sections (riser and downer) by baffles and is linked together via connecting ports. The solid particles are circulated in the ICFB reactor by unequal fluidizing velocities in the riser and downer sections. The solid particles in the downer section are operated with low fluidizing velocity in the bubbling fluidization regime. In the riser section, the solid particles are transferred to the downer section because of the high fluidizing velocity. In this study, the carbon dioxide reforming or dry reforming reaction of methane was investigated. Dry reforming is a reaction between carbon dioxide and methane, which are the major greenhouse gases, for producing syngas, hydrogen and carbon monoxide. The hydrodynamics and dry reforming reaction with different reactor designs were explored using two-dimensional Eulerian–Eulerian computational fluid dynamics simulation (CFD). The ICFB reactor with no elevation gas distributor, 90 cm of draft tube height, 2 cm of loop-seal length and 1 cm of gas outlet diameter gave the highest methane conversion of 15.4%. The analysis of variance (ANOVA) indicated that the length of loop-seal and the gas outlet diameter gave substantial negative and positive effects on methane conversion, respectively. The interaction between length of loop-seal and gas outlet diameter also had a significant effect on the methane conversion. Although the ICFB reactor in this study gave the methane conversion slightly lower than referenced literature conventional CFB reactor, the ICFB reactor is the reactor that can reduce construction cost and simplify complex reactor operation.

Tellurium segregation-induced intergranular corrosion of GH3535 alloys in molten salt

The catastrophic intergranular cracking had been observed in the low-Cr alloy components in the primary loop of molten salt reactor. Such cracking was related to the coupling effect of fission product tellurium (Te) and fuel salts, whose atomic-level mechanism has puzzled researchers for many decades. In this work, combining multi-scale characterization techniques and theoretical calculations, it was found that Te segregation at grain boundaries can reduce the diffusion barrier of active element along grain boundaries and cause the nano-voids, which further evolve into the intergranular cracking.

Development and Research of New Media Jet Aeration Scheme in a Loop Bioreactor Producing Microbiological Products

Abstract Jet aeration loop reactors are widely used in the chemical industry due to their high mixing intensity, possibility of significant saturation of medium with gases, and simplicity of scaling the processes when passing from laboratory equipment to industrial designs. However, to ensure the necessary amount of air in standard jet aeration schemes with ejectors, high pressure of the medium (up to 6 bars) is required in the reactor loop. This paper presents a newly developed scheme of jet aeration, based on mixing the pressure flows of medium and air supplied to the ejector by individual blowers. Experiment results of the new water aeration scheme showed that, for the formation of suspended matter in the ejector with an air content of 2–20 l per 1 l of water, a pressure of 0.3 bar in the reactor loop at an air pressure of 1 bar is sufficient. This allows usage of low-pressure pump and compressor, which form the basis for the energy consumption reduction and creates prerequisites for lowering the shear loads on a crop. The medium aeration rate during experiments was 0.220–0.266 vvm. It was found that, for each mode of joint operation of pump and compressor, it is possible to maintain a constant aeration rate, which is not significantly affected by the flowrate regulation of medium in loop.