Supercritical Reactor Research Articles

Research and analysis of new generation nuclear reactors in the world

The research of new nuclear reactors is gaining urgent importance worldwide due to the need for continuous improvement of technologies to ensure safety, efficiency, and emissions reduction. This is crucial in the context of climate change and rapid technological development, which demand constant updating and improvement of nuclear energy. The objective of the study was to analyse next-generation reactors worldwide and identify their advantages and potential prospects for the future. The research utilized statistical, comparative, and analytical methods. The results of the analysis considered contemporary technological and safety parameters related to the operation of such reactors, including their ability to optimize fuel usage, enhance operational safety, and effectively manage radioactive waste. As a result of the study, fourth-generation nuclear reactors were analysed, including fast neutron reactors using gas cooling, very high-temperature reactors, reactors using sodium as a coolant, fast neutron reactors with lead cooling, reactors where the reaction occurs in molten salt, and supercritical water-cooled reactors. Each of these reactors has its unique features that make them distinctive in their application. For example, gas-cooled reactors have high productivity due to their ability to achieve high temperatures without significant pressure. On the other hand, molten salt reactors offer flexibility in using different types of fuel, including spent fuel, and can help reduce the level of radioactive waste through the use of special materials. During the analysis, it was noted that fourth-generation reactors, using various cooling and reaction-slowing technologies, are characterized by high efficiency, low accident risk, and the ability to produce stable electricity. Improved methods of reaction control open up new possibilities for the efficient production of electricity and increased safety in nuclear energy. The practical significance of the research lies in the opportunity to enhance modern electricity production technologies and ensure greater safety and efficiency in the field of nuclear energy

Noise term modeling of dynamic Monte Carlo using stochastic differential equations

Dynamic Monte Carlo (DMC) method recently became a relevant tool to perform kinetic calculations even on the scale of power plants and further to simulate coupled neutronics — thermal-hydraulics problems. DMC offers time-dependent, high fidelity calculations in very detailed and complex geometries in return for the high computational need. DMC provides stochastic results which is an indescribable property by the mathematics of the deterministic neutron transport. The stochasticity of the results also rises the unsolved problem of stability and convergence during the nonlinear, coupled DMC and thermal-hydraulics simulation. This paper aims to introduce the Stochastic Differential Equations (SDEs) in the field of Monte Carlo neutron transport for connecting the DMC method and differential equation formalism. The derivation of a Non-Analog Monte Carlo (NAMC) model is shown for the Stochastic Point-Kinetics equation (SPKe) to determine a noise model capable to approximate one trajectory of DMC. The theoretical framework allows the reader to observe analytic solutions for a random variate, the expectation and the variance of the reactor power. Also, analytic formulas are given for a simple system of coupled reactor physics and thermal-hydraulics without feedback. The analytic variance of the Monte Carlo simulated reactor power were compared to the calculations of GUARDYAN (GpU Assisted Reactor DYnamic ANalysis) code in a subcritical, a critical and a supercritical reactor models.

Initial Multidisciplinary Study of Oxidized Chromium-Coated Zirconium Alloy for Fuel Cladding of SCW-SMR Concept: Weight-Gain and Thermal Conductivity Measurements and Coating Cost Evaluation

One of the challenges of small modular reactors (SMRs) in comparison with large reactors is the greater difficulty in achieving high burnups in smaller cores. With greater neutron leakage through the periphery, a key factor is the neutron economy of the fuel cladding. However, all large supercritical water-cooled reactor (SCWR) concepts have employed neutron-absorbing stainless steels and nickel-based alloys in order to meet all the requirements in terms of corrosion and thermalhydraulics. In order to achieve higher burnups and extend the time between refueling in a SCW-SMR, the use of chromium-coated zirconium alloy as a potential fuel cladding candidate has been explored. Chromium coatings up to a few micrometers thick have shown improved oxidation resistance of zirconium-based claddings under operating conditions relevant to SCWR concepts. In this study, Zr-2.5Nb alloy (UNS R60904) from pressure tube samples was coated using a physical vapor-deposition (PVD) method. Oxidation tests were performed on coated samples at 500 °C and approximately 25 MPa in a refreshed autoclave. The effects of the oxide on heat transfer and hydraulic resistance are also discussed in this study. Last, but not least, this study evaluates the coating cost of the fuel cladding with chromium in a vacuum plasma spray process.

An improved frequency-domain model for the supercritical flow instability analysis in vertical tubes

In a supercritical water reactor, the density wave oscillation (DWO) may occur due to the dramatic changes in the density of the working fluid in the pseudo-critical region. Based on the frequency-domain method, this study proposes an improved model to study DWO. First, the latest partition method is used in this study. Second, in the derivation of the transfer function in the light fluid region, the density change of the fluid is considered. Moreover, the different states of the outlet fluid are handled differently. We use the Nyquist criterion to determine the flow stability and obtain the boundary heat flux (BHF) of different working conditions. After model verification, it is proved that the model accuracy is improved. We analyze the influence of pressure, inlet enthalpy, mass flow rate, inlet throttling coefficient, and tube parameter on the instability boundary. At a certain inlet enthalpy, a minimum BHF of flow instability exists.

Construction of a feedback control system based on CFD simulations for the 64-element Canadian SCWR

The knowledge of the dynamic behaviors of the supercritical water in the fuel bundle of the nuclear reactor is necessary for the design of the control system. For this purpose, the transient full-scale computational fluid dynamics (CFD) simulations are used to obtain the dynamic input and output data sets of the 64-element Canadian supercritical water-cooled reactor (SCWR) in this study. Subsequently, the linear dynamic models of the reactor are obtained by the system identification technique and validated by the comparison with the results from CFD simulations. Based on the linear dynamic models, a feedback control system consists of three PI/PID controllers is constructed. The objective is to regulate the system back to the design point when the reactor is subjected to disturbances. Therefore, the performance of the designed feedback control system is also evaluated in this work. It shows that the designed control system can timely minimize the deviations and return the reactor back to the desired operating condition.

Technological readiness of alternative reactor concepts

Abstract. Some alternative reactor concepts are frequently discussed in the literature as relevant for the disposal of high-level radioactive waste. We present results on the technological readiness of and research and development on those reactor concepts from a project commissioned by the Federal Office for the Safety of Nuclear Waste Management on novel reactors (Pistner et al., 2023). The focus is on 7 technology lines and 10 specific reactor concepts within those lines. These are sodium-cooled fast reactors (SFRs), lead-cooled fast reactors (LFRs), gas-cooled fast reactors (GFRs), very high-temperature reactors (VHTRs), super-critical water-cooled reactors (SCWRs), molten-salt reactors (MSRs) and accelerator-driven systems (ADSs). For all technology lines considered, extensive research and development work has been taking place for several decades and in some cases since the middle of the last century. Depending on the technology line, technical test stands for individual phenomena have been built and operated and so have smaller experimental reactors and larger demonstration reactors. Nevertheless, until today no commercially competitive reactor concept exists in the field of alternative reactor concepts. The most extensive technical experience is available for the SFR and VHTR technology lines. However, proof of reliable operation under economic boundary conditions is still required for SFRs and VHTRs. For other technology lines, neither has technical feasibility in the form of a demonstration reactor been demonstrated so far (LFR, MSR) nor are more extensive findings from smaller experimental reactors available (GFR, SCWR, ADS). To plan, license, construct and operate such experimental and demonstration reactors, a period of at least 1 to 2 decades must be assumed for each reactor project and probably substantially more based on historical experience. It has been shown that developers' schedules are often characterized by overly optimistic assumptions, that developments are delayed by years or even decades, and that in many cases specific approaches are discontinued completely because the underlying technological difficulties could not be overcome. Thus, the time still required for the development of alternative reactor concepts is probably in the range of several decades. Against this background, it cannot be assumed that such reactor concepts will be used on a relevant scale by the middle of this century. Additionally we will present general results of this study regarding criteria such as safety, fuel supply and waste disposal, proliferation, and costs and point out advantages and disadvantages of those systems over light-water reactors.

Evolution of Recrystallization Texture in A286 Iron-Based Superalloy Thin Plates Rolled via Various Routes

The A286 iron-based superalloy has wide-ranging applications in replacing expensive nickel-based and cobalt-based superalloy to manufacture the turbine disk as well as the pipelines and valves of the fourth-generation supercritical water reactor (SCWR) working below 650 °C. The recrystallization texture importantly affects the mechanical properties and oxidation resistance of superalloys. However, very few investigations are available on the recrystallisation texture of A286 alloy. The current work reports the texture of A286 alloy fabrication using various rolling routes, including one-stage cold rolling with the rolling rates of 83% (Route I) and 91% (Route II), and two-stage cold rolling with the rolling rate of 83% + 50% (Route III). Route III is preferentially recommended to manufacture A286 alloy thin plates due to the obviously reduced work hardening as well as the weakened recrystallization texture and anisotropy; moreover, compared with other routes, the recrystallized grain size of Route III did not significantly coarsen. We also revealed the mechanism for the effect of cold rolling textures on the final recrystallization texture and the role of the deformation twin in {110} <115> grain. In this study, recrystallization texture develops through two independent mechanisms related to different deformation microstructures, including the recrystallization texture inherited from deformation texture and the recrystallization texture depended on deformation twins.

Technology Development for Hydrogen Production Using Nuclear Energy

Nuclear energy is a low-carbon energy source which can provide both electricity and high-temperature heat. Nuclear-based hydrogen production offers a unique opportunity to mitigate carbon emissions. Canadian Nuclear Laboratories (CNL) has been actively developing technologies for nuclear hydrogen production. This poster will highlight CNL’s capability and achievements on the Cu-Cl thermochemical process (CNL trademarked as HCuTECTM) and Solid oxide electrolysis (SOE).HCuTECTM has four main steps in the process: Electrolysis, Separation, Hydrolysis and Thermolysis. The maximum temperature of the HCuTECTM process is 530 oC, which is lower than other thermochemical processes and makes its coupling with some small modular reactors and the supercritical water nuclear reactor ideal since these reactors can supply heat above 600 oC. CNL has extensively studied these individual steps and demonstrated an integrated laboratory system (shown in Figure 1) with a hydrogen production capacity of 100 g/d. CNL is interested in scaling up the process with industrial partners.SOE is also being researched and advanced for hydrogen and clean fuel production at CNL. Current effort is focused on the materials development for oxygen conducting cells. Several CNL cells were made with in-house materials and tested on single cell stations. Preliminary results of CNL cells showed very promising performance for hydrogen production. In the meantime, CNL is assessing the economic benefit of different options to integrate SOE with high-temperature nuclear reactors for both hydrogen and clean fuel production. Figure 1

Research on prediction and analysis of supercritical water heat transfer coefficient based on support vector machine

In order to better perform thermal hydraulic calculation and analysis of supercritical water reactor, based on the experimental data of supercritical water, the model training and predictive analysis of the heat transfer coefficient of supercritical water were carried out by using the support vector machine (SVM) algorithm. The changes in the prediction accuracy of the supercritical water heat transfer coefficient are analyzed by the changes of the regularization penalty parameter C, the slack variable epsilon and the Gaussian kernel function parameter gamma. The predicted value of the SVM model obtained after parameter optimization and the actual experimental test data are analyzed for data verification. The research results show that: the normalization of the data has a great influence on the prediction results. The slack variable has a relatively small influence on the accuracy change range of the predicted heat transfer coefficient. The change of gamma has the greatest impact on the accuracy of the heat transfer coefficient. Compared with the calculation results of traditional empirical formula methods, the trained algorithm model using SVM has smaller average error and standard deviations. Using the SVM trained algorithm model, the heat transfer coefficient of supercritical water can be effectively predicted and analyzed.

Exploring detailed microcosmic mechanisms of sodium chloride deposition in supercritical water reactors from the aspect of solvation structure: A combination of MD and QM simulations

The dissolution characteristics and microscopic dissolution process of salt in supercritical water (SCW) are the fundamental influence of scale formation. In this paper, molecular dynamics (MD) and quantum mechanics (QM) were used to study the solubility and solvation structure of sodium chloride in supercritical water. The potential mean force (PMF), radial distribution function (RDF) and coordination number (CN) indicate that the solubility of sodium chloride in SCW decreases with the increase of temperature and the decrease of density. As the system density decreases from 0.3 g/cm3 to 0.1 g/cm3 and the temperature increases from 600 K to 900 K, the CN of water molecules in the solvation shell decreases from 2.89 to 1.01, which is consistent with the trend of the experimental results. The electrostatic potential distribution and intermolecular interaction of the solvation shell around Na+ were obtained by QM method. The solution free energy of Na+–3H2O in implicit solvent was obtained. In the condition range of this study, when the temperature is the lowest and the dielectric constant is the largest, the free energy of dissolution is the smallest, which is −0.0991 a.u. This study will lay a foundation for the study of salt deposition in SCW.

Results and lessons learned from the Generation IV SCWR-FQT comprehensive Monte Carlo computational benchmark

A joint European Canadian Chinese development of a supercritical water-cooled small modular reactor technology has been in progress since September 2020 in the framework of a Horizon 2020 project called ECC-SMART. A specific work package has been dedicated to studying the design- and safety-related neutronic parameters and reactor physics behavior of the SCW-SMR to support the pre-conceptual design process. Three Monte Carlo codes, viz., MCNP, OpenMC, and Serpent, were selected for pre-conceptual design applications and code-to-code comparison within the Gen-IV SCWR-FQT reactor physics computational benchmark. The effective multiplication factor, the axial power distribution within the fuel, the axial three-group neutron flux distribution, and the spatial distribution of the energy deposition due to neutron and photon interactions were determined. In this paper, results and lessons learned from this study are presented, and useful considerations are summarized to provide guidance in obtaining consistent results among the three Monte Carlo codes.

Investigation of Stress Corrosion Cracking Resistance of Irradiated 12Cr Ferritic-Martensitic Stainless Steel in Supercritical Water Environment

The supercritical water-cooled reactors (SWCR) belong to Generation IV of reactors. These reactors have a number of advantages over currently operating WWERs and PWRs. These advantages include higher thermal efficiency, a more simplified unit design, and the possibility of incorporating it into a closed fuel cycle. It is therefore necessary to identify candidate materials for the SWCR and validate the safety and effectiveness of their use. 12Cr ferritic-martensitic (F/M) stainless steel is considered a candidate material for SWCR internals. Radiation embrittlement and corrosion cracking in the primary circuit coolant environment are the main mechanisms of F/M steels degradation during SWCR operation. Here, the stress corrosion cracking (SCC) in supercritical water at 390 and 550 °C of 12Cr F/M steel irradiated by neutrons to 12 dpa is investigated. Autoclave tests of specially designed disk specimens in supercritical water were performed. The tests were carried out under different constant load (CL), temperature 450 °C, and pressure in autoclave 25 MPa. The threshold stress, below which the SCC initiation of irradiated 12Cr F/M steel does not occur, was determined.

The influence of duct wall shape on mass redistribution in supercritical water-cooled reactor fuel assembly

Based on the fuel assembly of supercritical water-cooled reactor, a 19-rod bundle with triangular arrangement is established and different shapes of duct wall are used to adjust the hydraulic diameter of the external sub-channel. The Reynolds stress model (RSM) and enhanced wall treatment were employed to simulate the fluid flow and heat transfer phenomena of supercritical water and the mass redistribution in fuel assembly. The result shows that the hydraulic diameter difference between internal and external sub-channels and the physical properties of supercritical water are the main factors affecting the mass redistribution phenomenon. The mass redistribution can be effectively controlled by adjusting the hydraulic diameter of the external sub-channel. As the hydraulic diameter of the external sub-channel decreases, the fluid accumulation to the external sub-channel gradually weakens and then reverses to the internal sub-channel accumulation. Meantime, comparing with the primary model, the overall heat transfer effect of internal and external sub-channel can be balanced and kept at a high level by weakening the mass redistribution. Besides, the circumferential non-uniformity of the outermost fuel rod is affected by the local hydraulic diameter and the mass redistribution. Mass redistribution is the dominant factor and the local hydraulic diameter can only start to affect the circumferential wall temperature until the mass redistribution is weakened to a certain extent.

CFD prediction of heat transfer at supercritical pressure with rough walls: Parametric analyses and comparison with experimental data

In this paper, a low-Reynolds number turbulence model developed by the authors in past activities for rough walls is used for predicting heat transfer at supercritical pressure in the presence of different degrees of surface finishing. The model, able to reproduce the typical trends of friction factors from classical data reported by Nikuradse and summarised in the Moody diagram, is based on a simple-minded description of the effect of wall protrusions through the boundary layer on turbulence production.Though prior validation of the model only on the basis of friction factor data did not assure any basis for achieving accuracy in heat transfer prediction, a sensitivity analysis is firstly presented in order to characterise the obtained predictions at variable values of the roughness parameter, in particular concerning the possible suppression of deteriorated heat transfer by roughened surfaces. These analyses are extended to different fluids, making use of a fluid-to-fluid similarity theory recently proposed by the authors in order to establish similar boundary conditions and predicted phenomena.The results obtained by these analyses can be considered interesting, especially in view of the design of supercritical water-cooled nuclear reactors; however, an assessment against experimental data was obviously necessary. Experimental carbon dioxide data published in a very recent archival paper were thus addressed and were found useful in this regard. Considering these data allowed to extend the above analysis to provide confirmation of the promising features of the model in comparison with wall temperature values obtained with different boundary conditions. The model here described appears promising not only for its capability to predict experimentally measured effects, but also for the perspective to be used in the study of the behaviour of purposely roughened surfaces reducing the probability of occurrence of deteriorated heat transfer.

Numerical simulation of the fluid flow and heat transfer of the supercritical water in the 64-element Canadian SCWR fuel bundle

ABSTRACT Canada has participated in the Generation IV nuclear energy systems focusing on the Supercritical Water-cooled Reactor (SCWR) system. The 64-element fuel rods channel is used in the Canadian SCWR. There are few publicly available experimental studies for the flow in the fuel channels with multiple fuel rods. To date, the CFD simulation reported in open literature for rod bundle flow in the 64-element SCWR is relatively scarce. The cladding surface temperature is of key importance in assessing the safety of the reactor. Since the inhomogeneities in the bundle cross-section can present complex flow phenomena, a CFD study can provide substantial insight into the flow physics. In this work, the full-scale CFD simulation of the supercritical water flow in the 64-element rod bundle was performed. The results suggest the possibility of the presence of gap vortices in the flow subchannels. Higher streamwise velocities and normal Reynolds stresses always exist at the center subchannel regions. The circumferential cladding surface temperature distribution is extremely non-uniform and there is a large difference between the maximum cladding surface temperatures for different fuel rods.

The mechanism of abnormal corrosive behavior of metallic materials in the pseudocritical region of supercritical water

To study the corrosive behavior of metal materials in the pseudocritical region of supercritical fluid and explain the abnormal mechanism in corrosive within the pseudocritical region, a corrosive experimental platform for supercritical fluids was built. Three different candidate materials for Supercritical Water Reactor (i.e. Q235 low carbon steel, 316 L austenite stainless steel and 3Cr13 martensitic steel) were selected and reacted for 200 h in the pseudocritical region (375.6 ℃, 22.5 MPa) and gas-like region (405.6 ℃, 22.5 MPa) respectively. The results show that 316 L exhibits abnormal corrosive exacerbation in the pseudocritical region with a corrosive rate 1.67 times that of the gas-like region. Considering the special variation of physical properties in pseudocritical region, the enhanced kinetic effect is further analyzed. It is found that the migration of corrosive media such as oxygen to the material matrix is accelerated, which aggravates the corrosive products shedding and freeing. The present study can provide a basis for improving the anti-corrosive performance of materials under supercritical fluid environments and providing a safe operation of the Supercritical Water Reactor.

Experimental investigation on kinetic model with reaction heat for coal gasification in supercritical water

An accurate kinetic model, including the reaction pathway and reaction heat, is crucial to reactor scale-up and process optimization for the supercritical water gasification (SCWG) of coal. However, the reported kinetic models seldom satisfy the mass and energy conservation law due to the proposed inappropriate reaction pathway and difficult determination of reaction heat in supercritical water (SCW). This work focuses on determining the reaction heat and establishing a kinetic model based on mass and energy conservation law. A reaction pathway combined with reaction heat determination by a lumped method was proposed to establish a kinetic model and to quantify the reaction heat of each reaction in the coal gasification process. The steam reforming reaction I and II in the model have the highest activation energy, 59.629 ± 10.281 kJ/mol and 63.249 ± 3.937 kJ/mol, respectively. Reaction heats of the reaction between volatile and SCW and the steam reforming reactions exceed + 600 kJ/mol. Moreover, due to the combined influence of endothermic and exothermic reactions, the entire gasification process shift from endothermic to exothermic as the gasification proceeds. This work provides a more accurate mathematical model for the SCW reactor.

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

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

Review of Euratom projects on design, safety assessment, R&D and licensing for ESNII/Gen-IV reactor systems

Five Euratom projects launched since 2017 in support of the development of ESNII/Generation-IV reactor systems are briefly presented in the paper in terms of key objectives, results, and recommendations for the future. These projects focus on various aspects of the following ESNII/Generation-IV systems: Sodium Fast Reactor, Gas Cooled Fast Reactor, Supercritical Water Cooled Reactor, and Molten Salt Fast Reactor. The paper does not consider EU projects focused on the Gen-IV reactor technologies based on the use of heavy metals as a coolant because these projects are reviewed in a different paper.

Atomistic insights into stress corrosion cracking of α-Fe in supercritical water: The coupling effect of hydrogen embrittlement and intergranular corrosion

Mechanistically understanding stress corrosion cracking (SCC) of Fe-based alloy under extremely high-temperature and high-pressure conditions is a key challenge for designing and optimizing structural materials of the supercritical water (SCW) reactors. The mechanism of SCC of α-Fe in supercritical water has been investigated by reactive force-field molecular dynamics simulations, and it is obtained that SCC is caused by the coupling effect of hydrogen embrittlement and intergranular corrosion. Hydrogen atoms in metals induce vacancy formation when the loaded stress is larger than the threshold stress, resulting in forming stable vacancy-hydrogen clusters. These stable vacancies pin grain boundary (GB) and act as stress concentrators under sufficient tensile stress, leading to the generation of more vacancies. A large number of vacancies in GB promote intergranular corrosion which directly results in materials failure.