Supercritical Reactor Research Articles

Four equation k-omega based turbulence model with algebraic flux for supercritical flows

Hydrocarbon and nuclear fuels will contribute substantially to the growing global energy portfolio for some time to come. At the same time atmospheric carbon levels and economic concerns are increasingly shaping the power generation environment. Supercritical fluid technologies are being investigated to address both concerns. For example, direct fired supercritical CO2 power cycles offer key advantages for carbon capture and thermal efficiency. Additionally, the supercritical water reactor Generation IV nuclear reactor allows higher power densities and more efficient operation than current pressurized or boiling water reactors. As these technologies are developed, a need exists for accurate, yet cost-effective modeling and simulation solutions to assist with design and safety calculations. Supercritical fluid properties are strongly dependent on temperature which results in a number of unique flow features. Among these features are the complex buoyancy and buoyancy-turbulence interaction behaviors exhibited by supercritical flows that are not found in liquid or gas flows. Therefore, standard turbulence models with current model coefficient values are not likely to be applicable to supercritical fluid flows.The subject of this paper is efficient modeling and simulation of supercritical flows to support nascent technologies growing to maturation and operational deployment. We present a framework for developing Reynolds-Averaged Navier-Stokes turbulence models specifically equipped for the challenges of supercritical flows. A novel formulation of the algebraic heat flux model of the buoyancy production of turbulence term is used with a traditional shear stress transport model. To produce a new turbulence model for supercritical channel flows, the empirical coefficients of the resulting four equation model were calculated from data previously published by other authors regarding upward supercritical flow through heated pipes. The presented SST kt-ωt approach was validated against heated tube experiments to show predictive capabilities in moderate flow conditions, where buoyancy effects are important but not dominating of inertial effects.

Stability analysis of the Supercritical Water Reactor by means of the root locus criterion

Abstract The Supercritical Water Reactor (SCWR) is a concept for an advanced nuclear reactor operating at high temperature (500 °C average core outlet temperature at nominal power) and at high pressure (25 MPa), which give the SCWR a thermal efficiency of about 45%. However, due to the strong variability of the water properties near the thermodynamic pseudocritical point, concerns are raised towards thermal-hydraulic instabilities. A simulation tool was developed in Matlab® from the perspective of linear systems, aimed at investigating the reactor stability and identifying potential regions of instability through a consolidated and relatively simple approach. A frequency-domain stability analysis of the SCWR is carried out with the root locus criterion, characterizing the system stability features over its entire operating power interval. The impact of the coolant flow rate on the stability is also studied. The results show that the system is stable over the whole investigated operational range. Finally, the dynamic behavior of the SCWR is compared to the Boiling Water Reactor (BWR), pointing out significant differences due to the different working points and design features of the two reactors. The results of this study could be a starting point for further research on the SCWR, providing the designers with important feedbacks for the optimization of the SCWR coolant circuit.

Behavior and susceptibility to stress corrosion cracking of a nickel‐based alloy in superheated steam and supercritical water

Stress corrosion cracking (SCC) behavior of Alloy 617 was studied to evaluate its suitability for Generation IV supercritical water reactor concept. A series of slow strain rate tensile (SSRT) tests at a constant strain rate of 5 × 10−7 s−1 were carried out in superheated steam and supercritical water over a pressure range of 0.1–25 MPa at 650 °C. SSRT test in dry N2 gas was also performed to compare the results in non‐corrosive environment with corrosive medium environment. Intergranular cracks were observed for all specimens regardless of the test environment. Alloy 617 showed susceptibility to SCC at the tested experimental conditions. The dominance of intergranular SCC was observed on the gage surface and fracture surface by using scanning electron microscope analysis. The effect of varying environmental conditions on SCC susceptibility is further discussed.

A new assessment method for the evaluation of supercritical heat transfer correlations, particularly with regard to the “multiple/no solutions” problem

The heat transfer coefficient is an essential measure in the predesign of supercritical water-cooled reactors. At supercritical pressures, the distinct heat transfer modes, normal, improved, and deteriorated, differ from those at subcritical pressures, resulting in unusual behavior of the heat transfer coefficient. Over 30 correlations have been published in the past decades in order to predict this abnormal behavior at supercritical pressures. Due to unsatisfactory results, in recent years the trend has been to propose more complex and implicit correlations. In addition, various efforts have been made to find the most accurate correlation for all three heat transfer modes. However, using different assessment methods, namely the recalculation method and the prediction method, contradictory results have been obtained.As a result, this paper presents a new assessment approach that evaluates the heat transfer correlations with regard to prediction accuracy as well as numerical stability. Based on the proposed assessment method, 11 correlations from the literature are evaluated against a database that contains more than 12,000 experimental data points. The results show, that all implicit correlations have multiple and/or unfeasible solutions. Furthermore, a detailed assessment of the prediction accuracy of the correlations indicates the need for more accurate correlations, especially in the vicinity of the pseudo-critical point.

Hydraulic resistance of in-tube cooling supercritical water accompanying out-tube pool boiling

The experiment was conducted by immersing a smooth horizontal tube in a pool tank to simulate the flow condition of Passive Residual Heat Removal System (PRHRS) in a SuperCritical Water-cooled Reactor (SCWR). Hydraulic resistance and friction factor of in-tube cooling supercritical water accompanying out-tube pool boiling were investigated in this study with test pressure ranging from 23 to 28 MPa and mass flux ranging from 600 to 1000 kg·m−2·s−1. The influence of pressure and mass flux on pressure drop in adiabatic and cooling flows was analyzed. This paper also discussed the effect of deceleration in the cooling flow and assessed various friction-factor correlations by employing the experimental data. Results showed that the friction factor of the adiabatic flow exists a steep “pit” approaching to pseudocritical region. A noticeable “Λ-shaped” profile was observed in the vicinity of the pseudocritical temperature, due to deceleration-effect of frictional pressure drop in the cooling flow. The deceleration factor of supercritical cooling flow led to the axial fluid element shrinkage and radial bulk fluid velocity parabolic distribution. Taking into account the effect of deceleration-factor, a modified correlation was proposed for in-tube cooling supercritical water accompanying out-tube pool boiling, of which the average error and root mean square error are −2.51% and 15.28% respectively.

Thermal-hydraulic analysis of a novel design super critical water reactor with Al2O3 nanofluid as a coolant

In this study the thermal-hydraulic analysis of the SCWR reactor with a new geometric design is discussed. In the new design, coolant and moderator circuits are separated and nanofluid has been used as coolant of the reactor core. In the performed sub-channel analysis, the conservation equations of mass, momentum and energy are solved numerically. Since in the new geometry the coolant and moderator circuits are separated, compared to previous plans, using the nanofluids with higher mass fraction of nanoparticles is applicable. In this article, the thermal-hydraulic analysis is performed for various mass fractions of Al2O3 nanoparticle and moderator velocities. Calculations show that the core outlet temperature grows up with increasing nanoparticle mass fraction. As it was expected, by increasing the nanoparticle mass fraction the core pressure drop will be raised, but in the proposed design this pressure drop is much lower.

A new method on fluid-to-fluid scaling for heat transfer in tubes at supercritical pressures

Supercritical fluid heat transfer has great significance for the thermal-hydraulic design and safety performance of the SCWR (Super-Critical Water-cooled Reactor). Experimental studies on heat transfer at prototypical conditions of SCWRs would suffer a huge cost on safety and economy, which strongly limits the research of supercritical water heat transfer. One solution is to adopt model fluid technique which has been applied in thermal and fluids engineering widely. Therefore, finding the equivalent relationship between the experimental data in model fluid and those in prototype fluid becomes a key issue. It means a fluid-to-fluid scaling method is necessary for converting experimental data obtained by using model fluid into the prototype fluid. In the present study, from the perspective of thermo-physical properties similarity and conservation equations similarity, considering the comprehensive influence of the experimental conditions and fluid types, a theoretical study on fluid-to-fluid scaling is performed and most importantly, a new scaling method is proposed. The new scaling method adopts a modification on thermo-physical properties. Several dimensionless numbers are obtained according to the dimensionless momentum and energy governing equations. A new dimensionless number En is derived to present the comprehensive influence of experimental conditions and fluid types. According to the theoretical analysis, an entire set of scaling laws applied for supercritical fluid heat transfer in circular tubes was obtained. Based on the heat transfer data of supercritical R134a, the new scaling method was validated, indicating the feasibility and accuracy of the proposed new scaling method.

Irradiation Issues and Material Selection for Canadian SCWR Components

The Canadian super critical water-cooled reactor (SCWR) concept requires materials to operate at higher temperatures than current generation III water-cooled reactors. Materials performance after radiation damage is an important design consideration. Materials that are both corrosion resistant and radiation damage tolerant are required. This paper summarizes the operating conditions including temperature, neutron flux, and residence time of in-core Canadian SCWR components. The focus is on the effects of irradiation on in-core components, including those exposed to a high neutron flux in the fuel assembly, the high pressure boundary between coolant and moderator, as well as the low-temperature, low-flux calandria vessel that contains the moderator. Although the extreme conditions and the broad range of SCWR in-core operating conditions present significant materials selection challenges, candidate alloys that can meet the performance requirements under most in-core conditions have been identified. However, for all candidate materials, insufficient data are available to unequivocally ensure acceptable performance and experimental irradiations of candidate core materials will be required. Research programs are to include out-of-pile tests on un-irradiated and irradiated alloys. Ideally, in-flux studies at appropriate temperatures, neutron spectrum, dose rate, duration, and coolant chemistry will be required. Characterization of the microstructure and the mechanical behavior including strength, ductility, swelling, fracture toughness, cracking, and creep on each of the in-core candidate materials will ensure their viability in the Canadian SCWR.

The status of the research on the heat transfer deterioration in supercritical fluids: A review

Nowadays, both experimental and computational research on the turbulent convective heat transfer to supercritical fluids is particularly active, especially because the actual poor comprehension and prediction of the possible heat transfer deterioration is limiting the design of new promising engineering applications. In this review, such applications, among which supercritical water-cooled nuclear reactors, supercritical CO2 power generation cycles, and oxygen/methane-fuel rocket engines, are firstly introduced. Then, after a phenomenological description of the heat transfer deterioration, the status of the research is analysed in details, highlighting the major advantages and limitations of both experimental and computational studies performed so far. The review demonstrates that experimental research is mostly focused on finding simple heat transfer correlations rather than detailed models. Also detailed numerical insight of the problem is still almost unexplored. The main conclusion is that new approaches, possibly integrating extensive experiments and computations, are needed to shed new light on the problem of heat transfer to supercritical fluids.

Fluid-to-fluid scaling of heat transfer to mixed convection flow of supercritical pressure fluids

Fluid-to-fluid scaling for supercritical heat transfer can effectively reduce the difficulty and cost of heat transfer experiments in supercritical boilers and supercritical water reactors and can reduce the number of experiments by converting experimental data of the model fluid to the prototype fluid in organic Rankine cycles. Currently, most existing scaling methods are only suitable for forced convection, while few are developed for mixed convection where buoyancy significantly affects the heat transfer. This paper attempts to extend the applicability of scaling method to mixed convection with the aid of computational fluid dynamic simulations. The scaling parameters were analyzed first and then the shear-stress transport k-ω model was used to analyze the supercritical heat transfer characteristics of water and R134a to provide further information for developing a dimensionless number. The results show that significant variations of properties and flow parameters occur in the layer of y+ = 5 to 100 and the axial velocity gradient in this layer changes in quite a similar manner to the wall temperature. Based on numerical results, the axial velocity gradient was used with a thermal resistance analogy to derive a new dimensionless number, Re−0.9πA, to scale the mass flux. Then, a set of fluid-to-fluid scaling laws were developed to predict the heat transfer to supercritical fluids. To validate the newly proposed scaling laws, well-developed correlations were used for forced convection flow and a direct validation method was developed for buoyancy-influenced flow. Results show that this new scaling method exhibits reasonable accuracy for both forced and mixed convection heat transfer with supercritical fluids.

Effect of ammonia on gasification performances of phenol in supercritical water

The solutions of phenol and the mixture of phenol and ammonia were gasified in a continuous supercritical water reactor to evaluate the influences of ammonia on the hydrogen production and the degradation pathways of phenol. The effects of temperature (560–640°C), reaction time (2–20s) and concentration of ammonia (500–2000mg/L) on gaseous distributions, gasification efficiencies and reactants removal efficiencies were investigated. In addition, the effects of ammonia on the kinds of intermediate products were analyzed. Results showed that the increasing temperature greatly promoted the hydrogen production and degradation of phenol. Longer reaction time gave a positive effect on gasification efficiencies in 10s, while the effect was little when the reaction time longer than 10s. Ammonia prevented the production of hydrogen and the degradation of phenol, which was mainly due to the generation of more stable nitrogen polycyclic compounds. The possible degradation pathways for the mixture of phenol and ammonia were proposed.

Optimization of the Canadian SCWR Core Using Coupled Three-Dimensional Reactor Physics and Thermal-Hydraulics Calculations

The Canadian supercritical water-cooled reactor (SCWR) design is part of Canada's Generation IV reactor development program. The reactor uses batch fueling, light water above the thermodynamic critical point as a coolant and a heavy water moderator. The design has evolved considerably and is currently at the conceptual design level. As a result of batch fueling, a certain amount of excess reactivity is loaded at the beginning of each fueling cycle. This excess reactivity must be controlled using a combination of burnable neutron poisons in the fuel, moderator poisons, and control blades interspersed in the heavy water moderator. Recent studies have shown that the combination of power density, high coolant temperatures, and reactivity management can lead to high maximum cladding surface temperatures (MCST) and maximum fuel centerline temperatures (MFCLT) in this design. This study focuses on improving both the MCST and the MFCLT through modifications of the conceptual design including changes from a 3 to 4 batch fueling cycle, a slightly shortened fuel cycle (although exit burnup remains the same), axial graded fuel enrichment, fuel-integrated burnable neutron absorbers, lower reactivity control blades, and lower reactor thermal powers as compared to the original conceptual design. The optimal blade positions throughout the fuel cycle were determined so as to minimize the MCST and MFCLT using a genetic algorithm and the reactor physics code PARCS. The final design was analyzed using a fully coupled PARCS-RELAP5/SCDAPSIM/MOD4.0 model to accurately predict the MCST as a function of time during a fueling cycle.

Numerical analysis of density wave instability and heat transfer deterioration in a supercritical water reactor

The present study numerically investigates Density wave oscillations (DWOs), a dynamic instability and captures Heat transfer deterioration (HTD) phenomenon for a Supercritical water reactor (SCWR) of the U. S. reference design. A Thermal-hydraulic (TH) model is employed to solve the nonlinearly coupled mass, axial momentum and energy conservation equations in a time domain using characteristics based implicit finite difference scheme accounting for the compressible flow dynamics of supercritical water. The analysis of DWOs is carried out to obtain the marginal stability thresholds. Next, the steady as well as transient safety analyses are carried out to determine the occurrence of HTD, access its severity and identify the safe operating conditions. Finally, the common operating conditions are predicted where the U. S. SCWR will be stable as well as safe.

Review of Gasification of Organic Compounds in Continuous-Flow, Supercritical Water Reactors

Supercritical water gasification (SCWG) is an emerging technology with applications in renewable energy and waste processing. Environmentally benign, SCWG relies on supercritical water to decompose organic molecules. Decomposition of alcohols, aromatic compounds, carbohydrates, and amino acids was studied. Significant discrepancies in decomposition rates between batch and continuous-flow reactors exist; this review considers studies conducted with continuous-flow reactors. Intermediate and final products were identified for most decomposition reactions with good agreement between studies. Arrhenius decomposition parameters were tabulated for the data found in the literature. Due to experimental complexities and reactor limitations, most studies address a narrow range of conditions and one to few reagents. A comprehensive data set for a variety of compounds and operating conditions using a fixed reactor design has not been collected. Experimental variability results in inconsistencies in reported kinetic r...

Determination of the optimal few-energy group structure for the Canadian Super Critical Water-cooled Reactor

Most deterministic neutron transport codes implement the multi-group approximation theory to solve the linear Boltzmann equation by which the neutron flux and the interaction cross-sections are averaged over discretized energy groups. In order to further reduce the computational cost for large scale and/or time dependent problems, this multi-group information undergoes additional energy group condensation into the so-called few energy group structure which may be used in a broad range of diffusion based full-core analyses. The accuracy and efficiency of few-group based computations is dependent on the number and the boundaries of the discrete energy group structure. Since the flux spectrum used in the homogenization process may not be known a priori and indeed may evolve in space, with burnup, and during transients, the optimal energy group structure depends on reactor type, design, operating conditions, fuel type, and composition. The Canadian Pressure Tube Super-Critical Water-cooled Reactor (PT-SCWR) is a Generation IV advanced reactor system that uses light water above its thermodynamic critical point as coolant and a plutonium-driven thorium fuel mixture. Considering that the anticipated flux spectrum for such a design deviates significantly from the thermal-neutron dominated CANDU designs, there may be a need for improvements in the number and boundaries of the few-group nuclear data. This paper presents a systematic methodology for the delineation of few-group energy structure for the Canadian PT-SCWR. The methodology used the SCALE (Standardized Computer Analysis for Licensing Evaluation) code package to examine the effect of energy group homogenization and determine the structure which minimizes the sensitivity of the results to energy group partitioning. As such, it utilizes normal and off-normal operating conditions to determine the effect of energy homogenization and determines the minimal group of energy boundaries which can accurately capture the lattice physics phenomena within the lattice cell over a wide range of operating and accident conditions.

Rate Theory for Dislocation Loops Evolution in AL-6XN Austenitic Stainless Steel under Proton Irradiation

The average size and density evolution of dislocation loops in AL-6XN austenitic stainless steel, a candidate fuel cladding material for supercritical water-cooled reactor, under proton irradiation were simulated through a rate theory model. The simulation results exhibit relatively good agreement with the experimental results at 563 K. The size and density of defect clusters are calculated under irradiation temperature between 550 K and 900 K and irradiation doses up to 15 dpa which satisfies the working condition in supercritical water-cooled reactor. The fast nucleation between self-interstitials happens at the initial stage of irradiation. The average size of dislocation loops increases while the average density of these loops reduces with the increasing temperature, and the average density approaches to a constant when irradiated at higher irradiation doses. The mechanism is discussed based on the variation of rate constants of defect reactions and the variation of the diffusion coefficients of interstitials and dislocation loops with dose and temperature.

Prediction of rate constants of important chemical reactions in water radiation chemistry in sub and supercritical water – non-equilibrium reactions

The rate constants for reactions involved in the radiolysis of water under relevant thermodynamic conditions in supercritical water-cooled reactors are estimated for inputs in simulations of the radiation chemistry in Generation IV nuclear reactors. We have discussed the mechanism of each chemical reaction with a focus on non-equilibrium reactions. We found most of the reactions are activation controlled above the critical point and that the rate constants are not significantly pressure dependent below 300 °C. This work will aid industry with developing chemical control strategies to suppress the concentration of eroding species.

Identification of safe and stable zone of operation in supercritical water reactor

Generation-IV international forum has identified thermal-hydraulic (TH) stability and safety issues as important challenges in the operation of supercritical water reactor (SCWR). In the present work, the scheme of the U. S. SCWR design is analyzed numerically for the TH static and dynamic stability, and safety concerns by employing an in-house TH model which solves nonlinearly coupled mass, axial momentum and energy conservation equations along with the pressure and specific enthalpy dependent thermodynamic equation of state using a characteristics-based implicit time-domain solution methodology to account for the compressibility of the supercritical water (SCW) dynamics. The applicability of the model for the aforementioned purposes has been verified against the state-of-the-art available results. Next, the model is used for the analysis of static instability due to Ledinegg excursion for the U. S. SCWR and results predict the possibility of the instability for the reactor is only at hypothetical operating conditions. The analysis of dynamic instability due to density wave oscillations (DWOs) is performed next. The investigations find its possibility in the relevant operating regime and the stability thresholds are identified to obtain marginal stability boundary (MSBs) for the reactor. Parametric studies are carried out next to determine the effect of exit pressure and coolant mass flow rate on the MSB. Further, the zone of safe operation from metallurgical point of view for the U. S. SCWR at normal and deteriorated heat transfer (DHT) regions is identified after determining the axial variation of heat transfer rate, clad wall temperature and various DHT parameters. Finally, a common stable and safe zone is obtained.

Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

The Canadian Supercritical Water-cooled Reactor (SCWR), a GEN IV reactor design, is a hybrid design of the well-established CANDU™ and Boiling Water Reactor with water above its thermodynamic critical point. Given the batch fueled design, control rods are used to manage the reactivity throughout the fuel cycle. This paper examines the consequences of a control rod drop accident (CRDA) for the Canadian SCWR. The asymmetry generated by the dropped rod requires an accurate 3-dimensional neutron kinetics calculation coupled to a detailed thermal-hydraulic model. Before simulating the CRDAs, the proper implementation of the 3D reactivity feedback was verified and various sensitivity studies were performed. This work demonstrates that the proposed safety systems for the SCWR core are capable of terminating the CRDA sequence prior to exceeding maximum sheath and centerline temperatures. In one instance involving a rod on the periphery of the core, the proposed trip setpoint (115% FP) was not exceeded and a new steady state was reached. Therefore it is recommended that the design also include provisions for a high-log rate and/or local Neutron Overpower Protection (NOP) trips, similar to existing CANDU designs such that reactor shutdown can be assured for such spatial anomalies.

Parametric study of CSR1000 thermal hydraulic stability

A parametric study is performed on thermal hydraulic stability of CSR1000 (a Chinese supercritical water-cooled reactor) using SCTHSAC (Single Channel Thermal Hydraulic Stability Analysis Code). This code uses RMM (system response matrix method) for analyzing the system stability in the frequency domain, a useful method to reveal detailed information about the operational thermodynamic parameters before and after onset of instability in the nuclear reactor. System response matrices are written for each node of the single coolant channel (by linearizing and perturbing conservation differential equations), which are solved for decay ratios. The maximum decay ratio represents stability of the whole system. The decay ratios are calculated for different values of reactor power and inlet orifice diameters. A parametric study of pressure, inlet temperature and moderator flow fraction is presented, which gives vital information about the thermal hydraulic stability bounds of CSR1000 in a sliding pressure startup scheme. The results show a similar trend with the stability charts of a typical SCWR, proving the novelty of RMM. Using this method, a stability surface is generated against power and flow fraction, giving a quick reference for designing a detailed startup scheme of CSR1000 in the future.