Thermal-hydraulic Instabilities Research Articles

A control strategy considering buildings’ thermal characteristics to mitigate heat supply-demand mismatches in district heating systems

In district heating systems (DHS), periodic heat supply-demand mismatches (HSDM) exacerbate issues such as thermal imbalance and hydraulic instability. To mitigate these issues, this paper proposes a supply water temperature correction strategy for heating stations that considers building's thermal characteristics. Firstly, buildings are classified into energy-saving levels based on comprehensive thermal characteristic Bc and heat load index. Secondly, simulation models are developed for buildings with different thermal characteristics, and response of indoor temperature to supply-demand mismatch rates was analyzed. Then, after determining the regulation sequence and supply water temperature of heating stations, a balanced control strategy is formulated for HSDM conditions. Finally, the proposed strategy was applied to a practical combined heat and power system (CHPS) in severe cold regions. The results showed that the studied heating stations were classified into 4 types. After regulation of both insufficient and excessive heating conditions, the indoor temperature guarantee rate increased by over 20 %; the ranges of thermal imbalance degree and hydraulic imbalance degree were reduced by more than 0.09 and 0.22, respectively; the SD of supply water pressure and supply-return pressure difference were reduced by over 0.17 and 0.11, respectively. The strategy provides a reference for global balanced control of DHS under similar HSDM conditions.

A unified approach to simulate thermal hydraulic instabilities and the mutual interactions using transient distributed model

In the present work, a 1-D thermal-hydraulic model is developed to simulate two phase flow boiling for a working coolant R134a circulating inside a single horizontal channel which includes a surge tank. The TH model is the first of its kind which solves the transient conservation equations based on distributive formulations without the incorporation of quasi-steady state approximation and still can simulate pure pressure drop oscillations (PDOs), Ledinegg excursion and density wave oscillations (DWOs), and the co-existence of those instabilities as well using a unified approach by employing appropriate external characteristic curves. The model is validated with the available experimental data and then, further used for intended purpose as mentioned above. Finally, the notable conclusions are drawn based on the simulations conducted. Results indicate that the initial equilibrium position in the negative slope region of the pressure drop vs. mass flux curve is unstable, leading to Ledinegg excursion or PDOs based on the steepness of the external pump characteristic curve. Increased surge tank volume extends PDOs. For Ledinegg excursion, the unstable position shifts to positive slope regions, and gets stabilized or destabilized depending on the occurrence of DWOs. DWOs are more likely at low mass flux and positive slope region. With a constant mass flow rate from an external pump, DWOs occur with a surge tank upstream of the heated section but not without it. Simulations demonstrate coexistence of PDOs, Ledinegg excursion, and DWOs in certain situations.

Effect of position of heat flux profile on the absorber surface of parabolic trough solar collector for direct steam generation

The overall performance of parabolic trough solar collector (PTSC) based power plants could be improved by introducing the Direct steam generation (DSG) in the receiver of the solar collector. However, the thermal-hydraulic instability induced in the DSG process is a severe issue for the commercial application of the technology. The concentrated solar flux falling on the dry portion of the absorber before or after solar noon generates a high circumferential thermal gradient in the stratified flow region. In this work, numerical analysis of thermo-hydrodynamics of DSG has been performed to study the effect of position of solar flux profile using CFD solver ANSYS Fluent 2020R1. The TPF in the solar collectors is modeled through two-fluid modeling approach. The inlet mass flow rate and operating pressure for PTSC are considered as 0.6 kg/s, and 100 bar, respectively. The solar beam radiations are considered as 750 W/m2 and 1000 W/m2. The obtained results revealed that temperature distribution at the absorber outer surface varies in the range of 585 K to 643 K. The maximum circumferential temperature difference is observed as 55.5 K. The volume fraction of vapor at the absorber outlet are found as 0.31 and 0.37 respectively for DNI 750 W/m2 and 1000 W/m2. The corresponding pressure losses are 316 Pa and 350 Pa, respectively. The obtained results could be employed to characterize the thermal behavior of the DSG solar collectors. The model is useful to configure the solar field operation for optimum performance.

Understanding of thermal-hydraulic instabilities and the mutual interactions at supercritical pressure

In the present work, a steady state and a simple transient models are integrated to simulate various thermal-hydraulic (TH) static and dynamic instabilities in a single vertical channel, subjected to time independent power, for supercritical fluid (SCF). The scheme of U.S. supercritical water reactor (SCWR) design is analyzed for different TH instabilities and the mutual interactions of such instabilities. An in-house steady state TH model, which solves a set of nonlinear coupled mass, momentum and energy conservation equations accounting for the pressure and enthalpy dependent thermodynamic equation of state, is implemented to find the solution at the initial equilibrium condition. The steady state model is then validated against the recently available experimental results. Next, the simple transient model is integrated with the steady state model to simulate Ledinegg excursion, pressure drop oscillations (PDOs), density wave oscillations (DWOs), and the mutual interactions of each other in presence as well as in absence of a surge tank at the upstream of the heated section. A number of simulations are carried out considering various typical pump (external) characteristic curves. Results provide the understanding what are the different operational criteria on which the Ledinegg excursion, PDOs and DWOs, or the interactions of the aforementioned TH instabilities will depend.

A multi-dimensional dataset for two-phase instability in low pressure natural circulation based on direct transient local measurement

Thermal hydraulic instability can challenge the control and the safety of nuclear reactors. This experimental research captures and investigates the instability in a low-pressure (1–10 atmospheres) natural circulation loop with an adiabatic chimney installed above the heated section. Direct transient local measurement of two-phase flow is performed at multiple axial locations along the flow channel with thermocouples and conductivity probes traversed by customized high-precision mechanisms. Based on the periodicity of oscillations, a post-processing method is proposed to align the data captured non-simultaneously to reconstruct the instability behavior in one representative period. The averaging operation over multiple oscillation cycles is performed to reduce the random errors. Cyclic phenomena are clearly demonstrated by the post-processed data, and rapid void fraction change induced by the flashing is observed to be significant in the driving mechanism for low pressure oscillations. Parametric effects of inlet subcooling, heat flux, inlet flow restriction and system pressure are studied within the experiment range, and qualitative results are explained and compared with existing literature. The ratio between the oscillation period and the travelling time of the fluid particle through the test section is compared with the theoretical value for the density wave oscillation. The new dataset contains directly measured transient multi-dimensional two-phase data along the flow channel for natural circulation under density wave oscillations. The generated dataset is valuable for the validation of system analysis codes and CFD codes in the simulation of natural circulation phenomena and dynamic two-phase flow.

On nuclear-coupled thermal-hydraulic instability analysis of Super-Critical-Light-Water-cooled-reactor (SCLWR)

The nonlinear stability analysis of a supercritical light water reactor (SCLWR) is presented using a nuclear-coupled thermal-hydraulic reduced-order model. The analytical model is developed by coupling 1. the point-kinetics equations with one group of delayed neutrons, 2. the fuel heat transfer and 3. a 1-D reduced order model which represents the heat absorption phenomenon during the coolant flow. Unlike the existing studies, which are limited to linear stability analysis, the primary objective of the work is to present the detailed nonlinear dynamics of the SCLWR system. The said goal is achieved at two levels. The first level is the linear stability analysis wherein the linear stability boundaries are shown in two sets of parameter space namely the two intrinsic reactivity feedbacks (Doppler reactivity feedback and density reactivity feedback) and the pseudo-phase-change number and pseudo subcooling number. The parametric effects show the sensitivity of the linear stability boundaries with the system parameters. In the second level, to discuss the nonlinear characteristics of the system, two types of Hopf bifurcations (subcritical and supercritical) are studied with the help of first Lyapunov coefficients of the system. Multiple numerical simulations are performed to verify the resultant limit cycle behavior associated with these bifurcations. Moreover, the occurrence of the generalized Hopf bifurcation is shown which represents the bifurcation between the subcritical Hopf and the supercritical Hopf regions. Further, inside the stable region of the linear stability boundary, the saddle-node bifurcation is found which represents the location of the turning point and the threshold of the globally stable region of the system.

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.

Experimental Study of Blowdown Event in a PWR-Type Small Modular Reactor

As one of the future directions of nuclear energy development, small modular reactor (SMR) designs meet the demands of safety, sustainability, and efficiency by eliminating circulating pumps and using natural circulation–driven flows to transfer fission energy to power. However, natural circulation–driven flows could be affected by two-phase-flow instability that may occur during accidental scenarios of pressurized water reactor (PWR)-type SMRs due to relatively small driving force. In view of the influence of two-phase-flow instability during accident transients for a PWR-type SMR, experiments are performed in a well-scaled test facility to investigate potential thermal-hydraulic flow instabilities during blowdown events. The test facility has a height of 3.44 m, and the operating pressure limit is 1.0 MPa. The scaling analyses ensure that the scaled phenomena, i.e., depressurization of the reactor pressure vessel (RPV) and emergency core cooling system valve actuation, could be accurately simulated in the test facility. Important thermal-hydraulic parameters including RPV pressure, containment pressure, local void fraction and temperature, pressure drop, and natural circulation flow rate are measured and analyzed during the blowdown events. Test results show that throughout the experiment the liquid level is always maintained above the heated core and the RPV pressure decreases. Oscillations of the natural circulation flow rate, water level, and pressure drop are observed during blowdown transients. Specific reasons for and mechanisms of the observed instability phenomena are discussed.

Bifurcation analysis of density wave oscillations in natural circulation loop

The present study explores the linear stability and bifurcation analysis of two-phase flow in the single heated channel for the natural circulation loop. The bifurcation analysis of this natural circulation system is limited to the detection of Hopf bifurcation, generalized Hopf bifurcation and turning point or limit point bifurcation of limit cycles. These natural circulation loops have several engineering applications and prone to thermal-hydraulic instability, which may lead to an operationally inefficient system. Stability boundary (linearly) obtained on a Npch- Nsub plane predicts the existence of limit cycles (a nonlinear phenomenon) for these loops and gives motivation for the further study on the detection of nonlinear phenomena. The subsequent results show a novel case of stability boundary (listed as Type A in the present work) emerging for higher subcooling numbers on Npch- Nsub parameter plane, which has not been noted in literature earlier and lies near to the Ledinegg stability boundary, which is a static instability. The characteristics of points nearby Type A stability boundary predicts that this boundary is different from the Ledinegg stability boundary. This Type A stability boundary divided by subcritical and supercritical Hopf bifurcation, which is separated by a generalized Hopf bifurcation (GH) point. Sustained oscillations in flow velocity can be observed near to the Hopf points on this boundary. These subcritical (hard and dangerous) and supercritical (soft and safe) Hopf bifurcations have been examined with time series graphs for all types (Type A-C) of stability boundaries present for the loops and occurring on the Npch- Nsub parameter plane. The two other type of boundaries known earlier is characterized here as Type B and C, respectively, which is basically Type I and Type II density wave oscillations. The transition of stability boundaries is discussed on the Npch- Nsub parameter plane, which shows that Type A stability boundary contains GH bifurcation coming from Type B stability boundary. The bifurcation of limit cycles originating from the Hopf points in the system has been examined and it is found that limit point bifurcation of limit cycles is occurring in the system, originating from the generalized Hopf points.

Numerical Analysis to Investigate the Effects of Thermal-Hydraulic Instabilities on Deterioration Heat Transfer and Wall Temperature in the CANDU Supercritical Water Reactor

The present work analyzes the thermal-hydraulic behavior of the CANDU supercritical water reactor (SCWR) using a 1-D numerical model. The possibility of a static instability, the Ledinegg excursion, is investigated, which reveals it can occur only in a hypothetical condition, far from the proposed operating regime of the CANDU SCWR. The investigation demonstrates the possibility of density wave oscillations (DWOs), a dynamic instability, in the operating regime of the CANDU SCWR and its marginal stability boundary (MSB) is obtained. The phenomenon of the deterioration in heat transfer is observed, and the related investigation shows that the strong buoyancy effect is responsible for its appearance inside the heating section of the channel of the CANDU SCWR core. The MSB is found to be inadequate in determining the safe operating zone of the reactor because the wall temperature can exceed the allowable limit from metallurgical consideration. The investigations also determine the safe as well as stable zone where the CANDU SCWR should operate in order to avoid the maximum temperature limit and DWOs.

Experimental investigation of natural circulation instability in a BWR-type small modular reactor

The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV). Specifically, the NMR is one third the height and area of a conventional BWR RPV with an electrical output of 50 MWe. Experiments are performed in a well-scaled test facility to investigate the thermal hydraulic flow instabilities during the startup transients for the NMR. The scaling analysis for the design of natural circulation test facility uses a three-level scaling methodology. Scaling criteria are derived from non-dimensional field and constitutive equations. Important thermal hydraulic parameters, e.g. system pressure, inlet coolant flow velocity and local void fraction, are analyzed for slow and fast normal startup transients. Flashing instability and density wave oscillation are the main flow instabilities observed when system pressure is below 0.5 MPa. And the flashing instability and density wave oscillation show different type of oscillations in void fraction profile. Finally, the pressurized startup procedure is recommended and tested in current research to effectively eliminate the flow instabilities during the NMR startup transients.

ИССЛЕДОВАНИЕ УСТОЙЧИВОСТИ РАСХОДА ТЕПЛОНОСИТЕЛЯ В ДВУХКОНТУРНОЙ АЭС С РЕАКТОРОМ ВВЭР СКД

The coolant flow rate stability in a two-circuit APP with a fast breeder reactor cooled by supercritical pressure steam is numerically analyzed. The investigation is done using the D-decomposition method in the plane of the stability-determining parameters based on a one-dimensional homogeneous model of heat and mass transfer processes in the APP circulation loop. The D-decomposition boundary is constructed using the methodology of numerical computation of the transfer function contained in the characteristic equation of the mathematical model. The flow rate stability in the first closed circuit of the APP, in the second circuit of the steam generator and in the parallel channels of the shield fuel assemblies of the reactor core is analyzed, accounting for the variation of energy release over different fuel assembly groups. The coolant flow rate stability in the nominal regime is found to be provided with a margin. Keywords: reactor core, coolant circulation loop, stability region, steam generator, supercritical steam pressure, flexible regulation mode of the turbine, coolant, thermal-hydraulic instability, shield fuel assembly.

CFD analysis of the dynamic behaviour of a fuel rod subchannel in a supercritical water reactor with point kinetics

The paper presents the analysis by a CFD code of coupled neutronic–thermal hydraulic instabilities in a subchannel slice belonging to a square lattice assembly. The work represents a further phase in the assessment of the suitability of CFD codes for studies of flow stability of supercritical fluids; the research started in previous work with the analysis of bare 2D circular pipes and already addressed 3D subchannel slices with no allowance for heat conduction or neutronic effects. In the present phase, a more realistic system is considered, dealing with a slice of a fuel assembly subchannel containing the regions of the pellet, the gap and the cladding and including also the effect of inlet and outlet throttling. The adopted neutronic model is a point kinetics one, including six delayed neutron groups with global Doppler and fluid density feedbacks.The response of the model to perturbations applied starting from a steady-state condition at the rated power is compared with that of a similar model developed for a 1D system code. Upward, horizontal and downward flow orientations are addressed making use of a uniform power profile and changing relevant parameters as the gap equivalent conductance and the density reactivity coefficient. A bottom-peaked power profile is also considered in the case of vertical upward flow. Though the adopted model can still be considered simple in comparison with actual details of fuel assemblies, the obtained results demonstrate the potential of the adopted methodology for more accurate analyses to be made with larger computational resources.

Developing a CFD methodology for the analysis of flow stability in heated channels with fluids at supercritical pressures

The paper presents the first results of a systematic methodology aimed at assessing the feasibility of analyses by CFD codes of flow instabilities in heated channels containing supercritical fluids. The research makes use of features presently available in CFD models, in the aim to move step-by-step from simple channel cases towards the analysis of more realistic fuel bundle subchannels.In the present step, basing on previous experience, the STAR-CCM+ code is adopted to solve flow stability problems in circular channels and fuel bundle slices without heating structures, in the aim to characterise the response of CFD models in the analysis of purely thermal–hydraulic instability phenomena. Some of the effects related to numerical discretisation, flow direction with respect to gravity and fluid properties are studied, comparing the stability thresholds identified by transient calculations with maps set up by in-house 1D codes developed and adopted in previous work. Both static and dynamic instabilities are observed, clearly showing the contiguity of these two kinds of phenomena as a function of inlet fluid subcooling.Conclusions are finally drawn about the promising features of CFD codes for such applications, sketching the lines of the work already going on in order to address more realistic reactor scale conditions.

Two-Phase Natural Circulation Loops: A Review of the Recent Advances

The wide range of applications of two-phase natural circulation loops (NCLs) demands intense analyses of boiling systems, and consequently a large number of experimental and theoretical research findings are available in the literature. This comprehensive review focuses on the state-of-the-art appraisal of the related technology, with a primary focus on studies carried out in the last two decades. General modeling strategies have been reported and compared. Different forms of thermal-hydraulic instabilities appearing in two-phase NCLs have been summarized, with discussions on possible ways of identification and their differences. In addition, studies on system instability are presented in a categorywise grouping following a chronological style in terms development of knowledge. Coupled nuclear instabilities are reported as well in some detail. Finally, this review identifies the gray areas where the information is not quite complete and suggests the required path of future research.

Flashing-Induced Instability Analysis and the Start-Up of Natural Circulation Boiling Water Reactors

To help achieve the necessary natural circulation flow, a fairly long chimney is installed in a natural circulation boiling water reactor (BWR) like the ESBWR. In such systems, hot water near the chimney exit could flash, thus leading to thermal-hydraulic instability during low-pressure start-up. A BWR stability analysis code in the frequency domain, named the Flashing-Induced STability Analysis for BWR (FISTAB), was developed in this work to address the issue of flashing-induced instability. The FISTAB code was benchmarked against the experimental results from the SIRIUS-N facility at the Central Research Institute of Electric Power Industry in Japan. Both stationary and perturbation results agreed well with the experimental observations.The proposed ESBWR start-up procedure under natural convection conditions was examined by the FISTAB code. It was confirmed that the examined operating points along the ESBWR start-up trajectory from TRACG simulation would be stable. Furthermore, to avoid the instability resulting from the transition from single-phase natural circulation to two-phase circulation, a simple criterion was proposed for the natural convection BWR start-up when the steam dome pressure is still low.

Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT

Abstract The IAEA’s reference research reactor MTR-10 MW has been modeled using the code MERSAT. The developed MERSAT model consists of detailed representation of primary and secondary loops including reactor pool, bypass, main pump, heat exchanger and reactor core with the corresponding neutronics and thermalhydraulic characteristics. Following the successful accomplishment of the steady state operation at nominal power of 10 MW, reactivity insertion accident (RIA) for three different initial reactivity values of $1.5/0.5 s, $1.35/0.5 s and $0.1/1.0 s have been simulated. The predicted peaks of reactor power, hot channel fuel, clad and coolant temperatures demonstrate inherent safety features of the reference MTR reactor. Only in case of the fast RIA of $1.5/0.5 s, where the peak power of 133.66 MW arrived 0.625 s after the start of the transient, the maximum hot channel clad temperature arrives at the condition of subcooled boiling with the subsequent void formation. However, due to the strong negative reactivity feedback effects of coolant and fuel temperatures the void formation persists for a very short time so that thermalhydraulic conditions remained far from exceeding the safety design limits of thermalhydraulic instability and DNB. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermalhydraulic system codes.

Safety analysis of the IAEA reference research reactor during loss of flow accident using the code MERSAT

Using the thermal hydraulic code MERSAT detailed model including primary and secondary loop was developed for the IAEA's reference research reactor MTR 10 MW. The developed model enables the simulation of expected neutronic and thermal hydraulic phenomena during normal operation, reactivity and loss of flow accidents. Two different loss of flow accident (LOFA) have been simulated using slow and fast decrease time of core mass flow. In both cases the expected flow reversal from downward forced to upward natural circulation has been successfully simulated. The results indicate that in both accidents the limit of onset of subcooled boiling was not arrived and consequently no exceed of design limits in term of thermal hydraulic instability or DNB is observed. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermal hydraulic system codes.

D111 TRAC-BF1コードの熱水力不安定現象への適用生の検討(OS10 混相流動)

D111 TRAC-BF1コードの熱水力不安定現象への適用生の検討(OS10 混相流動)

Thermal Hydraulic Flow Oscillation Characteristics in Multiformed Channels under Natural Circulation and Low-Pressure Conditions

Thermal hydraulic instabilities in vertical multicombined channels under natural circulation conditions have been experimentally investigated. The multicombined channels consist of two parts; a nonheated single-channel section installed downstream and a heated parallel-channel section installed upstream. The length of each channel section could be varied to investigate the flow oscillation characteristics of the instabilities. The flow oscillation modes at the parallel-channel section could be roughly classified into three different types according to the length ratio of each channel, such as in-phase (natural circulation oscillation), out-of-phase (geysering, density wave oscillation), and intermediate-phase. The objective of this study is to experimentally clarify the transport mechanisms of those flow oscillations. The resulting geysering could be further classified into Griffith's type and Aritomi's type according to the difference in input thermal conditions. Especially in case that the length of the single section was almost equal to that of the parallel section, the intermediate-phase flow oscillation was observed. The intermediate-phase flow oscillation was basically the out-of-phase flow oscillation due to the pressure fluctuation in the parallel channels, but the range of the flow oscillation periodically increased or decreased in a certain width due to the hydrostatic head fluctuation in the single channel.