Thermal-hydraulic Model Research Articles

Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

Abstract This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al2O3 nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

Numerical investigation of a plate-type steam generator for a small modular nuclear reactor

A numerical feasibility study was conducted to investigate the thermal–hydraulic characteristics of a steam generator with corrugated plates for a small modular reactor. Accordingly, a one-dimensional thermal–hydraulic analysis code was developed based on the existing state-of-the-art thermal–hydraulic models and correlations for corrugated plate heat exchangers. Subsequently, the pressure loss, heat transfer, and instability characteristics of the steam generator with corrugated plates were investigated according to the chevron angle and mass flux. Additionally, the characteristics of rectangular and disk-type corrugated plate steam generators with equivalent heat transfer areas were analyzed. The steam generator with disk-type corrugated plates exhibited better performance in terms of pressure loss and heat transfer rate than the rectangular type. In addition, when the mass flux decreased from the onset of boiling points, reverse gradients of the total pressure change were observed in both types. Thus, it was confirmed that Ledinegg instability could occur in the steam generator with corrugated plates. However, it was dependent on the chevron angle, and the optimal chevron angle to minimize instability was 45° under the conditions of the present analysis.

Probabilistic safety margin characterization of an integrated small modular reactor using MFM and adaptive polynomial chaos

Characterization of the safety margins can provide essential information for both licensee and regulatory commission to support decision-making, especially for the next generation of nuclear power plants which implement several novel designs. However, the traditional monte-carlo based best estimate plus uncertainty method is not appropriate for large scale application because of its high computational cost. Thus, a novel hybrid method that combines multilevel flow modelling (MFM) and generalized polynomial chaos (gPC)is proposed in this paper. In the framework of the proposed method, the MFM is utilized to model the functions of collaborative working passive systems of an integrated small modular reactor (iSMR) and perform causal reasoning for selecting relevant uncertain parameters, then the gPC is implemented to construct an orthogonal polynomials-based surrogate model for the thermal hydraulic model to perform fast sensitivity analysis and safety margin characterization. An adaptive sampling algorithm is also implemented to improve the efficiency and accuracy of the safety margin characterization. The proposed method is applied to a test case to analyze the sensitivity of uncertain parameters during an SBO accident of iSMR. Subsequently, the adaptive method is implemented for probabilistic safety margin characterization after considering a delayed core makeup tank (CMT) injection and the conditional core damage frequency is predicted. The result shows great efficiency of the proposed MFM combining adaptive gPC method.

Thermal-hydraulic models for the cooling of HTS power-transmission cables: status and needs

An overview of the main peculiarities of the models currently available in the literature for the thermal-hydraulic analysis of the cooling of the high critical temperature superconducting power transmission cables, mainly in nominal operating conditions, is performed. Several models address specific issues such as the temperature distribution along or across the cables, and their pressure drop, typically with a simplified approach. The verification and validation of the models has not been systematically addressed so far. From analysis of the available literature, the lack of a general model, capable to address thermal-hydraulic transients for the cooling of the different possible cable designs, with the capability to capture both the behaviour of the cable solid components and different cryogens, is highlighted. The main ingredients that such a general thermal-hydraulic model should not miss are presented, also based on the know-how on, and comparison to, similar tools for low critical temperature cables for nuclear fusion applications.

A Study on the Effect of Geometry and Operating Variables on Density Wave Oscillation in a Supercritical Natural Circulation Loop

Nowadays, a prime technology in generation IV nuclear reactors, the supercritical water reactor (SCWR), is the main object of focus. The current article aims to develop a thermal hydraulic numerical model for predicting density wave oscillation (DWO) in a supercritical water natural circulation loop (SCWNCL). A one-dimensional thermal hydraulic mathematical model was developed. The numerical model consists of nonlinear mass, momentum, and energy conservation equations, which were discretized by applying the implicit finite difference technique. The numerical model was validated with experimental results, and numerical simulations were carried out to find the marginal stability boundary (MSB) and draw the stability map for the loop. Further, the effects of geometry (i.e., diameter and hot leg length) and operating parameters (i.e., inlet system pressure and friction factor) on the density wave oscillation of the SCWNCL were analyzed.

Validation of neutronics and thermal-hydraulics coupling model of the RELAP5-3D code using the MSRE reactivity insertion tests

In designing a molten salt reactor, a calculation code that combines neutronics and thermal-hydraulics is required. For this purpose, it is realistic to verify and use a system code that has been conventionally developed. As the system code, the RELAP5-3D code which has been used by many researchers was selected. The only transient tests using a molten salt reactor were performed at Molten Salt Reactor Experiment (MSRE) in the Oak Ridge National Laboratory (ORNL). Experimental results have been reported in which the reactivity is applied stepwise during MSRE operation at 1, 5 and 8 MW. Although there are few examples of analyzing this data, the analysis up to now has revealed that there is a discrepancy in the reactivity values described in the report. It has been also clarified by the data analysis in the present study that the amount of applied reactivity should be opposite to be consistent with the previous results of the same author. Therefore, the values were corrected and analyzed. In RELAP5-3D, the entire system from the core to the air cooler was modeled. The RELAP5-3D input deck was created based on the MSRE system data described in ORNL's past reports and recently edited reports for benchmark analysis. For RELAP5-3D, since it is necessary to prepare binary files of working fluid properties, the thermophysical properties about molten fuel salt and molten coolant salt were generated and used.The RELAP5-3D code could trace the three reactor power transients measured in the experiment, in general. However, in the case of low reactor power, the undershoot after the peak became large when the power returned to the original power level. At high reactor power, the undershoot was small and the behavior was in good agreement with the experimental result. As a result, it is considered that the calculation using the neutronics and thermal-hydraulics coupling model of the RELAP5-3D code could be validated.

Model predictive and fuzzy logic controllers for reactor power control at Reaktor TRIGA PUSPATI

The Reaktor TRIGA PUSPATI (RTP) rely extensively on a core power control system for control reactivity to fulfilment the fundamental safety function for a nuclear reactor. It is technically challenging to operate within tight multiple parameter constraints and keep the core power output stable. At present, the power tracking performance of the system could be considered unsatisfactory which produces a relatively long settling time and high control effort. Hence, a study of a model predictive control (MPC) strategy by integrating the current Power Change Rate Constraint (PCRC) using fuzzy logic which is part of the core power control design is conducted. In this paper, the MPC design based on mathematical models of the reactor core included point kinetics model, thermal-hydraulic model, reactivity model and dynamic rod position model help to enhance core power control. The power tracking performance of the proposed control method and previous Feedback Control Algorithm-Fuzzy PCRC is compared via computer simulation with different power range operations. Overall, the results show the MPC-Fuzzy PCRC strategy provides a greater level of operational safety and optimum operation of the TRIGA reactor.

A thermal-hydraulic coupled simulation approach for the temperature and flow rate control strategy evaluation of the multi-room radiator heating system

The lack of efficient temperature and flow rate control strategy is one of the key factors in excessive energy consumption of the district heating (DH) system. The current researches on secondary network simulation and control usually tackle the hydraulic characteristics of the network and building thermal dynamic separately. However, for the multi-room radiator heating system, the hydraulic characteristics and building thermal dynamics are coupled. An effective coupled thermal-hydraulic model for simulation and control of multi-room radiator heating systems is essential. In this paper, a novel thermal-hydraulic coupled simulation approach is proposed. The temperature distribution and thermal capacity of the radiator, as well as the hydraulic characteristics of the radiator heating network, are considered. A numerical method is developed to solve the proposed model. Different control strategies are analyzed based on the model. Result shows that efficient terminal flow rate control can save 8.4%–13% energy. The proportional control is superior to the on-off control in maintaining indoor temperature, while their energy consumptions are the same. Additionally, the hydraulic coupling effect of the network between households and floors should be considered if the balance valves are not installed on each floor.

A whole-core steady-state thermal-hydraulic model for annular fuel type fluoride-salt-cooled reactors

A whole-core, steady-state thermal-hydraulic model is developed for the fluoride-salt-cooled small modular advanced high-temperature reactor (SmAHTR) that employs an annular fuel configuration. This pre-conceptual reactor design by Oak Ridge National Laboratory (ORNL) has the annular fuel and moderator pins arranged in a hexagonal layout. The FLiBe coolant flows from the bottom to the top of the core, parallel to the hexagonal bundle. The fuel and moderator pins in the core are discretized into finite volumes and the 3-D heat conduction equation is solved to obtain the temperature profile. Inter-fuel assembly conduction is also addressed. For this fuel assembly configuration, the coolant flows through two distinct regions – the hexagonal pin bundle and the annulus between the fuel pin and the tie rod. The fluid flow through the hexagonal bundles is modeled using the subchannel approach, in which the coolant region is discretized into corner, edge and interior subchannels and the resulting conservation equations are systematically solved. The 1-D mass, momentum and energy equations are solved for the annulus channels between the fuel pin and the tie rod. Pertinent closure models from the literature are employed to close the system of equations. A preliminary code-to-code comparison between the present model and a CFD model is also performed. The resulting thermal-hydraulic model can provide temperature, flow rate and pressure drop profiles for the different solid and fluid regions throughout the entire core. Whole-core thermal-hydraulic results for a representative power profile are presented and discussed.

Extension of the Circe Methodology to Improve the Inverse Uncertainty Quantification of Several Combined Thermal-Hydraulic Models

Extension of the Circe Methodology to Improve the Inverse Uncertainty Quantification of Several Combined Thermal-Hydraulic Models

Thermal-hydraulic model of a two-phase circuit with a compensation cavity

Thermal-hydraulic model of a two-phase circuit with a compensation cavity

Reduced Thermal-hydraulic Model of Flue Gases in Horizontal Fire Tubes with Helical Coil Inserts Employing a CFD Simulated Database

Reduced Thermal-hydraulic Model of Flue Gases in Horizontal Fire Tubes with Helical Coil Inserts Employing a CFD Simulated Database

Simulation optimization framework for dynamic probabilistic safety assessment

Probabilistic safety assessment (PSA) based on event trees and fault trees has been widely used in the risk assessment of nuclear power plants. A static approach by nature, PSA has limitations to consider dynamic scenarios with time-dependent sequences and interactions. In contrast to static-based PSA, dynamic PSA has been introduced as a complementary methodology that considers dynamic scenarios between the system and human operations by interfacing physical simulation with thermal-hydraulic models for risk assessment. However, the various research on dynamic PSA has a common challenge in that the number of dynamic scenarios to be simulated increases impractically. An approach is therefore necessary to manage the number of simulations for performing dynamic PSA efficiently. The objective of this paper is to propose a simulation optimization framework using an optimization algorithm to reduce, as much as reasonably achievable, the large number of dynamic scenarios to be evaluated. The optimization algorithm is proposed to optimize the large numbers of generated dynamic scenarios while maintaining accurate risk quantification in the performance of dynamic PSA. To demonstrate the application of the proposed framework to dynamic PSA, two case studies were conducted considering loss of coolant accidents.

Development of a special thermal-hydraulic component model for the core makeup tank

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

A Validation Roadmap of Multi-Physics Simulators of the Resonator of MW-Class CW Gyrotrons for Fusion Applications

For a few years the multi-physics modelling of the resonance cavity (resonator) of MW-class continuous-wave gyrotrons, to be employed for electron cyclotron heating and current drive in magnetic confinement fusion machines, has gained increasing interest. The rising target power of the gyrotrons, which drives progressively higher Ohmic losses to be removed from the resonator, together with the need for limiting the resonator deformation as much as possible, has put more emphasis on the thermal-hydraulic and thermo-mechanic modeling of the cavity. To cope with that, a multi-physics simulator has been developed in recent years in a shared effort between several European institutions (the Karlsruher Institut für Technologie and Politecnico di Torino, supported by Fusion for Energy). In this paper the current status of the tool calibration and validation is addressed, aiming at highlighting where any direct or indirect comparisons with experimental data are missing and suggesting a possible roadmap to fill that gap, taking advantage of forthcoming tests in Europe.

Benchmark analysis of in-vacuum vessel LOCA scenarios for code-to-code comparison

Computer modeling codes represent a crucial tool to support the design of future fusion plants. Since one of the most important functions that codes must achieve is assessing safety systems design, a verification and validation phase is required. In this framework, a code-to-code comparison among four codes has been carried within EUROfusion consortium between Safety Analyses and Environment and Balance of Plant work packages.In particular, an in-vessel Loss of Coolant Accident has been selected as a benchmark scenario for investigating thermal-hydraulic parameters of the vacuum vessel and its suppression system. This paper aims to compare the answer of different codes in terms of peak pressure within the vacuum vessel and its timing, the equilibrium pressure, and relief valve and rupture disk opening using a simplified thermal-hydraulic model of the EU-DEMO. In order to minimize the differences among the codes, the models have been kept to the simplest possible nodalization, successively increasing the model complexity.

Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models

The Fast Flux Test Facility (FFTF) is a liquid sodium-cooled nuclear reactor designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. In July 1986, a series of unprotected transients were performed to demonstrate the passive safety of FFTF. Among these, a total of 13 loss-of-flow-without scram (LOFWOS) tests were conducted to confirm the liquid metal reactor safety margins, provide data for computer code validation, and demonstrate the inherent and passive safety benefits of specific design features. In our preliminary work, we have performed relatively coarse modeling of the FFTF. To better predict the transient behavior of FFTF LOFWOS test #13, we modeled it using a more refined thermal-hydraulics model. In this paper, we simulate FFTF LOFWOS test #13 with the system safety analysis code SAC-3D according to the benchmark specifications provided by Argonne National Laboratory (ANL). The simulation range includes the primary and secondary circuits. The reactor core was modeled by the built-in 3D neutronics calculation module and the parallel-channel thermal-hydraulics calculation module. To better predict the reactivity feedback introduced by coolant level variations within the GEMs, a real-time macro cross-section homogenization processing module was developed. The steady-state power distribution was calculated as the transient simulation initial boundary conditions. In general, both the steady-state calculation results and the whole-plant transient behavior predictions are in good agreement with the measured data. The relatively large deviations in transient simulation occur in the outlet temperature predictions of the PIOTA in row 6. It can be preliminarily explained by the reason for neglecting the heat transfer between channels in this model.

Model prediction of the critical heat flux and thermal resistance of a rectangular thin-strip evaporator with distributed liquid supply capillary wicks for high heat flux cooling

• A rectangular thin-strip evaporator with distributed liquid supply capillary wicks (RTEDCW) is designed and the performance predicted using a thermal-hydraulic model. • CHF is attained when the liquid pressure drop in the wick equals the wick's capillary pumping capability. • The overall thermal resistance is the ratio of the wick superheat and the corresponding input heat flux. • CHF increases from 275 W/cm2 to 655 W/cm2 as the particle size increases from 30 mm to 100 mm. • The overall thermal resistance increases from 0.045 K/(W/cm2) to 0.175 K/(W/cm2) as the particle size increases from 30 mm to 100 mm. The heat transfer performance characteristics, mainly the upper limit of heat flux and the overall thermal resistance, of two-phase liquid-vapor-based heat spreaders primarily developed for concentrated-heat dissipation in microelectronics are dependent, predominantly, on the thickness and particle size of the evaporation wick as well as on the distribution of the high-permeable liquid supply channels, also known as arteries. In this study we develop a heat and mass transfer (thermal-hydraulic) model to predict the critical heat flux and the overall thermal resistance of a novel wick with rectangular thin-strip evaporator (with a large length-to-thickness ratio, L / t ≥ 3.5), with distributed liquid supply capillary wicks in a downward-facing orientation, each made of sintered copper particles, for microgravity applications. The liquid supply capillary wicks are made of 350-μm-diameter particle sizes. With the thermal-hydraulic model, we predict the CHF and overall thermal resistance for three evaporation wick particle sizes, i.e., 30, 60, and 100 μm, monolayer each, and show that both performance characteristics are primarily controlled by the hydraulic properties of the evaporation wick. The predicted CHF and minimum thermal resistance are 275 (12.3 °C superheat) and 0.045 K/(W/cm 2 ), 599 (56.7 °C superheat) and 0.095 K/(W/cm 2 ), and 655 W/cm 2 (114.4 °C superheat) and 0.175 K/(W/cm 2 ), for the 30, 60, and 100 μm particles, respectively. Finally, we compare the results of the present work with some of the existing numerical and experimental data for different thin wick designs found in the literature.

Comparative study on heat extraction performance of geothermal reservoirs with presupposed shapes and permeability heterogeneity in the stimulated reservoir volume

The stimulated reservoir volume (SRV) is critical to the heat extraction from an enhanced geothermal system (EGS). Being different from previous simple cube-shape SRV, this study proposes a new shape SRV with heterogeneous permeability to evaluate the heat extraction performance of an EGS. First, the new shape SRV is generated based on the location and density of microseismic events during multi-stage hydraulic fracturing. Other two SRVs with simple cube or complex shape but same volume are generated for comparison. Second, a fractal permeability is developed for the heterogeneous SRV domain. Third, a thermal-hydraulic (TH) coupling model is implemented for the heat transfer and water flow in these three SRVs. Finally, the EGS performances of these three SRVs with a stimulated volume of 0.405 km 3 are comparatively evaluated through numerical simulations. Numerical simulations reveal that the EGS with cube-shape SRV overestimates both output temperature and energy extraction. The heterogeneous permeability in the SRV domain leads to a dramatic increase of reservoir impedance. This reduces energy extraction efficiency and breakthrough time, thus shortening the lifetime of the EGS system. • A new shape SRV with heterogeneous permeability in HDR reservoir is proposed. • An evaluation system for EGS is established based on a TH coupling model. • EGS performance of the three presupposed SRVs is comparatively analyzed. • The feasibility of a megawatt EGS power plant in Gonghe Basin is discussed.

The interplay of Malm carbonate permeability, gravity-driven groundwater flow, and paleoclimate\xa0\u2013 implications for the geothermal field and potential in the Molasse Basin (southern Germany), a foreland-basin play

The Molasse Basin in Southern Germany is part of the North Alpine Foreland Basin and hosts the largest accumulation of deep geothermal production fields in Central Europe. Despite the vast development of geothermal energy utilization projects especially in the Munich metropolitan region, the evolution of and control factors on the natural geothermal field, more specifically the time-varying recharge and discharge governing groundwater and heat flow, are still debated. Within the Upper Jurassic (Malm) carbonate aquifer as the main geothermal reservoir in the Molasse Basin, temperature anomalies such as the Wasserburg Trough anomaly to the east of Munich and their underlying fluid and heat transport processes are yet poorly understood. We delineate the two end members of thermal–hydraulic regimes in the Molasse Basin by calculating two contrasting permeability scenarios of the heterogeneously karstified Malm carbonate aquifer along a model section through the Wasserburg Trough anomaly by means of two-dimensional numerical thermal-hydraulic modelling. We test the sensitivity of the thermal-hydraulic regime with regard to paleoclimate by computing the two Malm permeability scenarios both with a constant surface temperature of 9 °C and with the impact of paleo-temperature changes during the last 130 ka including the Würm Glaciation. Accordingly, we consider the hydraulic and thermal effects of periglacial conditions like permafrost formation and the impact of the numerous glacial advances onto the Molasse Basin. Thermal-hydraulic modelling reveals the effect of recurrent glacial periods on the subsurface targets of geothermal interest, which is minor compared to the effect of permeability-related, continuous gravity-driven groundwater flow as a major heat transport mechanism.