Natural Circulation Experiments Research Articles

Linear and Nonlinear Stability Analysis of Molten Salt Natural Circulation Loop

Abstract A molten salt natural circulation loop (MSNCL) has been setup to study the steady-state, transient, and stability characteristics of molten nitrate salt, mixture of sodium nitrate, and potassium nitrate in 60:40 ratios by weight. Natural circulation experiments in a temperature range of 290 °C–600 °C have been performed. A semi-analytical linear model was derived for stability analysis of rectangular natural circulation loops with conventional localized surface heating and cooling. The developed model includes the effect of wall thermal inertia, variable internal heat transfer coefficient, finite secondary side heat transfer coefficient at cooler, and heat losses for predicting the stability map. These effects are incorporated considering their significant role in modeling high-temperature molten salt-based natural circulation systems. The developed model has been first validated with the experimental data of water loop available in literature. The developed model is then validated with the experimental data generated in MSNCL. Validation of in-house developed nonlinear model has also been performed against the same experimental data. The comparison of both linear and nonlinear stability analysis with the experimental data shows good agreement and articulates the importance of various parameters which have been included in the developed model.

A fully‐implicit numerical algorithm of two‐fluid two‐phase flow model using Jacobian‐free Newton–Krylov method

AbstractJacobian‐Free Newton–Krylov (JFNK) method is a stable and high‐efficiency method to solve the multi‐physics coupling problem for the modeling and simulation (M&S) of nuclear reactors. However, for the two‐fluid two‐phase flow model, the large number of constitutive models for different flow regimes as well as their discontinuities between different flow regimes present a huge challenge to solve the equations with JFNK method. Nevertheless, in this research, a fully‐implicit numerical algorithm was proposed to solve the two‐fluid two‐phase flow model based on the JFNK method. The field equations were fully‐implicitly discretized based on the second‐order backward time and Van Albada high‐order spatial difference schemes. A semi‐implicit‐scheme‐based preconditioner was constructed to improve the calculation efficiency. The V‐shaped linear advection test, water faucet test and the oscillating manometer test were simulated to evaluate the accuracy of the numerical algorithm and the performance of the numerical treatments with phase appearance and disappearance. By simulating the Bartolomei subcooled boiling experiment, Becker and Bennett dry out and post‐dry out experiments, the performances of the developed numerical algorithm for single‐phase flow, two‐phase flow and transition from the single‐phase flow to the two‐phase flow were validated. The single‐phase natural circulation and Edwards blowdown experiments were simulated to study the capability of the current fully‐implicit numerical algorithm for slow and quick transients, respectively. The results demonstrate good performance of the proposed algorithm and validate the accuracy of such algorithm. The comparisons of numerical efficiency show that the fully‐implicit numerical algorithm spends more time than the semi‐implicit numerical algorithm because the calculation of the inverse precondition matrix is time consuming. But the numerical stability of the fully‐implicit numerical algorithm is not influenced by the time step. In the practical simulations based on the fully‐implicit numerical algorithm, a large time step can be used to obtain a stable prediction and comparable or higher calculation efficiency and accuracy with respect to the semi‐implicit numerical algorithm.

Thermal-Hydraulic System-Level Analysis of a Molten Salt Natural Circulation Loop

Molten salt reactors (MSRs) are a class of Generation IV nuclear reactors using molten salts as heat transfer fluids. MSRs bring a number of benefits, including low primary system working pressure, high working temperature, and enhanced safety due to the passive safety systems adopted. Although MSRs promise these benefits, a number of key technology needs, such as the accurate prediction of the thermal-hydraulic performance of the passive safety systems, which completely rely on natural circulation, are indispensable for MSR development, licensing, and future deployment. Therefore, this study develops the one-dimensional (1D) NAtural Circulation COde (NACCO) considering the buoyancy and radiative heat transfer effects in high-temperature molten salts for such predictions. The 1D code, developed using MATLAB, is then benchmarked with experimental data from three natural circulation flow experiments, where water, nitrate salt NaNO3-KNO3 (60–40 wt%), and fluoride salt LiF-BeF2 (66–34 mol%, FLiBe) were used as the working fluids. Our analysis shows that (1) the buoyancy and radiative heat transfer effects need to be considered for high-temperature molten salt natural circulation flows, while the radiative heat transfer effect is negligible for low-temperature water flows in the natural circulation experiments investigated, and (2) the 1D code NACCO predicts salt temperature profiles reasonably well, with less than 18°C and 25°C discrepancies from experimental data for the pipe centerline temperature of NaNO3-KNO3 and FLiBe up to 450°C and 750°C, respectively.

Experimental study on the natural circulation behavior of a full-scale PWR fuel assembly

During a loss-of-coolant accident in spent fuel pools, the fuel assembly will be exposed to the air environment and decay heat will be removed through the natural circulation of air. There are few experimental studies on the natural circulation behavior in the spent fuel pool. In this study, a natural circulation experiment in a full-size PWR fuel assembly was carried out by replacing the fuel rods with electrical heating rods. The temperature distributions at eight different heights as well as air flow rate were obtained during the heating-up process of the fuel assembly. When the temperature difference between the fuel assembly and the environment is less than 150 °C, the flow rate increases as the temperature difference increase. But after the temperature difference exceeds 150 °C, the flow rate no longer changes with the temperature difference. The results of this study can be used to verify the system codes and CFD models.

Development and verification of a three-dimensional spatial dynamics code for molten salt reactors

Molten Salt Reactor (MSR) has the characteristics of drift effect of delayed neutron precursors (DNP), direct heating of coolant and moderator cooled by fuel salt under most conditions, which bring significant differences from traditional reactors in dynamics calculation. To deal with the unique dynamics characteristics of MSR, a specific three-dimensional spatial dynamics code – TMSR3D has been developed, in which the fourth order standard nodal expansion method and the exponential transformation method are implemented to solve the neutron diffusion equations and the neutron kinetics equations respectively, the parallel multi-channel thermal hydraulics (TH) model is adopted to take into account the feedbacks of local temperature to group constants, and a convection term is introduced to the traditional balance equations for DNP to consider the influence of fuel salt flow. Besides, the least square fitting method is employed to parameterize the few-group homogenized parameters generated by the lattice code DRAGON5. The 3D-LRA core benchmark is first used to verify TMSR3D for solid-fueled reactor problem. Afterwards, the experimental data of the Molten Salt Reactor Experiment (MSRE) collected in the European project MOST are simulated, which includes the transients perturbed by protected fuel pump start-up and coast-down at zero power, and the natural circulation experiment. The numerical results indicate that the developed code is capable of providing a reasonable description for the neutronics-TH coupling phenomena of graphite-moderated channel-type MSRs.

An experiment‐based validation of a system code for prediction of passive natural circulation in sodium‐cooled fast reactor

To evaluate residual heat removal capacity of next generation China sodium-cooled fast reactor (SFR) and provide data for code validation, an integral natural circulation experiment was performed on a scaled water platform which had same structure with reactor primary system. A modular code THACS focusing on safety analysis of SFR was validated by this test. Some necessary development and verification for physical property, frictional loss, and heat transfer of water were carried out first. Whole simulation of an unprotected loss-of-flow (ULOF) case was divided into two phases and computation lasted until 5000 s of test time during second stage. Agreement in light of system temperature distribution in first stage provided a suitable initial condition for subsequent transient. Satisfied flow rate curve was reproduced with decaying pressure head and valve characteristic as input when pump was stopping. All natural circulation paths were predicted, including original flow in primary loop through main pipe and reversed flow in reactor vessel cooling system (RVCS). Transition to operative mode of direct heat exchanger (DHX) in direct reactor auxiliary cooling system (DRACS) and its cooling power were also forecasted successfully. This work means a significant advance for THACS in the field of engineering application of China SFR.

Validation and Uncertainty Quantification for Two-Phase Natural Circulation Flows Using TRACE Code

The work presents validation of the TRAC/RELAP Advanced Computational Engine (TRACE) code for natural circulation two-phase flow in a vertical annulus. Natural circulation experiments were recently conducted for a vertical internally heated annulus at the Multiphase Thermo-Fluid Dynamics Laboratory at the University of Illinois. The experimental matrix consists of 107 experiments with system pressure in the range of 145 to 950 kPa and heat flux up to 275 kW/m2. Void fraction, gas velocity, and interfacial area concentration were measured in five axial locations along the test section with six measurements of bulk liquid temperature and pressure. To validate the capability of the TRACE code under natural circulation flow conditions, a complete model of the experimental facility was created and validated using forced convection and single-phase natural circulation data. Sensitivity and uncertainty quantification were performed. The sensitivity to important simulation parameters was studied using Sobol’s variance decomposition and the Morris screening method. The sensitivity of boundary conditions on void fraction measurement was investigated. The sensitivity study has shown significant differences in model sensitivity between different experimental conditions. With heat flux being the most influential parameter for high-pressure cases without flashing and pressure, temperature and heat flux have a combined strong effect in the case of low-pressure experiments when flashing occurs. Additionally, higher uncertainty in void fraction prediction was observed for experimental conditions at low pressure with flashing.

Prediction of dryout-type CHF for rod bundle in natural circulation loop under motion condition

Abstract In nuclear engineering, the occurrence of critical heat flux (CHF) is complicated for rod bundle, and it is much more difficult to predict the CHF when it is in natural circulation under motion condition. In this paper, the dryout-type CHF is investigated for the rod bundle in a natural circulation loop under rolling motion condition based on the coupled analysis of subchannel method, a one-dimensional system analysis method and a CHF mechanism model, namely the three-fluid model for annular flow. In order to consider the rolling effect of the natural circulation loop, the subchannel model is connected to the one-dimensional system code at the inlet and outlet of the rod bundle. The subchannel analysis provides the local thermal hydraulic parameters as input for the CHF mechanism model to calculate the occurrence of CHF. The rolling motion is modeled by additional motion forces in the momentum equation. First, the calculation methods of the natural circulation and CHF are validated by a published natural circulation experiment data and a CHF empirical correlation, respectively. Then, the CHF of the rod bundle in a natural circulation loop under both the stationary and rolling motion condition is predicted and analyzed. According to the calculation results, CHF under stationary condition is smaller than that under rolling motion condition. Besides, the CHF decreases with the increase of the rolling period and angular acceleration amplitude within the range of inlet subcooling and mass flux adopted in the current research. This paper can provide useful information for the prediction of CHF in natural circulation under motion condition, which is important for the nuclear reactor design improvement and safety analysis.

Experimental Investigation of Natural Circulation During Air-Ingress Scenario in a Very High Temperature Reactor

In order to investigate air-ingress phenomena in a gas-cooled very high temperature reactor (VHTR), natural circulation experiments have been conducted in a helium flow loop after the injection of nitrogen into the lower plenum. A pair of helium analyzers were used to measure the nitrogen and helium concentrations in the lower plenum and upper plenum. The changes in the nitrogen concentration in the upper plenum were used to calculate the time required for the transport of nitrogen from the lower plenum to upper plenum through a riser flow channel made of graphite. The effect of system temperature and pressure on the rate of nitrogen transport has been studied extensively. Furthermore, a close examination of the graphite flow channel wall temperatures at different elevations showed small but sudden drops indicating the arrival of nitrogen at each elevation. From these data, the upward transport of nitrogen injected into the lower plenum under natural circulation conditions could be quantitatively investigated. The experimental findings indicate that the driving mechanisms for air transport through the reactor core of VHTR would result from both molecular diffusion and natural circulation. At low graphite temperatures in the riser, molecular diffusion is the dominating mechanism; however, as the riser temperature increases, natural circulation becomes dominant and the rate of nitrogen transport increases. Further, the time constants for these mechanisms have been calculated using a simplified species transport equation.

Analysis of natural circulation frictional resistance characteristics in a rod bundle channel under rolling motion conditions

In order to study the frictional resistance characteristics of a rod bundle channel under ocean conditions, a natural circulation experiment was performed in a 3 × 3 rod bundle channel under rolling conditions. The characteristics of the instantaneous frictional resistance coefficient in the rod bundle channel under rolling conditions are obtained by analyzing newly obtained experimental data. The experimental results show that the frictional resistance coefficient varies with the rolling conditions. The larger the rolling amplitude and frequency, the larger the average value of the frictional resistance coefficient, and the greater the fluctuation of the coefficient. It is also affected by the fluid property parameters. The hotter the fluid in the channel, and the faster the fluid velocity, which means the larger heating power and the steady state Reynolds number, cause the coefficient to decrease, and the fluctuation of the coefficient will also decrease. A new correlation function of the instantaneous frictional resistance coefficient under differing rolling conditions is developed and compared with these newly presented data. The correlation function is a good predictor of the frictional resistance coefficient and is in good agreement with the experimental results over a broader range of conditions than previous functions and is also applicable for both laminar and transitions fluid flow. This new correlation function introduces the viscosity correction factor based upon the previous research, which makes it more widely applicable in predicting the instantaneous frictional resistance coefficient under rolling conditions. This newly developed function should be useful for design considerations for passive cooling for sea-based nuclear reactor cores.

Uncertainty quantification method for RELAP5-3D© using RAVEN and application on NACIE experiments

The best estimate plus uncertainty (BEPU) method plays a key role in the development of the innovative Generation IV nuclear reactors, for the improvement of knowledge and the good evaluation of the safety margins for new phenomena. The aim of this paper is to validate an uncertainty quantification (UQ) approach using RAVEN code. RAVEN, developed at the Idaho National Laboratory, is a multipurpose probabilistic and uncertainty quantification framework, capable to communicate with any system code, implemented with an integrated validation methodology involving several different metrics. In this activity, a coupled calculation RELAP5-3D/RAVEN has been performed to assess the validity of the embedded UQ approach. The simulations have concerned two tests conducted on NACIE (NAtural CIrculation Experiment) facility, a non-nuclear loop-type system using Lead-Bismuth Eutectic (LBE) as coolant, realized at the ENEA Brasimone Research Centre (Italy) to support the development of the GEN-IV reactors. The experimental tests are aimed to investigate the phenomena related to the natural and gas enhanced circulation flow regimes of heavy liquid metals (HLM), and to test and validate the main components in a LBE environment. A numerical 1D model of NACIE facility has been realized and the post-test analysis has been carried out using RELAP5-3D© ver. 4.3.4. Based on these results, a statistical analysis has been performed using RAVEN computer code, investigating the capability of three probabilistic comparison metrics fully integrated in the code.

Development and validation of a thermal hydraulic transient analysis code for offshore floating nuclear reactor based on RELAP5/SCDAPSIM/MOD3.4

Offshore floating nuclear power plants (OFNPs) can effectively solve the offshore energy supply problem in marine resource development and island construction. Affected by ocean waves and other ocean conditions, the OFNPs can generate different kinds of ship motions, which can oscillate the thermal hydraulic parameters and threaten the reactor safety. In the present study, ocean condition theoretical models are established by considering the effects of three basic movement forms (static inclining, linear motion and rotation motion) as well as the coupled ship motions. A thermal hydraulic transient analysis code for OFNPs is developed by adding ocean condition theoretical models into the RELAP5/SCDAPSIM/MOD3.4 code. The experimental data obtained by zero power loading experiment and single-phase natural circulation experiment under rolling motion are used to verify the ocean condition theoretical models as well as the code modification strategy. Results show that the flow fluctuation behaviors caused by rolling motion can be well simulated by the developed code. The calculation capability of modified RELAP5 code under static inclining and heaving motion is also verified by comparing with RETRAN-02/GRAV code. Besides, the effects of the coupled ship motions on natural circulation system are studied by the modified RELAP5 code. Compared with the basic movement forms, the coupled ship motions can cause greater flow fluctuation and obviously reduce the core flow rate, which means the influence of the coupled ship motions is necessary to be considered in the safety analysis of OFNPs.

Transient safety analysis for simplified accelerator driven system with gas-lift pump

Owing to remarkable advantage for transmutation of the long-lived nuclear waste, Accelerator Driven System (ADS) is one of the most promising systems in nuclear power field. In this study, the transient analysis of a simplified lead bismuth alloy cooling accelerator driven system with gas-lift pump has been numerically simulated. By using point reactor neutron kinetics equation and related thermal-hydraulic models, TASAR (Transient Analysis of Simplified ADS Reactor) code is developed. The five kinds of transient cases are simulated: Starting gas-lift pump (SG), Unprotected loss of flow (ULOF), Beam interruption (BI), Unprotected transient over current (UTOC), Unprotected transient overpower (UTOP). Meanwhile, the calculation results of TASAR code are validated by the experimental data of NACIE (Natural Circulation Experiment) facility. The analysis is as following: In SG and ULOF conditions, the introduction of gas-lift pump has a remarkable influence of loop natural circulation capability. In BT and UTOC conditions, the short-time beam cutting off and a 100% increase of the external neutron source would not lead to the immediate fuel melting. In UTOP condition, a reactivity jump of 2$ would not lead to the damage of fuel and the cladding. The above numerical results show that, the accelerator and gas-lift pump influence the loop mass flow rate and core temperatures of research system in this paper greatly, and the simplified accelerator driven system with gas-lift pump has a very good safety performance.

Air Water Loop for investigaion of flow dynamics in a steam drum: Steady state two-phase natural circulation experiments and validation

Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type boiling water reactor that relies on two-phase natural circulation for heat removal from the nuclear core under normal operating as well as accidental conditions. In case of AHWR, the two-phase flow generated in the core is transported by vertical riser pipes to horizontal steam drums at the top. These horizontal steam drums provide sufficient surface area for the separation of steam water mixture purely based on gravity i.e. due to density difference between steam and water. To investigate this gravity separation phenomenon, an experimental facility known as Air-Water Loop has been designed. The facility has a scaled geometry of AHWR steam drum operating with Air-Water mixture as a working fluid. The required flow conditions for the experimental simulations are generated using air-water two-phase natural circulation in this facility. Steady state natural circulation experiments were performed where measurements on recirculation flow rates, two-phase and single phase pressure drop in various sections of the loop have been carried out. The present paper aims to describe the design aspects of the facility, pre-test calculations, steady state two-phase natural circulation experiments and assessment of measured experimental data. The experimental measurements have been predicted using a numerical model that considers various void fraction and pressure drop correlations from the literature for estimation of measured parameters.

The development and validation of a natural circulation analysis code for marine reactors

ABSTRACTA code PNCMC (Program for Natural Circulation under Motion Conditions) has been developed for natural circulation simulation of marine reactors. The code is based on one-dimensional two-fluid model in noninertial frame of reference. The body force term in the momentum equation is considered as a time dependent function, which consists of gravity and inertial force induced by three-dimensional ship motion. Staggered mesh, finite volume method, semi-implicit first order upwind scheme and Successive Over Relaxation (SOR) method are used to discretize and solve two-phase mass, momentum and energy equations. Single-phase natural circulation experiments under rolling condition performed in Institute of nuclear and new energy technology of Tsinghua University and two-phase natural circulation experiments under rolling condition performed by Tan and colleagues are used to validate PNCMC. The validation results indicate that PNCMC is capable to investigate the single-phase and two-phase natural circulation under rolling motion.

Flow characteristics of natural circulation in a lead–bismuth eutectic loop

Lead and lead-alloys are proposed in future advanced nuclear system as coolant and spallation target. To test the natural circulation and gas-lift and obtain thermal-hydraulics data for computational fluid dynamics (CFD) and system code validation, a lead–bismuth eutectic rectangular loop, the KYLIN-II Thermal Hydraulic natural circulation test loop, has been designed and constructed by the FDS team. In this paper, theoretical analysis on natural circulation thermal-hydraulic performance is described and the steady-state natural circulation experiment is performed. The results indicated that the natural circulation capability depends on the loop resistance and the temperature and center height differences between the hot and cold legs. The theoretical analysis results agree well with, while the CFD deviate from, the experimental results.

Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique because (i) it has an inclined cooling channel with downwards-facing heating surface, of which flow processes are not fully exploited, (ii) it is usually exposed to a low-pressure condition, where phase change causes dramatic changes in the flow, and (iii) the effects of a multidimensional flow are very large in the upper part of the core catcher. These features make computational analysis more difficult. In this study, the MARS code is assessed using the two-phase natural circulation experiments that had been conducted at the CE-PECS facility to verify the cooling performance of a core catcher. The code is a system-scale thermal-hydraulic (TH) code and has a multidimensional TH component. The facility was modeled by using both one- and three-dimensional components. Six experiments at the facility were selected to investigate the parametric effects of heat flux, pressure, and form loss. The results show that MARS can predict the two-phase flow at the facility reasonably well. However, some limitations are obviously revealed.

Coupled simulations of the NACIE facility using RELAP5 and ANSYS FLUENT codes

This work deals with the development and preliminarily assessment of a coupling methodology between a modified version of RELAP5/Mod3.3 STH code and FLUENT commercial CFD code, applied to the NACIE (natural circulation experiment) LBE (lead bismuth eutectic) experimental loop (built and located at the ENEA Brasimone research centre).The coupling tool is used to simulate experiments representative of both natural circulation conditions and isothermal gas enhanced (assisted) circulation. Furthermore, an accidental test reproducing an Unprotected Loss of Flow (ULOF) scenario is also simulated and the outcomes are presented. A preliminary sensitivity analysis has shown that, to guarantee a suitable numerical convergence, the assisted circulation tests require a time step one order of magnitude lower compared to natural circulation ones. The comparison between the RELAP5 stand-alone simulations and RELAP5/FLUENT coupled simulations proved the capability to simulate the thermal-hydraulic behaviour of a loop experimental facility for all the examined conditions.

Design of the Compact Integral Effects Test Facility and Validation of Best-Estimate Models for Fluoride Salt–Cooled High-Temperature Reactors

The capability to validate integral transient response models is a key issue for licensing new reactor designs. The Compact Integral Effects Test (CIET 1.0) facility reproduces the thermal-hydraulic response of fluoride salt–cooled high-temperature reactors (FHRs) under forced- and natural-circulation operation. CIET 1.0 provides validating data to confirm the predicted performance of the direct reactor auxiliary cooling system, used for natural-circulation–driven decay heat removal in FHRs, under a set of reference licensing basis events. CIET 1.0 uses a simulant fluid, DOWTHERM A oil, which, at relatively low temperatures (50°C to 120°C), matches the Prandtl, Reynolds, and Grash of numbers of the major liquid salts simultaneously, at ~50% geometric scale and heater power under 27% of prototypical conditions. CIET 1.0 has been designed, fabricated, filled with DOWTHERM A oil, and operated. Isothermal pressure drop tests were completed, with extensive pressure data collection to determine friction losses in the system. The project then entered a phase of heated tests, from parasitic heat loss tests to more complex feedback control tests and natural-circulation experiments, with the ultimate goal of validating best-estimate FHR models using RELAP5-3D and the novel one-dimensional FHR Advanced Natural Circulation Analysis (FANCY) code. This paper introduces the scaling strategy, design, and fabrication aspects, and start-up testing results from CIET 1.0. The CIET 1.0 model in RELAP5-3D and FANCY is detailed, and verification and validation efforts are presented. For various heat input levels and temperature boundary conditions, mass flow rates are compared between RELAP5-3D and FANCY results, analytical solutions when available, and experimental data, for both single and coupled natural-circulation loops. The study shows that both RELAP5-3D and FANCY provide excellent predictions of steady-state natural circulation in CIET 1.0, with mass flow rates within 13% of experimental data, suggesting that both codes are good candidates for design and licensing of FHR technology.

Experimental and analytical studies on heat transfer in a scaled intermediate heat exchanger with water

The present paper describes an experimental and analytical study in terms of the temperature distribution and heat transfer of an intermediate heat exchanger (IHX). An experimental apparatus of IHX with the primary heat transport system and the secondary heat transport system is made, and forced and natural circulation experiments are conducted using the apparatus. Temperature distributions inside the shell are measured using the fixed thermocouples as well as traversable thermocouples. The Nusselt numbers are investigated based on the measured temperature distributions. One-dimensional and three-dimensional computations are conducted under the forced circulation and natural circulation conditions in order to simulate the experimental results. One-dimensional computation is conducted during the transient cooling down process using a plant system code for the heat transport systems of primary and secondary loops. For the three-dimensional computations, a quasi-steady state condition is analyzed. Good agreement is obtained between the one-dimensional computational results and the experimental results for both flow patterns. The precise temperature distribution inside the shell is computed by the three-dimensional CFD code. Based on the calculated temperature distribution, the heat transfer coefficients are evaluated and converted into the Nu numbers, and are compared with the experimental results. As a result of the 3D computation, the Nu number levels off to a constant value as the Reynolds number decreases to the laminar flow condition.