Thermal-hydraulic Analysis Model Research Articles

Development of transient thermal-hydraulic analysis model for the two-phase loop thermosyphon

To comprehensively and accurately simulate and predict the flow and heat transfer characteristics of the two-phase loop thermosyphon (TPLT), a two-fluid two-phase model was developed. A six-equation set was formulated to account for two-phase flow, encompassing mass equations, momentum equations, and energy equations. The interactions between the fluids in terms of mass, momentum, and energy were also taken into account. To obtain numerical solutions for the partial differential equations within a staggered mesh framework, we employed the 2nd-order Lax-Wendroff scheme to obtain the discrete linear system, complemented by the inclusion of a flux limiter. The Semi-Implicit Method for Pressure-Linked Equation (SIMPLE) algorithm was employed to address the solution of the linear equation system. Data from a small-scale TPLT test was used to assess the accuracy and reliability of both the physical models and the numerical method proposed in this study. Simulations were performed under two working conditions of the thermosyphon with filling ratio of 45 % and 64 %. The simulation results exhibited good agreement with the experimental data, demonstrating the accuracy of the model. Furthermore, the model is capable of providing information on the complex two-phase flow inside the TPLT that cannot be measured experimentally.

Design of the Cooling Water System for the EAST Lower Divertor

The new lower tungsten divertor is a crucial component to achieve the targets of long pulse and high <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\beta $ </tex-math></inline-formula> H-mode operation of Experimental Advanced Superconducting Tokamak (EAST) device. The cooling water system (CWS) of lower divertor should have sufficient heat removal capability and be able to cope with pressure transients caused by the potential water hammer events, which makes the design challenging. In this study, the system process and details of main component are determined based on the design requirements. The thermal-hydraulic analysis model has been developed by Flowmaster (FLOWMASTER) to conduct the heat transfer analysis. Furthermore, the system water hammer transients under one pump trip condition have been analyzed. The results show that the flow distribution of each client is uniform, and the pressure and temperature of key points meet the design requirements. The maximum pressure fluctuation, 0.16 MPa, caused by the sudden closing of the check valve is within the acceptable range. The experiment data are in good agreement with the simulation values, which proves the correctness of the model and the feasibility of the design.

Model development and transient analysis of thermal stratification phenomenon in pool-type sodium-cooled fast reactors

Thermal stratification of pool-type SFR under transient accident conditions has a very significant impact on the performance of key components of the primary loop and the establishment of natural circulation of the residual heat removal system. In this paper, combined with the mechanism of thermal stratification, a three-dimensional (3-D) system analysis model is developed, which takes into account the energy source model, the 3-D convection model and the large space inlet-effect model. Based on the model development, the accuracy of 3-D thermal–hydraulic analysis model and the capability of thermal stratification analysis for SFR were fully verified. The validation results show that the method established in this paper takes into account accuracy and calculation efficiency. It can greatly reduce the amount of numerical grid, take up very little computing resources and greatly improves analyzing efficiency. And it can accurately simulate the thermal stratification phenomenon in the sodium pool. At the same time, the location with maximum temperature gradient of the inner barrel wall can be suggested in the process of thermal stratification. Furthermore, this study provides an important analytical method for the transient safety analysis of SFR, and has a great academic value and engineering significance.

Steady-State Thermal-Hydraulic Analysis of Steam Generator with Axial Economizer Based on Node Method

Based on the one-dimensional flow hypothesis, heat transfer hypothesis, and two-phase thermal equilibrium hypothesis, a lumped parameter method and a distributed parameter method are used to establish a one-dimensional steady-state thermal-hydraulic analysis model of the steam generator with axial economizer. Using C++ language programming, the calculation results are compared with the design values of the thermal-hydraulic parameters of a typical steam generator with axial economizer. The results show that the relative error of most of the overall parameter calculation results is within 3%, which verifies the rationality of the model. The analysis of the temperature, cavitation share, pressure and other parameters of the steam generator along the direction of the measured fluid flow trend consist well with the results of thermal hydraulics and qualitative mechanism analysis, indicating that the model and solution method established in this paper can accurately predict the steady-state thermal hydraulic parameter distribution of the steam generator with axial economizer.

A thermal-hydraulic analysis model of printed circuit heat exchangers for system simulation using Modelica

ABSTRACT Supercritical CO2 (S-CO2) Brayton cycle is one of the alternative energy conversion technologies for generation IV nuclear energy systems. Printed circuit heat exchanger (PCHE) is one of the key components in S-CO2 Brayton cycle. Although PCHE has dedicated steady-state models, transient models that can be applied to precise analysis of systems remain to be developed. The deviations brought by different flow and heat transfer correlations needs to be studied in simulations of PCHE, considering that the disparity of results obtained by various correlations under different working conditions are more obvious in iterative calculations. We developed a PCHE model in Modelica according to the existing S-CO2 flow and heat transfer correlations. To benchmark the model, steady-state results are compared with the results obtained by Fluent. The deviations caused by various flow and heat transfer correlations are analyzed. Transient analysis is performed to compare different control strategies when reducing the power of PCHE. According to the results of the transient simulation, PCHE used as a pre-cooler in S-CO2 Brayton cycle can quickly respond to transient variations. Both steady-state and transient results verified the reliability of the PCHE model, which suggests that the model can be utilized for the system analysis of S-CO2 Brayton cycle.

Concept design of primary heat transfer system for CFETR vacuum vessel

The Chinese Fusion Engineering Testing Reactor (CFETR) aims at providing technical support and data reference for future fusion demonstration reactor. The vacuum vessel primary heat transfer system (VV PHTS), playing a vital role in guaranteeing the safety operation of tokamak device, which is a part of the tokamak cooling water system. It will operate in three main modes: plasma operation, water baking and decay heat removing. The design of the VV PHTS should have three functions corresponding to the above modes. Based on the thermal-hydraulic requirements, the system process have been proposed. The concept design of the system has been performed including pipes and pressurizer. For the purpose of obtaining the parameters of other main components and predicting the system operating behaviour, the thermal-hydraulic analysis model has been developed by using AFT fathom mode. The steady-state results of all three operating modes such as velocity and pressure show that the designed VV PHTS and selected methodology are feasible.

Development of a model for thermal-hydraulic analysis of helically coiled tube once-through steam generator based on Modelica

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a critical component for compact small reactors. The present study developed and verified a model in Modelica language and simulated steady-state and transient situations of the model. The Modelica model takes double-pipe heat exchanger as a prototype and adopts counter flow design. Different heat transfer correlations and pressure drop correlations for helically coiled tubes are reviewed and applied to the model. The model is verified by comparing the simulation results with the design parameters of the Marine Reactor X steam generator, and with the results performed by other dedicated programs. After verification, a 100 MW H-OTSG is proposed and steady-state conditions of the four main operating conditions, i.e. 100 MW, 75 MW, 50 MW, 25 MW, are simulated, and analyzed. Transient analysis is then performed in order to investigate the effects of different control strategies of the H-OTSG. The H-OTSG model developed with Modelica can be utilized for the safety analysis of small reactors in the future.

Development of thermal-hydraulic analysis code of a helically coiled once-through steam generator based on two-fluid model

A thermal-hydraulic analysis model, coupling the flow and heat transfer of the primary and secondary sides, is developed to describe the thermal hydraulic behavior of OTSG (helically coiled Once-Through Steam Generator). This numerical model is developed based on the two-fluid model with distributed parameters method to predict all flow variables at each position along the tube. Correlations, validated a lot, are adopted to consider the effect of helical structure on the flow and heat transfer of both the primary and secondary sides in OTSG. The computational code THOSG (Thermal-Hydraulic analysis code of Once-through Steam Generator) for OTSG is then developed based on the numerical model with the modified SIMPLE algorithm. To benchmark the developed physic model and computer code, steady-state simulation of OTSG of IRIS reactor is conducted. Results are compared with RELAP5 code with respect to the design parameters. Comparisons indicate that the THOSG code can predict the thermal-hydraulic characteristics of OTSG well. In order to investigate the effect of changes in the input conditions on the output of the OTSG, transient analysis has been performed. Simulations of change in feedwater flow rate and temperature are conducted to assess the performance of OTSG. Both steady-state and transient results indicate that the THOSG code can be utilized for the design and performance analysis of OTSG. However, further analysis, coupling with the thermal-hydraulic behavior of the reactor core, is required to comprehensively evaluate the safety and reliability of OTSG design in nuclear power plant.

Regulation of the temperature in the ampoule channel with natural circulation of coolant

It has been shown by calculations that it is possible to extend considerably the capabilities for control of temperature conditions in an ampoule channel with natural coolant circulation, using the proposed hydraulic circuit layout, on samples during irradiation in the SM-3 reactor reflector cell by changing the circulation circuit geometry through the arrangement of a bypass heat removal line formed in the upper part above the flow limiter as compared to control only by changing the thermal conductivity of the gas gap in the channel body (through changing the gas pressure or composition). The ampoule channel test conditions, layout and simulation model for thermal-hydraulic analysis using the RELAP5/MOD3.2 code are presented. An investigation was conducted to study the effects of the bypass cooling circuit on the temperature conditions during irradiation of samples in an ampoule channel. The bypass flow rate change is achieved by varying the passage area of the flow limiter orifice. Options have been considered for filling the channel body gas gap with helium and a helium mixture. The calculation showed that the heat removed by the bypass line could reach 40% of the total heat released in the channel. With helium used in the channel body gap, the temperature conditions during irradiation are adjusted in a broader range (200–330 °С) than with a gas mixture of a lower thermal conductivity (279–330 °С), the major temperature variation taking place with the flow limiter orifice area being less than 0.2–0.3 cm2. Any further increase in the flow limiter orifice area does not lead to a major temperature change in the coolant flowing about the samples.

Design of coolant distribution system (CDS) for ITER PF AC/DC converter

The Poloidal Field (PF) converter unit, playing an essential role in the plasma shape and position control in vertical and horizontal direction, which is an important part of ITER power supply system. As an important subsystem of the converter unit, the coolant distribution system has the function to distribute the cooling water from ITER component cooling water system (CCWS) to its main components at the required flow rate, pressure and temperature. This paper presents the thermal hydraulic design of coolant distribution system for the ITER PF converter unit. Different operational requirements of the PF converter unit regarding flow rate, temperature and pressure have been analyzed to design the system process and arrangement. A thermal-hydraulic analysis model has been built to size the system and predict the flow rate and temperature distribution of the system under the normal operation. Based on the system thermal-hydraulic analysis results, the system pressure profile has been plotted to evaluate the pressure behavior along each client flow path. A CDS prototype for the ITER PF converter has been constructed and some experiments have been performed on it. A good agreement of the flow distribution and temperature behavior between the simulated and test results validate the proposed design methodology.

Thermal–Hydraulic Analysis of ITER Component Cooling Water System Loop 2B

ITER cooling water system includes Tokamak cooling water system, component cooling water system (CCWS), chilled water system and the heat rejection system. The CCWS is further divided into CCWS-1, CCWS-2A, CCWS-2B, CCWS-2C, and CCWS-2D loops to provide independent chemical control and to prevent galvanic corrosion among the different clients’ materials (e.g. Cu, Al). The CCWS-2B is responsible to remove heat load generated by coil power supply components and the neutral beam injectors and diagnostics system during all the phases: commissioning, testing and conditioning and plasma operation. A CCWS-2B thermal–hydraulic analysis model was developed, by using the AFT Fathom code, to conduct the steady state thermal–hydraulic analysis of the system. In this thermal–hydraulic analysis model, the critical path with the largest pressure loss was used to size the pump head has been identified and the pressure loss on the control valve were used to establish the required flow balance at each piping connection points. This paper presents the results of this thermal hydraulic analysis which was composed by required pump head of the CCWS-2B loop and the main thermal–hydraulic parameters for each client (i.e. flow rate, velocities, pressure drops and outlet temperatures).

Thermal behaviour analysis on ITER component cooling water system loop 2B

ITER cooling water system is composed by several cooling loops, the primary heat transfer loops that form the Tokamak Cooling Water System (TCWS), the secondary heat transfer loops that form the Component Cooling Water System (CCWS) and the Chilled Water System (CHWS) and a tertiary heat transfer loop which is the Heat Rejection System (HRS). The CCWS is further divided into CCWS-1, CCWS-2A, CCWS-2B, CCWS-2C, CCWS-2D depending on the water chemistry needs of clients and wetted area material. The component cooling water system loop 2B (CCWS-2B) has the function to remove heat load from coil power supply component, Neutral Beam Injectors (NBIs) system component and diagnostic system which are located in different buildings. As the total number of the client connections for the loop is a few hundreds, simplified thermal hydraulic analysis model has been developed to perform thermal analysis on the component cooling water system loop 2B. The curve of the cooling water temperature at client inlet and outlet during one cycle of the most demanding plasma operation scenario was obtained and the cooling water flow rate can meet the thermal removal requirement of client was also confirmed from this analysis. In addition, operation behaviour of the main heat exchanger for CCWS-2B in this thermal analysis was depicted for main heat exchanger selection purposes. This study has been carried out with the AFT Fathom code.

Mithrandir+: a two-channel model for thermal-hydraulic analysis of cable-in-conduit super-conductors cooled with helium I or II

Mithrandir [1]– a 1-D code for the analysis of thermal-hydraulic transients in cable-in-conduit super-conductors (CICC) with cooling channel – has been extended to helium II by including Gorter-Mellink heat transport. The code treats the general case of different thermodynamic properties and flow velocities for the helium in the cable bundle region and that in the cooling channel, which are coupled to conductor and jacket at different temperatures. A detailed validation of the code against a thermal-hydraulic experiment in helium II is presented, showing good agreement, and code convergence is demonstrated numerically. We present results of simulations of the QUench Initiation and Propagation Study (QUIPS) experiment, concentrating mostly on the initial phase of the quench. We show that the simulation is at least qualitatively in agreement with experimental data.

Neutronics and Thermal-Hydraulics Analyses of the Pellet Bed Reactor for Nuclear Thermal Propulsion

Neutronics and thermal-hydraulics design and analyses of the pellet bed reactor for nuclear thermal propulsion are performed based on consideration of reactor criticality, passive decay heat removal, maximum fuel temperature, and subcriticality during a water flooding accident. Besides calculating the dimensions of the reactor core to satisfy the excess reactivity requirement at the beginning-of-mission of 1.25 $ (K{sub eff} of 1.01), the TWODANT discrete ordinates code is used to estimate the radial and axial fission power density profiles in the core. These power profiles are used in the nuclear propulsion thermal-hydraulic analysis model (NUTHAM-S) to determine the two-dimensional steady-state temperature, pressure, and flow fields in the core and optimize the orificing in the hot frit to avoid hot spots in the core at full-power operation.