Once-through Steam Generator Research Articles

Simulation Research of Once-through Steam Generation Tube Rupture in an Improvement Model

Simulation Research of Once-through Steam Generation Tube Rupture in an Improvement Model

Comparing the dynamic response of U-tube and helical coil steam generators in small modular reactor flexible operation

A steam generator (SG) is one of the most important subsystems in a pressurized water reactor (PWR). As the interface between the reactor and the turbine for power/energy transform, its behavior dictates the response of the entire plant, especially when maneuvering the power output under dynamic operation. Generally, two types of SGs are most commonly employed in PWR designs: a recirculation-type U-tube steam generator (UTSG) and a once-through steam generator (OTSG). Among the PWR-type SMR concepts, some designs use a UTSG while several others rely on the helical coil steam generator (HCSG), which is a type of OTSG. These two designs differ in a variety of ways, namely construction, physical size for a given capability, thermodynamic characteristics, and control techniques – leading to a very different dynamic behavior under flexible operation. This paper examines the dynamic characteristics of a UTSG and a HCSG in the context of SMR flexible operation and highlights the key implications of choosing one against the other. Detailed moving boundary models are developed for both SGs. Simulation studies have been carried out in MATLAB/Simulink environment to compare their characteristics quantitatively under different loading conditions. The results have indicated that there are significant differences in dynamic behaviors between these two SGs during transient operations. Relatively larger pressure excursions in the UTSG during power maneuvers suggest more stringent considerations in equipment design and selection for safe operation.

A movable boundary model for helical coiled once-through steam generator using preconditioned JFNK method

Helical coiled once-through steam generator (H-OTSG) is the largest and the most complicated heat exchanger in the high-temperature gas-cooled reactor (HTGR), whose performance plays an important role in the nuclear power plant design and dynamic behavior analysis. A mathematical model of H-OTSG is developed based on the movable boundary method and solved by the advanced fully implicit Jacobian-free Newton-Krylov algorithm. The physical-based preconditioners are proposed and analyzed in this work, which achieves high-performance in solving the nonlinear steam generator system. In order to evaluate the performance of the movable boundary model, the fixed fine mesh model is also considered. The comparison is made between these two models from the perspective of the computational efficiency and accuracy. The simulation results are validated by the HTR-10 steam generator design data, showing that both models agree well with the design data, but have their own features. The fixed fine mesh model has its advantages in the accuracy, while the movable boundary model can realize the high computational efficiency.

Numerical analysis of liquid metal helical coil once-through tube steam generator

Much attention has been focused on the Liquid Metal Fast Reactor (LMFR) as one of the most promising concepts for GEN-IV reactors. Helical coil once-through tube steam generator (HCOTSG) is a proposed form of steam generator. Due to its unique advantages, HCOTSG is widely used in various reactor power systems, including LMFR. In this paper, a numerical approach is proposed to simulate heat transfer from the primary side (liquid metal) and secondary side (water/steam). The liquid lead-bismuth eutectic (LBE) is flowing in the shell side and pressurized water is flowing in the tube side. FLUENT CFD commercial code is adopted for the simulations. The robustness of the method is validated by comparing it with relevant experimental research, and the maximum error is less than 25%. Based on this method, the distribution of thermal-hydraulic parameters inside HCOTSG is obtained and the flow and heat transfer characteristics are analyzed. The influence of boundary conditions and geometric parameters on the flow and heat transfer in HCOTSG is analyzed. The comprehensive performance of different models is compared with the comprehensive performance evaluation index. The optimal geometric arrangement scheme for the working conditions is recommended. Compared with the reference model, the comprehensive performance can be improved by up to 14%. This study provides a numerical simulation method reference for the study of flow and heat transfer characteristics and structural design optimization of liquid metal HCOTSG.

Experimental & numerical study of different orifice configurations for helical coil Once Through Steam Generator

The main aim of the present study is to aid the selection of suitable orifice configuration for helical coil once-through steam generators being designed for SMRs. The selection of orifice is based on the pressure drop occurring in the orifice over the desired flow range, the minimum length required for the orifice, repeatability of the results, and ease of fabrication.In the present study, three different tube inlet orifice configurations are evaluated for a helical coiled Once Through Steam Generator. Orifice length is derived for all three configurations based on the orifice pressure drop required for avoiding flow instabilities. Orifice pressure drop required to avoid flow instabilities is estimated as factor k times the sum of two phase pressure drop and super-heating zone pressure drop of the tube side flow, where minimum value of k is taken as 2 (Han et al., 2019; Kang et al., 2007). Various power levels (10%–100%) are considered to obtain the governing length of the orifice. Pressure drop across the orifice over the flow range 10%–100% is estimated theoretically using various correlations available in the literature (Idelchik). Numerical simulations are carried out for all three configurations of the orifice. Flow through the orifice falls in the turbulent regime and hence, three different turbulent models viz. k-ε model, RNG k-ε model and Realizable k-ε model are compared for their capability to predict pressure drop across the orifice. Theoretical and Numerical results are verified with experimental pressure drop measurements over the flow range corresponding to a power range of 10%–100%. The effect of manufacturing tolerances on pressure drop is studied experimentally for all three configurations. Five nos. of each orifice configuration were manufactured and tested experimentally for pressure drop over 10%–100% flow range to see the effect of manufacturing tolerances.It is concluded from the study that irrespective of the orifice configuration, the minimum length required for the orifice is governed by the lowest power operation i.e, the lowest flow operation. Also, a comparison of various turbulence models viz. k-ε model, RNG k-ε model and Realizable k-ε model, reveals that, RNG model is best suited for orifice configurations that utilize sudden expansion and contraction for creating pressure drop. Whereas, all three turbulence models compared in the present study gives similar results for orifice configurations in which pressure drop occurs mainly due to wall friction and gradual but smooth change in flow direction. Orifice configuration-3 which utilizes a helical flow path is finally selected based on the selection criteria stated above.

Large-leak sodium–water reaction accident analysis of sodium-cooled fast reactor

A sodium-cooled fast reactor (SFR) is one of Generation IV nuclear systems. The once-through steam generator is important for heat transfer and acts as a barrier to separate the sodium and water. If the heat transfer tube breaks, a sodium–water reaction (SWR) occurs, and the integrity of the secondary loop may be destroyed. Thus, a dynamic model to analyze the SWR must be developed, especially for a large-leak SWR. A large-leak SWR analysis model was developed for an SFR, consisting of four modules: water/steam leakage rate, hydrogen bubble growth, pressure wave propagation, and SWR protection system action. A large-leak SWR accident was simulated using this model, and the process was investigated. Sensitivity analysis was performed for the critical parameters, and the parameters with a key influence on the large-leak SWR were determined. The results are significant for the design of the steam generator and the SWR protection system.

Mineralogical characterization and thermodynamic modelling of scales formed in once through steam generators

Extraction of hydrocarbons from oil sand deposits relies on technologies designed to lower bitumen viscosity, either by thermal energy or combination of thermal energy and solvents. Thermal energy is commonly delivered by steam. Once-through steam generators (OTSGs) are the most common class of generators used to generate steam from boiler feed water (BFW) at high pressure and temperature. Generated steam is used to mobilize the bitumen in the in situ reservoirs, and then water-bitumen emulsion is produced and processed, separating a hydrocarbon product and oil-free water, which subsequently will be reused to generate steam after treatment. Some of the aqueous constituents remaining or added during steam interaction with oil sands and subsequent treatment contribute to scale growth in OTSG tubing. Accumulation of the scale in OTSGs may lead to tubing failure due to overheating.In the present work we focused on inorganic scales – solid phases formed by components dissolved in BFW. We used a suite of geochemical, mineralogical and modelling techniques to characterize and model the inorganic components of scale samples provided by Canada's Oil Sands Innovation Alliance (COSIA) from OTSGs operators at multiple sites in Northeastern Alberta, Canada.We identified the composition of minerals in the scales as Na-, Ca-, and Mg-silicates with relatively low SiO44- content (inosilicates). Our modelling results allowed us to constrain the evolution of pressure, temperature, pH, mineral phases, and dissolved components in BFW as it undergoes heating in the OTSG to form steam and boiler-blowdown water. Modelling results are consistent with observed minerals and are promising for future management of scale formation processes.

Investigation of two-phase natural circulation with the SMART-ITL facility for an integral type reactor

A two-phase natural circulation test using SMART integral test loop (SMART–ITL) was conducted to explore thermo-hydraulic phenomena of two-phase natural circulation in the SMART reactor. Specifically, the test examined the natural circulation in the primary loop under a stepwise coolant inventory loss while keeping the core power constant at 5% of the scaled full power. Based on the test results, three flow regimes were observed: single-phase natural circulation (SPNC), two-phase natural circulation (TPNC), and boiler–condenser natural circulation (BCNC). The flow rate remained steady in the SPNC, slightly increased in the TPNC, and dropped abruptly and maintained in the BCNC. Using a natural circulation flow map, the natural circulation characteristic in the SMART–ITL was compared with those in pressurized water reactor simulators. In the SMART–ITL, a BCNC regime appeared instead of siphon condensation and reflux condensation regimes because of the use of once-through steam generators.

Model Development and Transient Analysis of the WCLL BB BOP DEMO Configuration Using the Apros System Code

Extensive modelling and analytical work has been carried out considering the water-cooled lithium–lead breeding blanket (WCLL BB) balance of plant (BOP) configuration of the demonstration power plant (DEMO) using the Apros system code, developed by VTT Technical Research Centre of Finland Ltd. and Fortum. Contributing to the BOP work package of the EUROfusion Consortium, the integral plant model for dynamic analyses of the WCLL BB configuration has been updated with special attention to primary system components. Following trends of relevant neutronics modelling, a new BB model has been implemented in 2020 with the aim to obtain higher resolution output data and a more realistic thermalhydraulic feedback from the primary system. Once-through steam generator user components have been built based on CAD models conceived by BOP partners. Transient analyses have been performed providing a better picture regarding the behaviour of main components, e.g., the BB and the OTSGs, whilst highlighting possible ways to optimise the control scheme of the plant.

Parametric sensitivity analysis of liquid metal helical coil once-through tube steam generator

Lead-bismuth Fast Reactor (LFR) has attracted attentions due to its intrinsic characteristics, such as sustainability, safety, and economics. Helical coil once-through tube steam generator (HCOTSG) is a proposed form of steam generator. Due to its unique advantages, HCOTSG is widely used in various reactor power systems, including LMFR. In this paper, FLUENT CFD commercial code is adopted to simulate heat transfer from the primary side (liquid metal) and secondary side (water/steam). The robustness of the method is validated by comparing it with relevant experimental research, and the maximum error is less than 25%. Based on this method, 9 models with different geometrical parameters were established to analyze the effects of geometrical parameters on the performance of liquid metal HCOTSG. The wall heat flux and comprehensive performance factor (En) of different models is compared. The optimal geometric arrangement scheme for the working conditions is recommended. The geometrical parameters of the tube side and shell side are analyzed as sensitivity parameters. Among the established geometric models, the HCOTSG has the best comprehensive performance when Pr /D is 0.167 and Pa /D is 0.137 under the working conditions in this paper. This study provides a numerical simulation method on structural design optimization of liquid metal HCOTSG.

Prediction method for thermal-hydraulic parameters of nuclear reactor system based on deep learning algorithm

The nuclear reactor core and the steam generator are the core components of nuclear reactor system, which directly determine the heat releasing and carrying capacity of nuclear reactor system. The traditional thermal-hydraulic design of nuclear reactor system generally selects the linear iterative mode, which requires the repeated “calculation-evaluation-correction”. The whole process is tedious and time-consuming, which also relies on the design experience. In recent years, the successful application of artificial intelligence technology in many fields has brought new enlightenment to the innovation for the thermal-hydraulic design of nuclear reactor system. In this paper, a prediction program for the thermal-hydraulic parameters of nuclear reactor system was established by means of deep learning algorithm, which was mainly composed of the neural network module, the activation function module, the error function module, the optimization algorithm module and the data initialization module. Then, the KLT-40S nuclear reactor core and the tube-in-tube once-through steam generator were selected as the analysis objects, and the corresponding input card of RELAP5/SCDAPSIM program was built. Based on the prediction program, the calculation results of RELAP5/SCDAPSIM program were studied, which could realize the rapid prediction for the thermal-hydraulic parameters of nuclear reactor system, and were in good agreement with the calculation results of RELAP5/SCDAPSIM program.

Investigation of Operation of Coil-Flow Steam Generator of Serpentine Type in Conditions of Low Ambient Temperatures

Stationary and mobile steam generators are widely used in low ambient temperatures, for example in areas described by the Köppen world map as subarctic. Such equipment is often used in oil and gas fields. At the moment, the existing standard boiler plants are outdated. The purpose of this work is an experimental study of a coiled-type direct-flow steam generator developed by the authors in the winter period at low ambient temperatures. The tasks to be solved to achieve the goal are associated with obtaining experimental data at different operating modes of the installation, their processing and the development of empirical coefficients of gas movement inside the coaxial cylinders of the steam generator. In addition, another task is to develop a theoretical basis for the obtained experimental data. Based on the results of the work, the dependences of pressure and temperature on fuel consumption in various modes were obtained. Statistical analysis was applied to the data obtained. The authors have developed equations for calculating the convective part in the process of radiant-convective heat transfer in coaxial gas ducts, taking into account the design features of a once-through coil-type steam generator. Finally, promising directions for further improving the efficiency of steam generators of this type are proposed.

A Method Study of Numerical Simulation on Flow Field of Spiral Tube in Steam Generator Shell Side

The lead based fast reactors often use integrated core design scheme, due to its compact structure, and the helical coil tube type once-through steam generators (OTSG) have been utilized in various reactors. In order to study the flow and heat transfer characteristics of the lead bismuth eutectic based coolant in the shell side of the helical coil tube OTSG in lead based fast reactors, a section-by-section solution method is adopted, and the wall heat flux fitting formula is used to simulate the flow field of shell side, with the help of FLUENT software. Finally, the correctness of the section-by-section solution method is verified, the flow and heat transfer characteristics of the shell side are analyzed, and the calculation data of velocity, temperature and pressure field are obtained. Its results lay a foundation for the next step of steam generator flow induced vibration analysis and high temperature stress calculation.

Development of a thermal-hydraulic analysis code for thermal sizing of once-through steam generators for SMR

A one-dimensional thermal-hydraulic analysis code was developed for the conceptual design and sizing of once-through steam generators (OTSGs) with application to small modular nuclear reactors. The OTSG in the model is represented by one characteristic double-tube heat exchanger with countercurrent flow. The secondary side of the OTSG is divided into subcooled liquid region, saturated two-phase mixture region, and superheated vapor region. Heat transfer and pressure drop correlations for each flow region were selected and implemented. To benchmark predictability of the developed computer code, the B&W OTSG whose design data is available in open literature was simulated. The results were compared both to the design data of the OTSG and to the analysis results obtained using a best-estimate thermal-hydraulic analysis code. The overall characteristics of the heat transfer area, pressure, and temperature distributions calculated with the developed code were in good agreement with the design data as well as the results from the best-estimate code. Lastly, a new concept of OTSG that incorporates feedwater circulation to reduce its volume for SMR application was simulated using the developed code. This demonstrates that the code can be used for design innovation and performance analysis of OTSGs.

Sizing of a tube inlet orifice of a once-through steam generator to suppress the parallel channel instability

Sizing the tube inlet orifice of a Once-Through Steam Generator (OTSG) is important to protect the integrity of the tubes from thermal cycling and vibration wear. In this study, a new sizing criterion is proposed for the tube inlet orifice to suppress the parallel channel instability in an OTSG. A perturbation method is used to capture the essential parts of the thermal-hydraulic phenomena of the parallel channel instability. The perturbation model of the heat transfer regime boundaries is identified as a missing part in existing models for sizing the OTSG tube inlet orifice. Limitations and deficiency of the existing models are identified and the reasons for the limitations are explained. The newly proposed model can be utilized to size the tube inlet orifice to suppress the parallel channel instability without excessive engineering margin.

Compact Steam Bottoming Cycles: Minimum Weight Design Optimization and Transient Response of Once-Through Steam Generators

Compactness and weight minimization are key aspects for successful and widespread implementation of waste heat recovery steam cycles in off-shore oil and gas platforms due to the site weight and volume footprint constraints. The power plant off-shore must be designed for flexibility in its operations to provide varying power demands across multiple time scales. Reliability of the heat and power production units is crucial. Within a case study in an off-shore platform in the Norwegian Continental Shelf, this article conducts design optimization of compact and low-weight steam cycles for power production from gas turbine exhaust and transient analysis of the core of heat recovery steam generators (HRSGs) via dynamic modeling and simulation, considering once-through steam generators (OTSGs) for the HRSGs. A method for simultaneous thermodynamic and heat exchanger geometry optimization design for bottoming cycles is applied, with the main objective being weight minimization and compactness of the cycle heat exchangers. Ten different optimal minimum weight bottoming cycle designs are provided by selecting ten different manufacturable tubes. The resulting bottoming cycle designs are compared in terms of weight, OTSG core weight distribution, heat transfer area, and footprint. The resulting bottoming cycle weight varies from 48.4 to ca. 87.10 ton for designs sensible for off-shore applications, and from 95.8 to 178.9 ton when selecting outer tube diameters typical of onshore applications. Smaller outer tube diameter selection in OTSG bundles is a key driver for low-weight and compact steam cycle designs. Three different designs representing light, normal, and heavy OTSG designs are compared by dynamic trajectory and response time analysis under transient scenarios by means of dynamic modeling and simulation. More compact and lighter designs respond faster to changes in the gas turbine (GT) operation upstream the OTSG. The results in this analysis indicate the need for feedforward control. Feedback control alone is probably not a good option due to the high OTSG open loop stabilization time and large sensitivity to GT exhaust gas variations. More compact and low-weight designs based on the OTSG can reduce potential challenges in controlling and stabilizing bottoming cycles for power production.

Study on deformation mechanism of Ti-2Al-2.5Zr alloy tube in the flattening test

The flattening test, using one pair paralleled plates to flatten tube structure radially, is generally employed to examine the ductility of tube used in once through steam generator (OTSG). This study focuses on the stress condition and deformation mechanism analysis on the concentrated stress regions of Ti-2Al-2.5Zr alloy tube during the flattening test. Firstly, the finite element analysis was performed using the commercial software ABAQUS to determine the stress condition. Secondly, the Electron Back-Scattered Diffraction (EBSD) was implemented to observe the microstructure evolution of Ti-2Al-2.5Zr alloy. Finally, Schmid law was employed to analyze the activated deformation mechanism under condition of the complex stress. It was found that the condition of the complex stress in stress concentration regions, including tension and compression regions, can be simplified into two directional stresses condition. In grains whose c-axis is nearly towards TD and ND, the strain is mainly accommodated by prismatic slip, whereas in grains whose c-axis deviates about 45° from ND to TD mainly by basal slip. The {10–12} extensive twin and corresponding parent grain orientation mainly relied on the stress condition. Additionally, the intergranular twin pairs connected at common grain boundaries (GBs) with high strain compatibility were found at low angle GBs, which were activated under a combination of macro stress and strain compatibility effect.

Accuracy and efficient solution of helical coiled once-through steam generator model using JFNK method

The Helical coiled tube Once-Through Steam Generator (H-OTSG) is key equipment in the High-Temperature Gas-cooled Reactor (HTGR) system, where the fission energy is transferred from the helium side to the water side, together with the complicated heat transfer and phase change phenomenon. In detail, the water is heated to the water-steam two-phase flow by the hot helium and further to the superheated steam. As a result, the H-OTSG is a multi-phases, multi-physics, and multi-domain highly nonlinear system due to the complex constitutive equations in the two-phase flow model. The accuracy and efficient solution of this highly nonlinear system is a key point for the successful simulation of the H-OTSG. In this work, the advanced Jacobian-Free Newton Krylov (JFNK) method is used instead of the traditional semi-implicit iteration method to pursue high computational performance. An efficient physical-based preconditioner is proposed, which is derived from the original semi-implicit based method. Moreover, the numerical technologies of the phase generation/disappearance and the global convergence algorithms of JFNK method are implemented and discussed. The two-phase drift-flux model is used to describe the two-phase flow in the waterside and the second-order accuracy temporal and spatial numerical scheme is employed to discrete the conservation equations. A new H-OTSG simulation code based on JFNK method is developed in this work and applied to analyze the steady and transient state behavior of the H-OTSG in the 10 MW high-temperature gas-cooled reactor (HTR-10). The simulation results show that the accuracy of the newly developed H-OTSG simulation code agrees well with the design data under different operating conditions. Furthermore, the computational performance of the JFNK method with the physical-based preconditioner is the best compared with the traditional Picard iteration method, Newton–Krylov method and JFNK method without preconditioner.

Start-up characteristics for the test facility of a prototype sodium heated OTSG

The test facility of the prototype sodium heated once-through steam generator (OTSG) is used to verify the accuracy and reliability of the design for the sodium heated OTSG. Besides, according to the test data, the structure and parameters of steam generator can be improved. In this paper, the start-up mode of the test facility of the prototype sodium heated OTSG was pre-analyzed by using the modified RELAP5 MOD4.0. Two kinds of modes to reach the steady state, one is step change mode the other is linear change mode. These two modes are adopted respectively to analyze the thermal hydraulic characteristics of prototype sodium heated OTSG under different conditions. The results show that the step change mode can be used to reduce the time of reaching steady state in low power starting. When the loop is running under high power condition, step change mode can reduce the temperature raise rate and increase the safety of the system.

Innovation Control Strategy Simulation Research of Once-through Steam Generator

The new type of nuclear power plant adopts once-through steam generator (OTSG) which can supply steam to the turbine, control and operation of OTSG is different from the traditional natural circulation steam generator, the control strategy of the traditional natural circulation steam generator is not applicable to OTSG. In order to meet the demand of the general operation of the integrated reactor, the operation control of OTSG is put forward. The paper proposes a double const control strategy that the average temperature of the primary circuit coolant and the secondary circuit steam pressure are kept constant, also proposesa coordinated control strategy considering reactor and turbine as a whole. The control strategy is different the traditional control method. In order to verify the control strategy, simulation experiments are carried out in a full-scale simulator system. The simulation results show that the proposed control strategy of OTSG in integrated reactor to meet the load demand, also can effectively guarantee the dynamic characteristics of pressure steam generator and related parameters of OTSG.