Steam Generator Design Research Articles

Vertically aligned Juncus effusus fibril composites for omnidirectional solar evaporation

Despite tremendous efforts to improve solar-driven steam generation, most solar evaporators are limited to unidirectional evaporation due to changes in the direction of solar irradiation under real-world conditions. Hence, the development of non-unidirectional evaporators, ideally allowing omnidirectional evaporation, remains a significant challenge. In this work, the first-ever vertically aligned activated carbon–juncus effusus (AC-JE) evaporator (VACJE) for omnidirectional evaporation is demonstrated. AC-JEs are produced by uniformly decorating activated carbon powder (a cost-lost material with broad efficient solar absorption) on the fibril skeletons of JE with interconnected network architectures. The surprising photothermal performance (the evaporation rate of 2.23 kg m−2 h−1 under 1 sun illumination) of the VACJE is attributed to the unique inbuilt interconnected network architecture of JE. This structural feature multiplies the available evaporation area to promote omnidirectional solar evaporation, while also providing excellent light absorption thermal management and salt-resistant. This unexpected finding reveals the hidden potential of JE as a low-cost biomass material and provides inspiration for the future design and development of high-performance solar-driven steam generators.

Tube inlet orifice design of a once-through steam generator for flow stabilization

A numerical study is conducted on the secondary side screw-type tube inlet orifice design of a once-through steam generator. An orifice length criterion for flow stabilization is derived by introducing the hydraulic resistance ratio of the orifice and the subcooled region to the two-phase and superheated regions. Various tube plugging conditions and power levels are considered, and the secondary coolant pressure at the tube outlet is adjusted to maintain a constant thermal power. Comprehensive numerical solutions are acquired to evaluate the minimum orifice length under various operating conditions. The results obtained show that a constant thermal power is maintained by properly adjusting the secondary coolant outlet pressure with a variation of the superheat degree and secondary coolant pressure drop when the steam generator operates at high power level. The steam generator performance is analyzed according to the tube plugging condition in terms of the degree of superheat, secondary side pressure drop, temperature distribution, and quality distribution. The secondary side outlet pressure curve for the constant thermal power operation is obtained, and the required minimum orifice length to suppress the flow oscillation below the allowable level is evaluated. The lowest power level results in the highest minimum orifice length, and non-plugging condition provides a limiting case for the orifice length criterion except near the 100 % power level.

Void fraction measurements of steam–water two-phase flow in vertical rod bundle: Comparison among different techniques

Void fraction is a fundamental parameter to characterize two-phase flows and is very important for the proper design and safety operation of steam generator in a nuclear power plant. In the present work, void fraction measurements of steam–water two-phase upward flow in a triangular-configured vertical rod bundle were conducted at pressures from 5 to 9 MPa and mass velocities from 100 to 350 kg·m−2·s−1. Local, chordal and average void fractions were measured by the optical probe (OP), gamma-ray densitometry (GD), and differential pressure (DP) methods respectively. The effects of pressure and mass velocity on void fractions were investigated in detail. The results showed that a non-uniform distribution of local void fractions could be clearly revealed. The average void fractions obtained by GD and DP were found to approach the volumetric quality under higher pressure and higher mass velocity conditions owing to the homogeneous flow trends. This also occurred under a high volumetric quality condition for the homogeneous vapor flow and low volumetric quality condition for the homogeneous bubble flow. The results obtained from the different techniques were compared for the same experimental conditions. It was found that the void fractions measured by both OP and DP agreed well with that of GD, with relative deviations less than 10% and 15% respectively. Besides, an assessment of four void fraction correlations was carried out to predict the present area-average void fraction data from GD. The results showed that both the Bestion correlation and the Chexal and Lellouche correlation predict the present data well when measured void fraction greater than 0.3.

Analysis of a steam line break accident of a generic SMART-plant with a boron-free core using the coupled code TRACE/PARCS

The Karlsruhe Institute of Technology (KIT) is engaged in safety-relevant investigations of a generic SMART-plant using available public data. Within this work, a boron-free core that fits into the reactor pressure vessel (RPV) of the SMART-plant has been developed and optimized at the beginning-of-life condition from the safety perspective that fulfils general regulatory requirements. In this paper, the analysis of the behavior of a boron-free core integrated into a generic SMART-plant under a steam line break (SLB) accident using the coupled code TRACE/PARCS is presented and discussed.The SLB-accident in a conventional PWR is an overcooling accident that could lead to re-criticality and return-to-power after the reactor SCRAM. It is characterized by an asymmetrical cooling behavior due to the break of one of the steam lines while others are intact leading to a strong radial non-symmetrical core power distortion. Hence, to capture this physical phenomenon, 3D neutronics and thermal-hydraulics codes are applied for the analysis.The performed investigation showed that the boron-free core does not experience any re-criticality and a return-to-power in comparisons with a conventional PWR due to the low severity impact of the SLB-accident. That is due to the following factors: (a) the unique helical steam generator (SG) design concept, in which there are eight SGs integrated within the RPV, and almost negligible coolant inventory of the secondary side; and (b) the flow mixing header assembly located around the core that is developed to reinforce the coolant mixing within the downcomer. The core decay heat has been proven to be removed passively after the reactor trip thanks to the establishment of natural circulation in both the primary and secondary-side, and the excellent performance of the passive residual heat removal system (PRHRS). Generally, it can be stated that the boron-free core integrated into the generic SMART-plant under an SLB-accident does not threaten the safety limits.Moreover, the SLB-accident is analyzed assuming the failure of all safety systems including the PRHRS. Such a hypothetical accident revealed that the grace time to reach core uncovery is about two hours without any human intervention and safety system actuation.

Supercritical steam generator design and thermal analysis based on HTR-PM

The high temperature gas-cooled reactor pebble-bed module (HTR-PM) being built by the Institute of Nuclear and New Energy Technology of Tsinghua University in China is a typical Generation IV nuclear power system. The subcritical steam generators in HTR-PM can be replaced by the supercritical steam generators to work with the current reactors and a supercritical steam turbine to improve the thermal efficiency, which is being presented in the third technical phase of the HTR-PM program in China. The supercritical high temperature gas-cooled reactor pebble-bed module system scheme is proposed based on HTR-PM. A supercritical steam generator with helically coiled tubes is designed. Parameters of the geometry and the design conditions are listed in detail. The heat transfer performance of the supercritical steam generator is analyzed, and the heat flux, the temperature variation and the heat transfer coefficient along the tube are shown. The effects of the transverse and longitudinal pitches between the tubes in the supercritical steam generator are very important. The tube length, tube bundle height, and the deviation of the tube length in various layers are compared with various transverse and longitudinal pitches and the optimal relative transverse and longitudinal pitches are illustrated.

Analytical Design of Helical Coil Steam Generator for Hot Temperature Gas Reactor

Reaktor Daya Eksperimental (RDE) is a non-commercial Hot Temperature Gas Reactor (HTGR) which is currently being developed by National Nuclear Power Agency of Indonesia (BATAN). Integrated helical coil steam generator attached in the RDE is selected to generate 10 MW steam. In order to achieve the rated power, steam pressure of 62 bar, steam temperature of 798.15K and mass flowrate of 3.54 kg/s are required as the output parameters to be produced by the steam generator. A compact design of a once through counter-flow helical coil steam generator is proposed. Analytical design of the helical coil steam generator is conducted to check the heat transfer area required to achieve the rated power. A simple thermodynamic analysis combined with empirical heat transfer coefficient for convective and boiling process inside the steam generator at constant pressure was performed. It is shown that with inlet water temperature of 418.15K and helium gas inlet of 973.15K at 4.27 kg/s and 34.2 bar as the heating fluid produced output helium gas temperature of 511.42K. Based on comparison with available similar existing HTGR reactor, the calculation overestimate heat transfer surface area by around 20%. Hence, the required area to achieve the rated power for the proposed design is estimated around 60.2-75.3 m2.

Vertical steam generators for VVER NPPs

Steam generators for NPPs are the important large-sized metal consuming equipment of nuclear power installations. Efficiency of steam generator operation determines the overall service life of the whole nuclear facility. The main aim of the current study is to analyze advantages and shortcomings of horizontal and vertical types of steam generator design. This analysis is aimed at the development of recommendations for designing advanced steam generators for future Russian units of NPPs with VVER reactors of increased power. Design solutions and fifty-year experience of operation of 400 steam generators of horizontal type accepted in Russia and of vertical type applied by Westinghouse, Combustion Engineering, Siemens, Mitsubishi, Doosan were analyzed within the framework of the present study. Advantages and drawbacks of both types of equipment determining the development of conditions of the operating processes were also identified and systematized. Currently NPPs equipped with VVER are characterized with extended surface area of containment shells due to the application of four-loop design configuration and horizontal-type steam generators. It was established that steam generator equipment of horizontal type is characterized by such inherent disadvantages of design, technological and operational nature as the following: 1) small height and volume of the vapor space above the evaporation surface reducing separation capabilities and the capacity of the equipment as a whole; 2) impossibility of organizing separate single-phase pre-boiling section. Because of the above, horizontal steam generators with dimensions permissible for railroad transportation and, for VVER-1200 with reactor vessel diameter equal to 5 m, by water transport as well, have exhausted the possibilities for further significant increase of the per unit electric power. The demonstrated advantages of vertical-type steam generators were as follows: 1) absence of stagnant zones within the second cooling circuit, and, consequently, of hold-ups in them; 2) uniformity of heat absorption efficiency of the heating surface ensuring, as well, improved conditions for moisture separation; 3) high degree of moisture removal from steam-water mixture due to the combination of moisture separating elements of chevron and swirl-vane types; 4) increased temperature drop with parameters of generated steam elevated by 0.3 – 0.4 MPa. Conclusion was made on the advisability of introduction of steam generators with vertical-type layout in the Russian nuclear power generation. Practical tasks that need to be addressed in order to ensure introduction of vertical steam generators at NPPs with high-power VVER reactors were formulated.

Modeling analysis on solar steam generator employed in multi-effect distillation (MED) system

Recently the porous bilayer wood solar collectors have drawn increasing attention because of their potential application in solar desalination. In this paper, a thermodynamic model has been developed to analyze the performance of the wood solar collector. A modeling analysis has also been conducted to assess the performance and operating conditions of the multiple effect desalination (MED) system integrated with the porous wood solar collector. Specifically, the effects of operating parameters, such as the motive steam temperature, seawater flow rate, input solar energy and number of effects on the energy consumption for each ton of distilled water produced have been investigated in the MED desalination system combined with the bilayer wood solar steam generator. It is found that, under a given operating condition, there exists an optimum steam generation temperature of around 145°C in the wood solar collector, so that the specific power consumption in the MED system reaches a minimum value of 24.88 kWh/t. The average temperature difference is significantly affected by the solar heating capacity. With the solar capacity increasing from 50 kW to 230 kW, the average temperature difference increases from 1.88°C to 6.27°C. This parametric simulation study will help the design of efficient bilayer wood solar steam generator as well as the MED desalination system.

Experimental flow visualization study using particle image velocimetry in a helical coil steam generator with changing lateral pitch geometry

A full size section from a helical coil steam generator with five concentric coils with a helix-to-rod diameter ratio of 127 was constructed to study shell side fluid flow. The geometry has a constant transverse pitch ratio (a = 2.96), but changing lateral pitch ratio (b = 0.6–0.9) across the image planes of interest. Particle Image Velocimetry (PIV) with an index of refractive matching fluid technique was used to capture images of the flow behavior at these planes of interest. Experiments focused on visualizing cross-flow at the entrance and center regions of the test section at each plane of interest at Re = 8500. Analysis was conducted to study average vertical and transverse velocity components, vorticity and streamline fields for the shell side flow. Averaged results showed characteristic flow structures common of staggered cross-flow behavior within tube and shell heat exchangers. Select linear heights based on rod geometry show a consistent average velocity pattern across all three planes of interest regardless of lateral pitch ratio. Average velocity components do not suggest a correlation with lateral pitch ratio as previous studies with constant lateral pitch ratio designs have shown. The constantly changing geometry of this particular helical coil steam generator design affects the development and dependence of vortex formation and motion.

Improvement of steam generator tube failure propagation analysis code LEAP for evaluation of overheating rupture

ABSTRACT As a part of safety assessment or design of steam generators of sodium-cooled fast reactors, it is necessary to evaluate the water leak rate under sodium–water reaction accident. The computer code called LEAP-II calculating a design basis water leak rate during long-time event progress including self-wastage, target-wastage, wastage-type failure propagation, water leak detection, and water/steam blowdown was developed for the prototype fast reactor in the past studies. In this study, a numerical analysis method to predict occurrence of overheating tube rupture was constructed and integrated into this code to expand its application range. The newly constructed method consists of the elemental analysis models for temperature distribution formed by a reacting jet, water-side thermal hydraulics, heat transfer at the tube wall, temperature and stress of the tube, and failure of the tube. Applicability of the method was investigated through the numerical analysis of the experiment on water vapor discharging into liquid sodium pool under the actual condition of the steam generator. The numerical analysis demonstrated that the method could provide the appropriately conservative result on the overheating-rupture-type failure propagation.

Design of header and coil steam generators for concentrating solar power applications accounting for low-cycle fatigue requirements

Concentrating solar power plants are experiencing an increasing share in the renewable energy generation market. Among them, parabolic trough plants are the most commercially mature technology. These plants still experience many challenges, one of which is the cyclic daily start-up and shut-down procedures. These pose new challenges to industrially mature components like the steam generator system, as frequent load changes might decrease their lifetime considerably due to cyclic thermo-mechanical stress loads. In this context, the header and coil design is a promising configuration to minimize the stresses.This paper presents a method to design the header and coil heat exchangers of the steam generator, taking into account low-cycle fatigue requirements, by defining minimum allowable heating rates for the evaporator and superheater. Optimal designs were obtained by minimizing the total water pressure drops and purchase equipment costs. A comparison with a sizing routine without accounting for low-cycle fatigue constraints was also conducted.The model was validated against the component data of a 55 MWe power plant, with a maximum deviation on the total area estimation of +2.5%. The results suggest that including the heating rate constraints in the design routine substantially affects the optimal design configuration, with a 41% cost increase for a 1 bar pressure drop. The optimal design for maximizing the lifetime of the components uses tube outer diameters of 38 mm and 50 mm and a low number of tubes per layer (4–10) for the superheater.

Thermal design and optimization of a heat recovery steam generator in a combined-cycle power plant by applying a genetic algorithm

The heat recovery steam generator (HRSG) is an important and critical equipment of a combined cycle power plant that connects the gas turbine system to the steam cycle. The thermal design and the optimization of the HRSG are important for achieving safe operation, high efficiency and low product cost in a combined cycle power plant. This paper deals with the comprehensive optimization of the thermal design and minimization of the cost of an HRSG using a genetic algorithm (GA). Based on actual and existing designs of HRSG in most combined cycle power plants, a water tube HRSG including two superheaters, one evaporator and one economizer is considered in the optimization. A comprehensive program was developed in Visual Basic for this purpose.The optimization variables include the fin tube arrangement (transverse pitch, longitudinal pitch, number of rows in flue gas direction, number of tubes attached on circumference of headers, in line or staggered tube pattern), fin tube specification (tube diameter, fin height, fin thickness, fin type (solid or serrated), fin per meter, segment width of serrated fin) and also approach point, pinch point, water and steam velocity. On the other hand, the pressure at the exit of the gas turbine, fin tube metal temperature, amount of desuperheater spray water flow, steam pressure drops, minimum guaranteed thermal efficiency of HRSG, gap between fin tubes and the overall dimensions of HRSG (length, width, height) have been considered as main constraints in the optimization.The developed method selects and provides the best geometric parameter and arrangement of finned tubes based on minimum capital cost and relevant defined constraints. However, any other objective function such as minimum flue gas pressure drops, minimum heat transfer surface area, maximum rate of heat transfer, etc. could be used as objective functions in the program and easily optimized.In order to test this method, the results of the optimized design have been compared to an existing HRSG of a combined cycle power plant. All design parameters and the HRSG arrangement could be easily determined based on any optimization strategy and constraints.

Low Cycle Fatigue Behavior of Steam Generator Tubes under Axial Loading.

Compared with the fatigue properties of the material (Inconel Alloy 690), the real fatigue lives of tubes are more meaningful in the fatigue design and assessment of steam generator (SG) tube bundles. However, it is almost impossible to get a satisfactory result by conducting fatigue tests on the tube directly. A tube with a uniform and thin wall easily fails near the clamping ends under cyclic loading due to the stress concentration. In this research, a set-up for fatigue tests of real tubes is proposed to overcome the stress concentration. With the set-up, low cycle fatigue tests were conducted in accordance with an existing fatigue design curve for Alloy 690. Strain control mode was applied with fully reversed push–pull loading under different strain amplitudes (0.15%, 0.2%, 0.3%, and 0.4%). A favourable result was obtained, and the low cycle fatigue behavior was investigated. The results showed that the fatigue life tested by the real tube was below the strain–life curve of Alloy 690 which was fitted by conventional solid specimens. A cyclic hardening behavior was found by the cyclic stress–strain curve when compared with the monotonic stress–strain curve.

Maximizing the power block efficiency of solar tower plants: Dual-pressure level steam generator

Solar tower plants (STP) are one of the most promising renewable technologies to substitute conventional power plants. To further increase its penetration in electricity markets, it is necessary to maximize its efficiency. This work addresses the challenge of increasing the inlet pressure up to 165 bar at the high pressure turbine (HPT) to improve the power block efficiency. Nowadays, these plants operate Rankine cycle with reheating using steam at 126 bar at the inlet of the HPT. This pressure is limited in current STP by the working temperatures of the molten salt, which range from 285 °C and 565 °C, the pinch point temperature difference in the evaporator and the current steam generator (SG) layout. A new steam generator design with dual-pressure level evaporation is proposed. The heat exchangers that form the SG are designed thermomechanically and the power cycle performance is analyzed using the first and second laws of thermodynamics. The results show that the novel dual-pressure level SG layout increases the power block efficiency from 44.14% to 44.64%. Assuming a market pricing scenario of two-tier tariff and a power purchase agreement price of 16.3c€/kWhe, the new SG layout yields an extra economic benefit of 623 k€ per year due to an increase of energy produced of 5.71 GWhe/year.

Thermal Design of the High-Temperature Steam Generator for the Nuclear Power Plant GT-MGR with Helical (Coil-in-Box) Pipe Bundles

Thermal Design of the High-Temperature Steam Generator for the Nuclear Power Plant GT-MGR with Helical (Coil-in-Box) Pipe Bundles

Numerical prediction of a flashing flow of saturated water at high pressure

Abstract Transient fluid velocity and pressure fields in a pressurized water reactor (PWR) steam generator (SG) secondary side during the blowdown period of a feedwater line break (FWLB) accident were numerically simulated employing the saturated water flashing model. This model is based on the assumption that compressed water in the SG is saturated at the beginning and decompresses into the two-phase region where saturated vapor forms, creating a mixture of steam bubbles in water by bulk boiling. The numerical calculations were performed for two cases of which the outflow boundary conditions are different from each other; one is specified as the direct blowdown discharge to the atmosphere and the other is specified as the blowdown discharge to an extended calculation domain with atmospheric pressure on its boundary. The present simulation results obtained using the two different outflow boundary conditions were discussed through a comparison with the predictions using a simple non-flashing model neglecting the effects of phase change. In addition, the applicability of each of the non-flashing water discharge and saturated water flashing models for the confirmatory assessments of new SG designs was examined.

Conceptual design of a bayonet tube steam generator with heat transfer enhancement using a helical coiled downcomer

Bayonet tube heat exchangers are a proposed solution for advanced nuclear reactor steam generators and heat removal systems, in particular in liquid metal cooled reactors. However, the performance of this heat exchangers is not very high due to the limited heat transfer surface; thus, many long bayonet heat exchangers are needed to remove significant power values. In the present paper, the use of a helical coiled downcomer inside the bayonet is proposed, instead of the traditional straight one, as a passive heat transfer enhancement method. The helical coiled downcomer acts on the annulus as a swirl flow device; for the evaluation of heat transfer and pressure drop it can be considered as a helical coiled wire or a helical coiled tape. The global thermal resistance between the heat source (external flow of primary fluid) and the heat sink (secondary fluid) is reduced, while the pressure drop is not excessively increased. The component behavior has been simulated using the thermal-hydraulic code RELAP5-3D, with lead and water respectively as primary and secondary coolants. A one-dimensional model has been adopted, modifying the heat transfer coefficient and the frictional pressure losses to properly simulate the helical coil effect with two different methods: the use of a heat transfer coefficient multiplication factor and the adoption of the swirl tube heat transfer mode. The component has been operated as a steam generator considering the parameters and the boundary conditions of ALFRED reactor proposed design. Two different downcomer tube dimensions have been considered (6–8 mm and 8–10 mm ID-OD) with five helical pitches (5, 7.5, 10, 15 and 20 cm), which have been selected taking into account the manufacturing constraints. The results show a significant improvement of the bayonet heat exchanger performance. The removed power increases more than 8% and the steam superheating increases around 23 K for the best configuration.

Optimal working fluid charge and degradation thresholds for a closed-circuit intermediate natural circulation heat transport loop

The study derives the optimal working fluid charge of a closed-circuit intermediate natural circulation heat transport loop based on the passive residual heat removal system (PRHRS) of SMART (System-integrated Modular Advanced ReacTor). The study finds that efficient natural circulation flow and heat transfer regimes can be maintained in a narrow band of working fluid charge mass owing to the small secondary side volume of the once-through helical-coil steam generator design of SMART. Overcharging condition is a severe passive system degradation mode that quickly leads to system failure. Overcharging is characterized by sustained two-phase flow instabilities, mainly liquid slugging in the hot leg of the PRHRS, and an increased operating pressure of the PRHRS coupled with a compensating increase in the operating pressure of the reactor coolant system (RCS). Increased pressures maintain the required temperature differential across the steam generator tubes and heat sink heat exchanger to achieve a prescribed quasi-steady state heat removal rate capacity. Undercharging condition is a benign degradation mode relative to overcharging and is characterized by a slow upward drift of the RCS pressure and temperature of the liquid water entering the primary side of the steam generator. Only the undercharged case of a completely dry steam generator secondary side leads to PRHRS failure-to-start via vapor lock.

A new dimensionless thermal hydraulics parameter for the heat exchangers

Thermal and hydraulic considerations are crucial for the design of an efficient heat exchanger (HX) which allow boiling. Steam generator is one of the biggest and most expensive components of most nuclear power plants. Therefore, design perfections of heat exchangers can lead to improved design of steam generators. Two-phase heat transfer coefficient (HTC) strongly depends on the prevailing flow regime for a given surface pattern as well. The enhancement in HTC is associated with an increase in frictional pressure drop. A single parameter for characterization of heat transfer surfaces based on thermal and hydraulic consideration doesn’t exist. A new dimensionless number for thermal and hydraulics characterization (DITCH) of the heat exchanger surface has been derived for evaluation of heat transfer surfaces. It captures the hydraulic behavior of coolant due to phase change and wall friction. Its value characterizes different flow regimes of two phase flow. Analysis of a selected data reveals that DITCH increases with quality ‘x’ and the mass flow rate ‘ṁ’. A steep increase up to x = 0.4 is followed by lesser slope. It is also observed that bubbly flow regime has a least value of this parameter whereas slug flow regime has a value less than 1. This ratio is nearly equal to one or higher for the annular flow region. It is therefore concluded that in annular flow regime, compared to the other flow regimes, heat transfer coefficient is getting significant and pressure drop is getting insignificant.

Numerical investigation of turbulent flow in an annular sector channel with staggered semi-circular ribs using large eddy simulation

Computational fluid dynamics (CFD) simulations are conducted to explore flow inside a channel with staggered semi-circular ribs attached to both the inner and outer annular walls. Large eddy simulation (LES) with the wall-adapting local eddy-viscosity (WALE) sub-grid scale (SGS) model is used to simulate turbulent flow in the computational domain. The flow channel is modeled on the test section of a simplified helical coil steam generator (HCSG) design. Owing to their compactness and higher heat transfer coefficient in comparison to straight tube steam generators, HCSGs are employed in recent designs of advanced light water reactors in the nuclear industry. A Reynolds number of 13,900 is considered based on the mean inlet velocity, channel height, and water properties. Results from LES are compared with both an unsteady Reynolds-Averaged Navier-Stokes (URANS) simulation and experimental data acquired from the particle image velocimetry (PIV) method. Flow features observed inside the channel are presented by visualizing contours of velocities, vorticity, and Reynolds stresses on a monitoring plane. Time-averaged profiles of velocities and Reynolds stresses at five monitoring lines are also presented to validate the flow field resolved by LES with PIV. Additionally, vortical structures of the flow are captured using the Q-criterion. This study aims at providing flow characteristics of a specifically designed flow channel and acquiring the scalability of the numerical experiment.