Critical Heat Flux Margin Research Articles

Enhancement of the critical heat flux for downward-facing saturated pool boiling on the reticular hollow shell structure surfaces

In-vessel retention (IVR) technology was currently an effective strategy for retaining molten materials in the reactor pressure vessel (RPV) when the nuclear power plants occurred severe accident, further ensuring the integrity of RPV. The effectiveness of IVR depended on the critical heat flux (CHF) of external reactor vessel cooling (ERVC). A novel type of reticular hollow shell structure (RHS) was designed to meet this need. Five RHS surfaces were designed and the pool boiling heat transfer performance were investigated in saturated deionized water, the experimental inclination angles were 5°, 30°, 45°, 60° and 90°. The boiling heat transfer coefficient (BHTC) and CHF were measured by steady-state heating and transient quenching methods. The experiment results showed that the CHF increased with the increase of inclination angles, and the CHF of all RHS surfaces were significantly enhanced compared with the plain surface. The maximum CHF value and the maximum CHF enhancement exhibited on surface 4, with a maximum CHF of 2992.2 kW/m2.When the inclination angle was 30°, the maximum CHF could reach 2338.6 kW/m2, which was 160.3% enhancement over the plain surface. Surface 1 and surface 3 had better BHTCs, and the maximum values were 178.3 kW/(m2·K) and 126.8 kW/(m2·K), respectively, which were much better than other structured surfaces. The RHS could form a gas–liquid conversion system, which was composed of cavities and alternating channels to increase the surface nucleation sitesand accelerate the circulation of cooling water, thus obtained the high CHF and high BHTC characteristics. RHS provided a large enough CHF margin to ensure that the RPV was continuously cooled and its integrity was maintained. Finally, the behavior of bubbles generation, coalescence, sliding and detachment in reticular hollow shell structure were systematically analyzed.

Critical heat flux analysis of divertor cooling flow channel in fusion reactor with CFD method

Situated at the bottom of the vacuum vessel, the divertor extracts heat and ash produced by the fusion reaction, minimizes plasma contamination, and protects the surrounding walls from thermal and neutronic loads. The vertical targets of divertor are designed to be able for up to 20 MW/m2 high heat flux. It is a great ordeal for both the material performance and the cooling ability. Critical heat flux margin is very crucial during the design of divertor. The ANSYS FLUENT is used in this paper to predict the critical heat flux on a monoblock structure with a twisted tape inside the tube. Numerical results are validated with the corresponding sets of experimental results. In this paper, CFD method used to predict critical heat flux of divertor cooling channel was first introduced. On the other hand, influence of inlet subcooling on critical heat flux is studied in detail. The inlet subcooling affect the critical heat flux much complicated for the single-side heated and swirl flow channel. Whether the influencing trend or the locations of critical heat flux occurrence are different under different inlet sub-cooling. The derivations between the simulation and experimental results were no more than 32%. This study proves the CFD tools can provide efficient help on the understanding of the critical heat flux phenomenon of complex construction.

Hydraulic assessment of cooling circuit for CFETR W/Cu divertor

Divertor, whose main functions are to sustain extremely high heat load and to exhaust impurity, is a key component in fusion reactor. CFETR divertor is currently under engineering design and technical validation phase. In this study, a cooling circuit design is proposed to obtain sufficient heat removal capability for CFETR divertor. The cooling circuit consists of two main modules which are outer target module and inner target module. Considering the complexity of the parallel cooling circuit concept, hydraulic characteristics is assessed and optimized for enough heat transfer capability and critical heat flux margin. Assessment of the hydraulic characteristics for the cooling circuit is performed by investigating flow velocity and local pressure on channels’ cross-sections. Uniform flow velocity of 10 m/s and acceptable critical heat flux margin higher than 1.4 in PFU coolants channels are obtained after optimization meeting the design requirements of the divertor.

Hydraulic assessment of an upgraded pipework arrangement for the DEMO divertor plasma facing components cooling circuit

In the context of the Work Package DIVertor (WPDIV) of the EUROfusion action, a research campaign has been carried out by University of Palermo in cooperation with ENEA to assess the thermal-hydraulic performances of the DEMO divertor cooling system, concentrating the attention on its 2019 Plasma Facing Components (PFCs) configuration, relevant to DEMO baseline 2017. The research activity has been performed following a theoretical-numerical technique based on the finite volume method and adopting the well-known ANSYS CFX CFD code. The PFCs cooling circuit thermal-hydraulic performances under nominal steady-state conditions, assessed mainly in terms of coolant total pressure drop, coolant axial flow speed and margin against Critical Heat Flux (CHF) distributions among the plasma-facing channels, have been evaluated with a CFD analysis to check their compliance with the corresponding limits. Results have highlighted serious critical issues, such as an intolerable total pressure drop (significantly higher than 1.4 MPa) as well as an insufficient margin against CHF onset (lower than 1.4) within all the PFU channels. Therefore, an optimization study has been performed to investigate the potential improvements of the PFCs cooling circuit thermal-hydraulic performances due to proper changes of its geometric configuration, focussing the attention on the inlet manifold branch. The study has allowed selecting the most effective cooling circuit configuration, that fulfils the maximum pressure drop requirement (Δp<1.4 MPa) while raising the minimum CHF margin within the PFU channels. Moreover, it would allow reducing the ex-vessel total pressure drop, while decreasing the average coolant flow velocity up to values lower than 10 m/s. Models, loads and boundary conditions assumed for the analyses are herewith reported and critically discussed, together with the main results obtained.

Assessment of critical heat flux margins on tungsten monoblocks of the ITER divertor vertical targets

Recent Critical Heat Flux (CHF) experimental results in the range of 40 MW/m2 obtained on small scale tungsten (W) monoblock mock-ups with reduced minimum W thickness above the coolant tube compared to the nominal 6 mm planned for ITER are presented. The CHF margin during ITER operations is further assessed, taking into account design and manufacturing aspects, demonstrating that an absorbed top surface heat flux of 20 MW/m2 should not be exceeded during ITER operations if the factor 1.4 margin to CHF, which must be respected for ITER monoblocks, is to be preserved.

Efficacy of endoscopic transpapillary drainage for patients with pancreatic duct disruption

In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to effectively remove big vapor mushrooms on the heating surface.

An experimental study of hypervapotron structure in external reactor vessel cooling

In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30mm by 61mm rectangular flow channel with a 200mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to effectively remove big vapor mushrooms on the heating surface.

Investigation of two-phase flow instabilities under advanced PWR conditions

The characteristics of two-phase flow instability under advanced pressurized water reactor (PWR) conditions are explored in this study. Density wave oscillations in vertical parallel tubes are experimentally investigated by employing three different test sections to observe the parametric effects of the axial power shape, unheated riser, and channel inlet throttling on the onset of flow instability (OFI). A critical heat flux (CHF) accompanied by flow oscillation is indicated under certain test conditions with high pressure and/or low inlet subcoolings. A frequency-domain linear stability analysis code, called ALFS_F, was developed on the basis of one-dimensional two-phase mixture equations and a reactivity feedback model with point neutron kinetics and lumped parameter heat conduction equations. The mass flux at OFI conditions and the oscillation periods are calculated by the ALFS_F to compare with the experimental data. The thermal margin characteristics are illustrated with a comparative analysis of OFI and CHF margins under the core conditions of SMART, which is an integral-type advanced PWR being developed at KAERI.

Optimized mass flow rate distribution analysis for cooling the ITER Blanket System

This paper presents the rationale to the optimization of water distribution in ITER blanket modules, meeting both Blanket System requirements and interface compliance requirements.The key challenging constraints include to: be compatible with the overall water allocation (3140kg/s for 440 wall mounted BMs); meet the critical heat flux margin of 1.4 in the plasma facing units; meet a maximum temperature increase of 70°C at the outlet of each single BM; and ensure that water velocity is less than 7m/s in all manifolds, and that the pressure drops of all BMs can be equilibrated. The methodology and the successful result are presented.

Russian development of enhanced heat flux technologies for ITER first wall

Recently the ITER first wall (FW) design has been significantly upgraded to improve resistance to electromagnetic loads, to facilitate FW panel replacement and to increase FW ability to withstand higher (up to 5MW/m2) surface heat loads. The latter has made it necessary to re-employ technologies previously developed for the now-abandoned port limiters. These solutions are related to the cooling channel with CuCrZr–SS bimetallic walls and hypervapotron type cooling regime, optimization of Be-tiles dimensions and Be to CuCrZr joining technique. A number of representative mockups were tested at high heat flux (HHF) at the Tsefey electron-beam facility to verify the thermo-hydraulic characteristics of the reference cooling channel design at moderate water flow velocities (V=1–3m/s, P=2–3MPa, T=110–170°C). The heat flux was gradually varied in the range of 1–10MW/m2 until the critical heat flux was registered. The mockups of hypervapotron structure demonstrated the required cooling efficiency and critical heat flux margin (1.4) at a water velocity of ≥2m/s. Dimensions of Be armor tiles strongly affect the thermo-mechanical stresses both in the CuCrZr cooling wall and at the Be–CuCrZr interface. Results of tile dimensions optimization (variable in the range 12mm×12mm×6 to 50mm×50mm×8mm) obtained by the HHF (variable in the range of 3–8MW/m2) experiments are presented and compared with analysis. It is shown that optimization of the tile geometry and joining technology provides the required cyclic fatigue lifetime of the reference FW design.

Prediction of Critical Heat Flux in Water-Cooled Plasma Facing Components Using Computational Fluid Dynamics

Several commercial computational fluid dynamics (CFD) codes now have the capability to analyze Eulerian two-phase flow using the Rohsenow nucleate boiling model. Analysis of boiling due to one-sided heating in plasma facing components (pfcs) is now receiving attention during the design of water-cooled first wall panels for ITER that may encounter heat fluxes as high as 5 MW/m2. Empirical thermalhydraulic design correlations developed for long fission reactor channels are not reliable when applied to pfcs because fully developed flow conditions seldom exist. Star-CCM+ is one of the commercial CFD codes that can model two-phase flows. Like others, it implements the RPI model for nucleate boiling, but it also seamlessly transitions to a volume-of-fluid model for film boiling. By benchmarking the results of our 3d models against recent experiments on critical heat flux for both smooth rectangular channels and hypervapotrons, we determined the six unique input parameters that accurately characterize the boiling physics for ITER flow conditions under a wide range of absorbed heat flux. We can now exploit this capability to predict the onset of critical heat flux in these components. In addition, the results clearly illustrate the production and transport of vapor and its effect on heat transfer in pfcs from nucleate boiling through transition to film boiling.This article describes the boiling physics implemented in CCM+ and compares the computational results to the benchmark experiments carried out independently in the United States and Russia. Temperature distributions agreed to within 10 °C for a wide range of heat fluxes from 3 MW/m2 to 10 MW/m2 and flow velocities from 1 m/s to 10 m/s in these devices. Although the analysis is incapable of capturing the stochastic nature of critical heat flux (i.e., time and location may depend on a local materials defect or turbulence phenomenon), it is highly reliable in determining the heat flux where boiling instabilities begin to dominate. Beyond this threshold, higher heat fluxes lead to the boiling crisis and eventual burnout. This predictive capability is essential in determining the critical heat flux margin for the design of complex 3d components.

Thermal-Fluid Characterizations of ZnO and SiC Nanofluids for Advanced Nuclear Power Plants

Nanofluids including metal or metal-oxide nanoparticles have shown improved thermal performance compared to water. Introducing zinc into a water-containing component of the primary system of a boiling water reactor or pressurized water reactor nuclear power plant is known to prevent the deposition of radioactive cobalt and to reduce the possible susceptibility of component materials to corrosion. Also, it is well known that silicon carbide (SiC) is a promising material for advanced reactors. Therefore, preparations of zinc oxide (ZnO) and SiC nanofluids have been used in diverse methods to reduce inconsistency of nanofluid performance. Thermal-fluid characterizations were carried out under the control of the preparation methods. In addition, indirect checks by contact angle measurements of depositions have shown that ZnO and SiC nanofluids as a self-recovering/healing coating solution can contribute to advanced nuclear safety systems in terms of the critical heat flux margin. The results show that more efforts are needed to develop the nanofluid formulation to tailor the desired properties.

COBRA-TF Analysis of the High Flux Reactor Hot Channel for a Postulated Large-Break Loss-of-Coolant Accident

In this study, calculations were performed to simulate a postulated large-break loss-of-coolant accident for the High Flux Reactor (HFR) cooling system using the COBRA-TF computer code. COBRA-TF has been chosen for this analysis since it has suitable and validated two-phase flow models and critical heat flux (CHF) correlations for channels having small hydraulic diameters. Calculations have been performed to determine the CHF margins for the HFR. Six types of calculations were performed to provide a range of CHF margins. All COBRA-TF calculations indicate that margin does exist to the CHF limit for the small-hydraulic-diameter highest-power HFR channel. The range of margin is 2.1 to 1.3 times the nominal power of the highest power channel, depending on the boundary conditions and CHF correlation used. The range of margin identified in the HFR analysis is consistent with the margin values used in commercial nuclear power plants.

A Methodology for Selecting High Thermal-Hydraulic Performance Fuel Configurations for Tightly Packed Epithermal Core Designs

Cylindrical fuel pins with wires are the design of choice for tightly packed fuel arrays. However, it is important to investigate novel fuel configurations in order to increase the thermal margins. Hence, new fuel designs have been studied for the epithermal option of the light water-cooled IRIS core. These designs are also of potential use in other tightly packed, epithermal advanced core designs.First, design equations have been used to determine number, height, and size of the principal features (clad, gap, fuel cross-sectional area) of the novel fuel configurations under investigation. Then, performance indices have been introduced to relate fuel geometrical characteristics to selected thermal-hydraulic parameters, such as pressure drop, critical heat flux (CHF), fuel centerline temperature, and clad surface temperature and stress distribution. Finally, variously shaped fuel configurations, including cylindrical, triangular, square, and hexagonal, have been ranked according to the performance indicators.The hexagonal fuel pins, both twisted and straight, proved to be good solutions for the epithermal tight core of the light water-cooled IRIS reactor, with performances comparable to those of the cylindrical fuel with wires. In particular, for water-to-fuel ratios ~0.33, the twisted hexagonal shape is the preferable design with a reduction of the total pressure drop by 16% and an increase of the CHF margin by 200%, compared to the traditional cylindrical pins with grids. Furthermore, the straight hexagonal shape allows flatter subchannel velocity profiles, wall shear stress, and wall temperature distributions. However, geometric constraints unfortunately do not allow application of the twisted hexagonal shape for smaller water-to-fuel ratios, which is a design regime of more favorable epithermal neutronics performance. In this regime, the cylindrical pins with wires are the solution of choice.

Critical heat flux analysis and R&D for the design of the ITER divertor

The vertical target and dump target of the ITER divertor have to be designed for high heat fluxes (up to 20 MW/m 2 over ≈10 s). Accommodation of such high heat fluxes gives rise to several issues, including the critical heat flux (CHF) margin which is a key requirement influencing the choice of cooling channel geometry and coolant conditions. An R&D programme was evolved to address the overall CHF issue and to help focus the design. It involved participation of the four ITER home teams and has been very successful in substantially expanding the CHF data base for one-sided heating and in providing more accurate experimental measurements of pressure drop (and derived correlations) for these geometries. This paper describes the major R&D results and the design analysis performed in converging on a choice of reference configuration and parameters which resulted in a CHF margin of ≈1.4 or more for all divertor components.

Consideration of critical heat flux margin prediction by subcooled or low quality critical heat flux correlations

The accurate prediction of the critical heat flux (CHF) margin which is a key design parameter in a variety of cooling and heating systems is of high importance. These margins are, for the low quality region, typically expressed in terms of critical heat flux ratios using the direct substitution method. Using a simple example of a heated tube, it is shown that CHF correlations of a certain type often used to predict CHF margins, expressed in this manner, may yield different results, strongly dependent on the correlation in use. It is argued that the application of the heat balance method to such correlations, which leads to expressing the CHF margins in terms of the critical power ratio, may be more appropriate.

CHF Experiments under Steady-State and Transient Conditions for Tight Lattice Core with Non-Uniform Axial Power Distribution

Abstract Steady-state and transient critical heat flux (CHF) experiments were performed using triangular pitched 7-rod assemblies with non-uniform axial power distributions under the maximum pressure of 15.5 MPa. The onset of steady-state CHF was predicted within the uncertainty of 10% with the KfK correlation using the local flow conditions calculated by the subchannel analysis code COBRA-IV-T. On the other hand, various mechanistic CHF models did not agree with the steady-state CHF data. The transient CHFs under the conditions of flow reduction, power increase or flow and power simultaneous variation were predicted with the quasi-steady-state method within approximately the same uncertainty as the steady-state CHF experiments. The predictive capability did not depend on the transient speed within 30%/s of the flow reduction rate and within 120%/s of the power increase rate. It was also revealed that there exists large CHF margins under the thermal hydraulic conditions simulating the locked rotor accident and the control rod cluster ejection accident of the double-flat-core type high conversion pressurized water reactor (HCPWR).

CHF Experiments under Steady-State and Transient Conditions for Tight Lattice Core with Non-Uniform Axial Power Distribution.

Abstract Steady-state and transient critical heat flux (CHF) experiments were performed using triangular pitched 7-rod assemblies with non-uniform axial power distributions under the maximum pressure of 15.5 MPa. The onset of steady-state CHF was predicted within the uncertainty of 10% with the KfK correlation using the local flow conditions calculated by the subchannel analysis code COBRA-IV-T. On the other hand, various mechanistic CHF models did not agree with the steady-state CHF data. The transient CHFs under the conditions of flow reduction, power increase or flow and power simultaneous variation were predicted with the quasi-steady-state method within approximately the same uncertainty as the steady-state CHF experiments. The predictive capability did not depend on the transient speed within 30%/s of the flow reduction rate and within 120%/s of the power increase rate. It was also revealed that there exists large CHF margins under the thermal hydraulic conditions simulating the locked rotor accident and the control rod cluster ejection accident of the double-flat-core type high conversion pressurized water reactor (HCPWR).

Experimental and Analytical Evaluations on Critical Heat Flux Under One Sided Heating Condition for Fusion Applications

AbstractIt is one of key issues to predict a critical heat flux (CHF) with the highly subcooled flow boiling and the CHF margin for the promising design of plasma facing components for the next generation fusion machines such as an International Thermonuclear Experimental Reactor and a Fusion Experimental Reactor. In particular, divertor plate is subjected to severe heat loads under a condition for one side heating. Because of no correlations predicting CHF for the highly subcooled flow region with heating on one side of the divertor plate, experimental data which were obtained under one sided heating condition have been evaluated by various existing CHF correlations which are based on a condition for uniform circumferential heating. As results, experimental CHF data of the straight tube are relatively good agreement with some correlations within an accuracy of −20 to +10 %, but no correlations are available in the CHF prediction of the externally-finned tube. Further experiments are necessary to evaluate...

General Features of Advanced Pressurized Water Reactors with Improved Fuel Utilization

The main feature of an advanced pressurized water reactor (APWR) with improved fuel utilization is a plutonium-enriched tight lattice core integrated in an unchanged primary system of a common pres...