Design considerations for Gardon heat flux sensors
Centerline loss, transient sink temperature and thermal conductivity design constantan considerations for thin foil copper-Gardon heat flux sensors
Ash, R. L.↗
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Results for “Transient heat flux”
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Results of the CHF-SERTTA In-Pile Transient Boiling Experiments at TREAT
To advance the current understanding of heat transfer between nuclear fuel rod cladding and the surrounding coolant under transient scenarios involving critical heat flux (CHF), the CHF-Static Environment Rodlet Transient Test Apparatus (CHFSERTTA), an in-pile experiment, was developed for and irradiated in the Transient Reactor Test (TREAT) facility at Idaho National Laboratory. The CHF-SERTTA experiment consists of a hollow, borated stainless steel rod submerged in a static water pool. The rod, heated via the (n, a) reaction in boron-10, was designed for investigating in-pile transient heat transfer between the rod surface and the coolant, avoiding the complexity of including the heat transfer between the fuel, gas gap, and cladding in typical light-water reactor (LWR) fuel rods. The instrumentation package for CHF-SERTTA focused on capturing those data that best describe the heat transfer behaviour between the rod and coolant. This paper outlines the CHFSERTTA experiment design and presents the initial results of the experiment campaign.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
A non-fueled nuclear-heated rod for in-pile transient boiling studies
Separate-effects boiling experiments have recently been conducted in the Transient Reactor Test Facility at Idaho National Laboratory to investigate transient heating and irradiation effects on cladding-to-coolant heat transfer. Specifically, transient critical heat flux (CHF) remains an important area of research, and better understanding of this phenomenon has potential for improving predictive models related to operational and safety limits. Consequently, this knowledge is expected to improve efficiency of light-water reactor operations. A novel borated nuclear-heated rodlet (BNHR) was designed to enable observation of transient cladding-to-coolant heat transfer phenomena. The final BNHR design takes a surrogate approach, wherein nuclear heating is induced by 10 B(n, α) reactions rather than derived from fissions in a fueled specimen. The structure of the BNHR consists of a hollowed out borated (B nat ~ 2.05 wt %) stainless steel tube with an hourglass-shaped outer surface, capped at both ends with non-borated stainless steel. This geometry allows for inner-rodlet instrumentation and generation of the highest nuclear heating rates near the center of the rodlet to ensure onset of boiling near instrumentation for real-time observation. A novel approach to measuring the nuclear energy deposition rate in the BNHR separate and apart from the influence of the coolant, termed the n-a thermometer, is also detailed in this paper. This device has demonstrated excellent repeatability, and measurements indicate predictive modeling results for energy deposition in the BNHR rod agree within a 10% margin of the experiment measurements. In conclusion, these results give confidence that the BNHR design has successfully met experiment objectives.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Design of separate-effects In-Pile transient boiling experiments at the $\mathrm{TREAT}$ Facility
The cladding-to-coolant heat transfer during rapid power excursions, such as reactivity-initiated accidents, remains a crucial area of uncertainty in nuclear reactor safety. This uncertainty impacts the ability to accurately predict fuel performance behavior for these conditions. Improving our understanding of transient cladding-to-coolant heat transfer will enhance our ability to model design basis accidents and could increase design and improve safety margins for the current commercial fleet and advanced reactors. To address these issues, the Critical Heat Flux Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA) experiment uses a novel approach with a borated stainless-steel heater rodlet that can replicate cladding temperatures experienced during a design-basis reactivity-initiated accident. The rodlet and experiment is instrumented to provide temperature and thermal-hydraulic conditions throughout the transient. The design of the rodlet allows for separate effect testing that eliminates complexities with a fuel and cladding specimen. We report the experiments have provided data that will be used to improve our understanding of boiling behavior, specifically critical heat flux, under transient conditions.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Active Divertor Heat Flux Control using Impurity Powder Dropper
Divertor plasma-facing components (PFCs) in a tokamak are typically designed to withstand average steady-state heat loads of about 5–10 MW/m 2 , a limit that applies to both solid and liquid lithium (LL) PFCs. Exceeding these design values can result in surface damage to tungsten PFCs or excessive lithium (Li) evaporation in liquid lithium divertor (LLD) PFCs. Since exceeding the divertor heat load limits has serious consequences, it is therefore prudent to develop a tool to reduce the divertor heat load and bring the heat load to within the design limit without affecting the plasma performance. Active low Z impurity injection such as Li has been suggested as a potential solution to mitigate excess heat flux as suggested previously, given that non-coronal radiation can be quite large ~ 20–30 MJ per mole of injected Li. Li is considered desirable for reducing the edge neutral recycling helping to improve plasma energy confinement. In this paper, we model the Impurity Power Dropper (IPD) to investigate its potential of divertor heat flux control. The IPD is typically located at the top of the tokamak device and uses a vertical drift tube of a few meters. In the 2 m drift tube case, the IPD powder is accelerated to ~ 6 m/sec before reaching the plasma with the upper divertor configuration, matching the condition for the in-board side pellet injection case. By modeling the IPD geometry we determined the IPD powder deposition profile, and thus the non-coronal radiation and ionization profiles in time as well. From the enhanced radiation power loss, it is therefore possible to reduce the divertor heat load using the divertor simulation code. In conclusion, the IPD divertor heat flux control can be tested in the facilities with IPD including ST-40, DIII-D, EAST, WEST and NSTX-U.
Active lithium injection↗
Recrystallization suppression through dispersion-strengthening of tungsten
Tungsten is the material of choice for the divertor region of future nuclear fusion reactors, an environment that will expose plasma-facing components (e.g. divertor, etc...) to high temperatures and transient high heat flux events. Under these conditions, recrystallization and grain growth of tungsten can occur, leading to undesirable microstructural and mechanical property changes. Therefore, there is a need to raise the recrystallization temperature of tungsten and limit the kinetics of the recrystallization and grain growth processes. In this work, we examine the effect of different types (TiC vs. TaC vs. ZrC) and different concentrations (1.1 vs. 5 vs. 10 wt.%) of dispersed second phase particles in a tungsten matrix on the high temperature performance. Additionally, the addition of second-phase particles effectively increases the temperature of and time for recrystallization and slow grain growth; however, the addition of a high weight fraction of particles alters the surface chemistry, which may impact subsequent plasma-surface interactions. These results show that the addition of small concentrations of dispersed particles can be effectively employed in tungsten to raise the upper operating temperature limit for tungsten in a fusion reactor.
36 MATERIALS SCIENCE↗
Effect of detachment on Magnum-PSI ELM-like pulses: direct observations and qualitative results
Abstract Conditions similar to those at the end of the divertor leg in a tokamak were replicated in the linear plasma machine Magnum-PSI. The neutral pressure in the target chamber is then increased to cause the target to transition from an attached to a detached state. Superimposed to this steady state regime, edge localised mode (ELM)-like pulses are reproduced, resulting in a sudden increase in plasma temperature and density, such that the heat flux increases transiently by half an order of magnitude. Visible light emission, target thermography, and Thomson scattering are used to demonstrate that the higher the neutral pressure the more energy is removed from the ELM-like pulse in the volume. If the neutral pressure is sufficiently high, the ELM-like pulse can be prevented from affecting the target and the plasma energy is fully dissipated in the volume instead (ID 4 in table 1). The visible light images allow the division of the ELM-plasma interaction process of ELM energy dissipation into 3 ‘stages’ ranging from no dissipation to full dissipation (the target plasma is detached). In the second publication related to this study, spectroscopic data is analysed with a Bayesian approach, to acquire insights into the significance of molecular processes in dissipating the plasma energy and particles.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Linear simulation of the stationary eddies in a general circulation model. I - The no-mountain model
A linear primitive-equation stationary-wave model is developed to describe Northern Hemisphere winter stationary eddies on a sphere, and simulation results are compared with the predictions of a flat-lower-boundary GCM in extensive graphs and maps. The linear model is shown to reproduce the general behavior of the GCM well, with 10-30-percent underestimation of amplitudes in the Northern Hemisphere extratropical upper troposphere. The response to global heating plus transient eddy flux convergences is then decomposed into responses to total tropical forcing (heating plus transients) and total extratropical forcing (upper and lower tropospheric transients and latent, sensible, and radiative heating). The quality of the simulation is found to deteriorate rapidly if transient forcing is omitted, although somewhat better results are obtained by substituting thermal damping for transient forcing. The importance of low-level-transient parameterization for theories of time-mean low-level flow is emphasized.
Nigam, Sumant↗
An investigation into the causes of stratospheric ozone loss in the southern Australasian region
Measurements of total ozone at Macquarie Island (55 deg S, 159 deg E) reveal statistically significant reductions of approximately twelve percent during July to September when comparing the mean levels for 1987-90 with those in the seventies. In order to investigate the possibility that these ozone changes may not be a result of dynamic variability of the stratosphere, a simple linear model of ozone was created from statistical analysis of tropopause height and isentropic transient eddy heat flux, which were assumed representative of the dominant dynamic influences. Comparison of measured and modeled ozone indicates that the recent downward trend in ozone at Macquarie Island is not related to stratospheric dynamic variability and therefore suggests another mechanism, possibly changes in photochemical destruction of ozone.
Lehmann, P.↗
Transient Convection Due to Imposed Heat Flux: Application to Liquid-Acquisition Devices
A model problem is considered that addresses the effect of heat load from an ambient laboratory environment on the temperature rise of liquid nitrogen inside an enclosure. This model has applications to liquid acquisition devices inside the cryogenic storage tanks used to transport vapor-free propellant to the main engine. We show that heat loads from Q = 0.001 to 10 W, with corresponding Rayleigh numbers from Ra = 109 to 1013, yield a range of unsteady convective states and temperature rise in the liquid. The results show that Q = 1 to 10 W (Ra = 1012 to 1013) yield temperature distributions along the enclosure height that are similar in trend to experimental measurements. Unsteady convection, which shows selfsimilarity in its planforms, is predicted for the range of heat-load conditions. The onset of convection occurs from a free-convection-dominated base flow that becomes unstable against convective instability generated at the bottom of the enclosure while the top of the enclosure is convectively stable. A number of modes are generated with small-scale thermals at the bottom of the enclosure in which the flow selforganizes into two symmetric modes prior to the onset of the propagation of the instability. These symmetric vertical modes transition to asymmetric modes that propagate as a traveling-wave-type motion of convective modes and are representative of the asymptotic convective state of the flow field. Intense vorticity production is created in the core of the flow field due to the fact that there is shear instability between the vertical and horizontal modes. For the higher Rayleigh numbers, 1012 to 1013, there is a transition from a stationary to a nonstationary response time signal of the flow and temperature fields with a mean value that increases with time over various time bands and regions of the enclosure.
Duval, Walter M. B.↗
Wettability and CHF limits of Accident-Tolerant nuclear fuel cladding materials in light water reactor conditions
Here, we present the results of experimental investigations aimed at evaluating the thermal–hydraulic performance of chromium-coated zircaloy, i.e., one of the most promising accident tolerant fuel (ATF) cladding material for light water nuclear reactors. Precisely, we investigate the wettability and critical heat flux (CHF) limits of chromium-coated and conventional zircaloy surfaces in prototypical reactor conditions. For both surface types, we measure the contact angle in a vapor-saturated atmosphere from ambient pressure to the operating pressure of pressurized water reactors (PWRs), i.e., ~15 MPa. We measure the ambient-pressure steady-state flow boiling CHF with a spatially uniform heat flux. We measure the high-pressure steady-state flow boiling CHF with a cosine shape heat flux (up to 20 MPa) and with a uniform heat flux (up to 15 MPa), also exploring the effect of CRUD deposits on the chromium-coated surface. Our results reveal that the chromium surface and the bare zircaloy surface have similar wettability and both become super-hydrophilic in PWR conditions, and that there is practically no difference in the steady-state CHF limits, both at low-pressure and high-pressure conditions, also when the chromium-coated surface is covered by a CRUD deposit. However, while the chromium-coating does not improve the CHF compared to the bare zircaloy surface, it improves the post-CHF behavior. The chromium coating prevents the reaction between zircaloy and steam, which results in the formation of a brittle zirconium oxide through the surface of the cladding. We also measure the transient CHF under exponentially escalating heat flux inputs of a nano-smooth and a rough surface mimicking a chromium-coated zircaloy cladding. Interestingly, the results of the transient heat flux tests suggest that the CHF limit for very short periods (i.e., fast transients) is independent of the surface finish, being the same for a rough chromium surface or a nano-smooth surface.
36 MATERIALS SCIENCE↗
Exact analytical solution to a transient conjugate heat-transfer problem
An exact analytical solution is found for laminar, constant-property, slug flow over a thin plate which is also convectively cooled from below. The solution is found by means of two successive Laplace transformations when a transient in the plate and the fluid is initiated by a step change in the fluid inlet temperature. The exact solution yields the transient fluid temperature, surface heat flux, and surface temperature distributions. The results of the exact transient solution for the surface heat flux are compared to the quasi-steady values, and a criterion for the validity of the quasi-steady results is found. Also the effect of the plate coupling parameter on the surface heat flux are investigated.
Sucec, J.↗
Progress in the measurement of SSME turbine heat flux with plug-type sensors
Data reduction was completed for tests of plug-type heat flux sensors (gauges) in a turbine blade thermal cycling tester (TBT) that is located at NASA/Marshall Space Flight Center, and a typical gauge is illustrated. This is the first time that heat flux has been measured in a Space Shuttle Main Engine (SSME) Turbopump Turbine environment. The development of the concept for the gauge was performed in a heat flux measurement facility at Lewis. In this facility, transient and steady state absorbed surface heat flux information was obtained from transient temperature measurements taken at points within the gauge. A schematic of the TBT is presented, and plots of the absorbed surface heat flux measured on the three blades tested in the TBT are presented. High quality heat flux values were measured on all three blades. The experiments demonstrated that reliable and durable gauges can be repeatedly fabricated into the airfoils. The experiment heat flux data are being used for verification of SSME analytical stress, boundary layer, and heat transfer design models. Other experimental results and future plans are also presented.
Liebert, Curt H.↗
On the subject of large-scale pool fires and turbulent boundary layer interactions
The role to which a realistic inflow turbulent boundary layer (TBL) influences transient and mean large-scale pool fire quantities of interest (QoIs) is numerically investigated. High-fidelity, low-Mach large-eddy simulations that activate low-dissipation, unstructured numerics are conducted using an unsteady flamelet combustion modeling approach with mutiphysics coupling to soot and participating media radiation transport. Three inlet profile configurations are exercised for a large-scale, high-aspect rectangular pool that is oriented perpendicular to the flow direction: a time-varying, TBL inflow profile obtained from a periodic precursor simulation, the time-mean of the transient TBL, and a steady power-law inflow profile that replicates the mean TBL crosswind velocity of 10.0 m/s at a vertical height of 10 m. Results include both qualitative transient flame evolution and quantitative flame shape with ground-level temperature and convective/radiative heat flux profiles. While transient fire events, which are driven by burst-sweep TBL coupling, such as blow-off and reattachment are vastly different in the TBL case (contributing to increased root mean square QoI fluctuation prediction and disparate flame lengths), mean surface QoI magnitudes are similar. Quadrant analysis demonstrates that the TBL configuration modifies burst-sweep phenomena at windward pool locations, while leeward recovery is found. In conclusion, positive fluctuations of convective heat flux correlate with fast moving fluid away from the pool surface due to intermittent combustion events.
97 MATHEMATICS AND COMPUTING↗
Sensitivity analysis of in-pile critical heat flux experiments in TREAT for characterization of RIA power-transient effects
A reactivity-initiated accident (RIA) is one type of postulated design basis accident (DBA) that can cause a departure from nucleate boiling (DNB) event in pressurized-water reactors (PWRs). A DNB occurrence and its consequences depend on the thermophysical properties of the fuel components and coolant, characteristics of the transient energy insertion into the fuel rod, and the onset of the critical heat flux (CHF) phenomenon. To leverage the restart of the Transient Reactor Test (TREAT) Facility, an effort is currently underway to better understand the cladding-to-coolant heat transfer mechanisms and the CHF phenomenon under fast-transient irradiation conditions. This paper characterizes the impact of power transients on the thermal-hydraulic behavior of a TREAT Facility reactor heater rodlet CHF experiment to provide the priority of parameters that need to be investigated for an improved CHF model. Sobol sensitivity analysis methods and the Reactor Excursion and Leak Analysis Program (RELAP5-3D) code were used to identify key input parameters on the uncertainty in the prediction of peak outer- and inner-surface temperatures of the heater tube, as well as the time of the DNB event. A series of sensitivity analyses revealed the total energy deposition on the tube and the transient effects of power pulse had large impacts on the maximum temperatures. The CHF multiplier had the largest impact on the time occurrence of CHF. The overall results show the energy deposition rate in the tube is the most influencing factor to the manifestation of CHF and the resulting thermal-hydraulic behaviors of the tube. The multiplier for the CHF, which is interpreted as the predicted CHF value, has the largest Sobol indices for the time of the CHF in all cases, since it directly determines the occurrence of CHF. It is inferred that the uncertainties in the thermal-hydraulic behaviors of fuels increase with respect to the key parameters as the power pulse becomes broader, and an accurate estimation of the energy deposition rate is required to reduce the uncertainty in the evaluation of the integrity of fuel if the CHF is expected to occur near the peak power. Therefore, the outputs are expected to provide rigorous interpretation of ongoing in-pile CHF experiments in the TREAT Facility reactor regarding thermal-hydraulic behavior of the fuel system aiming for a new transient in-pile CHF model.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Heat-Flux-Measuring Facility
Apparatus simulates conditions in turbine engines. Automated facility generates and measures transient and steady-state heat fluxes at flux densities from 0.3 to 6 MW/m(Sup2) and temperatures from 100 to 1,200 K. Positioning arm holds heat-flux gauge at focal point of arc lamp. Arm previously chilled gauge in liquid nitrogen in Dewar flask. Cooling water flows through lamp to heat exchanger. Used to develop heat-flux gauges for turbine blades and to test materials for durability under rapidly changing temperatures.
Liebert, Curt H.↗
Calculation of Critical Heat Flux Using an Inverse Heat Transfer Method to Support TREAT Experiment Analysis
Heat transfer between cladding and coolant during transient scenarios remains a critical area of uncertainty in understanding nuclear reactor safety. To advance the understanding of transient and accident scenarios involving critical heat flux (CHF), an in-pile experiment for the Transient Reactor Test (TREAT) facility at Idaho National Laboratory was developed. The experiment named CHF-Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA), consists of a hollow borated stainless steel heater rod submerged in a static water pool which is heated via the (n, a) reaction in boron-10. This paper presents a novel inverse heat transfer method to determine CHF by using the optimization and uncertainty software DAKOTA to calibrate a RELAP5-3D model of CHF-SERTTA to temperature measurements obtained from a thermocouple welded to the surface of the rod.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
The Maintainable Fusion Pilot Plant
The US fusion community has coalesced around the goal of building an FPP as described by the National Academies of Science, Engineering, and Medicine (NASEM). In addition to demonstrating the viability of the technologies necessary to operate such a plant, including demonstration of net energy and electricity production, NASEM found that a “fusion pilot plant will need to demonstrate the ability to efficiently perform remote maintenance and replacement in support of the design of a power plant, taking into account details of the consequences of the fusion environment, such as material activation and tritium retention in components.” Current designs of fusion demonstration reactors do usually foresee a regular exchange of their first wall modules, including the tritium breeding blankets. In the European Power Plant Conceptual Studies, it is assumed that a fusion reactor will need to change its divertor every 2 years and its first wall blanket module every 5 to 6 years to reach acceptable availability. Underlying this capability are remote-handling technologies to keep the outage for the exchange of these components short. There are many uncertainties in the remote-handling schemes, and most schemes are at a preconceptual level at best. In addition, the exchange of these components would either produce an enormous rad-waste stream or would require an enormous refurbishment activity with huge cost-prohibitive hot-cells. Past Fusion Nuclear Science Facility (FNSF) preconceptual studies have led to hot cell dimensions of an unbelievable size, likely costing tens of billions of dollars. Already at The Way (previously International Thermonuclear Experimental Reactor, ITER), hot-cells have become cost-prohibitive, demanding redesigns of the ITER first wall to reduce the toxic rad-waste/inventory. In this in-situ PFC repair project, a concept for a long-life, maintainable first wall module concept is developed and tested. This first wall concept relies on innovative remote handling to repair the first wall modules in-situ, avoiding costly refurbishments outside of the tokamak vessel. This approach was highlighted in the Fusion Energy Sciences Advisory Committee (FESAC) report on Transformative Enabling Capabilities for Efficient Advance Toward Fusion Energy. In general, the damage of the first wall armor is due to particle and radiation exposures. Load conditions vary from one fusion reactor design to another. In tokamaks, first wall Plasma Facing Components (PFCs) are exposed to far-Scrape-Off-Layer plasma fluxes, electromagnetic radiation, energetic CX neutrals, and potentially runaway electron beams. Protecting the first wall to the worst-case load conditions would require the design of a very thick first wall armor. Transient heat and particle fluxes due to disruptions or edge localized modes can lead to excessive heat loads resulting potentially in melting PFCs down to the cooling channel. Catastrophic events like these need to be avoided by appropriate disruption mitigation systems. However, failure of these systems will still put a first wall at an unacceptable risk. Hence, a first wall design needs to accommodate the occasional transient heat loads by introducing sacrificial limiters, which will absorb these transients.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗