A new experimental facility for materials irradiation and testing at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is being developed. Details of this facility have been presented before [1, 2]. A prototype of this facility, the Thermo-Syphon Test Loop (TSTL) has been built and experimental data have been obtained and analyzed [3, 4]. Pre-test calculations for this facility with the RELAP5-3D code [5] have been presented previously [6] as well as other calculations [7, 8] with the TRACE code [9]. The results of both codes were very different [7]. RELAP5-3D predicted much higher pressures and temperatures than TRACE. This paper compares calculated results with the TSTL experimental data.

Populations of small helium gas bubbles were introduced into a flowing mercury experiment test loop to evaluate mitigation of beam-pulse induced cavitation damage and pressure waves. The test loop was developed and thoroughly tested at the Spallation Neutron Source (SNS) prior to irradiations at the Los Alamos Neutron Science Center–Weapons Neutron Research (LANSCE–WNR) facility. Twelve candidate bubblers were evaluated over a range of mercury flow and gas injection rates by use of a novel optical measurement technique that accurately assessed the generated small bubble size distributions. Final selection for irradiation testing included two variations of a swirl bubbler provided by Japan Proton Accelerator Research Complex (J-PARC) collaborators and one orifice bubbler developed at SNS. Bubble populations of interest consisted of sizes up to 150μm in radius with achieved gas volume fractions in the 10−5–10−4 range. The nominal WNR beam pulse used for the experiment created energy deposition in the mercury comparable to SNS pulses operating at 2.5MW. Nineteen test conditions were completed each with 100 pulses, including variations on mercury flow, gas injection and protons per pulse. The principal measure of cavitation damage mitigation was pitting damage assessment on test specimens that were manually replaced for each test condition. Damage assessment was done after radiation decay and decontamination by optical and laser profiling microscopy with damaged area fraction and maximum pit depth being the more valued results. Damage was reduced by flow alone; the best mitigation from bubble injection was to one-third that of stagnant mercury. Other data collected included surface motion tracking by three Laser Doppler Vibrometers (LDV), test loop wall dynamic strain, beam diagnostics for charge and beam profile assessment, embedded hydrophones and pressure sensors, and sound measurement by a suite of conventional and contact microphones.

Proton beam experiments investigating cavitation damage in short pulse mercury spallation targets were performed at LANSCE - WNR in July of 2008. They included two main areas for investigation: damage dependence on mercury velocity using geometry more prototypic to the SNS target than previously employed and damage dependence on incident proton beam flux intensity. The flow dependence experiment employed six test targets with mercury velocity in the channel ranging from 0 to more than 4 m/s. Each was hit with 100 WNR beam pulses with peak proton flux equivalent to that of SNS operating at 2.7 MW.Damage dependence on incident proton beam flux intensity was also investigated with three intensity levels used on simple rectangular shaped targets without mercury flow. Intensity variation was imposed by focusing the beam differently while maintaining protons per pulse. This kept total energy deposited in each target constant. A fourth test target was hit with various beams: constant protons and varied spot size; constant spot size and varied protons. No damage will be assessed in this case. Instead, acoustic emissions associated with cavitation collapse were measured by laser Doppler vibrometer (LDV) from readings of exterior vessel motions as well as by mercury wetted acoustic transducers.This paper will provide a description of the experiment and present available results. Damage assessment will require several months before surface analysis can be completed and was not available in time for IWSMT-9. Published by Elsevier B.V.

The Spallation Neutron Source (SNS) is an accelerator-based neutron source at Oak Ridge National Laboratory (ORNL).The nuclear spallation reaction occurs when a proton beam hits liquid mercury. This interaction causes thermal expansion of the liquid mercury which produces high pressure waves. When these pressure waves hit the target vessel wall, cavitation can occur and erode the wall. Research and development efforts at SNS include creation of a vertical protective gas layer between the flowing liquid mercury and target vessel wall to mitigate the cavitation damage erosion and extend the life time of the target. Since mercury is opaque, computational fluid dynamics (CFD) may be used as a diagnostic tool to visualize the behavior of the liquid mercury and guide the experimental efforts. In this study, CFD simulations of three dimensional, unsteady, turbulent, two-phase flow of helium gas injection in flowing liquid mercury over smooth, vertically grooved and horizontally grooved walls are carried out with the commercially available CFD code Fluent-12 from ANSYS. The Volume of Fluid (VOF) model is used to track the helium-mercury interface. V-shaped vertical and horizontal grooves with 0.5 mm pitch and about 0.7 mm depth were machined in the transparent wall of acrylic test sections. Flow visualization data of helium gas coverage through transparent test sections is obtained with a high-speed camera at the ORNL Target Test Facility (TTF). The helium gas mass flow rate is 8 mg/min and introduced through a 0.5 mm diameter port. The inlet mercury mass flow rate is 51 kg/s and the predicted local mercury velocity is 0.9 m/s. In this paper, the helium gas flow rate and the local mercury velocity are kept constant for the three cases. Time integration of predicted helium gas volume fraction over time is done to evaluate the gas coverage and calculate the average thickness of the helium gas layer. The predicted time-integrated gas coverage over vertically grooved and horizontally grooved test sections is better than over a smooth wall. The simulations show that the helium gas is trapped inside the grooves.

An experiment to visualize small gas bubbles injected into mercury flowing in a test loop using proton radiography was conducted at the Los Alamos Neutron Science Center (LANSCE) in December 2006. Radiograph images of bubbles were obtained through two mercury thicknesses: 22 mm and 6 mm. Two jet bubblers and two needle bubblers were operated individually over a range of mercury flow speeds (0 - 1 m/s) and gas injection rates (0.1 - 500 sccm). Helium was most commonly used but Argon and Xenon were injected for limited test conditions. The smallest discernable bubbles were about 0.24 mm in diameter. Resolution was limited by image contrast which was notably improved with 6 mm of mercury thickness. Analysis of the radiograph images from jet bubbler conditions provided data on bubble size distribution and total bubble void fraction. In a few cases radiographs captured a large fraction of the injected gas, but generally 20 to 90% of injected gas was not captured in the images. In all more than 400 radiographs were made during the experiment in addition to several movies. Sound recordings of needle bubbler operation were also made and used to quantify bubble formation rate and size; these results aremore » compared to theoretical predictions. This paper describes the experiment goals, scope and equipment; key results are presented and discussed.« less

Intense proton beam-induced heating of the spallation neutron source mercury target will cause pressure spikes that lead to the formation of cavitation bubbles in the mercury. Erosion of the mercury container walls caused by violent collapse of bubbles could potentially limit its service lifetime. In-beam tests for a limited number of pulses (<1000 pulses for each test target) have demonstrated that pitting damage appears to be especially sensitive to beam intensity, surface treatment, and gas injection. Using results of off-line pressure pulse tests conducted for a million cycles or more to scale the results from limited in-beam tests, it is concluded that the mercury target will last at least two weeks at a time-averaged proton beam power level of 1MW. However, because of remaining uncertainties, it is concluded that further research and development and target design efforts are needed to verify these conclusions and extend the target to higher operating powers and longer lifetimes.

Previous experiments to examine corrosion resulting from thermal gradient mass transfer of type 316L stainless steel in mercury were conducted in thermal convection loops (TCLs) with an Hg velocity of about 1 m/min. These tests have now been supplemented with a series of experiments designed to examine the influence of increased flow velocity and possible cavitation conditions on compatibility. In one experiment, the standard TCL design was modified to include a reduced section in the hot leg that provided a concomitant increase in the local velocity by a factor of five. In addition, a pumped-loop experiment was operated with a flow velocity of about 1 m/s. Finally, a TCL was modified to include an ultrasonic transducer at the top of the hot leg in an attempt to generate cavitation conditions with corresponding extreme local velocity associated with collapsing bubbles. The results indicate that compatibility of type 316L/316LN stainless steel does not depend significantly on liquid metal velocity in the range of 1 m/min to 1 m/s. Benchtop cavitation experiments revealed susceptibility of 316L coupons to significant weight losses and increases in surface roughness as a result of 24 h exposure to 1.5 MPa pressure waves in Hg generated ultrasonically at 20 kHz. However, attempts to generate cavitation conditions on coupons inside the TCL with the ultrasonic transducer proved largely unsuccessful.

The R&D program for the Spallation Neutron Source (SNS) mercury target addresses issues associated with the quasi-static power handling and the effects of pressure waves created by the pulsed energy deposition. Three flow loops have provided the data necessary to confirm that mercury can transport the deposited proton beam power away from the target while also serving to adequately cool the stainless steel target vessel. Pressure pulse effects studies have considered (1) the disturbance of the flow, (2) the strains induced in the mercury container, and (3) the erosion caused by cavitation bubble collapse. Progress on these items including recent results of accelerator target tests and efforts to understand the erosion scaling is discussed.

The heat transfer coefficient is inferred from measurements for mercury flowing in a channel of cross-section 2 mm×40 mm with flow velocities from 1 m/s to 4 m/s and heat fluxes from 192 kW/m2 to 1.14 MW/m2. Mercury bulk temperatures vary from 67°C to 143°C. Inferred heat transfer coefficients agree with open literature tube data when compared on a Nusselt versus. Peclet number plot, with Nusselt numbers examined from 8 to 17 and Peclet numbers examined from 790 to 3070.

The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime.

A number of test facilities have been built in the last 7 yr to provide support to the Advanced Neutron Source Reactor (ANSR) design development and safety analysis. These facilities have provided required data and analytical model benchmarks critical to the project. This paper provides a summary of the results obtained from four test programs [corrosion testing, critical heat flux (CHF) and flow instability testing, flow blockage testing, and fuel plate stability testing].

Abstract A design calculation has been performed to determine thermal limits in support of an experiment in two-phase microchannel water cooling. Under the operating conditions (one atmosphere pressure and 23°C inlet temperature), the calculation predicts that the experimental channel can withstand a maximum surface temperature of 115°C and a heat flux up to 975 W/cm2 without exceeding the critical heat flux limit. The predicted results also indicate that a uniform heat flux along the channel in the two-phase domain can be achieved so that the heat losses from the experimental test section can be calculated in a straightforward manner.

The Advanced Neutron Source (ANS) reactor was designed to provide a research tool with capabilities beyond those of any existing reactors. One portion of its state-of-the-art design required high-speed fluid flow through narrow channels between the fuel plates in the core. Experience with previous reactors has shown that fuel plate damage can occur when debris becomes lodged at the entrance to these channels. Such debris disrupts the fluid flow to the plate surfaces and can prevent adequate cooling of the fuel. Preliminary ANS designs addressed this issue by providing an unheated entrance length for each fuel plate so that any flow disruption would recover, thus providing adequate heat removal from the downstream, heated portions of the fuel plates. As part of the safety analysis, the adequacy of this unheated entrance length was assessed using both analytical models and experimental measurements. The Flow Blockage Test Facility (FBTF) was designed and built to conduct experiments in an environment closely matching the ANS channel geometry. The FBTF permitted careful measurements of both heat transfer and hydraulic parameters. In addition to these experimental efforts, a thin, rectangular channel was modeled using the Fluent computational fluid dynamics computer code. The numerical results were compared withmore » the experimental data to benchmark the hydrodynamics of the model. After this comparison, the model was extended to include those elements of the safety analysis that were difficult to measure experimentally. These elements included the high wall heat flux pattern and variable fluid properties. The results were used to determine the relationship between potential blockage sizes and the unheated entrance length required.« less

To support the design of the Advanced Neutron Source (ANS), an experimental program was conducted wherein aluminum alloy specimens were exposed at high heat fluxes to high-velocity aqueous coolants in a corrosion test loop. The aluminum alloys selected for exposure were candidate fuel cladding materials, and the loop system was constructed to emulate the primary coolant system for the proposed ANS reactor. One major result of this program has been the generation of an experimental database defining oxide film growth on 6061 aluminum alloy cladding. Additionally, a data correlation was developed from the database to permit the prediction of film growth for any reasonable thermal-hydraulic excursion. This capability was utilized effectively during the conceptual design stages of the reactor. During the course of this research, it became clear that the kinetics of film growth on the aluminum alloy specimens were sensitively dependent on the chemistry of the aqueous coolant and that relatively small deviations from the intended pH 5 operational level resulted in unexpectedly large changes in the corrosion behavior. Examination of the kinetic influences and the details of the film morphology suggested that a mechanism involving mass transport from other parts of the test loop was involved. Such a mechanism would also be expected to be active in the operating reactor. This report emphasizes the results of experiments that best illustrate the influence of the nonthermal-hydraulic parameters on film growth and presents data to show that comparatively small variations in pH near 5.0 invoke a sensitive response. Simply, for operation in the temperature and heat flux range appropriate for the ANS studies, coolant pH levels from 4.5 to 4.9 produced significantly less film growth than those from pH 5.1 to 6. A mechanism for this behavior based on the concept of treating the entire loop as an active corrosion system is presented.