The Material Plasma Exposure eXperiment (MPEX) is a steady-state linear plasma device designed to expose neutron-irradiated materials to fusion divertor prototypic plasma conditions to perform plasma-material interaction (PMI) studies. The MPEX device will be capable of ion fluxes of 10(25) m(-2)s(-1), power fluxes up to 40 MW/m(2), ion fluences up to 10(31) m(-2), and operation at steady state for up to 10(6) s with magnetic fields up to 2.5 T. PMIs occur when the plasma directly impinges upon a surface. In the case of MPEX, this occurs primarily at the MPEX target. Observation of interaction using a variety of instruments is required. A PMI chamber was designed to enable these observations using currently envisioned diagnostics and to accommodate future instruments. The design includes 59 plasma-facing ports, 10 of which explicitly point at the target, and an additional 10 ports that can be used to assess the vacuum space in the chamber. The entire vacuum chamber is water-cooled and will experience sustained heat from plasma radiated power, microwaves, and neutral gas thermal loads. Because of the method of fabrication, this chamber has undergone significant manufacturability testing. The chamber includes provisions for chamber, plasma, and target diagnostics such as a residual gas analyzer, Thomson scattering lasers, and both visible and IR cameras. The design effort included operational testing of the autocoupler to ensure vacuum integrity and included machining and welding studies to verify that the tolerances required by the diagnostics could be held. Provisions have also been included to eventually accommodate a water-cooled target dump, an irradiated sample recovery module, and any other hardware needed to support future target designs and diagnostics.

The Material Plasma Exposure eXperiment (MPEX) at Oak Ridge National Laboratory is in the final design phase. MPEX will be capable of exposing neutron-irradiated materials to plasmas for the study of plasma-material interaction. This facility will provide information about the complex effects of plasmas on materials and contribute to examining new materials that can withstand high heat fluxes and high ion fluences for future fusion devices. MPEX plasma is heated by 70-GHz or 105-GHz electron Bernstein wave/electron cyclotron heating (ECH), and the high-frequency microwaves are prone to scattering microwave power, which can have detrimental effects, especially on diagnostic components. A large portion of the injected ECH power is expected to be absorbed by plasma, but the remainder requires that microwave absorbers be placed immediately upstream and downstream of the ECH launcher to minimize stray microwaves leaving the ECH region. These microwaves can inadvertently heat components that cannot be shielded or otherwise protected. The microwave absorber design is based on an array of pyramid-shaped ceramic tiles brazed to a water-cooled explosion-bonded heat sink and a stainless steel plate to produce one tile module. Computational fluid dynamics and structural analyses were performed to optimize and validate the design. Multiple test coupons were produced to validate the process for brazing the two different tile materials to the Glidcop AL-15 baseplate. The articles were tested to evaluate the reliability and thermal performance through exposure to an electron beam with a heat flux of up to 1.5 MW/m(2). Nondestructive testing was performed before and after testing to identify voids or separations that may have been introduced by the high heat flux. This paper discusses the details of high heat flux microwave absorber design, manufacturing details and associated challenges, and test results, demonstrating the effectiveness of the proposed design.

The Material Plasma Exposure eXperiment (MPEX) project seeks to design a steady-state linear plasma facility at Oak Ridge National Laboratory that will be used to study plasma-material interactions (PMIs) at fusion prototypic levels, supporting the evaluation and development of materials for the next generation of fusion devices. This study is focused on PMI exposure of small-size neutron-irradiated specimens, which are clamped onto an actively cooled component. A thermohydraulic evaluation of a new MPEX target assembly design to assess the appropriate operation during MPEX operation is presented. To further guide the design and assess the structural integrity of the components under expected loads, preliminary thermomechanical stress analyses were also conducted. To ensure good thermal contact between the components, thermal interface materials, such as silver flexible graphite, were used in the assembly.It was found that the maximum target temperatures of 1572, 1463, and 1315 K were obtained for Grafoil thicknesses of 0.61, 0.38, and 0.25 mm, respectively. The distribution of the axial deformation at high heat fluxes showed that there are no gaps between components, indicating good contact at material interfaces. Moreover, the contact pressure between the target and other components indicated that very good contact was established at these interfaces. The stress-strain conditions for the target will be further used to assess the appropriate operation during MPEX experiments and gain insight into materials science phenomena during PMI experiments.

The Material Plasma Exposure eXperiment (MPEX) vacuum pumping system is responsible for creating prototypic conditions in the plasma material interaction chamber that mimic those in a fusion reactor divertor region. Additionally, the vacuum system needs to minimize the pressure in the plasma heating region to improve the coupling of the electron cyclotron heating and ion cyclotron heating to the plasma, minimizing waste heat exhausted to high heat flux components. The final design of the system has been sized to comply with the vacuum pump operating environment and to reduce the number of unique pumps required while meeting performance requirements. Bounding cases with and without the plasma present have been developed in the pumping analysis, and an initial calculation has been performed based on the plasma pumping identified in Proto-MPEX operation though this result will remain unverified until MPEX operation.

The Material Plasma Exposure eXperiment (MPEX) has completed its design phase. MPEX will be a unique facility to investigate plasma material interactions (PMIs) under fusion prototypic divertor conditions in steady state. This includes plasma exposure conditions expected in a fusion reactor divertor. Materials to be investigated will include solids, liquids, and neutron pre-irradiated materials. Electron and ion heating will allow for a large operational domain ranging from erosion-dominated PMI conditions to deposition-dominated PMI conditions. An overview of the final design for all MPEX systems is given. In particular, it is shown how mission-driven project requirements have led to detailed design choices with innovative solutions. Examples are the water-cooled helicon window, the electron cyclotron heating launcher, the target holder and manipulator, and the autonomous decouplers.

The Material Plasma Exposure eXperiment (MPEX) device is a linear plasma device developed to perform plasma material interaction experiments under the conditions prototypic of a fusion reactor divertor. MPEX has multiple systems that must be precisely aligned to the plasma axis, including an electron cyclotron heating system that emits up to 400 kW of microwave power into the vacuum vessel. Five distinct systems require precise alignment on the MPEX device, thus requiring four bellows, all of which are adjacent to the plasma at a relatively high heat flux of approximately 47 kW/m2 and microwave power regions. The MPEX high heat flux bellows (HHFB) is designed to deliver 6 degrees of freedom positioning. The HHFB includes titanium–zirconium–molybdenum (TZM) inserts that are brazed into a Glidcop AL-15 body using a high-temperature braze alloy, thus blocking direct line of sight to an edge-welded bellows from the plasma and microwave screen to block microwaves from the bellows. A custom ConFlat knife edge is machined into the Glidcop AL-15, so the vacuum flanges do not need a braze or weld joint on the vacuum interface. Fingerstock or copper mesh is used to restrict microwave power from entering the interstitial space between the water-cooled Glidcop AL-15 body and edge-welded bellows. Glidcop AL-15 was selected as the material of choice for the water-cooled body because it can maintain mechanical integrity at elevated temperatures, and it also allows for a high-temperature braze. TZM was selected for its machinability and compatibility with vacuum and plasma requirements. A test article is also being considered for a similarly shaped component, the MPEX limiter, that will demonstrate the integrity of the braze joint under high thermal load. Results from this testing will be extrapolated to deduce the lifetime and integrity of the HHFB design.

The Material Plasma Exposure eXperiment (MPEX) device is a steady-state linear plasma device currently in the final design phase at the Oak Ridge National Laboratory. This device will reach ion fluences up to $10^{31}\,\,\text{m}^{-2}$ and will be used to expose neutron-irradiated materials to divertor-relevant plasmas and to study the effects of plasma-material interactions. These studies will elucidate the complex effects of plasmas with divertor candidate materials capable of withstanding high heat flux and high fluences for next-generation fusion devices. Bidirectional plasma will be generated using a high-power (200 kW) helicon source. Plasma will be confined by superconducting magnets. The last plasma-facing component on the upstream side of the MPEX device is the dump, which has been designed to intercept plasma and energetic particles. The dump will have a total heat load of 9.2 kW. A copper alloy (Glidcop AL-15) was selected for use in the water-cooled flange design because of its high thermal conductivity, its retaining strength at elevated temperatures, and its ability to be used in the high-temperature braze joints used in this application. Titanium-zirconium-molybdenum (TZM) tiles are brazed to the Glidcop AL-15 flange using high-temperature braze alloy. External water-cooling channels are used on the dump flange to prevent water leakage inside the vacuum space. This article discusses the details of the high heat flux dump design, including the computational fluid dynamics (CFD) and structural analyses performed to validate the design to meet the operational requirements of the MPEX device.

The Material Plasma Exposure eXperiment (MPEX) steady-state linear plasma facility is currently under design at Oak Ridge National Laboratory to expose target specimens to fusion divertor regimes. The neutron-irradiated target is actively cooled and remote handled in the MPEX facility for conducting plasma-material-interaction (PMI) experiments. In this study, the steady-state stresses in the target and target assembly system are investigated using two-dimensional (2-D) and three-dimensional (3-D) models to provide expected stresses/strains under the heat loads to which various system components would be exposed during MPEX operation. The calculated temperatures from the 2-D axisymmetric mechanical model were found to be in excellent agreement with those from the full 3-D thermohydraulic model, providing a strong model validation. Numerical simulation results for the steady-state mechanical model indicate nonuniform distributions for the temperature, stress, and deformation within the critical components. For the initial design, the deformation results indicate possible gap openings between contacting surfaces below the plasma-facing materials. To reduce the possibility of interfacial gap opening, the target assembly was slightly changed and evaluated using the 2-D stress model. Numerical simulation results indicate that the interfacial gap openings can be minimized without drastically changing the entire target assembly. The stress-strain conditions for the target will be further used to assess the appropriate operation during MPEX experiments and gain insight into materials science phenomena during PMI.

The Facility for Rare Isotope Beams (FRIB) under construction at Michigan State University is based on a 400kW heavy ion accelerator and uses in-flight production and separation to generate rare isotope beams. The first section of the fragment separator houses the rare isotope production target, and the primary beam dump to stop the unreacted primary beam. The experimental program will use 400kW ion beams from 16O to 238U. After interaction with the production target, over 300kW in remaining beam power must be absorbed by the beam dump. A rotating water-cooled thin-shell metal drum was chosen as the basic concept for the beam dump. Extensive thermal, mechanical and fluid flow analyses were performed to evaluate the effects of the high power density in the beam dump shell and in the water. Many properties were optimized simultaneously, such as shell temperature, mechanical strength, fatigue strength, and radiation resistance. Results of the analyses of the beam dump performance with different design options will be discussed. For example, it was found that a design modification to the initial water flow pattern resulted in a substantial increase in the wall heat transfer coefficient. A detailed evaluation of materials for the shell is in progress. The widely used titanium alloy, Ti–6Al–4V (wt%), is presently considered as the best candidate, and is the subject of specific tests, such as studies of performance under heavy ion irradiation.

Effective high-temperature thermal energy exchange and delivery at temperatures over 600°C has the potential of significant impact by reducing both the capital and operating cost of energy conversion and transport systems. It is one of the key technologies necessary for efficient hydrogen production and could potentially enhance efficiencies of high-temperature solar systems. Today, there are no standard commercially available high-performance heat transfer fluids above 600°C. High pressures associated with water and gaseous coolants (such as helium) at elevated temperatures impose limiting design conditions for the materials in most energy systems. Liquid salts offer high-temperature capabilities at low vapor pressures, good heat transport properties, and reasonable costs and are therefore leading candidate fluids for next-generation energy production. Liquid-fluoride-salt-cooled, graphite-moderated reactors, referred to as Fluoride Salt Reactors (FHRs), are specifically designed to exploit the excellent heat transfer properties of liquid fluoride salts while maximizing their thermal efficiency and minimizing cost. The FHR s outstanding heat transfer properties, combined with its fully passive safety, make this reactor the most technologically desirable nuclear power reactor class for next-generation energy production. Multiple FHR designs are presently being considered. These range from the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) [1] design originally developed by UC-Berkeley to the Small Advanced High-Temperature Reactor (SmAHTR) and the large scale FHR both being developed at ORNL [2]. The value of high-temperature, molten-salt-cooled reactors is also recognized internationally, and Czechoslovakia, France, India, and China all have salt-cooled reactor development under way. The liquid salt experiment presently being developed uses the PB-AHTR as its focus. One core design of the PB-AHTR features multiple 20 cm diameter, 3.2 m long fuel channels with 3 cm diameter graphite-based fuel pebbles slowly circulating up through the core. Molten salt coolant (FLiBe) at 700°C flows concurrently (at significantly higher velocity) with the pebbles and is used to remove heat generated in the reactor core (approximately 1280 W/pebble), and supply it to a power conversion system. Refueling equipment continuously sorts spent fuel pebbles and replaces spent or damaged pebbles with fresh fuel. By combining greater or fewer numbers of pebble channel assemblies, multiple reactor designs with varying power levels can be offered. The PB-AHTR design is discussed in detail in Reference [1] and is shown schematically in Fig. 1. Fig. 1. PB-AHTR concept (drawing taken from Peterson et al., Design and Development of the Modular PB-AHTR Proceedings of ICApp 08). Pebble behavior within the core is a key issue in proving the viability of this concept. This includes understanding the behavior of the pebbles thermally, hydraulically, and mechanically (quantifying pebble wear characteristics, flow channel wear, etc). The experiment being developed is an initial step in characterizing the pebble behavior under realistic PB-AHTR operating conditions. It focuses on thermal and hydraulic behavior of a static pebble bed using a convective salt loop to provide prototypic fluid conditions to the bed, and a unique inductive heating technique to provide prototypic heating in the pebbles. The facility design is sufficiently versatile to allow a variety of other experimentation to be performed in the future. The facility can accommodate testing of scaled reactor components or sub-components such as flow diodes, salt-to-salt heat exchangers, and improved pump designs as well as testing of refueling equipment, high temperature instrumentation, and other reactor core designs.

The need for high-temperature (greater than 600 degrees C) energy transport systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The fluoride-salt-cooled high-temperature reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process.The experimental facility described here has been constructed to support the development of the fluoride-salt-cooled high-temperature reactor concept. The facility is capable of operating at up to 700 degrees C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system, a trace heating system, and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop.The initial experiment is aimed at measuring fluoride-salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble-bed advanced high-temperature reactor.This paper describes the details of the loop design, auxiliary systems used to support the facility, inductive heating system, and facility capabilities. (C) 2013 Elsevier Ltd. All rights reserved.

The Facility for Rare Isotopes Beams (FRIB) at Michigan State University will use projectile fragmentation and induced in-flight fission of heavy-ion primary beams at energies of 200MeV/u and higher and at a beam power of 400kW to generate rare isotope beams for experiments in nuclear physics, nuclear astrophysics, and fundamental symmetries, as well as for societal needs. The Advanced Rare Isotope Separator (ARIS) has been designed as a three-stage fragment separator for the efficient collection and purification of the rare isotope beams of interest. A vertically bending preseparator (first stage) with production target and beam dump is fully integrated into a production target facility hot cell with remote handling. The new separator compresses the accepted momentum width of up to ±5% of the beam by a factor of three in the standard operational mode. Provisions for alternate operational modes for specific cases are included in the design. This preseparator is followed by two, horizontally-bending separator stages (second and third stages) utilizing the magnets from the existing A1900 fragment separator at the National Superconducting Cyclotron Laboratory (NSCL). These stages can alternatively be coupled to a single high-resolution separator stage, resulting in the flexibility to optimize the operation for different experiments, including momentum tagging and in-flight particle identification of rare isotope beams. The design of ARIS will be presented with an emphasis on beam physics characteristics, and anticipated operational modes will be described.

A forced-convection liquid-fluoride-salt loop is being constructed at Oak Ridge National Laboratory (ORNL). This loop was designed as a versatile experimental facility capable of supporting general thermal/fluid/corrosion testing of liquid fluoride salts. The initial test configuration is designed to support the Pebble Bed Advanced High-Temperature Reactor and incorporates a test section designed to examine the heat transfer behavior of FLiNaK salt in a heated pebble bed. The loop is constructed of Inconel 600 and is capable of operating at up to 700oC. It contains a total of 72 kg of FLiNaK salt and uses an overhung impeller centrifugal sump pump that can provide FLiNaK flow at 4.5 kg/s with a head of 0.125 MPa. The test section is made of silicon carbide (SiC) and contains approximately 600 graphite spheres, 3 cm in diameter. The pebble bed is heated using a unique inductive technique. A forced induction air cooler removes the heat added to the pebble bed. The salt level within the loop is maintained by controlling an argon cover gas pressure. Salt purification is performed in batch mode by transferring the salt from the loop into a specially made nickel crucible system designed to remove oxygen, moisture and other salt more » impurities. Materials selection for the loop and test section material was informed by 3 months of Inconel 600 and SiC corrosion testing as well as tests examining subcomponent performance in the salt. Several SiC-to-Inconel 600 mechanical joint designs were considered before final salt and gas seals were chosen. Structural calculations of the SiC test section were performed to arrive at a satisfactory test section configuration. Several pump vendors provided potential loop pump designs; however, because of cost, the pump was designed and fabricated in-house. The pump includes a commercial rotating dry gas shaft seal to maintain loop cover gas inventory. The primary instrumentation on the loop includes temperature, pressure, and loop flow rate measurement. Although techniques for all these measurements have improved, no commercial instrumentation was available for flow and pressure measurement that had been tested under these conditions. Instrumentation was tested and modified to meet both corrosion and temperature requirements. This paper discusses the issues encountered during the design and construction of the ORNL Liquid Salt Loop and should prove useful to those contemplating construction of similar high-temperature liquid-fluoride-salt facilities. « less

A small forced-convection molten-fluoride-salt loop is being constructed at Oak Ridge National Laboratory to examine the heat transfer behavior of FLiNaK salt in a heated pebble bed. Objectives of the experiment include reestablishing infrastructure needed for fluoride-salt loop testing, developing a unique inductive heating technique for performing heat transfer (or other) experiments, measuring heat transfer characteristics in a liquid-fluoride-salt-cooled pebble bed, and demonstrating the use of silicon carbide (SiC) as a structural component for salt systems. The salt loop will consist of an Inconel 600 piping system, a sump-type pump, a SiC test section, and an air-cooled heat exchanger, as well as auxiliary systems needed to pre-heat the loop, transport salt into and out of the loop, and maintain an inert cover gas over the salt. A 30,000 Hz inductive heating system will be used to provide up to 250 kW of power to a 15 cm diameter SiC test section containing a packed bed of 3 cm graphite spheres. A SiC-to-Inconel 600 joint will use a conventional nickel/grafoil spiral wound gasket sandwiched between SiC and Inconel flanges. The loop system can provide up to 4.5 kg/s of salt flow at a head of 0.125 MPa and operate at a pressure just above atmospheric. Pebble Reynolds numbers of up to 2600 are possible with this configuration. A sump system is provided to drain and store the salt when not in use. Instrumentation on the loop will include pressure, temperature, and flow measurements, while the test section will be instrumented to provide pebble and FLiNaK temperatures.