The Fluoride High-Temperature Reactor (FHR) technology promises many benefits including passive safety, proliferation-resistant waste forms, and improved economics. However, selection of reliable structural materials and identification of the possible degradation mechanisms for these is important for the licensure and the safe operation of FHRs. In order to address this task, the Georgia Tech led Integrated Research Project (IRP) hosted a Phenomena Identification and Ranking Table (PIRT) panel of experts to address degradation mechanisms and other materials related issues of importance to the FHRs. Materials, ones that come in contact with FLiBe or FLiNaK molten salts or other related environments like high temperature steam etc., were considered in this PIRT. Focus of this PIRT was the metallic alloys, especially the ones that are permitted for the construction of elevated temperature Class A components by the ASME code. Degradation mechanisms considered in this PIRT included chemical degradation, mechanical degradation, radiation degradation, and synergistic effect of these mechanisms that may negatively impact operations or cause some safety concerns for the major structural components of FHRs. Main components which were considered included vessel and primary piping, primary heat exchangers, steam generator vessel, steam generator tubes, intermediate loop piping, valves and pumps. Welds in all structural components were identified as an important class of material, which varies in composition and properties, and needs more attention. Importance of impurity control in molten fluorides considered for FHR was highlighted throughout PIRT panel discussions. This paper gives a summary of important results from the PIRT panel discussions and report.

A small liquid fluoride salt test apparatus has been constructed and testing has been conducted to examine the compatibility of silicon carbide (SiC), Inconel 600 and a spiral wound gasket material in FLiNaK, the ternary eutectic alkaline metal fluoride salt mixture. These tests were conducted to evaluate materials and sealing systems that could be used in fluoride salt systems. Three months of testing at 700 degrees C was conducted to assure that these materials and seals would be acceptable when operating under prototypic operating conditions. The SiC specimens showed little or no change over the test period, while the spiral wound gasket material did not show any degradation except that salt might have been seeping into the outermost spirals of the gasket. The Inconel 600 specimens showed regions of voiding which penetrated the specimen surface to about 250 mu m in depth. Analysis indicated that the salt had leached chrome from the Inconel surface, as was expected for this material. (C) 2016 Elsevier B.V. All rights reserved.

Abstract The Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride salt–cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS heat exchanger (DHX) and the natural draft heat exchanger (NDHX). To experimentally investigate the thermal performance of the DRACS, a scaled-down low-temperature DRACS test facility (LTDF) has been constructed. The design of the LTDF is obtained through a detailed scaling analysis based on a 200-kW prototypic DRACS design developed at The Ohio State University. The LTDF has a nominal power capacity of 6 kW. It employs water pressurized at 1.0 MPa as the primary coolant, water near the atmospheric pressure as the secondary coolant, and ambient air as the ultimate heat sink. Three accident scenarios simulated in the LTDF are discussed in this paper. In the first scenario, startup of the DRACS system from a cold state is simulated with no initial primary coolant flow. In the second scenario, a reactor coolant pump trip process is studied, during which a flow reversal phenomenon in the DRACS primary loop occurs. In the third scenario, the pump trip process is studied with a simulated intermediate heat exchanger in operation during the simulated core normal operation. In all scenarios, natural circulation flows are developed as the transients approach their quasi steady states, demonstrating the functionality of the DRACS. The accident scenarios in the prototypic FHR design corresponding to the simulated ones in the LTDF are also predicted by following a scaling-up process. The predictions show that at any time during the simulated transient, the salt temperatures will be higher than their melting temperatures and that therefore there will be no issue of salt freezing in the three projected accident scenarios. However, the scaled-up primary salt temperatures indicate that the prototypic DHX may have been undersized and may need to be redesigned.

The Direct Reactor Auxiliary Cooling System (DRACS) is a passive residual heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three natural circulation/convection loops that rely on buoyancy as the driving force and are coupled via two heat exchangers, namely, the DRACS heat exchanger and the natural draft heat exchanger. A fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during reactor normal operation, and to activate the DRACS in accidents when the reactor is shut down. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has been developed, which consists of a core scaling and a loop scaling. The consistency between the core and loop scaling is examined via the reference volume ratio, which can be obtained from both the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a scientific design of a scaled-down high-temperature DRACS test facility. (C) 2015 Elsevier B.V. All rights reserved.

Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity], a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O2 cannot be ignored, especially for the FHR, in which environment the product, SiO2, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.

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 principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

The Direct Reactor Auxiliary Cooling System (DRACS) is a passive heat removal system proposed for the Advanced High-Temperature Reactor (AHTR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops relying completely on buoyancy as the driving force. A fluidic diode has been proposed in the DRACS primary loop to maintain the passive feature. Fluidic diodes are passive flow control devices with low flow resistance in one direction and high flow resistance in the opposite direction. The fluidic diode is orientated such that during reactor normal operation the primary salt flow in the DRACS is restricted, thus preventing excessive heat loss from the reactor to the DRACS. However, when the DRACS is functioning during reactor accidents, the primary salt flow is in the forward flow direction of the diode that features low flow resistance. To investigate the reliability and thermal performance of the DRACS, a high-temperature DRACS test facility (HTDF) is being designed and constructed at The Ohio State University (OSU). In this HTDF, a conventional vortex diode has been proposed. In this paper, a detailed design process of the vortex diode for the HTDF is presented. Design parameters, such as the desired flow rates in and pressure drops across the fluidic diode, were first determined for both the forward and reverse flow directions, following which was the parametric CFD study of multiple vortex diodes with variant nozzle size, chamber size, and inlet flow rates. Flow structures inside the diode, and the effects of the nozzle size, chamber size, and Reynolds number on the Euler number were examined for both flow directions. Correlations of the forward and reverse Euler numbers and the diodicity were developed and used to develop a vortex diode design that would be applicable to the HTDF.

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.

This report documents the current status of fabrication and assembly planning for the magnetic bearing, canned rotor pump being used as a demonstration platform for deeply integrating I&C into nuclear power plant components. The report identifies material choices and fabrication sequences for all of the required parts and the issues that need to be either resolved or accommodated during the manufacturing process. Down selection between material options has not yet been performed. Potential suppliers for all of the necessary materials have also been identified. The assembly evaluation begins by logically subdividing the pump into modules, which are themselves decomposed into individual parts. Potential materials and fabrication processes for each part in turn are then evaluated. The evaluation process includes assessment of the environmental compatibility requirements and the tolerances available for the selected fabrication processes. A description of the pump power/control electronics is also provided. The report also includes exploded views of the modules that show the integration of the various parts into modules that are then assembled to form the pump. Emphasis has been placed on thermal environment compatibility and the part dimensional changes during heat-up. No insurmountable fabrication or assembly challenges have been identified.

Fluoride Salt-Cooled High-Temperature Reactors (FHRs) are a promising new class of thermal-spectrum nuclear reactors. The reactor structural materials must possess high-temperature strength and chemical compatibility with the liquid fluoride salt as well as with a power cycle fluid such as supercritical water while remaining resistant to residual air within the containment. Alloy N was developed for use with liquid fluoride salts and it possesses adequate strength and chemical compatibility up to about 700°C. A distinctive property of FHRs is that their maximum allowable coolant temperature is restricted by their structural alloy maximum service temperature. As the reactor thermal efficiency directly increases with the maximum coolant temperature, higher temperature resistant alloys are strongly desired. This paper reviews the current status of Alloy N and its relevance to FHRs including its design principles, development history, high temperature strength, environmental resistance, metallurgical stability, component manufacturability, ASME codification status, and reactor service requirements. The review will identify issues and provide guidance for improving the alloy properties or implementing engineering solutions.

The overall project objective is to demonstrate improved reliability and increased performance made possible by deeply embedding instrumentation and controls (I&C) in nuclear power plant components. The project is employing a highly instrumented canned rotor, magnetic bearing, fluoride salt pump as its I&C technology demonstration vehicle. The project s focus is not primarily on pump design, but instead is on methods to deeply embed I&C within a pump system. However, because the I&C is intimately part of the basic millisecond-by-millisecond functioning of the pump, the I&C design cannot proceed in isolation from the other aspects of the pump. The pump will not function if the characteristics of the I&C are not embedded within the design because the I&C enables performance of the basic function rather than merely monitoring quasi-stable performance. Traditionally, I&C has been incorporated in nuclear power plant (NPP) components after their design is nearly complete; adequate performance was obtained through over-design. This report describes the progress and status of the project and provides a conceptual design overview for the embedded I&C pump.

This interim report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for coating large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for coating inaccessible surfaces such as the interior surfaces of heat exchangers. The final project report will feature an experimental evaluation of the performance of the two selected cladding techniques.