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 liquid fluoride salt experiment has been constructed and used to acquire natural convection heat transfer data. The experiment used FLiNaK salt in a small cell that included a cylindrical electrical heater, 1.27 cm in diameter, oriented vertically in a FLiNaK bath. Thermocouples internal to the heater were used, along with FLiNaK temperature measurements and heater power measurements, to determine natural circulation heat transfer coefficients. These data were acquired for Rayleigh numbers ranging from 3 × 105 to 8 × 109 and salt temperatures from 560 °C to 640 °C. Test results show that measured heat transfer coefficients are consistent with conventional natural convection heat transfer coefficients for cylinders, but the average error can be as high as 20% using these correlations. Correlations developed by fitting the data for this experiment show much lower errors (<10%).

This study identifies the phenomena that impact the accuracy of the multiphysics analysis of fluoride salt-cooled high temperature reactors (FHR) and categorizes the level of coupling needed to model these phenomena with sufficient accuracy in the context of three scenarios, namely, normal operation, station blackout, and simultaneous withdrawal of all control rods. Such a study is a necessary step prior to developing or extending existing multiphysics modeling and simulation tools. The identified multi physics phenomena are categorized into the groups requiring "tight" and "loose" coupling. "Tight" coupling refers to phenomena that require iterative feedback between multiple physics modules/methods whereas "loose" coupling refers to phenomena that require only data from other physics that can be pre-calculated and shared in tabular or functional form. For the station blackout scenario, it was determined that no identified phenomena required tight coupling. For the rod withdrawal and normal operation scenarios, it is found that there is a large number of phenomena that require tight coupling between neutronics and thermal hydraulics. (C) 2018 Elsevier Ltd. All rights reserved.

The Advanced High Temperature Reactor (AHTR) is a conceptual design of a Fluoride salt-cooled High temperature Reactor (FHR) utilizing (LiF)-Li-7-BeF2 (66-34 mol%) as its primary coolant. To identify key phenomena that impose potential challenges on thermal hydraulics modeling and simulation of such a reactor system, a thermal hydraulics phenomena identification and ranking table (TH-PIRT) study was performed for the AHTR in a Department of Energy Nuclear Energy University Program (NEUP) integrated research project led by Georgia Institute of Technology. A panel of experts from regulators, industries, national laboratories, and academia was assembled for the study. In this paper, the TH-PIRT5 identified by the panel for two events, namely, station blackout and simultaneous withdrawal of all control rods, are summarized and discussed in detail. In addition, the key phenomena that warrant further study and research for AHTR analysis are identified to support the validation of thermal hydraulics system-level analysis codes and computational fluid dynamics simulation tools, as well as future FHR reactor licensing. Crown Copyright (C) 2018 Published by Elsevier Ltd. All rights reserved.

Tritium production rate in Fluoride salt-cooled High-temperature Reactors (FHRs) was estimated to be several orders of magnitude higher than that in Light Water Reactors (LWRs). Due to the high permeability of tritium at elevated temperatures, a double-wall heat exchanger design consisting of inner and outer tubes was proposed to significantly reduce the tritium permeation through the heat transfer surfaces to, ultimately, the environment. A coupled heat transfer and tritium mass transport model was developed for performance analysis of a double-wall Natural Draft Heat Exchanger (NDHX) design. Since there was no published experimental data available in the literature involving both heat transfer and mass transport simultaneously, these two sub-models, i.e., heat transfer sub-model and mass transport sub model, were benchmarked against available experimental data separately. For the heat transfer sub model, the discrepancies for the predicted temperatures and heat transfer coefficients compared with their individual experimental data are within 16% and 24%, respectively. For the mass transport sub model, the relative discrepancies between the model predictions and the experimental data are 23-44% at temperatures from 700 to 1000 degrees C (23-35% at the salt temperatures from 700 to 800 degrees C, between which the maximum salt temperature is expected in FHRs). This coupled heat transfer and mass transport model was then used to analyze a double-wall NDHX design for the Advanced High-Temperature Reactor (AHTR), a pre-conceptual FHR design developed by the Oak Ridge National Laboratory, from the following four tube configurations: 1) inner plain tube with outer plain tube (IPOP); 2) inner plain tube with outer fluted tube (IPOF); 3) inner fluted tube with outer plain tube (IFOP); and 4) inner fluted tube with outer fluted tube (IFOF). The results show that for the heat transfer performance, the IFOF design is slightly superior to the IPOF design and that both are significantly superior to the IFOP and IPOP designs. For the mass transport performance, the IFOP design is slightly superior to the IFOF design, and both significantly over perform the IPOP and IPOF designs. In addition, Non-dominated Sorting in Generic Algorithms (NSGA) was applied for the design optimization of a potential NDHX with the IFOF configuration for AHTR. (C) 2018 Elsevier Ltd. All rights reserved.

Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system proposed for Fluoride-salt-cooled High-temperature Reactors (FHRs). Benchmark study of decay heat removal capability of DRACS by natural circulation is critical for FHR safety analysis. Two scaled-down test facilities, a low-temperature DRACS test facility (LTDF) and a high-temperature DRACS test facility (HTDF), were designed following a scaling analysis and constructed to investigate thermal performance and heat removal capability of the DRACS. The LTDF uses water as a surrogate coolant for both the primary and DRACS loops. Two transient scenarios were carried out in the LTDF, namely, DRACS startup and primary coolant pump trip. In both of the scenarios, the experimental results demonstrate sufficient heat removal capabilities of the DRACS with natural circulations established in the system. These two scenarios were simulated using RELAP5/SCDASIM/MOD4.0 code and the code simulation results show good agreement with the LTDF experimental data. For the HTDF, molten salts FLiNaK and KF-ZrF4 are used respectively as the primary coolant and DRACS coolant. Thermodynamic and transport properties of FLiNaK and KF-ZrF4 were implemented into RELAP5/SCDAPSIM/MOD 4.0. Simulations of transient scenarios in the HTDF were also performed and the code results will be compared with the HTDF experimental data once available.

The Spallation Neutron Source (SNS) is a high-power linear accelerator built at Oak Ridge National Laboratory (ORNL) which incorporates the use of a flowing liquid mercury target. The Mercury Thermal Hydraulic Loop (MTHL) was constructed to investigate and verify the heat transfer characteristics of liquid mercury in a rectangular channel. This report provides a compilation of previously reported results from the water-cooled and electrically heated straight and curved test sections that simulate the geometry of the window cooling channel in the target nose region.

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.

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 Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor. (C) 2015 Elsevier B.V. All rights reserved.

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.

Fluoride salt-cooled high-temperature reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics and fully passive safety. This paper provides an overview of a technology development pathway for expeditious commercial deployment of first-generation FHRs. The paper describes the principal remaining FHR technology challenges and the development path needed to address the challenges. First-generation FHRs do not appear to require any technology breakthroughs, but will require significant technology development and demonstration. FHRs are currently entering early phase engineering development. As such, the development roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant; the lack of an approved licensing framework; the lack of qualified, salt-compatible structural materials; and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental results show similar trends as the computational fluid dynamics (CFD) results presented in this report; however, some differences exist that will need to be assessed in future studies. The results of this testing will be used to improve the diode design to be tested in the liquid salt loop system.

A study on the Fukushima Daiichi nuclear power station spent-fuel pool (SFP) at Unit 4 (SFP4) is presented in this paper. We discuss the design characteristics of SFP4 and its decay heat load in detail and provide a model that we developed to estimate the SFP evaporation rate based on the SFP temperature. The SFP level of SFP4 following the March 11, 2011, accident is predicted based on the fundamental conservation laws of mass and energy. Our predicted SFP level and temperatures are in good agreement with measured data and are consistent with Tokyo Electric Power Company evaluation results.