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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Renewability and sustainability aspects of nuclear energy

Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, {sup 233}U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO{sub 2}/RG‐PuO{sub 2}) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG‐PuO{sub 2} + 96 % ThO{sub 2}; 6 % RG‐PuO{sub 2} + 94 % ThO{sub 2}; 10 % RG‐PuO{sub 2} + 90 % ThO{sub 2}; 20 % RG‐PuO{sub 2} + 80 % ThO{sub 2}; 30 % RG‐PuO{sub 2} + 70 % ThO{sub 2}, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ∼ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG‐PuO{sub 2} fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MW{sub th} has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ∼160 kg {sup 233}U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ∼1.3.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Permeation of hydrogen in hastelloy C-276 alloy at high temperature

Tritium is generated by the interaction of neutrons with the lithium and beryllium in the molten salt reactors (MSRs), which use FLiBe as one of solvents of fluoride fuel. Tritium as by-product in the MSRs is an important safety issue because it could easily diffuse into environment through high temperature heat exchangers. The experimental technique of gas driven permeation has been used to investigate the transport parameter of hydrogen in Hastelloy C-276 which is considered as one of the candidate for structure materials. The measurements were carried out at the temperature range of 400-800 Celsius degrees with hydrogen loading pressures ranging from 5*10{sup 3} to 4*10{sup 4} Pa. The H diffusive transport parameters for Hastelloy C-276 follow an Arrhenius law in this temperature range. Regarding diffusivity and Sieverts' constant, Hastelloy C-276 has lower values compared with Ni201 alloy. The possible reason may be the trapping effects, which were formed by the alloying elements of Mo and Cr in the matrix. At the same time, the thin oxidation layer formed by the high Cr content could lead to a slower dissociation process of H{sub 2} at the surface. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated at a system power level of 2000 MW th , took about 3.5 years to reach full plateau power, and was capable of an End of Plateau burnup of 38.7 %FIMA if considering just the neutronic constraints in the system design; however, fuel performance constraints led to a maximum credible burnup of 12.1 %FIMA due to a combination of internal gas pressure and irradiation effects on the TRISO materials (especially PyC) leading to SiC pressure vessel failures. The optimal neutron spectrum for the thorium-fueled blanket options evaluated seemed to favor a hard spectrum (low but non-zero neutron multiplier thicknesses and high TRISO packing fractions) in terms of neutronic performance but the fuel performance constraints demonstrated that a significantly softer spectrum would be needed to decrease the rate of accumulation of fast neutron fluence in order to improve the maximum credible burnup the system could achieve.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF 2 ) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Diffusion, Thermal Properties and Chemical Compatibilities of Select MAX Phases with Materials For Advanced Nuclear Systems

The demands of Gen IV nuclear power plants for long service life under neutron irradiation at high temperature are severe. Advanced materials that would withstand high temperatures (up to 1000+ ºC) to high doses in a neutron field would be ideal for reactor internal structures and would add to the long service life and reliability of the reactors. The objective of this work is to investigate the chemical compatibility of select MAX with potential materials that are important for nuclear energy, as well as to measure the thermal transport properties as a function of neutron irradiation. The chemical counterparts chosen for this work are: pyrolytic carbon, SiC, U, Pd, FLiBe, Pb-Bi and Na, the latter 3 in the molten state. The thermal conductivities and heat capacities of non-irradiated MAX phases will be measured.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Fluoride-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model

In this report we present work performed in Fiscal Year 2022 that demonstrates the modeling and simulation of a fully coupled neutronics thermal hydraulics reference plant model for a fluoride-cooled high-temperature pebble-bed reactor. The multiphysics model is developed on the Nuclear Regulatory Commission’s Comprehensive Reactor Analysis Bundle (BlueCRAB) available on the Idaho National Laboratory’s high-performance computer, which natively and seamlessly couples Griffin, Pronghorn, and the BISON Multiphysics Object-Oriented Simulation Environment based applications. Griffin provides reactor physics capabilities, including depletion to the equilibrium core, k-eigenvalue, adjoint, and transient. The unique direct equilibrium core capability in Griffin is based on a streamline methodology to spatially deplete the pebbles into burnup groups. Pronghorn solves the porous medium equations for the fluid regions and conduction in the solid regions and incorporates a fluidic diode model to simulate the transition from forced to natural convection during accident scenarios. MOOSE modules solves thermal conduction problems for the pebbles and tristructural isotropic in the pebble-bed core, thus providing the fuel and moderator spatial fields for each pebble burnup group. The neutronics feedback relies primarily on fuel, moderator, and reflector temperatures as as well as the FLiBe salt density. Here, we present results for the uncoupled equilibrium core and perform comparisons to equivalent Monte Carlo models. The power distributions and kinetic parameters obtained with Griffin are consistent with those computed with Griffin. We demonstrate a noticeable improvement with the use of discrete ordinates method (SN) transport. The coupled steady-state equilibrium core provides the initial condition for two time-dependent problems: a control rod withdrawal event and an unprotected loss of flow event. In both cases, the reactor design is self-stabilizing and the solutions are consistent with the expected physics. Although this model is prototypical regarding BlueCRAB’s capabilities, its results are consistent with published work by Kairos Power and other research entities. Significant improvements to the model are planned in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Modeling of fission and activation products in molten salt reactors and their potential impact on the radionuclide monitoring stations of the International Monitoring System

Molten Salt Reactors (MSRs) are one of six Generation IV reactor designs currently under development around the world. Because of the unique operating conditions of MSRs, which include molten fuel and the continuous removal of gaseous fission products during operation, work was performed to analyze the potential impact of emissions on the International Monitoring System (IMS) of the Comprehensive Nuclear-Test-Ban Treaty (CTBT). Simulations were performed to predict the production of IMS-relevant radionuclides in four MSR designs operating under two scenarios – (1) a sealed reactor with releases only during operational shutdown and (2) continuous reprocessing or sparging of the fuel salt. From these production estimates the radioxenon and radioiodine signatures were extracted and compared to three current reactor designs (PWR, BWR, RBMK). In the cases where continuous reprocessing of the fuel salt occurred both the radioxenon and radioiodine signatures are nearly indistinguishable from a nuclear explosion. Estimates were also made of the potential emission rate of radioxenon for three reactor designs and it was found that MSRs have the potential to emit radioxenon isotopes at a rate of 1015 – 8×1016 Bq/d for 133Xe if no abatement is used. An assessment was also made of activation products using a candidate fuel salt (FLiBe) mixed with corrosion products for the Thorium Molten Salt Reactor (TMSR-LF1)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of {sup 7}LiF and BeF{sub 2} (FLiBe) possessing a boiling point above 1300 C and the figure of merit {rho}C{sub p} (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Measurement of neutron spectra in the experimental reactor LR-0

The measurement of fast neutron fluxes is important in many areas of nuclear technology. It affects the stability of the reactor structural components, performance of fuel, and also the fuel manner. The experiments performed at the LR-0 reactor were in the past focused on the measurement of neutron field far from the core, in reactor pressure vessel simulator or in biological shielding simulator. In the present the measurement in closer regions to core became more important, especially measurements in structural components like reactor baffle. This importance increases with both reactor power increase and also long term operation. Other important task is an increasing need for the measurement close to the fuel. The spectra near the fuel are aimed due to the planned measurements with the FLIBE salt, in FHR / MSR research, where one of the task is the measurement of the neutron spectra in it. In both types of experiments there is strong demand for high working count rate. The high count rate is caused mainly by high gamma background and by high fluxes. The fluxes in core or in its vicinity are relatively high to ensure safe reactor operation. This request is met in the digital spectroscopic apparatus. All experiments were realized in the LR-0 reactor. It is an extremely flexible light water zero-power research reactor, operated by the Research Center Rez (Czech Republic). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Progress in Development of Fluoride-salt-cooled High-Temperature Reactors (FHRs)

Fluoride-salt-cooled High-temperature Reactors (FHRs) are a new class of reactors that use graphite-matrix coated-particle fuel with a clean liquid salt coolant. The largest university FHR program is an Integrated Research Project (IRP) sponsored by the U.S. Department of Energy and others that is led by the Massachusetts Institute of Technology (MIT) and includes the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The year 2016 was the fifth year of this program. We report herein some of the major results of work done in 2016. The base-line FHR concept studied in this IRP uses (1) graphite-matrix coated-particle fuel with a pebble bed geometry and (2) a clean liquid salt coolant-flibe (a mixture of {sup 7}LiF and BeF{sub 2}). The combination of a very high-temperature fuel with failure temperatures above 1650 deg. C, a coolant boiling point above 1400 deg. C, and a coolant that traps most fission products implies the capability to assure no major fuel failure under severe accident conditions. There are a family of FHR concepts using different fuel geometries and different fluoride salts. The power conversion system is a Nuclear Air-Brayton Combined Cycle (NACC) with peak power provided by natural gas or a Firebrick Resistance Heated Energy Storage (FIRES) system. This enables a base-load reactor to provide variable electricity to the grid that substantially increases revenue relative to base-load nuclear reactors. Most of the results reported herein are applicable to most FHR concepts as well as liquid-fueled molten salt reactors (MSRs)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Coupled CFD-DEM Analysis of Molten Salt-Cooled Pebble-Bed Reactor Experiment

The development of new nuclear reactor designs is an area of ongoing investigation, and a focus of dedicated work for various laboratories around the world. Despite the steadfast and reliable service of Light Water Reactors (LWR), the ideal power reactor, with a combination of high operational temperature, passive safety, proliferation resistance, economic build and waste minimization remains elusive. One promising design that might address these diverse, and at times contradictory requirements, is the Molten Salt Reactor (MSR). In its most radical form, a liquid fueled (LF) MSR has the nuclear fuel dissolved in the molten-salt primary coolant itself, as demonstrated by Oak Ridge National Lab (ORNL) in 1965 to 1969. An alternative fuel arrangement is to retain conventional, solid fuel elements. A proposed method is to implement pebble fuel elements, as have been used in previous gas-cooled reactors, with the gas coolant being replaced with a molten salt. This is referred to as pebble-bed, molten-salt reactor (PB-MSR). The spherical fuel pebbles used within PB-MSRs contain TRISO particles dispersed throughout the pebble's graphite matrix. Many pebbles are constrained to form a porous bed within the core region of a reactor. Coolant molten-salt passes through the pebble bed to remove heat produced due to the nuclear fission taking place within the fuel pebbles' TRISO particles. Various benefits are apparent with the PB-MSR design. When considering the choice of fuel elements, they are a geometry that has been produced and utilized in a similar way in the past, within gas-cooled reactors. The fuel pebbles are also able to be added to, and removed from, the core during operation, enabling on-line refueling. When considering the coolant choice, molten-salts have much higher boiling points (1703 K at 1 atm. for FLiBe) than conventional coolants such as water (630 K at 150 atm.), and are capable of reaching these temperatures at atmospheric pressure. This in turn removes the need for high pressure containment that is present in light water reactors and gas-cooled reactors. The future implementation of new reactors, and the search for improvements in efficiency and safety in current designs, require the advancement of simulation tools, which this work ultimately aims to address.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Improved Heat Transfer and Volume Scaling through Novel Heater Design

In 2014, the University of California, Berkeley, Nuclear Engineering Thermal Hydraulics Laboratory completed construction of the Compact Integral Effects Test (CIET) facility, a scaled thermal hydraulics scaled integral effects test facility designed to model the Mark-1 pebble-bed fluoride-salt-cooled high-temperature reactor (Mk1 PB-FHR or simply Mk1). CIET uses an organic simulant fluid, Dowtherm A, to match the Prandtl, Reynolds, Froude, and Grashof numbers of important fluoride salt candidates for molten salt reactors, such as flibe. The FHR reactor core is simulated using a resistively heated tube, which heats the oil flowing through the annulus between the heated tube and a non-heated inner tube. However, the volume fraction of the coolant in the CIET heater was dramatically lower than the volume fraction of the primary coolant in the Mk1 PB-FHR design (3% in CIET vs. 17% in Mk1) because the CIET heater was initially scaled to match an earlier FHR design. Also, the maximum core heater power in CIET is limited by the melting point of Teflon components in CIET, approximately 330 deg. C. For conservatism, the CIET heater power is limited to keep the maximum surface temperature of the heater below 250 deg. C. At the nominal mass flow rate of 0.18 kg/s, this power was about 5 kW. A higher heat transfer coefficient from the heater wall to the oil was needed to increase the effective power delivered to the oil. In order to alleviate these two limitations, a new heater insert was designed and constructed for CIET. This new heater insert uses a twisted tape to augment heat transfer. The twisted tape is supported inside a perforated inner tube that has the same diameter as the original inner tube of the heater, in order to make it compatible with CIET's original heater design. The heat transfer and pressure drop properties of the new heater insert designs were tested in CIET and compared to the previous design in terms of pressure drop and convective heat transfer coefficient. The new heater inserts are intended to increase the range of power levels and mass flow rates that can be tested in CIET in order to augment understanding of FHR thermal hydraulic phenomena. Also, obtaining different data points for CIET response dynamics under different heater coolant residence times will be valuable in evaluating the Mk1 design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Neutronic design and fuel cycle analysis of a fluoride salt-cooled High Temperature Reactor (FHR)

The Fluoride salt-cooled High temperature Reactor (FHR) is a new Gen IV reactor concept that can operate under near atmospheric pressure circumstances and further enhance the inherent safety. In this study, an FHR core design with 165 megawatts of thermal output (MWth) is proposed. The reactor core employs tri-structural-isotropic (TRISO) particle fuel within prismatic graphite blocks as the basic fuel form, FLiBe (2 {sup 7}LiF-BeF{sub 2}) as the primary coolant, and a three-batch fuel cycle scheme. Sensitivity analyses on various parameters were performed to optimize the cycle length and neutronic parameters. The fuel cycle of this core design was evaluated in detail from 4 aspects: cycle length, power peaking factor (PPF), discharge burnup, and temperature coefficient. It was found that larger fuel channel pitch (FCP) would have a relatively harder neutron spectrum, and yield relatively longer cycle length, lower, better fuel temperature coefficient (FTC), and moderator temperature coefficient (MTC). In addition, burnable poison (BP), Er{sub 2}O{sub 3}, can effectively reduce PPF, hold down the multiplication factor, and more importantly it can improve the MTC. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

SiC compatibility

SiC is a relatively chemically stable ceramic that has potential use in the extreme environments present in nuclear reactors. Reactions with high temperature steam, He, hydrothermal, molten salts, and liquid metals are some of the possible environments that SiC could be exposed to. While it is very stable in air, aqueous environments and very low oxygen partial pressures can lead to SiC recession. Generally, SiC does not react with the FliNaK or FliBe molten salts, or liquid metals, however contaminants in these working fluids leached from other materials can cause reaction layers to occur. Most of these systems require additional testing under neutron irradiation to determine if there are any combined effects.

Mouche, Peter↗
DOE OSTI2017

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Core design and safety analysis

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor II (EBR-II) for sodium-cooled fast reactors. Engineering demonstrations have historically played a vital role in advancing the technology readiness level of reactor technologies. This paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Core design characteristics, fuel cycle performance, and safety analysis of the FHR DR preconcept have been evaluated. The FHR DR core design features a negative or negligible void coefficient throughout a reactor operating cycle. Both single-batch (cartridge) and multiple-batch fuel cycles can be demonstrated in the FHR DR. The single-batch cycle length of the FHR DR core is estimated at between 12-18 months, assuming the successful qualification of composite carbon (C/C) or silicon carbide (SiC/SiC) structural fuel block tie rod material. Fuel cycle performance of the FHR DR is similar to a high temperature gas-cooled reactor. Preliminary safety analysis of the FHR DR indicates that the reactor could be used to demonstrate the inherent safety characteristics of FHR designs.

Brown, Nicholas R.↗
DOE OSTI2018

Phenomena Identification and Ranking Table (PIRT) study for metallic structural materials for advanced High-Temperature reactor

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. Futhermore, 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. As a result, this paper gives a summary of important results from the PIRT panel discussions and report.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Experimental investigation of unsteady flow dynamics in therminol-66 using edge tone-induced oscillations

Recent research has demonstrated that flow oscillations induced by a wedge positioned opposite the jet outlet in edge tones can be effectively utilized for flow measurement purposes. This highlights the necessity for developing advanced instrumentation capable of precise flow measurements in experimental setups exposed to radiation and in challenging fluids such as molten salts. Here, this study contributes to the existing body of knowledge by conducting experiments using Therminol-66, an effective surrogate for FLiBe under appropriate temperature conditions for nuclear applications, across a broad range of Reynolds numbers from 17,700 to 555,000. Experiments were conducted with Therminol-66 at approximately 368.8 K under atmospheric pressure, with varying flow rates. A reliable correlation was established between the generated frequencies, corresponding flow rates, and device geometry parameters. The data were subjected to non-dimensional analysis to reveal the interrelationships among flow rate, fluid properties, device geometry, and their impact on the acoustic tones produced by the edge tone device. These findings contribute to the development of precise flow measurement instrumentation for use in challenging environments.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗