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Charles W Forsberg - One of the best experts on this subject based on the ideXlab platform.
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the advanced high Temperature Reactor high Temperature fuel liquid salt coolant liquid metal Reactor plant
Progress in Nuclear Energy, 2005Co-Authors: Charles W ForsbergAbstract:The Advanced High-Temperature Reactor is a new Reactor concept that combines four existing technologies in a new way: (1) coated-particle graphite-matrix nuclear fuels (traditionally used for helium-cooled Reactors), (2) Brayton power cycles, (3) passive safety systems and plant designs from liquid-metal-cooled fast Reactors, and (4) low-pressure liquid-salt coolants with boiling points far above the maximum coolant Temperature. The new combination of technologies enables the design of a large [2400- to 4000-MW(t)] high-Temperature Reactor, with Reactor-coolant exit Temperatures between 700 and 1000°C (depending upon goals) and passive safety systems for economic production of electricity or hydrogen. The AHTR [2400-MW(t)] capital costs have been estimated to be 49 to 61% per kilowatt (electric) relative to modular gas-cooled [600-MW(t)] and modular liquid-metal-cooled Reactors [1000-MW(t)], assuming a single AHTR and multiple modular units with the same total electrical output. Because of the similar fuel, core design, and power cycles, about 70% of the required research is shared with that for high-Temperature gas-cooled Reactors.
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molten salt cooled advanced high Temperature Reactor for production of hydrogen and electricity
Nuclear Technology, 2003Co-Authors: Charles W Forsberg, Per F Peterson, Paul S PickardAbstract:The molten-salt-cooled Advanced High-Temperature Reactor (AHTR) is a new Reactor concept designed to provide very high-Temperature (750 to 1000°C) heat to enable efficient low-cost thermochemical p...
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hydrogen nuclear energy and the advanced high Temperature Reactor
International Journal of Hydrogen Energy, 2003Co-Authors: Charles W ForsbergAbstract:Abstract Nuclear energy has been proposed as an energy source to produce hydrogen (H2) from water. An examination of systems issues in this paper indicates that the infrastructure of H2 consumption is now compatible with the production of H2 by nuclear Reactors. Alternative H2 production processes were examined to define the requirements such processes would impose on the nuclear Reactor. These requirements include supplying heat at a near-constant high Temperature, providing a low-pressure interface with the H2 production processes, isolating the nuclear plant from the chemical plant, and avoiding tritium contamination of the H2 product. A Reactor concept—the advanced high-Temperature Reactor—was developed to match these requirements for H2 production.
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advanced high Temperature Reactor for production of electricity and hydrogen molten salt coolant graphite coated particle fuel
Other Information: PBD: 21 Feb 2002, 2002Co-Authors: Charles W ForsbergAbstract:The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high Temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-Temperature gas-cooled Reactors (MHTGRs), such as the General Atomics gas turbine-modular helium Reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal Reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced Reactor concept that operates at essentially ambient pressures and at very high Temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large Reactor sizes [{approx}1000 Mw(e)] with passive safety systems to provide the potential for improved economics.
David Eugene Holcomb - One of the best experts on this subject based on the ideXlab platform.
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thermal hydraulics analysis of the advanced high Temperature Reactor
Nuclear Engineering and Design, 2015Co-Authors: Dean Wang, Graydon L. Yoder, David W Pointer, David Eugene HolcombAbstract:Abstract 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.
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an experimental test facility to support development of the fluoride salt cooled high Temperature Reactor
Annals of Nuclear Energy, 2014Co-Authors: Adam M Aaron, David Eugene Holcomb, Kevin R Robb, Burns Cunningham, David Fugate, Roger A Kisner, Fred J Peretz, J B Wilgen, Dane F WilsonAbstract:Abstract The need for high-Temperature (greater than 600 °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 °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.
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AHTR Refueling Systems and Process Description
2012Co-Authors: Venugopal Koikal Varma, David Eugene Holcomb, Eric Craig Bradley, Nathaniel M Zaharia, Eliott J CooperAbstract:The Advanced High-Temperature Reactor (AHTR) is a design concept for a central station-type [1500 MW(e)] Fluoride salt–cooled High-Temperature Reactor (FHR) that is currently undergoing development by Oak Ridge National Laboratory for the US. Department of Energy, Office of Nuclear Energy’s Advanced Reactor Concepts program. FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-Temperature power cycle, and fully passive decay heat rejection. The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. The AHTR is approaching a preconceptual level of maturity. An initial integrated layout of its major systems, structures, and components (SSCs), and an initial, high-level sequence of operations necessary for constructing and operating the plant is nearing completion. An overview of the current status of the AHTR concept has been recently published [1], and a report providing a more detailed overview of the AHTR structures and mechanical systems is currently in preparation. This report documents the refueling components and processes envisioned at this early development phase. The report is limited to the refueling aspects of the AHTR and does not include overall Reactor or power plant design information. The report, however, does include a description of the materials envisioned for the various components and the instrumentation necessary to control the refueling process. The report begins with an overview of the refueling strategy. Next a mechanical description of the AHTR fuel assemblies and core is provided. The Reactor vessel upper assemblies are then described. Following this the refueling path structures and the refueling mechanisms and components are described. The sequence of operations necessary to fuel and defuel the Reactor is then discussed. The report concludes with a discussion of the levels of maturity of the various SSCs to provide guidance for future technology developments. The conceptual design information presented in this report is very preliminary in nature. Significant uncertainty remains about several aspects of the process and even the radiation and mechanical performance of plate-type coated-particle fuel
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core and refueling design studies for the advanced high Temperature Reactor
2011Co-Authors: David Eugene Holcomb, Dan Ilas, Anselmo T Cisneros, Venugopal Koikal Varma, Ryan P Kelly, Jess C GehinAbstract:The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-Temperature Reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, Reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The Reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete Reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas Reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure drop across the core was restricted to no more than 1.5 atm during normal operation to minimize the upward force on the core. Also, the flow velocity in the core was restricted to 3 m/s to minimize erosion of the fuel plates. Section 3.1.1 of this report discusses the design restrictions in more detail.
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pre conceptual design of a fluoride salt cooled small modular advanced high Temperature Reactor smahtr
2011Co-Authors: S R Greene, Jess C Gehin, David Eugene Holcomb, Juan J Carbajo, Dan Ilas, Anselmo T Cisneros, Venugopal Koikal Varma, W R Corwin, Dane F Wilson, Graydon L. YoderAbstract:This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high Temperature Reactors (FHRs). A preliminary Reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-Temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The Reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal Reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.
Per F Peterson - One of the best experts on this subject based on the ideXlab platform.
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NEUTRONIC DESIGN OF THE PB-AHTR
2015Co-Authors: Massimiliano Fratoni, Ehud Greenspan, Per F PetersonAbstract:This study (1) investigates the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR); (2) compares the PB-AHTR neutronic performance against those of the alternative design options of He-cooled high Temperature Reactors using either pebble (PBMR) or prismatic (VHTR) fuel and against a liquid-salt cooled design version of the latter (LS-VHTR); (3) studies the possibility of incinerating TRU in the PB-AHTR feeding TRU from LWRs spent fuel. It is found that the optimal features a graphite-to-heavy metal ratio of ~360 and its reactivity coefficients are all negative. A comparison with the helium-cooled pebble-bed Reactor and with a prismatic-fuel Reactor that is cooled with either flibe or helium is also presented. It is found that the PB-AHTR offers similar discharge burnup as the other three designs. As compared to the gas-cooled pebble bed, the PB-AHTR uranium loading and energy generated per pebble are ~2.5 times higher. When loading TRU in the pebbles it is found that they can reach a burnup as high as 685 GWd/tHM and the core average reactivity coefficients are all negative. About 70 % of the initial load of HM is incinerated in a single pass
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neutronic and depletion analysis of the pb ahtr
2007Co-Authors: Massimiliano Fratoni, Ehud Greenspan, Per F PetersonAbstract:The PB-AHTR is a Pebble Bed Advanced High Temperature Reactor that is cooled with the liquid salt flibe (LiF-BeF{sub 2}) rather than helium. This study presents a preliminary neutronic and depletion analysis for the PBAHTR. The attainable burnup is determined as a function of uranium loading per pebble, power density and core dimensions. It is found that the optimal design for a 425 {mu}m UC{sub 0.5}O{sub 1.5} fuel kernel diameter, 10% enriched uranium, features a graphite-to-heavy metal ratio of {approx}360 and its reactivity coefficients are all negative. A comparison with the helium-cooled pebble-bed Reactor and with a prismatic-fuel Reactor that is cooled with either flibe or helium is also presented. It is found that the PB-AHTR offers similar discharge burnup as the other three designs. As compared to the gas-cooled pebble bed, the PB-AHTR uranium loading and energy generated per pebble are {approx}2.5 times higher. (authors)
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advanced csic composites for high Temperature nuclear heat transport with helium molten salts and sulphur iodine thermochemical hydrogen process fluids
Nuclear Science, 2004Co-Authors: Charles Forsberg, Per F Peterson, Paul S PickardAbstract:This paper discusses the use of liquid-silicon-impregnated (LSI) carbon-carbon composites for the development of compact and inexpensive heat exchangers, piping, vessels and pumps capable of operating in the Temperature range of 800 to 1100°C with high-pressure helium, molten fluoride salts, and process fluids for sulfur-iodine thermochemical hydrogen production. LSI composites have several potentially attractive features, including ability to maintain nearly full mechanical strength to Temperatures approaching 1400°C, inexpensive and commercially available fabrication materials, and the capability for simple forming, machining and joining of carbon-carbon performs, which permits the fabrication of highly complex component geometries. In the near term, these materials may prove to be attractive for use with a molten-salt intermediate loop for the demonstration of hydrogen production with a gas-cooled high Temperature Reactor. In the longer term, these materials could be attractive for use with the moltensalt cooled Advanced High Temperature Reactor, molten salt Reactors, and fusion power plants.
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molten salt cooled advanced high Temperature Reactor for production of hydrogen and electricity
Nuclear Technology, 2003Co-Authors: Charles W Forsberg, Per F Peterson, Paul S PickardAbstract:The molten-salt-cooled Advanced High-Temperature Reactor (AHTR) is a new Reactor concept designed to provide very high-Temperature (750 to 1000°C) heat to enable efficient low-cost thermochemical p...
Graydon L. Yoder - One of the best experts on this subject based on the ideXlab platform.
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a coupled heat transfer and tritium mass transport model for a double wall heat exchanger design for fhrs
Annals of Nuclear Energy, 2018Co-Authors: Sheng Zhang, Xiao Wu, Richard N Christensen, Graydon L. YoderAbstract:Abstract 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 °C (23–35% at the salt Temperatures from 700 to 800 °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.
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thermal hydraulics analysis of the advanced high Temperature Reactor
Nuclear Engineering and Design, 2015Co-Authors: Dean Wang, Graydon L. Yoder, David W Pointer, David Eugene HolcombAbstract:Abstract 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.
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pre conceptual design of a fluoride salt cooled small modular advanced high Temperature Reactor smahtr
2011Co-Authors: S R Greene, Jess C Gehin, David Eugene Holcomb, Juan J Carbajo, Dan Ilas, Anselmo T Cisneros, Venugopal Koikal Varma, W R Corwin, Dane F Wilson, Graydon L. YoderAbstract:This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high Temperature Reactors (FHRs). A preliminary Reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-Temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The Reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal Reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.
Dane F Wilson - One of the best experts on this subject based on the ideXlab platform.
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phenomena identification and ranking table pirt study for metallic structural materials for advanced high Temperature Reactor
Annals of Nuclear Energy, 2019Co-Authors: Preet M Singh, Dane F Wilson, Kevin J Chan, Chaitanya Deo, Vinay P Deodeshmukh, James R Keiser, Weiju Ren, T L Sham, Jinsuo ZhangAbstract:Abstract 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.
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an experimental test facility to support development of the fluoride salt cooled high Temperature Reactor
Annals of Nuclear Energy, 2014Co-Authors: Adam M Aaron, David Eugene Holcomb, Kevin R Robb, Burns Cunningham, David Fugate, Roger A Kisner, Fred J Peretz, J B Wilgen, Dane F WilsonAbstract:Abstract The need for high-Temperature (greater than 600 °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 °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.
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pre conceptual design of a fluoride salt cooled small modular advanced high Temperature Reactor smahtr
2011Co-Authors: S R Greene, Jess C Gehin, David Eugene Holcomb, Juan J Carbajo, Dan Ilas, Anselmo T Cisneros, Venugopal Koikal Varma, W R Corwin, Dane F Wilson, Graydon L. YoderAbstract:This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high Temperature Reactors (FHRs). A preliminary Reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-Temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The Reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal Reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.