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!ONFERENCE PROCEEDINGS Knoxville, TN 2020, April 6th – 9th Track 2: Nuclear Fission Power and Propulsion Technical Track Chair: Paolo Venneri Author Title Abou‐Jaoude, Abdalla PARAMETRIC EVALUATION OF ALTERNATIVE NUCLEAR PROPULSION CORES USING CURVED FUEL Angh, Caen DEVELOPMENT OF ZIRCONIUM CARBIDE FOR INERT MATRIX FUELS Cendro, Samantha simulation and experimental validation of aN inductively HeATED Solid‐core nuclear thermal rocket Chaiken, Max HEAT PIPE DEVELOPMENT FOR SPACE FISSION DEMONSTRATION MISSIONS Colgan, Nathan A HEAT EXCHANGER FOR HTGR WASTE HEAT REJECTION TO MARTIAN ATMOSPHERE INTERFACE AND SUBSURFACE CERAMIC BEHAVIOR IN MOLYBDENUM CERMETS FOR NUCLEAR Duffin, Taylor THERMAL PROPULSION PRE‐IRRADIATION CHARACTERIZATION OF INSTRUMENT PERFORMANCE FOR NUCLEAR THERMAL Floyd, Dan ROCKETS: TEST PLAN IRRADIATION TESTING OF MOLYBDENUM AND TUNGSTEN BASED CERMETS FOR USE IN NUCLEAR Gaffin, Neal THERMAL PROPULSION Hashkins, Justin Toward an in‐depth material model for Nuclear thermal propulsion fuel elements Hirschhorn, Jacob DEVELOPMENT OF A MULTISCALE MODEL FOR FUEL LOSS FROM NUCLEAR THERMAL PROPULSION Design of the In‐pile experiment Set (INSET) apparatus to support Nuclear Thermal Propulsion fuel and Howard, Richard component testing. Howe, Steve SPRINTR: ADVANTAGES OF FLAT PLATE VERSUS PRISMATIC NTR FUELS Activation analysis of subscale experimental testbed: towards simulating nuclear thermal propulsion Hutchins, Emily prototypic conditions for material testing Joyner, Russell LEU NTP ENGINE SYSTEM FOR FLIGHT DEMONSTRATOR FOR A MARS CREW MISSION NTP Kaffezais, Naiki The Ultra‐small Modular Reactor for Space Applications Neutronic Analysis of the Submersion‐Subcritical Safe Space (S4) Reactor for Deploying LEU Fuel System Kajihara, Takanori Using Serpent King, Jeffrey SHIELDING ANALYSIS FOR A MODERATED LOW‐ENRICHED URANIUM FUELED KILOPOWER REACTOR Krecicki, Matt QUANTIFICATION OF INTRA‐ELEMENT POWER PEAKING IN LOW ENRICHED NUCLEAR THERMAL Krecicki, Matt NEUTRONIC FEASIBILITY OF A LOW ENRICHED FAST SPECTRUM NUCLEAR THERMAL PROPULSION Lin, Ching‐Sheng DESIGN AND ANALYSIS OF A 250 MW PLATE‐FUEL REACTOR FOR NUCLEAR THERMAL PROPULSION Combined Cycle Nuclear Power and Propulsion: Reduction in Engineering Complexity to Enable Human Maydan, Jack Mars Mission Architectures in the 2020s Nikitaev, Dennis A LABORATORY TEST TO EVALUATE SEEDED HYDROGEN IN A NUCLEAR THERMAL ROCKET ENGINE THE SIRIUS‐1 NUCLEAR THERMAL PROPULSION FUELS TRANSIENT TEST SERIES IN THE IDAHO NATIONAL O'Brien, Robert LABORATORY TREAT REACTOR Rader, Jordan Dynamic Nuclear Thermal Rocket and Engine Modeling Raftery, Alicia FABRICATION OF UN‐MO CERMET NUCLEAR FUEL USING ADVANCED MANUFACTURING TECHNIQUES Rau, Adam REDUCED ORDER NUCLEAR THERMAL ROCKET ENGINE MODEL Shivprasad, A.P. HYDROGEN ABSORPTION BEHAVIOR OF Y‐10Ce Sikorksi, David REDUCED ORDER NUCLEAR THERMAL ROCKET ENGINE MODEL Snead, Lance PROCESSING TRISO‐BEARING ULTRAHIGH TEMPEARTURE CARBIDE FUELS FOR NUCLEAR THERMAL Sprouster, David MAGNESIUM OXIDE FOR COMPOSITE MODERATORS AND TRISO FUEL MATRICES NUCLEAR THERMAL PROPULSION SUBSCALE EXPERIMENTAL TESTBED FOR MATERIAL INVESTIGATIONS Steiner, Tyler USING THE OHIO STATE UNIVERSITY RESEARCH REACTOR Unger, Aaron DESIGN OF A RADIATION SHIELD FOR A LOW‐ENRICHED URANIUM SPACE NUCLEAR REACTOR initial comparison of reduced and higher order thermal hydraulic solvers for nuclear thermal propulsion Wang, Jim fuel element design Wood, Emily Alternatives for electrical power production from a nuclear thermal propulsion engine Nuclear and Emerging Technologies for Space Knoxville, TN, April 6 – April 9, 2020, available online at https://nets2020.ornl.gov PARAMETRIC EVALUATION OF ALTERNATIVE NUCLEAR PROPULSION CORES USING CURVED FUEL PLATES Abdalla Abou-Jaoude1, and Gilles Youinou1 1Idaho National Laboratory, Idaho Falls, ID 83415 404-455-1657 | [email protected] Nuclear Thermal Propulsion (NTP) holds the The FA are surrounded by a Be block that will need to potential of reducing travel times for deep space missions be cooled in a similar fashion to the Soviet RD-0140 (e.g. to Mars). Previous reactor core designs considered design.1,2 This is in tern surrounded by a Be reflector by the Rover/NERVA program relied on highly enriched containing rotating drums with enriched B4C acting as a uranium (HEU) fuel contained within a hexagonal neutron poison. An axial Be reflector is placed above the graphite matrix. An alternative layout is investigated in fuel, while no lower reflector is used in light of the high this paper. It consists of a circular assembly containing outlet coolant temperatures. concentric curved plates of UN fuel. These fuel assemblies MCNP6.1.0 was used to model the proposed NTP.3 are placed within a beryllium block and reflector. The fuel assembly geometries are modeled explicitly, Preliminary results indicate that many variations of this including the separating structure. For all the criticality design are viable, with high power to mass ratio and outlet calculations, 10,000 virtual particles are used with 700 temperatures. active cycles. I. DESIGN SPECIFICATIONS I.B. Sub-Design Specifications I.A. Introduction and Overview Four main sub-designs are considered in this The NTP core evaluated consists of circular fuel assessment. The main objective to quantify key design assemblies (FA) divided into three parts, each loaded with trade-offs between them. They are labeled A to D and are curved plates as highlighted in Figure 1. The total core summarized in Table I. power output is set at 250 MW. The outside diameter of TABLE I. Design specifications for the four sub-designs. each FA is fixed at 10.1 cm and the height at 80 cm. Design A Design B Design C Design D Assemblies are arranged in a hexagonal pattern. Each FA consists of: Fuel UN UN UN UN Clad Mo W W 184W • A UN fuel meat plate (0.8-3 mm thick) Moderator Be Be Be Be • Mo/W cladding (0.50/0.25 mm thick) Clad thick. 0.50 mm 0.25 mm 0.25 mm 0.25 mm • 3 Mo/W separators (0.50 mm thick) Fuel thick. 0.25 cm 0.25 cm 0.15 cm 0.09 cm • Hydrogen channel (0.75 mm thick) # plates 8 8 12 16 • Between 8 to 16 fuel plates max(Tclad) 2320 K 3000 K 3000 K 3000 K max(Tfuel) 3100 K 3100 K 3100 K 3100 K Volume 16% H2 16% H2 24% H2 32% H2 Fractions 54% fuel 65% fuel 51% fuel 38% fuel 30% stru. 19% stru. 25% stru. 30% stru. Additional design specifications will be defined in the following sections. This includes the FA pitch, the core radius, the minimum number of assemblies, the outlet temperature, flow rate, and the total core mass. These variables are strongly dependent on thermal and neutronic performances. II. PERFORMANCE EVALUATION Fig. 1. Illustration of the curved plate fuel assembly and II.A. Thermal Hydraulic Performance core layout for the proposed NTP. The driving factor for the thermal hydraulic performances inside the four designs are temperature limits. The maximum allowable temperature for Mo and II.B. Neutronic Performance W, are respectively 2320 K and 3000 K. This corresponds While the analysis of Section II.A determined the to 80% of their melting temperatures. The added margin is minimum number of assemblies, the final core layout is due to a reduction of mechanical properties in W/Mo at also affected by neutronic characteristics and the total fuel temperatures approaching their melting temperature (the mass required to reach criticality. To provide sufficient clad provides the structural integrity of the fuel). The margins, an eigenvalue above 1.02 is targeted for all cores maximum centerline temperature for UN is 3100 K with all control rod drums rotated out. (corresponds to its melting point). Figures 2&3 show a visualization of the MCNP6 The scoping study needs to compute an approximate models developed. The reflector and control drums are maximum plate power for a given design in order to assumed to extend to the length of the core, with void determine the maximum total power production in a given above them. An axial reflector (orange) with 90% Be assembly. The inner clad temperature can be computed by volume fraction is located above the core. It is modeled as Eq. 1. It is expressed as a function of the axial position z, a monolithic block at this stage in the analysis. All volume the heat transfer coefficient hH2, the heat flux q”, the bulk beneath the core is assumed to be occupied by the hydrogen H2 temperature TH2, the clad thermal conductivity (푘푐) and exhaust. The main change between the various designs is thickness 훿푐. the number of plates within each fuel assembly (as illustrated in Figure 4), the assembly pitch, and the overall core radius. 1 훿푐 푇푐(푧) = 푞′′(푧) ( + ) + 푇퐻2(푧) (1) ℎ퐻2 푘푐(푧) 2 ′′ 훿푓 훿푔 푇퐶퐿(푧) = 푞 (푧) ( + ) + 푇푐(푧) (2) 2푘푓 푘푔 The fuel centerline temperature can be computed in a similar fashion in Eq. 2 using the fuel/gap thickness (훿푓, 훿푔) and thermal conductivity ( 푘푓, 푘푔 ). For a given outlet temperature and clad thickness (based on manufacturing limits) the maximum heat generation as function of fuel thickness can therefore be deduced. All limiting thermal parameters are highlighted in Table II. A peak-to-average FA ratio of 1.35 was used in the calculations to find an estimate for the minimum number of FA needed. The results show that the maximum power generated in a single assembly can range from 11 MW to 33 MW, leading to H2 mass flow rates between 0.29 kg/s and 0.85 kg/s. Fig. 2. MCNP6 visualization of the XY cross-section for Design A. TABLE II. Assembly-level thermal limits for the four designs. Design: A B C D max(PFA) 11.0 MW 14.0 MW 23.2 MW 32.6 MW Min(#FA) 31 24 15 10 FA 푚̇ 0.29 kg/s 0.36 kg/s 0.60 kg/s 0.85 kg/s Vout 0.7 km/s 0.9 km/s 1.0 km/s 1.0 km/s av(q”’) 1.3 1.6 2.7 3.8 kW/cm3 kW/cm3 kW/cm3 kW/cm3 Tout 2100 K 2550 K 2550 K 2550 K The thermal calculations above were conducted Fig.

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