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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

MARVEL Reactor Fuel Performance Report

The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

High Temperature High Vacuum Mechanical Property Assessment of Zirconium Nuclear Fuel Cladding

This report presents the mechanical characterization of a specific Zry-4 cladding batch serving as the foundation for a diverse range of fuel performance research at Oak Ridge National Laboratory (ORNL). This effort supports research needs for the U.S. Department of Energy (DOE), particularly regarding evaluating accident tolerant fuel (ATF) cladding coating concepts, expanding understanding of cladding response to loss-of-coolant accidents (LOCA) transients, refining post-critical heat flux (CHF) limits (t@T), and upcoming irradiation campaigns. The central objective was to define the baseline performance of the substrate Zircaloy-4 (Zry-4) material leveraged across ORNL Advanced Fuel Campaign (AFC) efforts through controlled high-temperature vacuum tensile testing. This work begins to address gaps in existing models where implementation based on nominal heat-treatment labels, such as stress relief annealed (SRA), often fail to capture the interplay of recovery, recrystallization, and grain growth. To quantify this, data was benchmarked against the Pacific Northwest National Laboratory (PNNL) stress strain model to determine where this material falls in comparison to assumed values for materials in the same heat treatment regime. Analysis of the tensile data revealed that this specific SRA batch exhibits a transitional microstructural state best described by an effective cold-work (CW) parameter of 0.09, diverging from the previous estimation of 0.5 for SRA materials. Additionally, comparative testing of Cr coated specimens demonstrated no distinct difference in axial strength relative to the bare substrate. This suggests that the strengthening benefits of Cr coatings observed in burst scenarios are driven by residual stress mechanisms acting solely in the hoop direction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1973

Evaluation of polyimide/glass fiber composites for construction of light weight pressure vessels for cryogenic propellants

The application of polyimide resin as a matrix for glass filament-wound thin metal-lined pressure vessels was studied over a temperature range of (minus) 320 to 600 F. Keramid 601 polyimide was found to perform quite well over the entire range of temperature. Hoop stress values of 425 ksi were determined at 75 F which is equivalent to epoxy resin in similar structures. At -320 and 600 F, 125 and 80% of this strength was retained. Thermal ageing at 500 F for up to 50 hours was studied with severe reduction in strength, but there is evidence that this reduction could be improved. Another polyimide resin studied was P10PA which was found to have processing characteristics inappropriate for filament-winding. NOL ring tensile and shear data was determined from both resins with S-glass. Pressure vessel design, fabrication and test procedures are described in detail.

Petker, I.↗
NASA NTRS2002

Structural Health Monitoring of Composite Wound Pressure Vessels

The increasing use of advanced composite materials in the wide range of applications including Space Structures is a great impetus to the development of smart materials. Incorporating these FBG sensors for monitoring the integrity of structures during their life cycle will provide valuable information about viability of the usage of such material. The use of these sensors by surface bonding or embedding in this composite will measure internal strain and temperature, and hence the integrity of the assembled engineering structures. This paper focuses on such a structure, called a composite wound pressure vessel. This vessel was fabricated from the composite material: TRH50 (a Mitsubishi carbon fiber with a 710-ksi tensile strength and a 37 Msi modulus) impregnated with an epoxy resin from NEWPORT composites (WDE-3D-1). This epoxy resin in water dispersed system without any solvents and it cures in the 240-310 degrees F range. This is a toughened resin system specifically designed for pressure applications. These materials are a natural fit for fiber sensors since the polyimide outer buffer coating of fiber can be integrated into the polymer matrix of the composite material with negligible residual stress. The tank was wound with two helical patterns and 4 hoop wraps. The order of winding is: two hoops, two helical and two hoops. The wall thickness of the composite should be about 80 mil or less. The tank should burst near 3,000 psi or less. We can measure the actual wall thickness by ultrasonic or we can burst the tank and measure the pieces. Figure 1 shows a cylinder fabricated out of carbon-epoxy composite material. The strain in different directions is measured with a surface bonded fiber Bragg gratings and with embedded fiber Bragg gratings as the cylinder is pressurized to burst pressures. Figure 2 shows the strain as a function of pressure of carbon-epoxy cylinder as it is pressurized with water. Strain is measured in different directions by multiple gratings oriented in both axial and hoops directions.

Joseph Grant↗
NASA NTRS1989

RSRM nozzle fixed housing cooldown test

Flight 5 aft segments with nozzles were exposed to -17 F temperatures while awaiting shipment to KSC in February, 1989. No records were found which show that any previous nozzles were exposed to air temperatures as low as those seen by the Flight 5 nozzles. Thermal analysis shows that the temperature of the fixed housing, and forward and aft exit cone components dropped as low as -10 F. Structural analysis of the nozzles at these low temperatures show the forward and aft exit cone adhesive bonds to have a positive margin of safety, based on a 2.0 safety factor. These analyses show the normal and shear stresses in the fixed housing bond as low values. However, the hoop and meridinal stresses were predicted to be in the 4000 psi range; the failure stress allowable of EA913NA adhesive at -7 F. If the bonds did break in directions perpendicular to the surfaces, called bond crazing, no normal bond strength would be lost. Testing was conducted in two phases, showing that no degradation to the adhesive bonds occurred while the Flight 5 nozzles were subjected to subzero temperatures. The results of these tests are documented. Phase 1 testing cooled a full-scale RSRM insulated fixed housing to -13 F, with extensive bondline inspections. Phase 2 testing cooled the witness panel adhesive tensile buttions to -13 F, with failure strengths recorded before, during, and after the cooldown.

Bolieau, D. J.↗
DOE OSTI2023

Material and component developments for the DEMO divertor using fibre reinforcement and additive manufacturing

Within the research along the European Fusion Roadmap, water-cooled divertor PFCs are foreseen in the design of a first fusion demonstration power plant (DEMO) in order to provide reliable heat removal capability. In the frame of this concerted attempt, the Max Planck Institute for Plasma Physics is concentrating on the development and testing of composite materials based on tungsten (W, preferred armour material) and copper (Cu, preferred heat sink material). W fibres (W f ) as monofilaments and yarns as reinforcement play a central role in these investigations due to their extraordinary properties concerning ductility already at room temperature and high tensile strength. Recent investigations on the impact of radiation damage suggest that the fibres retain their ductility upon irradiation. W reinforced with W fibres (W f /W) allows to overcome the intrinsic brittleness of W. Quantitative mechanical fracture tests of W f /W confirm the basic mechanisms of fibre reinforcement and the increased resistance to mechanical fatigue. The good wettability of W with liquid Cu and the absence of any metallurgical solubility make up an ideal material pairing for composite production. W fibre-reinforced Cu (W f /Cu) cooling tubes provide a rather high thermal conductivity (> 250 W mK -1 ) and at least twice the strength of CuCrZr in hoop direction in the temperature range up to at least 500 °C. Very recent neutron irradiation experiments confirm the sustainment of ductility of the W f /Cu composite. Numerical simulations suggest that thermal stresses in W-Cu PFCs could be strongly reduced by tailoring the local W and Cu volume fraction. This 'freely' distributed material composition can be achieved by means of additively manufactured W skeletons consecutively infiltrated by Cu. Investigations with W preforms produced by Laser Beam Powder Bed Fusion and infiltrated by Cu demonstrate the feasibility of this approach while testing of specifically prepared specimen is ongoing.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Mechanics of the Ring Tension Test (RTT): A Finite Element-based Investigation

The ring tension test (RTT) is an experimental method for determining mechanical behavior in a material’s circumferential or hoop direction. It is a crucial test for testing anisotropic materials with tube geometry, such as nuclear fuel cladding or irradiated pipes. Several RTT configurations exist, each with their own advantages and disadvantages. However, this test is significantly more complex than traditional tensile testing and can be especially sensitive to small differences and inconsistencies in the test setup and geometry, ultimately affecting the derived mechanical properties. Previous research has focused on method development, and little work has been done on understanding the subtle differences between an ideal test and experiments, specifically when the tests are performed on highly irradiated materials in hot cells. In this work, a finite element-based investigation of the RTT is conducted. Two promising test configurations are investigated, comparing their ability to determine accurate material strengths through plastic deformation. Several non-ideal conditions and uncontrollable effects which are likely to occur during experimental testing such as machining tolerances, variations of specimen geometry from nominal dimensions, rotation of specimens and fixturing, and other test setup discrepancies are studied. The sensitivity of measured strengths to these conditions is presented. A mechanics-based approach to describing and correcting raw data to determine actual strengths is also included for one of the configurations, resulting in a robust correction method with highly accurate material strength measurements. Based on these analyses, the hemicylindrical mandrel configuration is recommended with a gauge region oriented at a 45° angle.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗