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NASA NTRS2015

Review of Nuclear Thermal Propulsion Ground Test Options

High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

Coote, David J.↗

DOE OSTI2016

Study of Convection Heat Transfer in a Very High Temperature Reactor Flow Channel: Numerical and Experimental Results

Very High Temperature Reactors (VHTRs) are one of the Generation IV gas-cooled reactor models proposed for implementation in next generation nuclear power plants. A high temperature/pressure test facility for forced and natural circulation experiments has been constructed. This test facility consists of a single flow channel in a 2.7 m (9’) long graphite column equipped with four 2.3kW heaters. Extensive 3D numerical modeling provides a detailed analysis of the thermal-hydraulic behavior under steady-state, transient, and accident scenarios. In addition, forced/mixed convection experiments with air, nitrogen and helium were conducted for inlet Reynolds numbers from 500 to 70,000. Our numerical results were validated with forced convection data displaying maximum percentage errors under 15%, using commercial finite element package, COMSOL Multiphysics. Based on this agreement, important information can be extracted from the model, with regards to the modified radial velocity and property gas profiles. Our work also examines flow laminarization for a full range of Reynolds numbers including laminar, transition and turbulent flow under forced convection and its impact on heat transfer under various scenarios to examine the thermal-hydraulic phenomena that could occur during both normal operation and accident conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRSDate not supplied

Preliminary Findings of the Experimental Development Unit Cold Flow Test for a Generation Zero Nuclear Propulsion Engine

Nuclear thermal propulsion (NTP) technology will greatly benefit human travel to Mars by significantly shortening transit times, improving crew safety, and providing more mission flexibility than traditional chemical rockets. As part of DRACO follow-on work to develop, build, and fly a generation zero NTP engine, a full scale flight-like experimental design unit (EDU) reactor was constructed to collect sufficient on-ground performance data to characterize flow induced vibrations (FIV) of critical reactor structures/components, inform development of the engine and reactor control algorithm, and collect pressure drop and flow distribution data across the reactor. The fluid conditions for the test program were designed to achieve system responses equivalent to that of an operational engine through all phases of engine operation including reactor startup, mainstage operation, reactor shutdown, and reactor cooldown. Over 100 tests were executed, flowing either GN2 or GHe through the EDU at varying flow rates and pressures. This experiment provided early validation of flow behavior and vibration risks before nuclear testing, boosted critical subsystem TRLs, informed design iterations, and reduced future test costs. The steady-state flow parameters for the experiment were modeled in Ansys Thermal Desktop, allowing rapid tuning and experiment-informed updates to a flight-like test matrix. The EDU dynamic environment was characterized with accelerometers, strain gauges, and high-frequency pressure transducers all sampled at 20 kHz. While many narrow-band oscillations were identified, no significant FIV occurred; the reactor structural responses tend to be enveloped by typical launch vehicle ascent vibration environments (defined up to 2 kHz), although significant energy is also present at higher frequencies.

Flow Induced Vibration↗

DOE OSTI2021

Materials and Fuels Complex Five-Year Mission Strategy (FY21- FY25)

The Idaho National Laboratory (INL), through its designated mission of advancing innovative nuclear energy solutions, is actively engaged in the research, development, demonstration and deployment of advanced nuclear technology, as well as in fostering private-public partnership for technology development. Key to the success of INL’s mission is the Materials and Fuels Complex (MFC), the only complex in the U.S. that hosts a world-class assemblage of facilities, capabilities and instruments for handling, testing, and characterizing radioactive materials. Driven by its mission/vision of “Engineering and Experiments that Drive the World’s Nuclear Energy Future,” MFC is at the center of INL’s – and indeed the Department of Energy’s – advanced nuclear technology development initiatives, providing essential capabilities such as engineering-scale high-assay low-enriched uranium (HALEU) fuel production, reactor demonstration facilities, post-irradiation examination, and transient irradiation testing. Furthermore, MFC provides an ideal environment for test beds that are utilized for research, development and demonstration (RD&D) activities on used fuel treatment, nuclear non-proliferation, forensics, and nuclear power sources used for space exploration missions conducted by the National Aeronautics and Space Administration (NASA).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Materials and Fuels Complex (FY2021-2025 Five-Year Mission Strategy)

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2021

Validating Intra-Plate Power Distributions in the Advanced Test Reactor

We report that the neutronics software, HELIOS, was validated in 2015 for performing core reload design and safety analysis of the Advanced Test Reactor. However, when HELIOS was benchmarked against historic fission-wire measurements (i.e., zero-power full-core measurements) a statistically resolved calculation-to-measurement bias was discovered. The azimuthal power along each fuel plate computed by HELIOS has consistently shown to under-predict measurements made by fission-wires in historic zero-power tests near the fuel element side-plates. It was hypothesized during the HELIOS software validation work, that this bias is attributable to local moderation in coolant vents in the side-plates axially just above and below the fission-wires on the fuel-plate edges. This work used detailed MCNP and MC21 models of the side-plate vents to test this hypothesis. By comparing the average azimuthal biases between HELIOS, and 2D and 3D MCNP models, and a 3D MC21 model, it was found that the HELIOS azimuthal bias is not due to the measurement.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Performance and Stability of HyBRID Process - 19115

The HyBRID (Hydrazine Base Reductive metal Ion Decontamination) process shows good decontamination performance on radioactively contaminated specimens gathered from Hanul nuclear power plant and from the HANARO research reactor. The effect of γ-ray irradiation on the decomposition of decontamination solution is investigated. In an oxidation step, KMnO{sub 4} decomposes entirely when the absorbed dose exceeds 20 KGy. The addition of Cu{sup 2+} ion to SP retards slightly the decomposition of KMnO{sub 4}. Less than 20 % of hydrazine in the HyBRID solution, however, decomposes at 20 KGy. To treat the secondary waste generated from the HyBRID process by filtration, it is necessary to demonstrate the integrity of polypropylene filters by γ-ray irradiation. The absorbed dose is in the range of 0 KGy to 80 KGy. The variation of tensile strength and the variation of surface morphology of the irradiated PP filters are investigated. The mechanical test results show that the filters can be used safely in the test range. Hydrazine is decomposed into nitrogen and water by hydrogen peroxide. Tests on the decomposition of hydrazine by a hydrogen peroxide solution are carried out by varying the solution pH in the temperature range between 50 deg. C and 70 deg. C. The decomposition of hydrazine becomes effective as the increase of pH and the repetitive injections of hydrogen peroxide, and hydrazine is satisfactorily decomposed. The effect of copper ion on the decomposition of hydrazine is negligible. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Fabrication and Testing of Micro Flux Monitors in the High Flux Isotope Reactor

Neutron dosimetry is a critical technique to measure the neutron flux, fluence, and energy spectrum of nuclear reactors and can be used to indicate fuel reloading, high-power operations, or changes in cycle length. This work presents results of a novel device named the micro flux monitor, which is a miniaturized neutron dosimeter made from small quantities (100s of nanograms) of Ti, Al, Ni, and Au metals deposited on Si and high-purity fused silica substrates using standard semiconductor fabrication techniques. Approximately 1,250 of these devices were produced, and 40 were irradiated in the pneumatic tube facility of the Neutron Activation Analysis Laboratory at the High Flux Isotope Reactor. Testing demonstrated that dosimeter metals can be deposited predictably and repeatedly in multiple geometries, including as quick response (QR) codes. Neutron flux measurements with these devices were within 3% of measurements made using standard flux monitors, demonstrating the efficacy of these devices. Devices were also tested at 500°C under a N2 atmosphere and showed minimal degradation, suggesting they could be deployed for neutron dosimetry measurements in high-temperature advanced reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Measure Effects of Molten Halide Salt on Creep Rupture Lifetime for 316H, 709 and 617

Ongoing commercialization of different molten salt reactor concepts can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy in molten fluoride salt. This task evaluated the creep behavior of candidate structural materials 316H, 709 and 617 for advanced nuclear reactors under molten halide salt environments at 650, 700 and 750°C respectively. Creep tests were conducted in air, Ar, fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts. The molten salt environment caused a 50% reduction in creep rupture lifetime for 316H in FLiNaK and an 80% reduction in creep rupture lifetime in NaCl-MgCl 2 compared to air exposures. The molten fluoride salt environment resulted in almost a 90% and 50% reduction in creep rupture time for 709 and 617 irrespective of the cover gas environment (air and Ar). Significant differences in the compositional changes between the salt- and air-facing surfaces were observed in the alloy subsurface which might have potentially lead to formation of phases detrimental to the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

GNF fuel technology update

In 2015, Global Nuclear Fuel (GNF) launched the GNF3 fuel design by inserting lead-use assemblies into two boiling water reactors (BWRs) in the U.S. A lead use assembly is a nuclear fuel assembly which is inserted in a reactor core to confirm its performance. A third set of GNF3 lead-use assemblies are scheduled to be inserted into another reactor in the spring of 2017. GNF3 is a robust next-generation 10*10 fuel design that will improve both fuel cycle and fuel reliability. In addition, GNF3 will be delivered in a fully assembled condition - forgoing the need to insert fuel channels over the bundles at the site. The process of shipping BWR fuel in a channeled configuration is complex and requires an understanding of the potential high-cycle and low-cycle fatigue conditions that can occur during shipping, as characterized by accelerometer data from ship tests. The key component in evaluating channeled fuel shipping is the spacer, and the GNF3 spacer fatigue performance was found to be approximately twice that of GNF2. While fuel reliability has traditionally been related to the avoidance of fuel leakers, for the past 15 years channel - control blade interference has been a close second for concern. To eliminate this problem, GNF3 will come standard with NSF (a zirconium alloy) channels that are effectively resistant to fluence gradient-induced bow and shadow corrosion-induced bow. In addition to the development of GNF3, GNF is evaluating the use of advancing 3D metal printing or additive manufacturing for manufacturing fuel bundle components. Additive manufacture offers the hope of more flexibility in design and shorter manufacturing development cycles. However, since there is little experience with this technology in the nuclear industry, GNF has been evaluating the use of additive manufacture to fabricate small scale components that are easily retrieved from the reactor for testing. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Capabilities Development for Transient Testing of Advanced Nuclear Fuels at TREAT

The TREAT facility is a unique capability at the Idaho National Laboratory currently being prepared for resumption of nuclear transient testing. Accordingly, designs for several transient irradiation tests are being pursued to enable development of advanced nuclear fuels and materials. In addition to the reactor itself, the foundation for TREAT’s capabilities also include a suite of irradiation vehicles and supporting infrastructure to provide the desired specimen boundary conditions while supporting a variety of instrumentation needs. The challenge of creating these vehicles, especially since many of the modern data needs were not historically addressed in TREAT experiment vehicles, has necessitated a sizeable engineering effort. This effort is currently underway and maturing rapidly. This paper summarizes the status, future plans, and rationale for TREAT experiment vehicle capabilities. Much of the current progress is focused around understanding and demonstrating the behavior of fuel design with enhanced accident tolerance in water-cooled reactors. Additionally, several related efforts are underway to facilitate transient testing in liquid sodium, inert gas, and steam environments. This paper discusses why such a variety of capabilities are needed, outlines plans to accomplish them, and describes the relationship between transient data needs and the irradiation hardware that will support the gathering of this information.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

AGC 3 Irradiated Material Properties Analysis

This report documents the analysis of the irradiated material property data from the Advanced Graphite Creep (AGC)-3 graphite specimens. This is the third in a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-3 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 800°C, beginning with irradiation Cycle 152B on November 28, 2012 and ending with Cycle 155B on April 12, 2014, with a total received dose range of 0.9–3.7 dpa. Larger creep and control specimens located more centrally in the capsule received a dose of 1.0 – 3.7 dpa. AGC-3 was designed to provide irradiation conditions similar to AGC-1 and AGC-2 capsules (similar graphite grades tested, specimen dimensions, mechanical loading conditions) but at a different nominal irradiation temperature of 800°C. AGC-3 was irradiated for a short duration to provide material property values at lower dose levels. AGC-4 will have a longer duration and provide material property values at higher dose levels. After irradiation, material property and dimensional strain measurements were conducted on all AGC-3 specimens (from 11 nuclear graphite grades) using the same equipment and approved standards as were conducted before irradiation. The specimen loading configuration for all graphite grades within AGC-3 followed a similar pattern as earlier AGC capsules to provide easy future comparison of all irradiated material property data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Status of New Models Hosted on the Virtual Test Bed (VTB) in 2022

The VTB hosts 20 distinct models, 8 of which were recently developed during 2022 and will be discussed in more detail here. This article provides an overview of the various models hosted and their capabilities. Part of the intent of the VTB is to showcase a ‘block’ style approach to modeling and simulation: if a capability is not showcased for a specific reactor type, it could, in theory, be easily ported/replicated from another. The breadth of advanced reactor types considered is representative of the large variety of concepts being proposed for demonstration.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Integration of real-time measurements and atomistic modeling to license nuclear components

An improved testing and data gathering method is described herein with reference to testing a new fuel, as an exemplary component to be tested for licensure purposes. The method includes generally: generating models for the new system; making samples and getting them accepted for a reactor; and testing the samples in a test or commercial reactor until the exposure time is reached for the expected cycle length of the fuel at final use. The method is preferably done concurrent to submitting a license application for commercial use of the new component to the relevant government body to expedite license testing review.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

2025 Intern Poster

The Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) houses the largest U.S. inert atmosphere hot cell for nuclear material research. Key features include the precision gamma scanning (PGS), Fuel Accident Condition Simulator (FACS), Neutron Radiography Reactor (NRAD), and the focus of this project, the Metallograph Loading Cell (MET Cell). The MET Cell performs tests on spent nuclear fuel, such as microhardness testing, microscopy, and neutron radiography. However, the MET Cell’s existing pressure and lighting control systems are outdated and inefficient, with inadequate documentation for system changes over time. This project aims to design a new automated control system for the MET Cell, ensuring longevity (minimum ten years), ease of troubleshooting/repair, and integration into the building monitoring system. The design process addressed challenges such as space restrictions, varied voltages within enclosures, sourcing new components, and security limitations. Compliance with NFPA 70, UL508A, MFC Physical Security, and INL Engineering standards was essential. The project involves repurposing an existing PLC to manage lighting and pressure control using digital and analog signals, simplifying wiring, and ensuring thorough documentation for future reference.

42 - ENGINEERING↗

DOE OSTI2019

Weibull and Bootstrap-Based Data-Analytics Framework for Fatigue Life Prognosis of the Pressurized Water Nuclear Reactor Component Under Harsh Reactor Coolant Environment

In general, the fatigue life of a safety critical pressure component is estimated using best-fit fatigue life curves (S-N curves). These curves are estimated based on underlying in-air condition fatigue test data. The best-fitting approach requires a large safety factor to accommodate the uncertainty associated with large scatter in fatigue test data. In addition to this safety factor, reactor component fatigue life prognostics requires an additional correction factor that in general is also estimated deterministically. This additional factor known as the environmental correction factor F en is to cater the effect of the harsh coolant environment that severely reduces the life of these components. The deterministic F en factor may also lead to further conservative estimation of fatigue life leading to unnecessary early retirement of costly reactor components. To address the above-mentioned issues, we propose a data-analytics framework which uses Weibull and Bootstrap probabilistic modeling techniques for explicitly quantifying the uncertainty/scatter associated with fatigue life rather than estimating the lives based on a best-fit based deterministic approach. We assume the proposed probabilistic approach would provide the first hand information for assessing the maximum and minimum effects of pressurized water reactor water on the reactor component. In the discussed approach, in addition to the probabilistic fatigue curves, we suggest using a probabilistic environment correction factor F en . Here, we assume the probabilistic fatigue curve and F en would capture the S-N data scatter associated with the bulk effect of material grades, surface finish, strain rate, etc. on the material/component fatigue life.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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 OSTI2026

WEC Irradiation Can Thermal Testing Report

With renewed interest for the next generation of small modular reactors and space-based nuclear reactors, metal hydrides have been a focus of research for moderating materials to reduce the need for highly enriched fuel to maintain small core sizes. Yttrium hydride (YH) is an attractive candidate for these metal hydrides due to its ability to retain hydrogen at elevated temperatures as well as the low neutron absorption cross section of yttrium metal. Given the necessity for reactors to operate at elevated temperatures for extended periods of time, understanding hydrogen retention in moderator channels in simulated operating thermal conditions is paramount in predicting reactor performance and the long-term stability of YH. Additionally, coating strategies on the moderator encapsulating will be explored to ascertain if it aids in retention of hydrogen within the capsule and moderator.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗