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NASA NTRS2005

Results of a 2000 hour wear test of the NEXIS ion engine

The Nuclear Electric Xenon Ion System (NEXIS) ion thruster was developed for potential outer planet robotic missions under NASA's Prometheus program. This engine was designed to operate at power levels ranging from 16 to over 20 kWe at specific impulses of 6000 to 7500 s for burn times of up to 10 years, satisfying the requirements of nuclear electric propulsion systems such as that on the proposed Prometheus 1 mission to explore the icy moons of Jupiter. State-of-the-art performance and life assessment tools were used to design the thruster. Following the successful performance validation of a Laboratory Model thruster, Development Model hardware was fabricated and subjected to vibration and wear testing. The results of a 2000-hour wear test are reported herein. Thruster performance achieved the target requirements and was steady for the duration of the test. Ion optics performance was similarly stable. Discharge loss increases of 6 eV/ion were observed in the first 500 hours of the test and were attributed to primary electron energy decreases due to cathode insert conditioning. Relatively high recycle rates were observed and were identified to be high-voltage-to-ground arcs in the back of the thruster caused by wire insulation outgassing and electron penetration through the plasma screen. Field emission of electrons between the accelerator and screen grids was observed and attributed to evolution of field emitter sites at accelerator grid aperture edges caused by ion bombardment. Preliminary modeling and analysis indicates that the NEXIS engine can meet mission performance requirements over the required lifetime. Finally, successful validation of the NEXIS design methodology, design tools, and technologies with the results of the wear test and companion performance and vibration tests presents significant applicability of the NEXIS development effort to missions of near-term as well as long-term interest for NASA.

Nuclear Electric Xenon Ion System (NEXIS)Thruster ↗

DOE OSTI2021

MULTI-MISSION THERMOELECTRIC GENERATOR ASSEMBLY TESTING AND LAUNCH OPERATIONS FOR MARS 2020

The Space Nuclear Power and Isotope Technologies (SNPIT) division at Idaho National Laboratory (INL) fuels, performs acceptance testing, and provides spacecraft integration support of Radioisotope Power Systems (RPS) in support of National Aeronautics and Space Administration (NASA) missions. Recently the SNPIT team completed assembly, testing and launch support of the Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) for the Mars 2020 Perseverance Rover mission.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2021

MULTI-MISSION THERMOELECTRIC GENERATOR ASSEMBLY TESTING AND LAUNCH OPERATIONS FOR MARS 2020

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2025

Performance Analysis and Simulaion of the Hydraulic Scram System in TREAT Reactor

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) serves a vital role in nuclear fuel safety research, enabling transient experiments that simulate reactivity excursions and accident scenarios. Central to these operations is the transient control rod drive system (TCRDS), which drives rapid motion of the transient control rods such that TREAT can simulate rapid power changes typical of reactor accidents. The reliability and performance of this system are critical for protecting both fuel specimens and reactor infrastructure. This study presents the initial phase of a two-year investigation into the dynamics and reliability of the TREAT hydraulic TCRDS. Conducted in collaboration with INL, the research employs a combined computational and experimental approach to analyze the system's response time, pressure transients, and potential failure modes. Emphasis is placed on understanding how fluid characteristics influence the TCRDS’s ability to achieve both rapid power changes and mechanical stability. The TRDS and the skid that powers it will be analyzed throughout this investigation. Computational modeling using computational fluid dynamics (CFD) will simulate the hydraulic response under varying conditions. In parallel, experimental testing planned at INL will validate these models and capture key performance metrics. This paper outlines the system design, analytical framework, and modeling strategies that form the foundation for later testing. Ultimately, this work aims to support improvements to the TCRDS’s design and reliability, contributing to the broader goal of enhancing nuclear fuel safety and sustaining TREAT’s mission as a premier nuclear fuel test facility.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2021

Los Alamos National Laboratory Floodplain Assessment for Fire Risk Mitigations at the Lower Slobbovia Testing Site in (TA-36)

The National Nuclear Security Administration (NNSA), a semi-autonomous agency within the U.S. Department of Energy (DOE), is proposing to take fire pre-suppression actions at Los Alamos National Laboratory (LANL) Technical Area (TA) 36. Constructing a fire break and a fuel break surrounding the Lower Slobbovia explosives testing site in addition to the current vegetation mowing practices would establish a safety perimeter to help prevent fires from spreading away from the firing point. NNSA has prepared this floodplain assessment in accordance with 10 Code of Federal Regulations (CFR) Part 1022 Compliance with Floodplain and Wetland Environmental Review Requirements (10 CFR 1022), which was promulgated to implement DOE requirements under Executive Order 11988 Floodplain Management (EO 1977). A floodplain is defined in 10 CFR 1022 as “the lowlands adjoining inland and coastal waters and relatively flat areas and flood prone areas of offshore islands,” and a base floodplain as “the 100-year floodplain, that is, a floodplain with a 1.0 percent chance of flooding in any given year (CFR 2003).” This floodplain assessment evaluates potential impacts to floodplain values and functions from implementation of the proposed action, identifies alternatives to the Proposed Action, and allows for meaningful public comment.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2019

MCNP Analysis fo the Aqua-SETH Experiment in TREAT

The versatile Transient Reactor Test (TREAT) Facility’s capability to deliver customized transient nuclear power excursions gives researchers and developers a multitude of specialized testing options. In conjunction with reactorprescribed delivery of neutron flux, experiment design requires specimen packaging to meet TREAT safety containment requirements and experimenter test objectives. These objectives may range from the very simple separate effects tests (generally dry capsule) to the more complex integral scale testing wherein prototypic boundary or specimen reactor environment conditions are created. The historic MARK-III sodium flow loop that achieved thermalhydraulic prototypic test conditions serves as an example of the more complex experiment design package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Expansion of the Fuel Motion Monitoring System

The Idaho National Laboratory (INL) Transient Reactor Test Facility (TREAT) Fuel Motion Monitoring System (FMMS) is a diagnostic tool used to visualize the movement and location of fuel within the nuclear reactor during transient experiments. TREAT is designed to test the behavior of nuclear fuel under accident conditions, such as a rapid power increase. The TREAT FMMS, also referred to as the “hodoscope” because of the system’s massive steel collimator component, is located on the North beam port of the reactor. While the FMMS system is capable of incorporating a total of 360 sensors, initial refurbishment efforts resulted in a array of 96 fast neutron detectors. This initial array was configured to provide a narrow (2 channels) full-length view of experiment vessels with an extended 4-channel–wide region in the center of the array capable of providing a full view of all initially planned experiments. Planning and efforts to expand the FMMS array began immediately following the restart of TREAT. An additional 96 detectors were evaluated, characterized, and prepared for installation at TREAT during the summer of 2018. Unfortunately, funding reductions forced the project towards a 5-year hiatus. In October of 2023 funding from the Department of Energy’s Nuclear Science User Facilities enabled the expansion to recommence. This report serves to document the activities performed to expand the FMMS detector array to 192 channels in order to provide a broader view of the larger and more sophisticated test capsules currently planned for future irradiation in the TREAT reactor. This expansion also included doubling the data acquisition capability, addressing the new heat load on the system, and synchronizing the time for all digital components.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1973

Flowing gas, non-nuclear experiments on the gas core reactor

Variations in cavity wall and injection configurations of the gas core reactor were aimed at establishing flow patterns that give a maximum of the nuclear criticality eigenvalue. Correlation with the nuclear effect was made using multigroup diffusion theory normalized by previous benchmark critical experiments. Air was used to simulate the hydrogen propellant in the flow tests, and smoked air, argon, or Freon to simulate the central nuclear fuel gas. Tests were run both in the down-firing and upfiring directions. Results showed that acceptable flow patterns with volume fraction for the simulated nuclear fuel gas and high flow rate ratios of propellant to fuel can be obtained. Using a point injector for the fuel, good flow patterns are obtained by directing the outer gas at high velocity long the cavity wall, using louvered injection schemes. Recirculation patterns were needed to stabilize the heavy central gas when different gases are used.

Kunze, J. F.↗

DOE OSTI2018

Combined Assay Systems for the Characterisation of Historical ILW at of the Dounreay Shaft and Silo-18140

In 1955 the Dounreay nuclear power development establishment was created on the northern coast of Scotland primarily to pursue the UK Government policy of developing Fast Breeder Reactor (FBR) technology. The site operated then by the United Kingdom Atomic Energy Authority (UKAEA) saw the development, test and operation of 3 nuclear reactors, two of them FBRs, alongside fabrication and reprocessing facilities for the FBR fuel elements. From the late 1950's until 1977 Intermediate Level radioactive Waste (ILW) generated on site was disposed of into a 65 m deep, 4.6 m diameter shaft. In January 1977, an explosion thought to be caused by hydrogen generated from sodium and potassium wastes reacting with water occurred in the upper part of the shaft, thus stopping any further waste disposal. A second ILW storage facility, the ILW-Silo, started operation in 1971 to deal with the increasing volume of intermediate level waste generated on site in the mid-1960's and continued receiving waste until 1998. The decommissioning of the Shaft and Silo facilities is a challenging engineering task due to the complexity and potential hazards linked with the retrieval and handling of historical waste. Following recovery it is intended to shred the majority of the waste and load it into 200-l assay drums. Sludges would be separated and solidified within a special annulus drum to make them suitable for assay. To support the decommissioning strategies put forward by Dounreay Site Restoration Limited (DSRL), Mirion Technologies Canberra UK Ltd have designed and completed the manufacture of a pair of combined non-destructive waste assay (NDA) systems. The two identical automated ILW counters are intended to provide a means to characterise the waste materials by performing active and passive neutron measurement as well as segmented high resolution gamma spectrometry (HRGS) of 200-l sacrificial waste drums. The active neutron technique is based on the Differential Die Away (DDA) technique using a pulsed Deuterium-Tritium neutron generator to interrogate fissile material present in the waste. The active neutron measurements support, as well as the passive neutron coincidence measurements, a matrix correction capability by use of an Add-a-Source mechanism. The publication will describe in broad lines the approach proposed by DSRL in the processing of the waste retrieved from both the Shaft and Silo facilities. It will also present a description of the overall design of the NDA counters, illustrating how the measurements to be performed will allow DSRL to safely characterise the retrieved waste materials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Equibiaxial Flexure Strength of a Superfine-Grained Nuclear Graphite

The strength of advanced graphite is reported in accordance with ASTM D7846-16 using a two-parameter Weibull distribution for uniaxial strength testing. The rule of thumb for Weibull distributions is to use a minimum of 30 strength measurements to have a high level of confidence in the Weibull characteristic strength and the Weibull modulus. These large sample sets for statistical confidence are easily obtained for as-manufactured graphite, but determination of the Weibull two-factor parameters on graphite that has been exposed to neutron irradiation is nearly impossible due to irradiation testing restraints on specimen size and space limitations in test reactors for the accommodation of specimens. These restrictions have resulted in irradiation programs that use sub-sized specimens to measure uniaxial strength change, but the specimen geometries still hinder the number of replicate specimens for each irradiation condition. In contrast, this work used an advanced ceramic standard, ASTM C1499-05, as the foundation for equibiaxial strength testing of a superfine-grained nuclear graphite to investigate the irradiation-induced strength change and any changes to the two-parameter Weibull distribution. The goal was to investigate the effect of specimen size on the two-parameter Weibull distribution parameters in support of the small specimens and geometry desired for irradiation. A study of the effect of a reduced sample population was also undertaken to investigate the reliability of using the two-parameter Weibull distribution to study changes to the materials properties caused by neutron irradiation damage. The results of this work suggest that these equibiaxial specimens would be ideal for statistically significant studies of the effects of irradiation on the Weibull characteristic strength (but not the modulus) and for providing a method for surveillance specimen campaigns for future operating commercial reactors.

Campbell, Anne↗

DOE OSTI2019

ENDF/B-VIII.0 Covariance Data Development and Testing for Advanced Reactors

This report documents the ENDF/B-VIII.0 covariance data development and testing performed at Oak Ridge National Laboratory performed under the auspices of the US Department of Energy Office of Nuclear Energy. This work supports modeling and simulation of advanced reactor designs in the US by analyzing, validating and improving the nuclear data that provide underlying physics for calculations. This report assesses nuclear data uncertainty information for neutron cross sections, provided in the form of covariance matrices, in the 2018 ENDF/B-VIII.0 library. A review and interpretation of some of the issues with the current covariance library is provided. A new approach to augment the ENDF/B-VIII.0 covariance matrix to accurately reflect uncertainties in the results of modeling and simulation reliant on the current nuclear data library is documented. Rigorous testing of the ENDF/B-VIII.0 covariance data is documented. Finally, the performance of the ENDF/B-VIII.0 covariance data for advanced reactor applications is assessed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

The Annular Core Research Reactor (ACRR) Description and Capabilities

The Annular Core Research Reactor (ACRR) is a unique, one-of-a-kind nuclear reactor facility operated by Sandia National Laboratories for the National Nuclear Security Administration that is considered by many to be a national treasure. There is no other research reactor in the world that has the attributes and capabilities comparable to that of the ACRR. The ACRR was specifically designed to meet the irradiation testing needs of the U.S. nuclear weapons program. The ACRR has four major attributes that make it unique: 1) a large dry central cavity; 2) an epithermal neutron flux; 3) a large pulsing capability; and 4) a fueled-ring external cavity with a larger dry cavity. This report presents the unique capabilities of the ACRR and documents some of the more important metrics associated with its operation. ACKNOWLEDGEMENTS The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Startup Test Plan and Predictions for Highly Enriched Uranium to Low-Enriched Uranium Fuel Conversion at the University of Missouri Research Reactor

Nonpower reactors licensed by the U.S. Nuclear Regulatory Commission require a startup test plan as part of any facility modification to verify operability prior to resumption of operations. In order to support conversion of the University of Missouri Research Reactor from the use of highly enriched uranium to low-enriched uranium (LEU) fuel, a startup test plan has been devised to measure certain reactor physics parameters for the initial all-fresh LEU core licensing documentation that will be submitted. These parameters include the approach to critical, primary coolant void coefficient of reactivity, flux trap void coefficient of reactivity, determination of flux trap sample reactivity worth, radial and axial thermal neutron flux mapping, control blade worth calibration, primary and pool coolant temperature coefficient of reactivity, and flux mapping of experimental positions. Here, predictions for these parameters made using the Monte Carlo N-Particle Version 5 (MCNP5) radiation transport code are reported. These predictions will support the startup tests by providing a baseline set of expectations and additional insight into the performance of the LEU core.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Direct local parametrization of nuclear state densities using the back-shifted Bethe formula

Level densities are often parametrized using the back-shifted Bethe formula (BBF) for nuclei that possess experimental data for s-wave neutron resonance average spacings and a complete discrete level sequence at low excitation energies. However, these parametrizations require the additional modeling of the dependence of the spin-cutoff parameter on excitation energy. Here, in this work, we avoid the need to model the spin distribution of level densities by using the experimental data to parametrize directly the state densities, for which the BBF does not depend on the spin-cutoff parameter. This approach allows for a local parameterization of state densities that is independent of the spin-cutoff parameter. We provide these parameters in a tabulated form for applications in nuclear reaction calculations and for testing microscopic approaches to state densities.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

A Critical Review of Heat Pipe Experiments in Nuclear Energy Applications

This critical review provides heat pipe (HP) experimental data sets that contain pertinent information regarding nuclear technology that may be beneficial to researchers. Heat pipes have been shown to have a tremendously positive impact on nuclear technologies and will continue to become a more prevalent technology as more nuclear reactor concepts embrace this robust technology. Most previous reviews may focus on only a specific HP design or application, and some are backdated. Herein, this critical review extends previous efforts; integrates and summarizes previously reported HP experimental efforts; and provides updates with recently reported results in the literature for HPs in all nuclear-related applications, including space power (thermal radiators, core cooling, and electricity production), microreactors (emergency core cooling, hybrid control rods, and reactor core cooling), and HP involvement in other nuclear-related technologies (spent fuel pool cooling). The two main objectives of this critical review are (1) to facilitate the development of HP codes by outlining some of the existing experimental data sets to validate their codes and directing developers to these efforts and (2) to provide comprehensive information regarding the vast applicability of HPs used in the nuclear industry, including the theory of operation and limitations to supplement researchers in the development of new ideas for potential applications in nuclear-related technologies. The review clearly shows extensive and diverse experimental data sets for HPs developed under diverse testing conditions depending on the available nuclear application for validation purposes. Thus, this critical review is oriented to providing attention to the existing efforts rather than determining gaps in HP research.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Electrical Arc Fault Particle Size Characterization

In this investigation a series of small-scale tests were conducted, which were sponsored by the Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES) and performed at Sandia National Laboratories (SNL). These tests were designed to better understand localized particle dispersion phenomena resulting from electrical arcing faults. The purpose of these tests was to better characterize aluminum particle size distribution, rates of production, and morphology (agglomeration) of electrical arc faults. More specifically, this effort characterized ejected particles and high-energy dispersion, where this work characterized HEAF electrical characteristics, particle movement/distributions, and morphology near the arc. The results and measurements techniques from this investigation will be used to inform an energy balance model to predict additional energy from aluminum involvement in the arc fault. The experimental setup was developed based on prior work by KEMA and SNL for phase-to-ground and phase-to-phase electrical circuit faults. The small-scale tests results should not be expected to be scale-able to the hazards associated with full-scale HEAF events. Here, the test voltages will consist of four different levels: 480V, 4160V, 6900V and 10kV, based on those realized in nuclear power plant (NPP) HEAF events.

36 MATERIALS SCIENCE↗

DOE OSTI2023

NCERC Provides Unique Opportunities for University Student Researchers

Five students and one faculty member - sponsored by the Defense Nuclear Nonproliferation (DNN, NA 22) university consortia - visited the National Criticality Experiments Research Center (NCERC) in July of 2023 to measure radiation signatures from Category I Special Nuclear Material (SNM) in a week-long measurement campaign organized by staff at Los Alamos National Laboratory. Participants included University of Florida, University of Michigan, and University of Illinois-Champaign Urbana. The measurement campaign was organized on behalf of the Consortium for Monitoring, Testing, and Verification (MTV), the Nuclear Science and Security Consortium (NSSC), and the Consortium for Enabling Technologies and Innovation (ETI).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2019

Standardized verification of the C YCLUS fuel cycle simulator

Many nuclear fuel cycle simulators can analyze transitions from once-through to advanced nuclear fuel cycles. Verification studies compare various fuel cycle analysis tools to test agreement and identify sources of difference. A recent verification study, Feng et al. (2016) established transition scenario test case specifications and accordingly evaluated national laboratory nuclear fuel cycle simulators, DYMOND, VISION, ORION, and MARKAL. This work verifies the performance of C YCLUS , the agent-based, open-source fuel cycle simulator, using the test case specifications in Feng et. al. We report, C YCLUS demonstrates agreement with the results from the previous verification study. Minor differences reflect intentional, detailed material tracking in the C YCAMORE reactor module. These results extend the example results in Feng et al. to further enable future verification of additional nuclear fuel cycle simulation tools.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗