Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “fast neutron”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
NASA NTRSDate not supplied

LEIA: An Investigation of Radiation Risks to Biology at the Lunar South Pole

Radiation and reduced gravity pose biological risks to crewed deep space exploration. At the cellular level, radiation damage can be amplified by reduced gravity. Empirical evidence on cellular responses to beyond low Earth orbit (BLEO) environments is imperative to develop effective countermeasures for crew health and in-space biomanufacturing. The Lunar Explorer Instrument for Space Biology Applications (LEIA) project is developing an instrument suite to be delivered to the south polar region of the Moon by the Commercial Lunar Payload Services (CLPS) program. This presentation will provide an overview of the LEIA hardware, experiments, and mission timeline. The LEIA instruments include the BioSensor, the ARES charged particle detector, and the Mini-FND. The BioSensor is an autonomous light emitting diode (LED)-based spectrophotometer and microfluidic incubator. The BioSensor activates yeast cultures and can measure cell growth, metabolic activity, and carotenoid production. The ARES is a Timepix-based charged particle radiation detector that measures dose, dose rate, and linear energy transfer spectra. The Mini-FND is a fast neutron detector that measures albedo neutron flux and energy spectra. Combined, these instruments will be used for yeast genetics experiments to quantify growth, metabolism, and synthetic biology-enabled production of human nutrients, while taking real time measurements of biologically relevant radiation exposure on the lunar surface. These data will be used to test the importance of selected DNA damage repair and reactive oxygen species defense pathways in mitigating cellular damage from lunar surface radiation.

Yeast↗
NASA NTRSDate not supplied

Yeast Strain Development and Hardware Testing in Preparation of a Lunar BioSensor

With Artemis missions underway, it is clear we are going back to the Moon to stay. Before sending Astronauts for long-duration missions, it is crucial to understand the technological and biomedical countermeasures needed to protect them before they get there. We can use knowledge gained from biological CubeSats to guide the next generation of experiments to support human habitation on the Moon. Lunar Explorer Instrument for space biology Applications (LEIA) is NASA’s latest BioSensor, adapted BioSentinel, the only CubeSat to travel Beyond Low Earth Orbit. BioSentinel launched on Artemis I and is currently >50 million kilometers from Earth (as of July 2024). LEIA aims to identify biological responses to the Lunar environment, which unprotected against would pose a threat to astronauts (cancer, cardiovascular disease, neurological impairment). The suite of instruments within LEIA detects Lunar radiation using two on-board radiation sensors (ARES charged particle detector, Mini-Fast Neutron Detector), then monitors real-time biological responses to the Lunar environment via an autonomous microfluidic system, fit with 3-LED emitter and detector boards and the alamarBlue metabolic indicator dye. LEIA will use a genetic approach in addition to synthetic biology to test counter-measure production in space, with the goal to inform and protect astronauts for future Moon missions. We have conducted preliminary tests in preparation for launch to the anticipated South Pole of the Moon, optimizing the biology (strain down-selection, desiccation tolerance, radiation sensitivity) and improving the hardware (including a blue LED to detect the beta-carotene countermeasure product). Our team will discuss these findings in several parts – an overview of the LEIA mission (Mark Settles), adapting flexible CubeSat platforms for deep-space applications (Sergio Santa Maria, Kira Rienecker), developing new technologies to support LEIA ground studies (Chinmayee Govinda Raj), and yeast strain development and hardware testing in preparation for LEIA (presented here).

synthetic biology↗
DOE OSTI2021

ECAR-3051 As-Run Physics Analysis For The AGC-3 Experiment Irradiated In The ATR

This Engineering Calculations Analysis Report (ECAR) documents the results of the Advanced Test Reactor (ATR) detailed physics analyses performed to calculate the displacements per atom (DPA) and the fast neutron fluence (E > 0.1 MeV) of the Advanced Graphite Creep (AGC) experiment, AGC-3, irradiated in the ATR East Flux Trap (EFT) (see Figure 1) during ATR Cycle 152B, Cycle 154B, Cycle 155A, and Cycle 155B. This ECAR also reports the neutron and photon heat rates for the materials of the AGC-3 experiment for ATR Cycle 152B. The results for these evaluations and analysis are reported herein.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

ECAR-2291 AS-RUN PHYSICS ANALYSIS FOR THE AGC-2 EXPERIMENT IRRADIATED IN THE ATR

This Engineering Calculations Analysis Report (ECAR) documents the results of the Advanced Test Reactor (ATR) detailed physics analyses performed to calculate the displacements per atom (DPA) and the fast neutron fluence (E > 0.1 MeV) of the Advanced Graphite Creep (AGC) experiment, AGC-2, irradiated in the ATR South Flux Trap (SFT) (see Figure 1) during ATR Cycle 149A, Cycle 149B, Cycle 150B, Cycle 151A, and Cycle 151B. This ECAR also reports the neutron and photon heat rates for the materials of the AGC-2 experiment for ATR Cycle 149B. The results for these evaluations and analysis are reported herein.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Uncertainty Quantification of Calculated Temperatures for the AGR 5/6/7 Experiment

This report documents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) 5/6/7 fuel irradiation experiment conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technologies? research and development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR 5/6/7 capsules, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The calculated fuel temperatures for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. This report covers temperature uncertainty results for each of the five AGR 5/6/7 capsules. To quantify the uncertainty of calculated temperatures determined using the ABAQUS finite element heat transfer code, this study identifies and analyzes model parameters of potential importance to the calculated temperatures of fuel compacts and thermocouples. The selection of input parameters for uncertainty quantification is based on the ranking of their influences upon temperature predictions. Thus, selected input parameters include those with high sensitivity and those with the largest uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of calculated temperatures. Measurement uncertainty, analysis of modeling assumptions, and expert judgment are used as the basis to quantify the uncertainty range for selected input parameters. The input uncertainties are dynamic, accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperatures and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish sufficiency of the time dependent first order (linear) expansion terms in constructing the temperature response surface. To achieve completeness, uncertainty propagation made use of pairwise noise correlations of model parameters. Furthermore, using an interpolation scheme over the input parameter domain, the analysis obtains time dependent sensitivity over the test campaign duration. This allows computation of uncertainty for the calculated fuel temperatures and the calculated graphite temperatures at thermocouple locations during the entire irradiation period.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Phonon-informed Neural Thermal Scattering (NeTS) Optimization for Crystalline Graphite and Beryllium Metal

Fast neutrons born from fission lose energy through scattering interactions in the process of slowing-down. As neutrons thermalize to the order of $k$ $b$ $T$ (where $k$ $b$ is the Boltzmann constant, and $T$ is the temperature of the medium), their de Broglie wavelength and energy approaches the order of inter-atomic spacing and quantized lattice vibrations, i.e., phonons. At thermal energies, the thermal scattering law (TSL), i.e., $S$($α, β$), captures crystal binding contributions to the total reaction rate, or cross section. This dimensionless material property describes the energy ($β$) and momentum ($α$) exchanges available in a medium. Currently, $S$($α, β$) is evaluated in the Full Law Analysis Scattering System Hub (FLASSH) code for discrete inputs and stored as ENDF/B File 7 for 0-phonon elastic (MT 2) and n-phonon inelastic (MT 4) processes. Further processing recasts $S$($α, β$) into cumulative distribution functions for sampling post-collision scattering kinematics. In practice, interpolation schemes are employed to access data between tabulated values. An improvement to this juncture of the nuclear data pipeline is supplying cross sections on-the-fly (OTF), as has been developed for the un-resolved resonance region to minimize non-physical interpolation errors. This capability may improve simulation accuracy for accident and transient analyses, where rapidly varying changes in temperature and pressure are difficult to predict beforehand. To do so, deep artificial neural networks (ANNs) can be employed which collapse non-linear, complex data into a lightweight dictionary of neural weights and biases. This has been successfully demonstrated for the hydrogen in light water $S$($α, β$) dataset in the form of a Neural Thermal Scattering (NeTS) module. In this work, the NeTS framework is extended to consider the impact of material-dependent dynamical features on optimal neural pre-processing and architecture design decisions, such as number of neurons per hidden layer, residual skip connections and neural depth. New NeTS modules for crystalline graphite and beryllium metal illuminate a novel correlation between dynamical nonlinearity and optimal neural parametrization when deploying $S$($α, β$) on-the-fly.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ATR NEXSHARE Fact Sheet

This fact sheet will provide information for the database of experimental facilities supported by the IAEA as outlined below: NEXSHARE: As part of the Nuclear Harmonization and Standardization Initiative (NHSI), the IAEA proposes to establish a Network for global cooperation and resource sharing for experiments and code validation between experimental facilities, SMRs design organizations, International Organizations and Technical Support Organizations (TSOs). The proposed Network, NEXSHARE, will be done in collaboration with the OECD/NEA. EXPERIMENTAL FACILITY DATABASE: As part of this activity, the IAEA is compiling a database of experimental facilities applicable to SMRs (including water cooled, high temperature gas cooled, molten salt and fast neutron spectrum reactors). This database will also form part of NEXSHARE and will be documented in an IAEA publication.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Insight into the Radiolytic Degradation Mechanism of TODGA

Partitioning and transmutation schemes, where americium is separated from other components of used nuclear fuel and burned in a fast neutron reactor, offer a path to dramatically decrease the space requirements for storage of nuclear fuel cycle wastes, allowing more efficient usage of potentially scarce storage resources.[1] Improvements in the efficiency of processes for separating americium from used nuclear fuels can thus have a significant impact on the costs of future nuclear fuel cycles. Understanding separation ligand radiation chemistry is important for development of new ligands and processes, as the ligand degradation products can have a deleterious effect on a separation process. Tetraoctyl diglycolamide (TODGA) has been studied as a ligand for lanthanide/minor actinide extraction for partitioning and transmutation schemes. However, the mechanism of initial energy transfer from products of solvent radiolysis to TODGA is still under debate. One proposed mechanism is attack by a n-dodecane radical cation resulting in electron abstraction and formation of a TODGA radical cation,[2] while later work has proposed hydrogen abstraction from sites adjacent to the central ether oxygen by other radical species.[3] The latter mechanism is proposed to result in degradation products from rupture of the ether bond. However, radiolytic degradation of TODGA in n-dodecane shows products that would result from rupture of all the backbone bonds. There is little information on what degradation products would be expected to form from the TODGA radical cation. In this work, we have begun to investigate this by examining the electron impact (EI) ionization mass spectrum of TODGA, which initially produces a TODGA radical cation in the gas phase. The EI spectrum shows fragments that would result from most backbone bond cleavages, similar to what is observed in the radiolysis of TODGA in n-dodecane. As the lifetime of the solvent radical cation decreases from n-dodecane to n-hexane,[4] irradiation of TODGA in n-hexane should have less solvent radical cations available for reaction with TODGA, and instead should favor reactions with longer-lived radicals. In contrast to n-dodecane, radiolytic degradation of TODGA in n-hexane produces only N,N-dioctylacetamide, which would result from rupture of the ether bond. The n-dodecane radical cation lifetime is long enough to produce TODGA radical cations but the n-hexane radical cation is mostly consumed before reaction with TODGA, so other radical processes that abstract hydrogen from TODGA dominate. This suggests that electron abstraction by the n-dodecane radical cation is an important component of TODGA radiolysis, although it does not rule out the presence of mechanisms involving hydrogen abstraction adjacent to the ether oxygen. References [1] J. Serp et al., Energies 2017, 10 (9), DOI: 10.3390/en10091445. [2] C. Zarzana et al., Solvent Extr. Ion Exch. 2015, 33 (5), 431–447, DOI: 10.1080/07366299.2015.1012885. [3] T. Koubský et al., Prog. Nucl. Energy 2017, 94, 208–215, DOI: 10.1016/j.pnucene.2016.07.010. [4] F. Sviridenko et al., Chem. Phys. Lett. 1998, 297 (3), 343–349, DOI: 10.1016/S0009-2614(98)01099-9.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

The Transient Reactor Test Facility (TREAT)

TREAT operated from 1959-1994, later refurbished & resumed operation in 2017 to support fuel safety testing Zircaloy-clad graphite/fuel blocks comprise core Virtually any power history possible within ~2000 MJ core transient energy capacity From milliseconds to minutes: Pulses, Ramps, LOCA Fuel motion monitoring system “hodoscope” observes fast neutrons emitted from specimens to track fuel relocation in real time Reactor also can be a neutron source to adjacent radiography facility Experiment vehicle does everything else Safety containment, specimen environment, and instrumentation

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Development of Two-Step Method for Fast Chloride MSR Neutronics

This work presents neutronics models of a small and large fast-spectrum molten chloride-salt reactor. The models are similar to designs being pursued by industry, and they may serve as generic preconceptual and simplified neutronics models that provide information for decision making in licensing-related areas. Here, the two models were created using Serpent, a Monte Carlo neutron transport code, and Moltres, a neutron diffusion core simulator tool. Specifically, this study focused on exploring the applicability of diffusion theory to fast molten salt reactor (MSR) models, the capabilities of an open-source, MSR-oriented simulation tool (Moltres), and optimal energy-group structures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Characterization of a boron-loaded deuterated liquid scintillator for fast and thermal neutron detection

The preparation and characterization of novel boron-loaded deuterated liquid scintillators (BLDLS) are presented for the first time to investigate the performance of a detector sensitive to both fast and thermal neutrons intended for future neutron spectroscopy measurements. Three deuterated toluene-based scintillation cocktails were produced, one without boron, one loaded with ortho-carborane (natural boron isotopic abundance), and one with 96 wt% 10B enriched ortho-carborane. General optical and material properties were analyzed to include the composition, density, index of refraction, photoluminescence emission spectra, and absorbance spectra in addition to the relative light yield, resolution, relative efficiency, and pulse-shape-discrimination performance for all three scintillators and a hydrogen-based EJ-309 scintillator to evaluate the effects of boron loading in comparison with a common commercially-available scintillator and feasibility for future neutron spectroscopy measurements. Finally, the detection of shielded neutron sources was explored to capitalize on the 10B thermal neutron capture reaction. The results showed loading the scintillators with boron deteriorated the PSD performance slightly compared to the unloaded deuterated scintillator and resulted in a relative light yield loss of 26% and 41% for the 10B enriched and natural boron BLDLS, respectively. Lastly, the BLDLS have a higher relative detection efficiency than the unloaded deuterated scintillator for shielded neutron sources, based on total count rate, a desirable characteristic for missions related to finding and characterizing illicitly-trafficked special nuclear material.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2023

Adaptation of Temperature Dependent Thermal Neutron Libraries for Fast On-The-Fly Monte Carlo Sampling at Arbitrary Temperatures

Thermal neutron scattering data is necessary for reactor simulations of thermal reactors, but the data storage required can be unwieldy. This is exacerbated when simulations require a large temperature range, as is the case for accident scenarios and reactor start-up. Previous work conducted at RPI has explored and demonstrated a proof of concept of On-The-Fly (OTF) data storage with light water and crystalline graphite based on ENDF/B-VII.1. This provides an updated data storage paradigm for incoherent inelastic cross-sections that allow the data to be used to generate necessary sampling Cumulative Density Functions (CDFs) for any desired temperature. The work presented in this paper extends the previous work by implementing the proof of concept based on ENDF/B-VIII.0, instead of ENDF/B-VII.1, and to include additional materials, such as beryllium and oxygen in beryllium oxide and the new reactor grades of porous graphite. Completion of this work allows a single thermal data file to represent any thermal scattering event for a given material OTF for the temperature of interest, which greatly reduces the data storage needed with only a marginal cost in the calculation of the CDFs OTF. This paper will review the basic methodologies in generating the OTF data, as well as present the updates and improvements made in the generation process. Application of the methodologies covered in this paper show a significant reduction in the storage requirements for inelastic incoherent scattering events. In which, the evaluated ENDF/B-VIII.0 thermal scattering library can be reduced from 563.1 MB to 10.9 MB for light water and 339.3 MB to 13.6 MB for Be in BeO.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Preliminary Evaluation of Driver Fuel Plates and Reduced Sodium Temperature to Increase Fast Test Reactor Experimental Neutron Irradiation Capabilities

All sodium-cooled fast test reactors operated so far have used essentially the same fuel geometrical ar-rangement, namely cylindrical fuel pins assembled into a triangular array in hexagonal fuel assemblies. Fuel compositions and dimensions may vary but the overall geometrical pattern remains identical. Fur-thermore, most sodium-cooled reactors have operated, and still operate, with core-average inlet and outlet sodium temperatures of, respectively, 350-400°C and 500-550°C. The present preliminary study revisits the standard fuel assembly geometry as well as operating tempera-tures to evaluate potential impact on fast test reactor neutron irradiation capabilities. The reference fuel is U-20Pu-10Zr. Pin (reference) and plate geometries are considered together with two sets of core-average sodium inlet/outlet temperatures: 350/500°C (reference) and 200/400°C. More prototypic environments could be obtained inside ad-hoc irradiation testing vehicles that are independent from the reactor primary coolant. Everything else being the same, fuel temperature is significantly lower in plates than in pins. Therefore, plates could in principle accommodate higher plutonium content, thus, reducing the need for uranium enrichment or allowing the use of lower quality plutonium (U-xPu-10Zr thermal conductivity and solidus temperature decrease as x increases, hence, necessitating additional thermal margins). For the reference pin configuration, lowering the inlet/outlet sodium temperatures from the reference 350/500°C down to 200/400°C provides additional thermal margins that can be used to increase the peak fast flux from about 4.5 × 10 15 n/cm 2 -s to 6 ×10 15 n/cm 2 -s for the same core power of 300 MW. Assuming 300 Equivalent Fuel Power Days (EFPD) of operation per calendar year, a (steel) test article could accu-mulate up to 75 dpa/year. The use of fuel plates provides even more thermal margins which may allow the peak fast flux to reach values as high as 8 × 10 15 n/cm 2 -s for both sets of inlet/outlet temperatures. In this environment, a test article could accumulate up to 95 dpa/year assuming 300 EFPD/year, hence greatly accelerating irradiation testing. Allowing a fast test reactor to operate over a wide range of inlet/outlet temperatures could add significant flexibility to its neutron irradiation capabilities. Steel present in the driver fuel assemblies is not expected to accumulate more than 90 displacements per atom (dpa) over its lifetime. Hence, steel alloys that are not suited for very high dpa but can operate over a wide range of temperatures (from low to high) would be appropriate for this application. Finally, because pin and plate assemblies fit on the same grid plate, a fast test reactor could in principle start-up with standard fuel pin assemblies and, later on, move to a plate-type Mark-II fuel to further in-crease its neutron irradiation capabilities while potentially reducing the level of uranium enrichment needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Neutron Response of the EJ-254 Boron-Loaded Plastic Scintillator

Organic scintillators doped with capture agents provide a detectable signal for neutrons over a broad energy range. This work characterizes the fast and slow neutron response of EJ-254, an organic plastic scintillator with 5 loading by weight. For fast neutrons, the primary mechanism for light generation in organic scintillators is n-p elastic scattering. To study the fast neutron response, the proton light yield of EJ-254 was measured at the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory. Using a broad-spectrum neutron source and a double time-of-flight technique, the EJ-254 proton light yield was obtained over the energy range of approximately 270 keV to 4.5 MeV and determined to be in agreement with other plastic scintillators comprised of the same polymer base. To isolate the slow neutron response, an AmBe source with polyethylene moderator was made incident on the EJ-254 scintillator surrounded by an array of EJ-309 observation detectors. Events in the EJ-254 target coincident with the signature 477.6 keV γ ray (resulting from de-excitation of the residual 7Li nucleus following boron neutron capture) were identified. Pulse shape discrimination was used to evaluate the temporal differences in the response of EJ-254 scintillation signals arising from γ-ray and fast/slow neutron interactions. Clear separation between γ-ray and fast neutrons signals was not achieved and the neutron capture feature was observed to overlap both the γ-ray and fast neutron bands. Taking into account the electron light nonproportionality, the neutron capture light yield in EJ-254 was determined to be 89.4±1.1 keVee.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Fast and Thermal Neutron Spectrum Dosimetry Measurements in the Advanced Test Reactor large-B and small-I Positions Following the Sixth Core Internals Change-out [Slides]

The Advanced Test Reactor (ATR) has a wide variety of irradiation positions that have had experiments that were developed by users from around the world. Most experiment irradiations rely on thoroughly benchmarked numerical models. However, some irradiation positions in ATR are not as well- characterized and are complicated by spectral perturbations from control cylinder orientation. The “small I” and “large-B” irradiation positions are located nearby control cylinders and suffer from these flux perturbations from control cylinder orientations that change during an irradiation cycle to maintain the desired core power distribution. Models predicted that these types of position would exhibit both spectral shifts and amplitude changes in neutron flux, but few measurements have been conducted to benchmark these predictions. Recently, requalification testing was performed to confirm the operational readiness of the ATR following the completion of the Core Internals Change-out (CIC). These tests provided a unique opportunity to validate the analytical methods used to simulate the ATR because nearly all components in the reactor were in a clean as-built state, significantly reducing modelling uncertainties. One subset of the testing included characterization of the fast and thermal neutron flux in the “small-I” and “large-B” positions using silver, cobalt, and nickel neutron dosimetry. In contrast to typical irradiation cycles, the control cylinders were held in position during the post-CIC nuclear testing. The specific activity of these dosimeter wires was measured following two low-power tests with different control cylinder positions.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2024

A317_Fast and Thermal Neutron Spectrum Dosimetry Measurements in the Advanced Test Reactor large-B and small-I Positions Following the Sixth Core Internals Change-out

The Advanced Test Reactor (ATR) has a wide variety of irradiation positions that have had experiments that were developed by users from around the world. Most experiment irradiations rely on thoroughly benchmarked numerical models. However, some irradiation positions in ATR are not as well- characterized and are complicated by spectral perturbations from control cylinder orientation. The “small I” and “large-B” irradiation positions are located nearby control cylinders and suffer from these flux perturbations from control cylinder orientations that change during an irradiation cycle to maintain the desired core power distribution. Models predicted that these types of position would exhibit both spectral shifts and amplitude changes in neutron flux, but few measurements have been conducted to benchmark these predictions. Recently, requalification testing was performed to confirm the operational readiness of the ATR following the completion of the Core Internals Change-out (CIC). These tests provided a unique opportunity to validate the analytical methods used to simulate the ATR because nearly all components in the reactor were in a clean as-built state, significantly reducing modelling uncertainties. One subset of the testing included characterization of the fast and thermal neutron flux in the “small-I” and “large-B” positions using silver, cobalt, and nickel neutron dosimetry. In contrast to typical irradiation cycles, the control cylinders were held in position during the post-CIC nuclear testing. The specific activity of these dosimeter wires was measured following two low-power tests with different control cylinder positions.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2025

Feasibility of Recycling Discharged Microreactor Heavy Metal in Light Water and Sodium-Cooled Fast Reactors: A Neutronics Analysis

Nuclear microreactors (MRs) offer unique advantages, such as rapid deployment, potability, low maintenance requirements, and operational flexibility. Their compact size makes them a promising solution for decentralized power generation, particularly in remote areas, military bases, and disaster-stricken regions. However, MRs face challenges, including unutilized fissile material at the end of life, economic inefficiency, increased heavy metal (HM) waste complicating disposal, and the accumulation of plutonium (Pu) with high 239 Pu concentrations raising proliferation risks. Here, this study investigated the neutronics feasibility of a novel three-stage fuel cycle where discharged HM from MRs is recycled and burned in light water reactors and sodium-cooled fast reactors. This approach converts discharged HM into valuable fuel, enhancing the efficiency of MR deployments while improving the safeguardability of their final waste products. Neutronics analysis demonstrated that the safety characteristics of reactor designs in each stage were minimally impacted by the proposed cycle. For two representative MR designs, a fast-spectrum MR with solid pellet fuel and a thermal-spectrum MR with TRISO (TRi-structural-ISOtropic) fuel compacts, the proposed fuel cycle reduced the uranium disposal mass flow rate by ~60%, decreased the 235 U enrichment of the discharge fuel to ~1 wt%, eliminated plutonium disposal, and increased the cumulative fuel burnup to ~580 gigawatt-day per metric ton of initial heavy metal (GWd/t-iHM) or 60% fissions per initial metal atom. Despite the significant differences between the two MR designs, the performance and infrastructure requirements of the developed fuel cycles were remarkably similar, indicating its generalizability to a broader class of MRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗