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This report describes the regulatory bases for the 2025 revision of Army Regulation (AR) 50–7 - Nuclear and Chemical Weapons and Materiel: Army Reactor Program and the new associated Department of the Army pamphlet (DA PAM) Army Reactor Program Procedures.
The Advanced Materials and Manufacturing Technologies (AMMT) program previously demonstrated an additively manufactured (AM) irradiation capsule (commonly referred to as a “rabbit”) from 316H stainless steel (SS) for insertion into the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL)1. This report details efforts to design and fabricate an AM pressure limiting structure (PLS) into one of the end caps of a rabbit capsule and qualify it for insertion into HFIR. The PLS includes a thin cylindrical rupture wall, a shield, and internal supports to facilitate printing and ensure mechanical integrity. Its overall dimensions are 9-mm tall and 10-mm in diameter— equivalent to about one-fourth of the size of a AAA battery. The PLS maintains safe internal operating pressures for a rabbit capsule while in the reactor. Although this application is specific to HFIR, the approach lends itself to further applications in industrial, aeronautical, advanced space and power generation environments. Several PLS rabbits capsules have been successfully designed, fabricated, pressure tested, and qualified for future insertion into the HFIR for irradiation and post-irradiation evaluation.
This report summarizes work performed under DOE Award DE-EE00010149 to support the reliable operation of a community-scale biochar reactor system in Kivalina, Alaska. The project focused on improving sanitation and waste management in a remote community by assessing the installed system, identifying spare parts, defining key performance indicators (KPIs), preparing operator and maintenance manuals, and developing mobile reporting tools for operational data and KPI tracking. The team also produced training materials and recorded videos to support operator onboarding and continuity. The project demonstrated progress in system readiness, documentation, and digital reporting, while also identifying challenges common to remote deployments, including travel constraints, upstream system failures, and local resource limitations. This work provides a practical framework for improving the operation, monitoring, and future replication of biomass reactor systems in remote communities.
Idaho National Laboratory (INL) recently conducted a workshop endorsed by the National Nuclear Security Administration (NNSA) under the auspices of the Defense Nuclear Nonproliferation R&D (DNN R&D) Office's Safeguards Portfolio. The workshop's primary objective was to foster an engaging dialogue among researchers, with a specialized emphasis on the safeguards pertaining to molten salt reactors. At INL, the installation of the state-of-the-art Molten Salt Thermophysical Examination Capability (MSTEC) is underway. This shielded argon glovebox facility, designed for both irradiated and non-irradiated actinide materials, represents the cutting edge of research infrastructure. MSTEC is poised to serve as a pivotal research platform in the realm of molten salt technology, with significant implications for safeguards applications. Participants of the workshop had the opportunity to tour the facilities, including the site where MSTEC is being installed, as well as to observe the INL's molten salt and pyroprocessing research hot cells. The event featured insightful presentations delving into molten salt chemistry and the MSTEC project. Each participating laboratory contributed to the discourse with presentations on their respective research efforts addressing safeguards in relation to molten salt reactors. The workshop culminated with a generative brainstorming session, where participants shared their thoughts on strategic integration with partner agencies, aiming to synergize efforts in advancing the field of nuclear safeguards.
Advanced nuclear systems for terrestrial microreactors and space applications require moderator materials with high thermal stability, hydrogen retention, and efficient neutron moderation. Yttrium hydride (YH x ) is a promising candidate, but its irradiation performance and hydrogen stability remain insufficiently understood. To address this, Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory, supported by the US Department of Energy’s Nuclear Science User Facilities program, have initiated an irradiation campaign in ORNL’s High Flux Isotope Reactor. The experiment employs six irradiation capsules containing YH x specimens (H/Y ≈ 1.9) designed for three target temperatures (300°C, 400°C, and 500°C) and two distinct neutron fast fluence levels. This report documents the irradiation test matrix, capsule design, specimen precharacterization, and experiment readiness for insertion. Planned postirradiation examinations will provide new insights into hydrogen retention and microstructural stability, advancing the understanding of YH x as a moderator for compact reactor applications.
We aimed to develop a catalytic membrane reactor (CMR) that couples water–gas shift (WGS) reaction with in-situ H 2 separation to produce blue H 2 and a CO 2 -rich stream from coal- and biomass-derived syngas. Our core approach employs high-permeance carbon molecular sieve (CMS) hollow-fiber membranes (HFMs) with strong gas separation ability, which we further integrate into catalytic membrane reactors to intensify H 2 production and CO 2 capture in a single unit. Three organizations with complementary skills collaborated to achieve the goal, including the University at Buffalo (UB), Los Alamos National Laboratory (LANL), and Trimeric Corporation (Trimeric).
In addition to producing electricity, advanced nuclear reactors can also serve as a source of process heat for a wide range of industrial or residential applications. In cases where the heat application requires low-temperature heat, waste heat from the balance of plant (BOP) can be utilized with minimal impact on electricity generation. For applications requiring higher-temperature heat, heat can be extracted from the BOP, though this will reduce the amount of electricity generated. In either case, the optimal BOP design will be case-specific and depend on multiple factors, including the reactor type and the thermal requirements of the heat application.
The Advanced Fuels Campaign (AFC) has initiated the Advanced Reactor Cladding (ARC) irradiation campaign to generate irradiation performance data for candidate fuel cladding concepts. The campaign includes a diverse set of ferritic/martensitic steels, oxide dispersion strengthened (ODS) alloys, FeCrAlbased alloys, coated materials, and welded cladding specimens produced through multiple US Department of Energy (DOE) programs and international collaborations. Three complementary experimental thrusts comprise the campaign: tensile testing (ARC Tensile) to rapidly screen candidate alloys, fracture toughness testing (ARC Fracture) to evaluate irradiation effects on crack resistance, and tubular weld testing (ARC Weld) to quantify irradiation-induced changes in the mechanical performance of end cap welds. This report documents the irradiation campaign design, including the selected materials, specimen types, irradiation matrix, and capsule designs for irradiation within the High Flux Isotope Reactor (HFIR). A total of 14 irradiation capsules were developed to achieve target irradiation temperatures between 300°C and 600°C and doses up to 30 dpa. Thermal analyses were performed using finite element methods to establish capsule geometries capable of achieving the desired specimen temperatures while accommodating differences in specimen geometry and material properties. The resulting capsule designs provide the basis for irradiation of the AFC-ARC experimental matrix and subsequent post-irradiation examination to assess the effects of neutron irradiation on advanced cladding materials.
Solvolytic extraction of lignin from biomass is a critical step in lignin-first biorefining, including the reductive catalytic fractionation (RCF) process. Key to optimal RCF processing is the ability to rapidly extract lignin from biomass at high delignification extents and transfer the lignin molecules to a catalyst surface in a time frame that minimizes lignin condensation reactions. Here, we use a flow-through reactor to study the effects of temperature (175–250 °C), residence time (9 to 36 min), and solvent composition (methanol and methanol–water) on lignin extraction and condensation. We evaluated three metrics at each condition: total delignification, delignification rate, and extent of condensation, the latter measured by a decrease in monomer yield for batch hydrogenolysis reactions of solvolysis liquor compared to batch RCF reactions. We observe that delignification is predominantly determined by temperature, while residence time dictates the lignin condensation extent. Moreover, the extent of both extraction and condensation increased in the methanol–water solvent system compared to that in the methanol system. Lignin extracted in methanol is stable up to 18-min residence times at or below 225 °C, while a majority of the lignin extracted in methanol–water is condensed with a 9-min residence time at 200 °C. These results can inform reactor designs and solvent selection for lignin-first biorefining processes that aim to physically separate the biomass and catalyst.
The research presented in this article describes progress in applying stochastic methods, uncertainty quantification, parametric studies, and variance-based sensitivity analysis (also known as Sobol sensitivity analysis) to a full-core model of a nuclear thermal propulsion (NTP) system simulated via the radiation transport code Griffin to simulate neutronics. Our goal is to develop a reduced-order (surrogate) model that can be rapidly sampled with perturbations to multiple input parameters. In this NTP system, reactivity and power feedback affect the rotation of control drums (CDs), which is itself controlled by a hybrid proportional-integral-derivative (PID) controller actuated by the power demand and reactivity feedback from the numerical model. This model uses reactor kinetic feedback (mean generation time [Λ] and effective delayed neutron fraction [ β eff ] from a transient Griffin simulation executed via Griffin’s improved quasi-static solver to provide the kinetic parameters) as inputs to functions that control the CD rotation angle. By investigating numerous stochastic approaches, we developed a dual-purpose surrogate model of the NTP system, using polynomial regression in the Multiphysics Object-Oriented Simulation Environment (MOOSE) Stochastic Tools Module (STM). The trained model can be rapidly sampled while simultaneously perturbing various input parameters, such as coefficients on the PID control or temperature (directly affecting the neutron cross section). The surrogate model delivers accurate (within 5%) results at speeds orders of magnitude faster (minutes, not days of computational time) than the base model. Once the surrogate model has been trained, distributions of the uncertain parameters can be changed at will to investigate the effects of perturbing multiple inputs as well as the effects of these inputs on the model output. For example, coefficients used in the PID control system may vary due to some type of physical interference, or uncertainty may exist in the temperature of the neutron cross sections in various regions of the reactor. A distribution can be placed on these parameters, and operational boundaries can be determined. The goal of this work is to support development of an advanced control system for operating CDs in a functioning NTP system. This work is a scoping study of the MOOSE STM.
Properties of high temperature materials for nuclear rocket reactor application - reactor concepts for low power lightweight nuclear power plant
Diode and reactor design and nuclear fuels including uranium carbide alloys, uranium dioxide and uranium dioxide cermets for industry proposed in-pile thermionic space reactors
Gas core reactor, discussing application of fissionable gas to heat hydrogen propellant in nuclear rocket reactor
A numerical study is performed in order to assess the performance of biomass pyrolysis reactors which utilize direct particle-wall thermal conduction heating. An idealized reactor configuration consisting of a flat-plate turbulent boundary layer flow with particle convection along the heated wall and incorporating particle re-entrainment is considered.
This report identifies the considerations related to the operation of space nuclear power reactors that would be utilized for In Situ Resource Utilization (ISRU) missions on the lunar surface. The objective of this study is to evaluate the operational complexities of reactors that would provide electrical and thermal power to a growing lunar ISRU and mining outpost/base.
A U.S. governmental interagency space reactor standards working group (SWG) was convened in April 2021 with a memo from the NASA Chief Engineer to address the limited standards and regulations in place specifically for space reactor design and safety. The convening memo requested support from the Standards Executives of six other federal agencies. The Standards Executives agreed to support a 6-month study to assess the need for consensus standards and identified SWG representatives from each agency. This report summarizes the SWG’s study, their findings, and their recommendations.
NASA aims to deploy a fission reactor on the lunar surface by 2030 to generate 100 kWe of power for at least 10 years without maintenance or refueling. To assess system and subsystem survivability according to those requirements, a model of the reactor pallet was created in the radiation transport code MCNP to determine the neutron fluences and gamma doses that the materials and systems would be exposed to. The model was based on the 2025 government design. The model indicated that the control drum motors, Brayton cycle engines, and pipes and valves would be exposed to neutron fluences of 1×10 15 – 1×10 16 n/cm 2 per year and gamma doses ranging from 10 to 500 Mrad per year. The electronics box housing the instrumentation and control elements is expected to receive 1×10 14 – 1×10 15 n/cm 2 per year and gamma doses in the range of 1–5 Mrad per year. These doses necessitate that the control drum motors, Brayton cycle engines, valve components, control electronics, and other components be radiation-hardened (>10 16 n/cm 2 and >10 Mrad gamma dose) or additionally shielded to survive the 10-year lunar mission.
When a neutron born in fission thermalizes to the order of $k$ $B$ $T$, it’s de-Broglie wavelength and energy approach the order of inter-atomic spacing and elementary lattice oscillations, respectively. $S$($a,β,t$) or the scattering law, uuantify these temperature-dependent crystallographic contributions to total cross section (or reaction rate). In a Monte Carlo analysis, cumulative distribution functions (CDFs) of $S$($a,β,t$) are loaded to memory from “A Compact ENDF” (ACE) files for stochastically selecting thermal scattered neutron trajectories. In this work, novel neural thermal scattering (NeTS) modules for $S$($a,β,t$) CDFs are designed, trained, serialized and embedded within SERPENT using Python’s limited C-API for on-the-fly deployment of crystalline graphite $S$($a,β,t$) sampling. Torchscript tracing and Numba just-in-time (JIT) compilation streamline neural inference on NVIDIA GPUs with CUDA libraries. Demonstrations of bare sphere thermalization of fast and thermal sources show excellent agreement between embedded NeTS in SERPENT and MCNP. With an explicit model of the reactor, NeTS can predict on-the-fly changes in TREAT neutron spectra as a function of local temperature, which can serve to improve transient and accident predictions in a multiphysics analysis framework. This framework can be further extended to account on-the-fly for changes in local graphitic microstructure to scattering cross sections, and outlines a novel coupling of modern machine learning with state-of-the-art reactor physics methods.