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The ongoing Nuclear Renaissance includes many next generation reactor designs and testing, this means translates to increased demand for Health Physicists. These opportunities are not limited to just reactors but the entire fuel cycle from mining and extraction from existing tailings, R&D support efforts at new start-up companies, to shielding modeling and design, to Recycling and Reprocessing, and eventually Decommissioning.
As we transition toward renewable energy resources and electrification, there will be an increasing demand for critical materials, including copper. While copper is currently mined in the US, processing capacity is not sufficient, and intermediate mining products are shipped to Asia for further processing. The goal of this research was to develop a transformative hydrometallurgical process for the production of copper from low-grade ores that would eliminate the need for smelting and would increase the domestic processing capacity in the US, The project initially focused on the electrochemical reduction of copper concentrate using vanadium, followed by the biological oxidation to produce a stream compatible with existing solvent extraction and electrowinning operations. We discovered an efficient electrochemical process that could produce copper salts from concentrate without the need for biological oxidation, and this technology has been licensed and spun-off into a start-up company. The microbes involved in the current state-of-the-art bioleaching processes were genetically modified to introduced a number of new traits, including increased sulfur oxidation, salt-tolerance, and binding of other critical metals such as cobalt, molybdenum, rhenium, and the rare earth elements. The project was extended to also explore the electrochemical oxidation of copper concentrate using cerium which would potentially eliminate production of hydrogen sulfide that occurs with the reductive leaching process. This technology was found to have slower kinetics, however could it still be developed as an alternative process for domestic copper production and has been found to be applicable to other critical minerals.
This report presents the development and validation of an initial steady-state multiphysics capability for molten salt reactors (MSRs) under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in Fiscal Year 2025. The framework integrates neutronics, thermal-hydraulics, species transport, and thermochemistry to capture the coupled dynamics of liquid-fueled systems. Implementation and testing were performed on two representative designs: the Molten Salt Reactor Experiment (MSRE), a thermal-spectrum, channeled-flow reactor, and the Lotus Molten Salt Reactor (L-MSR), a fast-spectrum, pool-type reactor. The modeling suite employs Griffin for reactor physics and depletion, Pronghorn and SAM for thermal-hydraulics, Thermochimica for chemistry, and Saline for thermophysical properties, with benchmarking and validation carried out against historical MSRE data, experimental flow-loop measurements, and reference depletion calculations from Monte Carlo codes. The framework demonstrated the ability to reproduce key reactor behaviors including temperature feedback, reactivity losses, delayed neutron precursor transport, xenon poisoning, and redox potential evolution. The results confirm the feasibility and accuracy of the coupled models in predicting steady-state and selected transient MSR behaviors. This latter ones are used in this report as a proxy indicating that the steady-state models from which the transient starts are accurate. For MSRE, validation showed good agreement with pump start-up and natural circulation tests, while for the L-MSR, benchmarking confirmed hydraulic calibration and consistency of neutronics–thermal coupling. The tools also provided new insights into species transport, noble metal deposition, and salt solidification dynamics. On the Xenon transport front, the code is validated against the steady state Xenon poisoining measurement and showed good agreement with the experimental value. Identified areas for future work include advanced void transport modeling, three-dimensional simulations, improved alloy corrosion models, and tighter integration with high-fidelity Monte Carlo codes. These developments provide a foundation for high-fidelity MSR simulations that can support reactor design optimization, safety assessments, and long-term operational strategies.
Lock-out tag-out (LOTO) instructions protect personnel from unex-pected re-energization or start-up of equipment that is under maintenance. Even though there is rigorous training associated with LOTO activities, incidents still occur. Idaho National Laboratory is researching and developing dynamic LOTO instructions, which supports automatic place keeping, personnel working on mul-tiple instructions, and multiple work groups completing the same instructions. Researchers identified user needs through subject matter expert interviews, ob-servations, and classroom training. Failing to identify potential hazards during LOTO can cause injury or death to personnel. Dynamic instructions help to guide the user seamlessly through the instructions. Combining planning and record keeping may lead to clearer, more applicable instructions which addresses the reason for compliance issues with current paper-based procedures.
The purpose of this ECAR is to structurally evaluate the MARVEL reactor support frame during the dry critical (DryCrit) start-up testing. The ECAR includes evaluation of the DryCrit configuration for tip-over of the frame during a seismic event, floor loading of the frame on the MFC-720 (TREAT) operating floor, stress analysis of the adapter ring, and stress analysis of the frame itself for gravity and seismic loads.
Polymer electrolyte membrane fuel cells have durability limitations associated with the startup and shutdown of the fuel cell, which is critical for real-world vehicle commercialization. During startup or shutdown, there exists an active region (hydrogen/air) and a passive region (air/air) between the cell inlet and outlet. Internal currents are generated in the passive region causing high-potential excursion in the cathode leading to accelerated carbon corrosion. In this study, a segmented cathode hardware is used to evaluate the effect of platinum loading on both cathode and anode, and carbon support material on degradation due to repeated series of startups or shutdowns. Here, in situ losses in the performance and electrochemical surface area were measured spatially, and ex situ analysis of the catalyst layer thickness and platinum particle size was performed to understand the effect of startup or shutdown on different membrane electrode assembly materials. Startup degrades the region near anode outlet more, while shutdown degrades the region near anode inlet more compared to the rest of the electrode. While various system mitigation strategies have been reported in the literature to limit this degradation, one materials mitigation strategy is to limit the anode's ability to reduce oxygen to water through increasing the ratio of platinum loading in the cathode to the anode, or by using a bi-functional catalyst.
We report that the physical mechanisms of the ohmic breakdown in a tokamak have been understood based on the classical Townsend avalanche theory. However, a new systematic theory (Yoo et al 2018 Nat. Commun.9 3523) recently demonstrated that electron avalanches during the ohmic breakdown are completely different from the Townsend avalanche due to strong self-generated electric fields. In this study, we elucidate the multi-dimensional effects of the self-generated electric field on plasma dynamics during the ohmic breakdown. We also propose a novel electromagnetic topology analysis method that can easily predict the overall plasma behavior and where the main plasma is generated. The topology analysis method is validated by a state-of-art particle simulation for various magnetic configurations. New physical insights into the complex electromagnetic topology would facilitate designing more reliable and optimized ohmic breakdown scenarios in future tokamaks, such as ITER and beyond.
The National Spherical Torus Experiment Upgrade (NSTX-U) will advance the physics basis required for achieving steady-state, high-beta, and high-confinement conditions in a tokamak by accessing high toroidal field (1 T) and plasma current (1.0 - 2.0 MA) in a low aspect ratio geometry (A = 1.6 - 1.8) with flexible auxiliary heating systems (12 MW NBI, 6 MW HHFW). This paper describes progress in the development of L- and Hmode discharge scenarios and the commissioning of operational tools in the first ten weeks of operation that enable the scientific mission of NSTX-U. Vacuum field calculations completed prior to operations supported the rapid development and optimization of inductive breakdown at different values of ohmic solenoid current. The toroidal magnetic field (BT0 = 0.65 T) exceeded the maximum values achieved on NSTX and novel long-pulse L-mode discharges with regular sawtooth activity exceeded the longest pulses produced on NSTX (tpulse > 1.8s). The increased flux of the central solenoid facilitated the development of stationary L-mode discharges over a range of density and plasma current (Ip). H-mode discharges achieved similar levels of stored energy, confinement (H98y,2 > 1) and stability (βN/βN-nowall > 1) compared to NSTX discharges for Ip ≤ 1 MA. High-performance H-mode scenarios require an L-H transition early in the Ip ramp-up phase in order to obtain low internal inductance (li) throughout the discharge, which is conducive to maintaining vertical stability at high elongation (κ > 2.2) and achieving long periods of MHD quiescent operations. The rapid progress in developing L- and H-mode scenarios in support of the scientific program was enabled by advances in real-time plasma control, efficient error field identification and correction, effective conditioning of the graphite wall and excellent diagnostic availability.
Cross-seam high voltage DC (HVDC) transmission in the United States can provide an efficient and flexible bulk power delivery highway between the interconnections. Recent studies have shown significant economic benefits of cross-seam HVDC transmission systems by taking the advantages of load diversity, frequency response, renewable energy diversity, and energy arbitrage in the different interconnections. This paper introduces a station-hybrid system scheme for cross-seam interconnections. Here, the proposed station-hybrid system combines the advantages of line commutated converter (LCC) and voltage source converter (VSC) technologies, thus enabling reliable and flexible bidirectional power flows across the interconnections. Control strategies of the station-hybrid system are investigated with a focus on power flow reversal control under normal operation conditions and grid support control under emergency conditions. The feasibility and effectiveness of the proposed station-hybrid system and corresponding control strategies are verified by simulations in PSCAD/EMTDC.
The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.
A reduced semi-empirical model using time-dependent axisymmetric vacuum field calculations is used to develop the prefill and feed-forward coil current targets required for reliable direct induction (DI) startup on the new MA-class spherical tokamaks, MAST-U and NSTX-U. The calculations are constrained by operational limits unique to each device, such as the geometry of the conductive elements and active coils, power supply specifications and coil heating and stress limits. The calculations are also constrained by semi-empirical models for sufficient breakdown, current drive, equilibrium and stability of the plasma developed from a shared database. A large database of DI startup on NSTX and NSTX-U is leveraged to quantify the requirements for achieving a reliable breakdown (Ip ~ 20 kA). It is observed that without pre-ionization, STs access the large E/P regime at modest loop voltage (Vloop) where the electrons in the weakly ionized plasma are continually accelerating along the open field lines. This ensures a rapid (order millisecond) breakdown of the neutral gas, even without pre-ionization or high-quality field nulls. The timescale of the initial increase in Ip on NSTX is reproduced in the reduced model provided a mechanism for impeding the applied electric field is included. Most discharges that fail in the startup phase are due to an inconsistency in the evolution of the plasma current (Ip) and equilibrium field or loss of vertical stability during the burn-through phase. The requirements for the self-consistent evolution of the fields in the weakly and full-ionized plasma states are derived from demonstrated DI startup on NSTX, NSTX-U and MAST. The predictive calculations completed for MAST-U and NSTX-U illustrate that the maximum Ip ramp rate (dIp/dt) in the early startup phase is limited by the voltage limits on the poloidal field coils on MAST-U and passive vertical stability on NSTX-U.
Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling and simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact not conservative and could be overestimating reactivity feedback effects that are closely tied to reactor safety. We conclude that there is indeed value in performing direct simulation of deformed meshes despite the increased computational expense. PROTEUS-SN is already part of the SHARP multi-physics toolkit where both thermal hydraulics and structural mechanical feedback modeling can be applied but this is the first comparison of direct simulation to legacy techniques for radial core expansion.
Operational characteristics represent the set of responses that a nuclear system exhibits during normal operation. Operators rely on this behavior to assess the status of the system and to predict the consequences of off-normal events. These characteristics largely refer to the relationship between power and system operating conditions. The static and dynamic behavior of a chain-reacting system, operating at sufficient power, is primarily governed by reactivity effects. The science of reactor physics has identified and evaluated a number of such effects, including Doppler broadening and shifts in the thermal neutron spectrum. Often these reactivity effects are quantified in the form of feedback coefficients that serve as coupling coefficients relating the neutron population and the physical mechanisms that drive reactivity effects, such as fissile material temperature and density changes. The operational characteristics of such nuclear systems usually manifest themselves when perturbations between system power (neutron population) and system operating conditions arise. Successful operation of such systems requires the establishment of steady equilibrium conditions. However, prior to obtaining the desired equilibrium (steady-state) conditions, an approach from zero-power (startup) must occur. This operational regime may possess certain limiting system conditions that must be maintained to achieve effective startup. Once steady-state is achieved, a key characteristic of this operational regime is the level of stability that the system possesses. Finally, a third operational regime, shutdown, may also possess limiting conditions of operation that must be maintained. This report documents the operational characteristics of a “generic” Accelerator Driven Fissile Solution (ADFS) system during the various operational regimes of startup, steady-state operation, and shutdown. Typical time-dependent behavior for each operational regime will be illustrated, and key system parameters, such as response times, will be quantified. A generalized linear systems analysis of steady-state operations will be performed to evaluate the level of stability of ADFS systems. This information should provide a basic understanding of typical ADFS system operational behavior, and facilitate the development of monitoring procedures and operator aids.
NREL will develop meshing and modelling scripts to automate exploration of the design space for Maiden Wave Energy's Wave Energy Converter ("WEC") device. This will aid Maiden Wave Energy to assess a range of different device configurations and explore how different variables in the system affect the concept's performance.
In this work, we present validation test results of fully coupled neutronics and thermal-hydraulics models of the Molten Salt Reactor Experiment (MSRE) against experimental data of the zero power pump transients and the natural circulation tests at low power. To capture the strong coupling between neutronics and thermal-hydraulics due to fuel circulation, and to account for the delayed neutron precursor (DNP) distribution, the porous media thermal-hydraulics solver Pronghorn was fully coupled to the spatial neutron dynamics code Griffin, which solves the neutron diffusion equation, and to the 0-D point kinetics solver Squirrel, using a 2-D homogenized representation of the MSRE. The validation test results show very good agreement with experimental data for both point kinetics and spatial dynamics simulations, capturing the strong feedback effect and DNP losses in the MSRE. The 0-D code Squirrel accurately predicted the time-dependent behavior in the MSRE given the steady-state spatial dynamics solution of Griffin.
The Tropospheric Emission Spectrometer (TES) is an infrared, high spectral resolution Fourier transform spectrometer with a 3.3 to 15.4 micron wavelength coverage. TES is a scanning instrument intended for determining the chemical state of the Earth's lower atmosphere (troposphere) from the surface to 30+ km. TES produces vertical profiles of important pollutant and greenhouse gases such as carbon monoxide, ozone, methane, and water vapor on a global scale every other day. TES was launched into orbit onboard NASA's earth Observing System Aura spacecraft on July 15, 2004 from Vandenberg Air Force Base, California.
The operation and maintenance of nuclear power plants, the nuclear fuel cycle in general, research laboratories and pharmaceutical, medical and industrial facilities generate low-level radioactive waste which, along with the historical radioactive waste from past nuclear activities, needs to be treated and stored, awaiting final disposal. Plasma technology offers a very effective way of treating this waste with a high volume reduction factor (VRF), free from organics, liquids and moisture, and meets without a doubt the acceptance criteria for safe storage and disposal. By means of a plasma beam of approximately 5000 deg. C, the inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes while the organic material is gasified and afterwards oxidized in an afterburner and purified in an off-gas cleaning system. This paper describes the principles of plasma, the different waste feed systems, off-gas treatment and operational experience and future plasma plants. More specifically, it presents the new full-scale plasma facility for the treatment of radioactive waste at the Kozloduy Nuclear Power Plant in Bulgaria for which construction and assembly was finalized end of August 2016. It also describes in more detail the Site Acceptance Tests (SAT) of this facility. This facility is designed and built as an EPC contract by the Joint Venture Iberdrola Ingenieria y Construccion (Spain) and Belgoprocess (Belgium). (authors)
The Charter for the Evaluation and Screening (E and S) of Nuclear Fuel Cycle Options requested by the US Department of Energy calls for the identification of promising fuel cycle options to help guide the future directions of research and development in this area. The final report for this study, which was produced by various national laboratory and industry experts, collected the expansive set of fuel cycle options into 40 evaluation groups (EGs). In support of this study, detailed steady-state physics and mass flow data for each of these groups were compiled into a set of Fuel Cycle Data Packages (FCDPs). The data in the FCDPs were then used to quantify the nine criteria detailed in the study to identify the most promising EGs. In addition to supporting the E and S study, data from these FCDPs can be used to benchmark fuel cycle simulation tools for equilibrium calculations. Once benchmarked, their dynamic modeling capabilities can support the continuing effort to determine the most effective approaches for transitioning to proposed advanced future fuel cycles. While steady-state conditions can be calculated directly in some cases, fuel cycle simulation tools allow for the transition scenarios to be simultaneously optimized on multiple parameters such as the fissile material availability, the amount of desirable material in storage, and the time required for transition. In addition to fuel cycle simulations where material availability is considered and is often a constraint to achieving transition, determining the technology needs to enable a transition to take place is also a key consideration (e.g., what technology is needed, when it is needed, and whether it is needed only during the transition and not for the equilibrium fuel cycle). This paper evaluates the material and technology issues and challenges associated with a transition from a uranium-based fuel cycle (such as that operating in the US today, represented by EG01) to a thorium-based fuel cycle, recognizing that thorium requires fissile material to start up. Based on the findings of the E and S report, the two most promising thorium fuel cycle options are considered in this report: (1) EG26, a thermal spectrum molten salt reactor (MSR), and (2) EG28, represented by a sodium fast reactor (SFR, fast spectrum) and a resource-renewable boiling water reactor (RBWR, intermediate spectrum). (authors)