113 Search Results

Interstellar Boundary Explorer (IBEX) observations of energetic neutral atom (ENA) fluxes from the heliosphere have greatly enriched our understanding of the interaction of the solar wind (SW) with the local interstellar medium (LISM). However, there has been recent controversy surrounding the inability of most ENA models to produce as high an intensity of ~0.5–6 keV ENAs as IBEX observes at 1 au, especially as a function of time. In our previous study (E. J. Zirnstein et al.), we introduced a new model that utilizes a data-driven magnetohydrodynamic simulation of the SW–LISM interaction to propagate pickup ions through the heliosheath (HS) after they are nonadiabatically heated at the heliospheric termination shock. E. J. Zirnstein et al. only simulated and analyzed IBEX observations from the direction of Voyager 2. In this study, we expand our model to include fluxes from the direction of Voyager 1, as well as in the low-latitude part (middle) of the ribbon (10° below the ecliptic plane). We show that the model results at Voyager 1 are consistent with E. J. Zirnstein et al.'s results at Voyager 2 in terms of a secondary ENA source contribution of ≲20% from both directions. Our results in the middle of the ribbon also reproduce the data, when including a time-dependent secondary ENA source. Finally, we demonstrate with our simulation that three large pressure waves likely merged in the VLISM and were observed by Voyager 1 as "pf2," while at least one of the wave's effects in the HS was observed by IBEX as a brief enhancement in ENA flux in early 2016.

The value of the recovered uranium (RU) from high assay low-enriched uranium (HALEU) used nuclear fuels was evaluated. Three utilizations of the recovered uranium were considered in this study, which include the cases that RU is used as a fissile material of nuclear fuel, RU is reused in the original advanced reactor after reenrichment, and RU is reused in conventional light water reactors after down-blending. In this study, the RU values were identified by comparing the cost of making a unit mass of fuel with RU versus the fuel cost with the equivalent fresh enriched uranium (EU). A series of bounding analyses for calculating the fuel costs were conducted using several selected reactor types, which include microreactors, advanced thermal reactors, and fast reactors having a burnup of 2 – 165 GWd/t (with residual U-235 content in discharged fuels of 0.8 - 19.6%). This study concludes that RU having a residual U-235 content higher than ~7% would cost less than the fresh EU. The affordability increases as the residual U-235 content in RU increases. For instance, the fuel cost with RU having the residual U-235 content of 19.6% is about 85% cheaper than the fuel cost with the equivalent fresh EU. This study observed that reusing RU after reenrichment in the original microreactor is impractical because the U-235 content in the re-enriched RU fuel would need to be higher than the limit for low-enriched uranium (<20%) to provide the same burnup performance due to parasitic absorption from U-236. It is noted that this study focused on the recovery of uranium only, and the value of other fissile materials (such as Pu) in the used nuclear fuel was not considered even though those are bred significantly in fast reactors. In addition, the impacts of uncertainties in the cost data and the value of RU of TRISO fuels were not evaluated in this study due to the limited information on the cost data uncertainties and the separation cost from TRISO fuels.

The Systems Analysis and Integration campaign assessed the pros and cons of high-assay low-enriched uranium (HALEU) utilization in advanced reactors and associated fuel cycles. The assessment was done for three example fuel cycles at equilibrium states: once-through, limited recycle, and continuous recycle (CR) starting with HALEU. Front- and back-end fuel cycle parameters and the Levelized Cost of Fuel (LCF), which is the Levelized Cost of Electricity excluding reactor cost, of the three example fuel cycles were calculated using a single Analysis Example Reactor. The pros and cons of HALEU utilization were assessed by normalizing the fuel cycle parameters and LCF to a unit of electricity generation (GWe-year) and comparing them with a Basis of Comparison. In this study, a sodium-cooled reactor with sodium-bonded metallic fuel having a burnup of ~100 GWd/t was used as the Analysis Example Reactor because its technology readiness level is high, and the burnup and fuel enrichment are in the middle of those ranges of advanced reactor concepts that are under development. The current once-through Light Water Reactors (OT-LWRs) with <5% low-enriched uranium and 50 GWd/t burnup were used as the Basis of Comparison. In addition, a series of sensitivity analyses was conducted by varying burnup, enrichment, fuel forms, and reactor types to capture the design variations in two once-through Advanced Reactor Demonstration Program (ARDP) reactors, Natrium with sodium-free metallic fuel having a burnup of ~150 GW/t and Xe-100 with Tristructural-Isotropic (TRISO) pebble fuel having a burnup of ~168 GWd/t.

The purpose of this study is to evaluate the nuclear waste attributes of Small Modular Reactors (SMRs) scheduled for deployment within this decade using available data and established nuclear waste metrics, with the results compared to a reference large Pressurized Water Reactor (PWR). The current fleet of commercial nuclear reactors in the U.S. is composed of 92 large Light Water Reactors (LWR) with an average electricity generating capacity of over 1,000 MWe each. These large LWRs built on-site in massive construction projects have been the mainstay of the industry for the last 50 years. However, new construction soon is expected to include several designs of smaller reactors primarily fabricated in factories and installed in the field in modules. Some of these SMRs will also be LWRs, while some will use other coolants such as liquid metals, molten salts or gases, and different types of fuels. The technologies and economics of SMRs have been the focus of many studies, but there has been only minimal information published on the amount of nuclear waste different types of SMRs are expected to generate and no reports focused on near-term-deployable designs. In this study, the nuclear waste attributes of three small reactors scheduled for near-term-deployment, VOYGR^TM (from NuScale Power), Natrium^{TM ab} (from TerraPower), and Xe-100 (from X-energy), were assessed by comparing nuclear waste metrics with those of a reference large Pressurized Water Reactor (PWR).

A series of fuel cycle scenarios studies were performed to inform on fuel cycle capacities and facilities needed for large-scale deployment of the Advanced Reactor Demonstration Program (ARDP) reactors and potential future evolutionary fuel cycle scenarios. The reactor deployment and evolutionary fuel cycle scenarios from the present to 2100 were developed based on the following assumptions: 1) achievement of a net-zero emissions economy in the United States by 2050, which requires a nuclear energy generation capacity of ~250 GWe by 2050, 2) the U.S. economic growth of 1% per year from 2051 to 2100, which results in ~340 GWe of nuclear energy capacity in 2100, and 3) commercial-scale recycling and high burnup fuel technologies are available after 2050. Thus, evolutionary fuel cycles with those advanced nuclear technologies start after 2050. A single once-through fuel cycle scenario was assumed from the present to 2050 to achieve a net-zero emissions economy in the United States, and the following four evolutionary fuel cycle scenarios from 2051 to 2100 were considered, 1) Once through fuel cycle with ARDP reactors (Natrium and Xe-100), 2) Once-through fuel cycle with Breed-and-Burn (B&B) fast reactors, 3) Recycling fuel cycle of used metallic fuel in fast reactors, and 4) Recycle fuel cycle of both used uranium oxide and metallic fuels in fast reactors. The projected front-end and back-end fuel cycle capacity demands are compared with the current domestic and global (if needed) fuel cycle capacities.

EuTe₄ is a van der Waals material exhibiting a charge density wave (CDW) with a large thermal hysteresis in the resistivity and CDW gap. In this paper, we systematically study the electronic structure and transport properties of EuTe₄ using high-resolution angle-resolved photoemission spectroscopy (ARPES), magnetoresistance (MR) measurements, and scanning tunneling microscopy (STM). Here, we observe a CDW gap of ~ 200meV at low temperatures that persists up to 400 K, suggesting that the CDW transition occurs at a much higher temperature. The ARPES intensity near the Fermi level shows large thermal hysteretic behavior, consistent with the resistivity measurement. The hysteresis in the resistivity measurement does not change under a magnetic field up to 7 T, excluding the thermal magnetic hysteretic effect. Instead, the surface topography measured with STM shows surface domains with different CDW trimerization directions, which may be important for the thermal hysteretic behavior. Interestingly, we reveal a large negative MR at low temperatures that can be associated with the canting of magnetically ordered Eu spins. Our results shed light on the understanding of magnetic, transport, and electronic properties of EuTe₄.

A coal-to-nuclear (C2N) transition means siting a nuclear reactor at the site of a recently retired coal power plant. Three overarching questions from the C2N transition guide this research: where in the United States are retired coal facilities located and what factors make a site feasible for transition; what factors of technology, cost, and project timeline drive investor economics over such a decision; and how will C2N impact local communities? The study team evaluated the siting characteristics of recently retired plants and those operating coal-fired power plant sites run by a utility or an independent power producer utilizing publicly available data to screen U.S. coal power plant sites to nuclear-feasible locations. After screening all retired coal sites to a set of 157 potential candidates and screening operating sites to a set of 237 candidates, the study team estimates that 80% of retired and operating coal power plant sites that were evaluated have the basic characteristics needed to be considered amenable to host an advanced nuclear reactor. For the recently retired plant sites evaluated, this represents a capacity potential of 64.8 GWe to be backfit at 125 sites. For the operating plant sites evaluated, this represents a capacity potential of 198.5 GWe to be backfit at 190 sites. This report evaluates a case study for the detailed impacts and potential outcomes from a C2N transition. Based on the nuclear technology choices and sizes evaluated to replace a large coal plant of 1,200 MWe generation capacity at the case study site, nuclear overnight costs of capital could decrease by 15% to 35% when compared to a greenfield construction project, through the reuse of infrastructure from the coal facility. Nuclear replacement designs can have a lower capacity size because nuclear power plants run at higher capacity factors than coal power plants. In the case study replacing coal capacity with 924 GWe of nuclear capacity, the study team found regional economic activity could increase by as much as $$\$$275 million$ and add 650 new, permanent jobs to the region of analysis. The evaluated site choice in the report is hypothetical for analysis purposes only and based on available data and documented assumptions. Consequently, the findings only inform at a general level. A community, investor, or other interested stakeholder can use these results to set up a detailed, in-depth analysis for a specific application of interest, such as evaluating a C2N transition of a specific coal power plant and a specific nuclear technology design. The report was subjected to independent peer reviews by experts in systems engineering and regional economic modeling to evaluate analysis and assumptions.

This paper compiles and analyzes refined results of the coordinated research project (CRP) on Neutronics Benchmark of China Experimental Fast Reactor (CEFR) Start-Up Tests conducted by the International Atomic Energy Agency (IAEA) since 2018. Twenty-eight research organizations participated with various code systems. The China Institute of Atomic Energy (CIAE) provided the benchmark specifications and participants conducted the benchmark through blind and refined phases. This paper presents the benchmark results on six experimental measurements, criticality, control rod worth, temperature reactivity coefficients, sodium void worth, assembly swap reactivities, and foil activations. Except for a few outlier results, the simulation results show good agreement with the measured data within 1-σ experiment uncertainties. (authors)

The role of nematic order for the mechanism of high-temperature superconductivity is highly debated. In most iron-based superconductors (IBSs) the tetragonal symmetry is broken already in the normal state, resulting in orthorhombic lattice distortions, static stripe magnetic order, or both. Superconductivity then emerges, at least at weak doping, already from the state with broken C₄ rotational symmetry. One of the few stoichiometric IBSs, lithium iron arsenide superconducts below 18 K and does not display either structural or magnetic transition in the normal state. Here we demonstrate, using angle-resolved photoemission spectroscopy, that even the superconducting state in LiFeAs is also a nematic one. We observe spontaneous breaking of the rotational symmetry in the gap amplitude on all Fermi surfaces, as well as unidirectional distortion of the Fermi pockets. Remarkably, these deformations are hardly visible above superconducting T_c. Our results demonstrate the realization of the phenomenon of superconductivity-induced nematicity in IBSs, emphasizing the intimate relation between them. Furthermore, we suggest a theoretical explanation based on the emergence of a secondary instability inside the superconducting state, which leads to the nematic order and s–d mixing in the gap function.

This paper discusses the reactor core design and modeling approaches for a specific molten salt fast reactor (MSFR) design that is started up with high-assay low-enriched uranium (HALEU) within a chloride fuel salt and focuses on the technical assumptions and design choices. These include the fuel salt properties, first core and make-up heavy metal requirements, fuel composition evolution, equilibrium fission product concentration, etc. This MSFR concept was developed as part of a wider study that sought to compare the fuel cycle performance tradeoffs between deploying a fleet of sodium-cooled fast reactors (SFRs) and deploying a fleet of MSFRs. MSFRs that use flowing liquid fuel salt and on-line salt processing are of particular interest because these technology characteristics are significantly different from those of SFRs and can lead to major differences in fuel cycle impacts. The goal of this study was not to draw general conclusions based on a direct comparison of two very specific point designs for an SFR and MSFR, but rather to use this example MSFR design and associated parametric studies to gain insight into the key technology characteristics that impact fuel cycle performance as it relates to resource utilization (i.e., uranium consumption). The MSFR reactor physics, fission product removal, and fuel salt mass flows were modeled using Argonne’s recently-developed molten salt reactor (MSR) analysis tools while the time-dependent fuel cycle performance of this specific reactor was compared to that of a representative SFR design using the fuel cycle system dynamics code DYMOND.