In December 2023, the United States pledged at a United Nations conference to triple nuclear energy by 2050. In May of 2025 a new administration announced plans to quadruple nuclear energy by 2025, setting a new target that requires understanding the key infrastructure and resources needed to achieve large-scale deployment. This study attempts to quantify “what it takes” to deploy an additional 200 GWe of nuclear capacity. It projects demand levels for workforce, fuel, uranium mining and enrichment, waste management, supply chain, land, licensing, and water. Demand growth is also contextualized. For example, the peak number of construction workers required is approximately 215,000, a 37 % increase in the current US utility construction workforce. Up to 80,000 MT/year of mined uranium will be needed, necessitating a 60-fold increase in US mining production or a doubling of global mining if sourced externally. Advanced reactor fuels, uranium mining, and nuclear-grade equipment production would need substantial expansion. Operational workforce, enrichment needs, and licensing submissions also face significant increases. Although water consumption may see relatively lower increases, securing water rights poses unique challenges. This study consolidates various demand metrics in a broad context, highlighting the necessity of early preparations, such as workforce training, to support a tripling of nuclear energy. While it outlines the demand-side requirements, it does not assess the difficulties of ramping up or other supply-side issues. The findings also shed light on the impact of reactor size and reactor technology on various demand metrics.

The molten salt reactor is one candidate among the Generation IV nuclear reactor designs, with its deployment relying on advanced computational tools to capture the unique behavior of the circulating fuel system. The Molten Salt Reactor Experiment (MSRE) provides valuable experimental data for validating these computational tools. This work develops a reactor transient benchmark based on the MSRE pump transient tests.Two computational models are evaluated in the benchmark: a simplified one-dimensional (1D) system-level model and a more detailed R-Z axisymmetric model using the porous medium approximation. The models are used to evaluate the impact of spatial resolution on predicted reactivity responses during the transient. Several impactful factors are examined during the benchmark evaluation, including the neutron diffusion multigroup energy structure, delayed neutron precursor (DNP) diffusion, DNP group structure, bypass flow, and transient flow rates.The reactivity predictions using the computational models are compared to the experimental data. The mean errors in the predicted reactivity responses ranged from 11 to 21 pcm (1 pcm = 10-5) for the pump startup transient and 5 to 13 pcm for the pump coastdown transient. These results indicate that the 1D model can provide adequate accuracy on MSRE pump transients with limitations in predicting the rate of reactivity at the early stage of the transient, while the higher-order model improves this capability by incorporating the influence of radial salt flow distribution and bypass flow on transient reactivity.4

The Molten Salt Reactor (MSR) Multiphysics Applications technical area in the Nuclear Energy Advanced Modeling and Simulation program has supported the development of 3D Monte Carlo models of the Molten Salt Breeder Reactor (MSBR) over the last couple of years. This MSBR model was previously run with the Shift Monte Carlo code to perform radiation shielding calculations in the reactor cell area. The MSBR is a 2250 MWth (1000 MWe) liquid-fueled molten salt reactor design developed at Oak Ridge National Laboratory in the 1970s. Determining the source terms from activated primary heat exchanger (HX) components is important because delayed and prompt neutron fluxes incident on these components affect the dose rate in the primary HX maintenance areas. This information can be used in the development of remote handling procedures required during shutdown for maintenance. A methodology has been developed and is proposed in this paper to quantify the activated source term from the primary HX components as a result of the movement of the delayed neutron precursors in flowing primary fuel salt through the primary HXs in the MSBR. The goal of this research is to evaluate the gamma dose rates in the maintenance hatches above the primary HX using the activated HX source terms. The study showed that the gamma dose rates are approximately two orders of magnitudes higher when accounting for the neutron activation from the movement of delayed neutron precursors through the HXs than when flowing fuel is not considered. Thus, the movement of delayed neutron precursors must be taken into account for accurately predicting the neutron activation of primary loop components.

To assess the modeling and simulation capabilities for radionuclide characterization, criticality, and shielding in the nuclear fuel cycle of non-light–water reactor (LWR) technologies, the US Nuclear Regulatory Commission (NRC) initiated a collaborative project between the NRC, Sandia National Laboratories (SNL), and Oak Ridge National Laboratory (ORNL) with the goal to demonstrate capabilities of MELCOR and SCALE to calculate accident scenarios in all stages of the nuclear fuel cycle for relevant non-LWRs. The first project task was to develop representative nuclear fuel cycles and identify potential hazards and accident scenarios in the individual fuel cycle stages based on publicly available information. Because the nuclear fuel cycle is not established for any non-LWR concept, many assumptions were made, and it is anticipated that the details of the fuel cycles will eventually look different.

The concrete biological shield (CBS) of light water reactors is exposed to high neutron radiation dose in the long term, which may lead to the degradation of the concrete's mechanical properties. Given the important shielding role of the CBS, it is necessary to investigate the irradiation effects at the structural scale and provide estimates of the damage extent from the wall's inner surface to study potential license renewals. For this purpose, we developed a mechanical model accounting for radiation-induced expansion, creep, and damage in concrete using the Grizzly finite element code, informed by ex-core neutron flux calculations using the VERA tool. The model was applied to a 3D CBS structure represented by the CBS wall, a steel liner, reinforcement bars, and a concrete base mat and evaluated damage at 40, 60, and 80 years of operation. The VERA model predicted a maximum fluence of approximately 2 x 1019 n cm-2 at 80 years of operation. The results showed that damage is highest at the inner surface of the CBS wall and gradually decreases with depth. It extends beyond the rebar after 60 years and reaches a depth of approximately 12 cm at 80 years.

The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) is a reactor simulation software. It offers unique capabilities by combining high-fidelity in-core radiation transport with temperature feedback by using MPACT (a deterministic neutron transport code) and COBRA-TF (a thermal-hydraulic code) with follow-on, fixed-source transport calculations using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides Shift with the fission source for follow-on ex-core calculations. These ex-core simulations can be set up to calculate detector responses, as well as the flux and fluence in ex-core regions of interest, such as the reactor pressure vessel, nozzle, and irradiated capsules. A Watts Bar Nuclear Plant Unit 1 (WBN1) ex-core model was developed, as described in this paper, and this model was used to perform coupon calculations. The results for the coupon flux calculations show close agreement with the reference values for cycle 1 produced by the two-dimensional Discrete Ordinates Transport (DORT) code and presented in a BWXT Services Inc. report. However, differences in the results (10%) seen in cycles 2 and 3 and the reasons for these differences are discussed in this paper. The VERA WBN1 model was also used to perform a vessel fluence calculation for cycle 1. Additionally, a collaboration between CASL and Duke Energy led to the first code-to-code validation of VERA for reactor ex-core applications that used a model for the Shearon Harris reactor. Results from this collaboration show excellent agreement between VERA and the Monte Carlo N-Particle Transport Code for the detector response calculations. The work performed under this collaboration is also detailed in this paper.

current full-size aluminide core configurations with 43 {approximately} 45 fuel elements. Adequacy of the overall equilibrium cycle approach has been verified through comparison with recent FNR experience in spent fuel discharge rates and simulation of reactor physics characteristics for two representative cycles. Fuel cycle studies have been performed to compare equilibrium cycle characteristics of silicide fuel designs, including burnable absorbers, with current aluminide fuel. These equilibrium cycle studies have established the feasibility of doubling the fuel element lifetime, with minimal perturbations to the control reactivity requirements and peak power density, by judicious additions of burnable absorbers to silicide fuel. Further study will be required to investigate a more practical silicide fuel design, which incorporates burnable absorbers in side plates of each fuel element rather than uniformly mixes them in fuel plates.

will serve as references for fuel fabrication and qualification activities. Recent advancements in modeling and simulation tools enable quick prototyping of fuel designs. Shift, a Monte Carlo neutron transport and depletion tool optimized for high-performance computing (HPC) architectures, is used for efficient fuel cycle and performance metrics calculations. The HFIR Steady State Heat Transfer Code (HSSHTC) is used to vet the thermal safety margin. Also, a new automation tool that connects all fuel design analysis steps, named Python HFIR Analysis and Measurement Engine (PHAME), has been developed to expedite the design study in an efficient and reproducible manner. Leveraging these tools, several candidate fuel designs were selected for varying fabrication complexity. This report provides design feature details for four selected HFIR LEU low density U3Si2-Al fuel designs and their corresponding performance and safety metrics. Nominal, best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities are provided for candidate fuel designs relevant to framing irradiation experiments to support fuel qualification efforts. Simulations show that the low density U3Si2-Al, with design features to enhance safety, can meet HEU core performance metrics and safety requirements if the reactor power is increased from 85 MW (HEU) to 95 MW (LEU) and if the active fuel length is increased from 50.80 cm (HEU) to 55.88 cm (LEU).

processor domains. In these ex-core calculations, MPACT is coupled to CTF and performs the in-core depletion and heat transfer calculation, followed by a fixed-source Shift transport calculation including ex-core regions to produce ex-core responses. The fission source, fuel pin temperatures, moderator temperature and density, boron concentration, and fuel pin depleted isotopic compositions can be transferred to Shift from the MPACT calculation. Specifically, VERA performs fluence calculations in the core barrel outward to the end of the pressure vessel and detector response calculations in ex-core detectors. It also performs the requested tallies in any user-defined ex-core regions. VERA takes advantage of the General Geometry (GG) package in Shift. This gives VERA the flexibility to explicitly model features outside the core barrel, including detailed vessel models, detectors, and power plant details. A very limited set of experimental and numerical benchmarks is available for ex-core simulation comparison. The Consortium for Advanced Simulation of Light Water Reactors has developed a set of ex-core benchmark problems to include as part of the VERA verification and validation set of problems. The ex-core capability in VERA has been tested on small representative assembly problems, multi-assembly problems, as well as quarter-core and full-core problems. VERA View has also been extended to visualize these vessel fluence results from VERA. This manual serves to present a guide to VERA users about the methodology behind ex-core calculations and the details of input, output, and analysis of results from these calculations. Details in this version of the manual are based on features in VERA 4.0.1.

providing an attractive selection of vehicle makes and models for car-buyers. The upfront costs will be high and could persist for a decade or more, delaying profitability until an adequate number of vehicles can be produced and moved into consumer markets. However, the potential rewards to the economy, environment, and national security are immense. Such a profound market transformation will require careful planning and strong, consistent policy incentives. Section 811 of the Energy Policy Act (EPACT) of 2005, Public Law 109-59 (U.S. House, 2005), calls for a report from the Secretary of Energy on measures to support the transition to a hydrogen economy. The report was to specifically address production and deployment of hydrogen-fueled vehicles and the hydrogen production and delivery infrastructure needed to support those vehicles. In addition, the 2004 report of the National Academy of Sciences (NAS, 2004), The Hydrogen Economy, contained two recommendations for analyses to be conducted by the U.S. Department of Energy (DOE) to strengthen hydrogen energy transition and infrastructure planning for the hydrogen economy. In response to the EPACT requirement and NAS recommendations, DOE's Hydrogen, Fuel Cells and Infrastructure Technologies Program (HFCIT) has supported a series of analyses to evaluate alternative scenarios for deployment of millions of hydrogen fueled vehicles and supporting infrastructure. To ensure that these alternative market penetration scenarios took into consideration the thinking of the automobile manufacturers, energy companies, industrial hydrogen suppliers, and others from the private sector, DOE held several stakeholder meetings to explain the analyses, describe the models, and solicit comments about the methods, assumptions, and preliminary results (U.S. DOE, 2006a). The first stakeholder meeting was held on January 26, 2006, to solicit guidance during the initial phases of the analysis; this was followed by a second meeting on August 9-10, 2006, to review the preliminary results. A third and final meeting was held on January 31, 2007, to discuss the final analysis results. More than 60 hydrogen energy experts from industry, government, national laboratories, and universities attended these meetings and provided their comments to help guide DOE's analysis. The final scenarios attempt to reflect the collective judgment of the participants in these meetings. However, they should not be interpreted as having been explicitly endorsed by DOE or any of the stakeholders participating. The DOE analysis examined three vehicle penetration scenarios: Scenario 1--Production of thousands of vehicles per year by 2015 and hundreds of thousands per year by 2019. This option is expected to lead to a market penetration of 2.0 million fuel cell vehicles (FCV) by 2025. Scenario 2--Production of thousands of FCVs by 2013 and hundreds of thousands by 2018. This option is expected to lead to a market penetration of 5.0 million FCVs by 2025. Scenario 3--Production of thousands of FCVs by 2013, hundreds of thousands by 2018, and millions by 2021 such that market penetration is 10 million by 2025. Scenario 3 was formulated to comply with the NAS recommendation: 'DOE should map out and evaluate a transition plan consistent with developing the infrastructure and hydrogen resources necessary to support the committee's hydrogen vehicle penetration scenario, or another similar demand scenario (NAS, 2004, p. 4).' Each of the scenarios was extensively discussed at the stakeholder meetings and each received support from industry. Although there was no consensus on a particular vehicle penetration rate, it was agreed that this set of scenarios is inclusive of industry expectations and could provide a basis to interpolate or extrapolate the results to other cases. The purpose of the DOE study was not to select any one scenario but to assess the costs and impacts of achieving each.

but no fuel temperature information is passed back to MPACT/CTF, and (2) a two-way approach (coupled) in which the fuel temperature is passed from Bison to MPACT/CTF. In both approaches, power and temperature distributions from MPACT/CTF are used to inform the Bison simulations for each rod in the core.The demonstrations presented here are the first integrated fuel performance simulations in VERA, which opens many possibilities for future work, including applications to accident-tolerant fuel efforts and transient simulations, which are of critical importance to CASL. These demonstrations also highlight the potential to move away from the current Bison-informed fuel temperature lookup table approach, which is the default in MPACT/CTF simulations, if performance improvements are made in the near future.