Trace-level impurity quantification in lead-cooled fast reactors using ICP-MS: Methodology and challenges

Solar power has innate issues with weather, grid demand and time of day, which can be mitigated through use of thermal energy storage for concentrating solar power (CSP). Nuclear reactors, including lead-cooled fast reactors (LFRs), can adjust power output according to demand; but with high fixed costs and low operating costs, there may not be sufficient economic incentive to make this worthwhile. We investigate potential synergies through coupling CSP and LFR together in a single supercritical CO2 Brayton cycle and/or using the same thermal energy storage. Combining these cycles allows for the LFR to thermally charge the salt storage in the CSP cycle during low-demand periods to be dispatched when grid demand increases. The LFR/CSP coupling into one cycle is modeled to find the preferred location of the LFR heat exchanger, CSP heat exchanger, sCO2-to-salt heat exchanger (C2S), turbines, and recuperators within the supercritical CO2 Brayton cycle. Three cycle configurations have been studied: two-cycle configuration, which uses CSP and LFR heat for dedicated turbocompressors, has the highest efficiencies but with less component synergies; a combined cycle with CSP and LFR heat sources in parallel is the simplest with the lowest efficiencies; and a combined cycle with separate high-temperature recuperators for both the CSP and LFR is a compromise between efficiency and component synergies. Additionally, four thermal energy storage charging techniques are studied: the turbine positioned before C2S, requiring a high LFR outlet temperature for viability; the turbine after the C2S, reducing turbine inlet temperature and therefore power; the turbine parallel to the C2S producing moderate efficiency; and a dedicated circulator loop. While all configurations have pros and cons, use of a single cycle offers component synergies with limited efficiency penalty. Using a turbine in parallel with the C2S heat exchanger is feasible but results in a low charging efficiency, while a dedicated circulator loop offers flexibility and near-perfect heat storage efficiency but increasing cost with additional cycle components.

• Developed dynamic models to simulate LFR-based nuclear propulsion systems. • Introduced the Ship-to-Shore concept, enabling clean energy transfer to ports. • Achieved significant CO 2 emission reduction: 24.5 kt/y during navigation. • Increased energy production by 697% through the Ship-to-Shore concept. • Provided a pathway for sustainable maritime propulsion aligned with IMO targets. This paper investigates the integration of IV generation small modular reactors into maritime vessels as a viable alternative to conventional fossil fuel-powered systems. To support maritime decarbonisation and evaluate the energy and economic performance of their onboard implementation, suitable mathematical models of energy systems powered by small modular reactors are developed. Among various nuclear technologies, a lead-cooled fast reactor (LFR) using mixed oxide fuel is modelled and implemented within a dynamic simulation environment. The model simulates the energy performance of a Helium Brayton cycle with a regenerator and reheating, powered by an LFR. This model can assess the temperature distribution of the lead coolant, the helium temperatures throughout the Brayton cycle, the generated mechanical and electrical power, thermodynamic efficiency, and waste heat. Following the development of the LFR-based model, it is integrated into a broader computational framework that enables the evaluation of the ship’s overall energy performance, including the energy demands of onboard utilities under different operating modes and climatic conditions. To demonstrate the potential of these systems, a LFR is simulated as a replacement for conventional fossil fuel engines onboard a containership. A representative route, reconstructed from real-time online data, is used to assess reactor performance under dynamic maritime conditions. Furthermore, a novel concept – ship-to-shore connection – is introduced, whereby surplus electricity produced by the onboard LFR is supplied to port grids, contributing to the decarbonisation of port facilities and surrounding areas. Simulation results indicate that small modular reactor systems could effectively nullify onboard fossil fuel consumption. For instance, avoided CO 2 emissions are estimated at 24.5 kt/y during navigation and 15.3 kt/y during port stays, including also emissions offset through the replacement of national thermoelectric power generation in ship-to-shore application. The implementation of the ship-to-shore connection concept results in an increase in energy production during port stays ranging from 572% to 697%, yielding substantial environmental and economic benefits.

The Westinghouse lead-cooled fast reactor (LFR) is a next-generation nuclear power plant whose primary mission is to reduce front-end capital costs and generate flexible and cost-competitive electricity for global markets, while offering mission versatility and satisfying the highest safety and sustainability standards. The LFR is an economical choice for carbon-free power generation to meet the ambitious goals of reducing carbon emissions. The plant utilizes passive safety systems for high reliability and public safety. The LFR testing program aims to fill the technology gaps in the key materials, components, and systems of the LFR plant. It provides demonstrations of LFR engineering that reduce the uncertainty of the LFR design and experimental data for the verification and validation of the modeling and simulation computer codes. With the support of the LFR phenomena identification and ranking table, the phenomena important to plant safety with insufficient states of knowledge have been identified and a Westinghouse testing plan developed to address them. The United Kingdom Advanced Modular Reactor Program Phase 2 was purposed to advance the level of design maturity toward eventual regulatory approval and deployment. Westinghouse and its partners have developed eight state-of-the-art test facilities within the program. Among them, the Passive Heat Removal Facility, Versatile Lead Facility, Lead Water Interaction Facility, and the Lead Freezing Facility are thermal-hydraulic testing facilities, with the other four facilities dedicated to testing lead corrosion and material mechanical behavior in lead. The design, development, and testing of the four thermal-hydraulic facilities are presented in this paper. A variety of computer codes ranging from system-level analysis, component-level analysis, and high-resolution computational fluid dynamics analysis, are utilized to support the development of the facilities. The analyses provide operational and accidental conditions in the LFR to identify major testing conditions and inform the facility design and testing matrix. With a significant amount of experimental data being generated with these highly instrumented test facilities, the computer codes and models will be benchmarked against the experimental results to improve their validation and verification status.

Overview of lead-cooled fast reactor activities

This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The roadmaps are generic, i.e., not specific to a particular vendor’s design, but vendor design information may have been used as a representative of the concept family. The SFR technology is mature enough for commercial demonstration by the early 2030s, and the remaining key activities and R&D needs over the next five years are generally related to the completion of qualification of fuel and structural materials, validation of reactor design codes and methods, and support of the licensing frameworks. The LFR’s technology is instead less-mature and will be at the engineering demonstration stage by the early 2030s. Key LFR technology development activities for the acceleration of the engineering demonstration of the LFR will focus on resolving remaining design challenges and demonstrating the viability of systems and components in the integral system, which will be done in parallel with addressing the gaps shared with SFR technology. As discussed in the Advanced Demonstration and Test Reactor Options Study, the R&D approaches and timelines presented here assume that, for the first module demonstration, vendors would pursue a two-step licensing process based on 10CFR Part 50.

In lead-cooled fast reactors, preventing corrosion of materials like stainless steel by liquid metal is crucial. The protective oxide film on stainless steel surfaces, formed to mitigate corrosion in lead/lead-bismuth systems, can, paradoxically, degrade heat transfer in cladding during high-oxygen concentration or normal operation. This study investigates the impact of corrosion on heat transfer, focusing on steady-state conditions. By analyzing temperature changes with and without corrosion, the study aims to understand the balance between oxide film formation and heat conduction. Additionally, it assesses corrosion effects under accident conditions to enhance safety in lead-cooled fast reactors. The findings contribute valuable insights for designing and operating cooling systems, ensuring the safety and reliability of nuclear energy systems.

The management of nuclear wastes has long been a problem that hinders the sustainable and clean utilization of nuclear energy since the advent of nuclear power. These nuclear wastes include minor actinides (MAs: 237Np, 241Am, 243Am, 244Cm and 245Cm) and long-lived fission products (LLFPs: 79Se, 93Zr, 99Tc, 107Pd, 129I and 135Cs), and yet are hard to be handled. In this work, we propose a scheme that can transmute almost all the MAs and LLFPs with a lead-cooled fast reactor (LFR). In this scheme, the MAs and the LLFPs are loaded to the fuel assembly and the blanket assembly for transmutation, respectively. In order to study the effect of MAs loading on the operation of the core, the neutron flux distribution, spectra, and the keff are further compared with and without MAs loading. Then the LLFPs composition is optimized and the support ratio is obtained to be 1.22 for 237Np, 1.63 for 241Am, 1.27 for 243Am, 1.32 for 79Se, 1.53 for 99Tc, 1.02 for 107Pd, and 1.12 for 129I, respectively, indicating that a self-sustained transmutation can be achieved. Accordingly, the transmutation rate of these nuclides was 13.07%/y for 237Np, 15.18%/y for 241Am, 13.34%/y for 243Am, 0.58%/y for 79Se, 0.92%/y for 99Tc, 1.17%/y for 107Pd, 0.56%/y for 129I. Our results show that a lead-cooled fast reactor can be potentially used to manage nuclear wastes with high levels of long-lived radioactivity.

A review of recent numerical and experimental research progress on CDA safety analysis of LBE-/lead-cooled fast reactors

Understanding and predicting nuclear fuel behavior is the cornerstone for fuel design and optimization, and the safe and economic operation of nuclear reactors. Due to the excellent neutronic economy and safe operational characteristics of lead/lead-bismuth-based liquid metal coolant, lead-cooled fast reactor (LFR) has received much attention and has been regarded as one of the top candidate reactors for Generation IV nuclear power plants. Mixed oxide (MOX) fuel is beneficial to improve fuel cycle utilization and consume weapons-grade plutonium, among which hollow MOX fuel has excellent heat transfer advantages, which can further improve fuel economy and safety, and has good development prospects. In this work, an annular uranium-plutonium mixed oxide (MOX) fuel operating in a liquid lead/lead-bismuth cooled fast reactor is modeled and simulated to predict its behavior under transient and steady-state operation. This model has several modules working in a fully coupled approach based on COMSOL Multiphysics, such as heat transfer, fuel burnup, oxygen redistribution, plutonium redistribution, porosity evolution, fission gas release, JOG growth, and grain size evolution. The modeling results were benchmarked with the existing codes and experimental data, and the results were in satisfactory agreement. The parameters analysis was carried out in this work. Consistent with the solid MOX fuel, the O/M ratio (or deviation from the stoichiometry of oxygen) significantly affects temperature evolution, fission gas migration and release behavior, and plutonium redistribution. Linear power also has a significant influence on fuel performance.

With the increasing global demand for sustainable energy, the importance of advanced nuclear technologies, such as fourth-generation reactors, has become increasingly prominent. Fourth-generation reactors often use non-water coolants, among which liquid lead and lead–bismuth eutectics (LBEs) are highly promising, making lead-cooled fast reactors (LFRs) a popular area of research internationally. On the basis of extensive analysis and comparison conducted previously, this article summarizes and analyzes the advantages, problems, and differences in lead and LBE as LFR coolants. Overall, both lead and LBE have excellent neutronic characteristics, good heat transfer performance, and chemical inertness, and make LFRs highly efficient in nuclear fuel utilization, inherently safe, and relatively simplified in design. However, both of them corrode materials severely and produce highly toxic 210Po, which are the problems that need to be considered for further engineering development. Moreover, LBE has a lower melting point, which allows a wider temperature range and lower insulation requirements in its design, making it easier to achieve engineering and miniaturization under the same conditions. Lead has a lower cost, is less corrosive to materials, and produces less 210Po, which makes it a more ideal coolant for future development.

Comparative analysis on high-efficiency energy conversion system with different working mediums matched with small lead-cooled fast reactor

Study of amorphous alumina coatings for next-generation nuclear reactors: High-temperature in-situ and post-mortem Raman spectroscopy and X-ray diffraction

Development of a control-oriented simulator for a LFR demonstrator

Separate effect experimental test and theoretical investigate of pressure wave propagation and pressure rising process upon SGTR accident in advanced LFRs

China initiative Accelerator Driven System (CiADS) combines a linac, spallation target and a Lead-cooled Fast Reactor (LFR) together, which is designed to transmute nuclear waste and accelerate the progress of China’s energy technology research towards the goal of carbon neutrality. A LFR uses helical wire-wrap spacers as positioning components to enhance crossflow mixing in the reactor core. To study the velocity distribution and crossflow characteristics in wire-wrapped rod bundle channels, a 2 : 1 magnified scale 7-pin bundle fuel assembly model was fabricated using polymathic methacrylate. Particle image velocimetry (PIV) and computational fluid dynamics (CFD) simulations were used to investigate the velocity distribution in the 7-pin bundle flow channels at Reynolds number of 1250~5000 in the x z plane and Reynolds number of 1500 and 2500 in the x y plane. The deviation between CFD simulation results and PIV experimental data was small, and the Reynolds Average Navier-Stokes model could accurately simulate the flow characteristics of the wire-wrapped fuel rod bundle channels. The maximum crossflow velocity caused by helical wires was about 40% of the axial bulk velocity. The normalized crossflow velocity at the subchannel interface varied approximately sinusoidally with the axial height. As the Reynolds number increased, the velocity distribution trend and the loss rate of axial velocity in flow channels remained essentially constant while the peak value of crossflow velocity increased. The contour images of velocities with different axial heights were obtained from the x y plane, and their velocity distribution had a certain periodicity. The axial velocity loss rate in each subchannel caused by wire-wrap spacer resistance was between 7.35% and 38.51%, and the axial velocity loss rates in inner subchannels were usually higher than those in edge subchannels.