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An Overview of Non-LWR Vessel Cooling Systems for Passive Decay Heat Removal (Technical Letter Final Report)
The following report provides a technical review of various reactor vessel cooling system (VCS) concepts under consideration for decay heat removal (DHR) in non-light water advanced reactor designs. This review focuses on ex-vessel designs, including the Reactor Cavity Cooling System (RCCS), the Reactor Vessel Auxiliary Cooling System (RVACS), and hybrid iterations, using both air and water cooling to achieve their heat removal function. Based on a literature review of publicly available sources published between 1979 and 2021, a technical summary is presented detailing existing and planned VCS design options, their applicability to specific reactor type, and review of authored research and development studies. Following an assessment of the availability of data and modeling tools, an evaluation was performed analyzing their likely performance during normal, degraded, and accident conditions, including reliability, stability, and longevity. With renewed consideration of air-based DHR systems by some US vendors, there is a need to fully understand the complexities inherent to natural circulation systems, and more importantly, how they may influence safety-related heat removal functions and system performance. For DHR systems that rely on air-based mode of cooling, this entails quantification of the impacts of low flow conditions during start-up, relative elevations of inlet and outlet ducts for below-grade installations, effects of the use of multiple parallel chimneys for redundancy, and impact of weather conditions at the plant site. For systems relying on water-based mode cooling, reliance on limited coolant inventory in an event of pipe break or extended duration accident scenarios, sensitivities related to stability of boiling flow and heat transfer conditions, and quantification of full-scale structural vibrations on balance of plant structures are among the complicating factors.
Human Factors for Advanced Reactors
Existing light water reactors in the U.S. are primarily large baseload electricity generating facilities. The concept of operations for these plants remains largely unchanged since the advent of commercial nuclear power—the main control room serves as the hub of plant activities and is staffed with multiple licensed operators who work in tandem under the shift supervisor, and staff such as field workers support the control room remotely. While newer plants have brought the advent of digital human-machine interfaces to replace earlier analog and mechanical instrumentation and controls, much of the control process remains unchanged and manual. It is simply a newer version of legacy concepts. Advanced reactors potentially bring considerable changes to the size, fuel type, automation, and staffing of nuclear power plants, necessitating a fundamental shift not just from analog to digital, but further from human to automation, from onsite to remote, from control to monitoring, and from many to few operators. Despite this multitude of parallel evolutions in reactor designs, many of the vendors developing the next generation of reactors represent smaller research and development enterprises. It is therefore not feasible to address all aspects of plant design at the same time. In particular, the competing design aspects of new reactors present a significant challenge to the development of robust and human factored systems at the plant. As vendors develop new reactor designs, much of the early focus is naturally on the fuel and reactor system technology. Looming behind these early advances is the daunting prospect of first-of-a-kind control concepts that have not yet been developed or validated. A failure to address the human element of reactor design early will lead to missed opportunities. The quickest development process is the replication of existing concepts of operations at legacy plants, even when such systems were long ago surpassed by better human-machine technologies outside the nuclear industry. Conversely, attempting to undertake novel concepts of operations late in the design life cycle of a plant could result in protracted development efforts and delays in licensing and deployment. This does not have to happen, and it is imperative that human factors be considered now, early in the design of new reactors.
Reactor Walkway Design for Gear Box Support Beam Removal
The walkway over the reactor is intended to be used in the core internal changeout (CIC) outage to replace the gear box suport beams. (GBSB) The current configuration of the in-tank platform prevents easily discharging the GBSB to the canal through the drop chute. The walkway's purpose is to increase operators access to the south irration deck, where the GBSB will be put into discharge cans, as well as the drop chute.
Challenges in the UK Generic Design Assessment Process - 17062
In the UK, Generic Design Assessment (GDA) is a voluntary step-wise joint assessment process undertaken by the Office for Nuclear Regulation (ONR) and the Environment Agency (EA) of a generic reactor design. The process focuses purely on the reactor design and does not consider any site specific issues. The key objective of GDA is for the regulators to consider the adequacy of claims, arguments and evidence (CAE) put forward by a requesting party (RP) to demonstrate the safety of its design in the UK context. This includes all CAE relating to management of spent fuel and radioactive waste. The CAE are generally applied to a reference design put forward by the RP that is substantially complete in terms of the overall concepts and functions of the major systems, having already been deployed elsewhere. GDA assessment encompasses a broad range of topic areas which include fuel and core design, management of spent fuel, radioactive waste management and decommissioning. Other technical disciplines also have valid interests in particular aspects of the management of spent fuel. This paper summarises the UK GDA process and the types of issues raised; in particular aspects of the proposed operation of liquid waste management systems and Spent Fuel Interim Storage are presented and the instructive resolution of the RP's original limited design scope to allow the regulators to carry out a meaningful assessment. Some examples are presented and useful lessons learned for future GDAs are discussed. (authors)
Possibility of nuclear pumped laser experiment using low enriched uranium
Possibility to perform experiments for nuclear pumped laser oscillation by using low enriched uranium is investigated. Kinetic analyses are performed for two types of reactor design, one is using highly enriched uranium and the other is using low enriched uranium. The reactor design is based on the experiment reactor in IPPE. The results show the oscillation of nuclear pumped laser in the case of low enriched uranium reactor is also possible. The use of low enriched uranium in the experiment will make experiment easier.
Depletion Analysis of a Generic Fast Spectrum Molten Salt Reactor Supporting Material Control and Accounting (Rev.1)
Advanced reactors are of interest for a variety of use scenarios, and there are numerous advanced reactor designs being considered, which includes molten salt reactors (MSRs). Depending on the design, advanced reactors may have more extensive material control and accounting (MC&A) processes than current Light Water Reactor (LWR) designs. This project evaluated the use of Monte Carlo simulations for generating information that could guide the development of MC&A approaches for MSRs. In particular, the current capabilities of MCNP version 6.3 internally coupled with CINDER’90 were assessed for a fast spectrum liquid fueled MSR with online fission product removal and refueling. The fission product removal and refueling was performed in batches, and a Python wrapper was developed to control the simulations and update the fuel composition.
Reactor Subsystem Simulation for Nuclear Hybrid Energy Systems
Nuclear Hybrid Energy Systems (NHES) could be a key part of the solution to achieving energy security, may provide reliable power availability even with increasing renewable energy penetration into the power grid, and may allow repurposing excess electricity in times of low demand. Incorporating a fission-based power source in a multi-output system (electricity and process heat) can offer significant advantages over carbon-based production sources, such as coal or natural gas, including reduction in atmospheric carbon emissions. In an integrated multi-output system, thermal energy from the nuclear reactor subsystem can be diverted to industrial applications in times of low electricity demand. High temperature, high quality heat from advanced reactor designs might be used for hydrogen production or synthetic fuel production (coal-to-liquid or natural gas-to-liquid processes). Low temperature heat from advanced reactor designs or the current and future fleet of light water reactors (LWRs) might be applied to desalination processes or be stored for later use in low temperature ammonia or hydrocarbon-based thermal cycles. The NHES research team is currently developing a dynamic simulation of an integrated hybrid energy system. A detailed simulation of proposed NHES architectures will allow initial computational demonstration of a tightly coupled NHES to identify key reactor subsystem requirements, identify candidate reactor technologies for a hybrid system, and identify key challenges to operation of the coupled system. This work will provide a baseline for later coupling of design-specific reactor models through industry collaboration. The modeling capability addressed in this report focuses on the reactor subsystem simulation. Many computational tools and methods are available for modeling a nuclear reactor, power conversion systems and associated process applications. One method is to develop a single program capable of modeling the entire system. Another is to combine two or more existing computer programs, each modeling a portion of the system for which it is validated, where the programs exchange information to compute the solution. The former method requires a lengthy development and verification and validation process. The latter method leverages industry standard software to reduce development and funding requirements. Preliminary system models have been developed by Idaho National Laboratory researchers and are currently being enhanced to assess integrated system performance given multiple sources (e.g., nuclear + wind) and multiple applications (i.e., electricity + process heat). Initial efforts to integrate a Fortran-based simulation of a small modular reactor (SMR) with the balance of plant model have been completed in FY12. This initial effort takes advantage of an existing SMR model developed at North Carolina State University to provide initial integrated system simulation for a relatively low cost. The SMR subsystem simulation details are discussed in this report.
Griffin: A MOOSE-based reactor physics application for multiphysics simulation of advanced nuclear reactors
Griffin is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based reactor physics application for multiphysics simulations of advanced reactor designs jointly developed by Idaho National Laboratory and Argonne National Laboratory. This paper summarizes the motivation, significance, architecture, design, and features of Griffin. Griffin offers flexible and extensible features to address the challenges associated with advanced reactor designs. These features range from fundamental particle transport to specific reactor physics tasks. The features cover a wide range including on-the-fly and traditional two-step cross-section generation methods, steady-state and transient transport solvers suitable for both heterogeneous and homogeneous models, high-fidelity depletion where thousands of isotopes can be tracked and low-fidelity depletion characterized by burnup, etc. The most fundamental aspect that sets Griffin apart from other reactor analysis codes is that it is developed based on the MOOSE framework. A modular development approach is strongly enforced, with multiphysics being an essential element considered since the beginning of Griffin’s development. Griffin links various MOOSE physics modules and couples to other MOOSE-based applications and non-MOOSE-based applications for multiphyiscs simulations. Griffin includes three modules: ISOXML for preparing and managing multigroup cross sections, radiation transport for solving the neutron transport equation, and reactor analysis for user-oriented reactor physics analysis functionalities. Griffin uses various finite element methods for spatial discretization, multigroup approximation for energy discretization and discrete ordinates method, spherical harmonics expansion method, and diffusion approximation for streaming direction discretization to solve the neutron transport equation. Griffin’s flexibility is evidenced through Griffin’s various applications to fast reactor, high-temperature reactor, pebble bed reactor, molten salt reactor, and microreactor designs. Griffin development follows the software quality assurance procedure for MOOSE-based applications and with software requirements consistent with the ASME NQA-1 standard. Griffin has been adopted into the reactor analysis system for the U.S. NRC and is in use at U.S. companies, universities and national laboratories.
Fast Modular Reactor Nuclear Design Parameters of Fuel Cycle and Power Distributions
This paper describes conceptual design of a 112 MW thermal (50 MW electric) Fast Modular Reactor (FMR) system operating at 7 MPa with inlet/outlet temperatures of 509/800 ºC. The reactor system includes the reactor core, fuel assemblies, fuel rods, reactor internals, reflector, neutron control system, flow control, and structural components. The nuclear design and analysis were conducted to search for a baseline core with a cycle length greater than 8-year and the power peaking factor less than 1.5 by adjusting the fuel assembly, reactor core, and reflector configurations. The neutronics calculations of the baseline core showed that a refueling interval of 9-year is achieved with a total peaking factor of 1.47 and a fuel rod average linear power of 3.6 kW/m.
Fast modular reactor nuclear design parameters of fuel cycle and power distributions
This paper describes conceptual design of a 112 MW thermal (50 MW electric) Fast Modular Reactor (FMR) system operating at 7 MPa with inlet/outlet temperatures of 509/800 Celsius degrees. The reactor system includes the reactor core, fuel assemblies, fuel rods, reactor internals, reflector, neutron control system, flow control, and structural components. The nuclear design and analysis were conducted to search for a baseline core with a cycle length greater than 8-year and the power peaking factor less than 1.5 by adjusting the fuel assembly, reactor core, and reflector configurations. The neutron transport calculations of the baseline core showed that a refueling interval of 9-year is achieved with a total peaking factor of 1.47 and a fuel rod average linear power of 3.6 kW/m. (authors)
Radiological Releases from Novel Fuel Forms in Advanced Reactors During Severe Accidents for Consequence Analyses
Various advanced reactor developers are exploring the potential for reductions in the size of physical security forces and emergency planning zones. These reductions are based on robust fuel forms and inherently safe reactor designs. However, such reductions in physical protection measures could increase the risk of sabotage. To assess the possibility of reducing these measures, sabotage-induced radiological consequence analyses were carried out. These analyses considered accident scenarios that were beyond design basis accidents and overly conservative (Shah, 2025a; Shah, 2025b; Shah and Hartanto, 2026), yielding very large release fractions. These fractions, which can be used to evaluate physical protection and emergency planning requirements, have been crudely determined and applied as demonstrations for a sodium-cooled fast reactor (SFR) (Shah and Hartanto, 2025a), a high-temperature gas-cooled reactor (HTGR) (Shah and Hartanto, 2025b), a heat pipe–cooled reactor (HPR) (Shah and Hartanto, 2025c), and a molten salt–cooled reactor (MSR) (Shah et al., 2026). A Sandia National Laboratories (SNL) team used MELCOR—a fully integrated severe accident analysis code—to demonstrate the code’s capability to analyze advanced (i.e., not light water–cooled) reactors (including a fluoride salt–cooled high-temperature reactor [FHR]) and calculate radiological releases to the environment during severe accidents (Wagner et al., 2022a, 2022b, 2022c, 2023a, and 2023b). Although the analyses were carried out to demonstrate MELCOR’s growing capability, the release source terms were estimated for advanced reactors, providing valuable insights into the accident progression and radiological releases. These findings from prior SNL studies, including estimated source terms and related sensitivity studies, were leveraged to derive source terms for postulated sabotage-induced accidents. Insights from these sensitivity studies informed the scaling of SNL’s estimated source terms for the defined accident scenarios. The derived release fractions for the severe accident scenarios for the respective reactor designs can be used to perform more nuanced dose consequence analyses to evaluate the reactors’ physical protection and emergency planning zone requirements. These analyses are in accordance with the risk-informed, performance-based approach proposed under 10 CFR Part 53. This study builds on the prior source term analyses and associated sensitivity studies by SNL to derive time-dependent and design-informed release fractions. Section 2 describes the diverse advanced reactor designs analyzed by the SNL team. Section 3 discusses the severe accident analyses, the release fractions calculated, and the limitations and assumptions of the demonstration project. Section 4 presents the release percentages derived for the hypothetical sabotage-induced severe accidents at the advanced reactors. Section 5 summarizes the study’s findings and conclusions.
Techno-Economic Assessment of the Factory Production of Small Modular Reactor
The demand for electricity continues to grow around the world while pressures remain high to reduce carbon emissions. Nuclear energy provides a strong and reliable baseload power component without a high output of greenhouse gases. Global nuclear energy capacity is increasing with 62 reactors under construction in 14 countries. Nuclear energy will be crucial as energy demands are met with a sustainable and diverse energy portfolio. Small modular reactors (SMRs) are small scale (less than 300 MWe) reactors that incorporate the same advanced technologies included in recent LR nuclear reactor designs. SMRs are designed to include the latest reactor technology including passive safety features. SMRs offer economic advantages over LRs such as lower up-front capital and easier integration into the existing power grid. The modular characteristic of SMRs would allow incremental additions to power plants years after its completion. The cost of constructing SMR power plants will determine their future in the energy industry. Three SMR reactors are in advanced stages of design in the United States; the Westinghouse 225 MWe, mPower 180 MWe and nuScale 45 MWe. SMRs are expected to have lower capital expenses of reactor components due to factory production. Current estimates predict these savings to reduce SMR capital 30-40%. Existing studies have estimated SMR capital using economies of scale and holistic discounts applied to an overall capital expense figure, but do not include the expense of an SMR factory facility. To better understand the actual costs associated with construction of an SMR power plant, capital components will be broken down according to its place in the construction phase. The expense of a factory facility will also be included with SMR capital expenses to determine its significance and benefits to overall capital expenses. Results from this study show that SMRs can compete with LRs in terms of startup costs. The capital cost of building a factory for SMRs was previously overlooked for SMR capital estimates. The annualized capital expenses expected for an SMR factory are very small, but offer significant discounts to the overall capital costs for SMRs. Case 3 shows that factory production can reduce SMR capital expenses to the range of 30%-40% when compared to Case 1 in this study. Factory savings of this magnitude would require factory production of several SMR reactor and generator components including the reactor pressure vessel, control rods, steam generation, and chemical and volume control system, containment, condenser and turbo-generator. SMRs will also benefit from lower indirect expenses due to a much smaller construction duration. It is expected that SMRs will have capital costs competitive with LRs. The differences between Case 2 and Case 3 in this study show that SMRs are heavily dependent on the cost reductions coming from factory production. The factory facility represents a small increase in capital expense, but accounts for a large reduction of cost for factory components. SMRs have the potential to meet growing energy demands in the United States and around the world, but their economic viability is dependent on a factory facility that will lower the capital expenses of reactor and generator components. (authors)
Establishing a Database of High Temperature Salt Properties for Nuclear Applications
The Fluoride-Salt-Cooled High-Temperature Reactor (FHR) is a new reactor concept employing a high temperature liquid salt fluoride coolant. The thermophysical property data for candidate salt coolants is very limited. The uncertainties of these properties are quite high and not well documented. For researchers exploring reactor designs such as the FHR, a database of high temperature liquid salts for nuclear applications is being established. Researchers can use the database, which will be updated with new experimental data, to find thermophysical properties and associated uncertainties for salt coolants to use in their analyses. This paper serves as an introduction of the database and highlights the main three coolants in the database that are considered for use in the FHR design. The salts considered for the Fluoride-Salt-Cooled High- Temperature Reactor (FHR) include LiF-BeF{sub 2} (flibe) and NaFZrF{sub 4} (nafzirf) as potential primary loop coolants and LiF-NaFKF (flinak) as a secondary loop coolant. Flibe is the primary candidate for the primary loop coolant. It has the lowest neutron cross section, is chemically stable with low corrosion rates in high-nickel alloy systems, and is weakly radioactive with the primary activation product tritium resulting in very low radiation levels in the primary loop. Flibe does have some disadvantages, which include toxicity of beryllium and the need to enrich lithium to at least 99.995% Li-7 to maintain a low nuclear cross section. For design analysis, the baseline salt used is nafzirf because it has less attractive neutronic and thermophysical properties. The benefits of nafzirf over flibe including no direct tritium production and lower cost. By designing a reactor to operate using nafzirf, it is assumed that the reactor will operate better using flibe as the coolant. Thus, one can consider the coolants flibe and nafzirf to be bounding primary salt coolants for reactor design. (authors)
Numerical Simulation Comparison of Two Reactor Configurations for Chemical Looping Combustion and Chemical Looping With Oxygen Uncoupling
Chemical looping with oxygen uncoupling (CLOU) is a carbon capture technology that utilizes a metal oxide as an oxygen carrier to selectively separate oxygen from air and release gaseous O 2 into a reactor where fuel, such as coal, is combusted. Previous research has addressed reactor design for CLOU systems, but little direct comparison between different reactor designs has been performed. This study utilizes Barracuda-VR® for comparison of two system configurations, one uses circulating fluidized beds (CFB) for both the air reactor (AR) and fuel reactor (FR) and another uses bubbling fluidized beds for both reactors. Initial validation of experimental and computational fluid dynamic (CFD) simulations was performed to show that basic trends are captured with the CFD code. The CFD simulations were then used to perform comparison of key performance parameters such as solids circulation rate and reactor residence time, pressure profiles in the reactors and loopseals, and particle velocities in different locations of the reactor as functions of total solids inventory and reactor gas flows. Here, using these simulation results, it was determined that the dual CFB system had larger range for solids circulation rate before choked flow was obtained. Both systems had similar particle velocities for the bottom 80% of particle mass, but the bubbling bed (BB) obtained higher particle velocities as compared to the circulating fluidized-bed FR, due to the transport riser. As a system, the results showed that the dual CFB configuration allowed better control over the range of parameters tested.
Transformational Challenge Reactor Safety Design and Radionuclide Retention Strategy
An integrated safety design and radionuclide (RN) retention strategy is developed to support the Transformational Challenge Reactor (TCR) demonstration. This demonstration aims to showcase viability for rapid deployment of a novel reactor by leveraging the advances in materials, manufacturing, and computational sciences through a highly integrated and agile design and development approach. This strategy provides a logical description and understanding of how RNs are contained within the facility. Rather than discussing fission product barriers individually between separate design and safety basis reports, this paper provides a consistent description and narrative to better facilitate regulatory interactions and focus safety design efforts. The principal barriers credited include the various coating layers in the tristructural isotropic (TRISO) fuel particle, the silicon carbide (SiC) matrix hosting the particles within the fuel element, the helium pressure boundary, and the confinement system. The choice and assumed performance of the credited barriers are highly conservative, which is a direct reflection of the low hazard that the TCR demonstration presents and the need to simplify and focus the safety review process accordingly. However, the strategy and the associated framework are generalized and may be adopted and tailored to support other advanced reactor demonstration efforts.
Design changes in SNAP-8 for reduced reactor operating temperatures
SNAP 8 reactor design changes for reduced operating temperatures
Nuclear design and experiments for a space power reactor
Nuclear reactor design as heat source for electric power generation in space