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At least 19 records

Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term – Trial Calculation

The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Containment margins in FFTF for postulated failure of in-vessel post-accident heat removal

This study was performed to complete the assessment of potential melt-through consequences for the postulated failure of post-accident heat removal following a hypothetical core disruptive accident (HCDA) in the FFTF. The SPRAY, SOFIRE, CACECO, HAA-3B, and COMRADEX codes were used to determine potential site boundary radiation doses. Parametric comparisons were made to evaluate the effects of natural hydrogen-oxygen recombination, reactor cavity liner failures and containment coolers. Containment pressure relief, vent/purge options, and the effects of exhaust filtration were also included in the study. The potential site boundary radiation exposures for all options considered were within 10 CFR 100 guideline values.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Validation Study in SAS4A Code in Simulated Mild TOP Condition

Model improvements and validation efforts for SAS4A code, which is the most advanced computer code for simulating the the primary phase of the Core Disruptive Accident (CDA) of MOX-fueled Sodium-cooled Fast Reactors, have been performed utilizing the experimental data from the CABRI programs by JAEA (Japan Atomic Energy Agency), IRSN (Institut de Radioprotection et de Surete Nucleaire), CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) and KIT (Karlsruhe Institute of Technology). In order to confirm validity of improved models of SAS4A, a systematic and comprehensive validation plan was defined. It was based on the PIRT (Phenomena Identification and Ranking Table) result for CDA of MOX-fueled fast reactors. The fuel failure position and time are the most important parameters, which dominate the analytical results in a reactor calculation. The failure position will affect the reactivity feedback due to the fuel distribution after failure and consequently would have an effect on the energy release in the initiating phase. This paper describes the preliminary result of validation study for SAS4A code in simulated Transient Over-Power (TOP) condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Status Report on Ex-Vessel Coolability and Water Management

Specific to BWR plants, current accident management guidance calls for flooding the drywell to a level of approximately 1.2 m (4 feet) above the drywell floor once vessel breach has been determined. While this action can help to submerge ex-vessel core debris, it can also result in flooding the wetwell and thereby rendering the wetwell vent path unavailable. An alternate strategy is being developed in the industry guidance for responding to the severe accident capable vent Order, EA-13-109. The alternate strategy being proposed would throttle the flooding rate to achieve a stable wetwell water level while preserving the wetwell vent path. The overall objective of this work is to upgrade existing analytical tools (i.e. MELTSPREAD and CORQUENCH - which have been used as part of the DOE-sponsored Fukushima accident analyses) in order to provide flexible, analytically capable, and validated models to support the development of water throttling strategies for BWRs that are aimed at keeping ex-vessel core debris covered with water while preserving the wetwell vent path.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Thermal-Hydraulic-Mechanical Modeling and Simulation of Sodium-Potassium–Cooled MARVEL Microreactor Core

The U.S. Department of Energy's Microreactor Program, with Idaho National Laboratory's development of a nuclear microreactor applications test bed named MARVEL, aims to support R&D for the deployment of small, transportable reactors across civilian, industrial, and defense sectors. The MARVEL microreactor, an 85-kWth thermal fission reactor, incorporates TRIGA nuclear fuel and a sodium-potassium eutectic as its primary coolant, designed for safety and efficiency, with natural circulation eliminating the risk of critical heat flux conditions. The reliance on natural circulation for primary cooling means the reactor avoids using fuel spacers to minimize core pressure drop, which could disrupt the primary coolant's natural flow. However, the reactor core’s tight P/D ratio of 1.05, in the absence of fuel spacers, could pose a risk of fuel rod contact and increased peak cladding temperatures. To ensure the reactor safety, this study conducted computational modeling and simulations to investigate the reactor's thermal-hydraulic-mechanical characteristics, including the reactor core heat transfer coefficients, the potential for rod-to-rod contact, and assessed its impact on peak cladding temperature and overall reactor safety. The computational analyses of the MARVEL microreactor core revealed that the thermal deformation of fuel rods under worst-case scenario may lead to the fuel rod contact, but the peak cladding temperatures will remain significantly lower than the safety criteria, ensuring the safety operation reactor without fuel spacers under normal operating conditions.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in the relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.

97 MATHEMATICS AND COMPUTING↗

Experimental Study of Molten Metallic Fuel Relocation in Sodium-Filled Fuel Structures

The Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) is being developed by Korea Atomic Energy Research Institute (KAERI). While their design incorporates inherent safety features demonstrated in the landmark Shutdown Heat Removal Test (SHRT) conducted at EBR-II including a large pool configuration and uranium alloy fuel, it is still necessary to evaluate metal fuel relocation and theirs consequences in case of postulated core disruptive accidents. Extensive data regarding fuel relocation in sodium from both in-pile and out-of-pile tests were obtained for oxide fuels in the 1960's - 1970's, while few studies of metallic fuel behavior were performed because of concerns that metallic fuel could not achieve high burnup due to the irradiation induced swelling problem. Following new metal fuel design development that mitigated the swelling problem, experiments regarding interaction of molten metal fuels and liquid sodium have been conducted by several researchers. Most studies that were conducted using metal simulants in various liquids concluded that the molten metals were heavily fragmented after liquid contact and the dominant factors governing the fragmentation were superheating and latent heat of fusion of the molten metal fuel. Gabor et al. conducted experiments using kilogram quantities of various molten uranium alloys in an open sodium pool configuration. They found that the metal fuel fragments were basically in forms of filaments and sheets with a high bed voidage in the order of 0.9. Their calculation indicated that the debris beds formed by relocated metal fuels in an open sodium pool would be largely coolable by conduction heat transfer; and even if deep beds were to form, convective heat transfer and boiling heat transfer could preclude further melt penetration. Although Gabor's tests are useful in evaluating the coolability of relocated metal fuels during a postulated accident in a SFR, they focused on late phase behavior in which the core material relocates downward through the core support plate into the inlet plenum where the structures are more open and thus consistent with the open pool configuration. There is still a knowledge gap related to nature of metal fuel fragmentation and relocation within the fuel assembly structures. Argonne National Laboratory (ANL) is planning to conduct experiments regarding metal fuel relocation in the core structures which mock up the PGSFR's fuel assembly design. In the present study, preliminary tests for demonstration of melting and injecting the metal fuel were conducted using copper. Metal fuel relocation experiments with single fuel pin test section are planned to be conducted for understanding the flow and freezing behavior of the molten metal fuel in a narrow sodium channel and for evaluating the coolability of the relocated fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Runaway electron seed formation at reactor-relevant temperature

Systematic variation of the pre-disruption core electron temperature (T e ) from 1 to 12 keV using an internal transport barrier scenario reveals a dramatic increase in the production of ‘seed’ runaway electrons (REs), ultimately accessing near-complete conversion of the pre-disruption current into sub-MeV RE current. Injected Ar pellets are observed to ablate more intensely and promptly as T e rises. At high T e , the observed ablation exceeds predictions from published thermal ablation models. Simultaneously, the thermal quench (TQ) is observed to significantly shorten with increasing T e —a surprising result. While the reason for the shorter TQ is not yet understood, candidate mechanisms include: insufficiently accurate thermal ablation models, enhanced ablation driven by the seed RE population, or significant parallel heat transport along stochastic fields. Kinetic modeling that self-consistently treats the plasma cooling via radiation, the induced electric field, and the formation of the seed RE is performed. Including the combined effect of the inherent dependence of hot-tail RE seeding on T e together with the shortened TQ, modeling recovers the progression towards near-complete conversion of the pre-disruption current to RE current as T e rises. Measurement of the HXR spectrum during the early current quench (CQ) reveals a trend of decreasing energy with pre-disruption T e . At the very highest T e (≈ 12 keV), ≈ 100% conversion of the thermal current to runaway current is found. The energy of this peculiar RE beam is inferred to be sub-MeV as it emits vanishingly few MeV hard x-rays (HXRs). These measurements demonstrate novel TQ dynamics as Te is varied and illustrate the limitations of treating the RE seed formation problem without considering the inter-related dependencies of the pellet ablation, radiative energy loss, and resultant variations of the TQ duration. Finally, if the observed shortening of the TQ with increasing T e extends to fusion-grade plasmas, than their propensity to form large quantities of RE seeds at high T e may be far worse than previously thought. Positively, the high T e scenario in DIII-D produces REs so prodigiously that it can serve as a meaningful new platform for demonstrating RE avoidance techniques.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Viability of Data Analytics to Ascertain Component Performance for Additive Manufacturing

The Transformational Challenge Reactor (TCR) program is leveraging additive manufacturing (AM) technologies to produce multiple nuclear components to be assembled into a fully functional microreactor core. AM was selected as the main manufacturing technology for TCR because it has the potential to disrupt the nuclear industry on two fronts: (1) it enables the manufacturing of very complex geometries with optimized and tailored material properties for the intended use of the component, opening up new options for reactor designs, and (2) it allows for a better understanding of the manufacturing process through real-time in situ monitoring, data analytics, and artificial intelligence, which can lead to a streamlined qualification and certification process. Within TCR, the development and deployment of a digital platform aims at addressing the latter opportunity. As part of this effort, the collection of pedigree datasets is vital, should these data be generated before, during or after the manufacturing process. This report focuses on the development of data analytics tools enabling qualitative assessment of the manufacturing processes to automate the identification of flaws in in situ monitoring imagery and to correlate those flaws to the resulting mechanical testing data. Through examples, the report gives an overview of the data collected and presents a path forward for the digital platform.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Monitoring for Additive Manufacturing Technologies: Report on Progress, Achievements, and Limitations of Monitoring Techniques

The Transformational Challenge Reactor (TCR) program is leveraging additive manufacturing (AM) technologies to produce multiple nuclear components to be assembled into a fully functional microreactor core. AM was selected as the main manufacturing technology of choice for TCR because it has the potential to disrupt the nuclear industry on two fronts: (1) it enables the manufacturing of very complex geometries with optimized and tailored material properties for the intended use of the component, opening up new options for reactor designs, and (2) it allows for a better understanding of the manufacturing process through real-time in situ monitoring, data analytics, and artificial intelligence that can lead to a streamlined qualification and certification process. Within TCR, the development and deployment of a digital platform aims at addressing the latter opportunity. As part of this effort, the collection of pedigree data sets is vital, should these data be generated before, during, or after the manufacturing process. This report focuses on the collection of manufacturing process data using systems instrumentation and monitoring. It presents an inventory of the various monitoring modalities installed on the manufacturing systems by the machine manufacturer or specifically designed by our team to fill information gaps. It also provides details on other relevant data streams available and leverageable to evaluate the manufacturing process on all classes of AM systems investigated during the first year of this program. Through examples, the report gives an overview of the data collected and presents a path forward for the digital platform.

42 ENGINEERING↗

A laser inverse Compton scattering diagnostic to study runaway electron dynamics during tokamak disruptions

The goal is to design a new diagnostic based on laser inverse Compton scattering (LICS) to study the dynamics of runaway electron formation and their subsequent loss during triggered disruptions in DIII-D. It is known that better measurements of the runaway electron population, throughout the evolution of a disruption, is critical for mitigating the impact of energetic electron beams especially in high plasma current discharges such as those in reactor relevant burning plasmas. The goal of the LICS approach being designed for DIII-D is to measure the timeresolved spatial profile and energy distribution of the runaway electrons while they are in the core of the tokamak plasma. During the last year, a code was developed to simulate the signals from a new laser scattering diagnostic for DIII-D, which will measure the relativistic electron population expected in DIII-D after disruptions triggered by argon pellet injection.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Demonstration of Tokamak Discharge Shutdown with Shell Pellet Payload Impurity Dispersal

The first rapid tokamak discharge shutdown using dispersive core payload deposition with shell pellets has been achieved in the DIII-D tokamak. Shell pellets are being investigated as a possible new path toward achieving tokamak disruption mitigation with both low conducted wall heat loads and slow current quench. Conventional disruption mitigation injects radiating impurities into the outer edge of the tokamak plasma, which tends to result in poor impurity assimilation and creates a strong edge cooling and outward heat flow, thus requiring undesirable high-$Z$ impurities to achieve low conducted heat loads. The shell pellet technique aims to produce a hollow temperature profile by using a thin, low-ablation shell surrounding a dispersive payload, giving a greatly increased impurity ablation (and radiation) rate when the payload is released in the plasma core. This principle was demonstrated successfully using 3.6 mm outer diameter, 40 μ m thickness diamond shells holding boron powder. The pellets caused rapid (<10 ms) discharge shutdown with low conducted divertor heat fluence (~0.1 MJ/m 2 ). Confirmation of massive release of the boron powder payload into the plasma core was obtained spectroscopically. Some evidence for the formation of a hollow temperature profile during the shutdown was observed. Furthermore, these first results open a new avenue for disruption mitigation research, hopefully enabling development of highly effective methods of avoiding disruption wall damage in future reactor-scale tokamaks.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

A sustained high-temperature fusion plasma regime facilitated by fast ions

Nuclear fusion is one of the most attractive alternatives to carbon-dependent energy sources. Harnessing energy from nuclear fusion in a large reactor scale, however, still presents many scientific challenges despite the many years of research and steady advances in magnetic confinement approaches. State-of-the-art magnetic fusion devices cannot yet achieve a sustainable fusion performance, which requires a high temperature above 100 million kelvin and sufficient control of instabilities to ensure steady-state operation on the order of tens of seconds. Here, in this study, we report experiments at the Korea Superconducting Tokamak Advanced Research device producing a plasma fusion regime that satisfies most of the above requirements: thanks to abundant fast ions stabilizing the core plasma turbulence, we generate plasmas at a temperature of 100 million kelvin lasting up to 20 seconds without plasma edge instabilities or impurity accumulation. A low plasma density combined with a moderate input power for operation is key to establishing this regime by preserving a high fraction of fast ions. This regime is rarely subject to disruption and can be sustained reliably even without a sophisticated control, and thus represents a promising path towards commercial fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Improving the modeling of near-wall interphase heat transfer in porous media models of Pebble Bed Reactors

Here, this work aims to improve capabilities for modeling localized effects in porous media models of Pebble Bed Reactors. The wall-channeling effect is the primary local phenomenon of interest in a PBR, where the presence of the reflector wall disrupts the pebble packing, causing the pebbles near the wall to pack less efficiently and creating large void regions. Accurate modeling of the near-wall region is important as it will affect core bypass flow and temperature predictions. Porous media models are commonly used for design scoping and plant-level simulations of PBRs. Although these models have some capabilities to model the near-wall region, the correlations that are available in porous media codes are often inaccurate when a multi-region model is used to discretize the near-wall region. This work employs a high-to-low analysis to study the accuracy of available interphase heat transfer closures. NekRS, a spectral element computational fluid dynamics code, is used to perform Large Eddy Simulations. These LES simulation results are compared to porous media model results from the Pronghorn porous media code. The friction term of the KTA drag closure is first improved, reducing the error in the prediction of the near-wall velocity from over 50% to less than 5%. This is combined with improvements to the form term from previous works to produce a drag closure that is capable of accurately modeling the wall-channeling effect across a variety of flow conditions. The Nusselt number predictions of several heat transfer correlations are compared to the high-fidelity results where it is found that the KTA heat transfer correlation is capable of accurately predicting the local Nusselt numbers that were determined in the high-fidelity simulation. Comparison of the radial solid temperature profiles, however, reveal discrepancies between NekRS and Pronghorn. It is discovered that the implementation of the interphase heat transfer coefficient that exists in many current porous media codes is not valid when local porosities are modeled. Instead, it is suggested that the interphase heat transfer coefficient should be dependent on the local porosity, the Nusselt number, and the local solid surface-to-volume ratio. Implementation of this change produces improvement in the agreement between the results obtained by NekRS and Pronghorn while using the KTA heat transfer correlation.

interphase heat transfer↗

Overview of the KSTAR experiments and future plan

The Korea Superconducting Tokamak Advanced Research (KSTAR) device has recently entered a new phase of operation following the installation of a tungsten mono-block divertor, providing a unique platform to investigate high-performance, long-pulse plasma scenarios for future reactors. The 2023–2024 experimental campaigns demonstrated significant progress in commissioning the tungsten divertor, developing plasma operation scenarios, and achieving real-time control. Comparative studies revealed increased core radiation losses and associated performance degradation relative to the carbon divertor environment. Alleviation strategies, such as optimized neutral beam injection timing, boron powder seeding, and impurity control through gas puffing, were shown to reduce tungsten accumulation and extend high-performance phases, including record H-mode operation exceeding 100 s. International collaboration enabled the first demonstration of high poloidal beta scenarios with an internal transport barrier on KSTAR. Advances in control included upgraded real-time resonant magnetic perturbation schemes and disruption forecasting. Furthermore, divertor detachment control using a new surrogate-model-based approach with real-time radiation imaging demonstrated active impurity and heat flux management. Supported by enhanced heating, current drive, and diagnostic systems, these achievements offer valuable insights into tungsten impurity behavior, transport physics, and control strategies for future reactors including ITER. Planned upgrades toward a full-tungsten wall and ITER-aligned real-time plasma control will further establish KSTAR as a leading experimental platform for developing operational scenarios required to achieve steady-state, burning plasma conditions.

KSTAR↗

Tokamak level performance in the optimized stellarator Wendelstein 7-X with stable peaked density profiles

Progress towards achieving net gain from nuclear fusion is typically characterized by the triple product 𝑛⁢𝑇⁢𝜏. Although stellarators have several benefits for a reactor, such as the inherent steady-state capability and the lack of disruptions, the maximum achieved triple product has historically lagged behind that of tokamaks. In the large optimized stellarator Wendelstein 7-X, a triple product of (1.10 ± 0.15) × 10 20 m −3 ⁢ keVs has recently been achieved and held stable for 1.9 s. This marks an important step in stellarator core plasma performance and puts it in line with 𝐻-mode performance in similar sized tokamaks. The progress in Wendelstein 7-X performance is due to the suppression of ion-temperature-gradient turbulence by strong density gradients. In the record case, these are maintained by a fine balance of the core particle transport and central fueling using neutral beam injection. The duration of the scenario was limited in time only by technical restrictions of the neutral beam injectors and control of the divertor heat loads. This paper presents the result in the context of the leading magnetic confinement devices along with details of the scenario.

Bannmann, Sebastian [Max Planck Institute for Plas↗

Regulatory and Licensing Strategy for Microreactor Technology

With electricity demand on the rise, an increase in the use of nuclear energy is needed in order to meet future energy demands. The use of nuclear energy will not only begin to positively address this demand, but will also mitigate the negative impacts of carbon-induced climate change and provide energy security to the United States. An appropriate option to address the increasing energy demand is the use of advanced reactor systems like microreactors. Microreactors, often referred to as special-purpose reactors or vSMRs, are neutronically simple, factory manufacturable, easily transportable, and designed to produce less than 20 MWt (for classification as Hazard Category 2 per 10 CFR 830, DOE STD 1027). These reactors are decentralized energy sources that have the ability to provide sustainable and affordable power to remote communities and to industrial users, while having self-contained geometry that requires very low maintenance care. In order for microreactors to be approved for operation, they must abide by and successfully complete a rigorous regulatory and licensing process monitored by the Nuclear Regulatory Commission (NRC), the Department of Energy (DOE), or the Department of Defense (DoD). Because microreactors are a relatively new and disruptive technology, there is no clear licensing path for them to follow. Current licensing frameworks are tailored specifically to large, traditional light water reactors (LWRs) that have been developed and understood by the nuclear industry for many years. Unlike LWRs, microreactors have major design differences, specifically in the materials, coolant, reflectors and potential technology applications used. They also have higher in-core operating temperatures, which have the potential to push the materials used past their allowable limits. Because they are much smaller, safety margins, the amount of space required for assembly, and the distances required to meet dose-based regulatory criteria can technically be altered to better fit their new size. The following strategies and recommendations are intended to assure that key licensing issues do not influence the path to microreactor development, demonstration, and commercialization: Nuclear Regulatory Commission • The NRC should address the lack of knowledge and support of risk informed decision-making concepts from staff members and continue to advance and communicate risk initiatives using their Risk Informed Steering Committee • More research should be done into whether or not microreactors, which are small Class 103 facilities, can be licensed using Class 104-type guidance (e.g., NUREG 1537) • The NRC should use frequency-consequence curves to incrementally reduce regulatory uncertainty and risk to facilitate future construction and operation of advanced reactor designs Department of Energy • The DOE should continue to implement their multi-step Phenomena Identification and Ranking Table process in order to help detect and recommend the major design and technology work needed for licensing future reactor systems • The DOE should consider a joint licensing approach with the NRC where microreactors first go through the DOE process, then finish with the NRC process, providing an advantage for both applicants and NRC staff Department of Defense • The DoD should invest in personnel with sufficient expertise, incorporate their military reactors into tactical power systems, and self-regulate using Section 91b of the Atomic Energy Act Transportation • Licensees should convert enriched uranium hexafluoride into metal or oxide at an enrichment facility and then transport them to a fabrication facility • To avoid transportation issues altogether, licensees should consider positioning the enrichment and fabrication facilities on the same site • The DOE should provide their support in developing new shipping packages for safe transport, criticality benchmark data needed to license high-assay low-enriched uranium fuel facilities and transport packages, and provide financial and technical help to the Department of Transportation; and, • The NRC should help transportation efforts by finalizing and developing guidance for implementing 10 CFR 73 and 10 CFR 74 specifically for Category 2 Special Nuclear Material.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗