The advancement of fusion energy, heralded as an innovative, environmentally sustainable, and clean alternative to traditional energy sources, necessitates a comprehensive reevaluation and enhancement of the standards and criteria used to assess the suitability of materials for fusion reactors. This study underscores the reforming and simplifying of the current regulatory framework applicable to activated materials for fusion energy. It highlights the unique environmental properties inherent to plasma systems, requiring the adoption of materials that may not fully align within an established conventional regulatory framework. Existing systems and structures governing regulations and compliance, largely modeled on the rules and standards established for fission reactors, may impose excessively stringent constraints that could impede the advancement and innovation of new technologies. In addition, fusion systems would produce significantly less long-term (>100,000 years) radioactive waste per unit energy generated than fission systems, so the existing regulatory framework based upon fission systems is unnecessarily conservative. By implementing a comprehensive framework that thoroughly accounts for the unique properties and behaviors of radionuclides, along with a detailed assessment of the environmental impacts of different materials, innovation can be responsibly advanced while maintaining safety standards and adherence to regulatory requirements. A judicious and restricted use of activated materials is recommended by the integration of advanced waste management strategies and a comprehensive understanding of these materials during operating conditions. Studies show that several activated products from candidate materials for fusion applications will not meet the existing strict activity limits, either as the main elements or as additions. For example, only about 7.1 % of the blanket’s front wall tungsten volume in the Steady State Tokamak Reactor (SSTR) could generate 2,000 times the amount of 192nIr above the allowed limit. The regulatory framework should consider relaxing the criteria for fusion-activated materials by allowing higher activity levels, as the fusion waste decays rapidly and most materials require isolation periods of less than 100 years. The proposed relaxed criteria achieve a balanced integration of optimal performance, enhanced safety measures, and environmental sustainability, thereby promoting the development and adoption of fusion technology as a reliable and viable energy source for the future.

In this contribution, we present a new package for creating fixed-source neutron source profiles for OpenMC from both simulated and experimental tokamak plasma states. Realistic fusion neutronics simulations ultimately require information on plasma equilibrium, plasma transport, and a realistic source geometry. In this work, a new flexible tool is presented that allows the user to input detailed plasma states to create OpenMC neutron sources for fixed-source problems. The toolkit accepts radially and poloidally varying plasma density and temperature profiles from either experimental or simulated plasma states. Either D-T or D-D reaction rates may be used in the case of predictive or experimental simulations, respectively. This process provides a realistic neutron source that is intrinsically coupled to tokamak plasma physics parameters.We demonstrate the framework using a source modeled after an experimental neutron profile from the Mega Ampere Spherical Tokamak-Upgrade (MAST-U). An additional feature of this coupling is the ability to self-consistently convert the Monte Carlo tallies using total volumetric neutron production instead of relying on a measured total neutron production rate typical of these simulations. Experimentally informed and flexible source definitions for neutronics modeling are crucial to streamlining the design process of any future fusion pilot plant studies.

The shutdown dose rate (SDR) is critical for developing a plan for safe operation, establishing a maintenance scheme, and guiding potential design changes in a fusion energy system. This study identifies lithium ceramic blanket materials that produce lower dose rates compared to the leading dual-coolant lead-lithium breeder blanket concept. The aim is to find an optimal balance between maintaining a manageable SDR and achieving improved tritium recovery and thermal performance with ceramic blankets. This study was conducted on a 22.5-deg symmetric sector model of the Fusion Energy System Studies-Fusion Nuclear Science Facility (FESS-FNSF) and presents effective dose calculations for key components of the reactor as a result of using various blanket materials. The deuterium-tritium operational phase of the FESS-FNSF will last approximately 2.75 years and be used to define the neutron source. A Rigorous 2-Step workflow was used in OpenMC to calculate SDRs after shutdown, 1 day, 1 year, and 100 years for several lithium ceramic breeders. Across all time intervals, the PbLi blanket consistently exhibited the lowest SDR in the plasma region, though it contributed more to SDR in the outboard and inboard regions. The LiAlO2 blanket produced the highest SDR throughout, while Li4SiO4, Li2TiO3, and Li8ZrO6 provided the lowest SDRs among the ceramics. Regarding the effect of the beryllium multiplier, the pure Be multiplier led to lower SDRs, followed by Be12Ti, and finally Be12V. For all blanket materials tested at these different time periods, the SDR was found to exceed the limit of 2 x 103 mu Sv/h, thus suggesting the need for remote or robotic maintenance.

Neutron transport calculations have been per-formed on advanced shielding materials. Metal hydrides andborohydrides were evaluated to find an alternative to tungstencarbide (WC), which is used in the in-vessel components. Thestudy was conducted using a 22.5 degrees sector and a detailed 360 degrees geometry of the Fusion Energy System Studies-Fusion NuclearScience Facility (FESS-FNSF) using OpenMC and FENDL-3.2bnuclear data library. The neutronics analysis in this article wasconcentrated on calculating total nuclear (neutron and photon)heating at the magnet and the radiation damage of the inboardvacuum vessel (IBVV). For example, yttrium hydride (YH2) andvanadium hydride (VH2) showed lower radiation damage valuescompared to WC and other metals. Whereas alternative shieldingmaterials did not show a significant change in the magnet nuclearheating

OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. It can perform fission simulations, such as fixed-source, k -eigenvalue, and subcritical multiplication calculations on models built using either a constructive solid geometry (CSG) or CAD representation. To explore the use of OpenMC for fusion activation analysis, a detailed model of the Fusion Neutronics Science Facility (FNSF) was first developed for comparisons against an existing SERPENT model. A 90 $^{\circ}$ model of FNSF in standard-triangle language (STL) CAD format was converted to CSG using each code’s built-in functions, and the geometries were validated by ensuring no cells overlapped and no particles were lost during simulations. The neutron fluxes were calculated and compared for multiple components close to the plasma. The results show differences mostly below 1% in fluxes and averaged 8% for activity and decay heat. The work described in this study tests the CAD-based geometry using the DagMC toolkit in OpenMC and compares the activation analysis of OpenMC to SERPENT code.

This study is focused on a comparison of experimental neutron flux data from the VR-1 reactor with results from Serpent calculation. Serpent data were calculated by the dynamic part of the code and the steady-state criticality calculation. The Serpent calculation was performed with the JEFF 3.3 and ENDF/B-VIII.0 nuclear data library. This paper expands previous work and focuses on the integral control rod worth. Comparison of the ENDF/B-VIII.0 and JEFF 3.3 nuclear data libraries with experimental data is especially valuable. The comparison shows better performance of JEFF 3.3 nuclear data instead of ENDF/B-VIII.0 for the transient calculation of the cases presented in this article. In addition, the comparison of static and dynamic reactivity provided an interesting difference, especially for small rod insertion. Moreover, the study confirms that the Serpent dynamic external source simulation mode provides an accurate description of reactor behavior, and that the shape of the integral control rod worth is close to one experimentally measured.

The goal of our research is to build upon the capability of RELAP5-3D to model molten lead systems. Molten lead has several potential uses in future advanced reactors, like the lead fast reactor or fusion reactors that utilize dual-coolant lead lithium blankets. This potential for use in future generations of reactors highlights the necessity of developing molten lead models to ensure that they can accurately predict thermohydraulic behavior. We have developed a RELAP5-3D model of the Lobo Lead Loop facility located at the University of New Mexico to verify the accuracy of RELAP5-3D via comparison to existing computational fluid dynamics results and analytical calculations. It was found that RELAP5-3D accurately calculated radiative heat transfer (within <1%) when compared to theoretical calculations. In addition, pressure drop calculations done in RELAP5-3D demonstrated reasonable agreement within 20 kPa, mostly within similar to 7% to 15%, when compared to the computational fluid dynamics model of the facility developed by the University of New Mexico, and captured the dependence of pressure drop on flow velocity accurately. Finally, a hypothetical loss-of-flow transient was imposed on the RELAP5-3D model to determine the feasibility of performing a similar experiment with the Lobo Lead Loop. It was found that such an experiment could be possible, as the RELAP5-3D model indicated that the temperatures of the fluid would not exceed the limiting temperatures of the structure (1658 K) nor the maximum temperature of the electromagnetic pump inlet (823 K). Although there are no experimental data to begin validation, the model will be readily available for future validation studies when the experimental data are generated, especially as the model continues to evolve over time. The results so far demonstrate a promising first step in the verification/validation of the RELAP5-3D model of the Lobo Lead Loop. The highlights from our research are as follows: 1. The Lobo Lead Loop facility at the University of New Mexico is a good candidate for molten lead system code validation. 2. The Lobo Lead Loop currently has extensive pressure drop results from a high-fidelity computational fluid dynamics model, which offers the opportunity for code-to-code verification of pressure drop in RELAP5-3D. 3. A RELAP5-3D model of the Lobo Lead Loop has been developed to begin verification studies and to prepare for potential validation studies. 4. Development of the model will continue throughout the future to prepare for potential validation studies.

The Fusion Energy System Studies Fusion Nuclear Science Facility (FESS-FNSF) concept represents a transitional step between ITER and a commercial fusion power plant. The FNSF is a conceptualized D-T fueled tokamak with 518 MW of fusion power that has been extensively used to explore and optimize design features. The energetic 14.1-MeV neutrons can produce significant localized heating and activations, and can cause damage to plasma-facing components, which can determine maintenance/outage scheduling needs and also impact the lifetime of the device as a whole. This study illustrates a neutronics analysis that was conducted on a 22.5-degree symmetric sector of the FNSF with the goal of understanding the neutron heating and radiation damage that can be characterized by quantifying the displacements per atom (dpa).Concurrently, this study also focused on the development of analysis capabilities by converting a three-dimensional computer-aided design model of the FNSF into MCNP6.2 input using the McCad code. Accordingly, some confirmatory results on tritium production and the tritium breeding ratio (TBR) are provided to support model validation. The results produced by MCNP6.2 simulations showed that the highest heating and damage occurred in the outboard region, which concentrated approximately 290 MW of the total nuclear heating, in contrast to 97 MW within the inboard region. These results are consistent with previous studies that employed earlier versions of the FNSF concept and different modeling approaches.This study also provides additional details on neutron wall loading, as well as total heating from neutrons and gammas, results which show the total heating of the device (16 sectors) is approximately 477.83 +/- 0.80% MW, indicating a neutron energy multiplication factor of 1.15. Additionally, the capability to calculate hydrogen and helium production, as well as dpa, is illustrated. Finally, the neutronics effects of using alternative materials to tungsten carbide were evaluated for the vacuum vessel, low-temperature shield, and structural ring components, which showed that compounds like YH2, Mg(BH4)(2), and ZrH2 could reduce the total heating on the magnet and also reduce the TBR.

Pebble bed reactor fuel has both high enrichment and low fuel density. Previous research by the authors has determined the peak possible keff by examining the effect of water infiltration of all possible densities in a variety of container sizes. The previous research has found that for larger containers, the established sub-critical limit of 0.95 if exceeded. These current research has examined two possible criticality mitigation methods. The examined container is the largest of the previously examined ones, the MPC-37. The first method involves adding 9 hollow stainless steel rods that span the length of the container into the model. 1 in the middle, and 8 at the midpoints between the center cylinder and the wall of the container. By displacing fuel pebbles in the most active regions of neutron multiplication, keff is most effectively reduced. In the second method, dummy pebbles comprised of stainless steel or boron carbide replace at random fuel pebbles in the container with various replacement fractions. The stainless steel displacement rods cut the value of peak keff from roughly 1.33 to 0.94, a difference of ~40,000 pcm at the cost of ~15% of the container's capacity. Dummy pebbles are roughly twice as effective, cutting keff down a similar amount with only a 7% replacement required for the stainless steel pebbles, and 3% replacement required for the boron carbide pebbles.

This paper presents a parametric study of the Fusion Energy System Studies-Fusion Nuclear Science Facility’s (FNSF’s) tritium breeding performance for several solid breeder concepts, neutron multiplying materials, and blanket materials, assuming volume fractions based on the most recent FNSF design as a realistically representative fusion facility. In this study, we initially surveyed the tritium breeding ratio (TBR) of several solid breeder concepts by employing a simplified but efficient one-dimensional (1-D) infinite cylinder reduced-order model (ROM). Parametric studies were performed with the ROMs for the full range of breeder-to-multiplier ratios to identify the optimum mixture compositions for each breeder type that would lead to a maximum TBR.These optimized breeder-multiplier combinations were then homogenized with FNSF blanket component materials to estimate their impacts on the TBR. Subsequently, as a validation step for the optimal designs, TBR calculations were performed using a more realistic modified 1-D ROM with inner and outer breeding regions, as well as with a fully detailed 22.5-deg three-dimensional (3-D) sector of the FNSF to assess the impact of geometry details on the TBR. The differences between the two 1-D models were negligible, while the ROMs were able to correctly predict trends and identify the maximum and minimum TBR cases, as well as show consistent biases relative to the results produced by the full 3-D, 22.5-deg sector for specific breeder/multiplier combinations.Solid breeder concepts such as Li2O, Li4SiO4, and Li8ZrO6 outperformed all others in this study in terms of TBR performance when combined with all the neutron multiplier materials selected. An underlying goal of this study was to develop and improve rapid and reliable ROMs to aid designers during parametric optimizations of highly complex and computationally expensive fusion models.

This study describes an application of the SERPENT 2 code with the TENDL-2017 nuclear data library and the latest available model features of the Fusion Energy System Studies-Fusion Nuclear Science Facility (FNSF), to evaluate the activation of components after shutdown at 1, 10, and 100 years, assuming a plant lifetime of 8.5 full-power years. The primary parameters evaluated include the specific activity, decay heat, and waste disposal rating (WDR). The specific activity and decay heat are calculated with SERPENT 2 using a 360-deg model of the FNSF, while the WDR is calculated and classified based on the waste disposal limits established by the U.S. Nuclear Regulatory Commission under 10 CFR 61.55 as well as by using the Fetter approach.A python-based script developed for a previous high-level waste classification and analysis study was implemented and adapted to this research to calculate the WDR by comparing nuclide concentrations to the values established in 10 CFR 61.55 to generate a waste classification for each component surveyed. As only three short-lived isotopes have limitations for classifications beyond Class A, of which only Ni-63 is present in appreciable quantities, there is a limit to the amount that short-lived isotopes contribute to the most significant waste analyzed here. In most cases, a handful of long-lived isotopes can be problematic, such as Ni-59 and Nb-94, for example, which are solely responsible for multiple Class C classifications.The results herein reported heavily depend on the specific materials and mass/volume fractions in the specific model used in this study, which has changed and evolved since the inception of the FNSF concept and past studies. Therefore, the more significant contributions of this study may be the development of a modeling and simulation toolkit and a strategy to perform these calculations, so to help evaluate and optimize future fusion facilities.

Abstract A numerical framework for modeling depletion and mass transport in liquid-fueled molten salt reactions is presented based on exponential time differencing. The solution method involves using the finite volume method to transform the system of partial differential equations (PDEs) into a much larger system of ordinary differential equations. The key part of this method involves solving for the exponential of a matrix. We explore six different algorithms to compute the exponential in a series of progression problems that explore physical transport phenomena in molten salt reactors. This framework shows good results for solving linear parabolic PDEs with each of the six matrix exponential algorithms. For large problems, the series solvers such as Padé and Taylor have large run times, which can be mitigated by using the Krylov subspace.

This study examines the production of 232U via neutron irradiation. Uranium-232 is considered for use as a tracer in nuclear fuel. However, a source of 232U is needed. This study examines the production of 232U via neutron irradiation of targets constructed out of either 231Pa or 230Th in the High Flux Isotope Reactor. HFIR targets were modeled in MCNP to determine 232U yield in protactinium and thorium targets. Flux tallies were used inside the targets to determine the neutron flux inside each target. This flux was then used in SCALE 6.2 ORIGEN to determine the 232U yield, as well as the buildup of the byproduct 233U. Several 230Th enrichments were examined to determine how 232U yield is affected by 230Th enrichment, as well as the effects of the presence of 232Th. The buildup of 228Th and 229Th in thorium targets was also examined, as these isotopes may impact the feasibility of recycling of thorium target materials.

Molten salt reactors (MSRs) are a class of next-generation nuclear reactors that have received recent industrial and research interest. A generalized species transport solver was implemented in the Virtual Environment for Reactor Applications (VERA) computing suite to extend this tool to analyze liquid fueled MSRs. This core simulator has been extended to model the transport of fission product gases into a collection of circulating gas bubbles with the purpose of removing the gases. This paper presents the governing species transport equation, along with various nuclear source terms. Development of the source term for phase migration is discussed, along with a simplified interfacial area tracking method. Finally, a case study on a simplified MSR loop is presented in which modeling parameters were varied to assess their impact on gas removal. The steady state results show that parameters such as bubble diameter, gas injection rate and mass transfer coefficient have a low to moderate effect on the fraction of xenon in the core region. Removal efficiency has the greatest effect on the fraction in the core region. After the pump bowl, bubble diameter has a minor effect on the fraction of xenon in the gas void. These results point out that increasing parameters such as mass transfer coefficient, gas injection rate, and removal efficiency drives the xenon into the circulating gas void, while decreasing bubble diameter also drives xenon into the gas void by increasing interfacial area. (c) 2021 Elsevier Ltd. All rights reserved.

We describe a novel fusion reactor blanket concept called GAMBL – GA Modular BLanket. This design concept exploits the advantages of SiC-based structures while minimizing their limitations. Material microstructures are tailored for different functions throughout the first wall and blanket, including the use of graded W/SiC composites to enhance heat transfer capabilities and resist erosion at the plasma-facing surface. The design contains a decoupled first wall and breeding blanket, which we show can provide adequate tritium breeding. By decoupling the breeding part of the blanket, we can operate at very low PbLi breeder pressure and support gravity loads in simple, radiatively-cooled structural beams that do not contain coolant. First wall performance capabilities are enhanced by eliminating constraints imposed by the deep blanket. We have performed initial fluid, thermal, mechanical and neutronics analysis of this concept to show its capabilities to meet the design requirements. R&D needs are mostly related to materials development. Except for unique manufacturing needs, no new facilities are required beyond existing or planned facilities for the US base program on steel-based DCLL blankets.

The isotope U-232 is being considered for additive to uranium fuel for use as a tracer. The U-232 decay chain has high energy gammas that can be used for tracer purposes. However, the presence and intensity of such gammas may increase effective dose rates to workers around such materials This study examines the dose rate from different uranium materials that varying amounts of U-232 has been added to. Several materials and their respective storage geometries are modeled for particle transport calculations. For each material and U-232 concentration, gamma source terms were generated using SCALE 6.2 ORIGEN. These source terms were then used in MCNP for each material and geometry. Upon analysis, it was determined that the baseline dose from enriched uranium dominates until the U-232 concentration reaches about 10-100 parts per trillion (ppt), after which the dose rate increases linearly.

Our research aims to build a multiphysics framework for transient analysis of the Dual-Coolant Lead-Lithium (DCLL) blanket design of the proposed Fusion Nuclear Science Facility (FNSF) for all Anticipated Operational Occurrences (AOOs) using Reduced Order Models (ROMs). This paper presents an assessment of using RELAP53D for transient thermal-hydraulic blanket analysis of the DCLL design. Preliminary validation studies for the RELAP5-3D properties were conducted using a representative vertical flow loop model and a comparison study using experimental data from thermal convection corrosion loops at ORNL. These studies inspected the thermal hydraulic response of Lead-Lithium Eutectic (PbLi) systems within the RELAP5-3D code; mainly pressure drop and heat transfer, as compared to analytical calculations and experimental data. The Magnetohydrodynamic (MHD) pressure drop effect for liquid metal under the effects of a uniform magnetic field was implemented into our models using an equivalent forms loss coefficient based on the phasic momentum equations. We verified our implementation of MHD pressure drop within RELAP5-3D using a systematic approach consisting of two studies. Models were developed to perform comparison studies using experimental data from the Argonne Liquid metal EXperiment (ALEX) facility and the Magnetohydrodynamic PbLi Experiment (MaPLE) facility. Using the developed validation basis, we built a simplified DCLL channel model within RELAP5-3D, inspected MHD pressure drop and performed transient temperature analysis. The DCLL channel model used radial heating data from the International Thermonuclear Experimental Reactor (ITER) which was normalized to the FNSF design. A representative startup transient was performed using a relative power curve based on current PWR power ramp limits. This study demonstrated that RELAP5-3D ROMs are capable of accurate transient analysis of the DCLL blanket. We have since utilized MCNP to develop heating profiles based on the FNSF design and are working toward implementing them into our DCLL model.

We assessed the hydrogen migration and redistribution model implemented in the BISON fuel performance code for component-level tritium transport applications in fusion and fission example scenarios. We developed BISON models of an ITER heat exchanger, light water reactor fuel cladding, and a fluoride salt cooled high temperature reactor heat exchanger. Reasonable agreement between reported values in reference studies and BISON predictions demonstrated the ability of the BISON models to predict tritium transport behavior through steel materials. Next, sensitivity and uncertainty analysis was used to understand key parameter sensitivities within the BISON model for evaluation of tritium migration and redistribution. . The sensitivity analysis showed that the diffusivity activation energy was the most important parameter and that the heat of transport was the least important parameter for tritium migration in steels. Overall, the capabilities of the BISON code for component level modeling of tritium transport are promising.

Working toward the development of commercial fusion reactor technology, RELAP5-3D was selected for the thermal-hydraulic analysis of the Dual Coolant Lead-Lithium (DCLL) blanket within the proposed Fusion Nuclear Science Facility (FNSF) design. Having been developed by the DOE with a validation basis for the licensing of commercial light water reactors (LWRs) [1], RELAP5-3D is an excellent candidate for developing a validation basis for commercial fusion systems. This will be done by performing safety analyses over all anticipated operational occurrences (AOOs), as similarly applied to current LWR safety analysis [2]. Transients that will be considered are changes in heat removal by the secondary system, changes in blanket flowrate, changes in blanket coolant inventory, and coolant behavior following changes in plasma power [2]. With this basis for analysis, we can determine which aspects of LWR safety analysis are relevant to this type of fusion reactor and define an appropriate set of AOOs for fusion safety analysis.