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Fratoni, Massimiliano

Publications and source records attributed to Fratoni, Massimiliano.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Uranium Mononitride (UN) Handbook

Uranium Mononitride (UN) is being reexamined as a high temperature advanced nuclear fuel in many reactor applications due to its high thermal conductivity, high melting point, and high fissionable uranium density, among other desirable properties. A thorough review of the experimental data for the material and irradiation properties has been collected and is presented in this handbook. As an outcome of this review, in many cases, it is apparent that additional experimental verification must be conducted to verify the available data and to verify the suggested empirical correlations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Efficient continuous Energy-Multigroup hybrid depletion scheme using the Shift Monte Carlo code. Part I: Energy condensation sensitivity analysis

Monte Carlo (MC) codes coupled to depletion solvers are increasingly used to provide high fidelity fuel cycle modeling capabilities. Here, these coupled depletion-MC tools produce accurate results in general but can experience nonphysical spatial oscillations when time steps are large or when a system’s dominance ratio approaches unity. Two substepping techniques have been developed previously to remedy and dampen these spatial oscillations without needing to reduce step sizes. The first approach relied on higher-order techniques to account for spectral changes within steps (extrapolation and interpolation techniques). The second approach used the first order perturbation (FOP) theory to account for the change in the one-group spatial flux distribution within steps. This paper develops a hybrid depletion methodology which, in a way, combines how the flux is handled in both substepping techniques. Specifically, the multigroup (MG) MC Shift code is used to update the flux distribution within steps rather than a one-group FOP solver. A fully reflected pincell is investigated, which is not spatially dependent in the MG representation. Thus, the analysis in this paper is an initial demonstration of hybrid depletion. An upcoming companion paper will focus on how the hybrid depletion dampens spatial oscillations. The hybrid depletion approach is verified to be consistent with previous constant extrapolation depletion (CED) methods. This paper finds that the hybrid CED exhibits some error in the eigenvalue and one group constants within macro steps. To address this discrepancy, a simple interpolation scheme (CELI) is investigated. This work found that CELI sufficiently addresses the discrepancy in spectrum for macro steps up to 100 days. Overall, this work demonstrates that the hybrid depletion method can significantly reduce the number of high fidelity MC executions in a MC-coupled depletion with an acceptable eigenvalue error.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2023

Nuclear Criticality Safety Pipeline Course with Hands-On Experimental Training at Lawrence Livermore's Inherently Safe Subcritical Assembly Training Center

The Nuclear Criticality Safety Divisions at Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL) have partnered with Prof. Massimiliano Fratoni of the University of California Berkeley to offer a semester long course on nuclear criticality safety. This course is part of a larger pipeline project among many of the Department of Energy (DOE) laboratories designed to stimulate student interest in the field of criticality safety. The course focuses on teaching the fundamentals of criticality safety, familiarity with national and consensus standards, and preparing criticality safety evaluations. Students also receive hands-on experience with special nuclear material by performing experiments with the Inherently Safe Subcritical Assembly (ISSA) at LLNL. Guest lectures are taught remotely and in-person by criticality safety engineers at LLNL and LANL, giving students an opportunity to interact with professionals in the field. The students complete a semester long project involving developing and writing a criticality safety evaluation. As universities tend to focus heavily on nuclear power and advanced nuclear reactor design, this course gives students a better understanding and perspective of what criticality safety entails. The goal of this pipeline course is to introduce students to criticality safety as another available field for nuclear engineers. It is also a way for criticality safety programs to identify talented students who have the interest and aptitude to work in criticality safety for hire upon graduation. LLNL and LANL have both hosted past students as summer students, participated in student's graduate projects, and hired students as criticality safety engineers. This has provided a unique opportunity for criticality safety programs to spot young talent with better retention outcomes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

The corrosion effects of neutron activation of 2LiF-BeF 2 (FLiBe)

The redox effect of activation reactions in molten salts has been previously qualitatively defined but never quantified. Here, we quantify the corrosion effects of neutron activation of FLiBe. Rates of change of activation products are computed for FHR and ARC. The impact of activation reactions on salt chemistry is quantitatively compared to other environmental effects of similar consequences on the salt. The results are applied to a redox control case-study and an irradiation-corrosion experiment. The results are presented in a manner useful to the reactor chemist and designer concerned with corrosion and chemistry control and tritium management in irradiated FLiBe, and design of salt-irradiation experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Neutronic impact of Autonomous Reactivity Control system inclusion in a breed-and-burn core

This work examines the impacts induced in a large breed-and-burn core through the inclusion of Autonomous Reactivity Control (ARC) systems from the neutronics perspective. A combination of numerical and analytical techniques are used to conservatively quantify the reactivity penalty, amount of additional tritium production, and depletion of the system absorber fluid. The reactivity penalty is found to be negligibly small, with the largest impact resulting from the removal of fuel material to make room for the ARC system tubes. Similarly, the depletion of the absorber fluid is found to be negligible compared to the amount of fluid present. It is found that the production of tritium within the active core is also small in comparison to that which is nominally produced, while tritium production within the lower ARC system reservoir may be of concern if not adequately considered during the assembly design phase. This issue is explored and a mitigation technique is proposed. Ultimately this study concludes that there are no issues from the steady-state neutronics perspective preventing incorporation of ARC systems into typical breed-and-burn cores.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Molten Salt Reactor Experiment Benchmark Evaluation

As one of the six advanced reactor concepts selected by the 2002 Generation IV roadmap as a technology meriting future research, Molten Salt Reactors (MSRs) attracted broad attention and multiple private and public entities are working towards its commercialization. Facing stringent regulatory requirements, validations of computational codes used to calculate and prove the safety characteristics of MSRs play a key role. This project developed the world-first, MSR-related reactor physics benchmark based on a series of zero-power experiments conducted at the Molten Salt Reactor Experiment (MSRE). The benchmark was successfully reviewed by the International Reactor Physics Experiment Evaluation Project (IRPhEP) committee and is available in the IRPhEP handbook starting form the 2019 edition. To benchmark experimental data from the MSRE, a three-dimensional high-fidelity model was developed using the Monte Carlo neutron transport code Serpent 2 and ad-hoc methods have been implemented to account for the unique feature of fuel salt motion. The calculated effective multiplication factor, k eff , for the first criticality experiment, when 235 U was progressively added to the fuel salt in order to achieve criticality with stationary salt and isothermal conditions, was 1.02132 (±3 pcm). The total uncertainty for experimental k eff was estimated to be 420 pcm. The calculated k eff is 2.154% larger than the experimental and benchmark model value, which is approximately five times the benchmark model uncertainty. It is to be noted that, for systems containing large volume of graphite (or other carbonaceous materials), Monte Carlo codes tend to overestimate the keff of the benchmark model by 1% to 2%. The bias is possibly due to uncertainties in the impurity content of the graphite blocks, uncertainty in the neutron capture cross section of carbon or uncertainty in the properties of nuclear-grade graphite. The calculated reactivity coefficient of 235 U concentration (0.2228 ± 0.0014, represented as the change of reactivity over the relative change of 235 U mass in loop) matches well with the experimental value (0.223 ± 0.006), strengthening the confidence on the accurate representation of the fuel salt composition in the MSRE benchmark. Additional reactivity effect calculations, including the control rod bank worth, reactivity effects of fuel circulation and isothermal and fuel temperature coefficients show good agreement with the experimental values (within 1σ) as well.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

A Python Package for Fuel Cycle and Waste Management Analysis - 19545

The future of nuclear power utilization in the United States may involve a transition from the once-through light water reactor fuel cycle to a so-called advanced fuel cycle. An advanced fuel cycle may offer improvements in natural resource utilization, reactor safety, and environmental impact. In addition, adoption of an advanced fuel cycle will result in changes in the amount, composition, and disposal of the nuclear wastes they produce. Characteristics intrinsic to the wastes themselves, such as mass and radioactivity at specific times, were used as waste management metrics in the Department of Energy 2014 Fuel Cycle Evaluation and Screening (FCES) study. Binning of these metrics informed fuel cycle according to relative potential benefit. However, waste properties alone cannot qualify management strategy. Therefore, comparing fuel cycles from the perspective of waste management can provide additional insights. To that end, an object-oriented Python package has been developed to perform fuel cycle waste management calculations. The package, named nwpy (short for Nuclear Waste analysis in Python), mimics the mass flow operations reported in the FCES study and interfaces with the ORIGEN-S code to assess metric data as reported in the FCES. In addition, this Python package determines the loading of waste streams into forms suitable for final disposal in a geological repository and uses such data for repository performance assessment. This work describes the Python package methodology, assumptions, and capabilities. As input data, the package requires information about the fuel cycle such as the mass of fuel discharged, the cooling time between reactor discharge and subsequent operations, and the separations process efficiencies and waste forms it produces. In addition, the isotopic composition of the reactor discharge stream, such as that available from the FCES database, is required. With that, it can perform prescribed mass flow and decay operations throughout the back end of the fuel cycle, the latter carried out by the ORIGEN-S code distributed with the SCALE6.2 software suite. The result is time-dependent data characterizing the waste on a per-package basis. Package loading for waste streams depends on the stream form (i.e. UNF vs. HLW, or glass HLW from aqueous reprocessing vs. ceramic HLW from electrochemical processing) and is based on literature review and engineering assumption. As an example of the application of the code, calculations for the areal footprint of direct-contact repositories for different fuel cycles is shown. Development is ongoing, and the package is expected to grow incorporating capabilities for waste management analyses such as waste transportation and radionuclide transport through the biosphere. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Assessment of the potential for criticality in the far field of a used nuclear fuel repository

This study aims to assess the likelihood for criticality in the far field of a repository for direct disposal of commercial light water reactor used nuclear fuel. Two models are used in combination: (1) a neutronics model to estimate the minimum critical masses of spherical, water-saturated depositions of fissile material; (2) a transport model to simulate the dissolution of waste packages arranged in an array and the subsequent transport of fissile solutes through fractured bedrock to a single accumulation location. The neutronics model shows that heavy metals from different types of used fuel present the same minimum critical mass behavior in the parameter space of initial enrichment and burnup, dictated largely by the fissile content. However, the magnitude of the minimum critical mass varies significantly within that parameter space, and secondary effects like the presence of absorbing nuclides play a minor role. Here, the transport model employs various subsurface transport scenarios, and for each scenario the mass of each isotope and the overall fissile content of the accumulation is reported from the time of canister failure up to one hundred million years at various distances from the repository edge. Taking the results of the two models in concert, it is shown that even if the accumulation of a critical mass is possible under conservative conditions, these conditions are unlikely to be present in the vicinity of a carefully engineered repository. Based on the results of each model, recommendations for risk mitigation in terms of waste characteristics and repository design are given.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Integrated FHR technology development: Tritium management, materials testing, salt chemistry control, thermal hydraulics and neutronics, associated benchmarking and commercial basis

This report summarizes the results of a 3-year Integrated Research Project (IRP) sponsored by the U.S. Department of Energy to address major technical challenges in the development of the Fluoride-salt-cooled High-temperature Reactor (FHR). The IRP universities included the Massachusetts Institute of Technology (MIT), the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The FHR is a new reactor concept less than 20 years old that combines (1) a clean fluoride salt coolant, (2) the graphite-matrix coated-particle fuel originally developed for High- Temperature Gas-Cooled Reactors and (3) passive decay heat removal systems from sodium fast reactors. The base-line design is a pebble-bed FHR that delivers heat between 600 to 700°C with a 7 Li 2 BeF 4 (flibe) salt coolant to the power cycle or industrial users. The Executive Summary describes what was accomplished and summarizes major conclusions. The main report contains stand-alone sections in each major area of work. Highlights are summarized below. Tritium. The coolant salt produces tritium necessitating understanding and control of tritium. Experimental work included measuring hydrogen uptake on carbon, tritium uptake during irradiation of graphite in 700°C salt and the permeability of tritium through metals with and without coatings. An FHR systems model was developed to predict tritium behavior and enable design of tritium control systems. The leading candidate for tritium removal from the flibe salt is carbon bed outside the reactor core. Thermal hydraulics and Neutronics. A 10-meter high FHR integral effects test facility, using Dowtherm A®, has been built and used to simulate the integral response of a representative FHR to improve understanding of FHR thermal hydraulics and validate thermal hydraulics codes used for the FHR. Neutronic analysis included code-to-code benchmarking, multiphysics modeling, uncertainty analysis for nuclear data uncertainties and shielding analysis. Twisted-tube heat exchangers were examined that may provide significant cost reductions because of their improved heat-transfer performance with liquid salts. The impacts on safety and design caused by uncertainties in physical properties of the salt were assessed with recommendations on where added physical property measurements are needed. Work began on understanding radiative heat transfer that becomes important in salt systems at higher temperatures. Materials. Static corrosion tests were conducted on graphite and materials in 700°C salt -- both in the UW laboratory and in the MIT reactor. In addition to static corrosion testing, a circulating salt corrosion loop was built and is operating. Methods to measure salt properties including redox were developed. The presence of carbon increases corrosion. With good redox control, the evidence indicates stainless steel can be used as a material of construction. Benchmarking. An additional series of workshops where held to enable benchmarking and comparison of results between different investigators. Commercialization. The FHR delivers higher average-temperature (~650°C) heat to the power cycle or industry than other reactors. This enables power cycles with lower-cost heat storage for variable electricity to the grid while the reactor operates at base-load -- providing a major competitive advantage by maximizing revenue from electricity sales. In October, 2016, a new startup company named Kairos Power was formed, and opened its first office in Jack London Square, Oakland, California, in January 2017. Kairos Power exists largely due to the work that was accomplished by this IRP and the previous DOE-funded FHR IRP that performed research on FHR technology from 2012 to 2015. In September, 2018, as this IRP is concluding, Kairos Power was staffed at 60 full time employees, working to develop technology derived from this IRP. Many of these employees worked with the IRP before joining Kairos Power 1 (https://kairospower.com). Other students and researchers from this IRP have joined other startup companies, including the Terrapower Molten Chloride Fast Reactor and the Terrestrial Energy development efforts.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

The Criticality Safety Studies of Joonhong Ahn

The recent passing of UC Berkeley professor Joonhong Ahn at the age of 57 has been a dear loss to the community of nuclear scientists at large. His contributions to studies on advanced fuel cycle development and the societal and ethical aspects of nuclear technology were known around the world and fostered a great reputation. He produced at least fifty peer-reviewed journal papers, authored or contributed to ten books, and organized a considerable number of international meetings. Ahn's studies led to his eminence in the area of nuclear waste management, where he focused on repository performance assessments along with their impact on society. These assessments included the development of analytical models for nuclide transport phenomena and evaluation of the possibility of criticality events imparting a dose to the biosphere. This paper will summarize his achievements in criticality safety along with developments before his passing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Feasibility of a Breed-and-Burn Molten Salt Reactor

This work addresses the feasibility and characteristics of the breed and burn (B and B) cycle for Molten Salt Reactors. In the general B and B cycle, natural Uranium or Thorium is used as fuel for a reactor which breeds a sufficient amount to maintain criticality at equilibrium without reprocessing. Interest in applying this scheme to an MSR has developed recently For the MSR B and B scheme considered here, fresh salt containing fertile material is continually fed into the core while salt with the composition of the homogeneous fuel is expelled without reprocessing. Through an analysis of the equilibrium conditions and startup requirements, the feasibility of such a cycle operating with various salts is assessed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗