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Corradini, Michael

Publications and source records attributed to Corradini, Michael.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Progress in Fast Modular Reactor Conceptual Design

The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO 2 ) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA®) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. Further, the reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of–forced cooling accident, which showed the peak cladding temperature at ~1600°C during the transient, which is below the current design limit of 1800°C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Fuel performance analysis of Cr-coated Zircaloy-4 cladding during a prototypical LOCA event using BISON

Here, deformation and failure of chromium (Cr) coated Zircaloy-4 (Zry-4) were studied in loss-of-coolant accident (LOCA) conditions using the BISON fuel performance code. The BISON validation model simulating Halden research reactor experiments was extended to include Cr coatings and higher rod internal pressures to simulate high-burnup fuel. The transient simulations show Cr coatings help relieve stress in the Zry-4 substrate during the transient, delaying the onset of high-temperature creep, which leads to ballooning and bursting of the cladding. A nominal Cr coating thickness of 30 µm delays clad failure by 26 seconds and increases the clad burst temperature by 40 K. A parametric study showed that time to failure and burst temperature both increase with coating thickness, and Cr-coated cladding offers burst resistance under a wide range of rod internal pressures simulating high-burnup fuel. Results indicate that a thin Cr coating provides resistance against ballooning and bursting of cladding during LOCA events.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

MELCOR Model Development of the General Atomics Fast Modular Reactor

General Atomics Electromagnetic Systems (GA-EMS) is currently developing a fast modular reactor (FMR) capable of providing safe, carbon-free electricity by 2035. As part of the development process, researchers at the University of Wisconsin-Madison (UWM) use MELCOR 2.2 to develop a simulation model of the proposed design to perform accident analyses. This paper summarizes the efforts of researchers at UWM in the creation of the FMR model. This model was built based on initial FMR design parameters. The design parameters will change as the design evolves and is refined.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Pool boiling critical heat flux studies of accident tolerant fuel cladding materials

Pool boiling experiments have been performed at atmospheric pressure on potential accident tolerant fuel (ATF) cladding materials (Zirlo®, Cr coated Zirlo®, FeCrAl coated Zirlo®, and monolithic SiC) to evaluate the Critical Heat Flux (CHF) points. Corrosion resistant coatings of Cr and FeCrAl alloy were deposited on flat Zirlo® samples using cold spray technology. The as-prepared samples after uniform surface polishing were subjected to autoclave tests at 360 °C water and 18.6 MPa for 360 h to simulate prototypic corrosion of the ATF cladding materials in LWR normal operation. Surface characteristics potentially influencing CHF such as surface morphology, roughness, static contact angle, and surface chemistry were characterized using a suite of characterization methods including scanning electron microscopy, 3D optical profilometry, optical contact angle measurements, and x-ray photoelectron spectroscopy (XPS). Thermo-physical properties of the samples such as density, thermal conductivity, and heat capacity were measured by differential scanning calorimetry and laser flash thermal diffusivity measurement. FeCrAl coatings showed CHF values comparable to bare Zirlo® samples, while slightly lower CHF values were observed for Cr coated samples and bulk SiC flats (produced by chemical vapor deposition, CVD). As-deposited Cr coatings showed 67% higher CHF than the as-polished Cr coatings due to its higher surface roughness level. The autoclave testing for the prepared samples generally increased the CHF except for the surface polished Cr coatings. Similar CHF values of bare Zirlo®, FeCrAl coating, and CVD SiC samples were observed after autoclave test but there were negligible change CHF of Cr coatings. XPS studies indicated that the formation or deposition of a few monolayers of hydrophobic carbonaceous species on surface can affect CHF values. The trends in CHF data are discussed in terms of evolution of the surface characteristics of the ATF materials.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

An Experimental Study of Design and Performance for the Water-Based Reactor Cavity Cooling System

The document summarizes the test activity conducted at the Texas A&M University water-cooled Reactor Cavity Cooling System experimental facility. Flow and temperature characterization of the risers' panel is conducted using advanced measurement techniques. Steady-state and transient scenarios are studied, analyzed, and discussed in the report. The experimental work is complemented with modeling and simulations using system level codes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Evaluation of critical heat flux of ATF candidate coating materials in pool boiling

The critical heat flux (CHF) of potential accident tolerant fuel (ATF) cladding materials and concepts, Cr and FeCrAl coatings on zirconium-alloy substrates, was investigated experimentally. A cold spray process, which involves the propulsion of high velocity powder particles on to the substrate surface to form a coating upon impact, was used to create the coatings. Surface parameters such as contact angle, surface roughness, and morphology that are associated with CHF in pool boiling were characterized. Non-hydrodynamic effects of newly developed ATF candidate coating materials on CHF were also evaluated. The effects of thermal properties and coating thickness and expected challenges for evaluation of CHF for new ATF cladding materials development are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Model validation using CFD-grade experimental database for NGNP Reactor Cavity Cooling Systems with water and air

This project has been focused on the experimental and numerical investigations of the water-cooled and air-cooled Reactor Cavity Cooling System (RCCS) designs. At this aim, we have leveraged an existing experimental facility at the University of Wisconsin-Madison (UW), and we have designed and built a separate effect test facility at the University of Michigan. The experimental facility at UW has underwent several upgrades, including the installation of advanced instrumentation (i.e. wire-mesh sensors) built at the University of Michigan. These provides highresolution time-resolved measurements of the void-fraction distribution in the risers of the water-cooled RCCS facility. A phenomenological model has been developed to assess the water cooled RCCS system stability and determine the root cause behind the oscillatory behavior that occurs under normal two-phase operation. Testing under various perturbations to the water-cooled RCCS facility have resulted in changes in the stability of the integral system. In particular, the effects on stability of inlet orifices, water tank volume have and system pressure been investigated. MELCOR was used as a predictive tool when performing inlet orificing tests and was able to capture the Density Wave Oscillations (DWOs) that occurred upon reaching saturation in the risers. The experimental and numerical results have then been used to provide RCCS design recommendations. The experimental facility built at the University of Michigan was aimed at the investigation of mixing in the upper plenum of the air-cooled RCCS design. The facility has been equipped with state-of-theart high-resolution instrumentation to achieve so-called CFD grade experiments, that can be used for the validation of Computational Fluid Dynanmics (CFD) models, both RANS (Reynold-Averaged) and LES (Large Eddy Simulations). The effect of risers penetration in the upper plenum has been investigated as well.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Multi-Functional Zirconium-Silicide Coatings on Zirconium-alloy for Improved Accident Tolerance

Recently, development of protective coatings on zirconium-alloy (Zr-alloy) fuel cladding in light water reactors has been extensively investigated to enhance safety margins in accident scenarios. One accident scenario is degradation of the reactor core under a coolant-limited condition, where a large amount of Zr-alloy claddings would react with high temperature steam, leading to oxidation, hydrogen production, and loss of intrinsic mechanical properties of the fuel cladding. Although, total replacement of Zr-alloy with SiC-SiCf composites or FeCrAl is being considered for the long-term, application of oxidation resistant coatings on zirconium-alloy is being investigated as a near-term solution to mitigate oxidation of Zr-alloy and increase the accident coping time. Coatings, such as chromium, Ti{sub 2}AlC, SiC, and Cr{sub 3}C{sub 2}- NiCr, deposited by various deposition techniques have been investigated. In addition, recent studies on modified heating surface for heat transfer enhancement provides basis on the optimal design of coated Zr-alloy claddings so that the new cladding can be achieve accident tolerance and heat transfer efficiency for both accident and normal operation conditions. Therefore, boiling heat transfer from the coated surface to water environment must be evaluated in parallel. In heat transfer performance, boiling heat transfer is characterized by nucleate boiling, critical heat flux (CHF), transition boiling, minimum film boiling temperature (T{sub min}), and film boiling. Zirconium-silicide (ZrSi{sub 2}) is a potentially attractive protective coating material for cladding applications. Although this material has not been studied extensively, it appears to be a promising candidate because of the high melting points, and the low neutron penalty, and the generally good oxidation resistance to transition metal silicides. It has been reported that oxidation of ZrSi{sub 2} powder in a flowing air environment at elevated temperatures produced ZrO{sub 2}, SiO{sub 2}, and ZrSiO{sub 4}, pointing to the possibility of achieving highly protective surface layers of SiO{sub 2} and ZrSiO{sub 4} that would protect the material from further oxidation. In this study, magnetron sputter deposition of ZrSi{sub 2} on zirconium-alloy flats and cylindrical rods was performed. The coated samples were oxidized in high temperature air (700 deg. C, 1000 deg. C and 1200 deg. C) to evaluate their oxidation resistance. The coated cylindrical rods were quenched in saturated water to observe transition boiling, T{sub min}, film boiling regime, while the coated flats were tested in pool boiling facility to investigate nucleation boiling and CHF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Orificed Flow Results for a Two-Phase Natural Circulation Loop

Researchers at University of Wisconsin - Madison have constructed a 1/4 scaled experimental facility to study the operation of a two-phase natural circulation decay heat removal system. The Reactor Cavity Cooling System (RCCS) is a decay heat removal safety system designed for high temperature generation IV reactors. The system removes decay heat by encircling the outside of the walls of the pressure/containment vessel with cooling channels. These channels are filled with either air in an open loop configuration or water in a closed loop configuration. Radiative and convective heat loads from the pressure vessel induce a buoyancy driven flow in the risers. In accident conditions, the water-based system reaches saturation and the resulting steam is vented off further removing heat until the water inventory is depleted. The water based system has been under investigation at University of Wisconsin - Madison for variations in performance under various flow restriction cases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Fuel-Coolant Interaction Results for High Energy In-Pile LWR Fuels Experiments

Fuel-coolant interaction (FCI) has long been a subject of interest in severe-accident reactor safety applications. Significant chemical interaction also frequently results at the high temperature associated with these events in a water environment. Historical rapid, high energy transient irradiation test results are reviewed for experiments performed in transient experimental reactors in Idaho. Though such events are not credible for any deployed nuclear reactor technology, the pressure waves produced by FCI on surrounding containment structures are of importance to advanced light water reactor designs and for transient experiment design considerations. A brief introduction to steam explosion and related studies is provided in the context of these experiments. Extensive reactivity initiated accident testing programs were carried out in the Transient REActor Test (TREAT) facility, the Special Power Excursion Reactor Test (SPERT) facility, and the Power Burst Facility (PBF) from the 1960’s through the late 1970’s. A variety of fuel forms including waterlogged rods were tested in capsules containing water coolant at standard temperature pressure. A few tests were performed at high pressure and temperature in the water loop of PBF. In many of these experiments, energy deposition was high enough to greatly exceed melting temperatures of the fuel, resulting in some fraction of the fuel being dispersed into the coolant. A purpose of many of these tests was to determine the nuclear-to-mechanical energy conversion ratio as well as the extent of metal-water reaction. Relevant data collected from these tests included transient energy deposition, pressure response of the coolant, water column velocity, total hydrogen generation, and resulting particle size distributions and particle characterization. In this paper, the mechanical configuration of each experiment facility is reviewed and discussed relative to the measured responses. The empirical results for source pressure and extent of metal-water reaction are presented in the context of the experimental conditions. The results of these experiments provide a unique understanding for behavior of fuel under severe accident conditions in a reactor environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Reactor Safety for Beyond Design Basis Events

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant's current design basis. The DOE has played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R and D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating as well as better characterize the safety performance of future U.S. nuclear power plants. In pursuing this area of R and D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary end user of the results from this DOE-sponsored work. The purpose of the Reactor Safety Technology R and D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet. The RST R and D activities have evolved from an initial coordinated international effort to assist in the analysis of the Fukushima accident progression and accident response into the following three areas of current work. Fukushima Forensics and Examination Plans: This R and D is focused on providing insights into the actual severe accident progression at Fukushima through planning and interpretation of visual examinations and data collection of in-situ conditions of the damaged units as well as collection and analysis of samples from the damaged reactors. Severe Accident Analyses: This R and D is focused on analyses using existing computer models and their ability to provide information and insights into severe accident progression that aid in the development of severe accident management guidelines (SAMG) and/or training operators on these SAMGs; an auxiliary benefit can be informing improvements in these models. Accident Tolerant Components: This R and D work is focused on analysis or experimental efforts for hardware-related issues, including systems, structures and components with the potential to prevent core degradation or mitigate the effects of beyond-design basis events. This presentation will review current and future R and D activities in this reactor safety area. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Water-cooled Reactor Cavity Cooling System Flow Analysis

A research team at University of Wisconsin - Madison designed and constructed a 1/4 scaled experimental facility to study natural circulation cooling in a water based reactor cavity cooling system (WRCCS) for decay heat removal in an advanced high temperature reactor. The facility is capable of natural circulation operation scaled for simulated decay heat removal (up to 45 kW input power, which is equivalent to 1.70 MW at full scale) and pressurized up to 2 bar pressure. The WRCCS facility has been used to study natural circulation flow behavior from transients due to evaporation of the water inventory from decay heat removal. The natural circulation flow is observed to be oscillatory with the mean flow increasing as the pressure head in the system drops due to evaporation of the water inventory during two-phase operation. A first principles model was developed to understand the behavioral limits and trends. Through the use of void fraction measurements and the model, we find that the mass flow rate in the system increases due to small increases in the time-averaged void fraction over the course of a four-hour boil-off experiment. This increase in void fraction causes an increase in the buoyancy head and mass flow. The model also determines the operational limits on the mass flow rate under two-phase operation that can be verified with extended tests in the WRCCS facility. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Development of Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

There is considerable interest in the development of coatings for zirconium-alloy fuel cladding used in light water reactors (LWR) to improve their oxidation resistance during high temperature accident conditions and to enhance performance under normal operating conditions. These coatings should also enhance the heat transfer characteristics of the cladding surface so that heat from the fuel rod is efficiently transferred into the water coolant. One potential candidate coating material is zirconium-silicide. Transition metal silicides have been generally shown to have outstanding oxidation resistance. Zirconium-silicide is particularly attractive because Zr is strongly oxophilic, so that the Zr-silicide layer is expected to bond more strongly to the inner, native ZrO{sub 2} layer. The relative proximity of the free energy of formation of oxides of Zr and Si can lead to alternating multilayered oxide structures with radiation damage resistance. Additionally, oxidation of zirconium silicide can produce excellent corrosion resistant glassy zircon (ZrSiO{sub 4}) and silica (SiO{sub 2}), possessing self-healing qualities. This paper reports results of our results of high temperature oxidation tests of bulk zirconium-silicide and coatings of zirconium-silicide deposited on zirconium-alloy test flats and cladding tube configurations using plasma vapor deposition (PVD) process. We also report our initial results on the experimental facility for studying boiling characteristics of coated cladding tubes in prototypic pressurized water conditions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗