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DOE OSTI2022

BISON Validation to In-Situ Cladding Burst Test and High Burnup LOCA Experiments

The process to develop and qualify nuclear fuels for commercial nuclear application requires fundamental material development, characterization, and design; out-of-pile testing on unirradiated materials; integral fuel rod irradiations, testing, and post irradiation examinations (PIEs); and transient analyses. The historical approach depends on the generation of large empirical datasets and series of integral fuel rod irradiations, and this approach ultimately takes approximately 20 years—or sometimes longer—to acquire data through extensive sequential testing. Thus, the qualification and eventual deployment of new fuel systems constitute a long and drawn-out process. However, recent technological advancements have provided researchers the opportunity to perform out-of-cell, in-situ measurements to assess material performance for the duration of the experiment. One such example of this capability is the use of digital image coordination and thermal imaging to assess Zircaloy cladding performance under a simulated loss-of-coolant accident (LOCA) transient condition. In general, the in-situ measurements provide high-fidelity strain, strain rates, and temperature surface maps. This is critical for the United States nuclear industry, which is actively developing a technical basis to support extending the peak rod average burnup from 62 to ~75 GWd/tU and the deployment of accident-tolerant fuel. However, the Nuclear Regulatory Commission, through their research information letter, outlined a number of technical issues for the industry to address before extending burnup. One topic of specific interest is understanding the cladding balloon and rupture geometry during the LOCA heatup phase. Leveraging these advanced in-situ capabilities, this work intends to use insitu data generated from a simulated LOCA in the Severe Accident Test Station at Oak Ridge National Laboratory to better understand high-temperature creep and its impact on Zircaloy balloon and rupture performance. This work used the BISON fuel performance code to compare BISON’s high-temperature creep model predictions to in-situ data and identify limitations and gaps within the model. Additionally, BISON will subsequently be used to simulate relevant high-burnup LOCA tests. The results will be analyzed and compared to the available post-test data in a manner consistent with the approach outlined in the Nuclear Regulatory Commission research information letter.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

Hu, Jianwei↗

DOE OSTI2019

Graphite Moderation in Fast Reactors

With the renewed interest in designing and building a fast test reactor in the U.S., several new reactor concepts have been emerging. Although the main mission of a fast test reactor would be to enable high fast flux irradiation capability, there could be added value for such a fast test reactor to also offer thermal neutron irradiation capabilities. As such, several of these concepts considered the use of a moderator material in a Sodium-cooled Fast Reactor (SFR), which is claimed to also offer benefits with respect of mass of fuel required. The work discussed in this paper focuses on exploring and assessing the benefits and challenges resulting from making use of thermalized neutrons to reduce the critical fuel mass. The analyses are performed for a graphite moderating material, using a simple core model to easily identify the phenomena observed. It is found that it is necessary to reduce the fuel enrichment near the moderated regions in order to avoid unacceptable power peaking. In the end, use of moderating materials outside a SFR core does not provide significant benefit in term of fuel volume reduction, but allows for a noticeable reduction of the mass of fissile material needed. However, it also implies a number of additional challenges which are highlighted in this paper.

core optimization↗

DOE OSTI2018

Economic viability of multiple algal biorefining pathways and the impact of public policies

This study presents an extensive systems-level multi-pathway sustainability assessment of algae biofuel production that demonstrates the necessity of high-value co-products, examines the impact of public policy scenarios, and identifies improvements and pathway directions required for economic viability. Engineering process models for several fuel and co-product pathways were leveraged to perform high fidelity techno-economic analysis. These pathways included: baseline hydrothermal liquefaction; protein extraction followed by hydro-thermal liquefaction; fractionation into high-value chemicals with fermentation followed by hydrothermal liquefaction for fuels; and a small-scale first-of-a-kind plant coupled with a wastewater treatment facility. From these models, it was shown that hydrothermal liquefaction as a fuel-only pathway is not economically viable. Likewise, the benefits of coupling with wastewater water treatment are insignificant compared to the impact ofreduced facility size resulting in increased capital costs. These models were also used to examine public policy scenarios, uniquely presenting their impact on the breakeven cost of fuel production and sensitivity to scenario assumptions. Specifically, depreciation type was shown to be irrelevant for write offs faster than 10 years. Due to discounting, short-term subsidies were found to capture 50% of the subsidy value in 6 years with an additional24 years required for full subsidy valuation. Integration of a carbon economy was shown to decrease biofuel production costs, particularly for the protein pathway due to the co-product accounting. Finally, a metric of normalized costs was used to compare algal biorefineries to corn and cellulosic ethanol production, showing that algal systems are uniquely different due to significantly higher capital costs, though operational costs are comparable. Overall, this work demonstrates that, to reach economic viability, algal biofuel production must either utilize higher value non-fuel co-products or achieve drastic reductions in capital costs.

09 BIOMASS FUELS↗

DOE OSTI2015

Casting Process Modeling and Validation

Computational simulations of gravity casting for metallic U-Pu-Zr nuclear fuel rods were performed with the free surface flow solver Flow3d. Two different casting apparatus designs were explored for studying the effects on the final cast rod profiles. Experimental data for the casting problem obtained from Idaho National Laboratory (INL) was validated with numerical simulations on one of the designs considered. The volume of the rods cast, length of the rods, and temperature profiles were predicted through the transient casting process and were found to match closely with experimental data. A sensitivity analysis with vent locations, pressure, mesh refinement, and effects of surface tension at the interface were explored. This work serves as a validation benchmark for the casting problem and provides guidelines when performing simulation exercises for modeling casting. This paper also provides the mathematical framework to enabling modeling of the casting process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Mathematical model of marine diesel engine simulator for a new methodology of self propulsion tests

As a vessel operates in the open seas, a marine diesel engine simulator whose engine rotation is controlled to transmit through propeller shaft is a new methodology for the self propulsion tests to track the fuel saving in a real time. Considering the circumstance, this paper presents the real time of marine diesel engine simulator system to track the real performance of a ship through a computer-simulated model. A mathematical model of marine diesel engine and the propeller are used in the simulation to estimate fuel rate, engine rotating speed, thrust and torque of the propeller thus achieve the target vessel’s speed. The input and output are a real time control system of fuel saving rate and propeller rotating speed representing the marine diesel engine characteristics. The self-propulsion tests in calm waters were conducted using a vessel model to validate the marine diesel engine simulator. The simulator then was used to evaluate the fuel saving by employing a new mathematical model of turbochargers for the marine diesel engine simulator. The control system developed will be beneficial for users as to analyze different condition of vessel’s speed to obtain better characteristics and hence optimize the fuel saving rate.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2024

Innovative Natural Gas Technologies For Efficiency Gain In Reliable And Affordable Thermochemical Electricity Generation (INTEGRATE)

Hyper is a cutting-edge system that pairs computational models that simulate the response of different systems with physical components such as sensors and actuators to transfer said calculated response to the HyPer facility. Computational models must operate in real-time in order to resolve the system response when coupled to physical hardware. Real-time is defined by the minimal HyPer system response time of Δt = 0.080 s. The integration of these models serves as a foundation for performance characterization based on real-world SOFC hardware and paves the way for comprehensive fuel flexibility studies. This report underscores the innovative approach to cyber-physical system deployment at NETL’s HyPer Facility, highlighting the critical role of computational models in enhancing the performance and adaptability of energy technologies.

03 NATURAL GAS↗

DOE OSTI2013

Study of charged particle motion in fields of different configurations for developing the concept of plasma separation of spent nuclear fuel

The concept of plasma separation of spent nuclear fuel in a plane perpendicular to the magnetic field in an electric potential of special configuration is developed. A specific feature of the proposed approach consists in using an accelerating potential for reducing energy and angular spread of plasma ions at the entrance to the separator chamber and a potential well for the spatial separation of ions with different masses. The trajectories of ions of the substance imitating spent nuclear fuel are calculated. The calculations are performed for azimuthal and axial magnetic fields and model electric field configurations corresponding to different geometries of the separator chamber. It is shown that, using magnetic fields with a characteristic strength of 1 kG and electric potentials of up to 1 kV inside a region with a linear size less than 100 cm, it is possible to separate ions of spent nuclear fuel with energies from 0.2 to 3 eV. The calculations were performed for a collisionless mode in the single-particle approximation. Possible variants of the experimental facility for plasma separation of spent nuclear fuel are proposed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2017

Dynamic Analysis of the Next Generation Molten-Salt Breeder Reactor System

The conceptual design of a 1000 MWe two-fluid molten-salt breeder reactor (MSBR) station was performed during 1966-1970 at Oak Ridge National Laboratory (ORNL) with the goal of using the U-233 fuel cycle. This modular design with four reactor modules has an electrical output of 250 MW/module. The abundance of Th-232 was attractive in converting it to U-233 in a thermal breeder reactor. The two-fluid reactor is graphite-moderated 'with a {sup 7}LiF-BeF{sub 2}-UF{sub 4} fuel salt circulated through the core and a {sup 7}LiF-ThF{sub 4}-BeF{sub 2} blanket salt circulated through separate flow channels distributed throughout the core, as well as in a surrounding under-moderated region'. This reactor design has the advantages of not producing Pu isotopes and easy chemical separation of U-233 in the blanket material. The low fuel cost and the graphite replacement cost are attractive for such a design, with an estimated cost of $141/kW at 1968-dollar value. Potential non-proliferation features of MSBR, the low costs of raw materials, and established chemical separation of fission products are important factors in further development of MSBRs as an alternative to current light water reactors. These reactors are often referred to as liquid fluoride thorium reactor (LFTR). The current research on the modeling of the MSBR is based on an earlier work on the preliminary dynamic model of the MSBR concept. Revisions are made to the original model. The physics parameters have been modified for the current modeling effort based on additional information including more recent values of model parameters from and several typographical misprints in equations and values. Details of this effort will be presented in a full follow-up paper. This work presents the detailed nodal models of reactor fluid core, blanket fluid loop, graphite moderator, fuel fluid heat exchanger, and the blanket fluid heat exchanger. A set of simulations has been performed to validate the model using the commercial software MATLAB{sup TM}-Simulink. Simulations were performed for the cases of source perturbation and reactivity perturbation. The model is inherently nonlinear. A description of the MSBR conceptual design, along with a tabulation of relevant system parameters, is presented. The results of simulation demonstrate stable behavior of the MSBR dynamics, temperature feedback effects, and load-following capability of the MSBR system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Mesoscale simulations of High Burnup Fuel Fragmentation: Applying phase-field modeling to understand high burnup fuel fragmentation during loss of coolant accident conditions

During a loss-of-coolant accident (LOCA) in a nuclear power plant, a rapid increase in temperature is experienced in the fuel. In commercial light water reactors (LWRs), the high burnup structure (HBS) forms in regions of the UO 2 fuel where local burnup is high, characterized by smaller grain sizes and large, overpressurized bubbles. During a LOCA transient, fuel in the HBS region is susceptible to fine fragmentation, where the fuel breaks up into micron-size fragments (sometimes referred to as pulverization). To better understand the mechanisms behind this phenomenon, mesoscale computational modeling using the Idaho National Laboratory code Marmot has been employed to simulate the process of HBS formation and response to a LOCA transient. The process of complete and partial HBS formation was demonstrated in prototypical LWR conditions. The simulated HBS microstructures from these simulations were passed to a phase-field model of fracture. To provide a more direct comparison with experiments, a set of experiments conducted by Studsvik on fine fragmentation/pulverization during LOCA conditions was simulated using the BISON fuel performance code. The temperature profiles from the BISON simulations were passed to a phase-field model to determine the gas bubble pressure as a function of time. These pressure histories were used in a phase-field model of fracture to determine under what conditions and in which types of bubbles fine fragmentation was likely to occur.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1987

Computer-based phosphoric acid fuel cell analytical tools Descriptions and usages

Simulation models have been developed for the prediction of phosphoric acid fuel cell (PAFC) powerplant system performance under both transient and steady operation conditions, as well as for the design of component configurations and for optimal systems synthesis. These models, which are presently computer-implemented, are an engineering and a system model; the former being solved by the finite difference method to determine the balances and properties of different sections, and the latter using thermodynamic balances to set up algebraic equations that yield physical and chemical properties of the stream for one operating condition.

Lu, C.↗

DOE OSTI2019

Evaluation of Aluminum-clad Spent Nuclear Fuel during Drying and Dry Storage

The DOE SNF Working Group ASNF Subgroup and the U.S. Nuclear Waste Technical Review Board have identified knowledge gaps regarding the management of aluminum-clad spent nuclear fuel (ASNF).[1, 2] The objectives of this research are to observe, sample and analyze the nature of aluminum surface corrosion and corrosion layer chemistry relevant to the ASNF performance over the duration of the dry storage segment of the fuel life cycle. The primary goal is to build an understanding of ASNF behavior, from reactor service and wet storage through drying and dry storage, with parallel testing of surrogate materials. The premise is that residual water affiliated with the aluminum oxide from corrosion during reactor service and wet storage, continues to influence ASNF performance during dry storage, and that even after successful drying the surface conditions of ASNF have yet to be satisfactorily characterized. The initial scope collected and analyzed specimens from irradiated non-fuel components of aluminum-clad Advanced Test Reactor (ATR) fuel elements that are routinely removed at the end of reactor service life. Techniques have been developed for remote material sampling from ATR elements from vented dry storage at the Irradiated Fuel Storage Facility (IFSF). These techniques involve 1) nipping off a small corner of the aluminum near where the end box had been removed and 2) scraping the exterior (non-fueled) plate surface to collect an oxide sample. This second effort targets ATR elements expected to have seen the longest residence in storage, both in the CPP-603 pool (with poor water quality) and in dry storage at IFSF, within the fuel handling scheduled to consolidate ATR at IFSF and receive the remaining inventory at CPP-666. Follow on work has been proposed to nip corners and collect scrapings immediately following ATR fuel drying, before placement in dry storage. Sample analysis will use multiple techniques to understand the chemical composition and morphology of the surface oxides. A simultaneous effort is progressing to develop surrogate materials to enable rigorous controlled experiments. Consistently produced, these surrogate materials can then be used to demonstrate limitations inherent to remote sampling and to validate performance models under relevant constraints. The combined efforts will substantiate ASNF performance and fuel management decisions regarding drying and packaging over a more complete range of life cycle conditions. [1] DOE Spent Nuclear Fuel Working Group, Aluminum-Clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 Years) Dry Storage, DOE/ID RPT#: 1575, prepared by Aluminum-Clad SNF Sub Working Group, June 2017. [2] U.S. Nuclear Waste Technical Review Board, Management and Disposal of U.S. Department of Energy Spent Nuclear Fuel, A Report to the United States Congress and the Secretary of Energy, December 2017, p. 143.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Safety Analysis of the Mo-99 Production Upgrade to the University of Missouri Research Reactor (MURR) with Highly-Enriched and Low-Enriched Uranium Fuel

The University of Missouri has been working in conjunction with Argonne National Laboratory (ANL) in the National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program to support conversion of the University of Missouri- Columbia Research Reactor (MURR®) from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. MURR is one of five U.S. High Performance Research Reactors (USHPRR) plus a critical facility that plans to convert to the use of LEU fuel. The purpose of this report is to document the safety margins for both HEU and LEU cores with the Mo-99 production upgrade proposed, designated as the 2017-RBM-99 device in this report. In addition to the 2017-RBM-99 device, the neutronics models also include descriptions of the experimental facilities and experiment loading that are currently typical for MURR and which are different from the experiments that were modeled in the analysis performed for the PSAR. Lastly, there have been changes to the LEU fuel element design with regard to the coolant channel thickness tolerances for the innermost and outermost coolant channels based on feedback from the fuel fabricator. This design change did not change the nominal dimensions of the coolant channels/fuel plates or fuel loading of the fuel plates in the element. Thus, for the purpose of the neutronics analysis for the 2017-RBM-99 upgrade, which is based on nominal dimensions and fuel loadings, the components of the LEU element are modeled the same as in the PSAR analysis. However, the design tolerance changes were incorporated in the calculations of the steady-state safety margins for both HEU and LEU. While the use of fewer significant figures has simplified fabrication, yet there were no negative impacts, including on safety margins, since the end channels were not limiting locations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Mitigation of Syngas Cooler Plugging and Fouling

This Final Report summarizes research performed to develop a technology to mitigate the plugging and fouling that occurs in the syngas cooler used in many Integrated Gasification Combined Cycle (IGCC) plants. The syngas cooler is a firetube heat exchanger located downstream of the gasifier. It offers high thermal efficiency, but its’ reliability has generally been lower than other process equipment in the gasification island. The buildup of ash deposits that form on the fireside surfaces in the syngas cooler (i.e., fouling) lead to reduced equipment life and increased maintenance costs. Our approach to address this problem is that fouling of the syngas cooler cannot be eliminated, but it can be better managed. The research program was funded by DOE using two budget periods: Budget Period 1 (BP1) and Budget Period 2 (BP2). The project used a combination of laboratory scale experiments, analysis of syngas cooler deposits, modeling and guidance from industry to develop a better understanding of fouling mechanisms and to develop and evaluate strategies to mitigate syngas cooler fouling and thereby improve syngas cooler performance. The work effort in BP 1 and BP 2 focused on developing a better understanding of the mechanisms that lead to syngas cooler plugging and fouling and investigating promising concepts to mitigate syngas cooler plugging and fouling. The work effort focused on the following: • analysis of syngas cooler deposits and fuels provided by an IGCC plant collaborating with this project; • performing Jet cleaning tests in the University of Utah Laminar Entrained Flow Reactor to determine the bond strength between an ash deposit to a metal plate, as well as implementing planned equipment modifications to the University of Utah Laminar Entrained Flow Reactor and the one ton per day, pressurized Pilot Scale Gasifier; • performing Computational Fluid Dynamic modeling of industrially relevant syngas cooler configurations to develop a better understanding of deposit formation mechanisms; • performing Techno-Economic-Analysis for a representative IGCC plant to investigate the impact on plant economics, in particular the impacts on the Cost of Electricity (COE), due to plant shutdowns caused by syngas cooler plugging and fouling and potential benefits to plant economics of developing strategies to mitigate syngas cooler fouling; and • performing modeling and pilot scale tests to investigate the potential benefits of using a sorbent (fuel additive) to capture the vaporized metals that result in syngas cooler fouling. All project milestones for BP 1 and BP 2 were achieved. DOE was provided a briefing on our accomplishments in BP1 and BP2 and our proposed plans for Budget Period 3 (BP 3). Based on our research the mitigation technology selected to investigate in BP 3 was the use of a sorbent that can be injected into the gasifier with the fuel slurry to capture vaporized metals that lead to the deposit formation in the syngas cooler. The work effort proposed for BP 3 would have focused on addressing concerns raised by gasification industry personnel for the impacts on gasifier performance of sorbent injection, so that at the end of BP 3 the use of sorbent injection would be at “pre-commercial” stage and ready for use in a Field Demonstration that could be funded by industry or DOE. A Budget Continuation Application (BCA) was submitted to obtain funding for BP3 DOE but DOE chose to not fund the proposed BP3 effort.

47 OTHER INSTRUMENTATION↗

DOE OSTI2013

Model Predictive Control-based Optimal Coordination of Distributed Energy Resources

Distributed energy resources, such as renewable energy resources (wind, solar), energy storage and demand response, can be used to complement conventional generators. The uncertainty and variability due to high penetration of wind makes reliable system operations and controls challenging, especially in isolated systems. In this paper, an optimal control strategy is proposed to coordinate energy storage and diesel generators to maximize wind penetration while maintaining system economics and normal operation performance. The goals of the optimization problem are to minimize fuel costs and maximize the utilization of wind while considering equipment life of generators and energy storage. Model predictive control (MPC) is used to solve a look-ahead dispatch optimization problem and the performance is compared to an open loop look-ahead dispatch problem. Simulation studies are performed to demonstrate the efficacy of the closed loop MPC in compensating for uncertainties and variability caused in the system.

Mayhorn, Ebony T.↗

DOE OSTI2015

Scaling relation for high-temperature biodiesel surrogate ignition delay times

High-temperature Arrhenius ignition delay time correlations are useful for revealing the underlying parameter dependencies of combustion models, for simplifying and optimizing combustion mechanisms for use in engine simulations, for scaling experimental data to new conditions for comparison purposes, and for guiding in experimental design. Here, we have developed a scaling relationship for Fatty Acid Methyl Ester (FAME) ignition time data taken at high temperatures in 4%O 2 /Ar mixtures behind reflected shocks using an aerosol shock tube: τ ign [ms] = 2.24 x 10 -6 [ms] (P [atm]) -.41 ($\phi$) 0.30 (C n ) -.61 x exp $$ \left(\frac{37.1 [kcal/mol]}{\hat{R}_u [kcal / mol K] T [K]}\right) $$ In addition, we have combined our ignition delay time data for methyl decanoate, methyl palmitate, methyl oleate, and methyl linoleate with other experimental results in the literature in order to derive fuel-specific oxygen-mole-fraction scaling parameters for these surrogates. In conclusion, in this article, we discuss the significance of the parameter values, compare our correlation to others found in the literature for different classes of fuels, and contrast the above expression’s performance with correlations obtained using leading FAME kinetic models in 4%O 2 /Ar mixtures.

09 BIOMASS FUELS↗

DOE OSTI2011

Analysis Model for Domestic Hot Water Distribution Systems: Preprint

A thermal model was developed to estimate the energy losses from prototypical domestic hot water (DHW) distribution systems for homes. The developed model, using the TRNSYS simulation software, allows researchers and designers to better evaluate the performance of hot water distribution systems in homes. Modeling results were compared with past experimental study results and showed good agreement.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2012

Model biases in high-burnup fast reactor simulations

A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗