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Challenging Common Assumptions of Thick-Wall Chamber Dynamics in Inertial Fusion Systems using MOOSE

As an increasing number of companies look toward commercial Inertial Fusion Energy (IFE) designs, there is a pressing need to understand the physics of thick-wall chamber gas dynamics. The thick liquid wall approach implements a renewable wall to mitigate the fusion target emissions, thereby reducing the radiation damage rate and significantly extending the lifetime of chamber structures, leading to increased plant availability and reduced waste streams in comparison to dry wall chamber designs. It is necessary, however, to assess the critical performance and safety aspects of these systems. For example, it is crucial to predict (1) where the mass ablated from the liquid walls will vent, which determines the placement of condensing surfaces; (2) debris propagation up the beam lines, which provides essential information for design and protection requirements; (3) peak pressures and impulse on chamber walls, which affect chamber structural design; and (4) momentum transfer to the liquid jets, which constrains the shape and positioning of the jets. In turn, the chamber design and its liquid walls affect shielding requirements, material activation, and tritium fuel cycle. Currently available simulation tools, however, are unable to accurately capture key thick-wall chamber dynamics. Significant assumptions are often made to simplify the system and reduce computational cost and modeling capability needs, but the impact of these assumptions on simulation predictions has not been evaluated. For example, no three-dimensional simulations can be found in the open literature to evaluate gas venting and momentum transfer to the jets with simulations using two-dimensional domains to represent complex three-dimensional geometries. Moreover, limited studies have been dedicated to jet breakup due to both turbulence and neutron heating, and no studies have been found that evaluate how jet breakup can impact shock-jet interaction. Furthermore, effects of radiative heat transfer have rarely been included for the hydrodynamic phase of shock propagation, and integration of proper equations of state in shock dynamics codes has been mostly exploratory. In this study, we use the flexible, high-fidelity Multiphysics Object-Oriented Simulation Environment (MOOSE) to model these complex phenomena and inform design and safety studies. Capabilities to model thick-wall chamber gas dynamics are being developed, and the impact of the assumptions listed above (i.e., two-dimensional vs three-dimensional, absence of jet breakout, no radiative heat transfer, and ideal gas behavior) are being quantified.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Development and Testing of an UltraBattery-Equipped Honda Civic Hybrid

The UltraBattery Retrofit Project DP1.8 and Carbon Enriched Project C3, performed by ECOtality North America (ECOtality) and funded by the U.S. Department of Energy and the Advanced Lead Acid Battery Consortium (ALABC), are established to demonstrate the suitability of advanced lead battery technology in hybrid electrical vehicles (HEVs). A profile, termed the “Simulated Honda Civic HEV Profile” (SHCHEVP) has been developed in Project DP1.8 in order to provide reproducible laboratory evaluations of different battery types under real-world HEV conditions. The cycle is based on the Urban Dynamometer Driving Schedule and Highway Fuel Economy Test cycles and simulates operation of a battery pack in a Honda Civic HEV. One pass through the SHCHEVP takes 2,140 seconds and simulates 17.7 miles of driving. A complete nickel metal hydride (NiMH) battery pack was removed from a Honda Civic HEV and operated under SHCHEVP to validate the profile. The voltage behavior and energy balance of the battery during this operation was virtually the same as that displayed by the battery when in the Honda Civic operating on the dynamometer under the Urban Dynamometer Driving Schedule and Highway Fuel Economy Test cycles, thus confirming the efficacy of the simulated profile. An important objective of the project has been to benchmark the performance of the UltraBatteries manufactured by both Furukawa Battery Co., Ltd., Japan (Furakawa) and East Penn Manufacturing Co., Inc. (East Penn). Accordingly, UltraBattery packs from both Furakawa and East Penn have been characterized under a range of conditions. Resistance measurements and capacity tests at various rates show that both battery types are very similar in performance. Both technologies, as well as a standard lead-acid module (included for baseline data), were evaluated under a simple HEV screening test. Both Furakawa and East Penn UltraBattery packs operated for over 32,000 HEV cycles, with minimal loss in performance; whereas the standard lead-acid unit experienced significant degradation after only 6,273 cycles. The high-carbon, ALABC battery manufactured in Project C3 also was tested under the advanced HEV schedule. Its performance was significantly better than the standard lead-acid unit, but was still inferior compared with the UltraBattery. The batteries supplied by Exide as part of the C3 Project performed well under the HEV screening test, especially at high temperatures. The results suggest that higher operating temperatures may improve the performance of lead-acid-based technologies operated under HEV conditions—it is recommended that life studies be conducted on these technologies under such conditions.

25 ENERGY STORAGE↗

The Safe Guards Analysis Toolbox for Non-Proliferation Modeling

As nuclear technology becomes more widely available outside of nuclear weapon armed nations, the importance of preventing the illicit development of nuclear weapons has increased dramatically. A system of nuclear safeguards prevents proliferation; stopping the diversion of nuclear material from civilian facilities for the construction of nuclear weapons. These safeguards function through a combination of accountancy measurements that confirm the locations and amounts of nuclear material, and physical security at every nuclear facility. Safeguards Analysis (SGA) is a computational tool-kit developed using the MATLAB Simulink software suite that simulates a system of nuclear safeguards. The code allows users to examine current safeguards systems for weaknesses and to pinpoint the best locations for improvement. Developed as a modular framework, SGA does not directly model material flows, and relies on an outside code for this information. Instead, it represents a series of safeguards measurements that represent the values inspectors would see. Furthermore, SGA is designed to be modular in design, allowing the user to construct as complex or simple a model as needed: simulating an entire fuel-cycle comprising multiple facilities to a single measurement. SGA's use of outside codes for its material flow information makes it applicable to any model available to the user, as opposed to previous work that applied only to a single scenario. This paper discusses the initial development of SGA and the testing of these modules. The tests described in this paper demonstrate that the different pieces of SGA currently work as intended. Future work on the toolbox will develop and test SGA against a more realistic and complicated scenario using the various tools available in SGA. SGA currently has the pieces to perform analysis on safeguards systems, but needs to be tested on large, multi-MBA, systems to ensure the code is capable of handling such large models, to see the effect large models have on calculation speed, and if further work is needed to improve code efficiency. (authors)

97 MATHEMATICS AND COMPUTING↗

Optimal velocity prediction for fuel economy improvement of connected vehicles

With the advancement of vehicle-to-vehicle and vehicle-to-infrastructure technologies, more and more real-time information regarding traffic and transportation system will be available to vehicles. This paper presents the development of a novel algorithm that uses available velocity bounds and powertrain information to generate an optimal velocity trajectory over a prediction horizon. When utilised by a vehicle, this optimal velocity trajectory reduces fuel consumption. The objective of this optimisation problem is to reduce dynamic losses, required tractive force, and completing trip distance with a given travel time. Sequential quadratic programming method is employed for this nonlinearly constrained optimisation problem. When applied to a GM Volt-2, the generated velocity trajectory saves fuel compared to a real-world drive cycle. The simulation results confirm the fuel consumption reduction with the rule-based mode selection and the energy management strategy of a GM Volt 2 model in Autonomie.

33 ADVANCED PROPULSION SYSTEMS↗

Development of predictive capability of cycle-to-cycle variation in dual-fuel engines using supercomputing-based computational fluid dynamics

Cycle-to-cycle variation (CCV) in internal combustion engines is expected due to variabilities in operating conditions such as in-cylinder fuel stratification, mixing behavior, manifold air pressure, injection timing etc. Generally, engine experimental studies provide averaged quantities such as cylinder pressure, emissions, indicated mean effective pressure (IMEP) from multiple cycles and their CCV of IMEP, and CCV of cylinder pressure from consecutive cycles for a given test condition. Capturing 3D spatial distributions of relevant measures such as fuel-air mixing, temperature, turbulence levels and emissions from such experiments is challenging. Computational Fluid Dynamics (CFD) is an alternative to experiments, that can be used effectively to understand the spatial and temporal distributions of parameters of interest.Quantitative understanding of CCV in dual-fuel (DF) engines will accelerate the development and deployment of engines that provide fuel flexibility, economic advantage and emission compliance. Earlier work [1] performed to develop an understanding of the physics of CCV in large-bore, medium-speed, dual-fuel engines was partially successful. That effort entailed a sparse design of experiment (DOE) of full-cylinder geometry, closed-cycle, dual-fuel CFD simulations in which variabilities within some of the identified global parameters were exercised. The Coefficient of Variation (COV) of cylinder pressure was promising compared with experimental COV. This reinforced the understanding of which global parameters are significant contributors to CCV in DF combustion.The authors have used several high-performance supercomputing facilities in the process of developing predicting capability of CCV in large-bore engines, while operating in dual-fuel, diesel and natural gas (NG), combustion mode [1, 2]. In the current study CCV influencing parameters in a diesel-natural gas dual-fuel engine was studied using a 4th order DOE consisting of 137 concurrent cases. These full-geometry, closed-cycle, CFD simulations were run on the Mira supercomputing facility at the Argonne National Laboratory (ANL). This paper describes key learnings of that study along with a validation of the DOE. Further, the CCV prediction methodology, which is applicable to simulation-based design and testing, is described.

Gubba, Sreenivasa Rao↗

Thermal Properties Measurement Report

The Thermal Properties Measurement Report summarizes the research, development, installation, and initial use of significant experimental thermal property characterization capabilities at the INL in FY 2015. These new capabilities were used to characterize a U 3 Si 2 (candidate Accident Tolerant) fuel sample fabricated at the INL. The ability to perform measurements at various length scales is important and provides additional data that is not currently in the literature. However, the real value of the data will be in accomplishing a phenomenological understanding of the thermal conductivity in fuels and the ties to predictive modeling. Thus, the MARMOT advanced modeling and simulation capability was utilized to illustrate how the microstructural data can be modeled and compared with bulk characterization data. A scientific method was established for thermal property measurement capability on irradiated nuclear fuel samples, which will be installed in the Irradiated Material Characterization Laboratory (IMCL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Multi-level computational exploration of advanced combustion engine operating strategies

Advanced combustion engine (ACE) research is typically carried out on single-cylinder research engines. These engines are designed to tightly control fueling and conditions at intake valve closure (IVC) and to precisely measure in-cylinder conditions and emissions. However, to be able to measure and control engine operation so precisely, these research engines typically do not feature intake and exhaust tracts that resemble those in production engines, specifically in regards to turbomachinery, heat exchangers, and exhaust gas recirculation (EGR) systems. For this reason, these research engines are effective for understanding in-cylinder combustion parameters such as heat release rate, burn duration, combustion efficiency, pollutant formation, and exhaust valve opening (EVO) conditions. Here, this paper applies high fidelity simulations to determine the feasibility of achieving a chosen single cylinder engine operating point on a production type homogeneous charge compression ignition (HCCI) engine, using a partial fuel stratification (PFS) strategy. To accomplish this, a Converge 3 dimensional (3D) – computational fluid dynamics (CFD) model of the experimental combustion chamber and intake and exhaust runners was created to simulate the experimental engine. This model was used to simulate an operating point achieved experimentally, as well as to determine the sensitivity of the operating point to variations in intake pressure, intake temperature, injection timing, injected mass, and EGR fraction. The results from these simulations were fed into a 1-dimensional engine simulation created in AVL Boost, featuring production-type intake and exhaust systems, including turbomachinery and heat exchangers necessary to create the required IVC conditions. This full engine simulation was used to assess the cycle efficiency of the engine at the experimental operating condition, and to assess whether changes to this operating point in intake temperature, intake pressure, direct injection timing, or fueling are beneficial to the cycle efficiency and engine-out emissions. In addition, the sensitivity of promising engine operating points to injection timing and injection mass are determined to evaluate the potential stability of these operating points.

33 ADVANCED PROPULSION SYSTEMS↗

Impact of the deployment schedule of fast breeding reactors in the frame of French act for nuclear materials and radioactive waste management

In the frame of the French Act of June 28, 2006 on 'a sustainable management of nuclear materials and radioactive waste' EDF R and D assesses various research scenarios of transition between the actual French fleet and a Generation IV fleet with a closed fuel cycle where plutonium is multi-recycled. The basic scenarios simulate a deployment of 60 GWe of Sodium-cooled Fast Reactors (SFRs) in two steps: one third from 2040 to 2050 and the rest from 2080 to 2100 (scenarios 2040). These research scenarios assume that SFR technology will be ready for industrial deployment in 2040. One of the many sensitivity analyses that EDF, as a nuclear power plant operator, must evaluate is the impact of a delay of SFR technology in terms of uranium consumptions, plutonium needs and fuel cycle utilities gauging. The sensitivity scenarios use the same assumptions as scenarios 2040 but they simulate a different transition phase: SFRs are deployed in one step between 2080 and 2110 (scenarios 2080). As the French Act states to conduct research on minor actinides (MA) management, we studied different options for 2040 and 2080 scenarios: no MA transmutation, americium transmutation in heterogeneous mode based on americium Bearing Blankets (AmBB) in SFRs and all MA transmutation in heterogeneous mode based on MA Bearing Blankets (MABB). Moreover, we studied multiple parameters that could impact the deployment of these reactors (SFR load factor, increase of the use of MOX in Light Water Reactors, increase of the cooling time in spent nuclear fuel storage...). Each scenario has been computed with the EDF R and D fuel cycle simulation code TIRELIRE-STRATEGIE and optimized to meet various fuel cycle constraints such as using the reprocessing facility with long period of constant capacity, keeping the temporary stored mass of plutonium and MA under imposed limits, recycling older assemblies first... These research scenarios show that the transition from the current PWR fleet to an equivalent fleet of Generation IV SFR can follow different courses. The design of SFR-V2B that we used in our studies needs a high inventory of plutonium resulting in tension on this resource. Several options can be used in order to loosen this tension: our results lead to favour the use of axial breeding blanket in SFR. Load factor of upcoming reactors has to be regarded with attention as it has a high impact on plutonium resource for a given production of electricity. The deployment of SFRs beginning in 2080 instead of 2040 following the scenarios we described creates higher tensions on reprocessing capacity, separated plutonium storage and spent fuel storage. In the frame of the French Act, we studied minor actinides transmutation. The flux of MA in all fuel cycle plants is really high, which will lead to decay heat, a and neutron emission related problems. In terms of reduction of MA inventories, the deployment of MA transmutation cycle must not delay the installation of SFRs. The plutonium production in MABB and AmBB does not allow reducing the use of axial breeding blankets. The impact of MA or Am transmutation over the high level waste disposal is more important if the SFRs are deployed later. Transmutation option (americium or all MA) does not have a significant impact on the number of canister produced nor on its long-term thermal properties. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Redwing: A MOOSE application for coupling MPACT and BISON

Fuel performance and whole core neutron transport programs are often used to analyze fuel behavior as it is depleted in a reactor. For fuel performance programs, internal models provide the local intra-pin power density, fast neutron flux, burnup, and fission rate density, which are needed for a fuel performance analysis. The fuel performance internal models have a number of limitations. These include effects on the intra-pin power distribution by nearby assembly elements, such as water channels and control rods, and the further limitation of applicability to a specified fuel type such as low enriched UO2. In addition, whole core neutron transport codes need an accurate intra-pin temperature distribution in order to calculate neutron cross sections. Fuel performance simulations are able to model the intra-pin fuel displacement as the fuel expands and densifies. These displacements must be accurately modeled in order to capture the eventual mechanical contact of the fuel and the clad; the correct radial gap width is needed for an accurate calculation of the temperature distribution of the fuel rod. Redwing is a MOOSE-based application that enables coupling between MPACT and BISON for transport and fuel performance coupling. MPACT is a 3D neutron transport and reactor core simulator based on the method of characteristics (MOC). The development of MPACT began at the University of Michigan (UM) and now is under the joint development of ORNL and UM as part of the DOE CASL Simulation Hub. MPACT is able to model the effects of local assembly elements and is able calculate intra-pin quantities such as the local power density on a volumetric mesh for any fuel type. BISON is a fuel performance application of Multi-physics Object Oriented Simulation Environment (MOOSE), which is under development at Idaho National Laboratory. BISON is able to solve the nonlinearly coupled mechanical deformation and heat transfer finite element equations that model a fuel element as it is depleted in a nuclear reactor. Redwing couples BISON and MPACT in a single application. Redwing maps and transfers the individual intra-pin quantities such as fission rate density, power density, and fast neutron flux from the MPACT volumetric mesh to the individual BISON finite element meshes. For a two-way coupling Redwing maps and transfers the individual pin temperature field and axially dependent coolant densities from the BISON mesh to the MPACT volumetric mesh. Details of the mapping are given. Redwing advances the simulation with the MPACT solution for each depletion time step and then advances the multiple BISON simulations for fuel performance calculations. Sub-cycle advancement can be applied to the individual BISON simulations and allows multiple time steps to be applied to the fuel performance simulations. Currently, only loose coupling where data from a previous time step is applied to the current time step is performed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

An assessment of the effect on Olkiluoto repository capacity achievable with advanced fuel cycles

Previously a few scenarios have been simulated for transition from thermal to fast reactor fleet in Finland in order to determine how much the transuranic inventory could be reduced with the partitioning and transmutation (P-T) technologies. Those calculations, performed with COSI6 code developed by CEA, are extended in the present study, in which the effect of P-T on the capacity of the planned final disposal repository at Olkiluoto (Finland) is evaluated by taking into account the created fission products and transuranic residuals from the reprocessing operations. The decay heat is assumed to be the most restrictive factor in defining the waste disposal packing density. The repository capacity evaluation of this study is based on the comparison of the decay heats produced by the deposited waste in various scenarios. The reference scenario of this article involves only Light Water Reactors (LWR) in an open fuel cycle. The capacity requirement of the geological repository is estimated in a few closed fuel cycle scenarios, all including actinide transmutation with Fast Reactors (FR). The comparison between the P-T scenarios and reference is based on the decay heat production of the deposited waste. The COSI6 code is used for simulations to provide the repository decay heat curves. Applying the closed fuel cycle would change the disposal concept and schedule, because of which it is not quite straightforward to assess the impact of P-T on the capacity. However, it can be concluded that recycling the transuranic nuclides probably decreases the required volume for the disposal, but thermal dimensioning analysis is needed for more specific conclusions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Computational Fluid Dynamics of rising droplets

The main goal of this study is to perform simulations of droplet dynamics using Truchas, a LANL-developed computational fluid dynamics (CFD) software, and compare them to a computational study of Hysing et al.[IJNMF, 2009, 60:1259]. Understanding droplet dynamics is of fundamental importance in liquid-liquid extraction, a process used in the nuclear fuel cycle to separate various components. Simulations of a single droplet rising by buoyancy are conducted in two-dimensions. Multiple parametric studies are carried out to ensure the problem set-up is optimized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

An investigation on the fuel savings potential of hybrid hydraulic refuse collection vehicles

Highlights: • Driving cycle acquisition in a refuse collection vehicle. • Vehicle modeling and validation for numerical simulations based on the measured driving cycle. • Fuel consumption analysis for a conventional diesel vehicle and a hybrid hydraulic vehicle. - Abstract: Refuse trucks play an important role in the waste collection process. Due to their typical driving cycle, these vehicles are characterized by large fuel consumption, which strongly affects the overall waste disposal costs. Hybrid hydraulic refuse vehicles offer an interesting alternative to conventional diesel trucks, because they are able to recuperate, store and reuse braking energy. However, the expected fuel savings can vary strongly depending on the driving cycle and the operational mode. Therefore, in order to assess the possible fuel savings, a typical driving cycle was measured in a conventional vehicle run by the waste authority of the City of Stuttgart, and a dynamical model of the considered vehicle was built up. Based on the measured driving cycle and the vehicle model including the hybrid powertrain components, simulations for both the conventional and the hybrid vehicle were performed. Fuel consumption results that indicate savings of about 20% are presented and analyzed in order to evaluate the benefit of hybrid hydraulic vehicles used for refuse collection.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Preliminary Insights on Potential Transitions to Closed Thorium-Based Fuel Cycles

Closed fuel cycles with comprehensive recycle strategies are often regarded as the logical and desired end-states of long-term nuclear energy strategies; such options offer potential improvements in terms of resource supply and sustainability, waste management, and potentially nuclear security. In the DOE-NE Fuel Cycle Option Campaign report, systems with closed recycle strategies (both uranium- and thorium-based) generally dominated strategies with less comprehensive recycle approaches across many performance metrics. That being said, developing momentum to transition from the current global implementation of once-through fuel cycles has proven challenging. Intermediate paths en route to closed fuel cycles have not been readily elucidated; this is especially true for thorium-based options. A comparatively extensive amount of work has been conducted on transitions to closed fuel cycles based on uranium and plutonium, generally based on an endpoint of sodium-cooled fast reactors (SFRs). In contrast, corresponding analyses for thorium-based options are limited. Early estimates, without the benefit of full-scale modeling, were performed in conjunction with the Molten Salt Fast Reactor (MSFR) design. There is some transition analysis work associated with the Japanese THORIMS-NES molten salt design, but the scope of this work was not directly comparable with the SFR-oriented analyses and considered the thorium fission reactors as an intermediate state en route to the global deployment of accelerator-driven systems. Others have considered thorium recently, but not in a closed-recycle context. To begin to resolve this gap, this paper describes a high-level analysis and initial perspectives on the topic of transitions to closed thorium-based cycles. The emphasis on this paper is not on the specific quantitative results that are described, although some of the overall takeaways from these trends may still be useful. Instead, this paper describes the need for such analyses to be conducted, describes the methodologies by which this can be accomplished, and reports an initial demonstration of these methodologies for a specific transition candidate scenario. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Non-algorithmic Capability Gaps for CYCLUS and CYCAMORE Transition Analyses

As part of NEUP-FY16-10512, fuel cycle transition scenarios were simulated using CYCLUS and existing CYCAMORE archetypes. The purpose of this paper is to identify current non-algorithmic gaps in the capabilities necessary for key transition scenarios. The gaps identified through this exercise mainly pertain to the greedy exchange model, and the manual, static parameter of fuel cycle facilities. The scenarios are from the Idaho National Laboratory Nuclear Fuel Cycle Evaluation and Screening Report (Hereby E&S Report). The transition scenarios begin with EG01 and transition to EG23, EG24, EG29, EG30, separately.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Spanning the Core Design Space with Binary Perturbations - A Mathematical Investigation

A reactor core design engineer, even one with many years of experience or one using advanced fuel pattern optimization software, will likely fine tune the pattern before finalizing the design. Fine tuning can further minimize local enthalpy peaks, optimize reactivity to extend burnup, and add additional margin to limits on reactivity coefficients or heat flux. For reactors that have reached an equilibrium fuel cycle, the number of feeds each cycle is typically the similar to previous cycles as is their placement in the core. It is then the engineer's responsibly to efficiently select the placement of burned assemblies to ensure optimal fuel utilization, satisfy safety limits, and meet cycle energy requirements. One of the engineer's most commonly used tools in this fine tuning investigation is to swap the positions of two assemblies, most likely two previously irradiated ('burned') assemblies since their reactivity is lower than fresh fuel. Such an operation is referred to in this work as a binary perturbation or a 'swap'. If quarter core symmetry is maintained, the perturbation translates to the other three quadrants. Likewise if eighth core symmetry is maintained, the perturbation translates to other seven octants. The question is then, how much of the core design space can binary perturbations possibly span? This work seeks to quantitatively define the space spanned by m number of binary perturbations as a function of variable size through combinatorial principles and computer simulation. The results of this study are useful in understanding the in-core fuel management problem faced by both professionals who employ hand manipulations and those who attempt stochastic solutions of the problem.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Two-Dimensional Neutronic and Fuel Cycle Analysis of the Transatomic Power Molten Salt Reactor

This status report presents the results from the first phase of the collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear, Nuclear Energy Voucher program. The TAP design is a molten salt reactor using movable moderator rods to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this design. Additional analyses of time step sizes, mass feed rates and enrichments, and isotopic removals provide additional information to make informed design decisions. This work further demonstrates capabilities of ORNL modeling and simulation tools for analysis of molten salt reactor designs and strongly positions this effort for the upcoming three-dimensional core analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Uncertainty Quantification and Sensitivity Analysis for Coupled VERA-CS and FRAPCON

The current nuclear power plant fleet in the U.S. is facing numerous challenges economically ranging from low natural gas price to the rapidly growing renewable energy sources such as wind and solar power. In order for the nuclear industry to keep the viability of its operating fleet, new innovations have to be introduced to lower the costs to operate the fleet. One area that could contribute to the cost reduction is to lower the fuel cycle cost and improve the plant safety through initiatives such as higher burnup fuel and accident tolerant fuel. At the same time, the nuclear industry has to address future challenges on the horizon such as load following and other flexible operating strategies as well as those associated with new regulations such as 10 CFR 50.69 and 10 CFR50.46c etc. All these require coupled analysis of core neutronics, thermal hydraulics and fuel performance, which necessitates the development of a multi-physics toolkit. Since different models, assumptions, and fidelities have gone into various computer codes, how to consistently propagate uncertainties in a multi-physics environment warrants careful evaluation. The U.S. DOE's Light Water Reactor Sustainability (LWRS) Program initiated a project to develop analytical capabilities to support the industry to address these challenges. The general idea behind the initiative was the development of an integrated multiphysics evaluation model. This integrated evaluation model is called LOTUS which stands for LOCA Toolkit for the U.S. Light Water Reactors. The focus of LOTUS is to establish the automation of the interface between the five disciplines. LOTUS framework ensures that uncertainties can be easily propagated to solve the problems forward and backward (inverse) with large numbers of simulations. The objective of this work is to perform uncertainty quantification (UQ) and sensitivity analysis (SA) on a coupling of CASL's Virtual Environment for Reactor Applications-Core Simulator (VERA-CS) and NRC's fuel performance code FRAPCON. The work presented herein is a subset of the overall LOTUS framework. The figures of merit (FOM) selected for analysis are minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and the gap conductance at peak power (GCPP). All FOM are obtained at the end of the of a depletion case for a half cycle lasting 263 days. The minimum is taken over all spatial location for each pin at the end of the half cycle. This value is calculated in VERACS alone and has been analyzed in previous work by Brown and Zhang. The MFCT is available in both VERA-CS and FRAPCON, but is calculated via different methods (see Model Inconsistencies section for more details). For this work, only the FRAPCON values are reported due to the code's maturity from extensive validation in fuel performance. Previous work includes SA on a standalone nuclear fuel performance code VTT-modified FRAPCON- 3.4 by Ikonen. The GCPP is the gap conductance at the location corresponding to maximum power in the fuel rod. This value is available in FRAPCON alone and highlights the capability of LOTUS to obtain unique fuel performance metrics for each pin in a core simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗