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At least 217 records · Page 12

Influence of Shear Strength Assumptions on BISON Debonding Simulations

Accurately predicting the thermomechanical response of buffer–IPyC debonding in TRISO fuel particles requires reliable mechanical property inputs for each coating layer, particularly the normal and shear strengths that influence interlayer delamination and stress concentrations. Micro tensile testing of AGR-2 fuel particles provided experimentally measured normal strengths for the buffer, IPyC, and buffer–IPyC interface; however, shear strength was not measured. As a result, BISON simulations of interface debonding must rely on assumed shear strength values, typically estimated as 20–40% of the measured ultimate tensile strength. This study evaluates how these assumed shear strength values influence cohesive zone model (CZM) predictions of buffer–IPyC separation in AGR 2 TRISO particles. Using micro tensile data from three AGR 2 compacts (2 1 3, 5 1 3, and 6 3 3), BISON simulations were performed with multiple shear strength assumptions to quantify their effect on radial and tangential stress evolution, debonding, and gap propagation. The results show that shear strength is a high sensitivity parameter: increasing the assumed shear strength significantly alters the stress distribution at the buffer–IPyC junction, shifts the predicted debonding location, and changes the extent of partial gap formation. While normal strength controls the initiation of interface separation, shear strength strongly influences the mode mixity of the failure process and the resulting stress concentrations transmitted to the IPyC and SiC layers. These findings highlight a critical gap in current TRISO mechanical characterization. Without experimentally measured shear strength, BISON simulations must rely on approximations that introduce uncertainty into predictions of coating layer integrity and fission product barrier performance. Future fuel qualification campaigns should therefore consider measurement of shear strength at the interlayer interfaces to reduce model uncertainty and improve the fidelity of TRISO fuel performance simulations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

BISON: A Flexible Code for Advanced Simulation of the Performance of Multiple Nuclear Fuel Forms [Slides]

As fuel vendors and designers pursue the development of advanced reactors or the increase in burnup limits for existing reactors, advanced computational tools are necessary to understand the fuel performance. BISON, a fuel performance code developed primarily at Idaho National Laboratory, which is built upon the Multiphysics Object-Oriented Simulation Environment (MOOSE), provides capabilities to analyze multiple nuclear fuel forms in a wide variety of dimensions. Since its inception, BISON has been used to investigate the performance of light-water reactor fuel rods, accident tolerant fuel concepts, metallic and mixed-oxide (MOX) fuels for fast reactors, plate fuels for research reactors, and tri-structural isotropic (TRISO) fuel particles. This talk will provide a history of BISON, highlights of major development milestones, tributes to key contributors, and applications of its use to various fuel forms in one-, two-, and three-dimensions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Towards a Benchmark Experiment with the Compact Nuclear Power Source

The Compact Nuclear Power Source (CNPS) was a high-assay low enriched uranium (HALEU) tristructural isotropic (TRISO)-fueled, graphite-moderated microreactor constructed in 1987 at Los Alamos National Laboratory. The reactor, conceived as a power source for short-range radar stations, was designed to be "walk-away safe," and was cooled by heat pipes and ambient air. Though the project was formally cancelled after the fuel and moderator material had been received, its potential to meaningfully advance the body of critical and integral data was evident, and critical experiments with a mock-up of the reactor proceeded at TA-18 until its disassembly in 1991. The reflector and some components of the core would later go on to see service as part of the New Production Reactor Modular High-Temperature Gas-Cooled Reactor (NP-MHTGR) critical experiments. In light of the progress made by groups like Westinghouse and X-Energy towards contemporary graphite-moderated microreactors, the system remains an attractive candidate for the basis of a benchmark experiment even today. Uncertainties remain, however, in some features of the system–inconsistencies in the dimensions and composition of the reactor’s components as described in literature and implemented in current computational models. The identification, quantification, and to the extent possible, the minimization of these uncertainties is a crucial task on the path towards creating a benchmark based on the CNPS system. The present work seeks to initiate this process with a survey of available documentation and improvements to the accuracy of CNPS neutronics models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Other Power Reactor Fuels

While oxide fuels have received the bulk of historic interest as power reactor fuels, a range of other uranium compounds and deployment architectures have also been considered and demonstrated to varying degrees. Metal alloys, technical ceramics, and oxycarbide or oxynitride particle fuels are candidates for various power reactor applications. Although in general these families of nuclear fuels have received more limited study than oxides, the benefits of uranium density, favorable thermophysical properties, and other factors have resulted in interest from both the traditional nuclear power industry as well as more recent entrepreneurial efforts. The dominant fuel chemistry issues for these fuel forms differ substantially from those of oxide systems. Synthesis routes, factors that affect fresh fuel properties, and critical aspects of fuel performance are unique to these systems and often dominated by the desired chemical behavior of uranium in each system. This chapter will briefly introduce metallic uranium alloys, specifically uranium-zirconium, non-traditional ceramic fuels such as carbides and nitrides, and oxycarbide/oxynitride particle fuels most familiar as the basis for TRISO fuel designs. A survey of major fuel chemistry issues is provided for each of these families of fuels, with a focus on contemporary challenges and active research avenues.

Nelson, Andrew↗

FY20 Verification of BISON Using Analytic and Manufactured Solutions

In 2010, the U.S. Department of Energy created its first Energy Innovation Hub, which is focused on developing high-fidelity and high-resolution Modeling and Simulation (M&S) tools for modeling of Light Water Reactors (LWRs). This hub, Consortium for Advanced Simulation of LWRs (CASL), has developed an LWR simulation tool called Virtual Environment for Reactor Applications (VERA). The multi-physics capability of VERA is achieved through the coupling of single-physics codes, including BISON, CTF, MPACT, and MAMBA. BISON is a fuel performance code which models the thermo-mechanical behavior of nuclear fuel using high performance M&S. It is capable of modeling traditional LWR fuel rods, fuel plates, and TRi-structural ISOtropic (TRISO) fuel particles. It can employ three-dimensional Cartesian, two-dimensional axisymmetric cylindrical, or one-dimensional radial spherical geometry. It includes empirical models for a large variety of fuel physics: temperature- and burnup-dependent thermal properties, fuel swelling and densification, fission gas production, cladding creep, fracture, cladding plasticity, and gap/plenum models. This document details a series of code verification test problems that are used to test BISON. These problems add confidence that the BISON code is a faithful representation of its underlying mathematical model. The suite of verification tests are mapped to the underlying conservation equations solved by the code: heat conduction, mechanics, and species conservation. Twenty-two problems are added for the heat conduction solution, two for the mechanics solution, and none for species conservation. Method of Manufactured Solutions (MMS) capability is demonstrated with three problems, and temperature drops across the fuel gap are tested.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Triso Analysis Tool For Predictive Source Terms

Source term modeling for TRi-structural ISOtropic (TRISO) fuel has been performed for previous reactor designs, but few are available in the open literature. Thus, there is a need to develop a simple, versatile, and mechanistic model of fission product release and transport in gas reactor cores that can be applied to a variety of reactors through user inputs and reactor-specific radionuclide inventories. To meet this need, the TRISO Analysis Tool for Predictive Source terms (TRISO-ATOPS) was developed. This model calculates the release of the key safety-important fission products by diffusion through the kernel, silicon carbide and graphite based on fuel and graphite temperatures in the reactor under normal operation. These releases from the fuel enter the coolant where they can plate-out on cooler surfaces. A clean-up model is included for designs with a coolant purification system for removing fission gases. This initial distribution of fission products in the reactor serves as an initial condition for potential releases under postulated accident conditions. From this initial condition, the model calculates the fission product release for any transient temperature profile, and the fission product releases can then be used to assess radiological dose to the workers and the public using conventional radiological dose tools. Data on the diffusion of fission products is based on historic German TRISO experiments and the more current Department of Energy Advanced Gas Reactor TRISO fuel development program. The example cases in this work demonstrate the flexibility of the model

Stoyer, Benjamin [Idaho National Laboratory (INL),↗

Fuel Performance Modeling Status Update and Potential Model Improvements

Fuel performance modeling status update and potential model improvements overview of TRISO fuel performance modeling codes PARFUME/BISON, AGR experiment support, potential modeling improvements, BISON smeared cracking model, BISON fission product source term, and AGR-3/4 reirradiation heating test.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Bison Analyses of Fuel Forms used in Small Modular Reactors

Small Modular Reactors (SMRs) are actively being pursued by the international nuclear industry. The Bison fuel performance code developed at Idaho National Laboratory is a tool capable of analyzing the thermo-mechanical response and species diffusion for a wide variety of fuel types under both normal and accident conditions. Two of the fuel types considered for a wide range of SMRs are TRISO fuel particles and metallic fuel. This paper provides an overview of the capabilities in Bison for these fuel types with a few example studies relevant for the SMR industry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Modeling Interface Debonding in Coated Fuel Particles with BISON

Here we introduce a meshed debonding model intended for use with tristructural isotropic (TRISO) fuel particles. The ability to better examine the potential effects of interface debonding between the coating layers will enable a host of investigations of the stress state when complex loading and interactions between coated fuel particles are taking place. In this work, the debonding model is discussed in detail. Then, simulations are presented to show the capabilities and limitations of the model. First, a traditional high-temperature gas reactor TRISO particle is considered. Then, a TRISO particle in a graphitic matrix is analyzed, followed by a TRISO particle in a silicon carbide matrix.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Candidate Core Designs for the Transformational Challenge Reactor

Early cycle activities under the Transformational Challenge Reactor (TCR) program focused on analyzing and maturing four reactor core design concepts: two fast-spectrum systems and two thermal-spectrum systems. A rapid, iterative approach has been implemented through which designs can be modified and analyzed and subcomponents can be manufactured in parallel over time frames of weeks rather than months or years. To meet key program initiatives (e.g., timeline, material use), several constraints—including fissile material availability (less than 250 kg of HALEU), component availabilities, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small (less than one cubic meter in volume) cores with near-term viability. The fast-spectrum designs did not meet the fissile material constraint, so the thermal-spectrum systems became the primary design focus. Since significant progress has been made on advanced moderator materials (YH x ) under the TCR program, gas-cooled thermal-spectrum systems using less than 250 kg of HALEU that occupy less than 1 m 3 are now feasible. The designs for two of these systems have been evolved and matured. In both thermal-spectrum design concepts, bidirectional coolant flow is used. Coolant flows down through YH x moderator elements and is reversed in a bottom manifold and core support structure, and then flows up though or around the fuel elements. The main difference between the two thermal-spectrum design concepts is the fuel elements—one uses traditional UO 2 ceramic fuel, and the other uses UN-bearing TRISO fuel particles embedded inside a SiC matrix. Finally, core neutronics and thermal performance for these systems are assessed and summarized herein.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

FPMS_XPeRT_INL_Poster

As nuclear energy expands and experimental programs increasingly rely on the facilities at Idaho National Laboratory (INL) for reactor and fuel testing, research capabilities must also expand to meet these demands. A new Fission Product Monitoring System (FPMS) has been deployed at the Advanced Test Reactor (ATR) at the Auxiliary Lead-out Experiment (ALE) House to support this expanding fuel testing mission. By tracking gaseous fission products releases from test fuel in near real-time, release rates, calculated from FPMS data, can be used to characterize the effectiveness of fuel cladding, especially for Tri-structural Isotropic (TRISO) fuel concepts. The new iteration of the FPMS supports up to 14 fission product monitors for online fission-product tracking via gamma-ray spectroscopy of the experiment’s effluent gas. Each monitor consists of a nominally 10% HPGe detector housed in a copper-lined lead shield with a warm gas trap. The new system features gamma-ray count rate information with a five-second temporal resolution and provides isotopic activity every five minutes, capable of resolving multiple overlapping fission product releases over a broad range of activities in near real-time. This work includes data from ATR cycle 175D data to demonstrate these capabilities. The hourly resolution data shows general trends and significant releases over the cycle, while the 5-minute resolution data allows for a more detailed examination of events due to unexpected particle releases.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

BISON: A Finite Element-Based Nuclear Fuel Performance Code

BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It is a multiphysics fuel analysis tool that solves fully-coupled thermomechanical problems. BISON is based on MOOSE and can efficiently solve problems using standard workstations or very large high-performance computers in a variety of different dimensions, including full 3D, 2D-RZ axisymmetric, layered axisymmetric 1D, and spherically symmetric 1D systems. It is developed by a team of scientists and engineers at Idaho National Laboratory and by collaborators. The development of BISON is supported by various funding agencies, principally the United States Department of Energy.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

INL ART AGR-5/6/7 PIE at Oak Ridge National Laboratory

Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under potential accident conditions. PIE fuel work from the final test irradiation (Advanced Gas Reactor [AGR]-5/6/7) is expected to commence at INL in early 2021, but the PIE preparations work began in FY2016. The work scope in this statement of work includes Oak Ridge National Laboratory (ORNL) providing project management and technical support to PIE-related activities; technical input to the moisture/air-ingress furnace design, fabrication, and equipment qualification; and technical support for development of equipment and techniques for planned PIE evolutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

INL ART AGR-5/6/7 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the last in the series of test irradiations, Advanced Gas Reactor (AGR)-5/6/7, will begin at Idaho National Laboratory (INL) in approximately September 2020, but because of the complexity of the experiment preparations for the PIE are going to begin in FY2016. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing project management and technical support to PIE-related activities, providing technical input to the moisture furnace design, providing technical support for development of equipment and techniques for planned PIE evolutions, and contributing general expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

INL ART AGR-5/6/7 PIE at Oak Ridge National Laboratory

Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under potential accident conditions. PIE fuel work from the final test irradiation (Advanced Gas Reactor [AGR]-5/6/7) is expected to commence at INL in early 2021, but the PIE preparations work began in FY2016. The work scope in this statement of work includes Oak Ridge National Laboratory (ORNL) providing project management and technical support to PIE-related activities; technical input to the moisture/air-ingress furnace design, fabrication, and equipment qualification; and technical support for development of equipment and techniques for planned PIE evolutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Initial results from electron microscopy from AGR 5/6/7 Capsule 1 fuel and

Initial results from electron microscopy from AGR 5/6/7 Capsule 1 fuel and applications of other microanalysis techniques to irradiated TRISO fuel. Includes detail on characterization of particles from AGR 5/6/7 Capsule 1, which include background and motivation, x-ray computed tomography on four particles from compact 1-7-9, SEM/EDS of particles from compacts 1-7-9 and 1-7-4, and Summary. Local thermal conductivity measurement on AGR-2 particles also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Model-based reconstruction for enhanced x-ray CT of dense tri-structural isotropic particles

Tri-structural isotropic (TRISO) fuel particles are a key component of next generation nuclear fuels. Using x-ray computed tomography (CT) to characterize TRISO particles is challenging because of the strong attenuation of the x-ray beam by the uranium core, leading to severe photon starvation in a substantial fraction of the measurements. Furthermore, the overall acquisition time for a high-resolution CT scan can be very long when using conventional laboratory-based x-ray systems and reconstruction algorithms. Specifically, when analytic methods such as the Feldkamp–Davis–Kress (FDK) algorithm are used for reconstruction, it results in severe streak artifacts and noise in the corresponding 3D volume, which makes subsequent analysis of the particles challenging. In this paper, we develop and apply model-based image reconstruction (MBIR) algorithms to improve the quality of CT reconstructions for TRISO particles to facilitate better characterization. We demonstrate that the proposed MBIR algorithms can significantly suppress artifacts with minimal pre-processing compared to conventional approaches. We also demonstrate that the proposed MBIR approach can obtain high-quality reconstruction compared to the FDK approach even when using a fraction of the typically acquired measurements, thereby enabling dramatically faster measurement times for TRISO particles.

Venkatakrishnan, Singanallur↗