Metallic fuel has an important historical significance in the development of nuclear reactors and continues to be relevant to the progression of advanced test and power reactors. A number of models, ranging from empirical to mechanistic, have been developed and implemented in various fuel performance codes to describe U-Zr and U-Pu-Zr fuel and typical fast reactor cladding materials. One challenge of using these models to simulate fuel performance is the inevitable tangling of coupled phenomena that can cloud proper implementation, calibration, and eventual utilization of new models. In an effort to provide a baseline capability that will facilitate the use of advanced models, new capabilities have been implemented into the fuel performance code BISON specific to metallic fuel simulations, ranging from materials properties, fission gas release and swelling calculations, coolant channel models, and cladding correlations. These models have been applied to the X441/X441A EBR-II experimental assembly data, a set of irradiated metallic UPuZr fuel rods of varying pin designs. The models implemented in BISON are able to capture the general trend of the expected response of the fuel and cladding to irradiation in EBR-II, especially when considering the spread in experimental measurements and the uncertainties inherited from the historical material models. Ultimately, the models outlined here provide the baseline capabilities on which new models can build upon in order to improve the prediction of metallic fuel performance simulations in off-normal designs or operations.

This paper seeks to introduce the latest design of the Extended Length Test Assembly-Cartridge Lead (ELTA-CL) with associated thermal-hydraulic (TH) assessment and related experiment activities to support the critical component development performed by the ELTA-CL team (Los Alamos National Laboratory, Westinghouse Electric Company, and the University of New Mexico). The goal of the ELTA-CL program is to develop and validate an experimental capability to perform irradiation experiments in the Versatile Test Reactor (VTR) addressing Lead Fast Reactor (LFR) technology gaps, in support of the commercial development of advanced lead-cooled fast reactor concepts. Through a design maturation process and parametric study, a conceptual design is proposed to meet the requirements for material and corrosion testing. Thermal-hydraulic characteristics for the conceptual design at desired operating conditions are assessed with systems-level (one-dimensional) and computational fluid dynamics (three-dimensional) simulations. Along with the conceptual design work, experimental activities for the development of critical components such as the pump and flowmeter are undertaken. From both the modeling study and the experimental results, the design requirements of the Phase 1 ELTA-CL (e.g., 500 degrees C and 2 m/s) are achievable with the current conceptual design. Additional design improvements and safety assessments at both steady-state and transient conditions for the final ELTA-CL design will be pursued.

The Versatile Test Reactor (VTR) is a sodium-cooled fast reactor designed to accelerate the design and approval of new nuclear material and reactor concepts by providing a high neutron fast flux environment on U.S. soil. To ensure that the reactor simultaneously achieves the target irradiation environment while maintaining sufficient margin to safety limits, supporting design analysis of the VTR has been performed using MCNP and TRACE. High-fidelity MCNP calculations have been performed that confirm design parameters, such as control rod worth and neutron and photon flux distributions, and provide needed reactivity coefficients for TRACE analyses. The MCNP simulations additionally provide fuel rod power profiles of interest to fuel performance designers and provide an excellent model for experimental cartridge design within the VTR core. TRACE simulations of several postulated transients, such as station blackout, loss of heat sink, and transient overpower, have been performed (results included here are limited to the transient overpower), and the obtained results confirm the robust safety behavior of the VTR. The TRACE simulations provide a valuable confirmatory transient analysis capability using a U.S. Nuclear Regulatory Commission-developed safety analysis tool incorporating inputs from the high-fidelity neutronic simulations performed with MCNP. Taken together, the confirmatory analysis capability provided by MCNP and TRACE serves to further strengthen the understanding of and confidence in the VTR's performance.

A material corrosion test loop ("Lobo Lead Loop") has been established at the University of New Mexico to investigate the compatibility of structural materials with flowing molten lead. The project aims to prequalify materials for Versatile Test Reactor (VTR) testing and support the development of computational models of flow accelerated corrosion in molten lead. The Lobo Lead Loop is designed to operate at high temperatures up to 700 degrees C and high mean flow velocities reaching 3 m/s within the specimen holder channels. Numerical simulations are utilized to design loop components with performance that allows for achieving the operation targets. Specimen holders are designed for (a) multi-material testing at 3 m/sec, (b) multi-velocity testing, (c) shear stress testing, and (d) high temperature (>= 600 degrees C) testing at low mass flow rates. The computational fluid dynamics models used are shown to produce results in agreement with experimental data for flow in pipes. The models are then used in parametric analyses to identify parameters of importance to specimen holder design and computation of pressure losses for flows along smooth and rough specimen walls. Specimen holder designs for the purposes outlined are presented along with performance curves. The maximum achievable flow rate is estimated based on the intersection of the pump performance curve with the system curve.

Based on available modeling and simulation capabilities of peridynamics module in MOOSE framework for oxide fuel, the overall goal of this project is to further develop the peridynamics capabilities for modeling metallic fuel. It includes two major tasks: 1) develop validated scheme to handle the shape tensor singularity due to insufficient active neighbors of a material particle in the peridynamic correspondence model for fracture problems, and 2) develop failure modeling scheme including failure criterion for metallic fuels. Before the peridynamics can be applied to model metallic fuel, the formulation instability of the peridynamic correspondence model should be addressed. The PI first worked on developing new stabilization method to improve the performance of the peridynamic correspondence model and reduce the possibility of getting a singular shape tensor while applying the model for fracture problems. The new stabilization scheme uses bond-associated weight function rather than bond-associated horizon. Compared to bondassociated horizon stabilized method, this new stabilization scheme has better performance with improved prediction accuracy and reduced free surface effect. Using this newly developed stabilization, materials models from BISON can be directly used in peridynamics for metallic fuels, such as fission rate and burnup dependent creep and swell models. Publication of this work is under preparation.

In order to accurately predict the performance of materials under dynamic loading conditions, models have been developed that describe the rate-dependent material behavior and irrecoverable plastic deformation that occurs at elevated strains and applied loads. Most of these models have roots in empirical fits to data and, thus, require the addition of specific parameters that reflect the properties and response of specific materials. In this work, we present a systematic approach to the problem of calibrating a Johnson-Cook plasticity model for 304L stainless steel using experimental testing in which the parameters are treated as dependent on the state of the material and uncovered using experimental data. The results obtained indicate that the proposed approach can make the presence of a discrepancy term in calibration unnecessary and, at the same time, improve the prediction accuracy of the model into new input domains and provide improved understanding of model bias compared to calibration with stationary parameter values.

This work is motivated by the need to analyze the behavior of nuclear fuels which under normal operating conditions build up stresses due to non-homogeneous thermal expansion, fission gas and solid product swelling among other phenomena that are simultaneously relaxed by creep and plastic flow. This report details the stress and tangent update equations for combined J2 based rate independent plasticity, timehardening creep and fission gas induced swelling e ects in a fully implicit hypo-elastic formulation involving two cases: pure creep without plasticity where the yield criterion has not yet been met and the combined e ect of both creep and plasticity beyond yield. Closed form expressions for the consistent material tangent to be used in both cases are derived which can be used in implicit codes and is expected to help in obtaining optimal convergence rates.

Purpose Partitioned analysis is an increasingly popular approach for modeling complex systems with behaviors governed by multiple, interdependent physical phenomena. Yielding accurate representations of reality from partitioned models depends on the availability of all necessary constituent models representing relevant physical phenomena. However, there are many engineering problems where one or more of the constituents may be unavailable because of lack of knowledge regarding the underlying principles governing the behavior or the inability to experimentally observe the constituent behavior in an isolated manner through separate-effect experiments. This study aims to enable partitioned analysis in such situations with an incomplete representation of the full system by inferring the behavior of the missing constituent. Design/methodology/approach This paper presents a statistical method for inverse analysis infer missing constituent physics. The feasibility of the method is demonstrated using a physics-based visco-plastic self-consistent (VPSC) model that represents the mechanics of slip and twinning behavior in 5182 aluminum alloy. However, a constituent model to carry out thermal analysis representing the dependence of hardening parameters on temperature is unavailable. Using integral-effect experimental data, the proposed approach is used to infer an empirical constituent model, which is then coupled with VPSC to obtain an experimentally augmented partitioned model representing the thermo-mechanical properties of 5182 aluminum alloy. Findings Results demonstrate the capability of the method to enable model predictions dependent upon relevant operational conditions. The VPSC model is coupled with the empirical constituent, and the newly enabled thermal-dependent predictions are compared with experimental data. Originality/value The method developed in this paper enables the empirical inference of a functional representation of input parameter values in lieu of a missing constituent model. Through this approach, development of partitioned models in the presence of uncertainty regarding a constituent model is made possible.

Early in its development, the primary emphasis of efforts to improve the Bison fuel performance code were focused on initial development of new capabilities. As Bison has matured and become adopted by a wider set of users, it has become clear that a number of robustness issues needed to be addressed to ensure that the code can reliably produce a converged solution. This need became particularly evident when a recent effort by Westinghouse Electric Company to simulate all of the fuel rods in a reactor using Bison resulted in approximately 30% of the models failing to run to completion because of lack of convergence or other errors that were encountered. It is important to emphasize that there was no evidence that Bison generated results that were incorrect for the set of models exercised. In many cases, issues such as those encountered could be addressed by adjusting solver-related parameters. However, there is no question that for Bison to become widely adopted for production work, it needs to be inherently far more robust than it was. Because of this, a concerted effort was undertaken to improve Bison's robustness, or its ability to reliably obtain converged solutions under a wider variety of conditions. As a result of this work, a number of high-priority areas for improvement for modeling light water reactor (LWR) fuel rods have been identified and addressed. As a result of this work, the number of failing Bison analyses in the aforementioned set of simulations for a full core has been reduced to a fraction of a percent. In addition, the robustness of Bison models using the VERA core simulator has been significantly improved. This is important because all of the Bison models representing individual fuel rods must run to completion successfully for the full core model to run to completion. This report provides summaries of the identified sources of robustness issues in Bison and the development that was done to resolve these issues. It also summarizes the addition of new capabilities to address deficiencies in Bison's ability to include needed aspects of LWR fuel behavior, and improvements that were made to improve Bison's solution efficiency.

The fuel cladding chemical interaction (FCCI) phenomenon is potentially the main factor restricting the application of metallic fuels in liquid sodium cooled fast reactors. The understanding of the lanthanide (Ln) transport behaviors in liquid Cs filled pores in U-Zr fuel is essential for understanding FCCIs. By using ab initio molecular dynamics, fundamental properties of the metallic system Cs-Ln, such as density of states and coordination number, have been studied. Then, the Ln diffusivities in liquid cesium and the solution viscosity were calculated. For validating the model, the viscosity of the pure liquid Cs which has been well measured is also calculated at three temperatures, which indicates the present model has a high accuracy in calculation of viscosity and self-diffusivity of Cs in liquid Cs.

It is recognized that the lanthanide fission products can enhance the fuel-cladding chemical interaction (FCCI), which is a key concern of using metallic fuels such as U-Zr in a sodium-cooled fast reactor. The present work conducts a critical review on the analysis of lanthanide functions in FCCIs. Available in-pile and out-of-pile data are first collected and analyzed, then the theories for lanthanide migration and redistribution in the fuel in operation are analyzed. For mitigating FCCIs, one of the effective method is applying fuel additives for immobilizing lanthanides. The review investigates four candidates of fuel additives (Pd, Sn, Sb and In), and it is concluded that Sb can be the best candidate among the four elements based on current available thermodynamic data and microstructure characterizations. Considering that FCCIs lead to formation of metallic alloys/compounds between lanthanides and steel cladding constituents (Fe, Cr, and Ni), the review also analyzes the thermodynamic data such as enthalpy of formation of the alloys and/or metallic compounds and identify the solidus temperature of each alloy.