Engineering topics

Pastore, G.

Publications and source records attributed to Pastore, G..

34 records · Page 2

DOE OSTI2016

PolyPole-1: An accurate numerical algorithm for intra-granular fission gas release

This paper describes the development of a new numerical algorithm (called PolyPole-1) to efficiently solve the equation for intra-granular fission gas release in nuclear fuel. The work was carried out in collaboration with Politecnico di Milano and Institute for Transuranium Elements. The PolyPole-1 algorithms is being implemented in INL's fuels code BISON code as part of BISON's fission gas release model. The transport of fission gas from within the fuel grains to the grain boundaries (intra-granular fission gas release) is a fundamental controlling mechanism of fission gas release and gaseous swelling in nuclear fuel. Hence, accurate numerical solution of the corresponding mathematical problem needs to be included in fission gas behaviour models used in fuel performance codes. Under the assumption of equilibrium between trapping and resolution, the process can be described mathematically by a single diffusion equation for the gas atom concentration in a grain. In this work, we propose a new numerical algorithm (PolyPole-1) to efficiently solve the fission gas diffusion equation in time-varying conditions. The PolyPole-1 algorithm is based on the analytic modal solution of the diffusion equation for constant conditions, with the addition of polynomial corrective terms that embody the information on the deviation from constant conditions. The new algorithm is verified by comparing the results to a finite difference solution over a large number of randomly generated operation histories. Furthermore, comparison to state-of-the-art algorithms used in fuel performance codes demonstrates that the accuracy of the PolyPole-1 solution is superior to other algorithms, with similar computational effort. Finally, the concept of PolyPole-1 may be extended to the solution of the general problem of intra-granular fission gas diffusion during non-equilibrium trapping and resolution, which will be the subject of future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

BISON Theory Manual The Equations behind Nuclear Fuel Analysis

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 solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact, and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Summary of BISON Development and Validation Activities - NEAMS FY16 Report

This summary report contains an overview of work performed under the work package en- titled “FY2016 NEAMS INL-Engineering Scale Fuel Performance (BISON)” A first chapter identifies the specific FY-16 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes major additional accomplishments, which in- clude: 1) publication of a journal article on solution verification and validation of BISON for LWR fuel, 2) publication of a journal article on 3D Missing Pellet Surface (MPS) analysis of BWR fuel, 3) use of BISON to design a unique 3D MPS validation experiment for future in- stallation in the Halden research reactor, 4) participation in an OECD benchmark on Pellet Clad Mechanical Interaction (PCMI), 5) participation in an OECD benchmark on Reactivity Insertion Accident (RIA) analysis, 6) participation in an OECD activity on uncertainity quantification and sensitivity analysis in nuclear fuel modeling and 7) major improvements to BISON’s fission gas behavior models. A final chapter outlines FY-17 future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

FUEL PERFORMANCE MODELING OF FECRAL CLADDING FAILURE DURING LOSS OF COOLANT CONDITIONS

Accident tolerant materials are defined as those that provide significantly increased response time in the event of an accident while providing similar or improved performance as the conventional UO2/Zircaloy-4 fuel rods during normal operation. One avenue for materials to demonstrate enhanced accident tolerance is to have improved reaction (e.g., oxidation) kinetics with steam, resulting in slower hydrogen (or other combustible gases) generation rate, while maintaining acceptable thermo-mechanical properties, fuel-clad interactions, and fission product behavior. Due to their improved reaction kinetics with steam compared to Zircaloy-4, iron-chromium-aluminum (FeCrAl) alloys are being considered as an accident tolerant cladding material. In order to determine whether FeCrAl alloys are indeed accident tolerant materials, comparisons between conventional zirconium-based claddings under accident conditions is required. To this end, we have developed a burst (failure) criterion for FeCrAl using existing experimental data for first-generation alloys. This model was incorporated into the BISON fuel performance code to perform comparative studies with Zircaloy-4 cladding tubes under non-oxidizing and oxidizing LOCA conditions. The separate effects LOCA conditions considered were adopted from the PUZRY and REBEKA test series. Due to the fact that FeCrAl alloys have a larger thermal absorption cross-section and higher stiffness than Zircaloy-4, cladding thickness will be smaller for FeCrAl than for Zircaloy-4. Three different cladding thicknesses were analyzed in this work. The results indicate that for both non-oxidizing (PUZRY) and oxidizing (REBEKA) LOCA conditions, FeCrAl has similar burst behavior (i.e., time to burst and pressure at burst) as Zircaloy-4. It should be noticed that this study used data for first-generation FeCrAl alloys to develop the burst criterion. Although no burst data?currently exists for C35M (a second-generation FeCrAl alloy under development at Oak Ridge National Laboratory), improved burst behavior is anticipated. As burst data for C35M becomes available the burst criterion will be updated and the simulations revisited to determine the accident tolerance of second-generation FeCrAl alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

3D modeling of missing pellet surface defects in BWR fuel

One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can be used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed in this paper. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Finally, parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding adjacent to the defect.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Analysis of a Loop for Fission Gas Release and Monitoring at the PULSTAR Reactor

Modern nuclear fuel performance analysis tools aim at establishing a predictive approach for examining macroscopic fuel behavior due to fuel gaseous swelling and the associated pellet cladding mechanical interaction (PCMI). In general, the macroscopic behavior is highly sensitive to understanding microscopically driven phenomena, which in this case is the release of fission gas (i.e., Kr and Xe) from the fuel. Previous studies identified the uncertainties associated with the intra-granular fission gas diffusion coefficient as most limiting for the accuracy of fission gas release calculations. In particular, it was concluded that, in order to improve fission gas release as well as gaseous swelling predictions using fuel performance codes, a better characterization of the diffusion coefficient is needed. This work explores the implementation of an on-line fission gas release and measurement loop at the North Carolina State University (NCSU) PULSTAR reactor. A unique characteristic of this loop is the ability to control the temperature of the fuel samples during irradiation to establish a more accurate correlation between the irradiation conditions and an observable such as the intragranular gas diffusion coefficient. Analysis was performed of an 'in-pile' fission gas release experiment that is currently under implementation at the NCSU PULSTAR reactor. Using MCNP6 calculations and COMSOL simulations the fission rate density and temperature distributions in UO{sub 2} microspheres were estimated. To explore the feasibility of detecting fission gas isotopes, estimates of R/B for {sup 138}Xe were obtained using the equivalent sphere model and combined with MCNP6 predicted B estimates to obtain R. Assuming detection in a typical gamma-ray spectrometry HPGe system, it was conclude that reasonable count rates will be achieved to allow extraction of release information. At this stage, accurate assessment is underway of the performance of various loop components to set the final design criteria in support of loop fabrication. (authors)

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DOE OSTI2016

Validating the BISON fuel performance code to integral LWR experiments

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Our results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. In the initial rod diameter comparisons they were unsatisfactory and have lead to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Isospin transport and reaction mechanism in nuclear reactions in the range 20–40 MeV/n

In recent years, many efforts have been devoted to the investigation of the isospin degree of freedom in nuclear reactions. Comparing systems involving partners with different N/Z, it has been possible to investigate the isospin transport process and its influence on the final products population. This can be then related to the symmetry energy term of the nuclear EOS. From the experimental point of view, this task requires detectors able to measure both charge and mass of the emitted products, in the widest possible range of energy and size of the fragments. With this objective, the FAZIA and GARFIELD+RCo apparatus have been used with success in some recent experiments.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2015

Advanced multiphysics coupling for LWR fuel performance analysis

Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics, particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is possible to use lower length scale models such as those used in the mesoscale MARMOT code to compute average properties, e.g. swelling or thermal conductivity. These may then be used by an engineering-scale model. Examples of this type of multiscale, multiphysics modeling are shown.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Summary of BISON Development and Validation Activities - NEAMS FY15 Report

The objective of the NEAMS ToolKit is to develop a “pellet-to-plant” simulation capability useful for predicting performance and safety for a broad range of nuclear reactor power systems. The NEAMS ToolKit has been organized into a Fuels Product Line (FPL) and a Reactor Product Line (RPL) and is modular in design. Within the FPL, a multiscale approach has been adopted in which simulations of fuel performance at the engineering scale are informed by material property and irradiation performance models developed from mesoscale simulations of microstructural evolution. The focus in this report is on development and validation of the engineering-scale fuel performance analysis tool within the FPL, which is BISON. This summary report contains an overview of work performed under the work package enti- tled “FY2015 NEAMS INL-Engineering Scale Fuel Performance & Interface with RPL Tools.” A first chapter identifies the specific FY-15 milestones, providing a basic description of the asso- ciated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes major additional accomplishments including 1) a BISON validation paper, 2) a multiphysics coupling paper, 3) application of BI- SON to Reactivity Insertion Accident analysis and 4) use of BISON for uncertainity analysis in fuel modeling. A final chapter outlines FY-16 future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Improving the Accuracy of PCMI Simulations with More Realistic Geometry and Material Models

Assessment of fuel performance codes is an endeavour that requires re-evaluation as new simulation capabilities become available. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, has been used to simulate a variety of experiments. Calculations from these simulations are compared to experimental measurements for the purpose of evaluating the progress of code development. One such experiment, from the Risø 3 Fission Gas Project, provides measurements of fission gas release and clad diameter. A subset of the Risø3 Fission Gas Project, which is referred to here, consisted of a base-irradiation period of a full-length fuel rod in the Quad-Cities-1 BWR. The base irradiation test was followed by a “bump test”, in which a sub-section of the original rod was subjected to a strongly non-uniform axial power profile. The bump test took place in Risø test reactor DR3 in a water-cooled HP1 rig under BWR conditions. This experiment was chosen because it allows for an evaluation of several aspects of BISON, including fully coupled thermo-mechanics, contact, and several nonlinear material models. A summary of simulation results and comparison to the experiment data has been published previously, however, this paper shows new results with recently added features to BISON. These new features include representing the pellets as discrete entities rather than a smeared column, including primary creep in the clad creep model, and the coupling of fuel gaseous swelling to fission gas release. Comparing to previous results, the addition of theses features has narrowed the gap between BISON calculations and experimental measurements for final clad diameter and fission gas release.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2015

MODELLING OF FUEL BEHAVIOUR DURING LOSS-OF-COOLANT ACCIDENTS USING THE BISON CODE

This work presents recent developments to extend the BISON code to enable fuel performance analysis during LOCAs. This newly developed capability accounts for the main physical phenomena involved, as well as the interactions among them and with the global fuel rod thermo-mechanical analysis. Specifically, new multiphysics models are incorporated in the code to describe (1) transient fission gas behaviour, (2) rapid steam-cladding oxidation, (3) Zircaloy solid-solid phase transition, (4) hydrogen generation and transport through the cladding, and (5) Zircaloy high-temperature non-linear mechanical behaviour and failure. Basic model characteristics are described, and a demonstration BISON analysis of a LWR fuel rod undergoing a LOCA accident is presented. Also, as a first step of validation, the code with the new capability is applied to the simulation of experiments investigating cladding behaviour under LOCA conditions. The comparison of the results with the available experimental data of cladding failure due to burst is presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Summary of BISON Development Activities: NEAMS FY14 Report

This summary report contains an overview of work performed under the work package entitled “FY2014 NEAMS INL-Engineering Scale Fuel Performance & Interface with RPL Tools.” A first chapter identifies the specific FY-14 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes substantial additional work including 1) efforts to improve numerical convergence and contact in BISON, 2) development of capability to simulate hydrogen behavior in Zircaloy cladding and 3) efforts to enhance collaborative work with the Halden Research Program. A final chapter briefly outlines planned future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Verification as a Foundation for Validation of a Nuclear Fuel Performance Code

Complex multiphysics simulations such as nuclear fuel performance analysis are composed of many submodels used to describe specific phenomena. These phenomena include, as examples, the relationship between stress and strain, heat transfer across a gas gap, and mechanical contact. These submodels work in concert to simulate real-world events, like the behavior of a fuel rod in a reactor. If a simulation tool is able to represent real-world behavior, the tool is said to be validated. While much emphasis is rightly placed on validation, model verification may be undervalued. Verification involves showing that a model performs as intended, that it computes results consistent with its mathematical description. This paper explains the differences between verification and validation and shows how validation should be preceded by verification. Specific verification problems, including several specific to nuclear fuel analysis, are given. Validation results are also presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Verification of the BISON fuel performance code

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Labo- ratory (USA) since 2009. The code is applicable to both steady and transient fuel behavior and is used to analyze 1D spherical, 2D axisymmetric, or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods and other well known fuel performance codes. Results from several assessment cases are reported, with emphasis on fuel centerline temperatures at various stages of fuel life, fission gas release, and clad deformation during pellet clad mechanical interaction (PCMI). BISON comparisons to fuel centerline temperature measurements are very good at beginning of life and reasonable at high burnup. Although limited to date, fission gas release comparisons are very good. Comparisons of rod diameter following significant power ramping are also good and demonstrate BISON’s unique ability to model discrete pellet behavior and accurately predict clad ridging from PCMI.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2014

BISON Software V&V Plan

The primary vision for the BISON development team is to deliver a nuclear fuel performance simulation tool that is used to provide a researcher or fuel designer with best estimate calculations of the highly coupled and nonlinear phenomena that govern nuclear fuel behavior. Accurately simulating nuclear fuel behavior is a challenging computational undertaking and verification and validation (V&V) play an important role in realizing this vision. The purpose of this V&V plan is to express the BISON team’s definition of the terms verification and validation, document what we have done regarding V&V, and outline what we plan to do.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗