Problem Specification for FY12 Modeling of UNF During Extended Storage

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the Advanced Modeling and Simulation Office (AMSO) of the US Department of Energy, Office of Nuclear Energy (DOE/NE) has invested in the initial extension and application of advanced nuclear simulation tools to address relevant needs in evaluating the performance of used nuclear fuel (UNF) during extended periods of dry storage. There are many significant challenges associated with the prediction of the behavior of used fuel during extended periods of dry storage and subsequent transportation. The initial activities are focused on integrating with the Used Fuel Disposition (UFD) Campaign of the DOE/NE and a demonstration that the Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) for modeling the mechanical state of the cladding after decades of storage. This initial focus will model the long-term storage of the UNF and account for the effect, and generation, of radially and circumferentially oriented hydride precipitates within the cladding and predict the end of storage (EOS) mechanical state (stress, strain) of the cladding. Predicting the EOS state of the cladding is significant because it (1) provides an estimate of the margin to failure of the cladding during nominal storage operation and it (2) establishes the initial state of the fuel for post-storage transportation. Because there are significant uncertainties associated with the storage conditions, hydride precipitate formation, and the beginning of storage (BOS) condition of the UNF, this will also allow for the development of a rigorous capability to evaluate the relative sensitivities of the uncertainties and can help to guide the experimental and analysis efforts of the UFD Campaign. This document is focused on specifying the problem that will be solved with AMPFuel. An associated report, documents the specifics of the constitutive model that will be developed and implemented in AMPFuel to account for the presence and predict the generation of the hydride precipitates. This report satisfies the deliverable for the DOE Office of Nuclear Energy, Advanced Modeling and Simulation Office, milestone M3MS-12OR0605083, 'Definition of Problem Specification,' which defines the problem to be solved that will satisfy milestone M2MS-12OR0605081, 'Demonstration of the Advanced Multi-Physics (AMP) Nuclear Fuel Performance code for modeling UFD.' This document should provide sufficient detail to model a high burnup pressurized water reactor (PWR) fuel rod to provide an estimate of the end of storage (EOS) mechanical state of the cladding. The fuel rod and irradiation history are based on seven cycles of irradiation in the CP and L H.B. Robinson nuclear reactor, which achieved a discharge burnup of 66.682 MWd/kgU. The fuel has been experimentally examined for storage conditions by Argonne National Laboratory for the NRC. In addition, we have compiled a list of key factors that have been shown to strongly influence the EOS state of the fuel and have identified baseline values and ranges of uncertainties that will be considered. The simulations that will be performed have been described in detail and include the modeling assumptions and boundary conditions

Software Design Document for the AMP Nuclear Fuel Performance Code

The purpose of this document is to describe the design of the AMP nuclear fuel performance code. It provides an overview of the decomposition into separable components, an overview of what those components will do, and the strategic basis for the design. The primary components of a computational physics code include a user interface, physics packages, material properties, mathematics solvers, and computational infrastructure. Some capability from established off-the-shelf (OTS) packages will be leveraged in the development of AMP, but the primary physics components will be entirely new. The material properties required by these physics operators include many highly non-linear properties, which will be replicated from FRAPCON and LIFE where applicable, as well as some computationally-intensive operations, such as gap conductance, which depends upon the plenum pressure. Because there is extensive capability in off-the-shelf leadership class computational solvers, AMP will leverage the Trilinos, PETSc, and SUNDIALS packages. The computational infrastructure includes a build system, mesh database, and other building blocks of a computational physics package. The user interface will be developed through a collaborative effort with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Capability Transfer program element as much as possible and will be discussed in detail in a future document

Analysis of the IFA-432, IFA-597, and IFA-597 MOX Fuel Performance Experiments by FRAPCON-3.4

Validation of advanced nuclear fuel modeling tools requires careful comparison with reliable experimental benchmark data. A comparison to industry-accepted codes, that are well characterized, and regulatory codes is also a useful evaluation tool. In this report, an independent validation of the FRAPCON-3.4 fuel performance code is conducted with respect to three experimental benchmarks, IFA-432, IFA-597, and IFA-597mox. FRAPCON was found to most accurately model the mox rods, to within 2% of the experimental data, depending on the simulation parameters. The IFA-432 and IFA-597 rods were modeled with FRAPCON predicting centerline temperatures different, on average, by 21 percent

Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162 billion degrees of freedom for 10 loading steps. The single radiation transport calculation required about 50% of the time required to solve the thermo-mechanics with a single loading step, which demonstrates that it is feasible to incorporate, in a single code, a high-fidelity radiation transport capability with a high-fidelity nuclear fuel thermo-mechanics capability and anticipate acceptable computational requirements. The results of the full assembly simulation clearly show the axial, radial, and azimuthal variation of the neutron flux, power, temperature, and deformation of the assembly, highlighting behavior that is neglected in traditional axisymmetric fuel performance codes that do not account for assembly features, such as guide tubes and control rods

C++ Coding Standards for the AMP Project

This document provides an initial starting point to define the C++ coding standards used by the AMP nuclear fuel performance integrated code project and a part of AMP's software development process. This document draws from the experiences, and documentation [1], of the developers of the Marmot Project at Los Alamos National Laboratory. Much of the software in AMP will be written in C++. The power of C++ can be abused easily, resulting in code that is difficult to understand and maintain. This document gives the practices that should be followed on the AMP project for all new code that is written. The intent is not to be onerous but to ensure that the code can be readily understood by the entire code team and serve as a basis for collectively defining a set of coding standards for use in future development efforts. At the end of the AMP development in fiscal year (FY) 2010, all developers will have experience with the benefits, restrictions, and limitations of the standards described and will collectively define a set of standards for future software development. External libraries that AMP uses do not have to meet these requirements, although we encourage external developers to follow these practices. For any code of which AMP takes ownership, the project will decide on any changes on a case-by-case basis. The practices that we are using in the AMP project have been in use in the Denovo project [2] for several years. The practices build on those given in References [3-5]; the practices given in these references should also be followed. Some of the practices given in this document can also be found in [6]

Problem Specification for FY12 Modeling of UNF During Extended Storage

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the Advanced Modeling and Simulation Office (AMSO) of the US Department of Energy, Office of Nuclear Energy (DOE/NE) has invested in the initial extension and application of advanced nuclear simulation tools to address relevant needs in evaluating the performance of used nuclear fuel (UNF) during extended periods of dry storage. There are many significant challenges associated with the prediction of the behavior of used fuel during extended periods of dry storage and subsequent transportation. The initial activities are focused on integrating with the Used Fuel Disposition (UFD) Campaign of the DOE/NE and a demonstration that the Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) for modeling the mechanical state of the cladding after decades of storage. This initial focus will model the long-term storage of the UNF and account for the effect, and generation, of radially and circumferentially oriented hydride precipitates within the cladding and predict the end of storage (EOS) mechanical state (stress, strain) of the cladding. Predicting the EOS state of the cladding is significant because it (1) provides an estimate of the margin to failure of the cladding during nominal storage operation and it (2) establishes the initial state of the fuel for post-storage transportation. Because there are significant uncertainties associated with the storage conditions, hydride precipitate formation, and the beginning of storage (BOS) condition of the UNF, this will also allow for the development of a rigorous capability to evaluate the relative sensitivities of the uncertainties and can help to guide the experimental and analysis efforts of the UFD Campaign. This document is focused on specifying the problem that will be solved with AMPFuel. An associated report, documents the specifics of the constitutive model that will be developed and implemented in AMPFuel to account for the presence and predict the generation of the hydride precipitates. This report satisfies the deliverable for the DOE Office of Nuclear Energy, Advanced Modeling and Simulation Office, milestone M3MS-12OR0605083, 'Definition of Problem Specification,' which defines the problem to be solved that will satisfy milestone M2MS-12OR0605081, 'Demonstration of the Advanced Multi-Physics (AMP) Nuclear Fuel Performance code for modeling UFD.' This document should provide sufficient detail to model a high burnup pressurized water reactor (PWR) fuel rod to provide an estimate of the end of storage (EOS) mechanical state of the cladding. The fuel rod and irradiation history are based on seven cycles of irradiation in the CP and L H.B. Robinson nuclear reactor, which achieved a discharge burnup of 66.682 MWd/kgU. The fuel has been experimentally examined for storage conditions by Argonne National Laboratory for the NRC. In addition, we have compiled a list of key factors that have been shown to strongly influence the EOS state of the fuel and have identified baseline values and ranges of uncertainties that will be considered. The simulations that will be performed have been described in detail and include the modeling assumptions and boundary conditions

Roadmap to an Engineering-Scale Nuclear Fuel Performance & Safety Code

Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied more on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO{sub 2} with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel-development activities. Realizing the full benefits of this approach will likely take some time. However, it is important that the developmental activities for modeling and simulation be tightly coupled with the experimental activities to maximize feedback effects and accelerate both the experimental and analytical elements of the program toward a common objective. The close integration of modeling and simulation and experimental activities is key to developing a useful fuel performance simulation capability, providing a validated design and analysis tool, and understanding the uncertainties within the models and design process. The efforts of this project are integrally connected to the Transmutation Fuels Campaign (TFC), which maintains as a primary objective to formulate, fabricate, and qualify a transuranic-based fuel with added minor actinides for use in future fast reactors. Additional details of the TFC scope can be found in the Transmutation Fuels Campaign Execution Plan. This project is an integral component of the TFC modeling and simulation effort, and this multiyear plan borrowed liberally from the Transmutation Fuels Campaign Modeling and Simulation Roadmap. This document provides the multiyear staged development plan to develop a continuum-level Integrated Performance and Safety Code (IPSC) to predict the behavior of the fuel and cladding during normal reactor operations and anticipated transients up to the point of clad breach

Analysis of the IFA-432, IFA-597, and IFA-597 MOX Fuel Performance Experiments by FRAPCON-3.4

Validation of advanced nuclear fuel modeling tools requires careful comparison with reliable experimental benchmark data. A comparison to industry-accepted codes, that are well characterized, and regulatory codes is also a useful evaluation tool. In this report, an independent validation of the FRAPCON-3.4 fuel performance code is conducted with respect to three experimental benchmarks, IFA-432, IFA-597, and IFA-597mox. FRAPCON was found to most accurately model the mox rods, to within 2% of the experimental data, depending on the simulation parameters. The IFA-432 and IFA-597 rods were modeled with FRAPCON predicting centerline temperatures different, on average, by 21 percent

Analysis of the IFA-432, IFA-597, and IFA-597 MOX Fuel Performance Experiments by FRAPCON-3.4

Validation of advanced nuclear fuel modeling tools requires careful comparison with reliable experimental benchmark data. A comparison to industry-accepted codes, that are well characterized, and regulatory codes is also a useful evaluation tool. In this report, an independent validation of the FRAPCON-3.4 fuel performance code is conducted with respect to three experimental benchmarks, IFA-432, IFA-597, and IFA-597mox. FRAPCON was found to most accurately model the mox rods, to within 2% of the experimental data, depending on the simulation parameters. The IFA-432 and IFA-597 rods were modeled with FRAPCON predicting centerline temperatures different, on average, by 21 percent