Evaluation of Reconfiguration and Damage of BWR Spent Fuel During Storage and Transportation Accidents


Structural Integrity Associates is participating in a Department of Energy (DOE) Integrated Research Projects (IRP) program focused on storage and transportation of used nuclear fuel (UNF). The project, entitled Cask Mis-Loads Evaluation Techniques, was awarded to a university-based research team in 2016 under the DOE Nuclear Fuels Storage and Transportation (NFST) project. The team is led by the University of Houston (U of H) and includes representatives from the University  of Illinois at Urbana-Champaign, the University of Southern California, the University of Minnesota, Pacific Northwest National Laboratory, and staff members from the Nuclear Fuel Technology and Critical Structures and Facilities divisions of SI. The primary objectives of NFST are to 1) implement interim storage, 2) improve integration of storage into an overall waste management system, and 3) prepare for large-scale transportation of UNF and high-level waste.  The goal of the cask mis-load project is to develop a probabilistically informed methodology, utilizing innovative non-destructive evaluation (NDE) techniques, determining the extent of potential damage or degradation of internal components of UNF canisters/casks during normal conditions of transport (NCT) and hypothetical accident conditions (HAC).

FIGURE 1. Cask, basket, and fuel assembly FE models for global dynamics analysis [derived from Ref.1]

An integral part of this methodology uses thermal-mechanical simulations to determine the fuel-assembly/fuel-rod response and subsequent reconfiguration for NCT and HAC. These analyses are comprised of two primary steps 1) global modeling and explicit-dynamics structural analysis, and 2) local fuel performance modeling and failure analysis. The results of these analyses will yield data defining the global forces acting on the fuel rods, spacer grid, and assembly distortions and deformation, and fuel rod failure probability/potential. Our team has conducted extensive work on Pressurized Water Reactor (PWR) fuel during past research sponsored by EPRI [1, 2, 3]. The focus of this new research effort will be high burnup Boiling Water Reactor (BWR) fuel and associated transportation casks.

The four primary objectives for the thermo-mechanical simulation work under the cask mis-load project are the determination of:

  • Global forces acting on the components of the fuel assemblies
  • Fuel rod, spacer grid, and overall assembly distortions and deformation under the defined conditions
  • Failure probability/potential of fuel rods under the defined conditions
  • The most likely fuel reconfiguration events

This data will ultimately provide feedback for the calibration and validation process linking the mock-up testing and NDE phases of the overall cask mis-load project methodology.

FIGURE 2. Integration of a fuel assembly model into the cask and basket model

To date, much work has been completed on the finite element (FE) model analyses needed for the cask level dynamics analyses. This model is based on design information for the Hi-STAR 100 cask, MPS-68 canister, basket, and surrogate GE-14 fuel assembly and rod models (the specific GE-14 design is proprietary). The FE model implementation is based on our prior work using 3D modeling for PWR fuel, assembly, and cask designs. Figure 1 provides several 3D views of prior cask, basket and assembly models developed for PWR fuel types. The BWR-based models are derived along the same path but incorporating the specific design aspects for the components noted above. The surrogate BWR fuel assembly model is integrated into the cask and basket model, as shown in Figure 2. As an example, Figure 3 shows the computed response of the fuel assembly and basket models at the apex of the 9-meter drop analysis.

FIGURE 3. Fuel assembly and basket displacement, deformation, and stress computed for the 9-meter drop analysis

As noted above, the global forces acting on the components of the fuel assemblies as determined with the dynamics analyses are transformed to determine the force and stresses exerted on individual fuel rods within the assembly. Single fuel rod FE model analyses are conducted to evaluate the distribution of stresses and deformation on the cladding. The preliminary results of these analyses are shown in Figure 4 representing the impingement of an external load, for example, at a grid spacer location, on the cladding surface.

Upcoming work will focus on the development of a Zr-2 radial hydride damage model which will be used to relate the effects of the evolution and concentration of radial hydrides during dry storage to the failure probability under the stress conditions defined by the global dynamics analyses. Under normal irradiation conditions, hydrides are formed within the cladding material as a result of the diffusion of hydrogen released from the

FIGURE 4. Deformation and stress from external impingement on a single fuel rod during the 9-meter drop analysis.

corrosion and oxidation of the exterior of the fuel rod. These hydrides generally form circumferentially in decreasing concentration from the exterior to the interior of the cladding (Figure 5). During the drying process, as spent fuel is prepared for encapsulation in dry storage casks, the orientation of the hydrides can change from circumferential to radial resulting in a mixed hydride structure such as shown in Figure 6 (the cladding material in these samples was artificially charged with hydrogen for testing). Hydride reorientation occurs as a result of the increased cladding stress from the higher internal pressure experienced by the fuel rod as the temperature rises (up to a maximum of 400 °C) during the drying process.

The presence of radial hydrides has a dramatic effect on the mechanical response of the cladding under load and can significantly increase the probability of cladding failure at much lower cladding strains. Ultimately, the Zr-2 radial hydride damage model will be used to inform the local fuel rod analysis and provide a more accurate determination of the probability of fuel rod failure resulting from the local impingement forces experienced during NCT and HAC.  This work is new and significant in that it is the first application of this analysis methodology for a BWR fuel design and represents a substantial advancement in the assessment of BWR UNF during normal and hypothetical transportation conditions.

FIGURE 5. Hydride structure and outer surface oxide for high burnup cladding with an average hydrogen concentration of 600 ppm [3]

FIGURE 6. Hydride structure of cladding samples prepared for mechanical testing: (a) circumferential versus (b) mixed circumferential + radial hydrides [3]




Contact Form