Measurement of Aerosol Transmission Through a Stress Corrosion Crack–Like Geometry

Abstract

The formation of a stress corrosion crack (SCC) in the canister wall of a dry cask storage system (DCSS) has been identified as a potential issue for the long-term storage of spent nuclear fuel. The presence of a SCC in a storage system could represent a through-wall flow path from the canister interior to the environment. Modern, vertical DCSSs are of particular interest due to the commercial practice of using relatively high helium backfill pressures (up to approximately 800 kPa) in the canister to enhance internal natural convection. This pressure differential offers a comparatively high driving potential for blowdown of any particulates that might be present in the canister in the event of a through-wall SCC.

In this study, the rates of gas flow and aerosol transmission of a spent fuel surrogate through an engineered microchannel with dimensions representative of a SCC were evaluated experimentally using coupled mass flow and aerosol analyzers. The microchannel was formed by mating two gauge blocks with a linearly tapering slot orifice nominally 13 μm (0.0005 in.) tall on the upstream side and 25 μm (0.001 in.) tall on the downstream side. The orifice is 12.7 mm (0.500 in.) wide by 8.89 mm (0.350 in.) long (flow length). Surrogate aerosols of cerium oxide (CeO₂) were seeded and mixed with either helium or air inside a pressurized tank. The aerosol characteristics were measured immediately upstream and downstream of the simulated SCC at elevated and ambient pressures, respectively.

The next iteration of testing involves replacing the engineered microchannel with lab-grown SCCs. Preliminary clean flow testing has been conducted on SCC samples provided by the Electric Power Research Institute. These data sets demonstrate a new capability to characterize SCCs under well-controlled boundary conditions. Preliminary testing efforts are focused on understanding the evolution in both the size and quantity of a hypothetical release of aerosolized spent fuel particles from failed fuel cladding into the canister interior, and ultimately, through a SCC.

Keywords:

• Spent nuclear fuel
• dry storage
• stress corrosion crack
• aerosol transmission

Acknowledgments

The authors would like to express their appreciation to Thad Vice, Beau Baigas, Ron Williams, and Greg Koenig for their assistance in performing these tests. Parallel modeling efforts by Yadu Sasikumar at Oak Ridge National Laboratory, Andrew Casella at Pacific Northwest National Laboratory, and Jessie Phillips at SNL are expected to aid in the interpretation of these tests and are eagerly anticipated.

Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s (DOE’s) National Nuclear Security Administration under contract DE-NA0003525.

This paper describes objective technical results and analysis. Any subjective views or opinions that might be expressed in the paper do not necessarily represent the views of the DOE or the U.S. government.

This paper has been authored by an employee of National Technology & Engineering Solutions of Sandia, LLC under contract DE-NA0003525 with the DOE. The employees own all rights, title, and interest in and to the paper and are solely responsible for its contents. The U.S. government retains and the publisher, by accepting the paper for publication, acknowledges that the U.S. government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this paper or allow others to do so for U.S. government purposes. The DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan https://www.energy.gov/downloads/doe-public-access-plan.

Disclosure Statement

No potential conflict of interest was reported by the author(s).

Additional information

Funding

This work was funded by the DOE, Office of Nuclear Energy Spent Fuel and Waste Disposition Research and Development Program.