Characterization of Thermal-Hydraulic and Ignition Phenomena in Prototypic, Full-Length Boiling Water Reactor Spent Fuel Pool Assemblies After a Postulated Complete Loss-of-Coolant Accident (NUREG/CR-7143, SAND-2007-2270)

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Publication Information

Manuscript Completed: October 2012
Date Published: March 2013

Prepared by:
E. R. Lindgren and S. G. Durbin

Sandia National Laboratory
Albuquerque, NM 87185

G. A. Zigh, Technical Advisor
A. Velazquez-Lozada, Project Manager

NRC Job Code Y6758

Prepared for:
Division of Systems Analysis
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington DC 20555-0001

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The NRC regulates the operation of the civilian nuclear power plant fleet by establishing and enforcing regulatory requirements for their design, construction, and operation. To protect the health and safety of the public and environment, the NRC requires all nuclear power plants to have a spent fuel pool where the used reactor fuel assemblies are allowed to cool for a number of years before being moved to interim or permanent storage. Spent fuel pools (SFP) are robust structures with an extremely low likelihood of a complete loss of coolant under traditional accident scenarios. However, in the wake of the terrorist attacks of September 11, 2001, the SFP accident progression was reconsidered and reevaluated using best-estimate accident codes.

In 2001, the NRC staff performed an evaluation of the potential accident risk in a SFP at decommissioning plants in the United States. This evaluation is documented on NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," (ML010430066). The study described a modeling approach of a typical decommissioning plant with design assumptions and industry commitments, thermal-hydraulic analyses performed to evaluate spent fuel stored in the SFP at decommissioning plants, risk assessment of SFP accidents, consequence calculations and the implications for decommissioning regulatory requirements. Some assumptions in the accident progression were known to be necessarily conservative, especially the estimation of the fuel damage. Consequently, the NRC continued SFP accident research by applying best-estimate computer codes to predict the severe accident progression following various postulated accident initiators. These code studies identified various modeling and phenomenological uncertainties that prompted a need for experimental confirmation. The present experimental program was undertaken to address thermal-hydraulic issues associated with complete loss-of-coolant accidents in boiling water reactor SFPs. All of the experiments and numerical simulations described in this report were performed at Sandia National Laboratories (SNL) in Albuquerque, New Mexico.

The objective of this project was to provide basic thermal-hydraulic data associated with a postulated SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce modeling uncertainties within the code.

The research summarized in this report achieved the stated objective and resolved several unexpected technical challenges related with thermocouple attachments and the choice of appropriate input power in the heated design test that would cause the ignition. The close coupling of the experimental and numerical programs allowed for rapid validation and improvement of the MELCOR whole pool calculations. Because of the success of this approach, this project will be used as a model for subsequent studies.

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