A scaling and experimental approach for investigating in-vessel cooling
- Fauske & Associates, Inc., Burr Ridge, IL (United States)
The TMI-2 accident experienced the relocation of a large quantity of core material to the lower plenum. The TMI-2 vessel investigation project concluded that approximately 20 metric tonnes of once molten fuel material drained into the RPV lower head. As a result, the lower head wall experienced a thermal transient that has been characterized as reaching temperatures as high as 1100{degrees}C, then a cooling transient with a rate of 10 to 100{degrees}C/min. Two mechanisms have been proposed as possible explanations for this cooling behavior. One is the ingression of water through core material as a result of interconnected cracks in the frozen debris and/or water ingression around the crust which is formed on internal structures (core supports and in-core instrumentation) in the lower head. The second focuses on the lack of adhesion of oxidic core debris to the RPV wall when the debris contacts the wall. Furthermore, the potential for strain of the RPV lower head when the wall is overheated could provide for a significant cooling path for water to ingress between the RPV and the frozen core material next to the wall. To examine these proposed mechanisms, a set of scaled experiments have been developed to examine the potential for cooling. These are performed in a scaled system in which the high temperature molten material is iron termite and the RPV wall is carbon steel. A termite mass of 40 kg is used and the simulated reactor vessels have water in the lower head at pressures up to 2.2 MPa. Furthermore, two different thicknesses of the vessel wall are examined with the thicker vessel having virtually no potential for material creep during the experiment and the thinner wall having the potential for substantial creep. Moreover, the experiment includes the option of having molten iron as the first material to drain into the RPV lower head or molten aluminum oxide being the only material that drains into the test configuration.
- Research Organization:
- US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Office of Nuclear Regulatory Research; Brookhaven National Lab. (BNL), Upton, NY (United States)
- OSTI ID:
- 467943
- Report Number(s):
- NUREG/CP-0157-VOL.2; CONF-9610202-Vol.2; ON: TI97004274; TRN: 97:008409
- Resource Relation:
- Conference: 24. water reactor safety information meeting, Bethesda, MD (United States), 21-23 Oct 1996; Other Information: PBD: Feb 1997; Related Information: Is Part Of Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity; Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)]; PB: 443 p.
- Country of Publication:
- United States
- Language:
- English
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