Published October 1989 | Version v1
Report

Melt Attack and Coolability Experiments (MACE) program elements and the scoping test

  • 1. Electric Power Research Institute, Palo Alto, CA (USA)
  • 2. Argonne National Lab., IL (USA)

Description

The most important issue in the ex-vessel progression of the severe accidents is whether sufficient energy (heat) can be removed from the discharged melt that the threat to containment integrity is avoided. This issue is pertinent for both the initial and the longer term (hours) interactions of the discharged melt with the BWR and PWR containments. For example, if the superheat of the melt discharged in the Mark-I drywell is removed during the initial interaction, the spreading of the melt could be stopped and the melt attack on the liner precluded. Similarly, for high pressure melt ejection (HPME) in a PWR containment cavity, effective heat removal in the initial melt-structure and/or melt-coolant interactions greatly reduces the direct containment heating loads. The initial/short-term coolability of the melt discharged from the vessel depends upon the containment configuration and the placement of water. Since the initial/short-term coolability of the melt in the PWR and BWR containments is so very much initial-condition dependent and plant specific, it is not the subject of the Melt Attack and Coolability Experiments (MACE) program. The MACE program addresses the generic question of the coolability of the discharged melt spread out to a part of the basemat floor under the reactor vessel. This situation results after the initial/short-term interactions of the melt with water and/or structures found in the BWR and PWR containments. The key issue then is whether the melt, if flooded with water, will quench and remain coolable, so that molten corium concrete interaction (MCCI) is avoided and containment integrity maintained

Additional details

Publishing Information

Imprint Title
Transactions of the seventeenth water reactor safety information meeting
Imprint Pagination
186 p.
Journal Page Range
p. 15.9-15.10.
Report number
NUREG/CP--0104

Conference

Title
17. water reactor safety information meeting.
Dates
23-25 Oct 1989.
Place
Rockville, MD (USA).

Optional Information

Secondary number(s)
CONF-8910222--Summs.