FOREVER Experiments on Thermal and Mechanical Behavior of a Reactor Pressure Vessel during a Severe Accident
- 1. Division of Nuclear Power Safety, Royal Institute of Technology, Brinellvagen 60, 10044 Stockholm (Sweden)
Description
This paper describes the FOREVER (Failure Of Reactor Vessel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The facility employs 1/10.-scale carbon steel vessels of 0.4 m diameter, 15 mm thickness and 600 mm height. Up to 20 liters of binary-oxide melts with 100-300 K superheat are employed, as a simulant for the prototypic corium melt, and internal heating is provided by electrical heaters of up to 20 kW power in order to maintain the vessel wall temperatures at 1100-1200 K. Auxiliary systems are designed to provide an overpressure up to 4 MPa in the test vessel. Thus, severe accident scenarios with RCS depressurization are modeled. Creep behavior of the three-dimensional vessel, formation of the gap between the melt pool crust and the creeping vessel, and mechanisms of the gap cooling by water ingression will be the subjects of study and measurements in the FOREVER experimental program. Scaling rationale as well as pre-test analyses of the thermal and mechanical behavior of the FOREVER test vessels are presented. (authors)
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Additional details
Publishing Information
- Imprint Title
- Proceedings of the Workshop on in-vessel core debris retention and coolability
- Imprint Pagination
- 409 p.
- Journal Page Range
- p. 252-260
- Report number
- NEA-CSNI-R--1998-18
Conference
- Title
- Workshop on in-vessel core debris retention and coolability
- Dates
- 3-6 Mar 1998
- Place
- Garching (Germany)
INIS
- Country of Publication
- Nuclear Energy Agency of the OECD (NEA)
- Country of Input or Organization
- Nuclear Energy Agency of the OECD (NEA)
- INIS RN
- 44026304
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- A CODES; ACCIDENT MANAGEMENT; CONTAINMENT SYSTEMS; COOLING; CORE FLOODING SYSTEMS; CORIUM; CREEP; FAILURES; HEAT FLUX; HEAT TRANSFER; M CODES; MECHANICAL PROPERTIES; MELTING; REACTOR ACCIDENTS; REACTOR CORES; REACTOR EXPERIMENTAL FACILITIES; REACTOR VESSELS; SCALING LAWS; TEMPERATURE DEPENDENCE; THERMAL HYDRAULICS
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; CONTAINERS; CONTAINMENT; ECCS; ENERGY TRANSFER; ENGINEERED SAFETY SYSTEMS; FLUID MECHANICS; HYDRAULICS; MANAGEMENT; MECHANICAL PROPERTIES; MECHANICS; PHASE TRANSFORMATIONS; REACTOR COMPONENTS; REACTOR PROTECTION SYSTEMS
Optional Information
- Notes
- 6 refs.