Lessons on in-vessel severe accidents from experiments at KfK and the INEL

Conference · Fri Jan 01 04:00:00 EST 1993 · Transactions of the American Nuclear Society; (United States)

OSTI ID:6911573

Hagen, S; Hofmann, P ^[1]; Allison, C ^[2]

---

1. Kernforschungszentrum Karlsruhe (Germany)
2. Idaho National Engineering Lab., Idaho Falls, ID (United States)

At the time of the Three Mile Island unit 2 accident, the severe fuel damage data base was very limited. However, as a result of international severe accident research programs, that data base has been greatly expanded to include experiments ranging from single rod tests in the NIELS facility at the Kernforschungszentrum, Karlsruhe, Germany (KfK), to the coupled reactor coolant system (RCS) thermal-hydraulics-assembly melting loss-of-fluid test (LOFT) LP-FP-2 test in the LOFT reactor at the Idaho National Engineering Laboratory (INEL). In addition to these tests at the extreme ends of the scaling spectrum, other experiments in the CORA facility at the KfK and the Power Burst Facility at the INEL have also made a substantial contribution to the severe fuel damage data base. The experiments performed at the KfK and the INEL have included different modes of heating from the electrically heated experiments in Germany to the fission and decay-heat-driven experiments in Idaho. The tests have included (a) different heating rates, peak bundle temperatures, and coolant conditions, (b) fresh, trace-irradiated, and previously irradiated (36,000 MWd/tonne U) fuel, and (c) different bundle designs, including fuel-only bundles, fuel bundles with Ag-In-Cd control rods, and fuel bundles with representative B[sub 4]C control blade/channel boxes. This paper discusses several key results from these experiments that are common across all the facilities.

• OSTI ID: 6911573
• Report Number(s): CONF-931160--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 69
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
ENRICHED URANIUM REACTORS
FLUID MECHANICS
FUEL ELEMENTS
FUEL RODS
GERMAN FR ORGANIZATIONS
HYDRAULICS
KERNFORSCHUNGSZENTRUM KARLSRUHE
LOFT REACTOR
MECHANICS
MELTING
NATIONAL ORGANIZATIONS
PHASE TRANSFORMATIONS
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH PROGRAMS
SAFETY
TANK TYPE REACTORS
TEST REACTORS
TESTING
THERMAL ANALYSIS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS