Modeling of the HDR E11. 4 containment test using the CONTAIN severe reactor accident code

Conference · Wed Jan 01 00:00:00 EST 1992 · Transactions of the American Nuclear Society; (United States)

OSTI ID:6660790

Green, J; Almenas, K ^[1]

---

1. Univ. of Maryland, College Park (United States)

During 1989, a series of severe reacotr accident tests was performed at the HDR test facility in Karlsruhe, Germany. This facility is a decommissioned light water nuclear power plant with more than 700 data measurement channels located throughout the containment. These tests, the E11 series, were designed to simulate conditions in a containment over long periods of time under scenarios that are not always considered during the design phase. The purpose of the tests was to clarify the concepts and procedures necessary for mitigation of these events. An added aspect of these tests is to allow verification and enhancement of computer simulation programs currently used for containment analysis. The CONTAIN servere reactor accident code is one such program developed by Sanida National Laboratory. The E11.4 test of the series included several unique events over a 56-h period. Ths fission product decay heat to the containment vessel from the primary system was simulated by a continuous external source of steam introduced into a lower compartment room. The actual loss-of-coolant accident (LOCA) was performed using the existing reactor vessel pressurized to 110 bar, which was allowed to blow down into the identical room as the external source. Later in the test an 85-15 vol % helium-hydrogen gas mixture was introduced into the containment to simulate the oxidation reaction of Zircaloy cladding, which can occur during the LOCA. The gas mixture had been added at two time frames during the test and with varying mass flow rates. Twenty-four gas measuring instruments were located throughout the containment building to assess the distribution of the hydrogen gas as well as the effect of noncondensable gases on the heat transfer to existing structures.

• OSTI ID: 6660790
• Report Number(s): CONF-921102-; CODEN: TANSAO
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 66; Conference: Joint American Nuclear Society (ANS)/European Nuclear Society (ENS) international meeting on fifty years of controlled nuclear chain reaction: past, present, and future, Chicago, IL (United States), 15-20 Nov 1992; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
CONTAINMENT
MATHEMATICAL MODELS
HDR REACTOR
AFTER-HEAT REMOVAL
C CODES
COMPARATIVE EVALUATIONS
FUEL-CLADDING INTERACTIONS
HYDROGEN
LOSS OF COOLANT
TESTING
TRANSIENTS
ACCIDENTS
BWR TYPE REACTORS
COMPUTER CODES
ELEMENTS
ENRICHED URANIUM REACTORS
EVALUATION
EXPERIMENTAL REACTORS
NONMETALS
POWER REACTORS
REACTOR ACCIDENTS
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors