Test Design, Results, and Archived Database for the OECD Lower Head Failure Program Integral Experiments

Abstract

In the event of a severe core meltdown accident, core material can relocate to the lower head of a pressurized water reactor (PWR) vessel resulting in significant thermal and pressure loads to the vessel. The potential for failure of the pressure vessel makes possible the release of core material to the containment. The objective of this experimental/analytical program is to characterize the mode, timing, and size of lower head failure (LHF) under severe accident conditions. The OECD Lower Head Failure (OLHF) project investigates lower head failure for conditions of low reactor coolant system (RCS) pressure (2–5 MPa) and prototypic through-wall temperature differential (ΔTW >200K). Low RCS pressure is motivated by the desire to use the data to develop models for assessing accident management strategies involving reactor pressure vessel (RPV) depressurization. Pressure transient is useful in assessing the effect of water injection as part of accident management strategy. Prototypic through-wall temperature differential, ΔTW, is of importance because of the need to provide data where stress redistribution in the vessel wall occurs (as a result of decreasing material strength with temperature). Test design and results for the four OLHF integral tests are reported and summarized in this paper. A short description of the test conduct and heating history is followed by a description of the vessel failure site, the vessel deformation, temperature profiles, stress state, and rupture dynamics for each test. Key observations and conclusions are summarized for each test. The ∼1/5 scale tests are extensively instrumented to provide temperature, pressure, and displacement data. The vessel surfaces are mapped both before and after the test to provide measurements of pre-test thickness, post-test thickness, and cumulative vessel deformation. Data has been assessed and qualified in data reports for each test. The data has been preserved in MSEXCEL™ spreadsheets with macro utilities to facilitate access and analysis of the data. As a result, there exists a well-archived, well-qualified database for model development and validation.

Development of failure evaluation method for BWR Lower head in severe accident; - Creep damage evaluation based on thermal-hydraulics and structural analyses -

“…4). These tendencies of Mises stress distributions are similar to the previous papers which focused on rupture analysis for lower head of reactors (Onizawa, 1993), (Humphphries, 2002), (Villanueva, 2011). In our analysis case, it is noted that the highest Mises stresses occur within the penetrations at outward of lower head.…”

“…After the Fukushima Daiichi (1F) nuclear power plant (NPP) accident due to the Pacific coast of Tohoku earthquake and Tsunami in March, 2011, Japan Atomic Energy Agency (JAEA) has been conducting various research activities for the early completion of the decommissioning of NPPs. Before the 1F accident, several researches regarding phenomena inside the reactor and failure behavior of the reactor due to severe accidents had been actively conducted in reaction to the Three Mile Island accident in 1979 (Onizawa, 1993), (Humphphries, 2002), (Tran, 2010), (Villanueva, 2010). However, it is difficult to assess rupture behavior of the lower head of RPV in boiling-water-type NPPs like 1F due to severe accident.…”