03/02652 Simulation of delayed neutrons using MCNP: Werner, C. J. Progress in Nuclear Energy, 2002, 41, (1–4), 385–389

Abstract

The primary containment of a liquid metal fast breeder reactor (LMFBR) is designed to allow removal of decay heat from a damaged core in the event of a hypothetical core disruptive accident (HCDA). The primary containment consists of the reactor vessel, the vessel head and the primary loops (piping, heat exchangers, pumps, etc.) and must retain functional integrity if decay heat requirements are to be met following an HCDA. As part of an evaluation program, an experimental program was carried out by Stanford Research Institute (SRI) under the auspices of the Hanford Engineering Development Laboratory (HEDL) and the US Atomic Energy Commission ∗∗ to investigate the structural integrity of the primary containment of the fast flux test facility (FFTF) under an HCDA of 150 MW sec. The main objective of the program was to determine whether the structural envelope, comprised of the reactor vessel, vessel head and primary piping, would remain intact so that in the post-accident period coolant flow and decay heat removal through the heat transport system would not be precluded. Another objective was to provide data for correlation with predictions of computer codes so that they can eventually be applied with confidence to full-scale reactors. The two main parts of the program were: (1) development of an explosive energy source for simulating the HCDA, and (2) investigation of hydrodynamic and structural effects of an HCDA. In the source development, instrumented 1 : 30 and 1 : 10 scale models of the fast test reactor (FTR) were employed without the vessel head. In the investigation of hydrodynamic and structural effects of an HCDA, instrumented 1 : 30 and 1 : 10 scale simple models of the FTR were employed, as well as two instrumented 1 : 30 scale complex models of the FTR and one instrumented 1 : 10 scale complex model of the FTR with three primary loops.The paper presents selected detailed results from the experiments on the 1 : 30 and 1 : 10 scale simple rigid and flexible models of the FTR to increase the data available for computer code development. Source development and most of the general results of the simple model experiments have already been presented. For convenience, the paper includes the general results. To achieve the objective of providing experimental data for assisting computer code validation the description includes the FTR model dimensions and material properties. The specific results generated include: simulated HCDA core pressures; pressures at the vessel wall, head and outlet pipe; strains in the vessel wall and head; water surface displacement behavior; and final vessel and core barrel deformations. Observations drawn from the experimental results address the following subjects: overall structural response; energy partition among vessel, core barrel, head and holding-down bolts; strain and deformation magnitudes; reproducibility; and scaling.