Jeffrey N Cardoni

Fukushima Daiichi accident study : status as of April 2012.

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission (NRC) and Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing severe accident modeling capability of the MELCOR code. MELCOR is the state-of-the-art system-level severe accident analysis code used by the NRC to provide information for its decision-making process in this area. The objectives of the project were: (1) collect, verify, and document data on the accidents by developing an information portal system; (2) reconstruct the accident progressions using computer models and accident data; and (3) validate the MELCOR code and the Fukushima models, and suggest potential future data needs. Idaho National Laboratory (INL) developed an information portal for the Fukushima Daiichi accident information. Sandia National Laboratories (SNL) developed MELCOR 2.1 models of the Fukushima Daiichi Units 1, 2, and 3 reactors and the Unit 4 spent fuel pool. Oak Ridge National Laboratory (ORNL) developed a MELCOR 1.8.5 model of the Unit 3 reactor and a TRACE model of the Unit 4 spent fuel pool. The good correlation of the results from the SNL models with the data from the plants and with the ORNL model results provides additional confidence in the MELCOR code. The modeling effort has also provided insights into future data needs for both model development and validation.

MELCOR Simulations of the Severe Accident at Fukushima Daiichi Unit 3

Abstract In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty due to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. The MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.

MELCOR Simulations of the Severe Accident at the Fukushima Daiichi Unit 1 Reactor

Abstract In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code and developing an understanding of the likely accident progression. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of MELCOR and the Fukushima models against plant data. In this study Sandia National Laboratories developed MELCOR 2.1 models of Fukushima Daiichi Units 1 (1F1), 2, and 3 as well as the Unit 4 spent fuel pool. This paper reports on the analysis of the 1F1 accident. Details are presented on the modeled accident progression, hypothesized mode of failures in the reactor pressure vessel (RPV) and containment pressure boundary, and release of fission products to the environment. The MELCOR-predicted RPV and containment pressure trends compare well with available measured pressures. Conditions leading up to the observed explosion of the reactor building are postulated based on this analysis where drywell head flange leakage is thought to have led to accumulation of flammable gases in the refueling bay. The favorable comparison of the results from the analyses with the data from the plant provides additional confidence in MELCOR to reliably predict real-world accident progression. The modeling effort has also provided insights into future data needs for both model development and validation.

Presentation of Fukushima Analyses to U.S. Nuclear Power Plant Simulator Operators and Vendors

This document provides Sandia National Laboratories’ meeting notes and presentations at the Society for Modeling and Simulation Power Plant Simulator conference in Jacksonville, FL. The conference was held January 26-28, 2015, and SNL was invited by the U.S. nuclear industry to present Fukushima modeling insights and lessons learned.

Fukushima Daiichi unit 1 uncertainty analysis--Preliminary selection of uncertain parameters and analysis methodology

Sandia National Laboratories (SNL) plans to conduct uncertainty analyses (UA) on the Fukushima Daiichi unit (1F1) plant with the MELCOR code. The model to be used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). However, that study only examined a handful of various model inputs and boundary conditions, and the predictions yielded only fair agreement with plant data and current release estimates. The goal of this uncertainty study is to perform a focused evaluation of uncertainty in core melt progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, vessel lower head failure, etc.). In preparation for the SNL Fukushima UA work, a scoping study has been completed to identify important core melt progression parameters for the uncertainty analysis. The study also lays out a preliminary UA methodology.

Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.