RCIC Governing Equation Scoping Studies
Abstract not provided.
Publications
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.
RCIC governing equation scoping studies for severe accidents
International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
A simplified but mechanistic governing equation for a reactor core isolation cooling (RCIC) system is developed to support investigations into severe accident mitigation strategies. Since the RCIC uses a single-stage impulse turbine, the model is essentially the application of Newton's Laws for a rotational system. Specifically, the control volume formulation of angular momentum conservation is used to derive an equation of motion that is simple enough to be implemented as user input for lumped parameter codes such as MELCOR. Preliminary testing of the RCIC equations and solution methodology has been completed. The equations are integrated into MELCOR input via control functions for scoping calculations; the derivation of the equations and solution methods are intentionally selected to facilitate this effort and the subsequent scoping calculations. The MELCOR model used for the test calculations contains simplified representations of the RCS and RCIC piping for a generic 2000 MW BWR. Scoping calculations of the accident scenario at Fukushima unit 2 are presented that show promising initial results. In conjunction with a literature review of RCIC turbine design, a key conclusion is established that the simplicity and pure-impulse design of the turbine facilitates computational modeling using simplified (lumped-parameter) momentum methods.
Risk Management for Sodium Fast Reactors
Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.
SNL BSAF Update
Abstract not provided.
?Smart Procedures?: Using dynamic PRA to develop dynamic context-specific accident severe management guidelines
Abstract not provided.
SOARCA Surry Power Station Uncertainty Analysis: Parameter Methodology and Insights
Abstract not provided.
"Smart Procedures": Using dynamic PRA to develop dynamic context-specific severe accident management guidelines (SAMGs)
Abstract not provided.
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.
SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis: Knowledge Advancement
This paper describes the knowledge advancements from the uncertainty analysis for the State-of- the-Art Reactor Consequence Analyses (SOARCA) unmitigated long-term station blackout accident scenario at the Peach Bottom Atomic Power Station. This work assessed key MELCOR and MELCOR Accident Consequence Code System, Version 2 (MACCS2) modeling uncertainties in an integrated fashion to quantify the relative importance of each uncertain input on potential accident progression, radiological releases, and off-site consequences. This quantitative uncertainty analysis provides measures of the effects on consequences, of each of the selected uncertain parameters both individually and in interaction with other parameters. The results measure the model response (e.g., variance in the output) to uncertainty in the selected input. Investigation into the important uncertain parameters in turn yields insights into important phenomena for accident progression and off-site consequences. This uncertainty analysis confirmed the known importance of some parameters, such as failure rate of the Safety Relief Valve in accident progression modeling and the dry deposition velocity in off-site consequence modeling. The analysis also revealed some new insights, such as dependent effect of cesium chemical form for different accident progressions. (auth)
"Smart SAMGs" Dynamic robust severe accident management guidelines
Abstract not provided.
Melcor simulations of the severe accident at the Fukushima Daiichi unit 1 reactor
Nuclear Technology
In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and US. 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 (IFI), 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.
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.
MELCOR/MACCS2 Analysis for BWR Mark-I Filtered Containment Venting
Abstract not provided.
Discrete Dynamic Event Tree Analysis of Small Modular Reactor Severe Accident Management
Abstract not provided.
LEVERAGING EXISTING TOOLS FOR SIMULATING OPERATOR PERFORMANCE IN DISCRETE DYNAMIC EVENT TREES
Abstract not provided.
MELCOR/MACCS2 Analysis for BWR Mark I Filtered Containment Venting
Abstract not provided.
Discrete Dynamic Event Tree Analysis of Small Modular Reactor Severe Accident Management
Abstract not provided.
LEVERAGING EXISTING TOOLS FOR SIMULATING OPERATOR PERFORMANCE IN DISCRETE DYNAMIC EVENT TREES
Abstract not provided.
SOARCA Peach Bottom Atomic Power Station long-term station blackout uncertainty analysis MELCOR parameters and probabilistic results
Abstract not provided.
Core Molten Pool Modeling
Abstract not provided.
Fukushima Accident Spent Fuel Pool Model in MELCOR V2.1
Abstract not provided.
MELCOR Simulations of the Severe Accident at the Fukushima 1F1 Reactor
Abstract not provided.
MELCOR 2.1 Simulations of Fukushima Unit 3
Abstract not provided.
MELCOR Source Term Calculation
Abstract not provided.
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