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  • SANDIA REPORT SAND2014-1170 Unlimited Release Printed February 2014

    Fukushima Daiichi Unit 1 Uncertainty Analysis -- Preliminary Selection of Uncertain Parameters and Analysis Methodology

    Jeff Cardoni and Donald Kalinich

    Prepared by Sandia National Laboratories Albuquerque, New Mexico 87185 and Livermore, California 94550

    Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. Approved for public release; further dissemination unlimited.

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    Issued by Sandia National Laboratories, operated for the United States Department of Energy

    by Sandia Corporation.

    NOTICE: This report was prepared as an account of work sponsored by an agency of the

    United States Government. Neither the United States Government, nor any agency thereof,

    nor any of their employees, nor any of their contractors, subcontractors, or their employees, make any warranty, express or implied, or assume any legal liability or responsibility for the

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    of their contractors.

    Printed in the United States of America. This report has been reproduced directly from the best

    available copy.

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    SAND2014-1170

    Unlimited Release

    Printed February 2014

    Fukushima Daiichi Unit 1 Uncertainty Analysis -- Preliminary Selection of Uncertain Parameters

    and Analysis Methodology

    Jeff Cardoni and Donald Kalinich

    Sandia National Laboratories

    P.O. Box 5800

    Albuquerque, New Mexico 87185-MS0748

    Abstract

    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.

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    ACKNOWLEDGEMENTS

    The authors would like to acknowledge the U.S. Department of Energy’s Office of Nuclear

    Energy for providing the funding which supported this work.

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    CONTENTS

    1 OVERVIEW ........................................................................................................................... 14

    1.1 Purpose of this Report ................................................................................................... 15

    1.2 Peach Bottom SOARCA Uncertainty Study and Comparison to Fukushima Daiichi.. 15

    1.3 Accident Sequence and Operator Mitigative Actions ................................................... 17

    1.4 Scope of Fukushima UA ............................................................................................... 17

    1.5 Report Structure Summary ........................................................................................... 18

    2 PLANT DESIGN .................................................................................................................... 20

    2.1 RPV and RCS ............................................................................................................... 20

    2.2 Containment .................................................................................................................. 22

    2.3 Safety Systems .............................................................................................................. 22

    3 PEACH BOTTOM SOARCA UNCERTAINTY ANALYSIS .............................................. 24

    4 BOUNDARY CONDITION UNCERTAINTIES .................................................................. 28

    4.1 Initiating and External Events – Associated Immediate System Response .................. 28

    4.2 Unknown System Behaviors and Failures .................................................................... 29

    4.2.1 Functioning of ICs .......................................................................................... 29

    4.2.2 Modes of RPV Depressurization and RCS Pressure Boundary Failure ......... 29

    4.2.3 Containment Failure and Self-Venting ........................................................... 30

    4.3 Operator Mitigative Actions ......................................................................................... 31

    4.3.1 Emergency Core Injection .............................................................................. 31

    4.3.2 Containment Venting ...................................................................................... 31

    4.4 Core Decay Heat and Radionculide Inventory ............................................................. 31

    5 CORE PHENOMONLOGICAL MODEL INPUT UNCERTAINTIES ................................ 38

    5.1 COR Package Sensitivity Coefficients ......................................................................... 38

    5.1.1 Molten Zircaloy Melt Break-through Temperature ........................................ 39

    5.1.2 Molten Clad Drainage Rate ............................................................................ 41

    5.1.3 Time Constants for Radial Debris Relocation ................................................ 42

    5.1.4 Sensitivity Coefficients for MELCOR Models for Core Blockages .............. 45

    5.1.5 Control Volume Temperature Distribution Models ........................................ 45

    5.1.6 RPV Lower Head Failure................................................................................ 50

    5.2 Additional Uncertain COR Package Input .................................................................... 54

    5.2.1 Thermal-Mechanical Weakening of Oxidized Cladding ................................ 54

    5.2.2 Candling Secondary Material Transport Parameters ...................................... 58

    5.2.3 Lipinski Dryout Model of a Submerged Debris Bed ...................................... 60

    5.2.4 Debris Quenching Input .................................................................................. 61

    5.3 Treatment of Core Blockages ....................................................................................... 62

    6 MELCOR COMPUTATIONAL UNCERTAINY ................................................................. 67

    6.1 Time Step Schemes for Reliable Transient Execution ................................................. 67

    6.2 Core Nodalization ......................................................................................................... 68

    6.3 MELCOR Numerical Uncertainty ................................................................................ 71

    7 PRELIMINARY UA CALCULATION PLAN...................................................................... 73

    8 SUMMARY ............................................................................................................................ 78

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    REFERENCES ............................................................................................................................. 80

    APPENDIX A COR

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