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Terry Turbopump Expanded Operating Band Full-Scale Integral Long-Term Low-Pressure Experiments – Preliminary Test Plan90685Terry Turbopump Expanded Operating Band Full-Scale Integral Long-Term Low-Pressure Experiments – Preliminary Test Plan, SAND2018-12965, D. OSborn, M. Solom, January 2019. 1/9/2019 2:52:02 PMSANDIA REPORT SAND2018-12965 Unlimited Release January 2019 Terry Turbopump Expanded Operating Band Full-Scale Integral Long-Term Low-Pressure Experiments – Preliminary Test Plan 231https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2019-01-13T07:00:00Z
The CORQUENCH Code for Modeling of Ex-Vessel Corium Coolability under Top Flooding Conditions Code Manual – Version4.1-beta221993The CORQUENCH Code for Modeling of Ex-Vessel Corium Coolability under Top Flooding Conditions Code Manual – Version4.1-beta, ANL-18/22, M. Farmer, August 2018.10/15/2018 4:53:00 PMANL -18/22 The CORQUENCH Code for Modeling of Ex-Vessel Corium Coolability under Top Flooding Conditions Code Manual – Version4.1-beta Nuclear Science and 7https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2018-08-06T06:00:00Z
Improvements in the reactor core isolation cooling (RCIC) pump model90681Improvements in the reactor core isolation cooling (RCIC) pump model, INL/EXT-18-44739, H. Zhang, J. O'Brtien, February 2018.3/1/2018 5:22:31 PMINL/EXT-18-44739 Improvements in the reactor core isolation cooling (RCIC) pump model Hongbin Zhang James O'Brien February 2018 U.S. Department of Energy Office of Nuclear 167https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2018-02-18T07:00:00Z
Enhancements to Engineering-scale Reactor Pressure Vessel Fracture Capabilities in Grizzly, INL/EXT-17-43427.221997Enhancements to Engineering-scale Reactor Pressure Vessel Fracture Capabilities in Grizzly, INL/EXT-17-43427, B. Spencer, W. Hoffman, W. Jiang, September 2017.9/28/2017 8:27:11 PMINL/EXT-17-43427 Light Water Reactor Sustainability Program Enhancements to Engineering-scale Reactor Pressure Vessel Fracture Capabilities in Grizzly Benjamin W. Spencer 8https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2017-09-18T06:00:00Z
Status Report on Ex-Vessel Coolability and Water Management, ANL/NE-16/18.90684Status Report on Ex-Vessel Coolability and Water Management, ANL/NE-16/18, M. Farmer, K. Robb, September 15, 2016.9/16/2016 2:32:14 AMANL/NE-16/18 Status Report on Ex-Vessel Coolability and Water Management NE Division About Argonne National Laboratory Argonne is a U.S. Department of 176https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2016-09-15T06:00:00Z
US Efforts in Support of Examinations at Fukushima Daiichi – 2016 Evaluations, ANL/LWRS-16/02.90686US Efforts in Support of Examinations at Fukushima Daiichi – 2016 Evaluations, ANL/LWRS-16/02, P. Amway, N. Andrews, W. Bixby, R. Bunt, M. Corradini, P. Ellison, M. Farmer, T. Farthing, M. Francis, J. Gabor, R. Gauntt, C. Henry, P. Humrickhouse, S. Kraft, R. Linthicum, W. Luangdilok, R. Lutz, D. Luxat, J. Maddox, C. Negin, c. Paik, M. Plys, J. Rempe, K. Robb, R. Sanders, R. Warchowiak, B. Williamson, August 2016.9/7/2016 3:01:01 PMANL/LWRS-16/02 Light Water Reactor Sustainability Program US Efforts in Support of Examinations at Fukushima Daiichi – 2016 Evaluations August 2016 U.S. Department of Energy 159https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2016-08-01T06:00:00Z
Development of Three-Field Poromechanics Capability in MOOSE, INL/EXT-16-38931.222013Development of Three-Field Poromechanics Capability in MOOSE, M. Sivaselvan, N. Oliveto, A. Whittaker, INL/EXT-16-38931, April 2016.6/2/2016 3:02:48 PMThis information was prepared as an account of work sponsored by an agency of the U.S. Government Neither the U.S. Government nor any agency thereof, nor any of their employees 7https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2016-04-01T06:00:00Z
Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors – Volume I, SAND2016-0232.90679Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors – Volume I, Sandia National Laboratories, SAND2016-0232, January 2106.1/13/2016 1:20:49 PMSevere Accident Analysis Department - 06232 Sandia National Laboratories Albuquerque, New Mexico 87185 and Livermore, California 94550 NOTICE: This report was prepared as an 200https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2016-01-19T07:00:00Z
Application of Nonlinear Seismic Soil-Structure Interaction Analysis for Identification of Seismic Margins at Nuclear Power Plants, INL/EXT-15-37382.90674Application of Nonlinear Seismic Soil-Structure Interaction Analysis for Identification of Seismic Margins at Nuclear Power Plants, A. Varma, J. Seo, J. Coleman, INL/EXT-15-37382, November 201511/30/2015 6:15:59 PMAmit H. Varma, Jungil Seo, and Justin Coleman This information was prepared as an account of work sponsored by an agency of the U.S. Government Neither the U.S. Government nor any 210https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2015-11-03T07:00:00Z
Scoping Study Investigating PWR Instrumentation during a Severe  Accident Scenario, INL/EXT-15-35940.90677Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario, J. Rempe, D. Knudson, R. Lutz, INL/EXT-15-35940, September 2015.9/17/2015 5:22:16 PMINL/EXT-15-35940 Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario Joy L. Rempe, Rempe and Associates, LLC Darrell L. Knudson, Idaho National 9https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2015-09-01T06:00:00Z
US Efforts in Support of Examinations at Fukushima Daiichi, J. Rempe, ANL/LWRS-15/2.90678US Efforts in Support of Examinations at Fukushima Daiichi, J. Rempe, ANL/LWRS-15/2, August 2015.8/20/2015 2:38:09 AMU.S. Department of Energy Office of Nuclear Energy This information was prepared as an account of work sponsored by an agency of the U.S. Government This report provides a basis for 132https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2015-08-01T06:00:00Z
Fukushima Daiichi Unit 1 Uncertainty Analysis – Exploration of Core Melt Progression Uncertain Parameters – Volume II, SAND2015-6612.90680Fukushima Daiichi Unit 1 Uncertainty Analysis – Exploration of Core Melt Progression Uncertain Parameters – Volume II, M. Denman, D. Brooks, SAND2015-6612, August 2015.8/6/2015 9:13:09 PMPrepared by Sandia National Laboratories Albuquerque, New Mexico 87185 and Livermore, California 94550 Sandia National Laboratories is a multi-program laboratory managed and 112https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2015-08-01T06:00:00Z
Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis, ANL-NE-15-4.90673Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis, R. Bunt, M. Corradini, P. Ellison, M. Farmer, M. Francis, J. Gabor, R. Gauntt, C. Henry, R. Linthicum, W. Luangdilok, R. Lutz, C. Paik, M. Plys, C. Rabiti, J. Rempe, K. Robb, R. Wachowiak, ANL-NE-15-4, March 31, 2015.3/31/2015 7:55:35 PMANL/NE-15/4 Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis Nuclear Engineering Division About Argonne National Laboratory Argonne is 6https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2015-03-31T06:00:00Z
Modular Accident Analysis Program (MAAP) – MELCOR Crosswalk Phase 1 Study, 3002004449, Electric Power Research Institute, Technical Update, November 2014.90675Modular Accident Analysis Program (MAAP) – MELCOR Crosswalk Phase 1 Study, 3002004449, Electric Power Research Institute, Technical Update, R. Wachowiak, EPRI 3002004449, November 2014.12/21/2014 12:18:42 AMELECTRIC POWER RESEARCH INSTITUTE 3420 Hillview Avenue, Palo Alto, California 94304-1338 ƒ PO Box 10412, Palo Alto, California 94303-0813 ƒ USA 800.313.3774 ƒ 650.855.2121 564https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2014-11-01T06:00:00Z
Report on Prevention Analysis Trial Method and Case Study Improvements222053Report on Prevention Analysis Trial Method and Case Study Improvements, D. Blanchard and R. Youngblood, M3LW-12IN0702012, April 2012.6/28/2012 5:44:35 PMThe INL is a U.S. Department of Energy National Laboratory operated by Battelle Energy Alliance INL/EXT-11-23479 Revision 1 Risk Informed Safety Margin Characterization Case Study 10https://lwrs.inl.gov/Reactor Safety Technologies/Forms/AllItems.aspxpdfFalsepdf2012-04-01T06:00:00Z