ML13196A432
| ML13196A432 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 07/15/2013 |
| From: | Atomic Safety and Licensing Board Panel |
| To: | |
| SECY RAS | |
| References | |
| 50-443-LR, ASLBP 10-906-02-LR-BD01, RAS 24821 | |
| Download: ML13196A432 (7) | |
Text
NextEra Energy Seabrook, LLC (Seabrook Station, Unit 1)
License Renewal Application NRC Staff Answer to Motion for Summary Disposition of Contention 4B ATTACHMENT 4B-N
R A ND ALL OW E N G AU NTT 55 Eastridge Rd l Edgewood, NM 87015 l +1 (505) 2860619 l rogauntt@gmail.com EMPLOYMENT HISTORY Manager Severe Accident Analysis Department 2004 Present Sandia National Laboratories
- Perform diverse research activities for DOE and NRC including analyses of severe accident progression in commercial power plants
- Manage a team of researchers and code developers who develop and apply the MELCOR severe accident analysis code and the MACCS atmospheric transport and consequence assessment code Temporary Assignment in U.S. Embassy in Tokyo supporting DOE 3/23/2011 4/27/2011 Fukushima Support and Response Sandia National Laboratories
- Provided in-country technical support and consultation to numerous stakeholders following the accidents at the Fukushima power plants, including support to the U.S. Military, U.S. NRC, DOS and DOE as well at Japanese organizations including TEPCO and the JNES.
- Supervised initial source term estimations for use in ongoing estimation of what releases from the accidents could have been
- Performed MELCOR analyses evaluating the potential damage states of each of the Fukushima reactors
- Performed MELCOR analyses focused on assessing potential for subsequent damage to the already crippled Fukushima cores in the event of strong seismic aftershocks to evaluate potential consequences of continued core damages 2003 2004 Distinguished Member of Technical Staff Technical Leader of the MELCOR Code development project 1995 2004
- Led a team of engineers in the development of the MELCOR severe accident computer code for the U.S. Nuclear Regulatory Commission
- Performed expert analyses of severe accident behavior using MELCOR aimed at characterizing anticipated fission product releases from accidents from nuclear power plants
- Performed numerous code validation and assessment efforts aimed at developing and improving physical models of core melt progression and fission product release and transport behavior Technical Leader of the Ex-Reactor Severe Accident Experiments Team 1991 1996 Sandia National Laboratories
- Led the design and construction of a test facility for performing high temperature melt progression experiments on BWR core structures
- Developed a novel and unique wire-fed melt furnace to provide high temperature core melt materials for simulating meltdown on Zircaloy cladding, channel box materials and BWR control blade materials
- Directed analysis support in the development of the MERIS porous media code for predicting BWR core degradation and melt progression behavior.
Technical Leader of the ACRR In-Pile MP Late Phase LWR Fuel 1988 1994 Damage Experiments Sandia National Laboratories
- Based on observations form the TMI-2 core examinations, developed a test program for characterizing Late Phase core melt progression behavior where fuel debris subsequently melts and thermally attacks underlying core blockages that formed earlier in the severe accident
- Performed in-pile experiments MP-1 and MP-2 producing phenomenological test data for modeling late phases of core degradation beginning with a degraded fuel debris bed overlying lower intact fuel rods and pre-formed core blockages using prototypic materials (enriched UO2, ZrO2 and control materials).
- Supervised neutronic analyses on MP experiments coupling with the neutron flux of the ACRR central irradiation cavity to ensure proper fission heating of the test package
- Tests were performed in the ACRR reactor using fission heating methods to simulate decay heat
- Prepared Safety Analysis Reports for the MP Experiments and presented to the ACRR Safety Review Board to assure QA and safe conduct of experiments capable of producing radiological releases
- Directed analysis support in the development of the DEBRIS porous media code for predicting late phase melt progression behavior.
Technical Leader of the ACRR In-Pile DFR(Damaged Fuel and 1984 1988 Relocation) Early Phase LWR Fuel Damage Experiments Sandia National Laboratories
- Motivated by the accident at TMI-2 and aimed at characterizing the phenomena involved in initial LWR core damage in a severe accident, including cladding oxidation and hydrogen generation, fuel liquefaction and relocation processes and transition from rod-like geometry to degraded core debris geometry
- Performed in-pile experiments DF-3 and DF-4 producing phenomenological test data for modeling early phases of core degradation beginning with a rod geometry in a small 9-rod test bundle. DF-3 included a Ag-In-Cd control rod and DF-4 was the first ever experiment to investigate a SS-B4C BWR control blade element.
- Played supporting engineer role in the conduct of earlier DF-1 and DF-2 experiments and supervised test report and post test analyses.
- Supervised neutronic analyses on DFR experiments coupling with the neutron flux of the ACRR central irradiation cavity to ensure proper fission heating of the test package
- Tests were performed in the ACRR reactor using fission heating methods to simulate decay heat
- Prepared Safety Analysis Reports for the DF-3 and DF-4 Experiments and presented to the ACRR Safety Review Board to assure QA and safe conduct of experiments capable of producing radiological releases
- Directed analysis support in the development of the MELPROG code for predicting fuel heatup, uncover and damage progression phenomena.
Technical Leader of the TRAN-GAP Experiment Project 1982 1986 Sandia National Laboratories
- Motivated by safety issues raised concerning the Liquid Metal Fast Breeder Reactor (LMFBR) project to be built at Clinch River, these experiments were aimed at characterizing the behavior of reactor fuel that has melted during a Hypothetical Core Disruptive Accident (HCDA) and become ejected from the core region to the breeder blanket region. At question was whether the ejected fuel would freeze in the blanket region or if the accident would TRANsition to a renewed critical condition.
- Performed in-pile experiments GAP-1 and Gap-3 wherein more than 2 kg of enriched UO2 was melted in a prompt pulse of neutrons from the ARCC central irradiation cavity and ejectd under pressure into prototypic LMFBR blanket region structures.
- Played supporting engineer role in the conduct of earlier TRAN 1 - 4 experiments that explored molten fuel ejection into core fuel pin geometry
- Supervised neutronic and transient thermal analyses on TRAN experiments coupling with the neutron flux of the ACRR central irradiation cavity to ensure proper fission heating/melting of the fuel materials using neutrons from the ACRR operated in pulse mode at over 30,000MW per pulse.
- Prepared Safety Analysis Reports for the GAP-1 and GAP-2 Experiments and presented to the ACRR Safety Review Board to assure QA and safe conduct of experiments capable of producing radiological releases
- Directed analysis support in the development of the PLUGM code for predicting fuel fuel motion and freezing in fuel pin geometry.
ACTIVITIESAND AWARDS
- 2011 USDOE Secretarial Honors Award for Japan Earthquake and Tsunami Disaster Response Teams - this is the highest honor awarded to DOE Laboratory contractors, awarded for ongoing support of DOE and Japan during the Fukushima post-accident crisis management period.
- 2003 Promoted to the position of Distinguished Member of Technical Staff, the highest level of recognition open to members of the Sandia Technical Staff
- Maintained Membership in American Nuclear Society (ANS) and American Society of Mechanical Engineers (ASME) and regularly supporting organization and reviews for technical conferences of the ANS and ASME, including ANS, NURETH and ICONE.
PUBLICATIONS
- See Attached List
Title Reference Author Year ANS Winter Meeting, Jeff Cardoni, Don MELCOR Simulations of the Severe Fukushima Embedded Kalinich, Jeff Cardoni Accident at the Fukushima 1F3 Reactor Topical, San Diego and Randall Gauntt 2012 ANS Winter Meeting, Jesse Phillips, Don MELCOR Simulations of the Severe Fukushima Embedded Kalinich, Jeff Cardoni Accident at the Fukushima 1F2 Reactor Topical, San Diego and Randall Gauntt 2012 ANS Winter Meeting, Randall O. Gauntt, Don MELCOR Simulations of the Severe Fukushima Embedded Kalinich, Jeff Cardoni Accident at the Fukushima 1F1 Reactor Topical, San Diego and Jesse Phillips 2012 Fukushima Daiichi Accident Study (Status as of April 2012) SAND2012-6173 Randall Gauntt, et al. 2012 State of the Art Reactor Consequence Project, Volume 2, Surry Integrated Analyses NUREG/CR 7110 Vol 2 Randall O. Gauntt, et al 2012 State of the Art Reactor Consequence Project, Volume 1, Peach Bottom Integrated Analyses NUREG/CR 7110 Vol 1 Randall O. Gauntt, et al 2012 Accident Source Terms for Light-Water D.A. Powers, M.T.
Nuclear Power Plants Using High- Leonard, R.O. Gauntt, Burnup or MOX Fuel SAND2011-0128 R.Y. Lee, and M. Salay 2011 Scott G. Ashbaugh, Assessment of Severe Accident Source Kenneth C. Wagner, Terms in Pressurized-Water Reactors Pamela Longmire, with a 40% Mixed-Oxide and 60% Low- Randall O. Gauntt, Enriched Uranium Core Using MELCOR Andrew S. Goldmann, 1.8.5 SAND2008-6665 and Dana A. Powers 2010 Synthesis of VERCORS and Phebus Data in Severe Accident Codes and Applications SAND2010-1633 Randall O. Gauntt 2010 MELCOR 1.8.5 Modeling Aspects of Fission Product Release, Transport and Deposition SAND2010-1635 Randall O. Gauntt 2010 Randall O. Gauntt, Analysis of Main Steam Isolation Valve Tracy Radel, Michael A.
Leakage in Design Basis Accidents Using Salay, and Donald A.
MELCOR 1.8.6 and RADTRAD SAND2008-6601 Kalinich 2008 K. C. Wagner, R. O.
A New Air Oxidation Model for Safety Gauntt, E. R. Lindgren, Analysis SAND2007-0597P and S. Durbin 2007 Accident Source Terms for Boiling Mark T. Leonard, Water Reactors with High Burnup Cores Randall O. Gauntt and Calculated Using MELCOR 1.8.5 SAND2007-7697 Dana A. Powers 2007 NEW MODEL IMPROVEMENTS IN The 11th International Larry L. Humphries, MELCOR 1.8.6 TO SIMULATE THE Topical Meeting on Nuclear Randall O. Gauntt, FORMATION AND BEHAVIOR OF Reactor Thermal-Hydraulics Randall K. Cole, Jamie MOLTEN POOLS IN SEVERE ACCIDENTS (NURETH-11) Popes Palace E. Cash 2005
Conference Center, Avignon, France, October 2-6, 2005.
The 11th International Topical Meeting on Nuclear AN UNCERTAINTY ANALYSIS FOR Reactor Thermal-Hydraulics HYDROGEN GENERATION IN STATION (NURETH-11) Popes Palace BLACKOUT ACCIDENTS USING MELCOR Conference Center, Avignon, 1.8.5 France, October 2-6, 2005. Randall O. Gauntt 2005 Severe accident analysis of a PWR K. Vierow, Y. Liao, J.
station blackout with the MELCOR, Science Direct, Nuclear Johnson, M. Kenton, R.
MAAP4 and SCDAP/RELAP5 codes Engineering and Design Gauntt 2004 R. O. Gauntt, R. K. Cole, C. M. Erickson, R. G.
Gido, R. D. Gasser, S. B.
Rodriguez, and M. F.
MELCOR Computer Code Manuals, Vol. Young, Scott 3: Demonstration Problems, Version NUREG/CR-6119, Vol. 3, Ashbaugh, Mark 1.8.5 May 2001 Rev. 0; SAND2001-0929P Leonard, and Adam Hill 2001 N. E. Bixler, R K. Cole, Recent MELCOR and VICTORIA Fission M. F. Young, R. 0.
Product Research at the NRC SAND99-0179C Gauntt, J.H. Schaperow 1999 R.O. Gauntt, R.K Cole, S.A. Hedge, S.B.
Rodriguez, RL. Sanders, MELCOR Computer Code Manuals, RC. Smith, D.S. Stuart, Primer and Users Guides, Version 1.8.4 NUREG/CR-6119, VoL 1, Rev. RM. Summers, M.F.
July 1997 1; SAND97-2398 Young 1997 R.O. Gauntt, R.K. Cole, S.A Hedge, S.B.
Rodriguez, R.L.
MELCOR Computer Code Manuals, Sanders, RC. Smith, Reference Manuals, Version 1.8.4 July NUREG/CR-6119, Vol. 2, D.S. Stuart, RM.
1997 Rev. 1; SAND97-2398 Summers, MF. Young 1997 NEW MODEL IMPROVEMENTS IN MELCOR 1.8.6 TO SIMULATE THE FORMATION AND BEHAVIOR OF NUREG/CR-6527; SAND97- R. O. Gauntt, L. L.
MOLTEN POOLS IN SEVERE ACCIDENTS 1039 Humphries 1997 Stephen A. Slutz, Randall O. Gauntt, and JOURNAL OF PROPULSION Gary A. Harms, Thomas Thermal Radiation in Gas Core Nuclear AND POWER, Vol. 10, No. 3, Latham, Ward Roman, Reactors for Space Propulsion May-June 1994 and Richard J. Rodgers 1994 Randall 0. Gauntt, Stephen A. Slutz, and IN-REACTOR TESTING OF THE CLOSED Gary A. Harms, Thomas CYCLE GAS CORE REACTOR -THE S. Latham, Ward C.
NUCLEAR LIGHT BULB CONCEPT- SAND92-1462C Roman, and Richard J. 1992
Rodgers Stephen. A. Slutz, Randall. 0. Gauntt, and Gary. A. Harms, Thermal Radiative Transport in Gas Thomas Latham, Ward Core Nuclear Reactors for Space Roman, and Richard J.
Propulsion SAND92-2193J Rodgers 1992 THE DEBRIS MODULE: AN EFFECTIVE TOOL FOR THE ANALYSIS OF MELT R.D. Gasser, S.S.
PROGRESSION IN LWRs SAND90-2712C Dosanjh, R.O. Gauntt 1990 RESULTSOF THE DF-4BWR R. O. Gauntt and R.D.
CONTROLBLADE-CHANNELBOXTEST SAND90-2716C Gasser 1990 MELPPROG Debris Meltdown Model S.S. Dosanjh and R.O.
and Validation Experiments SAND88-0263C Gauntt 1988 MELPROG Analysis of the Debris T.J. Heames and R.O.
Formation Experiment DF-1 SAND87-2330C Gauntt 1987 CONTRIBUTIONS FROM THE ACRR IN-PILE EXPERIMENTS TO THE P. ROYL, W. BREITUNG, UNDERSTANDING OF KEY PHENOMENA E.A. FISCHER, G.
INFLUENCING UNPROTECTED LOSS OF Nuclear Engineering and SCHUMACHER, R.O.
FLOW ACCIDENT SIMULATIONS IN Design 100 (1987) 387-408, GAUNTT and S.A.
LMFBRs North Holland, Amsterdam WRIGHT 1986 R.D. Gasser, C.P. Fryer, R.O. Gauntt, A.C.
Damaged Fuel Experiment DF-1, Results NUREG/CR-4668; SAND86- Marshall, K.O. Reil, K.T.
and Analyses 1030 Stalker 1986 The DF-4 Fuel Damage Experiment in ACRR with a BWR Control Blade and NUREG/CR-4671; SAND86- R.O. Gauntt, R.D.
Channel Box 1443 Gasser, L.J. Ott 1986