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Characterization of solids in the Three Mile Island Unit 2 Reactor defueling water: Addendum

Description: Shortly after ORNL/TM-10362 was issued, it was discovered that a series of 31 figures had been inadvertently omitted. These figures, which consist of scanning electron microscope (SEM) photographs and energy-dispersive X-ray fluorescence scans, provide significant information about the results obtained in the tests performed with water sample W3. This Addendum includes these figures. Details of and comments on the SEM photographs may be found in ORNL/TM-10362.
Date: March 1, 1988
Creator: Campbell, D.O.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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TMI-2 lower head creep rupture analysis

Description: The TMI-2 accident resulted in approximately 40% of the reactor's core melting and collecting on the lower head of the reactor pressure vessel. The severity of the accident has raised questions about the margin of safety against rupture of the lower head in this accident since all evidence seems to indicate no major breach of the vessel occurred. Scoping heat transfer analyses of the relocated core debris and lower head have been made based upon assumed core melting scenarios and core material … more
Date: August 1, 1988
Creator: Thinnes, G.L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Thermal behavior of molten corium during TMI-2 core relocation event

Description: During the TMI-2 accident, a pool of molten corium formed in the central region of the core and was contained by solidified crusts. Failure of the crust surrounding the molten material, at approximately 224 min, resulted in a relocation of an estimated 20-25 tons of molten corium through peripheral fuel assemblies in the east side of the vessel, as well as through the core barrel assembly (CBA) at the periphery of the core. This paper presents the results of an analyses carried out to investiga… more
Date: January 1, 1988
Creator: Anderson, J. L. & Sienicki, J. J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Modeling of thermal hydraulic behavior and fission product releases in degraded cores

Description: When core material reaches melting conditions severe degradation of the core geometry occurs. Data available on the core behavior in a severely degraded state suggest that extensive blockage of the flow channels would occur. If a sufficient bypass is available for the gas flow, such as in the LOFT LP-FP-2 test, severe retardation of the hydrogen and fission product sources from the degraded channel is suggested from the available data. This phenomena is expected to occur in an LWR core and shou… more
Date: January 1, 1988
Creator: Sharon, A.; Ellison, P.G.; Henry, R.E.; Kenton, M.A. & Hammersley, R.J. (Fauske and Associates, Inc., Burr Ridge, IL (USA))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Timing of the Three Mile Island Unit 2 core degradation as determined by forensic engineering

Description: Unlike computer simulation of an event, forensic engineering is the evaluation of recorded data and damaged as well as surviving components after an event to determine progressive causes of the event. Such an evaluation of the 1979 Three Mile Island Unit 2 accident indicates that gas began accumulating in steam, generator A at 6:10, or 130 min into the accident and, therefore, fuel cladding ruptures and/or zirconium-water reactions began at that time. Zirconium oxidation/hydrogen generation rat… more
Date: January 1, 1988
Creator: Henrie, J.O. (Hydrogen Control, Inc., Panguitch, UT (USA))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Progress in understanding of direct containment heating phenomena in pressurized light water reactors

Description: Progress is described in development of a mechanistic understanding of direct containment heating phemonena arising during high-pressure melt ejection accidents in pressurized water reactor systems. The experimental data base is discussed which forms the basis for current assessments of containment pressure response using current lumped-parameter containment analysis methods. The deficiencies in available methods and supporting data base required to describe major phenomena occurring in the rea… more
Date: January 1, 1988
Creator: Ginsberg, T. & Tutu, N.K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Overview of the CONTAIN LMR code

Description: The use of containments in proposed US Liquid Metal Reactors (LMR) will require that a computational tool be available for safety analysis. As a result of an international cooperative effort, such a tool has been produced in the CONTAIN-LMR computer code. This is a version of the CONTAIN code that has been enhanced with extra capabilities for LMR applications. CONTAIN is the NRC's best-estimate code for the evaluation of the conditions that may exist inside a reactor containment building during… more
Date: January 1, 1988
Creator: Carroll, D.E.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

Description: Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temper… more
Date: January 1, 1988
Creator: Hyman, C.R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Results of EPRI/ANL DCH investigations and model development

Description: The results of a series of five experiments are described addressing the severity and mitigation of direct containment heating. The tests were performed in a 1:30 linear scale mockup of the Zion PWR containment system using a reactor-material corium melt consisting of 60% UO/sub 2/, 16% ZrO/sub 2/, 24% SSt at nominally 2800C initial temperature. A ''worst-case'' type test involving unimpeded corium dispersal through an air atmosphere in a closed vessel produced an atmosphere heatup of 323K, equ… more
Date: January 1, 1988
Creator: Spencer, B. W.; Sienicki, J. J.; Sehgal, B. R. & Merilo, M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Evaluation of selected samples from the TMI-2 core

Description: Core-bore samples from the K9 and N12 locations in the TMI-2 core were examined for microstructural and microchemical features. The purpose of the examinations was twofold, first to determine core temperatures at known elevations in the core, and second to obtain insight into materials interactions that lead to core degradation. The temperature at the /approximately/50-cm elevation in the N12 location, a control rod position, was estimated to have been /approximately/960/degree/C and dropping s… more
Date: December 1, 1988
Creator: Liu, Y. Y.; Neimark, L. A. & Jackson, W. D.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Containment and release management

Description: Reducing the risk from potentially severe accidents by appropriate accident management strategies is receiving increased attention from the international reactor safety community. Considerable uncertainty still surrounds some of the physical phenomena likely to occur during a severe accident. The USNRC, in developing its research plan for accident management, wants to ensure that both the developers and implementers of accident management strategies are aware of the uncertainty associated with … more
Date: January 1, 1988
Creator: Lehner, J.R. & Pratt, W.T.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Experimental studies on melt spreading, bubbling heat transfer, and coolant layer boiling

Description: Melt spreading studies have been undertaken to investigate the extent to which molten core debris may be expected to spread under gravity forces in a BWR drywell geometry. The objectives are to determine the extent of melt spreading as a function of melt mass,melt superheat, and water depth. These studies will enable an objective determination of whether or not core debris can spread up to and contact containment structures or boundaries upon vessel failure. Results indicate that the most impor… more
Date: January 1, 1988
Creator: Greene, G. A.; Finfrock, C.; Klages, J.; Schwarz, C. E. & Burson, S. B.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Fuel relocation mechanism based on microstructures of debris

Description: Argonne National Laboratory (ANL) has performed a number of examinations to determine the microstructure and micro-chemistry of samples of debris from the TMI-2 reactor. These examinations have been a small part of the overall effort to gain an understanding of the TMI-2 accident. As a result of these overall efforts, a general scenario of the response of the core components has been established. In this paper we will describe the microstructure and micro-chemistry of debris from the lower plen… more
Date: May 1, 1988
Creator: Strain, R. V.; Neimark, L. A. & Sanecki, J. E.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Accident Management for Severe Accidents

Description: The management of severe accidents in light water reactors is receiving much attention in several countries. The reduction of risk by measures and/or actions that would affect the behavior of a severe accident is discussed. The research program that is being conducted by the US Nuclear Regulatory Commission focuses on both in-vessel accident management and containment and release accident management. The key issues and approaches taken in this program are summarized. 6 refs.
Date: January 1, 1988
Creator: Bari, R. A.; Pratt, W. T.; Lehner, J.; Leonard, M.; Disalvo, R. & Sheron, B.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Spreading of molten corium in Mk I geometry following vessel meltthrough

Description: A one-dimensional, multicell, Eulerian computer code is under development to predict the gravity-driven spreading dynamics and thermal interactions of a molten corium layer flowing horizontally over a concrete substrate. The code is compared to recent experiments in which molten mixtures of iron and aluminum oxide flowed over concrete in the presence and absence of water. Results are presented from scoping calculations for the Mk I BWR system investigating the spreading-induced penetration imme… more
Date: January 1, 1988
Creator: Sienicki, J. J.; Farmer, M. T. & Spencer, B. W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Low-pressure cutoff for melt dispersal from reactor cavities

Description: A series of debris dispersal experiments using a 1/42-scale model of the Surry cavity without internal structure was performed. The experimental data showed that in addition to the Kutateladze number (based upon the area averaged gas velocity), the fraction of melt simulant retained within the cavity was also dependent upon the hole size and the density of the driver gas. A correlation that fit all the data reasonably in the region of the debris dispersal threshold was used to develop a low-pre… more
Date: January 1, 1988
Creator: Tutu, N.K.; Ginsberg, T. & Finfrock, C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Results of fission product release from intermediate-scale MCCI (molten core-concrete interaction) tests

Description: A program of reactor-material molten core-concrete interaction (MCCI) tests and related analyses are under way at Argonne National Laboratory under sponsorship of the Electric Power Research Institute (EPRI). The particular objective of these tests is to provide data pertaining to the release of nonvolatile fission products such as La, Ba, and Sr, plus other aerosol materials, from the coupled thermal-hydraulic and chemical processes of the MCCI. The first stages of the program involving small … more
Date: January 1, 1988
Creator: Spencer, B. W.; Thompson, D. H.; Fink, J. K.; Gunther, W. H. & Sehgal, B. R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

Description: High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that th… more
Date: September 1, 1988
Creator: Pilch, M.; Tarbell, W.W. & Brockmann, J.E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Analysis of the thermal response of a BWR Mark-I containment shell to direct contact by molten core materials

Description: This study was undertaken to evaluate the thermal response of a BWR Mark-I containment shell in the event of an accident severe enough for molten core materials to fall into the cavity beneath the rector vessel and eventually come into direct contact with the shell. An existing ORNL three-dimensional transient heat transport computer code, HEATING-6, was used for a specific 2-D case (and variations) for which representative melt/shell boundary conditions required as input were available from ot… more
Date: January 1, 1988
Creator: Kress, T.S. & Cleveland, J.C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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