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A Heat Transfer Model for a Stratified Corium-metal Pool in the Lower Plenum of a Nuclear Reactor

Description: This preliminary design report describes a model for heat transfer in a corium-metal stratified pool. It was decided to make use of the existing COUPLE model. Currently available correlations for natural convection heat transfer in a pool with and without internal heat generation were obtained. The appropriate correlations will be incorporated in the existing COUPLE model. Heat conduction and solidification modeling will be done with existing algorithms in the COUPLE. Assessment of the new model will be done by simple energy conservation problems.
Date: August 1, 1999
Creator: Sohal, Manohar Singh & Siefken, Larry James
Partner: UNT Libraries Government Documents Department

Experimental observations of the breakup of multiple metal jets in a volatile liquid

Description: A postulated severe loss of coolant accident in a nuclear reactor can lead to significant core damage due to residual heat generation. Subsequently, melted core materials (i.e.; corium) could migrate downward and impinge upon the lower head of the reactor pressure vessel (RPV). During this relocation, the complexity of the reactor structure could segregate the molten corium into various flow paths. A perforated flow plate could readily provide the mechanism to segregate the molten corium. The resulting small (a few cm) diameter melt streams, driven by gravity, could then penetrate the remaining coolant in the RPV and cause any of the following events: impingement of the high temperature melt streams on the lower head could breach the RPV; re-agglomeration of the corium melt on the lower head could influence the coolability of the debris bed; {open_quotes}pre-mixing{close_quotes} of the melt streams with the coolant could lead to a vapor explosion; or, sufficient quenching of the melt streams by the coolant could produce a stabilized debris bed. An overview of the thermo-science issues related to core-melt accidents is presented by Theofanous. Thus, insight into the melt stream breakup mechanisms (i.e.; interfacial conditions, fragmentation, and geometric spacing) during the melt-coolant interactions is necessary to evaluate the plausibility, and characteristics, of these events. Molten Fuel Stream Breakup Simulation (MFSBS) experiments have been performed at Argonne National Laboratory in which simulant materials were used to determine a `boiling` jet breakup length correlation and to visualize the melt fragmentation mechanisms during the penetration of a single molten metal jet into a volatile liquid. The goal was to characterize the hydrodynamics of the corium-water interactions in a postulated core melt accident. The present experiment closely follows the procedures of the MFSBS.
Date: July 1, 1995
Creator: Marciniak, M.J.
Partner: UNT Libraries Government Documents Department

Characterization of jet breakup mechanisms observed from simulant of molten fuel penetrating coolant. Technical progress report, 1989--1990

Description: The objective of the proposed experiments is to replicate approximately, by injecting low melting point metal alloys into Freon-11 and liquid nitrogen, the dispersal of corium streams in water. To first gain a better understanding of the corium dispersal process to be simulated, experimental data from the CCM experiments, in which the injection of streams of molten corium into water was studied, was interpreted in cooperation with Argonne National Laboratory (ANL) staff. The results of these experiments are discussed briefly below. This is followed by a description of the preparations made to date for the present simulant experiments.
Date: December 31, 1990
Creator: Jones, B.G.
Partner: UNT Libraries Government Documents Department

A Heat Transfer Model for a Stratified Corium-Metal Pool in the Lower Plenum of a Nuclear Reactor

Description: This preliminary design report describes a model for heat transfer in a corium-metal stratified pool. It was decided to make use of the existing COUPLE model. Currently available correlations for natural convection heat transfer in a pool with and without internal heat generation were obtained. The appropriate correlations will be incorporated in the existing COUPLE model. Heat conduction and solidification modeling will be done with existing algorithms in the COUPLE. Assessment of the new model will be done by simple energy conservation problems.
Date: August 1, 1999
Creator: Sohal, M. S. & Siefken, L. J.
Partner: UNT Libraries Government Documents Department

Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-6 test data report : thermal hydraulic results, Rev. 0.

Description: The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure? (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium ({approx} {phi} 30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength is being addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus measures the fracture strength of the crust while it is either at room temperature or above, the latter state being achieved with a heating element placed below the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale ...
Date: June 28, 2011
Creator: Lomperski, S.; Farmer, M. T.; Kilsdonk, D. & Aeschlimann, B. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Electrometallurgical treatment of TMI-2 fuel debris

Description: Argonne National Laboratory (ANL) has developed an electrometallurgical treatment process suitable for conditioning DOE oxide spent fuel for long-term storage or disposal. The process consists of an initial oxide reduction step that converts the actinide oxides to a metallic form, followed by an electrochemical separation of uranium from the other fuel constituents. The final product of the process is a uniform set of stable waste forms suitable for long-term storage or disposal. The suitability of the process for treating core debris from the Three Mile Island-2 (TMI-2) reactor is being evaluated. This paper reviews the results of preliminary experimental work performed using simulated TMI-2 fuel debris.
Date: August 1, 1997
Creator: Karell, E.J.; Gourishankar, K.V. & Johnson, G.K.
Partner: UNT Libraries Government Documents Department

Status and future direction of the melt attack and coolability experiments (MACE) program at Argonne National Laboratory.

Description: The Melt Attack and Coolability Experiments (MACE) program has been underway at Argonne National Laboratory addressing the ability of water to quench and thermally stabilize a molten core concrete interaction (MCCI) when the interaction is flooded from above. In this program, which has been sponsored by the EPRI-headed Advanced Containment Experiments (ACE) international consortium, large scale reactor material integral effects experiments have been conducted, in parallel with related modeling efforts. Plans are currently being developed for continued utilization of the MACE facility under the sponsorship of the Nuclear Energy Agency (NEA) to achieve the following objectives: (i) resolution of the ex-vessel debris coolability issue through a redirected program which focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests; and (ii) address remaining uncertainties related to long-term two-dimensional MCCI under dry cavity conditions. In terms of the ex-vessel debris coolability issue, separate effects tests are planned to provide data on key melt coolability mechanisms identified in MACE integral effects tests. The results of these tests will provide both confirmatory evidence and test data to support development of validated models for extrapolation to plant conditions. In terms of dry cavity conditions, reactor material tests are planned to address remaining uncertainties related to long-term 2-D MCCI; in particular, lateral vs. axial power split. This paper describes the essential elements of the program to address these two remaining important LWR safety issues.
Date: February 2, 2001
Creator: Farmer, M.T.; Spencer, B.W.; Binder, J.L. & Hill, D.J.
Partner: UNT Libraries Government Documents Department

Quick look data report for COMET Test U2

Description: Investigations are underway at Forschungszentrum Karlsruhe (FZK) addressing methods to terminate and stabilize a core melt accident situation ex-vessel. In this approach, the molten core-concrete interaction (MCCI) begins erosion of the concrete, and after erosion proceeds to some modest depth, it exposes and unseals an array of tubes. The tubes are connected to a water reservoir pressurized by static water head. Upon unsealing, the tubes direct a flow of water into the bottom of the corium layer. The water is forced up through the melt, cooling the melt and causing it to solidify in a form that allows continued permeation and heat removal by the water. Thus, the accident progression can be halted, and the debris may be permanently cooled. The key aspect of the passive ex-vessel core retention approach described above is the ability of water injected at the bottom of a corium melt layer to quench the melt forming a coolable debris bed in the process. This process has been tested using iron-alumina thermite as a corium simulant with promising results. As a part of a collaborative research agreement between FZK and the US DOE, two scoping tests are being conducted at Argonne National Laboratory to test the FZK core retention concept using real reactor materials. The second of these two tests, denoted COMET Test U2, was successfully conducted on December 17, 1997. The objectives of this data report are to: summarize the experiment facility and operating procedure for COMET Test U2, and present the test data.
Date: January 8, 1998
Creator: Farmer, M. T.; Spencer, B. W.; Kilsdonk, D. J. & Aeschlimann, R. W.
Partner: UNT Libraries Government Documents Department

Experiments on Corium Dispersion after Lower Head Failure at Moderate Pressure

Description: Concerning the mitigation of high pressure core melt scenarios, the design objective for future PWRS is to transfer high pressure core melt to low pressure core melt sequences, by means of pressure relief valves at the primary circuit, with such a discharge capacity to limit the pressure in the reactor coolant system to less than 20 bar. Studies have shown that in late in-vessel reflooding scenarios there may be a time window where the pressure is indeed in this range, at the moment of the reactor vessel rupture. It has to be verified that large quantities of corium released from the vessel after failure at pressures <20 bar cannot be carried out of the reactor pit, because the melt collecting and cooling concept of future PWRs would be rendered useless. Existing experiments investigated the melt dispersal phenomena in the context of the DCH resolution for existing power plants in the USA, most of them having cavities with large instrument tunnels leading into subcompartments. For such designs, breaches with small cross sections at high vessel failure pressures had been studied. However, some present and future European PWRs have an annular cavity design without a large pathway out of the cavity other than through the narrow annular gap between the RPV and the cavity wall. Therefore, an experimental program was launched, focusing on the annular cavity design and low pressure vessel failure. The first part of the program comprises two experiments which were performed with thermite melt steam and a prototypic atmosphere in the containment in a scale 1:10. The initial pressure in the RPV-model was 11 and 15 bars, and the breach was a hole at the center of the lower head with a scaled diameter of 100 cm and 40 cm, respectively. The main results were: 78% of melt mass ...
Date: September 21, 1999
Creator: BLANCHAT,THOMAS K.; GARGALLO,M.; JACOBS,G.; MEYER,L. & WILHELM,D.
Partner: UNT Libraries Government Documents Department

CONTAIN code analyses of direct containment heating (DCH) experiments: Model assessment and phenomenological interpretation

Description: Models for direct containment heating (DCH) in the CONTAIN code for severe accident analysis have been reviewed and a standard input prescription for their use has been defined. The code has been exercised against a large subset of the available DCH data base. Generally good agreement with the experimental results for containment pressurization ({Delta}P) and hydrogen generation has been obtained. Extensive sensitivity studies have been performed which permit assessment of many of the strengths and weaknesses of specific model features. These include models for debris transport and trapping, DCH heat transfer and chemistry, atmosphere-structure heat transfer, interactions between nonairborne debris and blowdown steam, potential effects of debris-water interactions, and hydrogen combustion under DCH conditions. Containment compartmentalization is an important DCH mitigator in the calculations, in agreement with experimental results. The CONTAIN model includes partially parametric treatments for some processes that are not well understood. The importance of the associated uncertainties depends upon the details of the DCH scenario being analyzed. Recommended sensitivity studies are summarized that allow the user to obtain a reasonable estimate of the uncertainties in the calculated results.
Date: May 12, 1995
Creator: Williams, D.C.; Griffith, R.O.; Tadios, E.L. & Washington, K.E.
Partner: UNT Libraries Government Documents Department

Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

Description: The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications.
Date: April 26, 2005
Creator: Anderson, M.; Corradini, M.; Bank, K.Y.; Bonazza, R. & Cho, D.
Partner: UNT Libraries Government Documents Department

OECD MCCI project Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-1 test data report : thermal hydraulic results. Rev. 0 September 20, 2002.

Description: The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium ({approx}{phi}30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results ...
Date: May 23, 2011
Creator: Lomperski, S.; Farmer, M. T.; Kilsdonk, D. J.; Aeschlimann, R. W. & Basu, S.
Partner: UNT Libraries Government Documents Department

Drying tests conducted on Three Mile Island fuel canisters containing simulated debris

Description: Drying tests were conducted on TMI-2 fuel canisters filled with simulated core debris. During these tests, canisters were dried by heating externally by a heating blanket while simultaneously purging the canisters` interior with hot, dry nitrogen. Canister drying was found to be dominated by moisture retention properties of a concrete filler material (LICON) used for geometry control. This material extends the drying process 10 days or more beyond what would be required were it not there. The LICON resides in a nonpurgeable chamber separate from the core debris, and because of this configuration, dew point measurements on the exhaust stream do not provide a good indication of the dew point in the canisters. If the canisters are not dried, but rather just dewatered, 140-240 lb of water (not including the LICON water of hydration) will remain in each canister, approximately 50-110 lb of which is pore water in the LICON and the remainder unbound water.
Date: December 31, 1995
Creator: Palmer, A.J.
Partner: UNT Libraries Government Documents Department

Neutron transport in random media

Description: The survey reviews the methods available in the literature which allow a discussion of corium recriticality after a severe accident and a characterization of the corium. It appears that to date no one has considered the eigenvalue problem, though for the source problem several approaches have been proposed. The mathematical formulation of a random medium may be approached in different ways. Based on the review of the literature, we can draw three basic conclusions. The problem of static, random perturbations has been solved. The static case is tractable by the Monte Carlo method. There is a specific time dependent case for which the average flux is given as a series expansion.
Date: August 1, 1996
Creator: Makai, M.
Partner: UNT Libraries Government Documents Department

Status report on severe accident material property measurements

Description: Measurements of selected material properties of molten reactor core material (corium) were made. The corium used was a mixture of UO{sub 2}, ZrO{sub 2} and Zr, with oxygen content being a parameter to reflect different stages of zirconium oxidation. The mixtures used were representative of typical in-vessel melt sequences. For most measurements, the UO{sub 2}/ZrO{sub 2} mass ratio was 1.51, representative of VVER/440 melt compositions and melt compositions of most US BWRs. Measurements were made of the solidus/liquidus temperatures of corium compositions using a Differential Thermal Analysis technique. Observation of the solubility of unoxidized Zr in the oxide phase was made by metallographic analysis of solidus/liquidus melt samples. The results of laminar flow corium spreading tests in one dimension were used to estimate the viscosity of corium compositions. Measured solidus and liquidus temperatures for compositions representative of Zr oxidation of 30, 50 and 70% were compared with those obtained form a phase diagram provided by Kurchatov Institute. It was found that experimental measurements agreed well with the phase diagram values at 70% oxidation, but the measured solidus temperatures were higher than those on the phase diagram and the measured liquidus temperatures were lower than those on the phase diagram at 30 and 50% oxidation. From a microstructure examination it was determined that there was no global segregation into distinct metal and oxide phases during the cooldown of a sample in which there was initially 70% Zr oxidation. Therefore it is concluded that Zr metal is soluble in the oxide phase under molten conditions. Viscosity estimates were made for compositions representative of Zr oxidation of 30, 50 and 70% by fitting the results of spreading tests to Huppert`s equation. It was found that, at a temperature of 2500 C, the viscosity varied by three orders of magnitude over this range of ...
Date: June 1, 1997
Creator: Farmer, M.T.; McUmber, L.; Spencer, B.W. & Aeschlimann, R.W.
Partner: UNT Libraries Government Documents Department

Potential for AP600 in-vessel retention through ex-vessel flooding

Description: External reactor vessel cooling (ERVC) is a new severe accident management strategy that involves flooding the reactor cavity to submerge the reactor vessel in an attempt to cool core debris that has relocated to the vessel lower head. Advanced and existing light water reactors (LWRs) are considering ERVC as an accident management strategy for in-vessel retention (IVR) of relocated debris. In the probabilistic risk assessment (PRA) for the AP600 design, Westinghouse credits ERVC for preventing vessel failure during postulated severe accidents with successful reactor coolant system (RCS) depressurization and reactor cavity flooding. To support the Westinghouse position on IVR, DOE contracted the University of California--Santa Barbara (UCSB) to produce the peer-reviewed report. To assist in the NRC`s evaluation of IVR of core melt by ex-vessel flooding of the AP6OO, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform: An in-depth critical review of the UCSB study and the model that UCSB used to assess ERVC effectiveness; An in-depth review of the UCSB study peer review comments and of UCSB`s resolution method to identify areas where technical concerns weren`t addressed; and An independent analysis effort to investigate the impact of residual concerns on the margins to failure and conclusions presented in the UCSB study. This report summarizes results from these tasks. As discussed in Sections 1.1 and 1.2, INEEL`s review of the UCSB study and peer reviewer comments suggested that additional analysis was needed to assess: (1) the integral impact of peer reviewer-suggested changes to input assumptions and uncertainties and (2) the challenge present by other credible debris configurations. Section 1.3 summarized the corresponding analysis approach developed by INEEL. The remainder of this report provides more detailed descriptions of analysis methodology, input assumptions, and results.
Date: December 1, 1997
Creator: Rempe, J.L.; Knudson, D.L.; Allison, C.M.; Thinnes, G.L. & Atwood, C.L.
Partner: UNT Libraries Government Documents Department

IFCI 7.0 Models and Correlations

Description: The Integrated Fuel-Coolant Interaction Code (IFCI) is a best-estimate computer program for analysis of phenomena related to mixing of molten nuclear reactor core material with reactor coolant (water). The stand-alone version of the code, IFCI 7.0, has been designed for analysis of small- and intermediate-scale experiments in order to gain insight into the physics (including scaling effects) of molten fuel-coolant interactions. The code's methods, models, and correlations are being assessed. This report describes the flow regime, friction factor, and heat-transfer models used in the current version of IFCI (IFCI 7.0).
Date: May 1, 1999
Creator: Reed, A.W.; Schmidt, R.C. & Young, M.F.
Partner: UNT Libraries Government Documents Department

Resolution of the direct containment heating issue for all Westinghouse plants with large dry containments or subatmospheric containments

Description: This report uses the scenarios described in NUREG/CR-6075 and NUREG/CR-6075, Supplement 1, to address the direct containment heating (DCH) issue for all Westinghouse plants with large dry or subatmospheric containments. DCH is considered resolved if the conditional containment failure probability (CCFP) is less than 0.1. Loads versus strength evaluations of the CCFP were performed for each plant using plant-specific information. The DCH issue is considered resolved for a plant if a screening phase results in a CCFP less than 0.01, which is more stringent than the overall success criterion. If the screening phase CCFP for a plant is greater than 0.01, then refined containment loads evaluations must be performed and/or the probability of high pressure at vessel breach must be analyzed. These analyses could be used separately or could be integrated together to recalculate the CCFP for an individual plant to reduce the CCFP to meet the overall success criterion of less than 0.1. The CCFPs for all of the Westinghouse plants with dry containments were less than 0.01 at the screening phase, and thus, the DCH issue is resolved for these plants based on containment loads alone. No additional analyses are required.
Date: February 1996
Creator: Pilch, M. M.; Allen, M. D. & Klamerus, E. W.
Partner: UNT Libraries Government Documents Department

Nuclear analysis of the chornobyl fuel containing masses with heterogeneous fuel distribution.

Description: Although significant data has been obtained on the condition and composition of the fuel containing masses (FCM) located in the concrete chambers under the Chernobyl Unit 4 reactor cavity, there is still uncertainty regarding the possible recriticality of this material. The high radiation levels make access extremely difficult, and most of the samples are from the FCM surface regions. There is little information on the interior regions of the FCM, and one cannot assume with confidence that the surface measurements are representative of the interior regions. Therefore, reasonable assumptions on the key parameters such as fuel concentration, the concentrations of impurities and neutron poisons (especially boron), the void fraction of the FCM due to its known porosity, and the degrees of fuel heterogeneity, are necessary to evaluate the possibility of recriticality. The void fraction is important since it introduces the possibility of water moderator being distributed throughout the FCM. Calculations indicate that the addition of 10 to 30 volume percent (v/o) water to the FCM has a significant impact on the calculated reactivity of the FCM. Therefore, water addition must be considered carefully. The other possible moderators are graphite and silicone dioxide. As discussed later in this paper, silicone dioxide moderation does not represent a criticality threat. For graphite, both heterogeneous fuel arrangements and very large volume fractions of graphite are necessary for a graphite moderated system to go critical. Based on the observations and measurements of the FCM compositions, these conditions do not appear creditable for the Chernobyl FCM. Therefore, the focus of the analysis reported in this paper will be on reasonable heterogeneous fuel arrangements and water moderation. The analysis will evaluate a range of fuel and diluent compositions.
Date: October 14, 1998
Creator: Turski, R. B.
Partner: UNT Libraries Government Documents Department

CONTAIN code analyses of direct containment heating (DCH) experiments

Description: In some nuclear reactor core melt accidents, a potential exists for molten core debris to be dispersed into the containment under high pressure. Resulting energy transfer to the containment atmosphere can pressurize the containment. This process, known as direct containment heating (DCH), has been the subject of extensive experimental and analytical programs sponsored by the US Nuclear Regulatory Commission (NRC). DCH modeling has been a major focus for the development of the CONTAIN code. In support of the peer review, extensive analyses of DCH experiments were performed in order to assess the CONTAIN code`s DCH models and improve understanding of DCH phenomenology. The present paper summarizes this assessment effort.
Date: June 1, 1995
Creator: Williams, D.C.; Griffith, R.O.; Tadios, E.L. & Washington, K.E.
Partner: UNT Libraries Government Documents Department

Transient quenching of superheated debris beds during bottom reflood

Description: The experimental data suggest that for small liquid supply rate and low initial particle temperature, the bed quench process is a one-dimensional frontal phenomenon. The bed heat flux is constant during most of the duration of the quench period. The range of conditions which display one-dimensional frontal cooling characteristics is identified as the deep bed regime of bed quenching, and a limiting mathematical model was developed to describe the observed behavior. For large liquid supply rate and high initial bed temperature, the bed quench process is a complex phenomenon. Under these conditions, the bed heat flux displays a nonuniform time dependence. In order to characterize this shallow bed regime, it was necessary to develop a detailed transient model of the coolant-debris interaction. This model, while developed for the shallow bed regime, also applies to the deep bed regime. Numerical computations clearly demonstrate the importance of developing a general reliable model for the solid-fluid heat transfer coefficients.
Date: January 1, 1984
Creator: Tutu, N.K.; Ginsberg, T.; Klein, J.; Schwarz, C.E. & Klages, J.
Partner: UNT Libraries Government Documents Department