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Iodine transport analysis in the ESBWR.

Description: A simplified ESBWR MELCOR model was developed to track the transport of iodine released from damaged reactor fuel in a hypothesized core damage accident. To account for the effects of iodine pool chemistry, radiolysis of air and cable insulation, and surface coatings (i.e., paint) the iodine pool model in MELCOR was activated. Modifications were made to MELCOR to add sodium pentaborate as a buffer in the iodine pool chemistry model. An issue of specific interest was whether iodine vapor removed from the drywell vapor space by the PCCS heat exchangers would be sequestered in water pools or if it would be rereleased as vapor back into the drywell. As iodine vapor is not included in the deposition models for diffusiophoresis or thermophoresis in current version of MELCOR, a parametric study was conducted to evaluate the impact of a range of iodine removal coefficients in the PCCS heat exchangers. The study found that higher removal coefficients resulted in a lower mass of iodine vapor in the drywell vapor space.
Date: March 1, 2009
Creator: Kalinich, Donald A.; Gauntt, Randall O.; Young, Michael Francis & Longmire, Pamela
Partner: UNT Libraries Government Documents Department

Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

Description: Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment.
Date: December 31, 1995
Creator: Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V. & Xiang, J.Y.
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

Development of the BWR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments

Description: The objectives of the Boiling Water Reactor Experimental Analysis and Model Development for Severe Accidents (BEAMD) Program at the Oak Ridge National Laboratory (ORNL) are: (1) the development of a sound quantitative understanding of boiling water reactor (BWR) core melt progression; this includes control blade and channel box effects, metallic melt relocation and possible blockage formation under severe accident conditions, and (2) provision of BWR melt progression modeling capabilities in SCDAP/RELAP5 (consistent with the BWR experimental data base). This requires the assessment of current modeling of BWR core melt progression against the expanding BWR data base. Emphasis is placed upon data from the BWR tests in the German CORA test facility and from the ex-reactor experiments [Sandia National Laboratories (SNL)] on metallic melt relocation and blockage formation in BWRs, as well as upon in-reactor data from the Annular Core Research Reactor (ACRR) DF-4 BWR test (conducted in 1986 at SNL). The BEAMD Program is a derivative of the BWR Severe Accident Technology Programs at ORNL. The ORNL BWR programs have studied postulated severe accidents in BWRs and have developed a set of models specific to boiling water reactor response under severe accident conditions. These models, in an experiment-specific format, have been successfully applied to both pretest and posttest analyses of the DF-4 experiment, and the BWR severe fuel damage (SFD) experiments performed in the CORA facility at the Kernforschungszentrum Karlsruhe (KfK) in Germany, resulting in excellent agreement between model prediction and experiment. The ORNL BWR models have provided for more precise predictions of the conditions in the BWR experiments than were previously available. This has provided a basis for more accurate interpretation of the phenomena for which the experiments are performed. The experiment-specific models, as used in the ORNL DF-4 and CORA BWR experimental analyses, also provide a basis for ...
Date: January 1, 1994
Creator: Ott, L. 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

Effects of neutron streaming and geometric models on molten fuel recriticality accidents

Description: A postulated fast reactor accident which has been extant for many years is a recriticality following partial or complete core melting. Independently of the cause or probability of such a situation, certain cases can be defined and some facets of the dynamic history of these cases can be described with more than enough accuracy for safety considerations. Calculations were made with the PAD code for systems with 10 vol percent voids and varying reactivity insertion rates. Additionally, two distinct geometric and equation of state models were investigated in conjunction with a model which accounted for possible neutron streaming reactivity effects. Significant results include fission and kinetic energy, temperatures and pressures. (auth)
Date: October 1, 1975
Creator: McLaughlin, T.P.
Partner: UNT Libraries Government Documents Department

Analysis of Postulated Core Meltdown of an SRP Reactor - Final Report

Description: An analysis was made to determine the consequences of a postulated accident in which the core of a Savannah River Plant reactor melts down following the loss of coolant. The study was made to determine (1) the potential damage to the reactor building that could impair its integrity for confining activity and (2) the need for additional facilities to prevent the activity confinement system from being overheated by the decay heat in the debris. A preliminary report on this analysis was issued previously. The sequence of events during and following the loss of coolant has now been studied in more detail, and a computer program has been written and used to investigate transient heating effects. This is the final report of the analysis and presents the conclusions.
Date: October 1, 1970
Creator: Durant, W.S. & Brown, R.J.
Partner: UNT Libraries Government Documents Department

Interpretation of the XR2-1 experiment and characteristics of the BWR lower plenum debris bed

Description: The Ex-Reactor (XR) experiments have been conducted to advance the understanding of BWR severe accident melt progression events. The XR2-1 experiment addresses the fate of the initial large (code-predicted) movements of molten metals from the upper core to the lower core and core plate region. For this question, which has ramifications for blockage formation in the core region, the XR2-1 test results provide significant and perhaps definitive insights. Nevertheless, some events that occurred during this test are creatures of the special features of the test apparatus, and there is a potential for misconceptions with respect to the direct applicability of some of the results. This paper describes the conclusions that can be drawn from the XR2-1 experiment results and identifies those areas (such as fuel pellet stack collapse and core plate integrity) where care must be taken not to misconstrue the test events. Another important area where much recent work has been performed is the effort to analyze the potential for maintaining core debris within the reactor vessel lower plenum by cooling of the outer vessel wall. One of the first steps in such an analytical endeavor is to attempt to establish the pattern of energy transfers into the wall inner surface. As a prerequisite to determination of this pattern, it is necessary to first consider the nature of the debris within the lower plenum. Too often is an easily represented homogeneous circulating liquid pool incorporated without adequate consideration of the true material conditions. Basic considerations of the relative quantities of materials present, the potentials for eutectics formations, and the associated melting points dictate otherwise. This paper offers some insights as to the true nature of the lower plenum debris and discusses the need for some relatively simple experiments that would contribute much toward the basic understanding necessary for accurate ...
Date: November 1997
Creator: Hodge, S. A. & Ott, L. 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

Single tube meltdown incident

Description: In connection with design of rear face fittings for the plant-expansion study currently being conducted we have been asked to determine if rear face pressurization is required for safety reasons and if so, how much. Pressurization of the rear face piping would be used to provide sufficient reverse flow to prevent process tube burnout in the event of complete loss of coolant supply to a single tube by virtus of a front connector failure. Consideration of the effectiveness of rear face pressurization, however, requires a more general look at the problem of single tube meltdown than that provided by considering front fitting failure alone.
Date: February 1, 1960
Creator: Trumble, R.E.
Partner: UNT Libraries Government Documents Department

Information-theoretic approach to uncertainty importance

Description: A method is presented for importance analysis in probabilistic risk assessments (PRA) for which the results of interest are characterized by full uncertainty distributions and not just point estimates. The method is based on information theory in which entropy is a measure of uncertainty of a probability density function. We define the relative uncertainty importance between two events as the ratio of the two exponents of the entropies. For the log-normal and log-uniform distributions the importance measure is comprised of the median (central tendency) and of the logarithm of the error factor (uncertainty). Thus, if accident sequences are ranked this way, and the error factors are not all equal, then a different rank order would result than if the sequences were ranked by the central tendency measure alone. As an illustration, the relative importance of internal events and in-plant fires was computed on the basis of existing PRA results.
Date: January 1, 1985
Creator: Park, C.K. & Bari, R.A.
Partner: UNT Libraries Government Documents Department

TMI-2 analysis using SCDAP/RELAP5/MOD3.1

Description: SCDAP/RELAP5/MOD3.1, an integrated thermal hydraulic analysis code developed primarily to simulate severe accidents in nuclear power plants, was used to predict the progression of core damage during the TMI-2 accident. The version of the code used for the TMI-2 analysis described in this paper includes models to predict core heatup, core geometry changes, and the relocation of molten core debris to the lower plenum of the reactor vessel. This paper describes the TMI-2 input model, initial conditions, boundary conditions, and the results from the best-estimate simulation of Phases 1 to 4 of the TMI-2 accident as well as the results from several sensitivity calculations.
Date: November 1, 1994
Creator: Hohorst, J.K.; Polkinghorne, S.T.; Siefken, L.J.; Allison, C.M. & Dobbe, C.A.
Partner: UNT Libraries Government Documents Department