Publications

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MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications.

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications.

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light-water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories (SNL) for the U.S. Nuclear Regulatory Commission (NRC) as a second-generation plant risk assessment tool and the successor to the Source Term Code package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include thermal-hydraulic response in the reactor coolant system (RCS), reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; and fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 2.0, released to users in September 2008. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR User's Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package. Volume 3, MELCOR Assessment Problems, presents a portfolio of test and sample problems consisting of both analyses of experiments and of full plant problems. These analyses will be repeated with future releases of MELCOR in order to provide a metric on code predictions as new versions are released.

In an effort to improve the current state of the art in fire probabilistic risk assessment methodology, the U.S. Nuclear Regulatory Commission, Office of Regulatory Research, contracted Sandia National Laboratories (SNL) to conduct a series of scoping tests to identify thermal and mechanical probes that could be used to characterize the zone of influence (ZOI) during high energy arc fault (HEAF) testing. For the thermal evaluation, passive and active probes were exposed to HEAF-like heat fluxes for a period of 2 seconds at the SNL<U+2019>s National Solar Thermal Test Facility to determine their ability to survive and measure such an extreme environment. Thermal probes tested included temperature lacquers (passive), NANMAC thermocouples, directional flame thermometers, modified plate thermometers, infrared temperature sensors, and a Gardon heat flux gauge. Similarly, passive and active pressure probes were evaluated by exposing them to pressures resulting from various high-explosive detonations at the Sandia Terminal Ballistic Facility. Pressure probes included bikini pressure gauges (passive) and pressure transducers. Results from these tests provided good insight to determine which probes should be considered for use during future HEAF testing.

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Three large open pool fire experiments involving a calorimeter the size of a spent fuel rail cask were conducted at Sandia National Laboratories Lurance Canyon Burn Site. These experiments were performed to study the heat transfer between a very large fire and a large cask-like object. In all of the tests, the calorimeter was located at the center of a 7.93-meter diameter fuel pan, elevated 1 meter above the fuel pool. The relative pool size and positioning of the calorimeter conformed to the required positioning of a package undergoing certification fire testing. Approximately 2000 gallons of JP-8 aviation fuel were used in each test. The first two tests had relatively light winds and lasted 40 minutes, while the third had stronger winds and consumed the fuel in 25 minutes. Wind speed and direction, calorimeter temperature, fire envelop temperature, vertical gas plume speed, and radiant heat flux near the calorimeter were measured at several locations in all tests. Fuel regression rate data was also acquired. The experimental setup and certain fire characteristics that were observed during the test are described in this paper. Results from three-dimensional fire simulations performed with the Cask Analysis Fire Environment (CAFE) fire code are also presented. Comparisons of the thermal response of the calorimeter as measured in each test to the results obtained from the CAFE simulations are presented and discussed.

For certification, packages used for the transportation of plutonium by air must survive the hypothetical thermal environment specified in 10CFR71.74(a)(5). This regulation specifies that 'the package must be exposed to luminous flames from a pool fire of JP-4 or JP-5 aviation fuel for a period of at least 60 minutes.' This regulation was developed when jet propellant (JP) 4 and 5 were the standard jet fuels. However, JP-4 and JP-5 currently are of limited availability in the United States of America. JP-4 is very hard to obtain as it is not used much anymore. JP-5 may be easier to get than JP-4, but only through a military supplier. The purpose of this paper is to illustrate that readily-available JP-8 fuel is a possible substitute for the aforementioned certification test. Comparisons between the properties of the three fuels are given. Results from computer simulations that compared large JP-4 to JP-8 pool fires using Sandia's VULCAN fire model are shown and discussed. Additionally, the Container Analysis Fire (CAFE) code was used to compare the thermal response of a large calorimeter exposed to engulfing fires fueled by these three jet propellants. The paper then recommends JP-8 as an alternate fuel that complies with the thermal environment implied in 10CFR71.74.

This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety analysis capabilities of the advanced-reactor community for directly relevant scenarios. Beyond the focus on the thermally-interacting and smoldering sodium pool fires, experimental and analysis capabilities for sodium spray fires have also been developed in this project.

The objective of this work is to perform an uncertainty quantification (UQ) and model validation analysis of simulations of tests in the cross-wind test facility (XTF) at Sandia National Laboratories. In these tests, a calorimeter was subjected to a fire and the thermal response was measured via thermocouples. The UQ and validation analysis pertains to the experimental and predicted thermal response of the calorimeter. The calculations were performed using Sierra/Fuego/Syrinx/Calore, an Advanced Simulation and Computing (ASC) code capable of predicting object thermal response to a fire environment. Based on the validation results at eight diversely representative TC locations on the calorimeter the predicted calorimeter temperatures effectively bound the experimental temperatures. This post-validates Sandia's first integrated use of fire modeling with thermal response modeling and associated uncertainty estimates in an abnormal-thermal QMU analysis.

Four large-scale open pool fire experiments were performed with well-characterized boundary and initial conditions. Results presented include a general description of test observations, wind measurements, fire plume topology, fuel recession and heat release rates, incident heat flux to the pool, surrounding terrain, and calorimeters. All initial and boundary condition data required as necessary inputs to computation models are also presented. The large physical scale, the experimental design, the use of independent measurement techniques, and the attention to data quality provide a unique dataset to support numerical fire model validation. Copyright © 2008 International Association For Fire Safety Science.

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A series of fire benchmark water suppression tests were performed that may provide guidance for dispersal systems for the protection of high value assets. The test results provide boundary and temporal data necessary for water spray suppression model development and validation. A review of fire suppression in presented for both gaseous suppression and water mist fire suppression. The experimental setup and procedure for gathering water suppression performance data are shown. Characteristics of the nozzles used in the testing are presented. Results of the experiments are discussed.

Four Well-Characterized Open Pool fires were conducted by Fire Science and Technology Department. The focus of the Well-Characterized Open Pool fire series was to provide environmental information for open pool fires on a physics first principal basis. The experiments measured the burning rate of liquid fuel in an open pool and the resultant heat flux to a weapon-sized object and the surrounding environment with well-characterized boundary and initial conditions. Results presented in this report include a general description of test observation (pre- and post-test), wind measurements, fire plume topology, average fuel recession and heat release rates, and incident heat flux to the pool and to the calorimeters. As expected, results of the experiments show a strong correlation between wind conditions, fuel vaporization (mass loss) rate, and incident heat flux to the fuel and ground surface and calorimeters. Numerical fire simulations using both temporally- and spatially-dependant wind boundary conditions were performed using the Vulcan fire code. Comparisons of data to simulation predictions showed similar trends; however, simulation-predicted incident heat fluxes were lower than measured.

The measurement of heat flux in hydrocarbon fuel fires (e.g., diesel or JP-8) is difficult due to high temperatures and the sooty environment. Un-cooled commercially available heat flux gages do not survive in long duration fires, and cooled gages often become covered with soot, thus changing the gage calibration. An alternate method that is rugged and relatively inexpensive is based on inverse heat conduction methods. Inverse heat-conduction methods estimate absorbed heat flux at specific material interfaces using temperature/time histories, boundary conditions, material properties, and usually an assumption of one-dimensional (1-D) heat flow. This method is commonly used at Sandia.s fire test facilities. In this report, an uncertainty analysis was performed for a specific example to quantify the effect of input parameter variations on the estimated heat flux when using the inverse heat conduction method. The approach used was to compare results from a number of cases using modified inputs to a base-case. The response of a 304 stainless-steel cylinder [about 30.5 cm (12-in.) in diameter and 0.32-cm-thick (1/8-in.)] filled with 2.5-cm-thick (1-in.) ceramic fiber insulation was examined. Input parameters of an inverse heat conduction program varied were steel-wall thickness, thermal conductivity, and volumetric heat capacity; insulation thickness, thermal conductivity, and volumetric heat capacity, temperature uncertainty, boundary conditions, temperature sampling period; and numerical inputs. One-dimensional heat transfer was assumed in all cases. Results of the analysis show that, at the maximum heat flux, the most important parameters were temperature uncertainty, steel thickness and steel volumetric heat capacity. The use of a constant thermal properties rather than temperature dependent values also made a significant difference in the resultant heat flux; therefore, temperature-dependent values should be used. As an example, several parameters were varied to estimate the uncertainty in heat flux. The result was 15-19% uncertainty to 95% confidence at the highest flux, neglecting multidimensional effects.