Publications

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The U.S. Nuclear Regulatory Commission, in concert with industry, continues to explore the effects of fire on electrical cable and control circuit performance. The latest efforts, which are currently underway, are exploring issues related to fire-induced cable failure modes and effects for direct current (dc) powered electrical control circuits. An extensive series of small and intermediate scale fire tests has been performed. Each test induced electrical failure in copper conductor cables of various types typical of those used by the U.S. commercial nuclear power industry. The cables in each test were connected to one of several surrogate dc control circuits designed to monitor and detect cable electrical failure modes and effects. The tested dc control circuits included two sets of reversing dc motor starters typical of those used in motor-operated valve (MOV) circuits, two small solenoid-operated valves (SOV), one intermediate size (1-inch (25.4mm) diameter) SOV, a very large direct-acting valve coil, and a switchgear/breaker unit. Also included was a specialized test circuit designed specifically to monitor for electrical shorts between two cables (inter-cable shorting). Each of these circuits was powered from a nominal 125V battery bank comprised of 60 individual battery cells (nominal 2V lead-acid type cells with plates made from a lead-cadmium alloy). The total available short circuit current at the terminals of the battery bank was estimated at 13,000A. All of the planned tests have been completed with the data analysis and reporting currently being completed. This paper will briefly describe the test program, some of the preliminary test insights, and planned follow-on activities.

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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.

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Fire probabilistic risk assessment (PRA) methods utilize data and insights gained from actual fire events in a variety of ways. For example, fire occurrence frequencies, manual fire fighting effectiveness and timing, and the distribution of fire events by fire source and plant location are all based directly on the historical experience base. Other factors are either derived indirectly or supported qualitatively based on insights from the event data. These factors include the general nature and intensity of plant fires, insights into operator performance, and insights into fire growth and damage behaviors. This paper will discuss the potential methodology improvements that could be realized if more complete fire event reporting information were available. Areas that could benefit from more complete event reporting that will be discussed in the paper include fire event frequency analysis, analysis of fire detection and suppression system performance including incipient detection systems, analysis of manual fire fighting performance, treatment of fire growth from incipient stages to fully-involved fires, operator response to fire events, the impact of smoke on plant operations and equipment, and the impact of fire-induced cable failures on plant electrical circuits.

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This paper summarizes the results of a Phenomena Identification and Ranking Table (PIRT) exercise performed for nuclear power plant (NPP) fire modeling applications conducted on behalf of the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES). A PIRT exercise is a formalized, facilitated expert elicitation process. In this case, the expert panel was comprised of seven international fire science experts and was facilitated by Sandia National Laboratories (SNL). The objective of a PIRT exercise is to identify key phenomena associated with the intended application and to then rank the importance and current state of knowledge of each identified phenomenon. One intent of this process is to provide input into the process of identifying and prioritizing future research efforts. In practice, the panel considered a series of specific fire scenarios based on scenarios typically considered in NPP applications. Each scenario includes a defined figure of merit; that is, a specific goal to be achieved in analyzing the scenario through the application of fire modeling tools. The panel identifies any and all phenomena relevant to a fire modeling-based analysis for the figure of merit. Each phenomenon is ranked relative to its importance to the fire model outcome and then further ranked against the existing state of knowledge and adequacy of existing modeling tools to predict that phenomenon. The PIRT panel covered several fire scenarios and identified a number of areas potentially in need of further fire modeling improvements. The paper summarizes the results of the ranking exercise.

In power generating plants, switchgear provide a means to isolate and de-energize specific electrical components and buses in order to clear downstream faults, perform routine maintenance, and replace necessary electrical equipment. These protective devices may be categorized by the insulating medium, such as air or oil, and are typically specified by voltage classes, i.e. low, medium, and high voltage. Given their high energy content, catastrophic failure of switchgear by means of a high energy arcing fault (HEAF) may occur. An incident such as this may lead to an explosion and fire within the switchgear, directly impact adjacent components, and possibly render dependent electrical equipment inoperable. Historically, HEAF events have been poorly documented and discussed in little detail. Recent incidents involving switchgear components at nuclear power plants, however, were scrupulously investigated. The phenomena itself is only understood on a very elementary level from preliminary experiments and theories; though many have argued that these early experiments were inaccurate due to primitive instrumentation or poorly justified methodologies and thus require re-evaluation. Within the past two decades, however, there has been a resurgence of research that analyzes previous work and modern technology. Developing a greater understanding of the HEAF phenomena, in particular the affects on switchgear equipment and other associated switching components, would allow power generating industries to minimize and possibly prevent future occurrences, thereby reducing costs associated with repair and downtime. This report presents the findings of a literature review focused on arc fault studies for electrical switching equipment. The specific objective of this review was to assess the availability of the types of information needed to support development of improved treatment methods in fire Probabilistic Risk Assessment (PRA) for nuclear power plant applications.

This report documents the results of a Phenomena Identification and Ranking Table (PIRT) exercise performed at Sandia National Laboratories (SNL) as well as the experimental and modeling program that have been designed based on the PIRT results. A PIRT exercise is a structured and facilitated expert elicitation process. In this case, the expert panel was comprised of nine recognized fire science and aerosol experts. The objective of a PIRT exercise is to identify phenomena associated with the intended application and to then rank the current state of knowledge relative to each identified phenomenon. In this particular PIRT exercise the intended application was sodium fire modeling related to sodium-cooled advanced reactors. The panel was presented with two specific fire scenarios, each based on a hypothetical sodium leak in an Advanced Breeder Test Reactor (ABTR) design. For both scenarios the figure of merit was the ability to predict the thermal and aerosol insult to nearby equipment (i.e. heat exchangers and other electrical equipment). When identifying phenomena of interest, and in particular when ranking phenomena importance and the adequacy of existing modeling tools and data, the panel was asked to subjectively weigh these factors in the context of the specified figure of merit. Given each scenario, the panel identified all those related phenomena that are of potential interest to an assessment of the scenario using fire modeling tools to evaluate the figure of merit. Each phenomenon is then ranked relative to its importance in predicting the figure of merit. Each phenomenon is then further ranked for the existing state of knowledge with respect to the ability of existing modeling tools to predict that phenomena, the underlying base of data associated with the phenomena, and the potential for developing new data to support improvements to the existing modeling tools. For this PIRT two hypothetical sodium leak scenarios were evaluated for the ABTR design. The first scenario was a leak in the hot side of the intermediate heat transport system (IHTS) resulting in a sodium pool fire. The second scenario was a leak in the cold side of the IHTS resulting in a sodium spray fire.

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Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior.

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Under traditional fire PRA methods, operating experience was used primarily to support statistical analysis of fire frequencies for specific plant locations and/or specific classes of fire ignition sources. While this application of the data continues, recent efforts to improve fire PRA methods, tools, and data are drawing more widely on insights from operating experience. This paper will describe some of the ways in which operating experience is being used to support fire PRA development activities.

One effect noted during the March 1975 fire at the Browns Ferry plant is that fire-induced cable damage caused a range of unanticipated circuit faults including spurious reactor status signals and the apparent spurious operation of plant systems and components. Current USNRC regulations require that licensees conduct a post-fire safe shutdown analysis that includes consideration of such circuit effects. Post-fire circuit analysis continues to be an area of both technical challenge and regulatory focus. This paper discusses risk perspectives related to post-fire circuit analysis. An opening background discussion outlines the issues, concerns, and technical challenges. The paper then focuses on current risk insights and perspectives relevant to the circuit analysis problem. This includes a discussion of the available experimental data on cable failure modes and effects, a discussion of fire events that illustrate potential fire-induced circuit faults, and a discussion of risk analysis approaches currently being developed and implemented.

This report documents state-of-the-art methods, tools, and data for the conduct of a fire Probabilistic Risk Assessment (PRA) for a commercial nuclear power plant (NPP) application. The methods have been developed under the Fire Risk Re-quantification Study. This study was conducted as a joint activity between EPRI and the U. S. NRC Office of Nuclear Regulatory Research (RES) under the terms of an EPRI/RES Memorandum of Understanding [RS.1] and an accompanying Fire Research Addendum [RS.2]. Industry participants supported demonstration analyses and provided peer review of this methodology. The documented methods are intended to support future applications of Fire PRA, including risk-informed regulatory applications. The documented method reflects state-of-the-art fire risk analysis approaches. The primary objective of the Fire Risk Study was to consolidate recent research and development activities into a single state-of-the-art fire PRA analysis methodology. Methodological issues raised in past fire risk analyses, including the Individual Plant Examination of External Events (IPEEE) fire analyses, have been addressed to the extent allowed by the current state-of-the-art and the overall project scope. Methodological debates were resolved through a consensus process between experts representing both EPRI and RES. The consensus process included a provision whereby each major party (EPRI and RES) could maintain differing technical positions if consensus could not be reached. No cases were encountered where this provision was invoked. While the primary objective of the project was to consolidate existing state-of-the-art methods, in many areas, the newly documented methods represent a significant advancement over previously documented methods. In several areas, this project has, in fact, developed new methods and approaches. Such advances typically relate to areas of past methodological debate.

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Smoke can cause interruptions and upsets in active electronics. Because nuclear power plants are replacing analog with digital instrumentation and control systems, qualification guidelines for new systems are being reviewed for severe environments such as smoke and electromagnetic interference. Active digital systems, individual components, and active circuits have been exposed to smoke in a program sponsored by the U.S. Nuclear Regulatory Commission. The circuits and systems were all monitored during the smoke exposure, indicating any immediate effects of the smoke. The major effect of smoke has been to increase leakage currents (through circuit bridging across contacts and leads) and to cause momentary upsets and failures in digital systems. This report summarizes two previous reports and presents new results from conformal coating, memory chip, and hard drive tests. The report describes practices for mitigation of smoke damage through digital system design, fire barriers, ventilation, fire suppressants, and post fire procedures.

Smoke is known to cause electrical equipment failure, but the likelihood of immediate failure during a fire is unknown. Traditional failure assessment techniques measure the density of ionic contaminants deposited on surfaces to determine the need for cleaning or replacement of electronic equipment exposed to smoke. Such techniques focus on long-term effects, such as corrosion, but do not address the immediate effects of the fire. This document reports the results of tests on the immediate effects of smoke on electronic equipment. Various circuits and components were exposed to smoke from different fields in a static smoke exposure chamber and were monitored throughout the exposure. Electrically, the loss of insulation resistance was the most important change caused by smoke. For direct current circuits, soot collected on high-voltage surfaces sometimes formed semi-conductive soot bridges that shorted the circuit. For high voltage alternating current circuits, the smoke also tended to increase the likelihood of arcing, but did not accumulate on the surfaces. Static random access memory chips failed for high levels of smoke, but hard disk drives did not. High humidity increased the conductive properties of the smoke. The conductivity does not increase linearly with smoke density as first proposed; however, it does increase with quantity. The data can be used to give a rough estimate of the amount of smoke that will cause failures in CMOS memory chips, dc and ac circuits. Comparisons of this data to other fire tests can be made through the optical and mass density measurements of the smoke.

There now exist close to 20 years of history in the application of Probabilistic Risk Assessment (PRA) for the analysis of fire risk at nuclear power plants. The current methods are based on various assumptions regarding fire phenomena, the impact of fire on equipment and operator response, and the overall progression of a fire event from initiation through final resolution. Over this same time period, a number of significant fire incidents have occurred at nuclear power plants around the world. Insights gained from US experience have been used in US studies as the statistical basis for establishing fire initiation frequencies both as a function of the plant area and the initiating fire source.To a lesser extent, the fire experience has also been used to assess the general severity and duration of fires. However, aside from these statistical analyses, the incidents have rarely been scrutinized in detail to verify the underlying assumptions of fire PRAs. This paper discusses an effort, under which a set of fire incidents are being reviewed in order to gain insights directly relevant to the methods, data, and assumptions that form the basis for current fire PRAs. The paper focuses on the objectives of the effort, the specific fire events being reviews methodology, and anticipated follow-on activities.

This report provides an overview of the fire safety experiments performed under the sponsorship of the German government in the containment building of the decommissioned pilot nuclear power plant known as HDR. This structure is a highly complex, multi-compartment, multi-level building which has been used as the test bed for a wide range of nuclear power plant operation safety experiments. These experiments have included numerous fire tests. Test fire fuel sources have included gas burners, wood cribs, oil pools, nozzle release oil fires, and cable in cable trays. A wide range of ventilation conditions including full natural ventilation, full forced ventilation, and combined natural and forced ventilation have been evaluated. During most of the tests, the fire products mixed freely with the full containment volume. Macro-scale building circulation patterns which were very sensitive to such factors as ventilation configuration were observed and characterized. Testing also included the evaluation of selective area pressurization schemes as a means of smoke control for emergency access and evacuation stairwells.

This report presents the results of three fire endurance tests and one ampacity derating test set of the fire barrier system Thermo-Lag 330-1 Subliming Coating. Each test was performed using cable tray specimens protected by a nominal three-hour fire barrier envelope comprised of two layers of nominal 1/2 inch thick material. The fire barrier systems for two of the three fire endurance test articles and for the ampacity derating test article were installed in accordance with the manufacturer`s installations procedures. The barrier system for the third fire endurance test article was a full reproduction of one of the original manufacturer`s qualification test articles. This final test article included certain installation enhancements not considered typical of current nuclear power plant installations. The primary criteria for fire endurance performance evaluation was based on cable circuit integrity testing. Secondary consideration was also given to the temperature rise limits set forth in the ASTM E119 standard fire barrier test procedure. All three of the fire endurance specimens failed prematurely. Circuit integrity failures for the two fire endurance test articles with procedures-based installations were recorded at approximately 76 and 59 minutes into the exposures for a 6 inch wide and 12 inch wide cable tray respectively. Temperature excursion failures (single point) for these two test articles were noted at approximately 65 and 56 minutes respectively. The first circuit integrity failure for the full reproduction test article was recorded approximately 119 minutes into the exposure, and the first temperature excursion failure for this test article was recorded approximately 110 minutes into the exposure.

Sandia National Laboratories has, for several years, been engaged in the performance of both fire safety and electrical equipment qualification research under independent programs sponsored by the US Nuclear Regulatory Commission. Recent comparisons between electrical cable thermal damageability data gathered independently in these two efforts indicate that a direct correlation exists between certain of the recent cable thermal vulnerability information gathered under equipment qualification conditions and thermal damageability in a fire environment. This direct correlation allows for a significant expansion of the data base on estimated cable thermal vulnerability limits in a fire environment because of the wide range of cable types and products that have been evaluated as a part of the equipment qualification research. This paper provides a discussion of the basis for the derived correlation, and presents estimated cable thermal damage limits for a wide range of generic cable types and specific cable products. The supposition that a direct correlation exists is supported through direct comparisons of the test results for certain specific cable products. The proposed supplemental cable fire vulnerability data gained from examination of the equipment qualification results is presented. These results should be of particular interest to those engaged in the evaluation of fire risk for industrial facilities, including nuclear power plants.

Sandia National Laboratories has performed a series of experiments under the sponsorship of the US Nuclear Regulatory Commission (USNRC) to assess the fire performance of thermally aged electrical cables. Two measures of fire performance were evaluated, namely, (1) the vulnerability of cables to thermal damage and (2) the flammability of cables. In each case, direct comparisons were made between experiments involving unaged (i.e., new off the reel) cables and cables subjected to accelerated thermal aging. The results were evaluated from the perspective of fire risk. It was found that thermal aging did cause changes in the thermal damageability of the cables tested; however, the changes observed are not considered risk significant. Large-scale fire tests demonstrated a clear decrease in material flammability due to thermal aging. Thus, it was concluded that the use of cable thermal damage and flammability information based on the testing of unaged cable samples is an acceptable risk assessment practice. Indeed, in the case of flammability, this is a conservative practice.

This report documents the findings of an experimental investigation of the effects of thermal aging on the fire damageability of electric cables. Two popular types of nuclear qualified cables were evaluated. For each cable type, both unaged (i.e., new off the reel) and thermally aged samples were exposed to steady-state elevated temperature environments until conductor-to-conductor electrical shorting was observed. Plots of the time to electrical failure versus the exposure temperature were developed and thermal damage thresholds were determined. For one cable type, the thermally aged cables were less vulnerable to thermal damage than were the unaged samples as demonstrated by an increase in the thermal damage threshold for the aged samples, and an extended survival time at exposure temperatures above the damage threshold for aged samples compared to unaged samples. For the second cable, the threshold of thermal damage was lowered somewhat by the aging process, an indication of an increased vulnerability to thermal damage due to aging. However, for the higher temperature exposures, no statistical difference between the damage times for aged and unaged cable samples was noted. For both cable types, the changes in the thermal damage threshold observed were not considered significant in terms of fire risk. 4 refs., 9 figs., 8 tabs.

An investigation of the impact of thermal aging on the flammability of two common types of nuclear grade electrical cables has been performed. Four large-scale flammability tests were performed with each of the two cable types tested in both an unaged (i.e., new off the reel) and a thermally aged (artificially aged) condition. In all cases, the fire was observed to consume virtually all of the combustible cable jacket and insulation material present. However, for both cable types tested, the thermal aging process caused a decrease in the cable flammability as demonstrated by decreases in the rate of fire growth, peak fire intensity, total heat released and near fire temperatures. This result is consistent with past cable aging studies because it has been observed that the thermal aging process will drive off certain of the more volatile constituents of a polymeric material. Presumably, when these aged materials are subjected to a fire, the evolution of volatile combustible gases is reduced as compared to the unaged materials, and hence, flammability is reduced. The results of these tests indicate that, at least for the two cable types tested, the evaluation of cable flammability using unaged cable samples will remain a conservative indicator of cable flammability in a thermally aged condition.

This report summarizes the results and conclusions generated by the US Nuclear Regulatory Commission sponsored Fire Protection Research Program at Sandia National Laboratories. Efforts conducted from the programs inception in 1975 through 1987 are discussed. The individual efforts are discussed within a framework based on specific areas of investigation. Early efforts are presented in the context of investigations of specific regulatory concerns. Later efforts are presented within the context of an integrated investigation of fire safety issues. This integrated approach considers the fire safety issue in terms of (1) source fire characterization, (2) detection and suppression system effectiveness, (3) room effects, (4) equipment response, and (5) room-to-room fire effects. The report provides a complete bibliography of reports and journal articles generated as a result of these efforts with a cross-reference listing of major reports to specific efforts. 98 refs., 23 figs., 20 tabs.

Nuclear power plants present the fire protection community with unique challenges. In addition to the traditional concerns of life safety and property loss prevention, nuclear safety analysts must also be concerned with the impact of fires on the safe operability of the nuclear reactor. Safe shutdown equipment must be protected from fire damage. When nuclear power plants were first designed and built, fire safety considerations were based primarily on the same criteria applied to general industrial facilities, primarily those concerning life safety and property loss prevention. This practice continued until 1975 when the Brown's Ferry nuclear reactor site experienced a severe cable tray fire. The fire burned for over seven hours, due in part to the reluctance of on-site personnel to use water on the fire for fear of shorting out critical electrical circuits. 4 figs.

The fire risk scoping study was sponsored by the US Nuclear Regulatory Commission and performed at Sandia National Laboratories. The study was initiated as a result of previous USNRC-sponsored fire research efforts that had identified certain fire risk issues that had not been addressed in previously completed commercial nuclear power plant fire risk analyses. The specific objectives of this study were: (1) to review and requantify fire risk scenarios from four fire probabilistic risk assessments in light of updated data bases made available as a result of USNRC-sponsored Fire Protection Research Program and updated computer fire modeling capabilities, (2) to identify potentially significant fire risk issues that have not been previously addressed in a fire risk context and to quantify the potential impact of those identified fire risk issues where possible, and (3) to review current fire regulations and plant implementation practices for relevance to the identified unaddressed fire risk issues. 9 refs., 3 tabs.