TREXR is an organized, searchable collection of information that describes the hundreds of experiments conducted on nuclear reactor fuels in the Transient Reactor Test (TREAT) facility beginning in 1960. The experiments generally investigated the response of nuclear fuel samples to severe conditions similar to those associated with reactor accidents. The TREAT reactor was specifically designed and operated to provide such conditions. Investigation, understanding, and predictability of the transient response of nuclear fuels is important in supporting the design and licensing of safe nuclear reactors.
The experiments performed in TREAT span a wide range of fuel types and exposure conditions. Correspondingly, the information provided by the collection of experiments is broad. The goal of TREXR development has been to provide a convenient means of extracting from that large body of information selected information that is pertinent to the particular interests of researchers, reactor designers, and nuclear regulatory authorities. Such interests often center on particular combinations of fuel types, reactor accident scenarios, and fuel-behavior phenomena. TREXR is designed to help its users find such specific information as provided via experiments previously performed in TREAT. TREXR is also useful in providing information useful in the planning, preparation, and execution of future experiments in TREAT.
The database takes account of the variety of types of information that were generated both in the experimental programs and in related analytical programs that applied the experimental data to general fuel behavior and reactor safety evaluations. The information falls into several categories that include:
With reactor-fuel developers, safety analysts, and future TREAT experimenters in mind, the database was developed in a way that will help provide answers to such questions as the following (stated generically):
To accomplish this, TREXR users are presented with four general categories of experiment-related information. Each category is sub-divided into one or two levels of sub-categories presenting options to help users develop queries which will quickly narrow the searches to the relevant experiments, the types of information being sought, the types of documents that contain the information, etc. The categories and sub-categories are shown below, with the number of sub-category options indicated in parentheses. These options are presented on the "Search for Tests" page.
The name of the experimenter organization is also included. The user can also search for a test directly by name.
Technical information pertaining to the experiments has typically been described in limited-distribution experiment reports, conference proceedings, published journal articles, technical meeting transactions, and unpublished information. Links to word-searchable electronic files are provided in TREXR for many of the documents. Because many of the documents have US federally-imposed or contractually-limited restrictions on their dissemination or are regulated by copyright considerations, access to many of the links is restricted to users having authorization to access those links.
Individual documents in the database document collection are characterized in terms of the types of information they contain, as listed below. The query options provided to users include these document-content categories, so that searches can be narrowed to documents that contain only the selected types of content.
The TREAT reactor, designed, built, and operated by Argonne National Laboratory at a site that is now part of the Idaho National Laboratory, is world-renowned in its capability and history of providing large, high-level, controlled bursts of neutrons sufficient to heat test-fuel samples at rates simulating reactor accidents. The test-fuel samples are generally located within robust containers positioned at the center of the TREAT reactor core. A computer-controlled power rise in the TREAT reactor core creates the neutron intensity ("flux") within the core needed to cause a rapid rise in the fission rate in test-fuel samples. The internal, fission heating within the test samples is programmed to cause test-sample heating representative of what might occur in the core of a certain type of reactor undergoing a type of severe off-normal or accident transient of concern. The TREAT core is designed to safely experience such power transients without damaging itself, even though the transients may cause great damage to the test-fuel sample in the experiment.
Information regarding the behavior of test-fuel samples during TREAT experiments is typically obtained by observations and measurements taken during the experiment and also by observations and measurements taken of the fuel sample after the experiment.
Most of the experiments performed in TREAT between 1960 (when the experimentation began) and 1994 (when TREAT utilization was suspended for more than 20 years) supported the development of sodium-cooled fast reactors (SFRs), although many tests on fuels for light-water reactors (LWRs) and other reactor types were also performed. The data from the tests were used to identify key phenomena and to guide the development of analytical models of the transient behavior of nuclear fuels, addressing such phenomena as micro-structural changes in fuel pins, spatial redistributions of fuel constituents, fuel-cladding interactions, cladding failure thresholds, gross fuel melting, pre- and post-failure fuel motions, fuel-coolant-interaction effects, and post-failure coolant-channel blockages as well as the complex inter-play among combinations of those phenomena. Fuel-samples tested ranged from single un-irradiated fuel pellets tested in water to bundles of seven highly-irradiated full-length SFR fuel pins tested in flowing sodium. In many cases, TREAT test results provided the accepted validation basis for published reactor safety analyses.
The characteristics of TREAT that make it superbly capable for a wide variety of transient nuclear fuel experiments are described below.
The reactor core consists of a 19x19 array of fuel and reflector assemblies, which are each nominally 10-cm (4-in.) square and 2.7-m (8.8-ft) long (Figure 1). The assemblies contain a 1.2-m (4-ft) active fuel region with 0.6-m (2-ft) reflector regions above and below. Fuel assemblies are removed from the core to accommodate emplacement of experiment hardware (typically loops or capsules). The TREAT reactor fuel is a dilute mixture of fine particles of highly-enriched UO2 dispersed in graphite and carbon. The graphite and carbon fuel matrix absorbs heat rapidly, providing an essentially instantaneous large, negative temperature coefficient of reactivity. The core is air cooled with the cooling system designed to remove the heat generated during steady-state operation or following transient operation.
Surrounding the array is a permanent graphite reflector 0.6-m (2-ft) thick, around which is a 1.5-m (5-ft) thick concrete biological shield. Concrete shielding blocks above and around the core can be removed or reconfigured as needed to permit lateral access to the core for experimental purposes. Limited access to the core is also available from the bottom (Figure 2).
Fast-moving control rods (called "transient rods," having travel speeds up to 350 cm/s) are used to generate power transients in the core. Power and energy generated in test-fuel samples are related to the reactor core power and energy using a parameter called the power coupling factor (PCF). The value of the PCF for a particular experiment strongly depends upon the neutronic characteristics of the test-fuel and in-core experiment vehicle, as well as upon the core loading and the positions of the control rods at the beginning of the power transient. Administrative limits on core temperature result in limits on test-fuel sample power and energy. For transients limited only by reactivity feedback due to core temperature rise, the maximum-allowed core energy and peak power are approximately 2.5-GJ and 19 GW, respectively.
The TREAT automatic reactor control system provides for open- and closed-loop computer control, allowing for the generation of a wide variety of power-transient shapes. The control system also provides real-time, automated decision-making processes controlled by communication between sensors in the test vehicle and the reactor control system. Examples of feedback control that have been used are (a) initiation of a planned coolant-flow coast-down in the experiment loop upon attainment of a specified reactor power or after a specified elapsed time during a transient and (b) initiation of a specified type of TREAT transient-rod motion at a specified time interval after detection of a major change in coolant flow rate in the experiment.
Various test-vehicle designs have been used in TREAT experiments, depending upon fuel-sample characteristics and experiment objectives. Some vehicles contained single pins in a dry (no coolant) or stagnant-coolant environment (Figures 3 and 4). Other vehicles provided for flowing coolant next to test samples. The latter vehicles are called loops and are typically are of two designs: "package" loops that fit entirely within a shielded handling cask and can be inserted and removed from the reactor as a single unit (Figure 5) and non-package loops where much of the loop hardware is located outside the reactor and must be connected and disconnected from the in-core part before and after an experiment (Figure 6).
Historically, experiments generally fell into two groups. "Phenomenological" tests were performed in dry or stagnant-coolant capsules and investigated basic physical processes occurring during rapid overheating of fuel samples (such as molten fuel-coolant interactions). "Integral" tests used flowing-coolant loops and investigated complex interactions among multiple simultaneously-occurring phenomena (such as during prototypic simulations of hypothetical accidents involving multi-pin geometry). Both unirradiated and pre-irradiated fuel samples were tested. Sample sizes varied from individual fuel pellets to seven-pin bundles of fuel pins having 1 m-high fuel zones. Although most tests were on sodium-bonded uranium-alloy fuel, various gas-bonded ceramic fuels, including uranium oxide, uranium sulfide, uranium carbide, and thorium-based oxide were also studied.
The central importance of fuel motion in transient fuel behavior studies led to the development of the TREAT fast-neutron-hodoscope soon after TREAT began operations. This instrument enables real time tracking of the location of fissile material within an experiment by counting collimated fission neutrons emitted from the fuel. The hodoscope provided the primary source of data used to validate certain fuel motion models.
Evaluation of new reactor designs and new fuel types in the future will require a suitable database regarding fuel behavior and reactor system safety performance. As has been the case throughout the history of reactor development, an essential component of the database will need to be generated through in-reactor transient testing. Data from such testing have been found to be important both to assist the development of analytical models and codes that describe fuel and core behavior, and later to validate those models and codes and to provide confirmatory evidence for licensing.
It is intended to continue expanding the TREXR database not only as additional documents pertaining to historic tests are discovered but also as additional TREAT experiments are performed in the future.
TREXR was developed by the Nuclear Engineering Division of Argonne National Laboratory, USA. Argonne National Laboratory's work was supported by the U.S. Department of Energy, Assistant Secretary for Nuclear Energy, Office of Nuclear Energy, under contract DE-AC02-06CH11357.
Contributors to the TREXR development include Art Wright, Paul Froehle, Carolyn Tomchik, Tanju Sofu, Katherine Hunnicutt, Heather Connaway, Aaron Oaks, Ted Bauer, Keith Bowers, Justin Brothers, Andrew Cartas, Kristin Kubelsky, Matthew McColgan, Stacy Mo, Travis Mui, David Rhodes, Eric Rose, Adam Stensland, and Brian Strebel.
*Figures on this page are from unpublished internal Argonne ARC program records.