• Completed baseline tests on the Mod 3 facility to establish performance parameters of the system in evaluating different phases of a LOCA. Nine published NRC reports summarized these tests. • Tested the capability of a prototype optical probe to characterize water mixing behavior in the vessel downcomer. It was successful.

• Completed and published the results of studies scaling Semiscale MOD-3 to PWR, LOFT to PWR and Semiscale MOD-1 to LOFT. Results confirmed the rationale earlier employed in scaling LOFT from PWR and Semiscale from LOFT. • Published the report on Semiscale tests and analyses performed in support of the Three Mile Island evaluation effort.

Separate-Effects Experiments

Major test facilities for NRC's separate effects experiments are the Thermal Hydraulic (Blowdown) Test Facility (THTF) at Oak Ridge National Laboratory, Tennessee; FLECHT SEASET* at Westinghouse in Pittsburgh; the Two-Loop-TestApparatus (TLTA) and Sector Steam Test Facility (SSTF) at General Electric in California, and the steam-water mixing facilities at Battelle Memorial Institute in Ohio. Bench-scale tests and instrument development programs also are being conducted at several university laboratories. A table summarizing the location and capabilities of these facilities is on p. 151, 1977 NRC Annual Report.

Two Loop Test Apparatus (TLTA). TLTA tests during 1979 showed that the injection of emergency core cooling water cools the bundle during the early phases of the postulated accident more effectively than previously believed. Studies of later phases of the LOCA and tests of other, more probable accident sequences are planned for 1980 and beyond.

Steam Water Mixing Tests. The studies of steamwater mixing effects on the penetration of cooling water in models of PWR vessels conducted over the past five years in the small 1/15 and 2/15 scale models at Battelle Columbus Laboratories and at Creare, Inc. were largely completed during fiscal year 1979. A major result has been the accumulation of additional evidence of the conservatism in models used in the licensing process. Knowledge gained from these smallscale tests now will be used in the planning and analysis of full-scale penetration tests to be conducted in the Upper Plenum Test Facility in Germany. (See "3D Program," under "Research Support," later in this chapter.)

• Full Length Emergency Cooling Heat Transfer Separate Effects/System Effects Tests.

Counter Current Flow Limit (CCFL) Refill/Reflood Program. This project, sponsored jointly by NRC, the Electric Power Research Institute (EPRI), and General Electric, was initiated in 1979 and features the Sector Steam Test Facility (SSTF), a full-scale model of a 30 degree sector of a boiling water reactor (BWR). Tasks undertaken to date include investigations of the distribution of cooling water sprayed over the top of a core and how that cooling water penetrates fuel bundles. A modeling effort is in process to aid the development of a BWR version of the TRAC code (see "Code Development," below).

FLECHT-SEASET Program. This program is described in detail on p. 184, 1978 NRC Annual Report. In 1979, Westinghouse completed separate effects studies of both steam generator and core behavior during reflood, and continued the studies of core blockage using a small 21-rod bundle. The latter program will be expanded to use a 161-rod facility to investigate bypass as well as blockage geometry, and to permit comparisons of results with core blockage data from the Cylindrical Core Test Facility in Japan (see "3D Program," below). These tests should clarify some uncertainties regarding the conservatism of heattransfer-rate and flow-regime criteria now used in licensing regulations. The system components (core rod bundles and steam generators) which have been studied separately will be integrated to investigate system interactions during various post-accident cooling tests. These system tests will focus on heat transfer mechanisms that are important in both large and small-break LOCAs. Steam heat transfer data from the 161-rod facility were used in TMI-related activities.

PWR Blowdown Heat Transfer Program (BDHT). The PWR Blowdown Heat Transfer Program at Oak Ridge was redirected in 1979 toward gaining a better understanding of fluid conditions in a PWR core during slower depressurization such as that characteristic of TMI. After modifications, ORNL ran a series of tests with the electrically heated 7x7 fuel-rod bundle (Bundle No. 2) to simulate various transient conditions. Initial tests using Bundle No. 3 were run in December 1979. This bundle features much-improved instrumentation and should produce a considerably improved understanding of pressure, flow, power, and temperature behavior. Tests with Bundle No. 1 were described on p. 183, 1978 Annual Report.

Model Development Studies. NRC continues to use small scale tests to study the various phenomena associated with steam-water flow. Each test is directed toward a particular effect to produce improved models or better data for input to code calculations. In 1979, Argonne National Laboratory completed experiments on the effect of controlled oscillations on reflood heat transfer; Massachusetts Institute of

Technology completed work on spacer-grid effects on reflood and on natural circulation flows; Northwestern University continued its experiments on the rate of steam condensation, and Brookhaven National Laboratory continued its study of vapor generation rates in depressurization situations. In addition, Rensselaer Polytechnic Institute and Argonne National Laboratory undertook the development of models dealing with void fractions in reactor coolant flow during accidents, and the State University of New York at Stony Brook began work on models for the amount and character of droplet flow in and above a reactor core.

Instrument Development. Test facilities require many sophisticated instruments which are not commercially available. NRC's advanced instrumentation program, described in the 1977 and 1978 Annual Reports, is designed to fill this void. Progress during 1979 was reported in July at the Reactor Safety Instrumentation Review Group Meeting (NUREG/CP-0007). A summary of that report

follows:

A technique called "pulsed neutron activation" was developed by Argonne National Laboratory to measure the velocity and density of steam/water flow. The technique will be used as a standard calibration for other instruments. Sandia Laboratories developed a portable neutron generator as part of the U.S. contribution to the international "3D" program (see below). Flow-measuring instruments, including turbine, drag disk and gamma densitometers have been developed and improved by INEL, and impedance

A July 1979 progress report on the NRC advanced instrumentation program described the pulsed neutron activation (PNA) technique for measuring velocity and density of steam-water flow. The technique involves: activating fluid with a neutron pulse from a portable neutron source (left); detecting activated nuclei a known distance downstream, and plotting the transit-time; determining average mass flow velocity and the average density of the fluid, and deriving mass flow from the latter two averages. The torus detector (right) features sodium iodide scintillators mounted to photomultiplier tubes. The plot at center is reproduced from PNA measurement of an air-water mixture. PNA equipment has been installed at the Semiscale and the LOFT facilities, and will be used in tests with other devices in 1980.

probes have been developed by ORNL. Other advanced instrumentation for flow measurements include optical probes, ultrasonic densitometers, and a stagnation probe, developed by INEL for measurements in LOFT. Rensselaer Polytechnic Institute, State University of New York at Stony Brook, Lehigh University, and Northwestern University continued their development of void fraction probes, film probes, special thermocouples, laser doppler techniques, and pitot tubes to better measure the thermal hydraulic properties of water and steam under accident conditions. So far, all of this advanced instrumentation technology has been applied in research facilities. Evaluation of its applicability to commercial power plants continued in 1979 and into 1980, particularly in light of an important lesson learned from the Three Mile Island accident-the need for instrumentation to measure water levels in reactor cores.

FUEL BEHAVIOR

NRC's fuel behavior research programs in 1979 included cladding experiments, in-reactor tests, fuel meltdown and fission product transport tests, and fuel code development. These are described below. Data produced in these programs in previous years contributed significantly to improved understanding and analyses of the TMI-2 accident, particularly in the areas of oxidation, embrittlement and ballooning of the cladding and the release of fission gases from the fuel. As technical data from post-TMI analyses became available, the research staff began to recast some of those programs toward the study of fuel behavior when fuel rods are uncovered or severely damaged. In some tests, fuel assemblies will be allowed to boil dry to ascertain whether or how best they can be cooled. In others, the release and transport of fission products from damaged fuel will be studied. Some of the facilities and techniques which will be used for these TMI-related projects are described in the following summary of ongoing fuel-behavior research activities.

Cladding Experiments

Multirod Burst Test (MRBT) Program. The MRBT program at ORNL has two main objectives. The first is to better define the deformation behavior of unirradiated Zircaloy cladding under conditions postulated for a loss-of-coolant accident (LOCA). The second is to provide a data base for use in assessing geometrical changes which occur in fuel rod cladding and the extent of coolant flow restrictions that the changes (such as ballooning and rupture) might produce.

Data accumulated in the MRBT program have come from unheated-shroud single-rod tests and from heated-and unheated-shroud multirod (4 x 4 bundle)

Tube burst test equipment at the Battelle Columbus Laboratory in Ohio is used to determine changes in fuel rod cladding burst properties resulting from irradiation in a commercial power reactor. Above, technician (seated) operates the tube burst console which programs, reads and records test parameters, while a technician (right) uses remote manipulators to load a radioactive specimen into the test stand in a hot cell. Below, a specimen is mounted in one of two test fixtures. A furnace for conducting elevated temperature burst tests is shown above the test specimen.

tests in which the shroud temperature at the time of burst was no closer than 80°C from the fuel bundle temperature, and in which the rods were not restricted in their outward movement. The importance of closely simulating these thermal and confinement factors is not clear at this time, but the data base will be increased by investigating these concerns in tests. One conclusion that has held up throughout the testing is that, at a given temperature, the temperature gradients determine the extent of deformation, i.e., the more uniform the temperature distribution, the greater and more uniform the deformation.

Mechanical Properties of Zircaloy. Work in 1979 at Argonne National Laboratory produced information which will lead to better criteria for predicting damage to cladding embrittled by oxidation at high temperatures. These new criteria are stated in terms of measurable mechanical properties which can be used to define the oxidation limits that will permit the cladding to survive such conditions as thermal shock, impact and physical damage, and the loads on cladding resulting from the interaction of reflooding water and

steam.

In-Reactor Testing

Power Burst Facility (PBF). At the PBF in Idaho, research continued in 1979 on the behavior of fuel rods under various operating and accident conditions. (See page 154, 1977 Annual Report.) The "reactivity initiated accident" (RIA), described on p. 188, 1978 Annual Report involves a burst of power generated when a control rod is ejected from the core. Early in fiscal year 1979, two RIA experiments were performed in the PBF test reactor. In the first test, two fresh fuel rods and two pre-irradiated fuel rods were exposed to a power burst that in earlier experiments had done

substantial structural damage to the test fuel. In some regions of the pre-irradiated rods the fuel swelled more than expected, and the swollen fuel and debris from adjacent parts of a rod blocked the flow shroud more than expected.

The second PBF RIA test was performed with four pre-irradiated rods previously exposed to a lower power burst near the level at which cladding could be expected to fail and to release fission products. Only one of the four rods failed, but a series of small longitudinal cladding cracks appeared which resembled the kind of pellet-cladding interaction which usually induces cladding failure. Since this rod had not been opened after pre-irradiation, and the companion rods had been, these effects will be studied in future RIA damage threshold tests using unprocessed preirradiated rods.

Later in the year, three LOCA blowdown tests were performed to confirm that planned LOFT tests with unpressurized rods could be run without jeopardizing the fuel bundles.

The final two PBF tests of 1979 used both fresh and pre-irradiated rods, pressurized to match the fill gas pressures of commercial rods when new and at or near the end of life, to examine the "ballooning" associated with cladding rupture with the peak temperatures in the 1050°K-1500°K (1430°F-2240°F) range. In one test, circumferential ballooning as high as 50 percent was observed. (NOTE: More than 70 percent ballooning of four adjacent rods is required to block a flow channel in a commercial power reactor, and some cooling water flow through any cladding ruptures is possible even then.) The second test was performed in the last months of fiscal year 1979, and the rods had not been examined at year's end.

Fuel Meltdown/Fission Product
Release and Transport

Fission Product Release and Transport research continued at two laboratories:

• Oak Ridge National Laboratory completed an experimental program to determine the quantity and type of fission products which might escape during reactor accidents. In four hightemperature tests, segments of irradiated commercial fuel rods were subjected to peak temperatures (1200° to 1600°C) well above those predicted for successful control of a LOCA. These tests, conducted in a flowing-steam environment, indicated that the release of fission products iodine and cesium increased tenfold and of krypton by a factor of five within the temperature range 1350°C to 1400°C. The results of these tests were helpful in estimating the fuel temperatures attained during the TMI-2 accident.

• Battelle Columbus Laboratories published a user's manual for the TRAP-MELT code, which models the transport behavior of fission products in primary coolant systems of LWRS in various accident conditions, including fuel meltdown. Core-Melt Research. Sandia Laboratories continued investigating the potential for thermal explosions, if

A combined experimental/analytic program at Argonne National Laboratory, aimed at better understanding the effects of fission gas releases on irradiated fuel, employs this direct electric heating apparatus (left) and predictive models of the GRASS family of computer codes. In the experimental effort, fuel pellets are heated to temperatures similar to nuclear heating, then studied in various ways, including the use of scanning electron microscope photography. The photograph reveals an interconnected network of tunnels on the grain boundaries of irradiated UO, fuel.

molten core materials were to contact water, with a series of experiments dealing with the efficiency of converting the thermal energy of the melt into mechanical energy. In a series of 48 tests using molten iron/alumina with masses up to 27Kg, the maximum efficiency was measured at 1.34 percent, a factor of about 20 less than the maximum theoretical efficiency for thermal interactions.

Another Sandia Laboratories investigation explored the interaction of molten core materials and concrete, producing important data on the gases and aerosols generated, the penetration rate of the melt into the concrete, and the rate of fission product evolution from the melt. This information was used to develop an advanced melt/concrete computer code called CORCON.

Fuel Behavior Codes

Fuel Rod Analysis Program (FRAP). Information from the PBF, LOFT (see above) and Halden Reactor Project (see p. 189, 1978 Annual Report) programs is used in developing and assessing NRC codes "FRAPCON," used for the steady-state analysis of fuel rod response during normal reactor operation, and FRAP