NRC programs at the Idaho and Oak Ridge National Laboratories include the development of instrument systems for international (3D) activities in Germany and Japan. Some instruments and testing activities under these programs are: (1) An in-core guidetube impedance measurement assembly developed at Oak Ridge for the PKL facility in West Germany. (2) Testing the assembly on a steam-water test stand at Oak Ridge. (3) Instrumented spoolpiece, featuring 11 transducers, developed by EG&G in Idaho for the PKL. Four such devices have been sent to Germany and eight to Japan for the Cylindrical Core Test Facility. (4) Design development and calibration of instrumentation for the German Upper Plenum Test Facility (UPTF) entails testing in a "see-through" representation of the UPTF at Oak Ridge. A similar facility is used at EG&G. (5) A video camera can be coupled to the EG&G Imaging Optical Probe to record the flow of high-temperature and pressure steam-water in nuclear reactor piping. Engineering Center at Canoga Park, California. Results appeared to confirm the conservatism of the present model (i.e., that the flow through the valve is greater than predicted in the NRC analysis model). However, tests may still be needed. Program Support As part of NRC's participation in the 3D international program (see p. 199, 1978 Annual Report), researchers at the Idaho National Engineering Laboratory and the Oak Ridge National Laboratory began the loan of advanced instruments to West German and Japanese counterparts for use in their facilities, and scientists at Los Alamos began using the TRAC code to analyze those test facilities. During 1979, the Japanese completed the large 2000-rod electrically heated Cylindrical Core Test Facility, and checkout tests have been completed. Work continues in Japan on the Slab Core Test Facility, expected to be operational in 1980. Both facilities will provide improved understanding of steam/water behavior in a simulated nuclear core. In Germany, researchers continued design work on the Upper Plenum Test Facility, which will experimentally model the behavior of steam and water droplets in the upper plenum of a PWR during refill and reflood stages of a postulated LOCA. This cooperative program provides NRC with better understanding of the physics of accident behavior and an advanced accident-prediction code (TRAC) at about one-third the cost of doing the whole job alone. Technical Support Under the Technical Support Program, NRC shares sponsorship with the Department of Energy of the Nuclear Safety Information Center at Oak Ridge and the National Energy Software Center at Argonne in Illinois. Nuclear Safety Information Center. The Nuclear Safety Information Center (NSIC) at Oak Ridge provides a focal point for safety information on reactors and other nuclear facilities. Technical experts who are cognizant of the literature in each area of specialty provide replies to questions from NRC, DOE and the nuclear community. Information is provided to nonexempt customers on a cost recovery basis. Seventeen reports, in addition to the bimonthly review, Nuclear Safety, were published in 1979. The NSIC also gave significant support to the ACRS, and its consultants during its review of the Licensee Event Reports. National Energy Software Center. The National Energy Software Center (NESC) at Argonne National Laboratory is partially funded by NRC to make NRCsponsored computer codes available to the public. Between October 1978 and September 1979 the Center distributed 1180 copies of the software packages in response to requests from NRC and DOE offices, the Nuclear Energy Agency Data Bank in France, other U.S. government agencies, universities, and commercial and industrial organizations. On September 30, 1979, the NESC list of software packages available for distribution contained 45 items (codes) specifically sponsored by NRC. Water Reactor Safety Information Meeting. The NRC held its sixth Water Reactor Safety Research Information Meeting November 6-9, 1978, at the National Bureau of Standards, Gaithersburg, MD. One hundred sixteen papers were presented, including 12 invited papers dealing with foreign safety research programs. More than 700 persons attended including 175 foreign visitors. Advanced Reactor Research NRC's Advanced Reactor Safety Research program focuses on two reactor concepts: High Temperature Gas-cooled Reactors (HTGR), and Liquid Metalcooled Fast Breeder Reactors (LMFBR). Fiscal year 1979 activities in each of these areas are outlined below. HIGH TEMPERATURE GAS-COOLED The President's fiscal year 1980 budget eliminated funds for gas-cooled reactor research, and NRC programs in that area were targeted for termination in the first several months of that year. Some tasks had been discontinued by mid-1979. Four DOE laboratories were affected by the terminations. To assure an orderly termination and provide all possible allowances for possible future resumption of the research, guidelines were provided for actions in which: (1) items of particular significance to the operation of the Fort Saint Vrain reactor (FSV) in Colorado were identified; (2) distinctions were made between programs which would be costly to discontinue in terms of data-loss, restart expertise, etc., and those which could be resumed fairly easily, and (3) the impact on contractor personnel would be minimized, with assurances that cadres of expertise can be maintained at each laboratory in case of a decision to resume the research. Within these guidelines, a number of programs could be carried out, although at very low levels of activity. Some examples: the metals and graphite programs were kept going at Brookhaven National Laboratory; transient analysis and seismic core modeling of Fort Saint Vrain continued at Los Alamos; low-level efforts continued at Oak Ridge on FSV-related heat transfer, and at Battelle's Pacific Northwest Laboratory (PNL) on graphite inspection techniques. Other than these greatly curtailed activities, however, NRC research in advanced reactor concepts will be confined to the development of computer codes and models for future use in safety investigations. LIQUID METAL-COOLED FAST The LMFBR program is subdivided into five areas: analysis, safety test facilities, materials interactions, aerosol release and transport, and systems integrity. Progress during 1979 in each of these sub-programs is discussed below. Analysis Program Most of the work in this area was performed at Argonne, Brookhaven, Los Alamos, and Sandia Laboratories, as follows: Argonne National Laboratory completed an analysis of critical experiments dealing wth LMFBR safety using the VIM Monte Carlo code and the Zero Power Reactor-9 (ZPR-9) facility. Results should aid in validating the neutronics computer codes used in LMFBR accident analysis. Other ANL code work resulted in improved calculational efficiencies (by factors up to five) using the COMMIX (Component Mixing) and BODYFIT-1 (Boundary-Fitted Transformation) codes. ANL work in the cooperative studies with EURATOM and the United Kingdom featured calculations with the SAS3D/EPIC code to quantify consequences of various accident phenomena, as well as a study of fuel-pin behavior. At Brookhaven National Laboratory, work on the "Super System Code" (SSC) continued during 1979 (see p. 172, 1977 NRC Annual Report), and a version modeling the Fast Flux Test Facility reactor (FFTF) was completed. Startup tests planned for the FFTF were precalculated and will be compared with operating data when it becomes available next year. Los Alamos Scientific Laboratory's analysis program on hypothetical core disruptive accidents (HCDA) in breeder reactors has been shifted from rapid, energetic accidents to those that develop more slowly. A key concern in such hypothetical accidents is the transition phase in which the core begins to melt and core materials begin to move. LASL completed the first consistent analysis of the transition phase using the SIMMER computer code. SIMMER combines calculations of neutronics, fluid dynamics, thermodynamics and the interactions of these factors with one another during an accident. Also the portion of the code that calculates neutronics phenomena was completely revised and the revised code was tested against experiments performed at Argonne. The fluid dynamics and thermodynamic models of SIMMER had been tested earlier with generally good results. The code was made available in 1979 for use in the United Kingdom, Germany, and the European Economic Community Research Center in Italy. Sandia Laboratories completed a preliminary version of a computer code called CONTAIN, for use in analyzing the responses of advanced reactor containment systems to postulated accident threats. The code will compute the structural and radiological interactions when core material drops from a primary reactor vessel onto the containment floor, and will assess the character of the residual mass. The University of Arizona completed the BRENDA code used for dynamic similation of transients in loop type LMFBRs. Safety Test Facilities Following the upgrading of the Annular Core Research Reactor (ACRR) at Sandia Laboratories, NRC's safety test facility work has consisted of installing a new diagnostics system in ACRR, and of implementing the ACRR-CABRI collaboration in fastreactor safety experiments. (Both activities were described in the 1978 NRC Annual Report, p. 202.) Materials Interactions Experiments and analytical model development on the energetics of severe accidents and on the meltthrough potential of post-accident core debris continued at Sandia Laboratories in 1979. Analysis of previous ACRR experiments on fuel pellets under accident conditions showed that rapid fuel swelling resulted from fission gas production, and this could not be explained by existing analytical models. As a result, new models of this phenomenon were developed and an improved ACRR series of experiments on fuel disruption (FD-2) was begun. Experiments on the disruption of irradiated fuel and its sweep-out from the core by coolant in LMFBR accidents will use the ACRR's new fuel-motion diagnostics system. A possible spin-off from this work is a suggested series of tests on light water reactor fuel to examine details of fuel failure in conditions such as TMI-type accidents. In Prompt-Burst Energetics work at Sandia, experiments on the damage potential of severe power excursions (prompt bursts) were resumed in the upgraded ACRR. In these experiments an LMFBR fuel pin contained in a sodium-filled capsule is placed in the ACRR experiment cavity and exposed to an intense, short burst of neutrons that melts and may even partially vaporize the fuel. The resulting pressure and mechanical damage potential are measured and used in constructing analytical models for assessing the threat of such power excursions to the integrity of reactor vessels and piping. The experiments are showing considerably less damage potential than previously considered possible. Aerosol Release and Transport Tests of sodium/uranium oxide aerosols in the Nuclear Safety Pilot Plant (NSPP) at Oak Ridge (see p. Mixed Oxide Aerosol Concentration - Run 301 As part of the Aerosol Release and Transport program at Oak Ridge, uranium-oxide (U,O,) aerosols and sodium-oxide (Na,O) aerosols are mixed to permit study of their interactions and behavior as a function of time. (The U,O, aerosols are produced by burning uranium in a plasma torch and the Na,O aerosols by burning sodium in pools or sprays.) Mixing in the Nuclear Safety Pilot Plant (NSPP) permits study of behavior under conditions in which both aerosols coexist in the secondary containment and are presumed to act together. The experiment diagramed here involved the introduction of Na,O aerosols into an existing concentration of U,O, aerosol, causing a marked increase in the U,O, removal rate due to gravitational settling. The experiment tended to confirm that sodium oxide aerosols decrease the concentration of nuclear fuel (uranium) aerosols in a vessel under postulated LMFBR accident conditions. Another study at ORNL involves the transport of UO/sodium aerosols through overlying sodium. This study uses the Fuel Aerosol Simulant Test (FAST) facility, and trial tests using water instead of sodium to establish facility characteristics were under way at the end of 1979. Analytical assistance is provided to this program by the University of Virginia. Systems Integrity Debris-bed behavior modeled in the first three inreactor tests at Sandia Laboratories, using the Annular Core Research Reactor (see p. 176, 1977 Annual Report and p. 204, 1978 Annual Report) was used in coolability analyses of the Three Mile Island accident. Following that work, a fourth test (of a planned 16-test matrix) dealing with coolability as a function of sub-cooling was performed, with results indicating that self-rearrangement of the debris bed enhances its coolability. A special series of large-scale sodium concrete interaction tests was completed in support of the NRC's final safety evaluation of the Fast Flux Test Facility (FFTF). The results of the test confirmed the staff position regarding FFTF containment margins. Another test program was initiated to study the interaction of molten fuel materials with candidate materials that could be used in place of concrete to contain core debris from a postulated core meltdown. A large fuel melt test facility is under construction at Sandia. The facility will be used to conduct tests containing up to one half ton of molten fuel in contact with structural materials to confirm analytical methods for predicting containment system response under postulated accident conditions. General Reactor NRC's General Reactor Safety Research comprises three areas: site safety research, mechanical engineering research, and structural engineering research. These were described in the 1978 NRC Annual Report (see pp. 206-208). SITE SAFETY RESEARCH Site safety research is generic research directed toward estimating the effects on nuclear facilities of earthquakes, floods, and tornadoes and other severe phenomena, understanding the distribution of those severe natural phenomena, and providing information on meteorology affecting the atmospheric dispersion of radionuclides under postulated accident conditions. Geology and Seismology A magnitude-4 earthquake occurred about six miles from the Maine Yankee nuclear power plant near Wiscasset, Maine on April 18, 1979. Following that event, Boston College and the Maine State Geologist's Office cooperated in installing a dense array of portable seismographs to record aftershocks which would locate the causative fault. The aftershock pattern suggested a north-south trend clustered around the epicenter of the main shock. The last aftershock was recorded on June 21, 1979. In the Charleston, S.C., region, high-precision vertical seismic reflection profiling has revealed a probable fault in deep subsurface rock near the center of the large 1886 earthquake. (See p. 206, 1978 Annual Report.) This is the first direct evidence of a possible causative geologic structure for that earthquake. Studies are continuing to determine the extent, history of movements, and tectonic relationships of the fault in order to assess its potential earthquake hazard. In other 1979 activities under this program: • A new arrangement for direct cooperation with the Canadian Department of Energy, Mines and Resources will add 12 high quality seismograph stations to "look" southward into areas of interest in the Northeastern U.S. This also will improve U.S. capabilities to evaluate earthquake regions along the St. Lawrence River and in the northwest extension of the problematic "BostonOttawa seismic trend. • Summaries and interpretations of known data bearing on earthquake hazard assessment in areas of the Northeastern U.S., in the New Madrid, Mo., region and in the midcontinent region were published during 1979. • Studies of the response of soil foundations to earthquake motions are important to earthquake design of power plants and other structures. In 1979 NRC-supported studies of the foundations of important accelerograph stations were completed. Other geotechnical studies resulted in publication of a technical manual describing equipment and operations for determining dynamic soil properties in place. Meteorology and Hydrology Projects in this research field included the following: Severe Storms. Damage surveys of the December 4, 1978 tornado that struck Bossier City, La. and the tornado that devastated Wichita Falls, Tx., on April 10, 1979, were performed. Six 750-pound wide-flange steel beams, 18- and 24-ft. long, were hurled up to 300 yards in the Bossier City tornado, two of them penetrating the ground about eight feet. The most significant aspect of the Wichita Falls tornado was its size-up to one mile wide and more than 40 miles long. Information compiled from these surveys provide authoritative data against which to evaluate the design criteria developed for nuclear power plants and fuel cycle facilities. Flooding. A research program was initiated in 1979 to quantify the safety margins used in flood-related design criteria for nuclear power plants. Flood probabilities, as a function of geographic location, and with particular reference to coastal phenomena, will be determined. A numerical simulation of the November 1975 tsunami in Hawaii was completed during the year, and the storm surge and wave height associated with the passage of Hurricane David along the east coast of Florida in September were measured. Atmospheric Diffusion. The NRC-supported atmospheric dispersion research program featured fullscale field tests and wind tunnel simulations of building-wake-dispersion characteristics; the use of gaseous tracer and lidar technologies to measure vertical diffusion over different terrains, and determinations of thermal performance of cooling and fixedspray ponds used as heat sinks. Planning also began late in the year for a comprehensive field and modeling program to study atmospheric diffusion in a complex shoreline environment. MECHANICAL ENGINEERING NRC's new Mechanical Engineering Research Program, initiated in 1978, provides the licensing staff with improved methods for evaluating the safety and structural integrity of systems, components, and equipment under normal and accident conditions in terms of margins of safety and probabilities of failure. Major sub-programs are: Seismic Safety Margins Research Program (SSMRP). This multi-discipline program at Lawrence Livermore Laboratory is designed to estimate the conservatisms in the seismic safety requirements stated in the NRC licensing standard review plan, and to improve those requirements. The approach is to develop a probabilistic methodology that can realistically estimate the behavior of buildings and components of a nuclear plant during an earthquake. The first phase of the program will be completed in 1980. Nonlinear System Modeling Program. A simplified computer code for the analysis of piping systems was completed; a mathematical model of a simplified mechanical system (typical of those in a nuclear plant) was validated; and design charts were issued for use in describing the motion of mechanical equipment, piping and components. There is a need for further research to better characterize the dynamic response of valves in nuclear plants and to better model and scale mechanical systems and equipment. PARET Program. PARET is a systems identification technique to determine frequencies, mode shapes and |