[en] The Large-Scale Seismic Test (LSST) Program at Hualien, Taiwan is a follow-on to the soil-structure interaction (SSI) experiments at Lotung, Taiwan. The planned SSI studies will be performed at a stiff soil site in Hualien, Taiwan that historically has had slightly more destructive earthquakes in the past than Lotung. The objectives of the LSST project is as follows: (1) to obtain earthquake-induced SSI data at a stiff soil site having similar prototypical nuclear power plant soil conditions; (2) to confirm the findings and methodologies validated against the Lotung soft soil SSI data for prototypical plant condition applications; (3) to further validate the technical basis of realistic SSI analysis approaches; and (4) to further support the resolution of USI A-40 Seismic Design Criteria issue. These objectives will be accomplished through an integrated and carefully planned experimental program consisting of: soil characterization, test model design and field construction, instrumentation layout and deployment, in-situ geophysical information collection, forced vibration test, and synthesis of results and findings

[en] The paper briefly discusses the background of soil-structure interaction research and identifies the nuclear industry's need for a realistic, experimentally qualified soil-structure interaction analysis methodology for nuclear power plant design to reduce excessive conservatism and stabilize the licensing process. EPRI research and joint research efforts between EPRI and Niagara Mohawk Power Corporation, Taiwan Power Company, and the Japanese Century Research Institute for Electric Power Industry are outlined. As a result of these and other research efforts, improvement in soil-structure interactions methodologies is being realized

[en] One aspect of source term consideration is the integrity of reactor containment. The key question to be answered is how will the containment fail given pressure and temperature histories associated with postulated, low probability, degraded core accident scenarios. The failure mode has significant ramifications. If the failure mode were a sudden gross one, such as a sudden rupture of containment wall, large quantities of fission products might be released to the environment immediately with potentially large public health effects. However, if the failure mode were some localized leakage which would lead to containment depressurization, the release of fission product would be gradual and limited and public health effects would be very small. To address this issue, a research program has been undertaken by EPRI at Construction Technologies Laboratories and Anatech International, Inc. The former has the testing responsibility, and the latter has the scope of developing an analytical model for failure mode prediction. This paper summarizes the efforts to date with emphasis on experimental findings

[en] One of the important problems in designing and predicting the responses of a fuel rod assembly is the understanding and quantification of the interactive forces and effects between fuel rods and their mechanical constraints. Consider a typical fuel rod assembly where fuel rods are laterally constrained by spring tabs. The precompression of the spring introduces a compressive force on the rod, which, in part, determines the wearing of the clad, the frictional effects, and the total response of the fuel rod assembly. While various analyses were carried out for this kind of problem, one important physical parameter has often been overlooked in the formulation. This parameter is related to creep and relaxation, and, in particular, the coupled behavior between them. (orig.)

[en] The US Nuclear Regulatory Commission and the Electric Power Research Institute have jointly sponsored a piping research project. The intent of this research effort, which involves design, analysis, fabrication, erection, and dynamic testing of two three-dimensional piping systems, is to provide the following benefits: (1) to expand the limited controlled data base on damping in piping systems at response levels at and above operating basis earthquake (OBE) stress levels; (2) to stimulate recognition of safety margins implicit in ASME code rules for class 2/3 piping by demonstrating the existence of large design margins in piping and support systems when subject to seimsic load far beyond its SSE design stress limit; and (3) to obtain a data base for benchmarking computer methods for analysis of pressurized piping systems with representative support systems and for response levels below and above piping yielding as well as including pipe support failure. In-depth study of the first of two planned test configurations is underway. The tests include: (1) a six-inch diameter main with no branch lines, and (2) a six-inch diameter main with three-inch branch lines

[en] In this paper, the most recent test results on piping response associated with pipe rupture, namely pipe whip impact and pipe rupture and depressurization, are reported. The tests performed do not represent any prototypical design configuration. The main objective was to generate a data base for validating the highly-nonlinear methodology required for realistic pipe rupture induced response evaluation and to quantify the conservatism in current simplified industry practice in analyzing pipe rupture related situations

[en] Containment and piping integrity research at EPRI is briefly described in the paper. The containment integrity research focuses on the determination of failure modes and load-carrying capabilities of concrete reactor containments under internal pressures beyond design basis. The objective of this research is to provide a more realistic basis for overall reactor risk studies. Results to date show potential of liner crack in discontinuity regions and stable crack propagation (with arrest) due to liner-concrete interaction. This finding supports the hypothesis that concrete containment will leak before break under severe core type of loading. The piping integrity research focuses on demonstrating and quantifying the true failure mechanism of dynamically loaded piping. Tests performed to date show that piping will fail in a fatigue rachetting manner rather than static plastic-collapse which is the basis of current ASME code equations. Additionally, test results suggest that at least an order of magnitude excess margin exists in the current standards. Based on these findings, improved rules for piping design and evaluation will be developed and proposed

[en] EPRI has sponsored an experimental program in the pipe whip impact and pipe rupture and depressurization areas. Sixteen pipe whip tests were performed with 3 in Schedule 80 (or 10) carbon steel pipes impacting on rigid target or concrete slab. The major testing parameters include distance, impact location, pipe rupture location, and concrete slab thickness and strength. The piping crushing at impact correlates with impact force and target response behavior. Conservatism was established by comparing measured and calculated impact forces. The pipe rupture and depressurization tests were carried out using 6 in stainless steel and carbon steel pipes under either PWR or BWR fluid conditions. These tests are of axial crack with initial machined-in surface flaw. It was found that pipe rupture would occur only if a long unstable through-wall crack was embedded in a sufficiently long unstable part-through crack (in the pipe wall). All other flaw configuration tested led to pipe leakage only. Reaction forces were measured which show conservatism of simplified method for fully ruptured condition. No good crack propagation information was obtained. (orig.)

[en] Piping System damping research at EPRI is briefly described in this paper. The focus of the research is to derive a set of technically defensible piping system damping values which can be included in Appendix N, Section III, Division I of the ASME Boiler and Pressure Vessel Code for piping dynamic analysis. Engineering judgments and regression analyses were employed to derive damping functional relationship by first establishing a uniform experimental data base, then de-aggregating the data to sort out significant contributing parameters and finally performing regression analysis. Results to date show support density, support type, insulation, frequency, attached equipment, diameter and weight are significant parameters controlling damping. A formal recommendation to ASME and guidelines of using recommended damping in various piping analysis methods remains to be accomplished

[en] The amount of data on aging is significant but scattered. Much of the data may not be relevant for the concrete structures in the nuclear industry because of differences in the materials, mix design, and curing regime. Indiscriminate regression using all the existing data may result in an unrealistic projection of the mechanical properties. For this reason, a data base specific to the types of concretes that were used in nuclear power plants in the 1960s and 1970s is being collected. New tests with concrete specimens used in five nuclear power plant projects have been performed, so that data for ages up to 19 years will be available. The new data include compressive strength, elastic modulus and Poisson's ratio evolution over time. It was observed that the Committee 209 recommendation provides a conservative measure of strength gain for this class of concrete. (author)