[en] This paper provides an overview of the methodology and the results of a II/I seismic systems interaction study at the Comanche peak steam electric station (CPSES). This II/I study combines the positive attributes of several different resolution approaches including earthquake data, seismic analysis and dynamic impact assessments. The results of this study provides basic insights into: the number and types of potential II/I interactions which exist in a typical nuclear power plant, the appropriate methods to be used to resolve II/I seismic interaction issues, and the quantity and types of conditions within a plant which pose credible II/I interactions and could require either a relocation or a redesign

[en] Based on test experience at Comanche Peak Unit 1, an acoustic emission data evaluation matrix for piston check valves has been developed. The degradations represented in this matrix were selected based on Edwards piston check valve failure data reported in the Nuclear Plant Reliability Data System. Evidence to support this matrix was collected from site test data on a variety of valve types. Although still under refinement, the matrix provides three major attributes for closure verification, which have proven useful in developing test procedures for inservice testing and preventing unnecessary disassembly

[en] A test program has been undertaken to identify trendable acoustic parameters and spectra characteristics, obtained during operating flow conditions, which are capable of indicating check valve degradation. Efforts were made to quantify changes in degradation sensitive acoustic parameters and spectral characteristics for various common types of check valves with artificially induced degradations. Acoustic data were acquired during steady-state operating flow with a single accelerometer and a portable, microprocessor-based, vibration data collector. Acoustic parameters and frequency spectra were calculated by the data collector and transferred to a computer database for analysis and trending. This methodology can be implemented in order to augment conventional, commercially available check valve diagnostic systems that can be difficult to apply on a regular basis, and typically require cycling a valve's state in order to analyze degradation

[en] This Supplement provides the results of the staff's evaluation and resolution of approximately 400 technical concerns and allegations in the mechanical and piping area regarding construction practices at the Comanche Peak facility. This report does not address the Walsh/Doyle allegations regarding deficiencies in the pipe support design process and the new allegations recently received by the staff

[en] This is a companion paper to the Hughes/Edwards paper presented at EPRI '96, entitled open-quotes Don't Put Your Troubles Down the Drain.close quotes That paper presents a detailed account of liquid radwaste processing performance at Texas Utilities' Comanche Peak Steam Electric Station from 1993 to 1995. This paper looks at the same period from a historical and philosophical (and subjective) perspective

[en] Generic Letter 89-10 issued by the United States Nuclear Regulatory Commission recommends periodic verifications of the capabilities of motor operated valves (MOVs) to perform their design basis functions. At TU Electric's Comanche Peak nuclear power plant, the present MOV verification methodology requires diagnostic testing wherein operational readiness is demonstrated by satisfying test procedure acceptance criteria. The test procedure acceptance criteria account for differences in MOV performance between ''static'' and ''design basis differential pressure'' conditions, and provide margin for reasonably anticipated increases in ''valve factors'' over the valve's service life. Consequently, it is sufficient to perform testing under static conditions to periodically verify continued MOV operational readiness under design basis differential pressure (DP) conditions. TU Electric has determined that intervals varying from two to six refueling cycles (three to nine years) are appropriate between verification tests. The maximum duration between verification tests is a function of: (1) the relative importance of each MOV to plant safety, (2) each MOV's design margin, and (3) the maximum duration between maintenance activities which require subsequent static verification testing. In some cases, the implemented switch settings may necessitate a duration between static verification tests less than the maximum duration justified by engineering. This paper describes how the TU Electric program addresses this long term aspect of GL 89-10. Present and future plans are presented and discussed

No abstract available

[en] This article discusses the control of unacceptable vibration levels that persisted for about two years in a 400 F heater drain piping system at Comanche Peak's 1100-MW Unit 1. The system in this PWR nuclear power plant is made up of approximately 300 ft of 8-in., Schedule 40 piping that runs from the steam generator heater to two 10-ft-diameter horizontal drain tanks. During start-up and shut-down operations, flashing of the 400 F water to steam occurs at the intakes of the tanks. This flashing caused severe vibration in the 8-in. piping from the tanks to about 80 ft upstream. Peak displacements of approximately three inches were measured. Viscous dampers were installed to solve the problem

No abstract available

No abstract available