Search

Flow-induced vibrations of reactor core components have been a major cause of failure of reactor internals in many nuclear power plants. To address this issue international standards (viz., ASME-OM-05) require monitoring of structural integrity of reactor core in nuclear power plants. A reactor internals vibration monitoring system (IVMS) has been developed for nuclear power plant surveillance. The system detects the core barrel motion and flow-induced vibrations of reactor internals by analyzing the inherent fluctuations (reactor noise) present in the neutron flux signals from ex-core neutron detectors. Before application of the IVMS in nuclear power plants the system hardware and methodology/software has been extensively tested on the research reactor through a series of reactor noise measurements. In the noise measurements the neutron flux signals were correlated with the signals from vibration sensors mounted on reactor structure and control rods drive mechanisms. The frequency spectra of reactor power signals obtained with IVMS had shown small oscillations in the neutron flux signals at well-defined frequencies. There was a strong correlation of these fluctuations with the establishment of coolant flow through the core. With the help of elaborate experiments using the IVMS the cause of oscillations in neutron flux signals was attributed to the flow-induced vibrations of control rods, and the particular vibrating control rod was identified. The magnitude of the displacement of vibrating control rod was calculated from the measured power spectral density of neutron power signals. It has been shown that ex-core neutron noise measurements are more sensitive in determining the dynamic behavior of reactor internals than the vibration sensors mounted at remote, out-of-the core locations. Index Terms--Control rod, data acquisition, flow induced vibration, neutron noise, temperature fluctuations.

Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS. [ABSTRACT FROM PUBLISHER]