Vibration Analysis of Primary Inlet Pipeline Of Pakistan Research Reactor-1 during Steady State and Transient Conditions

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    • Hayat T. HAYAT T.
    • Reactor Experiment Group, Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology


Pakistan Research Reactor-1 (PARR-1) has been converted to low enriched uranium (LEU) fuel and upgraded to a maximum power level of 9MW. During conversion and upgradation most of the reactor systems were modified and several additional facilities were provided. Major changes were carried out in the cooling system. These include installation of new primary pumps, set of heat exchanger assemblies, a cooling tower and piping system. Vibration tests on these systems were performed to check the integrity and detect any abnormality in normal operation. Results of the tests performed on the primary pumps and core support structure comply with the requirements for pre-operational and initial startup vibration testing of nuclear power plant (NPP) system. Attempt is now being made to analyze vibrations of primary piping.<BR>The Primary Water Inlet Pipeline (PW-IPL) is of stainless steel conveying demineralized water from holdup tank to the reactor pool. The section of the pipeline from heat exchangers to the valve pit is hanger supported, and the rest from valve pit to the reactor pool is embedded. Vibration of the PW-IPL may be categorized into steady state and transient. The reactor pumps mainly contribute the steady state vibrations, while transient vibrations arise due to instant closure of check valves (water hammer). The ASME Boiler and Pressure Vessel code provides data only about the limits of acceptable vibrations and stresses related to the primary static stress due to steady state vibrations. However, due to complexity and diversity in the pipe structure, stresses related to the transient vibrations are neglected in the code.In this paper steady state and transient vibration behavior of PW-IPL of PARR-1 has been analyzed. In the analysis vibration data was used for comparison with the allowable limits, estimations of pressure wave velocity, deflection, natural frequency, tensile and shear load on hanger support, and to obtain the ratio of maximum combine stress to the allowable.


  • Journal of Nuclear Science and Technology  

    Journal of Nuclear Science and Technology 35(2), 148-157, 1998-02-25 

    Atomic Energy Society of Japan

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