The main purpose of this research is to investigate the effect of friction in the thermal stress of Reactor Coolant System (RCS) of VVER-1000. RCS is a large system connecting reactor vessel, steam generators and RC Pumps. During the heat-up of reactor, the RCS expand and during cool-down of reactor, it contracts. Because of the heavy weight of reactor and steam generator, the friction at the support of RCS affects the thermal stress of RCS. In this paper how much support friction contributes to the development of thermal stress is assessed in order to investigate the thermal stress and effect of support friction. A quarter-symmetry model of VVER-1000 RCS is developed in ANSYS and meshed with hexahedral elements to ensure better solution accuracies. The model includes reactor vessel, steam generator and reactor coolant pump. Internals of reactor vessel, steam generators and RCPs are represented by point mass to simplify the model. Temperature of inside surface of hot-leg side of reactor vessel to inlet side of steam generator is assumed same uniform hot-leg temperature, and the temperature of inside surface of outlet side of steam generator to reactor vessel is uniform cold-leg temperature. All outside surface are assumed insulated. The analysis includes neither transient thermal loading nor dynamic loadings. The analysis results show that friction at support brings little effect on the peak thermal stress. The peak thermal stress occurs at hot-leg nozzle of reactor pressure vessel and it approached near yield stress. If load combination is included the localized total stress at hot-leg nozzle could go over the yield stress. This peak stress could affect fatigue life in a long run. A recommendation is made that a detailed fatigue analysis of VVER-1000 RCS is necessary.
- Nuclear Engineering Division
Effect of Support Friction in Thermal Stress of VVER-1000 RCS for NOP Condition
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Nguyen, TQ, & Namgung, I. "Effect of Support Friction in Thermal Stress of VVER-1000 RCS for NOP Condition." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Plant Systems, Structures and Components; Codes, Standards, Licensing and Regulatory Issues. Prague, Czech Republic. July 7–11, 2014. V001T03A026. ASME. https://doi.org/10.1115/ICONE22-30904
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