Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian Pressurised Heavy Water Reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results. The present paper highlights the analysis results for Prestressed Concrete Containment Vessel (PCCV) tested at Sandia National Labs, USA in a Round Robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd = design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete-tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd.
- Nuclear Engineering Division
Assessment of Ultimate Load Capacity for Pre-Stressed Concrete Containment Vessel Model of PWR Design With BARC Code ULCA
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Basha, SM, Singh, RK, Patnaik, R, Ramanujam, S, Kushwaha, HS, & Raj, VV. "Assessment of Ultimate Load Capacity for Pre-Stressed Concrete Containment Vessel Model of PWR Design With BARC Code ULCA." Proceedings of the 10th International Conference on Nuclear Engineering. 10th International Conference on Nuclear Engineering, Volume 1. Arlington, Virginia, USA. April 14–18, 2002. pp. 551-561. ASME. https://doi.org/10.1115/ICONE10-22187
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