Study of Pipe Break in a Nuclear Reactor

Sweden has 10 nuclear power reactors providing nearly half of its electricity. Due to the current concern for optimal energy sources, many Swedes consider nuclear power a good option when competitiveness and environmental impact are taken into account. As nuclear power carries the risk of toxic pollution, the safe, efficient operation and maintenance of the nuclear reactors is very important.

Because of its reliability and solution effectiveness, ADINA is used in many reactor studies in Sweden. In this News we feature one such study of the effect of a pipe break in a pressurized water reactor. The study was conducted by Onsala Ingenjörsbyrå (Onsala Engineering AB, Sweden) which performed the work under a contract from Ringhals AB. Figure 1 below shows the nuclear reactor considered.

Fig. 1. Pressurized water reactor studied

The geometry for the entire model was created in Pro/Engineer, while the mesh and the boundary condition sets were created in ANSA. The complete model consists of more than 800,000 solid, structural and fluid elements with 55,000 nodes in contact, a total of 1.6 million degrees of freedom. The model is depicted in Figure 2 below, and part of the mesh used to model the reactor vessel is shown in Figure 3.

Fig. 2. Finite element model of reactor

Fig. 3. Part of the mesh consisting of solid, structural and fluid elements
used to model the reactor vessel

The model was originally intended to calculate the loads, due to pipe breaks in the system, on the bolts holding in place the baffle plates that are used to direct the flow of coolant within the reactor vessel. The model was later extended to include the entire primary system in order to calculate the response of the whole system to pipe breaks.

The finite element model was imported to the AUI via the Nastran format, then completed in the AUI for subsequent solving by ADINA. Finally, the AUI was used to post-process the results.

The above movie shows the pressure distribution in the reactor vessel due to a pipe break in the cold leg of the reactor coolant loop (depicted by the breaking gray line at the left side of the movie). Figure 4 below shows the pressure distribution at a cross section of the reactor vessel just after the break.

Fig. 4. Pressure plot in section of reactor vessel

With the results of the analysis, Onsala engineers were able to predict the structural response of the reactor vessel due to the pipe break (Figure 5 below) as well as the maximum bolt forces.

Fig. 5. Predicted plastic regions in reactor vessel

While already yielding many useful results, of course, the model can be further developed for additional analyses.

This study illustrates, only to some degree, the powerful capabilities available in ADINA for dynamic linear and nonlinear analyses, including fluid-structure interactions. Many different types of analyses can be performed in a very effective and reliable manner, a requirement particularly important in studies of reactors.

Courtesy of Onsala Ingenjörsbyrå AB