Fatigue Analysis For A Reactor Feed Pressurization Drum

A fatigue analysis has been done for a reactor feed pressurization drum. The vessel operates under cyclic pressure ranging from 25 psig to 200 psig (1.7 – 13.8 bar), with a design life of more than 5 million cycles. Additionally, it is subjected to 200 full-range cycles from atmospheric pressure up to a design pressure of 310 psig (21.4 bar). All operations are conducted at a constant temperature of 130 ⁰F (54.4 ⁰C). The fatigue analysis is performed in accordance with the ASME VIII division 2 part 5.5.3 design code.

Based on the drawings provided by the client, a CAD model was created in Abaqus, which was used to run the finite element analysis (FEA). The vessel was modeled with sufficient detail to capture the complete geometry, including nozzles and non-structural attachments.

The next phase involved the discretization of the model and the specification of the different materials, boundary conditions, and cyclic loads. A proper mesh size is important, especially at critical areas, however, computational efficiency should also be taken into account. For that reason, two approaches from the ASME code were considered:

  • Peak stress evaluation: For regions with abrupt geometry changes, detailed local refinement is required, along with a convergence study to ensure accurate peak stress capture.
  • Stress linearization: If local details are not modeled, stress linearization is needed to exclude peak components, allowing for a less refined mesh. The linearized stresses must then be adjusted using a fatigue stress reduction factor (FSRF) as specified by the ASME code.
nozzle, local mesh refinement, FEA

Nozzle 1 local mesh refinement around the external notch and internal rounding areas. Structured mesh with a maximum mesh size of 4mm (0.157 in) around the nozzle and 2mm (0.079 in) along the notch and rounding

After that the model was verified by means of a mesh convergence check and the computed stresses were validated against simple hand calculations involving the axial and hoop stress in a cylinder.

σ a = P D 4 t
σ h = P D 2 t
Equivalent stress range values along the cylindrical section of the pressure vessel. The scale color shows how most of the stresses at the vessel far from the support lugs are in between 50-57MPa.

Equivalent stress range values along the cylindrical section of the pressure vessel. The scale color shows how most of the stresses at the vessel far from the support lugs are in between 50-57MPa.

Stress range values at nuclear source support, measured in Pa. Even without linearizing, with a max. stress range of 47 MPa (6.82 KSI) stress amplitudes would pass for a FSRF of 4.

Stress range values at nuclear source support, measured in Pa. Even without linearizing, with a max. stress range of 47 MPa (6.82 KSI) stress amplitudes would pass for a FSRF of 4.

The analysis showed that all pressure-bearing elements exhibit stress amplitude levels well below the 100 MPa (14.5 ksi) limit for the designed number of cycles, as specified in the S-N curve, thereby indicating no risk of fatigue. Moreover, the non-structural components that are welded to the outer face of the shell exhibit a relatively low-stress range. This means that the vessel is approved for fatigue according to the ASME VIII division 2 part 5.5.3 design code.

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