Fracture assessment for reactor circuit (FRAS)

the pipe model. Some dimensions were altered and the new version of Abaqus finite element code was taken in use. The new helpful features of latest Abaqus versions concerning this field of study were reviewed. More non-linearities were included in the models to make them more realistic. Earthquake was chosen as a new load type and it was preliminarily studied with the existing pipe model. Two international conference papers were written and presented.

Figure 2.6.5.1 Vertical forces due to pipe break as functions of time in the nonlinear spring element (cases N1 and N3) or in the restraint modelled with solid elements (case N2)

Computational Fluid Dynamics (CFD) calculations of the Vattenfall thermal mixing experiment were performed. Various linear and non-linear Reynolds-Averaged Navier-Stokes (RANS) turbulence models were tested. The profiles of temperature, velocity and turbulence quantities were qualitatively quite correct but the thermal mixing was clearly too low with all turbulence models. CFD calculations of thermal stratification in a horizontal feedwater pipe and nozzle of a boiling water reactor were also performed. Three different transients, where the feedwater has relatively low flow rate and temperature, were calculated. For the transient with lowest flow rate and temperature, steady state stratification was found only in the nozzle area. For the other transients, practically no stratification was obtained.

Figure 2.6.5.2 Instantaneous temperature field [K] in the Vattenfall experiment obtained with Reynolds stress turbulence model.

Weld residual stress (WRS) assumptions from relevant fitness-for-service procedures were implemented in numerical simulations to reactor circuit components: RPV nozzles, safe-ends and connecting pipes. The models were created and elastic-plastic analyses performed with suitable FEA code. Here as-welded state WRS assumptions were considered. With the FEAs it was clarified how WRS distributions alter/decrease over the years during plant operation due to various yearly transient load cases. Crack growth in welds including the WRS distributions have been examined with fracture mechanics based analysis tool VTTBESIT, with stress distributions taken from results of the FEM analyses.

The development work on advanced fracture mechanical assessment methods consisted of finite element assessment tools and fracture mechanics analysis methods based on material characterisation, damage mechanisms models and structural performance. The aim was to increase understanding on the behaviour of postulated initial flaws (surface-flaws) and, on the other hand, on environment assisted material damage like irradiation embrittlement and stress- corrosion cracking. Engineering tools for fracture analyses were studied and utilized together with submodelling technique that utilizes detailed models around the crack region together with the global model of the structural component.The reactor circuit integrity was further studied by analysing cracked T-junctions with 3D FEM model. The solution for integrity assessment of T- junction was sought by using finite element software (ABAQUS, ANSYS codes) and integrating its application to existing geometric and loading data, e.g., plant database. A case computation for a T- joint was performed.

Transferability of fracture mechanics test data associated with different levels of specimen’s constraint was to be investigated by performing fracture mechanics tests using specimens with both deep and shallow surface notches. The task includes fracture mechanics tests on selected materials using surface cracked specimens at various degrees of tension and bending. This should provide input for future development of FEM analysis methods taking into account constraint effects in structural analysis. Numerical work was performed to assess the fracture toughness transition between standard CT and SENB -type specimens and 3D surface cracks for cleavage initiation and propagation using the WST cleavage fracture model. The assessment clarifies how fracture toughness is affected by different crack types as well as 'realistic' crack features (such as asymmetric crack front). Micromechanical modelling of cleavage fracture was to be performed using multi-scale models and the Master Curve method. The WST model implementation was finalized. Development of models for predicting cleavage fracture toughness will be accomplished using micromechanical modelling methods. The development of sub-modelling techniques includes a comparison of crack growth results obtained with and

without sub-modelling in some typical NPP applications, such as RPV nozzle and safe end. The limits of sub-modelling techniques were defined in the first part of the study.

Results of fracture resistance measurements (JR curves) on different materials using various specimen types (e.g., CT, SENB) were thoroughly analysed in order to define the measuring capacity of specimen. Results show that although measuring capacity according to ASTM E1820 – 08 standard is double compared to the previous -05 version it is still very conservative.

Crack growth seems not to be limiting factor for using small specimens but J-integral, see Figure 2.6.5.3 below.

Figure 2.6.5.3. Fracture resistance curves and standard specimen size limitations showing the conservatism of standard requirements.

The final results from the IAEA coordinated and sponsored research project Phase 8 "Master Curve Approach to Monitor Fracture Toughness of Reactor Pressure Vessels in Nuclear Power Plants" were collected and issued in the IAEA report series in 2010 (IAEA-TECDOC-1631).

The project was a continuation to the CRP Phase 5 and aimed at clarifying the open issues addressed in this and the previous CRPs on applying the Master Curve methodology to irradiated reactor pressure vessels. The project Phase 8 comprised three topic areas: 1) Effects of test specime size, geometry and constraint, 2) loading rate effects and qualification of impact fracture toughness testing and 3) Master Curve shape for highly embrittled RPV materials.

Deliverables in 2009

Modelling methods (ABAQUS) for different types of supports and restraints were delineated and critical accidents cases reviewed. Stiffness of pipes and its supports were calculated with FEM and substituted with simpler special purpose elements. A pipe break was further simulated with different kinds of models which included more non-linearities. The results were compared with each other and their reliabilities evaluated. Two international conference papers were written and presented. Research report has been prepared.

Fracture resistance curves - SE500HR

0 200 400 600 800 1000 1200 1400 1600 1800 2000

0 0,5 1 1,5 2 2,5 3 3,5 4 4,5 5

Crack extension (mm) J (kJ/m²)

CT_A1 CT_A2 CT_A3 CT_A4 CT_A5 CT_B1 CT_B3 CT_C2 CT_C5 SENB_B1 SENB_B2 SENB_B3 SENB_C1 SENB_C2 Series15

deltaJmax ASTM E1820-01

Computational Fluid Dynamics (CFD) calculations have been carried out. Experimental data for thermal mixing in a T-joint has been acquired from Vattenfall. Calculations of the test case have been performed with different meshes and turbulence models. Comparison with the experimental data has been done and a report is complete.

A conference paper and three research reports have been prepared on weld residual stresses.

Weld residual stress (WRS) assumptions from relevant fitness-for-service procedures were implemented in numerical simulations to reactor circuit components: RPV nozzles, safe-ends and connecting pipes. The models were created and elastic-plastic analyses performed with suitable FEA code. Here as-welded state WRS assumptions were considered. Crack growth in welds including the WRS distributions have been examined with fracture mechanics based analysis tool VTTBESIT, with stress distributions taken from results of the FEM analyses.

The conference paper on limitations and application of the Master Curve method for RPVs (CRP-8 topic area 3) was presented in ASME PVP 2009. The article on Master Curve methodology was prepared and submitted.

A detailed test matrix on ATOM Probe and PA characterisation of irradiated materials was agreed with Tohoky University. Samples representing weld 501 material in eight different IAIA-conditions and ten different model alloys in irradiated conditions were prepared by VTT with EDM. The samples were transported to Japan in February 2009. First data is expected to be available in 2010.

2.6.6 Influence of material, environment and strain rate on environmentally assisted