Elisabeth Keim

Numerical weld modeling — a method for calculating weld-induced residual stresses☆

Abstract In the past, weld-induced residual stresses caused damage to numerous (power) plant parts, components and systems (Erve, M., Wesseling, U., Kilian, R., Hardt, R., Brummer, G., Maier, V., Ilg, U., 1994. Cracking in Stabilized Austenitic Stainless Steel Piping of German Boiling Water Reactors — Characteristic Features and Root Causes. 20. MPA-Seminar 1994, vol. 2, paper 29, pp.29.1–29.21). In the case of BWR nuclear power plants, this damage can be caused by the mechanism of intergranular stress corrosion cracking in austenitic piping or the core shroud in the reactor pressure vessel and is triggered chiefly by weld-induced residual stresses. One solution of this problem that has been used in the past involves experimental measurements of residual stresses in conjunction with weld optimization testing. However, the experimental analysis of all relevant parameters is an extremely tedious process. Numerical simulation using the finite element method (FEM) not only supplements this method but, in view of modern computer capacities, is also an equally valid alternative in its own right. This paper will demonstrate that the technique developed for numerical simulation of the welding process has not only been properly verified and validated on austenitic pipe welds, but that it also permits making selective statements on improvements to the welding process. For instance, numerical simulation can provide information on the starting point of welding for every weld bead, the effect of interpass cooling as far as a possible sensitization of the heat affected zone (HAZ) is concerned, the effect of gap width on the resultant weld residual stresses, or the effect of the ‘last pass heat sink welding’ (welding of the final passes while simultaneously cooling the inner surface with water) producing compressive stresses in the root area of a circumferential weld in an austenitic pipe. The computer program feresa (finite element residual stress analysis) was based on a commercially available abaqus code (Hibbitt, Karlsson, Sorensen, Inc, 1997. abaqus users manual, version 5.6), and can be used as a 2-D or 3-D FEM analysis; depending on task definition it can provide a starting point for a fracture mechanics safety analysis with acceptable computing times.

Life management of reactor pressure vessels under pressurized thermal shock loading: deterministic procedure and application to Western and Eastern type of reactors

Abstract The integrity of the reactor pressure vessel (RPV) has to be maintained throughout the plant life and it is also one of the main considerations regarding the plant life extension. Adequate approaches to the RPV integrity assessment provides a basis for plant-safe operation and for timely implementation of preventive and corrective measures if necessary. A substantial part of the RPV integrity assessment is related to the pressurized thermal shock (PTS) analysis. The safety of the RPV during such a severe loading condition has to be proven in the design stage and during life of the component. Irradiation programs allow for an adequate description of the material response and by means of analytical and numerical methods the structural analysis can be performed. The PTS-analysis of the RPV is a multidisciplinary effort and involves among others detailed thermal hydraulic analysis and structural analysis including fracture mechanic assessment. Operational load situations during accident conditions are dominated by pressure and temperature transients during an emergency core cooling event following a loss of coolant accident relevant for PTS. Cold water is injected into the cold legs and falling into the downcomer of the RPV. Complex thermal hydraulic situations occur and the fluid temperatures in the downcomer as well as the RPV wall-to-fluid heat transfer coefficient are highly non-linear. To determine the spatial fluid temperature and wall-to-fluid heat transfer coefficient distribution in the downcomer and the respective thermal stresses with adequate accuracy, fluid–fluid mixing scenarios are taken into account. Plume cooling during a loss of coolant accident of the RPV leads to a locally increased axial stress distribution in the plume region compared to the axial symmetric cooling. The higher axial stresses make the circumferential crack more relevant for the safety assessment. The transient temperatures and stresses in the RPV-wall are calculated by means of 3D finite element (FE) calculations and are performed for given transients. The safety assessment deals with a safety margin against the material toughness curve or the determination of the allowable brittle to ductile transition temperature. Using the global thermal hydraulic parameters calculated with adequate system codes, the fluid temperature and RPV-wall-to-fluid heat transfer coefficient distribution in the downcomer has been calculated with the Siemens code KWU-MIX. In the following fracture mechanics calculation, two flaw types have been considered, axial and circumferential cracks. By means of 3D FE calculations the sub-surface flaws were investigated and from the results of the fracture parameters safety margins could be derived. In this report the Siemens procedure is applied as an example to an Eastern plant WWER-1000 type, but it has been successfully carried out for different types of RPV also: Western 3 and 4-loop plants (German and French type) where the results are reported in [Proc. ASME PVP Conf., (1999), Boston] [7] and [NUREG/CR-6651 (ORNL/TM-1999/231), Oak Ridge National Laboratory, (1999)] [5] . Although all considered pressure vessels are different in their construction: the French type is constructed with a thermal shield between RPV wall and internals, the Eastern type has two rings of injection nozzles, which leads to a complex thermal hydraulic situation the general procedure could be applied. Following results have been obtained: –plume cooling leads to increased axial stresses in the plume region –therefore circumferential flaws are more severe than axial flaws –safety margins for axial flaws are higher than for circumferential flaws.

Reconstitution techniques qualification and evaluation to study ageing phenomena of nuclear pressure vessel materials (RESQUE)

The RESQUE project aims at optimising and normalising reconstitution techniques and is now in its final phase. The project belongs to the AGE-cluster, that also involves the REFEREE project being used as an input to RESQUE. At FISA 97 the reference data on non-reconstituted specimens were presented together with a set of recommendations on temperature measurements (WP1, WP2). Now, the results on the quality and limiting conditions of the reconstitution weld seam are discussed. The combination of this information leads to a set of recommendations for optimised reconstitution parameters that allow to qualify reconstitution equipment and methodology (WP3). The recommendations on the minimum insert length for impact and three-point bend fracture toughness testing have been established (WP4). Recommendations on dimensional tolerance deviations were put forward (WP5) and series of tests have been performed on selected reconstituted irradiated specimens (WP6). All work packages have been summarised. The overall information is being recapitulated in a Proposal for Code of Practice for Reconstitution of Irradiated C v -type Specimens (WP7).

Load-carrying capacity and crack resistance of a cladding by the Sigma-oscillating wire technique

If cracks are postulated in the ferritic base material beneath the austenitic cladding, their initiation and propagation under hypothetical loading cases is influenced by the load carrying capacity of the cladding. The toughness of the KKS-RPV cladding was assessed by means of elastic-plastic fracture mechanics methods. Sub-sized tensile and bend specimens were fabricated by reconstitution technique from broken halves of standard ISO-V Charpy specimens, representing crack extension in radial and circumferential direction. They were tested and evaluated and further analyzed with the Gurson model. For temperatures relevant to the loss of coolant and upset conditions analyzed, a sufficient toughness of the cladding in terms of J and CTOD resistance curves could be shown.

Linear elastic stress intensity factors for cracks in nuclear pressure vessel nozzles under pressure and temperature loading

Abstract Stress intensity factors for two different nozzle geometries and different crack sizes are evaluated for pressure and thermal loading utilising three-dimensional elastic finite element models. The results are compared to available experimental data and a procedure is proposed to estimate the maximum of the stress intensity factor for arbitrary crack size and loading conditions.

Determination of critical circumferential through-wall crack sizes regarding load reduction by increased flexibility in piping systems

Piping systems in power stations are not only loaded by the system pressure and force controlled bending moments e.g. due to the weight of the system, but also by additional bending moments caused by the thermal expansion of the hot system and prescribed motions, as in the case of an earthquake, resulting in displacement controlled moments. To determine critical circumferential through-wall crack sizes in pipes, a method based on the J-integral has been developed, which takes into account the load reduction due to the increased flexibility of a flawed section. The method is presented and compared with results by the finite element method and experimental data.

Comparison of Compact, Reconstituted and Subsize Charpy Specimens

During the operation time of nuclear power plants change of material properties of the reactor pressure vessel may occur due to ageing or irradiation effects. A common tool to investigate this change of material properties is the determination of the shift in transition temperature by means of the Charpy impact test. Suitable index temperatures are derived from the test data of tested specimens of surveillance programs or from test data of specimens cut out of templets of RPVs. Due to lack of material, subsize and/or reconstituted specimens will be often used. In this paper investigations on reconstituted specimens are discussed and compared to results of compact (original) and subsize impact specimens. From the experimental data base the conclusion can be drawn that valid index temperatures for safety analysis can be derived from the test data of reconstituted specimens with minimum insert lengths of 10mm. Also from the test data of subsize specimens valid index temperatures can be obtained, if corresponding correlation functions to the original conventional ISO Charpy test data are applied.

Brittle fracture analysis of the Kozloduy Unit 1 RPV

The safety of the RPV of the Bulgarian NPP Kozloduy Unit 1 was analysed within EC-financed contracts according to a pressurized-thermal-shock- (PTS-) procedure applied in Germany (Erve, M., Hertlein, R., 1991. Post SMiRT Seminar No 11, August 1991), considering the most relevant transients and taking into account the actual embrittlement in the core weldment. The paper reports on the main aspects of the PTS-procedure, determining the acceptable transition temperature (T a K -evaluation) to exclude brittle fracture, and compares the main results with the fluence related transition temperature (T F K ) of the material got from sampling from the weldment concerned. Testing of the toughness properties by small size Charpy-V-notch specimens revealed only a small irradiation effect in comparison to the properties after the recovery annealing performed in 1989. This could be explained by the fact that only small values of Cu-content in the weld metal were confirmed, thus balancing the expected influence of the relatively high P-content. The main conclusion is: assuming a defect size of 10 x 60 mm, the evaluation shows, for KNPP 1 after the 18th cycle for the screening transient, a sufficient margin in the T a K -value to the actual material properties and-from the technical point of view-thus, recovery annealing is not necessary for the time being. Further embrittlement of the RPV will be covered by an additional surveillance program with samples accelerated re-irradiated in a Russian NPP. Proper operator actions during PTS events can further improve the situation with respect to loading of the RPV during transients, thus increasing the safety margins.

Critical circumferential through-wall cracks according to the unloading of the cracked section under displacement-controlled bending load

Abstract The evaluation of critical circumferential through-wall crack lengths in piping is usually performed by the flow stress concept, plastic limit load method or GE-EPRI procedures. Most of these methods treat the secondary stresses, especially those caused by bending moments resulting from restrained thermal expansion, as force-controlled loads. In reality, there is a movement of the piping into the direction of the prescribed displacement and, therefore, a relaxation of the cracked section, which is due to the local rotation of the cracked section. Instead of the bending moment originating from the elastic analysis of the piping system there will be a reduced bending moment, the load decreases, the real critical through-wall crack lengths due to this displacement-controlled loading are larger than those predicted by the load controlled methods. A corresponding analytical procedure taking into account this relaxation was developed and validated by a comparison with experiments as well as finite element calculations. The procedure can be used for the evaluation of the safety of piping systems (e.g. leak-before-break analyses), if the usual methods based on force-controlled loads give unrealistic, conservative results.

Safety analysis of reactor pressure vessels considering irradiation effects and improved temperature calculations for PTS

Abstract The safety analysis of reactor pressure vessels has to take into account all parameters, including design, material, fabrication, inspection, loading and service conditions. This safety analysis for pressurized thermal shock consists of the following: materials analysis (toughness, irradiation); defect analysis (inspection during fabrication and while in service); thermal hydraulic analysis (fluid-fluid mixing scenario, fluid-fluid mixing experiments); fracture mechanics analysis (evaluation of material, defects and loads). The practice with respect to this procedure is described in this paper.