K. Kussmaul

New observations on the crack growth rate of low alloy nuclear grade ferritic steels under constant active load in oxygenated high-temperature water

Abstract Within the scope of reactor safety research attempts have been made over several decades to determine corrosion-assisted crack growth rates. National and international investigations have been performed on both an experimental and an analytical basis. A compilation of internationally available experimental data for ferritic steels exhibits a scatter of crack growth rates of up to 5 decades. This was one of the reasons for commencing further experimental investigations focused on the evaluation of corrosion-assisted crack growth rates. These experimental studies were performed under constant, active, external load on 2T-CT specimens of the materials 20 MnMoNi 5 5 with 0.009 and 0.020% S (similar to A508 Cl.3), 22 NiMoCr 3 7 with 0.006% S (similar to A508 Cl.2) and 17 MnMoV 6 4 with 0.017% S. The tests were carried out in deionized oxygenated high-temperature water (240°C; 0.4 and 8.0 ppm O 2 ). For K I values up to 60 MPa m 1 2 , the experimental results showed no significant dependence between corrosion-assisted crack growth rates and the stress intensity factor, the oxygen content of the medium or the sulphur content of the steel. Here it is important to note, that in this K I region the high crack growth rates after the onset of cracking due to loading are decreasing and finally come to a standstill after a short period of time as compared with operational times of plants. Consequently, the determination of crack growth velocities as corrosion-assisted crack advance divided by the test duration, so far practised worldwide, results in wrong crack growth rate values in the above-mentioned range of loading up to 60 MPa m 1 2 . Based on a test duration of 1000 h, the average crack growth rates are below 10 −8 mm s −1 for K I ≤ 60 MPa m 1 2 . When applied to a single start-up and service period of one year, this would formally lead to an average crack growth rate of 2·10 −9 mm s −1 (equivalent to 0.06 mm per year). At K I values between 60 and 75 MPa m 1 2 the average corrosion-assisted crack growth rates increase significantly. It can be observed experimentally that the crack propagates during the whole period of the test. Consequently the calculation of crack growth velocities as corrosion-assisted crack advance divided by the test duration as mentioned earlier can be applied as a first estimate. Finally, for K I values ≥ 75 MPa m 1 2 high crack growth rates up to 10 −4 mm s −1 can be observed. In this region the average crack growth rates are also in quite good agreement with a theoretically based crack growth model.

Formation and growth of cracking in feed water pipes and RPV nozzles

Abstract The observation of numerous small and large cracks in ferritic feed water pipes of boiling (BWR) and pressurized water reactors (PWR) in the last few years has led to basic research into the causes of cracking and the crack growth mechanisms. In horizontal feed water pipe sections connected to nozzles of reactor pressure vessels (RPV) of BWRs as well as of steam generators (SG) of PWRs, circumferential macro and micro cracks were detected. These cracking phenomena could be observed in base material of pipes as well as in weld seam regions. The examination of the stress state displayed that the cracked pipe regions have been exposed to a number of cyclic thermal transients (thermal shock, flow stratification) during start-up (hot stand-by) and shut-down periods of the plants. During thermal transient periods, local and global cyclic stresses in the referred pipe cross sections have been induced which in interaction with the influence from environment (in operation as well as in shut-down periods) and local geometrical imperfections led to the initiation and formation of macro and micro cracks. In the reactor water clean-up system of BWR through which reactor water is fed from the RPV to the main feed water line, two longitudinally welded elbows have been detected to be severely cracked. Both elbows have been subjected to an internal pressure corresponding to RPV and additionally to a relevant “in-plane” bending moment. These longitudinal cracks were found to be started from the inner elbow surface. In one case the longitudinal crack was situated in the base material and was enlarged to leakage. In the second elbow the longitudinal crack was located in the heat affected zone (HAZ) of a longitudinal weld. In both cases the macro cracks started either from corrosion pits located in defective areas of the magnetic protection layer or from geometrical notches (weld root). The semi-elliptic small cracks got linked to more extended shallow cracks. Formation and growth mechanism of these cracks have been studied at the MPA Stuttgart in laboratory under simulated operation conditions which were held as realistic as possible compared with those in nuclear power plants. The results of experimental studies in laboratory as well as conclusions based on the above mentioned cracking phenomena in piping have been used as basic information for a realistic design of large scale (RPV) thermal shock experiments under operation conditions. The formation and growth mechanism of these cracks and their detection by means of NDE during thermal transients at the inner surface of RPV nozzle and at the adjacent cylindrical areas of RPV shell will be described.

Some conclusions with regard to the safety assessment of cracked components drawn from the research programme “integrity of components” (FKS II) atthe present state

Abstract The continuation of the research program “Integrity of Components”, Phase II, mainly deals with further evaluation and assessment of material properties and the application of data from small standard specimens to large scale specimens and components. This includes the use of advanced numerical methods to check the transferability of fracture mechanics parameters with regard to the type of load and degree of multiaxiality on the failure behaviour of fracture mechanics specimens with component-like dimensions. Further points of interest are the relationship between upper shelf toughness and load-bearing capacity, the influence of neutron irradiation on the properties, and the effect of corrosion on cyclic crack growth.

Statistical evaluation of post-yield fracture mechanics properties on the basis of the notched bar impact test

Abstract By means of the test results it could be shown that a correlation exists between the ductile fracture mechanics parameters for crack initiation J i and J Ic and the notch impact energy, in which initially only values from the upper shelf of the notch energy were taken into consideration. Owing to a statistical evaluation and with the aid of the relationships indicated, the user can chose the probability with which the value selected is to be situated within the range in question by which the width of the scatter is presupposed. By consideration of all the distributed values, specifically the J i , J Ic and J 50 impact energy values and the material characteristic values from the tensile test, it could be demonstrated on two vessels made of modified 22 NiMoCr 3 7 (60 J on the upper shelf of the notch impact energy) and 20 MnMoNi 5 5 (200 J on the upper shelf of the notch impact energy) respectively with axial external cracks that the experimental instability load can be assigned to the upper bound of the scatter band formed from the crack initiation values J i ( J Ic ).

Determination of crack initiation and crack resistance using compact specimens, and comparison with results of numerical simulation of dynamically loaded wide plates

Abstract Static and dynamic crack resistance curves for the ferritic steel 20 MnMoNi 5 5 and the austenitic steel X6 CrNi 18 11 were determined for compact tension specimens, using a modified key curve method which contains a numerical calculation of the key curves. The method is described and verified under quasi-static loading. The J integral evaluation according to the ASTM standard (which at present is only applicable for ferritic steels under quasi-static loading) did not show in any case a considerable difference compared with the energy release rate approach. The numerical simulation of tests was carried out for wide plates made of austenitic material on the 12 MN high speed tensile testing machine at MPA Stuttgart.

Cyclic and transient thermal loading of the HDR reactor pressure vessel with respect to crack initiation and crack propagation

Abstract In a series of thermal loading tests at the HDR reactor pressure vessel – thermal stratification, cyclic thermal shock and pressurized thermal shock – the methods applied in safety analysis had to become qualified by a continuous intercomparison of calculated results and experimental data. Above all the complex boundary conditions of the HDR-tests offer a close approximation to the original components, so that they provide a real assessment of the transferability. The results of the thermal mixing tests indicated that during cold water inflow into the RPV longitudinal strains build up in the cylindrical wall which dominate over that in circumferential direction. During the cyclic thermal fatigue tests incipient crack formation in the cladding as well as the behaviour of crack propagation in the cladding and in the base material was analyzed. In the pressurized thermal shock tests, the nozzle region and the cylinder wall in the incipient crack condition were loaded by long cooling streaks. Even in the aggravated loading condition as the result of a routed cold water streak no remarkable indications of crack growth were noticed. In both cases, cyclic and pressurized thermal shock loading, the expected crack propagation was overpredicted by the fracture mechanical methods used. The non-destructive examination methods used were able to locate all of the cracks but they mostly overpredicted the actual crack depth.

Diagnostics of piping by ambient vibration analysis

Abstract The presented examples show the possibility to obtain actual information on the stiffness distribution and constraint parameters of a piping system by a non-destructive method without demounting hangers. A strategy for model updating by means of ambient vibration analysis is being developed to determine the real constraints of a piping system quantitatively. Hereby the amplitude dependency of the hanger stiffness must be investigated carefully.

A contribution to the numerical and experimental determination of residual stresses in welds

Abstract The knowledge and results obtained from the residual stress analysis of a single pass weld carried out during Phase I of the Integrity of Components Research Project (FKS) have been used as a basis for extension to a multi-pass weld within the scope of Phase II. For this purpose, the Finite Elemente Programme SMART [1] has been modified and a model weld test was conducted to verify the calculated results [2–5]. Furthermore, extensive and expensive material investigation were necessary in order to make the appropriate material properties available for the calculations.

Experience in the replacement of safety related piping in German boiling water reactors

Abstract In the German boiling water reactors (BWR) of the 69 series and their forerunner plant, high-strength low-alloy ferritic materials were used for a large number of pipings both inside and outside the pressure boundary (PB). The choice of this type of material led to comparatively thin-walled piping which, at that time, had been designed and manufactured in accordance with the codes and standards applying in the Federal Republic of Germany. Due to material properties resulting from production in a conventional manner, design features which did not sufficiently meet the requirements for nondestructive testability, and defects caused during processing, mainly in the area of circumferential welds, the ferritic pipings inside the PB were replaced in the course of a plant upgrading by new piping designed and manufactured according to the basis safety concept. The improvements and experience gained during backfitting of five German BWR plants are part of the German safety strategy and can be summarized as follows: 1. (1) Exclusion of large fractures on the basis of an optimized quality level for the piping. 2. (2) Elimination of need for the pipe whip restraints which existed in the former piping. 3. (3) Limited reduction of the former scope of inservice inspections, mainly as a consequence of improved weld quality and optimized weld performance. 4. (4) Reduction of personnel radiation exposure, e.g. by reduced number of welds and by manufacture of welds using automatic equipment, as well as by improved nondestructive testing. 5. (5) Availability values for backfitted BWR comparable to German PWR values. The pipings made of stabilized austenitic materials, which are arranged inside and outside the containment of the BWR plants, were not replaced since their quality level has been proved to be sufficient even on the basis of the present standards.

Influence of repair welding on cyclic thermal shock behaviour of a RPV nozzle corner

Abstract In piping as well as in RPV-nozzles and in surrounding parts of the cylindrical vessel shells, repeated cold water injections have been observed leading to crack initiation and subcritical crack growth. In the framework of the German Reactor Safety Research Programme, cyclic thermal shock tests have been carried out at the RPV of the decommissioned HDR (Heissdampfreaktor) near Frankfurt applying extremely conservative temperature transients. The crack propagation under access of oxygenated pressurized water was evaluated by fracture mechanics methods. The validation by non-destructive examination, as well as by destructive testing at a trepan, proved the fracture mechanics results to be conservative. The depression at the nozzle corner caused by the trepan removal has been repaired by ‘temper bead’ welding without stress relief heat treatment, similar to the ‘half bead’ technique of the ASME Boiler and Pressure Vessel Code. The subsequent thermal shock stressing (1200 cycles) under similar parameters demonstrated, with respect to the weld repair: 1. (1) no material separation extending into the ferritic RPV-wall; 2. (2) the quality of repair welding is comparable to the quality of the as-delivered condition manufactured in the 1960s in a more conventional manner. The positive experience derived from this work may be helpful in the future in dealing with any cases of cracking in thick-walled large diameter vessels which could require repair welding when a post-weld stress relief heat treatment is not possible.