Reactor Pressure Vessel Modification Approaches

Stress corrosion cracking of Alloy 600 has lead to the modification and replacement of many nuclear power plant components. Among these components are the Bottom Mounted Nozzles (BMN) of the Reactor Pressure Vessel (RPV). Modifications of these components have been performed on an emergent basis. Since that time, Framatome ANP has developed state-of-the-art modification methods for the repair of BMNs using the Electrical Power Research Institute (EPRI) managed Materials Reliability Program (MRP) attributes for an ideal repair as a basis for evaluation of modification concepts. These attributes were used to evaluate the optimal modification concepts and develop processes and tooling to support future modification activity. This paper details the BMN configurations, modification evaluation criteria, several modification concepts, and the development of the tooling to support the optimal modification scenarios.

Predicting Speed of Crack Running in Functionally Graded Material

A theoretical and computational methodology for the analysis of the functionally graded material (FGM) is introduced, and its application is made to the problem of a dynamically propagating crack running transversely in the FGM, where the intensity of the estimated crack-tip severity is managed to keep in valance with the graded material toughness in the FGM during the propagation. To detect the crack-tip severity, an integral fracture parameter, T*, is used. The crack is propagated so that the value of T* is equated to the prescribed varying critical values of T* for the graded material. Emphasis is placed on the use of a fuzzy inference technique in order to control the crack speed, which is deduced from a few T* values immediately preceding the current crack position. As to describing the constitutive law for the FGM, micro-spherical particles of arbitrary size in mesoscale are considered to be randomly dispersed in the matrix medium. By assuming that the volume fraction of the inclusion is continuously varied from 0 to 100 percent in the material, the grading is modeled. For modeling the constitutive law for the FGM composite media of thermo-elastoplasticity, a closed form SCC-LRM constitutive model describing the nonlinear material mechanics of the particle-dispersed medium is used. The model is based on the self-consistent scheme and uses Eshelby’s equivalent inclusion method. Unprecedented analytical results of predicting the crack speed of a crack running transversely in the FGM plate are obtained. In some cases of material grading, apparent crack arresting is observed as the crack runs into the metal rich area of the FGM.

Experimental Investigation of Ductile Tearing Properties for API X70 and X80 Pipeline Steels

This work presents an investigation of the ductile tearing properties for API 5L X70 and X80 pipeline steels using experimentally measured crack growth resistance curves (J-R curves). Testing of the pipeline steels employed compact tension (C(T)) fracture specimens to determine the J-R curves based upon the unloading compliance method using a single specimen technique in accordance with the ASTM E1820 standard procedure. Conventional tensile tests and Charpy V-Notch tests were also performed to determine the mechanical and impact properties for the tested materials. Severe splitting running parallel with the crack propagation path with varied lengths was observed in all tested fracture specimens, particularly for the API X80 material. The occurrence of splits makes the determination of JIc and resistance curves more difficult, as delamination of interfaces positioned normal to the crack front decreases the effective thickness of the test piece, inducing plane stress conditions deep inside the specimen. This experimental characterization provides additional toughness and mechanical data against which the general behavior of X70 and X80 class pipeline steel can be compared.

Impact of PWSCC and Current Leak Detection on Leak-Before-Break Acceptance

This paper assesses the effect of using primary water stress corrosion cracking (PWSCC) crack morphology parameters (roughness, number of turns, and actual flow path/pipe thickness) in determining the difference in the leakage crack length, and how the difference in the leaking crack lengths changes typical margins from past LBB submittals and published reports. Several past LBB submittal cases were selected; in addition, cases from generic LBB reports published by EPRI were also selected. The results of the analyses showed that the past submittals by nuclear steam system supply (NSSS) companies frequently used the surface roughness comparable to an air-fatigue crack with no turns and the actual flow path equal to the thickness of the pipe. This condition would give the shortest possible leakage flaw length. The roughness, number of turns, and actual flow path to thickness ratio for PWSCC cracks were determined from photomicrographs of service-removed cracks. When using the PWSCC crack morphology parameters that corresponded to the crack growing parallel to the long direction of the dendritic grains (V.C. Summer and Ringhals cases), then the leakage flaw length increased 69 percent over the air-fatigue crack length at the same leak rate. Using the same critical crack length as was used in the initial LBB submittals and the published documents, the margins on the crack length changed from 1.77 to 6.0 for the initial submittals (which we also reproduced) to 0.88 to 2.74 from our calculations for a PWSCC crack. If the crack grew in the buttered region of the bimetallic weld, then based on metallographic sections from service-removed flaws, there would be a more tortuous flow path. For this crack condition, in all but one case, the margins on the normal operating versus N+SSE crack lengths were below the safety factor of two required for LBB approval. The average margin decreased from 3.39 for the air-fatigue crack to 1.55 for the PWSCC crack growing transverse to the long direction of the dendritic grains. This was about an additional 20 percent decrease in the margin from the case of having the PWSCC grow parallel to the long direction of the dendritic grains. These results show that LBB is difficult to satisfy for PWSCC susceptible pipe using the current SRP 3.6.3 LBB approach. This LBB assessment did not consider the possible development of a long circumferential surface crack, which would be more detrimental to LBB behavior. Such cracking behavior would violate the LBB screening criterion.

Fatigue Crack Growth Properties of 16MnR and 316L Steels at High Temperature

Fatigue failures often take place in high temperature pressure vessels and equipment because of fluctuation of pressure and temperature. Fatigue crack growth properties of materials at high temperatures are very important for safety assessment of high temperature equipment. A series of fatigue crack growth tests were carried out, and fatigue crack growth rates were determined at 25∼500°C for typical steels 316L and 16MnR. The laws of fatigue crack growth of two materials at different temperatures and the effect of temperature on fatigue crack growth rates were studied. The results show that the crack growth rates increase with temperature for 316L steel. Both the exponent n and constant C for Paris law change with temperature. The fatigue cracks of 16MnR propagate at 150 °C and 300 °C more slowly than at room temperature and 425 °C. The fatigue crack growth rate at 425 °C is the highest for temperature range of 25–425 °C.

Integrity Assessment of Cracked Piping Components: Experimental, Analytical and Numerical Studies

Integrity assessment of piping components is very essential for safe and reliable operation of power plants. Over the last several decades, considerable work has been done throughout the world to develop a methodology for integrity assessment of pipes and elbows, appropriate for the material involved. However, there is scope of further development/improvement of issues, particularly for pipe bends, that are important for accurate integrity assessment of piping. Considering this aspect, a comprehensive Component Integrity Test Program was initiated in 1998 at Bhabha Atomic Research Centre (BARC), India. In this program, both theoretical and experimental investigations were undertaken to address various issues related to the integrity assessment of pipes and elbows. Under the experimental investigations, fracture mechanics tests have been conducted on pipes and elbows of 200–400 mm nominal bore (NB) diameter with various crack configurations and sizes under different loading conditions. Tests on small tensile and three point bend specimens, machined from the tested pipes, have also been done to evaluate the actual stress-strain and fracture resistance properties of pipe/elbow material. The load-deflection curve and crack initiation loads predicted by non-linear finite element analysis matched well with the experimental results. The theoretical collapse moments of throughwall circumferentially cracked elbows, predicted by the recently developed equations, are found to be closer to the test data compared to the other existing equations. The role of stress triaxialities ahead of crack tip is also shown in the transferability of J-Resistance curve from specimen to the component.

ENPOWER: Investigations by Neutron Diffraction and Finite Element Analyses on Residual Stress Formation in Repair Welds Applied to Ferritic Steel Plates

The European collaborative research project ENPOWER (Management of Nuclear Plant Operation by Optimizing Weld Repairs) has as one of its main objectives the development of guidelines for the application of repair welds to safety critical components in nuclear power plants. In this context letter box repair welds applied to thin ferritic steel plates to simulate repair of postulated shallow cracks have been manufactured for the purpose of experimental and numerical analysis of welding residual stresses. Two specimens have been procured, one of them prepared in accordance with a standard welding procedure, while in the second case a different procedure was followed in order to obtain extended martensite formation in the heat affected zone. Residual stresses have been determined in both specimens by neutron diffraction at the High Flux Reactor of the Joint Research Centre in Petten, The Netherlands. In parallel Institut de Soudure in France has performed a full 3-d analysis of the residual stress field for the standard welding case taking into account the materials and phase transformations. The experimental data obtained for both specimens clearly suggest that the non-conventional welding procedure rendered higher maximum stress values. In the case of the standard welding procedure numerical and experimental data show a reasonable qualitative agreement. The maximum stress value was in both cases found in the same region of the material — in the base metal just underneath the weld pool — and in both cases found to be of similar magnitude (∼800 MPa found in neutron diffraction and ∼700 MPa found in numerical analysis). In this paper the experimental and numerical approaches are outlined and the obtained results are presented. In addition an outlook is given to future work to be performed on this part of the ENPOWER project. A main issue pending is the application of an optimized advanced post weld heat treatment in one of the two cases and the subsequent numerical and experimental determination of its impact on the residual stress field. At the same time further evaluation of the materials transformations due to welding is pursued.

Application of Modified Damage Potential (ψ-Integral) to Predict Crack Initiation in Welded Pipes and Experimental Verification

The mesh dependency of Rice and Tracey cavity growth factor (CGF) is overcome by integrating the CGF over a process zone surrounding the crack tip. This integral represents a modified damage potential. The critical value of the integral for crack initiation in weld material has been determined analyzing a welded CT specimen and comparing the computed crack initiation J with experimentally measured J-initiation value. This critical value is then employed to predict crack initiation load in 8” and 12” welded pipes having different measure of through-wall cracks at the center to predict the loads under four point bending loads. The computed values are compared with the experimentally measured values. A close agreement between the computed crack initiation loads with the experimentally measured values justifies the usefulness of the present modified damage potential.

Estimation of Master Curve Based RTTo Reference Temperature From CVN Data

The ASME code cases N-629 and N-631 permits the use of a Master Curve-based index temperature (RTTo ≡ T0 + 19.4°C) as an alternative to traditional RTNDT-based methods of positioning the ASME KIc, and KIR curves. This approach was adopted to enable use of Master Curve technology without requiring the wholesale changes to the structure of the ASME Code that would be needed to use all aspects of Master Curve technology. For the brittle failure analysis considering irradiation embrittlement additionally a procedure to predict the adjustment of fracture toughness for EOL from irradiation surveillance results must be available as by NRC R.G. 1.99 Rev. 2 e.g.: ART = Initial RTNDT + ΔRTNDT + Margin. The conservatism of this procedure when RTNDT is replaced by RTTo is investigated for western nuclear grade pressure vessel steels and their welds. Based on a systematic evaluation of nearly 100 different irradiated material data sets, a simple relation between RTToirr, RTToref and ΔT41JRG is proposed. The relation makes use of the R.G. 1.99 Rev. 2 and enables the minimizing of margins, necessary for conventional correlations based on temperature shifts. As an example, the method is used to assess the RTTo as a function of fluence for several German pressure vessel steels and corresponding welds. It is shown that the method is robust and well suited for codification.

Application of Surface Stress Improvement Technologies to Nuclear Power Plant Components

The susceptibility of Alloy 600 to Primary Water Stress Corrosion Cracking (PWSCC) has proven detrimental to several nuclear power plant components. Repair, modification or replacement of the components to mitigate the effects of PWSCC on Alloy 600 has been deemed necessary. In some cases, repair or replacement of plant components can be exorbitantly expensive; therefore, modification of the components is necessary to keep the plant operable. A form of modification is surface stress improvement, which alters the stress state of the material. Changing the stress state of the material eliminates one of the contributing factors required for the propagation of PWSCC. This paper discusses the application of surface stress improvement technologies to commercial nuclear power plant components and provides insight to where these technologies can be employed in the future.