In the last 7 years, the incidences of cracking in Alloy 600 control rod drive mechanism (CRDM) tubes and their associated welds have increased significantly. The cracking mechanism has been attributed to pressurized water stress corrosion cracking (PWSCC) and has been shown to be driven by welding residual stresses and operational stresses in the weld region. During this time period, both the industry and the US Nuclear Regulatory Commission have been conducting detailed welding simulation analyses to predict the magnitude of these stresses in both the weld and tube material. To this point, a direct comparison of these analysis methodologies and results has not been made. In this paper, weld residual stress results from U.S. industry (conducted by Dominion Engineering) and the U.S. NRC (conducted by Engineering Mechanics Corporation of Columbus) for a steep angle (53 degrees) CRDM nozzle are compared. This comparison was performed for different yield strength tube materials, however only the low yield strength results are presented in this paper. The comparison illustrates the effect of weld analyses assumptions and suggests that simplifications in the analyses, i.e., lumping weld passes or material property assumptions, may lead to high predicted weld residual stresses.

The Atucha II nuclear power plant is a unique pressurized heavy water reactor being constructed in Argentina. The original plant design was by KWU in the 1970’s using the then German methodology of break preclusion, which assumed that the largest break-opening area would be 10-percent of the cross-sectional area of the largest pipe diameter. That philosophy was used for the design of the emergency core cooling system in the 1970’s. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The construction is progressing with initial start-up in 2011. Since the 10-percent of the cross-sectional area is a smaller ECCS design requirement than the normally assumed double-ended-guillotine break, the safety evaluation of the plant for beyond design basis seismic loading of the nuclear plant was a regulatory requirement. This overview paper describes a Robust LBB Evaluation that was conducted in great detail to assess the safety aspects of the piping system under beyond design basis seismic loading and the implications to the ECCS. Key aspects involved: • Static and dynamic material property testing, • Determination of weld residual stresses, • Determination of crack sizes that might evolve by worst case SCC growth rates under weld residual stresses and normal operating stresses, • Determination of leakage rates as a function of time with the upper-bounding crack growth rates, • Development of seismic hazard curves for the site, • Development of FE models of the containment building and primary NSSS system within the building, • Determination of normal operating stresses, SSE stresses and 10−6 seismic stresses using worst case soil foundation assumptions, • Evaluation of flaw behavior for circumferential cracks using the shapes from the natural crack growth. • Evaluation of margins on the critical flaw size and times to leakage, and • Standard LBB analyses, as well as Transition Break Size evaluations. The key result from this effort was that even with all the normal operating plus 10−6 seismic event loading, the pipe system behaved more like it was displacement-controlled than load-controlled. The displacement-controlled behavior made the pipe much more flaw tolerant, and it was found that a DEGB was not possible because the flaw could never reach the critical flaw size without greatly surpassing the leakage and water make-up capacity of the plant. Since there are many details in this multi-year effort, only the key points will be summarized in this paper while other details will be the topics of other papers.

During the last year, defects had been located by ultrasonic testing in three of the pressurizer nozzle dissimilar metal (DM) welds at the Wolf Creek nuclear power plant. Understanding welding residual stress is important in the evaluation of why and how these defects occur, which in turn helps to determine the reliability of nuclear power plants. The analysis procedure in this paper included not only the pass-by-pass welding steps, but also other essential fabrication steps of pressurizer surge nozzles. Detailed welding simulation analyses have been conducted to predict the magnitude of these stresses in the weld material. Case studies were carried out to investigate the influences to main weld stress fields with different boundary conditions, material strength, weld sequencing, as well as simulation of the remaining piping system stiffness. A direct comparison of these analysis methodologies and results has been made in this paper. Weld residual stress results from nuclear industry (conducted by Dominion Engineering, Inc.) and the US NRC (conducted by Engineering Mechanics Corporation) are also compared.

Computational weld modeling is challenging because many of the processes of welding are highly nonlinear. Material melts and re-solidifies, very high transient thermal gradients are experienced, non-linear temperature dependent plastic straining and phase transformations can occur, among other sources of nonlinearity. Moreover, for weld modeling to have practical advantages in industrial production, computational solution times must be manageable since an optimum weld design of large, complex fabrications requires numerous separate analyses. Weld modeling technology is now advanced to where it can have an important impact on numerous fabricated structures. These include nuclear power plant components in commercial nuclear plants and nuclear ship structures, including Aircraft Carriers, Submarines, and Destroyers. The benefits of weld modeling include: • Significantly reduced Fabrication Costs. • Life cycle cost reduction from improved corrosion, and fatigue performance and damage reduction enhancement. • Elimination of non-valued added re-work fabrication costs. • Improve readiness by speeding the time from conception to service for new designs. • Outreach program to continue paradigm shift improvements in welded fabrications. Weld distortion control must be performed on three dimensional models. Here, extensive full-scale experiments have validated the accuracy and predictive power of models. It can be used to reduce fabrication cost and improve quality by minimizing and controlling distortions. Several application examples are presented to illustrate how to apply this tool in welded structure design and manufacture.

Small indications were found in one replacement reactor pressure vessel head (RPVH) mock-up being fabricated from Alloy 690 material and compatible weld metals, Alloy 52/152. The mockups were non-destructively examined and the lowest number of cracks found was five and the highest number was 22. There are numerous indications with some of them quite long (50 mm) in length. The source of these weld fabrication cracks is unknown. However, from experience with other difficult to weld materials, the source can range from slag inclusions in the weld metal to hot cracking during the weld deposition process. Hot cracking includes solidification cracking (weld), liquation cracking (HAZ), and ductility dip cracking (DDC). The indications were mostly circumferential in orientation (with respect to the nozzle axis) but some were axial. This paper includes two parts. The first part includes the welding residual stress analysis of RPVH using Alloy 52/152 metal and provides comparison with similar Alloy 82/182 welds. Alloy 82/182 was the material used in the original dissimilar metal welds in these heads. Primary Water Stress Corrosion Cracking (PWSCC) can occur in the primary coolant system when the welds are exposed to water, tensile stress, and temperature (usually higher than 250 C). PWSCC rates are higher in Alloy 82/182 material due to its lower chromium content compared with the replacement alloy. The results for both center hole (0-degree) and side hill (53-degree) nozzles will be discussed. The second part deals with assessment of multiple small cracks in the reactor pressure vessel head penetration nozzles. The finite element alternating method (FEAM) was used for calculating stress intensity factors for cases where multiple cracks exist. More than twenty cracks, which were inserted based on field measurements, are considered in the analyses for both center hole and side hill nozzles. It is observed that the overall stress trends are similar to those without adding cracks. However, cracks introduce more local stress fluctuations. The magnitude of the local fluctuation can be around 100MPa. Limit analysis was also conducted. A new finite element model with a voided-out weld region was used to simulate loss of structural capacity due to multiple flaws. The voided out volume effects on the structural integrity and future performance of RPVH were examined. Discussions based on weld residual stress, multiple flaw analysis and limit analysis conclude the paper.