An ASME Section III Task Group (TG) was formed in 2012 to develop alternate rules for the design assessment of Section III Class 1 nuclear components subject to fatigue service with environmental effects. Specifically, a flaw tolerance approach is being investigated based on similar methodology to that found in ASME Section XI Nonmandatory Appendix L. A key initial task of the TG (which reports to the Section III Working Group on Environmental Fatigue Evaluation Methods) was to develop and solve a detailed sample problem. The intent of the sample problem was to illustrate application of proposed rules, which will be documented as a Section III Code Case with a supporting technical basis document. Insights gained from round robin solution of the sample problem are presented and discussed in this paper. The objective of documenting the findings from the sample problem are to highlight the observed benefits and limitations of the proposed procedures, particularly how rules typically associated with in-service experience might be adapted into design methods. The sample problem is based on a heavy-walled stainless steel nozzle that meets cumulative fatigue usage requirements in air (i.e., usage factor, U, without reactor water environment effects less than unity), but fails to meet usage factor requirements when environmental fatigue effects are applied. The sample problem demonstrates that there is a class of problems dominated by severe thermal transients where fatigue initiation is predicted based on elastic methods including environmental effects, but fatigue crack propagation results are acceptable. Preliminary conclusions are drawn based on the results of the sample problem, and the next steps are also identified.
Section III, Division 1 and Section VIII, Division 2 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code provide procedures for demonstrating shakedown using elastic-plastic analysis. While these procedures may be used in place of elastic analysis procedures, they are typically employed after the elastic analysis and simplified elastic-plastic analysis limits have been exceeded. In using the Section III, Division 1 and Section VIII, Division 2 procedures for elastic-plastic shakedown analyses, three concerns are raised. First, the Section III, Division 1 procedure is vague, which can result in inconsistent results between analysts. Second, the acceptance criteria contained in both procedures are vague, which can also result in inconsistent results between analysts. Lastly, differences in the procedures and acceptance criteria can result in demonstration of component elastic-plastic shakedown under Section III, Division 1 but not under Section VIII, Division 2. The authors presume that the ASME Code intends to provide similar design and analysis conclusions, which may not be a correct assumption. To demonstrate these concerns, a nozzle benchmark design subject to a representative thermal and pressure transient was evaluated using the two Code elastic-plastic shakedown procedures. Shakedown was successfully demonstrated using the Section III, Division 1 procedure. However, shakedown could not be demonstrated using the Section VIII, Division 2 procedure. The conflicting results seem to indicate that, for the nozzle design evaluated, the Section VIII, Division 2 procedure is considerably more conservative than the Section III, Division 1 procedure. To further assess the conservative nature of the Section VIII, Division 2 procedure, the nozzle benchmark design was evaluated using the same thermal transient, but without a pressure load. While shakedown was technically not observed using the Section VIII, Division 2 acceptance criteria, engineering judgment concluded that shakedown was demonstrated. Based on the results of all the evaluations, recommendations for modifications to both procedures were presented for consideration.
Fatigue usage factor evaluations including the effects of reactor water environment have been performed in numerous nuclear plant license renewal efforts. A large number of these evaluations have used the environmental fatigue penalty factor, Fen, approach prescribed in various regulatory documents. The Fen equations require input of strain rate, but the prescribing documents do not provide methodology or criteria for the quantification of the strain rate to be input. As a result, numerous approaches have been offered and studied. This paper presents an approach used by Westinghouse to include strain rate in an automated calculation of Fen based on the modified rate approach (MRA) to integrated strain rate applications. The starting point of the approach is ASME Code Section III NB-3200 fatigue analysis. With environmental fatigue evaluations in new plant designs now emerging in ASME Code criteria, strain rate considerations remain part of the discussion. The intent of this paper is to provide further insight into this process.
Nuclear power plants need to safely and efficiently remove their reactor vessel closure head assembly during plant outages. This is accomplished by lifting the closure head assembly out of the reactor vessel cavity and placing it on the closure head stand. In order for nuclear power plants to remove their closure head assembly, the United States Nuclear Regulatory Commission has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis [1]. The goal of head drop analyses is to qualify the maximum drop height in air per plant procedures. A significant percentage (greater than 30%) of the closure head assembly’s mass is comprised of components attached to the top of the head (such as: lifting fixtures, a missile shield, air cooling systems, and control rod drive housings). The analytical consideration of large deflection, plastic deformation, and local failure of these components can potentially change the energy imparted to the vessel during impact due to their energy-absorbing capacities during the drop event. This paper contains a sensitivity study to determine the benefits of modeling closure head assembly components, using nonlinear structural behavior. The guidelines of Nuclear Energy Institute Initiative NEI 08-05 [2] are followed for this study.
Fatigue usage factor evaluations including the effects of reactor water environment have been performed in numerous nuclear plant license renewal efforts. A large number of these evaluations have used the environmental fatigue penalty factor, Fen, approach prescribed in various regulatory documents. The Fen equations require input of strain rate, but the prescribing documents do not provide methodology or criteria for the quantification of the strain rate to be input. As a result, numerous approaches have been offered and studied. This paper presents an approach used by Westinghouse to include strain rate in an automated calculation of Fen based on the modified rate approach to integrated strain rate applications. The starting point of the approach is ASME Code Section III NB-3200 fatigue analysis. With environmental fatigue evaluations in new plant designs now emerging in ASME Code criteria, strain rate considerations remain part of the discussion. The intent of this paper is to provide further insight into this process.
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