Proposal of Determination Procedure About Fatigue Safety Factors for Arbitrary Best Fit Curves

Fatigue safety factors cover not only the scattered range of fatigue test results obtained from test specimens, but also the difference in the fatigue lives between the test specimens and vessels / piping in plants. Therefore necessary safety factors have a relation to the accuracy of a best fit curve which is determined by the test results and the component’s conditions. This paper describes a determination procedure of fatigue safety factors for a new best fit curve so that designers can determine adequate safety factors for arbitrary best fit curves and material databases. This determination procedure is explained using a 316FR fatigue database, and the safety factors for new 316FR best fit curve are shown. These safety factors are estimated based on the accuracy of the best fit curve, and are calculated for each surface finish in the designed vessels and piping. Therefore the adoption of accurate best fit curves in limited conditions will result in smaller safety factors than the present code.

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

The Sensitivity of Pressurized Thermal Shock Analysis Results to Alternative Models for Weld Flaw Distributions

The United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project in 1999 to determine if improved technologies can provide a technical basis to reduce the conservatism in the current regulations for pressurized thermal shock (PTS) while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license renewal considerations. During the PTS re-evaluation study, an improved risk-informed computational methodology was developed that provides a more realistic characterization of PTS risk. This updated methodology was recently applied to three commercial PWRs. The results of this study provide encouragement that a technical basis can be established to support a relaxation of current PTS regulations. One significant model improvement applied in the PTS re-evaluation study was the development of flaw databases derived from the non-destructive and destructive examinations of material from cancelled reactor pressure vessels (RPV). Empirically-based statistical distributions derived from these databases and expert illicitation were used to postulate the number, size, and location of flaws in welded and base metal (plate and forging) regions of an RPV during probabilistic fracture mechanics (PFM) analyses of RPVs subjected to transient loading conditions such as PTS. However, limitations in the available flaw data have required assumptions to be made to complete the risk-based flaw models. Sensitivity analyses were performed to evaluate the impact of four flaw-related assumptions. Analyses addressed: 1) truncations of distributions to exclude flaws of extreme depth dimensions, 2) vessel-to-vessel differences in flaw data, 3) large flaws observed in weld repair regions, and 4) the basis for estimating the number of surface breaking flaws. None of the four alternate weld flaw models significantly impacted calculated vessel failure frequencies or invalidated the tentative conclusions derived from the USNRC PTS re-evaluation study.

Investigation of Stress Indices and Directional Loading of Eccentric Reducers

Stress indices and stress intensification factors are used in the design of piping systems that must meet the requirements of ASME Section III for Class 1 and Class 2 systems. This study reviews the present values for eccentric reducers and provides new test data for comparison, which takes into account the directionality of the loading. Suggestions are presented which significantly improve the evaluation of reducers.

Design of an Optimum I-Beam Reinforcement for an API 620 Tank

The objective of this paper is to analyze an existing American Petroleum Institute (API) 620 Tank [10]. The API Tank had failed in the field. The tank is analyzed without reinforcement and with an optimum I-Beam reinforcement. The API Tank is used to store chemicals used in today’s industry. The initial over-all dimensions of the API Tank are determined from the capacity of the stored chemicals. The design function is performed using the ASME Code See VIII Div 1. The API Tank design is broken up into (a) bottom plate, (b) shell section with 9 mm thickness, (c) shell section with 8 mm thickness, (d) shell section with 7 mm thickness, (e) shell section with 6 mm thickness, (f) shell section with 5 mm thickness, (g) top head with 5mm thickness, (h) bolts, and (i) reinforcement ring. The designed dimensions are used to recalculate the stresses for the complete API Tank. The dimensioned API Tank without reinforcement is modeled first using STAAD III finite element software. The stresses from the finite element software are obtained. Next the API Tank with I-Beam reinforcement was modeled using STAAD III finite element software. Ten different I-Beams were considered for the present analysis. The main objective of this paper was to find the optimum I-Beam that resulted in safe reinforced configuration. Optimum I-Beam was considered to be the one that resulted in similar stresses for the beam as well as the tank. This assures elastic matching between the beam and the tank. The design is found to be safe for the I-Beam reinforced configuration considered.

CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions for Lifetime Evaluation: Methodology and Qualification Task of the EDF Numerical Tools

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behaviour of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA transients. This paper presents the Research and Development program started at E.D.F on the CFD determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermalhydraulic tools N3S and Code_Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. We first explain the recent improvement of the thermalhydraulic analysis with the global definition of the SBLOCA transient and the local analysis in the downcomer. Then, the qualification task of the EDF numerical tools is described. In order to reach this purpose, we have investigated several configurations related to an injection of cold water and focused our analysis particularly in the cold leg but also in a the downcomer. Two experiment test cases have been studied. A comparison between experiment and numerical results in terms of temperature field is presented. On the whole, the main purpose of the numerical thermalhydraulic studies is to accurately estimate the distribution of fluid temperature in the downcomer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margins.

Qualification of a Jacketed Vessel Using Finite Element Analysis

The main objective of this paper is to improve a jacketed vessel. The jacketed vessel is usually chosen to heat the contents of the vessel. The chamber or annulus contains fluid under pressure to heat the inner vessel contents. The initial over-all dimensions of the vessel are based on the capacity of the stored liquid. The design was in accordance with the ASME Boiler & Pressure Vessel Code, Section VIII, Div 1. The jacketed vessel bottom head and jacket bottom head are being improved to withstand internal and external design pressures. Bottom head of the jacket can be reinforced in one of the three ways, namely: (1) rings which are radial (these rings also create flow for the fluid); (2) attachment of the rings to the bottom jacket head with stays, since rings cannot be physically welded to the bottom jacket; or (3) there is a possibility, the new bottom head and jacketed head combination can be cast, but that would not be economically feasible. This leads to the following six configurations considered in this paper and they are: (1) internal pressure of 50 psi, (2) external pressure + vacuum pressure of 65 psi, (3) reinforcement with 5 rings with external pressure of 65 psi, (4) rings welded with the bottom jacket head with external pressure of 65 psi, (5) welded with stays on ring location (stay diameter of 1 inch) with external pressure of 65 psi, and (6) welded with stays on ring location (stay diameter of 1.5 inch) with external pressure of 65 psi. The pattern of stays chosen for this analysis is one of uniform distribution on ring locations, which are radially situated. The design dimensions based on Code sizing are used to recalculate the stresses for the jacket vessel. The dimensional jacketed vessel is modeled using STAAD III Finite Element Analysis (FEA) software. The design is found to be safe for the specific configuration considered herein with stays.

Considerations in the Design and Analysis of an ASME Section VIII, Div. 2 Reactor Support Skirt

This paper describes the considerations employed in the finite element analysis of a relatively “short” support skirt on a hydrocarbon reactor vessel. The analysis is accomplished in accordance with ASME B&PV Code, Section VIII, Division 2 alternate rules in conjunction with the guidelines outlined in WRC Bulletin 429. This provides a sound basis for the classification of the calculated stress intensities. The support skirt is capable of sustaining the deadweight load in addition to resisting the effects of thermal displacements, wind loadings, overturning moments from external piping loads on the attached hydrocarbon reactor vessel, and friction between the skirt base plate and concrete foundation. The displacement and thermal boundary conditions are well defined and discussed in detail. The effects of multiple scenarios for the displacement boundary conditions are examined. The skirt design also employs a hot-box arrangement whereby the primary mode of heat transfer is by radiation. A discussion of the two-part analysis is included and details the interaction between the heat transfer analysis and the subsequent structural analysis. The heat transfer finite element analysis is utilized to determine the temperatures throughout the bottom of the vessel shell and head, as well as the integrally attached support skirt. Of prime importance during the analysis is the axial thermal gradient present in the skirt from the base plate up to and slightly beyond the skirt-to-shell junction. While the geometry of the subject vessel and skirt is best described as axisymmetric, the imposed loadings are a mixture of axisymmetric and non-axisymmetric. This combination lends itself to the judicious selection and utilization of the harmonic finite element and properly chosen Fourier series representation of the applied loads. Comparison of the thermally induced axial stress gradient results from the FEA to those obtained by the closed form beam-on-elastic-foundation are also tendered and discussed. Finally, recommendations are included for the design and analysis of critical support skirts for large, heavy-wall vessels.

Presentation of a Numerical 3D Approach to Tackle Thermal Striping in a PWR Nuclear T-Junction

This paper presents a numerical study to tackle thermal striping phenomena occuring in piping systems. It is here applied to the Residual Heat Removal (RHR) bypass system. A large Eddy Simulation (L.E.S.) approach is used to model the turbulent flow in a T-junction. The thermal coupling between the Finite Volume CFD Code_Saturne and the Finite Element thermal code Syrthes, gives access to the instantaneous field inside the fluid and the solid. By using the instantaneous solid thermal fields, mechanical computations (as presented in (Stephan et al 2002)) are performed to yield the instantaneous mechanical stresses seen by the pipework T-junction and elbow.

Deformation Relaxation: A Finite Element Optimization Method for Tee

In this paper, an optimization method for a tee is proposed. A finite element model of a tee is built and its reliability is demonstrated by experiment. Analysis models with different fillet radii are analyzed. The results show that changing the fillet radius can not distinctly change the stress intensity level. Therefore, deformation relaxation, a new method for lowering stress intensity levels, is proposed. The method’s core is that the original shape of the tee is modified by deformation displacement. The computed results show that the decrease of the highest stress intensity is approximately in proportion to the modification. When the maximum modification is up to 10% of the outer diameter of main pipe, the highest stress intensity decreases by 52.2%.