Nuclear Piping Components Research Articles

Effect of cross section on B2 stress index for Nuclear Pipe Bends subjected to Closing Bending

Nuclear piping components have been designed against plastic loads using the guidelines provided by section III, NB-3652 of Boiler & Pressure vessel code, ASME. For safe operation in the power plants, the structural integrity assurance of pipe bends is essential. The B2 stress index is used in the design equation for evaluating the primary stresses due to moment loading in the piping components. The code based B2 stress index is expressed in bend characteristic only. In real practice, consideration of bend angle, bend radius, nominal pipe thickness, and material properties of the pipe bends are needed. This paper presents a detailed investigation of the combined effects of variable wall thinning and ovality on B2 stress indices for pipe bends under closing moment using three dimensional finite element nonlinear analyses. Twice elastic slopemethodology (TES) was employed to get the collapsing moment from load deflection curve drawn for each model. From finite element analysis, it is observed that, the wall thinning produces a negligible effect on the B2 stress index, whereas the ovality produces significant effect.

Comparative analysis of deterministic and probabilistic fracture mechanical assessment tools

Abstract Uncertainties in material properties, manufacturing processes, loading conditions and damage mechanisms complicate the quantification of structural reliability. Probabilistic structure mechanical computing codes serve as tools for assessing leak- and break probabilities of nuclear piping components. Probabilistic fracture mechanical tools were compared in different benchmark activities, usually revealing minor, but systematic discrepancies between results of different codes. In this joint paper, probabilistic fracture mechanical codes are compared. Crack initiation, crack growth and the influence of in-service inspections are analyzed. Example cases for stress corrosion cracking and fatigue in LWR conditions are analyzed. The evolution of annual failure probabilities during simulated operation time is investigated, in order to identify the reasons for differences in the results of different codes. The comparison of the tools is used for further improvements of the codes applied by the partners.

적외선열화상 시험에서 위상잠금모드 적용에 따른 배관 감육결함 검출능력 개선 평가

The lock-in mode infrared thermography (IRT) technique has been developed to improve the detection capability of defects in materials with high thermal conductivity, and it has been shown to provide better detection capability than conventional active IRT. Therefore, to investigate the application of this technique to nuclear piping components, lock-in mode IRT tests were conducted on pipe specimens containing simulated wall-thinning defects. Phase images of the wall-thinning defects were obtained from the tests, and they were compared with thermal images obtained from conventional active IRT tests. It showed that the ability to size the detected wall-thinning defects in piping components was improved by using lock-in mode IRT. The improvement was especially apparent when detecting short and narrow defects and defects with slanted edges. However, the detection capability for shallow wall-thinning defects did not improve much when using lock-in mode IRT.

원전 배관 감육 결함 검사를 위한 IR 열화상시험 조건 결정

This study conducted infrared (IR) thermography tests using pipe and plate specimens with artificial wall-thinning defects to find an optimal condition for IR thermography test on the wall-thinned nuclear piping components. In the experiment halogen lamp was used to heat the specimens. The distance between the specimen and the lamp and the intensity of halogen lamp were regarded as experimental parameter. When the distance was set to 1~2 m and the lamp intensity was above 60 % of full power, a single scanning of IR thermography detected all artificial wall-thinning defects, whose minimum dimension was $2{\Theta}

Safety assessment of austenitic steel nuclear power plant pipelines against stress corrosion cracking in the presence of hybrid uncertainties

Stress corrosion cracking (SCC) is an important degradation mechanism to be considered for safety assessment of nuclear piping components made of austenitic steels, especially in the heat-affected zones. Damage due to SCC occurs in a susceptible material, in a corrosive environment, in the presence of high temperature and high applied/residual stresses. The operating conditions and the environmental conditions show variations during the lifetime of the power plant. Also, there will be variations in micro-structural properties of the material of piping components. These variations should be taken into account while assessing the safety of the piping component against SCC. This can be accomplished by treating the relevant variables as random or fuzzy depending upon the source and type of uncertainty. In this paper, an attempt has been made to compute the fuzzy failure probabilities of a piping component against SCC with time, using an approach combining the vertex method with the Monte Carlo simulation technique. The initiation and propagation stages of stress corrosion cracks are modelled using a modified PRAISE approach. The degree of sensitisation, material fracture toughness, yield strength, ultimate strength and applied stress are considered as random variables, while operating temperature and oxygen concentration are considered as fuzzy variables. The R6 procedure is used in the computation of the fuzzy failure probabilities. The usefulness of the approach is demonstrated through an example problem.

Probabilistic Failure Analysis of Austenitic Nuclear Pipelines against Stress Corrosion Cracking

Stress corrosion cracking (SCC) is an important degradation mechanism to be considered for failure assessment of nuclear piping components made of austenitic steels. In this paper, an attempt has been made to compute the failure probabilities of a piping component against SCC with time using Monte Carlo simulation (MCS) technique. The initiation and propagation stages of stress corrosion cracks are modelled using the general methodology recommended in PRAISE modified by using the recommendations given by ASM for more rational modelling of stress field around cracks for estimating their growth with time. Degree of sensitization, applied stress, time to initiation of SCC, initial crack length, and initiation crack growth velocity are considered as random variables. An attempt has been made to study the stochastic propagation of stress corrosion cracks with time, using MCS technique. The trend of the distribution of crack depths at the initial stages obtained from simulation are compared and is found to be in satisfactory agreement with the relevant experimental observations reported in the literature. The failure probabilities are computed using two different failure criteria, namely (a) based on net-section stress and detectable leak rate as recommended in PRAISE and (b) based on R6 approach (using R6-option 1 curve as the failure assessment diagram). The procedure presented in the paper is general and the usefulness of the same is demonstrated through an example problem.

B 2 and C2 Stress Indices for Large-Angle Bends

B 2 and C2 stress indices are needed for evaluation of nuclear piping components. These indices are used in calculation of stress terms corresponding to moment loading. The formulas (B2=1.3/h2/3 and C2=1.95/h2/3) given in the ASME Boiler and Pressure Vessel Code for calculating these stress indices for elbows and bends are independent of bend angle. Author’s earlier work indicated that the values obtained by these formulas are conservative for bend angle ⩽90 deg. The objective of the present study was to investigate the values of B2 and C2 stress indices for elbows with bend angles larger than 90 deg and to evaluate the limit of the applicability of the ASME Code formulas in terms of the bend angle. For this purpose, elbows with bend angles ranging from 90 to 180 deg have been investigated in detail. Elbow sizes studied ranged from 2.5 to 20 NPS. Finite element models have been used for detailed analysis. To cover the worst case of stresses in the elbows, several hundred load cases were analyzed by varying the direction of the moment loading. It was found that the three independent loading cases corresponding to three orthogonal axes do not bound the worst-case stress in the elbows. The results of the analysis show that the calculated value of B2 stress index increases with bend angle; but its value, even for 180-deg bend angle, remains below the ASME Code value with a margin of at least 15%. Considering all analyzed elbows and bend angles from 90 to 180 deg, the calculated maximum value of C2 stress index is found to be within 2 percent of the ASME Code value. The maximum increase in the value of B2 stress index is 12.5 percent as the bend angle is increased from 90 to 180 deg. The corresponding increase in the C2 stress index is less than 8%. Based on these results and authors’ earlier work, a modification to ASME Code formulas is proposed for calculation of more realistic values of the stress indices for bends up to 180 deg.

Bend Angle Effect on B2 and C2 Stress Indices for Piping Elbows

B 2 and C2 stress indices are used for calculation of stress terms corresponding to moment loading in evaluation of nuclear piping components. The formulas given in the ASME Boiler & Pressure Vessel Code for calculating these indices for elbows and curved pipes are independent of the bend angle. The objective of this investigation was to study the effect of the bend angle on the values of B2 and C2. For this purpose, elbows with sizes ranging from 2.5 to 20 in. have been studied. The values of B2 and C2 have been calculated by detailed analysis using finite element models. Each elbow size has been analyzed for various bend angles (e.g., 15, 30,…90 deg and above). The results of this study show that the values of B2 and C2 based on the ASME Code formulas are conservative and more so if the bend angle is less than 90 deg. The values of B2 and C2 decrease as the bend angle is decreased. The reduction in the values is very significant for smaller bend angles. Simple relations have been established to express the dependence of B2 and C2 on the bend angle. Based on these results, a modified form of ASME Code formulas is proposed to account for the bend angle effect.

Fracture analyses of surface-cracked pipes and elbows using the line-spring/shell model

The line-spring/shell model proposed by Rice and Levy (1972) and further modified by Parks and co-workers (1981, 1982, 1995) provides an attractive and inexpensive alternative to analyzing surface-cracked structures. This model idealizes the complex three-dimensional surface crack problem in a plate- or shell-like structure as a more tractable two-dimensional problem. Though used in some previous studies, this model has largely been ignored in the study of surface-cracked nuclear piping components. While it is true that the line-spring method has some restrictions, it can provide reasonably accurate predictions of important fracture parameters of cracked components within the domain of its applicability. Through some detailed studies of surface-cracked pipes and elbows, the predictive nature of the line-spring model is demonstrated by comparing its predictions with that of a three-dimensional finite element model as well as with experimental measurements. Excellent comparisons between the predictions of the line-spring/shell and three-dimensional finite element models are shown not only using the three-dimensional finite element calculations performed by the author, but also using the three-dimensional calculations performed by two other independent investigators. It is hoped that the results shown in this paper invalidate the criticisms and reestablish the usefulness and reasonable accuracy of the line-spring mode, within its domain of applicability, in the analyses of surface-cracked structures.

Contributions of the ORNL Piping Program to Nuclear Piping Design Codes and Standards

Since 1967, ORNL Piping Program has been engaged in providing information for the development of stress indices to be used in the analysis of piping components for nuclear power plants. This effort has surveyed the piping manufacturing industry and analyzed that industry’s products; has combed the technical literature for pertinent engineering data; has performed theoretical and experimental analysis of nuclear piping components; and has defined, tested, and improved indices for the stress-index method of analysis for piping components. This paper briefly reviews the history of piping-analysis standards; outlines the philosophy of the stress-index method of analysis; and explains some of the specific contributions made by the ORNL program to the Codes and Standards. Current and future work is also noted.