Nuclear Power Plant Piping Systems Research Articles

Thermomechanical and isothermal fatigue behavior of 347 and 316L austenitic stainless tube and pipe steels

Many components of nuclear power plant piping systems are made of austenitic stainless steels. These structures undergo degradation by thermomechanical loading caused by thermal transients and stratifications. In scientific literature, most of the studies deal with this problem under isothermal fatigue conditions, which is different from the typical service conditions. In addition, less attention has been paid to thermo-mechanical fatigue ( TMF). This work aims to understand and compare the cyclic deformation behavior during TMF and isothermal fatigue ( IF) testing in air for two of the most commonly used grades of austenitic stainless tube and pipe ( TP) steel, the non-stabilized TP 316L and the niobium-stabilized TP 347, under light water reactor relevant temperature conditions. Three types of tests, i.e. in-phase, out-of-phase TMF in the 100–340 °C temperature range, and IF tests at the maximum temperature T max of TMF were performed. All the tests were carried out under total strain control, for two different mechanical strain amplitudes (Δ ε mech /2 = 0.3% and 0.5%). Results revealed that irrespective of the tested strain amplitude and material, in-phase TMF showed a higher lifetime over out-of-phase TMF and IF at T max . A crossover in fatigue lifetime was observed in TP 316L between out-of-phase TMF and IF test at T max with decreasing strain amplitude. In TP 347, out-of-phase TMF and IF test at T max showed similar fatigue lifetimes for both strain amplitudes. Finally, three lifetime prediction models were selected and compared to evaluate the possibility of estimating TMF lifetime by using IF data. Satisfactory predictions were obtained from all the prediction approaches.

04/01291 Markov models for evaluating risk-informed in-service inspection strategies for nuclear power plant piping systems: Fleming, K. N. Reliability Engineering & System Safety, 2004, 83, (1), 27–45

04/01291 Markov models for evaluating risk-informed in-service inspection strategies for nuclear power plant piping systems: Fleming, K. N. Reliability Engineering & System Safety, 2004, 83, (1), 27–45

Database development and uncertainty treatment for estimating pipe failure rates and rupture frequencies

Estimates of failure rates for nuclear power plant piping systems are important inputs to Probabilistic Risk Assessments (PRA) and risk informed applications of PRA. Such estimates are needed for initiating event frequencies for Loss of Coolant Accidents and internal flooding events and for risk informed evaluations of piping system in-service inspection programs. A critical issue in the estimation of these parameters is the treatment of uncertainties, which can exceed an order of magnitude deviation from failure rate point estimates. Sources of uncertainty include failure data reporting issues, scarcity of data, poorly characterized component populations, and uncertainties about the physical characteristics of the failure mechanisms and root causes. A methodology for quantifying these uncertainties using a Bayes' uncertainty analysis method was developed for the EPRI risk informed in-service inspection program and significantly enhanced in subsequent applications. In parallel with these efforts, progress has been made in the development of pipe failure databases that contain the quantity and quality of information needed to support piping system reliability evaluations. Examples are used in this paper to identify technical issues with previous published estimates of pipe failure rates and the numerical impacts of these issues on the pipe failure rates and rupture frequencies are quantified.

A synthesis formulation of explicit damping matrix for non-classically damped systems

In computing the dynamic response of a connected system with multiple components having dissimilar damping characteristics, often referred to as non-classically damped system such as nuclear power plant piping systems supported by stiff structures, one needs to define the system-level damping based upon the damping information of components. This paper presents a general formulation for computing rigorous explicit damping matrices for multiply connected, non-classically damped, coupled systems. The algorithm derived was based on a synthesis process developed by Hurty [AIAA J. 3 (4) (1965)], in which three different kinds of damping were utilized to describe the energy dissipation of the system. It was also shown that the system damping developed by the synthesis approach could reduce to damping formulations available in the literature such as those of Gupta et al. [Seismic response of non-classically damped systems. Nucl. Eng. Design, vol. 91, 1986] and Pajuhesh et al. [Determination of composite damping matrices. In: Proceedings on Fifth World Conference on Earthquake Engineering, vol. 1, 1973] when the same assumptions were applied for the interface connectivity. The method presented in this paper was used in developing the rigorous system matrices for the BNL solution to a set of benchmark problems analyzed in the NRC-BNL benchmark program for the evaluation of the state-of-the-art methods and computer programs for non-classically damped systems [Benchmark Program for the Evaluation of Methods to Analyze Non-Classically Damped Coupled Systems (2000)]. The details of the NRC-BNL benchmark program are presented in a companion paper [A NRC-BNL Benchmark Evaluation of Seismic Analysis Methods for Non-Classically Damped Coupled Systems (in press)].

Effect of measurement volume size on turbulent flow measurement using ultrasonic Doppler method

This paper presents the results of an investigation on the effects of measurement volume size on the mean velocity profile and the Reynolds stress for fully developed turbulent pipe flows. The study employs the ultrasonic velocity profile method, which is based on the ultrasonic Doppler method. The ultrasonic Doppler method offers many advantages over conventional methods for flow rate measurement in the nuclear power plant piping system. This method is capable of measuring the instantaneous velocity profile along the measuring line and is applicable for opaque liquids and opaque pipe wall materials. Furthermore, the method has the characteristic of being non-intrusive. Although it is applicable to various flow conditions, it requires a relatively large measurement volume. The measurement volume of the present method has a disk-shape determined by the effective diameter of the piezoelectric element and the number of the wave cycles of the ultrasonic pulse. Considering this disk-shaped measurement volume and expressing the time-averaged velocity in a truncated Taylor series expansion around the value at the center of the measuring control volume, the value of the velocity can be obtained. The results are then compared with the data obtained from DNS and LDA measurements. The result shows that the effect of the measurement volume size appears in the buffer region and viscous sublayer.

Markov models for evaluating risk-informed in-service inspection strategies for nuclear power plant piping systems

Abstract As part of an EPRI sponsored research project to develop technology for risk informed in-service inspection evaluations, new methods and databases were developed to predict piping system reliability. The methods include a Markov modeling technique for predicting the influence of alternative inspection strategies on piping system reliability, and Bayes' uncertainty analysis methods for quantifying uncertainties in piping system reliability parameters. This article describes these methods and associated databases needed for their quantification with particular emphasis on the application of the Markov piping reliability model. Insights are developed regarding reliability metrics that should be used in Probabilistic Risk Assessments for estimating time dependent frequencies of loss of coolant accidents and internal flooding events. The methodology for developing estimates of all the input parameters of the piping reliability models is described including the quantitative treatment of uncertainties in risk informed applications. Examples are presented to demonstrate the practical aspects of applying the Markov model and developing the inputs needed for its quantification.

802 ステンレス鍛鋼品の疲労き裂進展特性について

LBB evaluation guide in Nuclear Power Plants (NPP) was developed based on material properties of austenitic stainless steel piping in Japan. On the other hand, we use stainless steel forging in piping systems of NPPs. On the LBB evaluation, both the estimation of fatigue crack propagation and evaluation of crack instability play important roles. So we prepared stainless steel forging to obtain material properties such as strength, fracture toughness and fatigue crack propagation data in the simulated primary coolant of PWR. As for the strength and fracture toughness, stainless steel forging is as same as stainless steel pipes. On the fatigue crack propagation, experimental data of stainless steel forging is slower than the estimation formula in LBB evaluation guide. Taking these data of stainless steel forging account, we can treat stainless steel forging same as stainless steel pipes on LBB evaluation

Long-term fatigue control of nuclear power plant piping systems on the basis of in-service measurements

In principle, the stress and fatigue analyses during plant design are based on specked design loads which are classified into different levels (normal operation, upset conditions, etc.). Since the level B conditions, which rarely occur during actual service life, may essentially contribute to the fatigue usage factor, these design predictions are deemed to be pessimistic and conservative. A much more realistic stress and fatigue analysis can be performed on the basis of in-service measurements, which then influences the schedule and scope of the in-service inspections by non-destructive examinations (NDEs). For that purpose, the German nuclear power plants (NPPs) are equipped with in-service monitoring systems. Experience shows that the actual loading of the safety and fatigue relevant systems, for example surge lines, main coolant lines and spray lines, is more complex than assumed during design: temperature stratifications and transient loads vary in their level and combination. In some cases the design loads do not cover these events'; in other cases they are too conservative. This paper presents an approach for long-term fatigue control as used in German NPPs, illustrated by examples for surge and spray lines.

Effects of ductility on seismic response of piping systems and their implication on design and qualification

This study is concerned with the inelastic seismic response of nuclear power plant piping systems. Two systems are examined. The first one is an idealized four-equal-span pipe run and the second one consists of two configurations modified from an existing pipe run. Detailed finite element seismic time history analyses are performed using the Ansys computer program. By varying the various geometrical and physical parameters, calculations are made for a total of 76 cases. The results show that ductility generally contributes to reducing the response of piping systems. An empirical relation between the support load reduction factor and support ductility demand is given and a chart and simple procedures are suggested for the design and qualification of piping supports taking ductility into consideration.

Determination of limits for smallest detectable and largest subcritical leakage cracks in piping systems

Nuclear power plant piping systems — those still in their original as-built condition as well as upgraded designs — are subject to safety analysis. In order to limit the consequences of postulated piping failures, the basic safety concept incorporating rupture preclusion criteria is applied to specific high-energy piping systems. Leak-before-break analyses are also conducted within the framework of this concept. These analyses serve to determine the potential consequences of jet and reaction forces due to maximum subcritical leakage cracks while also establishing the minimum crack sizes that would be reliably detectable by the leakage rates resulting from these cracks. The boundary conditions for these analyses are not clearly defined. Using various examples as a basis, this paper presents and discusses how the leak-before-break concept can be applied.

Uncertainties in piping frequency analysis

Abstract A fuzzy set method for the estimation of uncertainties in computed natural frequencies of nuclear power plant piping systems is developed. Results are compared with the currently used probability-based approach of uncertainty analysis. Agreement is good. The fuzzy set method developed in this paper has applications in the seismic risk assessment of nuclear power plants.

Failure Mechanisms in Nuclear Power Plant Piping Systems

Research Papers Failure Mechanisms in Nuclear Power Plant Piping Systems S. H. Bush S. H. Bush Pacific Northwest Laboratory, Richland, WA 99352 Search for other works by this author on: This Site PubMed Google Scholar Author and Article Information S. H. Bush Pacific Northwest Laboratory, Richland, WA 99352 J. Pressure Vessel Technol. Nov 1992, 114(4): 389-395 (7 pages) https://doi.org/10.1115/1.2929244 Published Online: November 1, 1992 Article history Received: March 16, 1992 Revised: June 29, 1992 Online: June 17, 2008

Further development of the load coefficient method, lcm, for rational seismic design of nuclear facilities

The purpose of this paper is to present the results of a study conducted to compare the results of the Load Coefficient Method, LCM, proposed for seismic load determination, to modal analysis and the equivalent static load methods as defined in Section 3.7.2 of the U.S. Nuclear Regulatory Commission Standard Review Plan. The comparison is conducted using a number of nuclear power plant piping systems which used response spectra modal analysis input in their original design. The real piping systems studied are considered to be representative of ASME Section III nuclear Class 2 and 3 piping systems required to be designed to resist currently defined seismic loadings. Section 2 of this paper provides numerical comparisons of the application of LCM, Response Spectrum and Equivalent Static Load Methods.

A Frequency Approach to Mathematical Modeling of a Nuclear Power Plant Piping System

A methodology is presented for identification of a nuclear power plant piping system, which employs mathematical description in the form of transfer function matrix, frequency domain technique for estimation of system dynamic parameters, statistical technique for verification of model configuration and evaluation of parameter estimates, adaptive approach for current model updating. Model applications for estimation and monitoring of forcing functions, displacements, and stresses due to transient processes and steady state vibrations in the piping system are proposed. Methodology is illustrated by numerical examples.

Trends in the developments of new pipe systems for power plants

The necessity to save costs for building nuclear power plants whilst maintaining the high standard of safety requirements led, in the Federal Republic of Germany, through an increase in the stability of specifications to standardisation. As a result, the costs and schedules for nuclear power plant piping systems, which represent a schedule critical and thus a cost critical part of the nuclear power plant construction, are controllable and steerable. The principle of a basic, safe design offers the nuclear piping industry also for the future the opportunity for increasing efficiency, provided that sufficient pre-planning is available to ensure an orderly execution of piping, and that the available potential of experience of the specialized companies is used.

Stochastic analysis of structures and piping systems subjected to stationary multiple support excitations

AbstractA stochastic method has been developed for seismic analysis of structures and piping systems subjected to multiple support excitations. In either the time or the frequency domain, mean and extreme values of structural and piping system response can be found, including the effects of cross‐correlations of modal response and cross‐correlations of multiple support excitations. Stationary white noise and stationary filtered white noise ground excitations are used. A computer program has been developed to carry out the stochastic seismic analysis. Results for a realistic nuclear power plant structure and piping system with and without modal cross‐correlations and support excitation cross‐correlations are compared. From these results, it is concluded that neglecting cross‐correlations can lead to large errors. The stochastic method reported is shown to be more accurate than the response spectrum method and more economical than the time‐history method; therefore, it is recommended for seismic analysis of nuclear power plants.

Background for Changes in the 1981 Edition of the ASME Nuclear Power Plant Components Code for Controlling Primary Loads in Piping Systems

Section III of the ASME Boiler and Pressure Vessel Code contains simplified design formulas for placing bounds on the plastic deformations in nuclear power plant piping systems. For Class 1 piping a simple equation is given in terms of primary load stress indices (B1 and B2) and nominal pressure and bending stresses. The B1 and B2 stress indices reflect the capacities of various piping products to carry load without gross plastic deformation. In this paper, the significance of the indices, nominal stresses, and limits given in the Code for Class 1 piping and corresponding requirements for Class 2 and Class 3 piping are discussed. Motivation behind recent (1978–1981) changes in the indices and in the associated stress limits is presented.

Unstable fracture from stable fibrous crack growth -3rd report

Stress corrosion cracks are often found in chemical or nuclear power plant piping systems. One of the concepts that assess the safety of these piping systems is the Leak Before Break (LBB) concept. Namely, even if ductile mode crack initiates and grows, the plant is thought to be safe, when the growth of the crack is stable. The purpose of this paper is to establish the LBB assessment method for piping systems with cracks.A criterion is proposed which estimates the tearing instability at the onset of penetration from a surface cracked plate in tension. The criterion is based on Crack Tip Opening Angle (CTOA)-R curve concept.Two kinds of experiments were carried out. By stable carck growth tests, the specimen geometry dependence of J-and CTOA-R curve was investigated. J-R curve was found to be dependent on specimen geometry. On the other hand, CTOA-R curve was relatively insensitive to specimen geometry, in addition, it could be estimated from surface notched specimens.Unstable fracture tests were carried out in order to investigate the validity of the criterion proposed here. Two kinds of fractures were observed from the tests using surface notched plates with highly compliant loading system, i. e., instability at the onset of penetration and instability after stable crack growth. And it was found that the proposed criterion could predict whether the instability occurred at the onset of penetration or not.

Comparison of multiple support excitation solution techniques for piping systems

Design and analysis of nuclear power plant piping systems exposed to a variety of dynamic loads often require multiple support excitation analysis by modal or direct time integration methods. Both methods have recently been implemented in the computer program KWUROHR for static and dynamic analysis of piping systems, following the previous implementation of the multiple support excitation response spectrum method (see papers K6/15 and K6/15a of the SMiRT-4 Conference). The results of multiple support excitation response spectrum analyses can be examined by carrying out the equivalent time history analyses which do not distort the time phase relationship between the excitations at different support points. A frequent point of discussion is multiple versus single support excitation. A single support excitation analysis is computationally straightforward and tends to be on the conservative side, as the numerical results show. A multiple support excitation analysis, however, does not incur much more additional computer cost than the expenditure for an initial static solution involving three times the number, L, of excitation levels, i.e. 3 L static load cases. The results are more realistic than those from a single support excitation analysis. A number of typical nuclear plant piping systems have been analyzed using single and multiple support excitation algorithms for: (1) the response spectrum method, (2) the modal time history method via the Wilson, Newmark and Goldberg integration operators and (3) the direct time history method via the Wilson integration operator. Characteristic results are presented to compare the computational quality of all three methods.

Unstable fracture from stable fibrous crack growth-2nd report

Stress corrosion cracks are often found in chemical or nuclear power plant piping systems. One of the concepts that assess the safety of these piping systems is the Leak Before Break (LBB) concept. Namely, even if ductile mode crack initiates and grows, the plant is thought to be safe, when the growth of the crack is stable. The purpose of this paper is to establish the LBB assessment method for piping systems with cracks.A criterion is proposed which estimates the tearing instability at the onset of penetration from a surface cracked plate in tension. The criterion is based on Crack Tip Opening Angle (CTOA)-R curve concept.Two kinds of experiments were carried out. By stable carck growth tests, the specimen geometry dependence of J-and CTOA-R curve was investigated. J-R curve was found to be dependent on specimen geometry. On the other hand, CTOA-R curve was relatively insensitive to specimen geometry, in addition, it could be estimated from surface notched specimens.Unstable fracture tests were carried out in order to investigate the validity of the criterion proposed here. Two kinds of fractures were observed from the tests using surface notched plates with highly compliant loading system, i. e., instability at the onset of penetration and instability after stable crack growth. And it was found that the proposed criterion could predict whether the instability occurred at the onset of penetration or not.