Influences of the dynamic strain aging on the J–R fracture characteristics of the ferritic steels for reactor coolant piping system

The leak-before-break (LBB) design of the piping system for nuclear power plants has been based on the premise that the leakage due to the through-wall crack can be detected by using leak detection systems before a catastrophic break. The piping materials are required to have excellent J–R fracture characteristics. However, where ferritic steels for reactor coolant piping systems operate at the temperatures where dynamic strain aging (DSA) could occur, the fracture resistance could be reduced with the influence of DSA under dynamic loading. Therefore, in order to apply the LBB design concept to the piping system under seismic loading, both static and dynamic J–R characteristics must be evaluated.Materials used in this study are SA516 Gr.70 for the elbow pipe and SA508 Cl.1a for the main pipe and their welding joints. The crack extension during the dynamic and the static J–R tests was measured by the direct current potential drop (DCPD) and the compliance method, respectively. This paper describes the influences of the dynamic strain aging on the J–R fracture characteristics with the loading rate of the pipe materials and their welding joints.     

Piping research overview

The USNRC Piping Review Committee (PRC) was formed in 1983 with a charter to review NRC piping criteria, to recommend changes to this criteria, and to identify areas that would benefit from future research. This overview will outline the NRC-sponsored research being conducted to address those PRC recommendations concerning the design of nuclear piping systems to withstand dynamic loads. A key element of this research is the joint EPRI/NRC “Piping and Fitting Reliability Research Program.” This program consists of dynamic capacity testing of piping at the system, component, and specimen levels, plus analyses needed to support recommendations for changes to the ASME Code. As part of NRC's contribution to the EPRI/NRC program, a pipe system capacity test will be conducted at ETEC. The “Nonlinear Piping Response Prediction” project at HEDL is evaluating nonlinear response prediction techniques with differing degrees of complexity and will compare the various analytical results both with each other and with physical benchmarks such as the ETEC test. An ORNL project is developing nozzle design guidance that will provide a more realistic basis for evaluating the higher nozzle loads that will result from the more flexible piping systems design that are being considered. INEL will evaluate high frequency damping by considering the existing high frequency data and by conducting high frequency/high stress tests on two piping systems. LLNL is now conducting studies to more completely assess the uncertainties in the seismic response of building structures and piping systems. As a follow-on to the research efforts reported in NUREG/CR-3811, BNL will conduct additional studies to improve combinational procedures for piping response spectra analyses.     

Observations on the effect of post-weld heat treatment on J-resistance curves of sa-106b seamless piping welds

Ontario Hydro has developed a leak-before-break (LBB) methodology for application to large diameter piping (21, 22 and 24 inch) Schedule 100 SA106B heat transport (HT) piping as a design alternative to pipe whip restraints and in recognition of the questionable benefits of providing such devices. Ontario Hydro's LBB approach uses elastic-plastic fracture mechanics (EPFM).In order to assess the stability of HT piping in the presence of hypothetical flaws, the value of the material J-integral associated with crack extension (J— R curve) must be known. In a material test program J-resistance curves were determined from various pipe heats and four different welding procedures that were developed by Ontario Hydro for nuclear Class 1 piping. The test program was designed to investigate and quantify the effect of various factors such as test temperature, crack plane orientation and welding effects which have an influence on fracture properties. An acceptable lower bound J-resistance curve for the piping steels and welds were obtained by machining maximum thickness specimens from the pipes and weldments and by testing side-grooved compact tension specimens. This paper addresses the effect of test temperature and post-weld heat treatment on the J-resistance curves from the welds.The fracture toughness of all the welds at 250°C was lower than that at 20°C. Welds that were post-weld heat treated showed high crack initiation toughness, J_lc, rising J-resistance curves and stable and ductible crack extension. Non post-weld heat treated welds, while remaining tough and ductile, showed comparatively lower J_Ic, and J-resistance curves at 250°C. This drop in toughness is possibly due to a dynamic strain aging mechanism evidenced by serrated load-displacement curves. The fracture toughness of non post-weld heat treated welds increased significantly after a comparable post-weld heat treatment.The test procedure was validated by comparing three test results against independent tests conducted by Materials Engineering Associates (MEA) of Lanham, Maryland. The J_Ic and J-resistance curves obtained by Ontario Hydro and MEA were comparable.     

The steamhammer problem: dynamic shock loading of critical reactor piping systems

In the course of both pre-operational testing and power operation of commercial nuclear power plants, relatively large amplitude transient vibrations of steam piping systems have been experienced with damage to the piping supports in at least one recent case. These transient vibrations result from ‘steamhammer’ or dynamic shock loading induced by pressure and momentum transient conditions generated in the piping by sudden changes to the flow conditions, such as are produced by sudden valve opening or closure. In particular, vibrations have been experienced in by-pass and discharge lines as a result of relief valve discharge, and in main steam lines as a result of sudden main stop valve closure. Piping in both BWR and PWR reactor systems has been found to be susceptible to these conditions.This paper is concerned with the evaluation of the pressure and momentum transients resulting from sudden valve operation, and the determination of the dynamic response of the piping to the induced transient loading. The characteristics of the transient conditions existing immediately following both sudden valve opening and closure as encountered in BWR and PWR plants are discussed. The procedures used to calculate the transient time history functions are outlined. The derivation of the loading induced in the piping by the pressure and momentum transients is discussed and the determination of the dynamic response of the piping is presented. The procedures described in the paper are illustrated by actual examples from BWR and PWR plants, and the significance of steamhammer effects relative to other loading conditions is discussed.     

Development of a database of pipe fracture experiments

Over the last 35 years, researchers worldwide have conducted hundreds—if not thousands—of pipe fracture experiments. In the early years, researchers focused their attention on studying the failure pressure and crack propagation behavior of axially cracked pipe loaded by internal pressure. The earliest work was sponsored by the oil and gas industry and, as such, involved relatively thin-walled, low toughness carbon steel pipes. This work was eventually followed up by efforts in the USA and Germany on nuclear piping with axial cracks. In recent years, attention has turned to understanding the behavior of circumferentially cracked nuclear piping subjected to both pressure and bending loads. The loading histories for these experiments range from the relatively simple case of quasi-static, monotonic displacement control to the more complex cases of dynamic cyclic loading, and pipe system experiments. In this paper, two of the leaders in this research, i.e. Battelle in the USA and MPA Stuttgart in Germany, have collaborated to develop a database of pipe fracture experiments. The database includes data from other organizations as well as the data from Battelle and MPA. In addition, as part of this paper, an example of how the database was used to assess the failure pressure of axially cracked pipe is given.     

Status of boiling water reactor structural integrity programs at the General Electric Company

The General Electric Company is actively involved in a variety of programs whose objective is to evaluate and improve the technology related to structural integrity of Boiling Water Reactors. This work covers many technical disciplines that fall broadly under the title of Applied Mechanics and includes the technologies of solid mechanics, structural dynamics, elasticity, plasticity, and fluid dynamics.General Electric programs are directed at qualification of the predictive methodology, design bases and criteria used to evaluate and confirm the structural design margins of the Boiling Water Reactor Nuclear Steam Supply System. This includes the reactor pressure vessel, piping, fuel, core structures, and auxiliary equipment within General Electric's scope of supply. Focus of these programs includes identification, measurement and refinement of operational and postulated accident-type loads; prediction of structure and component response; study of damage mechanisms; and development of failure criteria, evaluation methods and design rules.In this report, some specific activities and accomplishments during the last few years are summarized in order to provide a broad overview of the wide range of technical issues and programs undertaken. These activities have been grouped into areas of: I — Dynamic Modeling and Structural Analysis; II — Fatigue and Fracture Evaluations; III — Structural Capability Tests; and IV — Flow Induced Vibration Experiments.     

PWR safety/relief valve blowdown analysis experience

The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively.General approaches for generating forcing functions from thermalfluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the “acceleration or wave force” method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawback are discussed.Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.     

Standardization of piping supports in nuclear power plants using pretested components

The article reports on the “standard piping-support catalog”, which was used in connection with the Convoy-nuclear-power-plants in Fed. Rep. Germany, under special consideration of: (i) cost reduction for planning and installation, (ii) suitability tests (type testing for standardized supports), and (iii) application of CAD. The advantages using the piping support catalog are summarised.     

Low cycle fatigue and ductile fracture for Japanese carbon steel piping under dynamic loadings

Dynamic fracture behavior of circumferentially cracked pipe is important to evaluate the structural integrity of nuclear piping from the viewpoint of the LBB concept under seismic conditions. Fracture tests have been conducted for Japanese carbon steel (STS410) circumferentially through-wall cracked pipes that are subjected to monotonic or cyclic bending loads at room temperature. In the monotonic-loading tests, the maximum load to failure increases slightly with increasing loading rate. The failure cycles can be expressed simply by ratio of the load amplitude to the plastic collapse load. Fracture analysis has been also conducted to model the pipe tests. A new equation for calculating ΔJ for a circumferentially through-wall cracked pipe subjected to bending has been proposed. The failure cycles under cyclic loads are satisfactorily evaluated using an elastic-plastic fracture mechanics parameter ΔJ.     

Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems

The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the “force equivalent area” (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques.     

Droplets deposition in steam piping connecting steam generator and steam turbine in nuclear plant

A numerical investigation is carried out for turbulent droplet-laden flow of saturated steam produced in a steam generator (SG) that feeds steam turbine (ST) through a long and multi-bend steam piping. The main purpose of the study is to analyze deposition of droplets that form a wall film in the piping system. Commercial CFD code StarCD is used for the solution of turbulent flow field of droplet-laden steam. Turbulence is treated using k–ω model of turbulence. Wall film formation is solved by additional conservation equations. Two tasks were performed: parametric study of the deposition in a 90° elbow positioned with different orientation and the deposition in a more complex piping system. This system starts with outlets from steam generator with five mouthpieces leading to a collector pipe and connecting the steam piping leading to a steam turbine. The steam piping consists of three straight segments of pipes and two 90° elbows in the total length 17 m. The diameter of the steam piping is 0.425 m. Results of the simulations show where droplets deposit and where a liquid separator should be placed to drain away the water film and to avoid droplets from entering the steam turbine.     

Hybrid fibre-reinforced concrete under repeated loading

Fibre-reinforced concrete containing several volume fractions in different ratios of two types of fibre—polypropylene and steel—were tested under repeated loading. The mechanical properties of speciments—cubes 150 mm × 150 mm × 150 mm (for compressive strength), prisms 100 mm × 100 × 400 mm (for flexural strength) and short cylinders 150 mm long and 60 mm in diameter (for impact strength)—have been experimentally investigated before and after cyclic loading after a curing of 28 days.Mix proportions were design according to DIN 1045 with maximum aggregate size 8 mm and grading curve B8. Portland cement PC 40 in the amount of 450 kg m⁻³ was applied and the water-to-cement ratio was 0.55. The workability of mixes was measured by the Vebe method and regulated by the plasticizing admixture Ligoplast Na. The maximum hybrid fibre volume fraction (polypropylene + steel) was 1.0%. The dynamic forces generated in a Schenck testing machine with a frequency of 16 Hz had a sinusoidal waveform varying between 0.7 and 0.1 of the static mechanical characteristics. The number of cycles in all tests was 10⁵. The residual MOR in the static four-point bending test and working force-deflection diagram were also obtained. The impact properties after repeated loading in compression were tested by means of the falling-weight test. Relationships between the composition of fibre composites with different amounts of polypylene (0.2, 0.3 and 0.5 vol.%) and steel fibre content (0.5, 0.7 and 0.8 vol.%) were obtained and the technological properties of the mixes as well.     

Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A “analysis of flaw indication” for the application to a PWR primary piping. Results of the analysis are discussed in detail.     

Determination of cycle-dependent material properties for a design concept on the basis of limited plastic strains

For unlikely dynamic service conditions, such as pressure surges or earthquakes, analysis of piping integrity in terms of strains is appropriate. Strain controlled tensile/compressive fatigue tests are meant to indicate whether the limited plastic strains caused during the few load cycles occurring as a result of these service conditions shall be admissible. The fatigue behaviour is influenced by parameters, such as frequency, resp. strain rate, temperature and material, resp. material condition and should be taken into these considerations. With the help of the experimentally determined cycles to fracture or crack initiation, it is possible to extend design curves in the technical code in the area of low cycles to fracture, e.g. crack initiation (N_I < 100).     

Dynamic loading of containment during blowdown: review of experimental data from marviken and brunsbüttel

This paper deals with the problem of the dynamic loading of the pressure suppression type containment due to pressure fluctuations which are induced during blowdown as a result of steam condensation in the suppression pool. It reviews the experimental data on pressure fluctuations and on the corresponding dynamic response of the structures submerged in the water charge. Two series of containment response experiments were performed on real pressure suppression systems. One series of measurements was performed within the framework of relief valve vent tests in the Brunsbüttel Power Plant (Federal Republic of Germany) in 1974. Another series of measurements was carried out within the framework of the blowdown tests performed at the Marviken Power Plant (Sweden) in 1972–1973.For the measurements of pressure fluctuations in the suppression pool strain gage-based pressure transducers were used. The dynamic response of the loaded structure was measured by strain gages or by piezoresistive (in some cases piezoelectric) accelerometers. The signal was amplified by carrier-frequency amplifiers (or loading amplifiers). The amplified signals were recorded on magnetic tape (FM recording). A two-step evaluation procedure was used. In the first step the time history plots of the variables directly measured (such as pressure, bending strain and acceleration) were plotted and interpreted. In the second evaluation step the cross-power spectral density function of different signal pairs was calculated, plotted and interpreted. For evaluation of the fluctuating pressure field in the suppression pool the combination was chosen of a reference pressure signal (measured at a fixed position called ‘reference position’) with a second pressure signal (measured at a different, arbitrary position) in most cases to perform the cross-spectra analysis. The set of cross-spectral density functions obtained by analyzing all signal combinations chosen was used to determine the power spectra, RMS amplitudes and phases of the corresponding pressure waves impinging on the structure submerged in the suppression pool. For the evaluation of pressure field—structural response relations one pressure signal was combined with one response signal (bending strain or acceleration) to give the input for cross-spectral analysis. By the combination of two response signals (bending strain—bending strain or acceleration—acceleration) for cross-spectral analysis the modal shapes of the responding structure was determined.     

New leak detection technique using ceramic humidity sensor for water reactors

Leak-before-break (LBB) approach has been considered for its application to the main steam line (MSL) of Korean Next Generation Reactor (KNGR) — an advanced pressurized water reactor under development. Unlike the primary water leakage, the MSL leak detection must be based on principles other than radioactivity measurements. Among potential options that are being considered as indicators of leakage, it is believed that humidity at the proximity of the piping system is an effective one. A ceramic-based humidity sensor was developed, which can be qualified for LBB applications. The ceramic material, sintered and annealed MgCr₂O₄–TiO₂, is shown to increase its electrical conductivity upon water vapor adsorption without any negative impact due to gamma radiation over the entire temperature range of interest. In the plant applications, the sensor array can be positioned in the annulus between the piping and surrounding insulation. By the analysis of humidity distribution in the annulus, a leak rate of 1 l h⁻¹ can be detected within an hour when the distance between two adjacent sensors does not exceed 1 m. In order to minimize the number of signal wires, the use of AC impedance technique is shown to be advantageous. In this paper, the results of the development and the performance characterization of ceramic humidity sensor for the LBB application to the MSL of KNGR are summarized.     

Fatigue studies on carbon steel piping materials and components: Indian PHWRs

This paper describes the results of fatigue studies on carbon steel piping materials and components of Indian Pressurized Heavy Water Reactors (PHWRs). The piping components include pipes and elbows, of outer diameter 219 mm, 324 mm and 406 mm, made of carbon steel (SA333 Gr.6 grade) material. Tests on actual pipes and elbows with part through notch were carried out to study the behaviour of crack growth under cyclic loading for different pipe sizes, notch aspect ratios, stress ratios, etc. During the tests, numbers of cycles for crack initiation from blunt notch were recorded with an accuracy of 0.1 mm. In conjunction with component tests, the experimental studies were also conducted on standard specimens to understand the effect of different variables such as size (thickness), type of specimen and components (elbow and pipe), welding, stress ratio, notch orientation on fatigue crack growth rate. The fatigue crack growth curve (da/dN versus ΔK) obtained from three-point bend specimen and pipe was compared with that given in ASME Section XI. The comparison shows that da/dN versus ΔK curves obtained from the specimen and pipe tests are nearly same. The analytical predictions for crack initiation and crack growth for the tested components were compared with experimental results. Such comparisons validate the modeling procedure for crack initiation and growth.     

Cyclic crack growth evaluation of a 20MnMoNi55 piping steel in high-oxygen reactor water

Samples of a low alloy steel piping material taken from the full scale corrosion fatigue test loop of the Heissdampfreaktor (HDR) plant have been tested at 240°C in high oxygen reactor water. The small-scale specimens (CT25) were exposed to a similar loading spectrum to that which has been used in the full-scale corrosion fatigue tests at the HDR-plant. During the autoclave tests cyclic crack growth rates were determined. Fracture surface investigations were performed not only for the laboratory test specimens but also for the fracture surface of a sample taken from the HDR test loop piping after the full scale test. In this paper the autoclave testing results and fracture surface observations are presented and compared to those obtained in the HDR piping tests.     

Modelling the fluid structure interaction produced by a waterhammer during shutdown of high-pressure pumps

Measurements of an experiment in a pipe system with pump shutdown and valve closing have been performed in the nuclear power plant KRB II (Gundremmingen, Germany). Comparative calculations of fluid and structure including interaction show an excellent agreement with the measured results. Theory and implementation of the fluid structure interaction (FSI) and the results of the comparison are described. The following measurements have been compared with calculations: (1) experiments in Delft, Netherlands to analyse the FSI; and (2) experiment with pump shutdown and valve closing in the nuclear power plant KRB II has been performed. It turns out, that the consideration of the FSI is necessary for an exact calculation of ‘soft’ piping systems. It has significant application in current waterhammer problems. For example, water column closure, vapour collapse, check valve slamming continues to create waterhammers in the energy industry. An important consequence of the FSI is mostly a significant increase of the effective structural damping. This mitigates—so far in all KED’s calculations the FSI has taken into account—an amplification of pipe movements due to pressure waves in resonance with structural eigenvalues. To investigate the integrity of pipe systems pipe stresses are calculated. Taking FSI into account they are reduced by 10–40% in the actual case.     

DEMT experimental and analytical studies on seismic isolation

This paper reviews the experimental and theoretical studies performed at CEA/DEMT related to the overall behavior of isolated structures. The experimental work consists of the seismic shaking-table tests of a concrete cylinder isolated by neoprene sliding pads, and the vibrational tests on the reaction mass of the TAMARIS seismic facility. The analytical work consists of the development of procedures for dynamic calculation methods: for soil—structure interaction where pads are placed between an upper raft and pedestals, for time-history calculations where sliding plates are used, and for fluid—structure interaction where coupled fluid and structure motions and sloshing modes are important.Finally, this paper comments on the consequences of seismic isolation for the analysis of fast-breeder reactor (FBR) vessels. The modes can no longer be considered independent (SRSS Method leads to important errors), and the sloshing increases.