Crack stability analysis of low alloy steel primary coolant pipe

Energy Technology Data Exchange (ETDEWEB)

Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

1997-04-01

At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

Fatigue crack growth rate studies on pipes and pipe welds made of austenitic stainless steel and carbon steel

International Nuclear Information System (INIS)

Arora, Punit; Singh, P.K.; Bhasin, Vivek; Vaze, K.K.; Pukazhendhi, D.M.; Gandhi, P.; Raghava, G.

2011-01-01

The objective of the present study is to understand the fatigue crack growth behavior in austenitic stainless steel and carbon steel pipes and pipe welds by carrying out analysis/predictions and experiments. The Paris law has been used for the prediction of fatigue crack growth life. To carry out the analysis, Paris constants have been determined for pipe (base) and pipe weld materials by using Compact Tension (CT)/Three Point Bend (TPB) specimens machined from the actual pipe/pipe weld. Analyses have been carried out to predict the fatigue crack growth life of pipes/pipe welds having part through cracks on the outer surface. In the analyses, Stress Intensity Factors (K) have been evaluated through two different schemes. The first scheme considers the 'K' evaluations at two points of the crack front i.e. maximum crack depth and crack tip at the outer surface. The second scheme accounts for the area averaged root mean square stress intensity factor (K RMS ) at deepest and surface points. In order to validate the analytical procedure/results, experiments have been carried out on full scale pipe and pipe welds with part through circumferential crack. Fatigue crack growth life evaluated using both schemes have been compared with experimental results. Use of stress intensity factor (K RMS ) evaluated using second scheme gives better fatigue crack growth life prediction compared to that of first scheme. (author)

Research on Buckling State of Prestressed Fiber-Strengthened Steel Pipes

Science.gov (United States)

Wang, Ruheng; Lan, Kunchang

2018-01-01

The main restorative methods of damaged oil and gas pipelines include welding reinforcement, fixture reinforcement and fiber material reinforcement. Owing to the severe corrosion problems of pipes in practical use, the research on renovation and consolidation techniques of damaged pipes gains extensive attention by experts and scholars both at home and abroad. The analysis of mechanical behaviors of reinforced pressure pipelines and further studies focusing on “the critical buckling” and intensity of pressure pipeline failure are conducted in this paper, providing theoretical basis to restressed fiber-strengthened steel pipes. Deformation coordination equations and buckling control equations of steel pipes under the effect of prestress is deduced by using Rayleigh Ritz method, which is an approximation method based on potential energy stationary value theory and minimum potential energy principle. According to the deformation of prestressed steel pipes, the deflection differential equation of prestressed steel pipes is established, and the critical value of buckling under prestress is obtained.

Fracture properties evaluation of stainless steel piping for LBB applications

International Nuclear Information System (INIS)

Kim, Y.J.; Seok, C.S.; Chang, Y.S.

1997-01-01

The objective of this paper is to evaluate the material properties of SA312 TP316 and SA312 TP304 stainless steels and their associated welds manufactured for shutdown cooling line and safety injection line of nuclear generating stations. A total of 82 tensile tests and 58 fracture toughness tests on specimens taken from actual pipes were performed and the effect of various parameters such as the pipe size, the specimen orientation, the test temperature and the welding procedure on the material properties are discussed. Test results show that the effect of the test temperature on the fracture toughness was significant while the effects of the pipe size and the specimen orientation on the fracture toughness were negligible. The material properties of the GTAW weld metal was in general higher than those of the base metal

Fracture properties evaluation of stainless steel piping for LBB applications

Energy Technology Data Exchange (ETDEWEB)

Kim, Y.J.; Seok, C.S.; Chang, Y.S. [Sung Kyun Kwan Univ., Suwon (Korea, Republic of)

1997-04-01

The objective of this paper is to evaluate the material properties of SA312 TP316 and SA312 TP304 stainless steels and their associated welds manufactured for shutdown cooling line and safety injection line of nuclear generating stations. A total of 82 tensile tests and 58 fracture toughness tests on specimens taken from actual pipes were performed and the effect of various parameters such as the pipe size, the specimen orientation, the test temperature and the welding procedure on the material properties are discussed. Test results show that the effect of the test temperature on the fracture toughness was significant while the effects of the pipe size and the specimen orientation on the fracture toughness were negligible. The material properties of the GTAW weld metal was in general higher than those of the base metal.

A parametric study of residual stresses in multipass butt-welded stainless steel pipes

Energy Technology Data Exchange (ETDEWEB)

Brickstad, B. [SAQ Inspection Ltd., Stockholm (Sweden); Josefson, L. [Chalmers Univ. of Technology, Goeteborg (Sweden). Div. of Solid Mechanics

1996-06-01

Multipass circumferential butt-welding of stainless steel pipes is simulated numerically in a non-linear thermo-mechanical FE-analysis. In particular, the through-thickness variation at the weld and heat affected zone, of the axial and hoop stresses and their sensitivity to variation in weld parameters are studied. Recommendations are given for the through thickness variation of the axial and hoop stresses to be used when assessing the growth of surface flaws at circumferential butt welds in nuclear piping system. 31 refs, 12 tabs, 54 figs.

The behavior of welded joint in steel pipe members under monotonic and cyclic loading

International Nuclear Information System (INIS)

Chang, Kyong-Ho; Jang, Gab-Chul; Shin, Young-Eui; Han, Jung-Guen; Kim, Jong-Min

2006-01-01

Most steel pipe members are joined by welding. The residual stress and weld metal in a welded joint have the influence on the behavior of steel pipes. Therefore, to accurately predict the behavior of steel pipes with a welded joint, the influence of welding residual stress and weld metal on the behavior of steel pipe must be investigated. In this paper, the residual stress of steel pipes with a welded joint was investigated by using a three-dimensional non-steady heat conduction analysis and a three-dimensional thermal elastic-plastic analysis. Based on the results of monotonic and cyclic loading tests, a hysteresis model for weld metal was formulated. The hysteresis model was proposed by the authors and applied to a three-dimensional finite elements analysis. To investigate the influence of a welded joint in steel pipes under monotonic and cyclic loading, three-dimensional finite elements analysis considering the proposed model and residual stress was carried out. The influence of a welded joint on the behavior of steel pipe members was investigated by comparing the analytical result both steel pipe with a welded joint and that without a welded joint

78 FR 45271 - Welded Stainless Steel Pressure Pipe From Malaysia, Thailand, and Vietnam

Science.gov (United States)

2013-07-26

... Stainless Steel Pressure Pipe From Malaysia, Thailand, and Vietnam Determination On the basis of the record... reason of imports from Malaysia, Thailand, and Vietnam of welded stainless steel pressure pipe, provided... contained in USITC Publication 4413 (July 2013), entitled Welded Stainless Steel Pressure Pipe from Malaysia...

The PEACE PIPE: Recycling nuclear weapons into a TRU storage/shipping container

International Nuclear Information System (INIS)

Floyd, D.; Edstrom, C.; Biddle, K.; Orlowski, R.; Geinitz, R.; Keenan, K.; Rivera, M.

1997-01-01

This paper describes results of a contract undertaken by the National Conversion Pilot Project (NCPP) at the Rocky Flats Environmental Technology Site (RFETS) to fabricate stainless steel ''pipe'' containers for use in certification testing at Sandia National Lab, Albuquerque to qualify the container for both storage of transuranic (TRU) waste at RFETS and other DOE sites and shipping of the waste to the Waste Isolation Pilot Project (WIPP). The paper includes a description of the nearly ten-fold increase in the amount of contained plutonium enabled by the product design, the preparation and use of former nuclear weapons facilities to fabricate the components, and the rigorous quality assurance and test procedures that were employed. It also describes how stainless steel nuclear weapons components can be converted into these pipe containers, a true ''swords into plowshare'' success story

The effect of cyclic and dynamic loads on carbon steel pipe

International Nuclear Information System (INIS)

Rudland, D.L.; Scott, P.M.; Wilkowski, G.M.

1996-02-01

This report presents the results of four 152-mm (6-inch) diameter, unpressurized, circumferential through-wall-cracked, dynamic pipe experiments fabricated from STS410 carbon steel pipe manufactured in Japan. For three of these experiments, the through-wall crack was in the base metal. The displacement histories applied to these experiments were a quasi-static monotonic, dynamic monotonic, and dynamic, cyclic (R = -1) history. The through-wall crack for the third experiment was in a tungsten-inert-gas weld, fabricated in Japan, joining two lengths of STS410 pipe. The displacement history for this experiment was the same history applied to the dynamic, cyclic base metal experiment. The test temperature for each experiment was 300 C (572 F). The objective of these experiments was to compare a Japanese carbon steel pipe material with US pipe material, to ascertain whether this Japanese steel was as sensitive to dynamic and cyclic effects as US carbon steel pipe. In support of these pipe experiments, quasi-static and dynamic, tensile and fracture toughness tests were conducted. An analysis effort was performed that involved comparing experimental crack initiation and maximum moments with predictions based on available fracture prediction models, and calculating J-R curves for the pipe experiments using the η-factor method

Investigation of the development and optimisation of cutting loads for the cutting of steel pipes with typical properties and material properties for nuclear power stations

International Nuclear Information System (INIS)

Schumann, S.; Freund, H.U.; Hollenberg, K.; Horning, W.; Esser, H.J.

1987-04-01

The aim of the project was to develop a type of cutter loading for the cutting of thickwalled steel pipes by explosive technique which, due to its construction and cutting performance, is suitable for use when dismantling pipelines in shutdown nuclear power stations. The loading sleeve is built up of individual linear elements and can be placed as a polygon (e.g. octagon) around pipes of different diameters. A steel pipe with dimensions 610 mm diameter x 36 mm wall thickness (live steam pipe of a German BWR of a new type) was completely and accurately cut using a cutting load sleeve with 1.84 kg of explosive. The great tamping of the cutting loader type developed, minimises the quantity of explosive required and reduces the air shock or blast wave peak pressure to about 30% compared to a charge without tamping. The distance at which the value of peak pressure of the blast wave of 1 bar (which could cause damage) is exceeded, is reduced to 3.0 metres compared to 5.3 metres for an untamped charge of the same cutting power. (orig./HP) [de

Experiment on electrolysis decontamination of stainless steel pipes

International Nuclear Information System (INIS)

Wang Dongwen; Dou Tianjun; Zhao Yujie

2004-01-01

A new electrolytic decontamination method used metal balls as conducting anode was investigated. The influences of current density, solution property and diameter of pipes on efficiency of electrolytic decontamination were examined and the efficiency of this method was compared with that of common electrolytic method under the same experimental conditions. Decontamination of samples of stainless steel pipes contaminated by plutonium was performed. Experimental results indicate that decontamination of stainless steel pipes contaminated by plutonium can be achieved at the optimum conditions of greater than 0.2 A·cm -2 current density, 5% sulfuric acid electrolyte and 5 min electrolysis. This method can be used in the decontamination of a wide variety of decommissioned metal materials. (author)

High nitrogen stainless steels for nuclear industry

International Nuclear Information System (INIS)

Kamachi Mudali, U.

2016-01-01

Nitrogen alloying in stainless steels (SS) has myriad beneficial effects, including solid solution strengthening, precipitation effects, phase control and corrosion resistance. Recent years have seen a rapid development of these alloys with improved properties owing to advances in processing technologies. Furthermore, unlimited demands for high-performance advanced steels for special use in advanced applications renewed the interest in high nitrogen steels (HNS). The combination of numbers of attractive properties such as strength, fracture toughness, wear resistance, workability, magnetic properties and corrosion resistance of HNS has given a unique advantage and offers a number of prospective applications in different industries. Based on extensive studies carried out at IGCAR, nitrogen alloyed type 304LN SS and 316LN SS have been chosen as materials of construction for many engineering components of fast breeder reactor (FBR) and associated reprocessing plants. HNS austenitic SS alloys are used as structural/reactor components, i.e., main vessel, inner vessel, control plug, intermediate heat exchanger and main sodium piping for fast breeder reactor. HNS type 304LN SS is a candidate material for continuous dissolver, nuclear waste storage tanks, pipings, etc. for nitric acid service under highly corrosive conditions. Recent developments towards the manufacturing and properties of HNS alloys for application in nuclear industry are highlighted in the presentation. (author)

Ductile fracture behavior of 6-inch diameter type 304 stainless steel and STS 42 carbon steel piping containing a through-wall or part-through crack

International Nuclear Information System (INIS)

Shibata, Katsuyuki; Ohba, Toshihiro; Kawamura, Takaichi; Miyazono, Shohachiro; Kaneko, Tadashi; Yokoyama, Norio.

1986-05-01

The double ended guillotine break philosophy in the design base accident of the nuclear power plant is considered to be overly conservative from the view point of piping design. Through the past experiences and developments of the fabrication, inspection, and operation of nuclear power plants, it has been recognized that the Leak-Before-Break (LBB) concept can be justified in the LWR pressure boundary pipings. In order to verify the LBB concept, extensive experimental and theoretical works are being conducted in many countries. Furthermore, a revised piping design standard, in which LBB concept is introduced, is under preparation in Japan, U.S.A., and European countries. At JAERI, a research program to investigate the unstable ductile fracture behavior of LWR piping under bending load has been carried out as a part of the LBB verification researches since 1983. This report summarizes the result of the ductile fracture tests conducted at room temperature in 1983 and 84. The 6-inch diameter pipes of type 304 stainless steel and STS 42 carbon steel pipe with a through-wall or part-through crack were tested under bending load with low or high compliance condition at room temperature. Pipe fracture data were obtained from the test as regards to load- displacement curve, crack extension, net section stress, J-resistance curve, and so on. Besides, the influence of the compliance on the fracture behavior was examined. Discussions are performed on the ductile pipe fracture criterion, flaw evaluation criterion, and LBB evaluation method. (author)

Prospects of weldable steels for nuclear power engineering

International Nuclear Information System (INIS)

Pilous, V.

1985-01-01

In nuclear power plants with WWER reactors a medium-alloyed CrNiMoV steel is considered for the pressure vessel and a MnNiMoV steel for the primary pipes, the pressurizer and other systems. The chemical composition of both steels is given and briefly discussed are the results of tests carried out within a study of the weldability of the steels. Attention is also devoted to the causes of cracks under austenite-based overlays occurring when medium-alloyed CrNiMoV steels are overlaid with strip electrodes using high thermal input submerged arc welding, and in the process of heat treatment. It appears that austenitic overlays reduce the life span by 5 to 15% as compared with the basic steel. If, however, the overlay is not part of the cross section critical with regard to strength, the reduced life span need not be considered and both types of steel will be suitable for primary circuits of nuclear power plants because they guarantee the required mechanical and physical properties of the welded joints. (Z.M.)

Reliability analysis of stainless steel piping using a single stress corrosion cracking damage parameter

International Nuclear Information System (INIS)

Guedri, A.

2013-01-01

This article presents the results of an investigation that combines standard methods of fracture mechanics, empirical correlations of stress-corrosion cracking, and probabilistic methods to provide an assessment of Intergranular Stress Corrosion Cracking (IGSCC) of stainless steel piping. This is done by simulating the cracking of stainless steel piping under IGSCC conditions using the general methodology recommended in the modified computer program Piping Reliability Analysis Including Seismic Events, and by characterizing IGSCC using a single damage parameter. Good correlation between the pipe end-life probability of leak and the damage values were found. These correlations were later used to generalize this probabilistic fracture model. Also, the probability of detection curves and the benefits of in-service inspection in order to reduce the probability of leak for nuclear piping systems subjected to IGSCC were discussed for several pipe sizes. It was found that greater benefits could be gained from inspections for the large pipe as compared to the small pipe sizes. Also, the results indicate that the use of a better inspection procedure can be more effective than a tenfold increase in the number of inspections of inferior quality. -- Highlights: • We simulate the pipe probability of failure under different level of SCC damages. • The residual stresses are adjusted to calibrate the model. • Good correlations between 40-year cumulative leak probabilities and D σ are found. • These correlations were used to generalize this probabilistic fracture model. • We assess the effect of inspection procedures and scenarios on leak probabilities

Modeling of residual stress mitigation in austenitic stainless steel pipe girth weldment

International Nuclear Information System (INIS)

Li, M.; Atteridge, D.G.; Anderson, W.E.; West, S.L.

1994-01-01

This study provides numerical procedures to model 40-cm-diameter, schedule 40, Type 304L stainless steel pipe girth welding and a newly proposed post-weld treatment. The treatment can be used to accomplish the goal of imparting compressive residual stresses at the inner surface of a pipe girth weldment to prevent/retard the intergranular stress corrosion cracking (IGSCC) of the piping system in nuclear reactors. This new post-weld treatment for mitigating residual stresses is cooling stress improvement (CSI). The concept of CSI is to establish and maintain a certain temperature gradient across the pipe wall thickness to change the final stress state. Thus, this process involves sub-zero low temperature cooling of the inner pipe surface of a completed girth weldment, while simultaneously keeping the outer pipe surface at a slightly elevated temperature with the help of a certain heating method. Analyses to obtain quantitative results on pipe girth welding and CSI by using a thermo-elastic-plastic finite element model are described in this paper. Results demonstrate the potential effectiveness of CSI for introducing compressive residual stresses to prevent/retard IGSCC. Because of the symmetric nature of CSI, it shows great potential for industrial application

Development of Structural Health Monitoring System for pipes in Nuclear Power Plants

International Nuclear Information System (INIS)

Eom, H. S.; Choi, Y. C.; Shin, S. H.; Youn, D. B.; Park, J. H.

2010-01-01

Structural health monitoring (SHM) has becoming an important issue in the maintenance of various structures such as large steel plates, vessels, and pipes in nuclear power plants. There are important factors to be considered in developing an SHM system. With consideration of these factors, we have developed a computerized multi-channel ultrasonic system that can handle array transducers and generate a high-power pulse for online SHM of the plates and pipes. The proposed system is compact but has all the necessary functions for SHM of important structure such as pipes and plates in a NPP

Development and testing of restraints for nuclear piping systems

International Nuclear Information System (INIS)

Kelly, J.M.; Skinner, M.S.

1980-06-01

As an alternative to current practice of pipe restraint within nuclear power plants it has been proposed to adopt restraints capable of dissipating energy in the piping system. The specific mode of energy dissipation focused upon in these studies is the plastic yielding of steels utilizing relative movement between the pipe and the base of the restraint, a general mechanism which has been proven as reliable in several allied studies. This report discusses the testing of examples of two energy-absorbing devices, the results of this testing and the conclusions drawn. This study concentrated on the specific relevant performance characteristics of hysteretic behavior and degradation with use. The testing consisted of repetitive continuous loadings well into the plastic ranges of the devices in a sinusoidal or random displacement controlled mode

Examination of the X-ray piping diagnostic system using EGS4 (measuring the thickness of a steel pipe with rust)

International Nuclear Information System (INIS)

Kajiwara, G.

2001-01-01

In a series of papers entitled 'Examination of the X-ray piping diagnostic system using EGS4' presented the proceedings of the EGS4 users' meetings, I discussed the possibility of measuring the thickness of piping walls with rust. In the present paper, I describe, based on our earlier results, how the thickness of steel pipes with rust can be measured. I conducted EGS4 simulation to measure the thickness of a combination of steel and rust and made an energy absorption diagram for this combination. The equivalent thickness of steel was obtained through experiments and the system operation. The thickness of the steel determined by using the diagram agreed well with the actual steel thickness obtained by the experiments. In the future, we will focus on how to automate this measurement procedure and how to use the same procedure to measure the thickness of pipes filled with water. (author)

Real-time corrosion control system for cathodic protection of buried pipes for nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Kim, Ki Tae; Kim, Hae Woong; Kim, Young Sik [School of Materials Science and Engineering, Andong National University, Andong (Korea, Republic of); Chang, Hyun Young; Lim, Bu Taek; Park, Heung Bae [Power Engineering Research Institute, KEPCO Engineering and Construction Company, Seongnam (Korea, Republic of)

2015-02-15

Since the operation period of nuclear power plants has increased, the degradation of buried pipes gradually increases and recently it seems to be one of the emerging issues. Maintenance on buried pipes needs high quality of management system because outer surface of buried pipe contacts the various soils but inner surface reacts with various electrolytes of fluid. In the USA, USNRC and EPRI have tried to manage the degradation of buried pipes. However, there is little knowledge about the inspection procedure, test and manage program in the domestic nuclear power plants. This paper focuses on the development and build-up of real-time monitoring and control system of buried pipes. Pipes to be tested are tape-coated carbon steel pipe for primary component cooling water system, asphalt-coated cast iron pipe for fire protection system, and pre-stressed concrete cylinder pipe for sea water cooling system. A control system for cathodic protection was installed on each test pipe which has been monitored and controlled. For the calculation of protection range and optimization, computer simulation was performed using COMSOL Multiphysics (Altsoft co.)

Real-time corrosion control system for cathodic protection of buried pipes for nuclear power plant

International Nuclear Information System (INIS)

Kim, Ki Tae; Kim, Hae Woong; Kim, Young Sik; Chang, Hyun Young; Lim, Bu Taek; Park, Heung Bae

2015-01-01

Since the operation period of nuclear power plants has increased, the degradation of buried pipes gradually increases and recently it seems to be one of the emerging issues. Maintenance on buried pipes needs high quality of management system because outer surface of buried pipe contacts the various soils but inner surface reacts with various electrolytes of fluid. In the USA, USNRC and EPRI have tried to manage the degradation of buried pipes. However, there is little knowledge about the inspection procedure, test and manage program in the domestic nuclear power plants. This paper focuses on the development and build-up of real-time monitoring and control system of buried pipes. Pipes to be tested are tape-coated carbon steel pipe for primary component cooling water system, asphalt-coated cast iron pipe for fire protection system, and pre-stressed concrete cylinder pipe for sea water cooling system. A control system for cathodic protection was installed on each test pipe which has been monitored and controlled. For the calculation of protection range and optimization, computer simulation was performed using COMSOL Multiphysics (Altsoft co.)

Modelling of the viscoelastic behaviour of steel reinforced thermoplastic pipes

NARCIS (Netherlands)

Kruijer, M.P.; Warnet, Laurent; Akkerman, Remko

2006-01-01

This paper describes the analysis of the time dependent behaviour of a steel reinforced thermoplastic pipe. This new class of composite pipes is constructed of a HDPE (high-density polyethylene) liner pipe, which is over wrapped with two layers of thermoplastic tape. The thermoplastic tapes are

The precision cutting control research of automotive stainless steel thin wall pipe

Directory of Open Access Journals (Sweden)

Jin Lihong

2015-01-01

Full Text Available Stainless steel thin-walled tube are widely used in automobile industry at present, but as a result of thin wall pipe is poor strength and poor rigidity,which lead to deformation, shaped differencer and other problems in the process, it is hard to ensure the processing quality of parts. This paper proposes a method of thin stainless steel thin wall pipe cutting process in vehicle, greatly improved the problems and technical difficulties in the traditional process, the main research is about the cutting system and the hydraulic fixture design, obtained under low cost circumstances, it can realize high precision stainless steel pipes, high degree of automation to automatic cutting,simplified operation steps at the same time, increased the applicability of the system, provided a kind of advanced stainless steel thin wall pipe cutting device for the small and medium-sized enterprises.

Elevated temperature mechanical properties of line pipe steels

Science.gov (United States)

Jacobs, Taylor Roth

The effects of test temperature on the tensile properties of four line pipe steels were evaluated. The four materials include a ferrite-pearlite line pipe steel with a yield strength specification of 359 MPa (52 ksi) and three 485 MPa (70 ksi) yield strength acicular ferrite line pipe steels. Deformation behavior, ductility, strength, strain hardening rate, strain rate sensitivity, and fracture behavior were characterized at room temperature and in the temperature range of 200--350 °C, the potential operating range for steels used in oil production by the steam assisted gravity drainage process. Elevated temperature tensile testing was conducted on commercially produced as-received plates at engineering strain rates of 1.67 x 10 -4, 8.33 x 10-4, and 1.67 x 10-3 s-1. The acicular ferrite (X70) line pipe steels were also tested at elevated temperatures after aging at 200, 275, and 350 °C for 100 h under a tensile load of 419 MPa. The presence of serrated yielding depended on temperature and strain rate, and the upper bound of the temperature range where serrated yielding was observed was independent of microstructure between the ferrite-pearlite (X52) steel and the X70 steels. Serrated yielding was observed at intermediate temperatures and continuous plastic deformation was observed at room temperature and high temperatures. All steels exhibited a minimum in ductility as a function of temperature at testing conditions where serrated yielding was observed. At the higher temperatures (>275 °C) the X52 steel exhibited an increase in ductility with an increase in temperature and the X70 steels exhibited a maximum in ductility as a function of temperature. All steels exhibited a maximum in flow strength and average strain hardening rate as a function of temperature. The X52 steel exhibited maxima in flow strength and average strain hardening rate at lower temperatures than observed for the X70 steels. For all steels, the temperature where the maximum in both flow

Microstructure and Mechanical Properties of J55ERW Steel Pipe Processed by On-Line Spray Water Cooling

Directory of Open Access Journals (Sweden)

Zejun Chen

2017-04-01

Full Text Available An on-line spray water cooling (OSWC process for manufacturing electric resistance welded (ERW steel pipes is presented to enhance their mechanical properties and performances. This technique reduces the processing needed for the ERW pipe and overcomes the weakness of the conventional manufacturing technique. Industrial tests for J55 ERW steel pipe were carried out to validate the effectiveness of the OSWC process. The microstructure and mechanical properties of the J55 ERW steel pipe processed by the OSWC technology were investigated. The optimized OSWC technical parameters are presented based on the mechanical properties and impact the performance of steel pipes. The industrial tests show that the OSWC process can be used to efficiently control the microstructure, enhance mechanical properties, and improve production flexibility of steel pipes. The comprehensive mechanical properties of steel pipes processed by the OSWC are superior to those of other published J55 grade steels.

Shock Wave Speed and Transient Response of PE Pipe with Steel-Mesh Reinforcement

Directory of Open Access Journals (Sweden)

Wuyi Wan

2016-01-01

Full Text Available A steel mesh can improve the tensile strength and stability of a polyethylene (PE pipe in a water supply pipeline system. However, it can also cause more severe water hammer hazard due to increasing wave speed. In order to analyze the influence of the steel mesh on the shock wave speed and transient response processes, an improved wave speed formula is proposed by incorporating the equivalent elastic modulus. A field measurement validates the wave speed formula. Moreover, the transient wave propagation and extreme pressures are simulated and compared by the method of characteristics (MOC for reinforced PE pipes with various steel-mesh densities. Results show that a steel mesh can significantly increase the shock wave speed in a PE pipe and thus can cause severe peak pressure and hydraulic surges in a water supply pipeline system. The proposed wave speed formula can more reasonably evaluate the wave speed and improve the transient simulation of steel-mesh-reinforced PE pipes.

75 FR 53714 - Stainless Steel Butt-Weld Pipe Fittings From Japan, Korea, and Taiwan

Science.gov (United States)

2010-09-01

...)] Stainless Steel Butt-Weld Pipe Fittings From Japan, Korea, and Taiwan AGENCY: United States International... stainless steel butt-weld pipe fittings from Japan, Korea, and Taiwan. SUMMARY: The Commission hereby gives... butt-weld pipe fittings from Japan, Korea, and Taiwan would be likely to lead to continuation or...

76 FR 76437 - Certain Welded Stainless Steel Pipe From Korea and Taiwan

Science.gov (United States)

2011-12-07

... Welded Stainless Steel Pipe From Korea and Taiwan Determination On the basis of the record \\1\\ developed... antidumping duty orders on certain welded stainless steel pipe from Korea and Taiwan would be likely to lead to continuation or recurrence of material injury to an industry in the United States within a...

Dimensional control of buttwelding pipe fitting for nuclear power plant Class 1 piping systems

International Nuclear Information System (INIS)

Rodabaugh, E.C.; Moore, S.E.; Robinson, J.N.

1976-11-01

Dimensional controls of wrought steel buttwelding fittings are examined from the standpoint of design adequacy. A fairly large number of fittings were purchased from different manufacturers. The dimensions of each fitting were measured and correlated along with additional information obtained from the manufacturers in an effort to establish ''standard'' shapes. This information and a critical examination of the present ANSI standards is used to develop a ''Supplementary Standard.'' The Supplementary Standard is intended to provide improved dimensional control and more complete design information for fittings used in Class 1 nuclear power plant piping systems

Steel Fibers Reinforced Concrete Pipes - Experimental Tests and Numerical Simulation

Science.gov (United States)

Doru, Zdrenghea

2017-10-01

The paper presents in the first part a state of the art review of reinforced concrete pipes used in micro tunnelling realised through pipes jacking method and design methods for steel fibres reinforced concrete. In part two experimental tests are presented on inner pipes with diameters of 1410mm and 2200mm, and specimens (100x100x500mm) of reinforced concrete with metal fibres (35 kg / m3). In part two experimental tests are presented on pipes with inner diameters of 1410mm and 2200mm, and specimens (100x100x500mm) of reinforced concrete with steel fibres (35 kg / m3). The results obtained are analysed and are calculated residual flexural tensile strengths which characterise the post-cracking behaviour of steel fibres reinforced concrete. In the third part are presented numerical simulations of the tests of pipes and specimens. The model adopted for the pipes test was a three-dimensional model and loads considered were those obtained in experimental tests at reaching breaking forces. Tensile stresses determined were compared with mean flexural tensile strength. To validate tensile parameters of steel fibres reinforced concrete, experimental tests of the specimens were modelled with MIDAS program to reproduce the flexural breaking behaviour. To simulate post - cracking behaviour was used the method σ — ε based on the relationship stress - strain, according to RILEM TC 162-TDF. For the specimens tested were plotted F — δ diagrams, which have been superimposed for comparison with the similar diagrams of experimental tests. The comparison of experimental results with those obtained from numerical simulation leads to the following conclusions: - the maximum forces obtained by numerical calculation have higher values than the experimental values for the same tensile stresses; - forces corresponding of residual strengths have very similar values between the experimental and numerical calculations; - generally the numerical model estimates a breaking force greater

A simplified LBB evaluation procedure for austenitic and ferritic steel piping

International Nuclear Information System (INIS)

Gamble, R.M.; Wichman, K.R.

1997-01-01

The NRC previously has approved application of LBB analysis as a means to demonstrate that the probability of pipe rupture was extremely low so that dynamic loads associated with postulated pipe break could be excluded from the design basis (1). The purpose of this work was to: (1) define simplified procedures that can be used by the NRC to compute allowable lengths for circumferential throughwall cracks and assess margin against pipe fracture, and (2) verify the accuracy of the simplified procedures by comparison with available experimental data for piping having circumferential throughwall flaws. The development of the procedures was performed using techniques similar to those employed to develop ASME Code flaw evaluation procedures. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austentic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials

A serviceability approach for carbon steel piping to intermittent high temperatures

International Nuclear Information System (INIS)

Ratiu, M.D.; Moisidis, N.T.

1996-01-01

Carbon steel piping (e.g., ASME SA-106, SA-53), is installed in many industrial applications (i.e. diesel generator at NPP) where the internal gas flow subjects the piping to successive short time exposures at elevated temperatures up to 1,100 F. A typical design of this piping without consideration for creep-fatigue cumulative damage is at least incomplete if not inappropriate. Also, a design for creep-fatigue, usually employed for long-term exposure to elevated temperatures, would be too conservative and will impose replacement of the carbon steel piping with heat-resistant CrMo steel piping. The existing ASME Standard procedures do not explicitly provide acceptance criteria for the design qualification to withstand these intermittent exposures to elevated temperatures. The serviceability qualification proposed is based on the evaluation of equivalent full temperature cycles which are presumed/expected to be experienced by the exhaust piping during the design operating life of the diesel engine. The proposed serviceability analysis consists of: (a) determination of the permissible stress at elevated temperatures, and (b) estimation of creep-fatigue damage for the total expected cycles of elevated temperature exposures following the procedure provided in ASME Code Cases N-253-6 and N-47-28

A simplified LBB evaluation procedure for austenitic and ferritic steel piping

Energy Technology Data Exchange (ETDEWEB)

Gamble, R.M.; Wichman, K.R.

1997-04-01

The NRC previously has approved application of LBB analysis as a means to demonstrate that the probability of pipe rupture was extremely low so that dynamic loads associated with postulated pipe break could be excluded from the design basis (1). The purpose of this work was to: (1) define simplified procedures that can be used by the NRC to compute allowable lengths for circumferential throughwall cracks and assess margin against pipe fracture, and (2) verify the accuracy of the simplified procedures by comparison with available experimental data for piping having circumferential throughwall flaws. The development of the procedures was performed using techniques similar to those employed to develop ASME Code flaw evaluation procedures. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austentic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials.

A Study on Effect of Local Wall Thinning in Carbon Steel Elbow Pipe on Elastic Stress Concentration

International Nuclear Information System (INIS)

Kim, Jong Sung; Seo, Jae Seok

2009-01-01

Feeder pipes that connect the inlet and outlet headers to the reactor core in CANDU nuclear power plants are considered as safety Class 1 piping items. Therefore, fatigue of feeder pipes should be assessed at design stage in order to verify structural integrity during design lifetime. In accordance with the fatigue assessment result, cumulative usage factors of some feeder pipes have significant values. The feeder pipes made of SA-106 Grade B or C carbon steel have some elbows and bends. An active degradation mechanism for the carbon steel outlet feeder piping is local wall thinning due to flow-accelerated corrosion. Inspection results from plants and metallurgical examinations of removed feeders indicated the presence of localized thinning in the vicinity of the welds in the lower portion of outlet feeders, such as Grayloc hub-to-bend weld, Grayloc hub-to-elbow weld, elbow-to-elbow, and elbow-to-pipe weld. This local wall thinning can cause increase of peak stress due to stress concentration by notch effect. The increase of peak stress results in increase of cumulative usage factor. However, present fatigue assessment doesn't consider the stress concentration due to local wall-thinning. Therefore, it is necessary to assess the effect of local wall thinning on stress concentration. This study investigates the effect of local wall thinning geometry on stress concentration by performing finite element elastic stress analysis

Experimental and numerical study of steel pipe with part-wall defect reinforced with fibre glass sleeve

International Nuclear Information System (INIS)

Mazurkiewicz, Lukasz; Tomaszewski, Michal; Malachowski, Jerzy; Sybilski, Kamil; Chebakov, Mikhail; Witek, Maciej; Yukhymets, Peter; Dmitrienko, Roman

2017-01-01

The paper presents numerical and experimental burst pressure evaluation of the gas seamless hot-rolled steel pipe. The main goal was to estimate mechanical toughness of pipe wrapped with composite sleeve and verify selected sleeve thickness. The authors used a nonlinear explicit FE code with constitutive models which allows for steel and composite structure failure modelling. Thanks to the achieved numerical and analytical results it was possible to perform the comparison with data received from a capacity test and good correlation between the results were obtained. Additionally, the conducted analyses revealed that local reduction of pipe wall thickness from 6 mm to 2.4 mm due to corrosion defect can reduce high pressure resistance by about 40%. Finally, pipe repaired by a fibre glass sleeve with epoxy resin with 6 mm thickness turned out more resistant than an original steel pipe considering burst pressure. - Highlights: • Numerical and experimental burst pressure evaluation of steel pipe was performed. • Seamless hot-rolled steel pipe with and without corrosion defect were considered. • Local reduction of pipe wall thickness from 6 to 2.4 mm reduces resistance by 40%. • Pipe repaired by a 6 mm fibre glass sleeve was more resistant than an original pipe.

Development of structural steels for nuclear application

Energy Technology Data Exchange (ETDEWEB)

Hong, Jun Hwa; Chi, S. H.; Ryu, W. S.; Lee, B. S.; Kim, D. H.; Kim, J. H.; Oh, Y. J.; Byun, T. S.; Yoon, J. H.; Park, D. K.; Oh, J. M.; Cho, H. D.; Kim, H.; Kim, H. D.; Kang, S. S.; Kim, J. W.; Ahn, S. B.

1997-08-01

To established the bases of nuclear structural material technologies, this study was focused on the localization and improvement of nuclear structural steels, the production of material property data, and technology developments for integrity evaluation. The important test and analysis technologies for material integrity assessment were developed, and the materials properties of the pressure vessel steels were evaluated systematically on the basis of those technologies, they are microstructural characteristics, tensile and indentation deformation properties, impact properties, and static and dynamic fracture toughness, fatigue and corrosion fatigue etc. Irradiation tests in the research reactors were prepared or completed to obtain the mechanical properties of irradiated materials. The improvement of low alloy steel was also attempted through the comparative study on the manufacturing processes, computer assisted alloy and process design, and application of the inter critical heat treatment. On the other hand, type 304 stainless steels for reactor internals were developed and tested successfully. High strength type 316LN stainless steels for reactor internals were developed and the microstructural characteristics, corrosion resistance, mechanical properties at high temperatures, low cycle fatigue property etc. were tested and analyzed in the view point of the effect of nitrogen. Type 347 stainless steels with high corrosion resistance and toughness for pipings and tubes and low-activated Cr-Mn steels were also developed and their basic properties were evaluated. Finally, the martensitic stainless steels for turbine blade were developed and tests. (author). 242 refs., 100 tabs., 304 figs.

Development of structural steels for nuclear application

International Nuclear Information System (INIS)

Hong, Jun Hwa; Chi, S. H.; Ryu, W. S.; Lee, B. S.; Kim, D. H.; Kim, J. H.; Oh, Y. J.; Byun, T. S.; Yoon, J. H.; Park, D. K.; Oh, J. M.; Cho, H. D.; Kim, H.; Kim, H. D.; Kang, S. S.; Kim, J. W.; Ahn, S. B.

1997-08-01

To established the bases of nuclear structural material technologies, this study was focused on the localization and improvement of nuclear structural steels, the production of material property data, and technology developments for integrity evaluation. The important test and analysis technologies for material integrity assessment were developed, and the materials properties of the pressure vessel steels were evaluated systematically on the basis of those technologies, they are microstructural characteristics, tensile and indentation deformation properties, impact properties, and static and dynamic fracture toughness, fatigue and corrosion fatigue etc. Irradiation tests in the research reactors were prepared or completed to obtain the mechanical properties of irradiated materials. The improvement of low alloy steel was also attempted through the comparative study on the manufacturing processes, computer assisted alloy and process design, and application of the inter critical heat treatment. On the other hand, type 304 stainless steels for reactor internals were developed and tested successfully. High strength type 316LN stainless steels for reactor internals were developed and the microstructural characteristics, corrosion resistance, mechanical properties at high temperatures, low cycle fatigue property etc. were tested and analyzed in the view point of the effect of nitrogen. Type 347 stainless steels with high corrosion resistance and toughness for pipings and tubes and low-activated Cr-Mn steels were also developed and their basic properties were evaluated. Finally, the martensitic stainless steels for turbine blade were developed and tests. (author). 242 refs., 100 tabs., 304 figs

SCC-induced failure of a 304 stainless steel pipe

International Nuclear Information System (INIS)

Tapping, R.L.; Disney, D.J.; Szostak, F.J.

1993-01-01

On 1991 January 12, a 304 Stainless Steel (SS) suction line in the AECL-Research NRU reactor failed, shutting down the reactor for approximately 12 months. The pipe, a 32 mm schedule 40 304 stainless steel line exposed to D 2 O at temperatures ≤35 degrees C had been in service for approximately 20 years, although no manufacturing data or composition specifications were available. The failure and resultant leak resulted in a small loss of D 2 O moderator from the reactor vessel. The pipe cracked approximately 180 degrees C around the circumference of a weld. This failure was unexpected and hense a thorough metallographic examination was carried out on the failed section, on the rest of the line (Line 1212), and on representative samples from the rest of the reactor in order to assess the integrity of the remaining piping

Ice plugging of pipes using liquid nitrogen

International Nuclear Information System (INIS)

Twigg, R.J.

1987-03-01

This report presents a study on the ice plugging of pipe using liquid nitrogen, and is based on a literature review and on discussions with individuals who use the technique. Emphasis is placed on ferritic alloys, primarily carbon steels, in pipe sized up to 60 cm in diameter and on austenitic stainless steels in pipe sizes up to 30 cm in diameter. This technique is frequently used for leak testing in nuclear facilities

Mechanical properties of roll extruded nuclear reactor piping

International Nuclear Information System (INIS)

Steichen, J.M.; Knecht, R.L.

1975-01-01

The elevated temperature mechanical properties of large diameter (28 inches) seamless pipe produced by roll extrusion for use as primary piping for sodium coolant in the Fast Flux Test Facility (FFTF) have been characterized. The three heats of Type 316H stainless steel piping material used exhibited consistent mechanical properties and chemical compositions. Tensile and creep-rupture properties exceeded values on which the allowable stresses for ASME Code Case 1592 on Nuclear Components in Elevated Temperature Service were based. Tensile strength and ductility were essentially unchanged by aging in static sodium at 1050 0 F for times to 10,000 hours. High strain rate tensile tests showed that tensile properties were insensitive to strain rate at temperatures to 900 0 F and that for temperatures of 1050 0 F and above both strength and ductility significantly increased with increasing strain rate. Fatigue-crack propagation properties were comparable to results obtained on plate material and no differences in crack propagation were found between axial and circumferential orientations. (U.S.)

Utilizing clad piping to improve process plant piping integrity, reliability, and operations

International Nuclear Information System (INIS)

Chakravarti, B.

1996-01-01

During the past four years carbon steel piping clad with type 304L (UNS S30403) stainless steel has been used to solve the flow accelerated corrosion (FAC) problem in nuclear power plants with exceptional success. The product is designed to allow ''like for like'' replacement of damaged carbon steel components where the carbon steel remains the pressure boundary and type 304L (UNS S30403) stainless steel the corrosion allowance. More than 3000 feet of piping and 500 fittings in sizes from 6 to 36-in. NPS have been installed in the extraction steam and other lines of these power plants to improve reliability, eliminate inspection program, reduce O and M costs and provide operational benefits. This concept of utilizing clad piping in solving various corrosion problems in industrial and process plants by conservatively selecting a high alloy material as cladding can provide similar, significant benefits in controlling corrosion problems, minimizing maintenance cost, improving operation and reliability to control performance and risks in a highly cost effective manner. This paper will present various material combinations and applications that appear ideally suited for use of the clad piping components in process plants

78 FR 65272 - Circular Welded Carbon Steel Pipes and Tubes From Thailand: Final Results of Antidumping Duty...

Science.gov (United States)

2013-10-31

..., Saha Thai Steel Pipe (Public) Company, Ltd. (Saha Thai), and Pacific Pipe Company Limited (Pacific Pipe... Pipe (Public) Company, Ltd 0.00 Pacific Pipe Company Limited * * No shipments or sales subject to this... parties to comment on the Preliminary Results. Saha Thai, Wheatland Tube Company, and United States Steel...

Detection of wall thinning of carbon steel pipe covered with insulation using Pulsed Eddy Current technique

International Nuclear Information System (INIS)

Park, Duckgun; Kishore, M. B.; Lee, D. H.

2013-01-01

The test sample is a ferromagnetic carbon steel pipe having different thickness, covered with a 10 cm plastic insulation laminated by 0.4 mm Al plate to simulate the pipelines in NPPs. The PEC Probe used for the wall thinning detection consists of an excitation coil and a Hall sensor. The excitation coils in the probe is driven by a rectangular bipolar current pulse and the Hall-sensor will detects the resultant field. The Hall sensor output is considered as PEC signal. Results shows that the PEC system can detect wall thinning in an insulated pipeline of the NPPs. Local wall thinning in pipelines affects the structural integrity of industries like nuclear power plants (NPPs). In the present study a pulsed eddy current (PEC) technology to detect the wall thing of carbon steel pipe covered with insulation is developed

76 FR 66893 - Certain Circular Welded Carbon Steel Pipes and Tubes From India, Thailand, and Turkey; Final...

Science.gov (United States)

2011-10-28

...] Certain Circular Welded Carbon Steel Pipes and Tubes From India, Thailand, and Turkey; Final Results of... circular welded carbon steel pipes and tubes from India, Thailand, and Turkey, pursuant to section 751(c..., Thailand, and Turkey. See Antidumping Duty Order; Certain Welded Carbon Steel Standard Pipes and Tubes from...

The application of low frequency longitudinal guided wave mode for the inspection of multi-hole steel floral pipes

International Nuclear Information System (INIS)

Liu, Z H; Xie, X D; Wu, B; Li, Y H; He, C F

2012-01-01

Shed-pipe grouting technology, an effective advanced supporting method, is often used in the excavation of soft strata. Steel floral pipes are one of the key load-carrying components of shed-pipe grouting supporting structures. Guided waves are a very attractive methodology to inspect multi-hole steel floral pipes as they offer long range inspection capability, mode and frequency tuning, and cost effectiveness. In this contribution, preliminary experiments are described for the inspection of steel floral pipes using a low frequency longitudinal guided wave mode, L(0,2). The relation between the number of grouting holes and the peak-to-peak amplitude of the first end-reflected signal was obtained. The effect of the grouting holes in steel floral pipes on the propagation velocity of the L(0,2) mode at 30 kHz was analyzed. Experimental results indicate that the typical grouting holes in steel floral pipe have no significant effect on the propagation of this mode. As a result, low frequency longitudinal guided wave modes have potential for the non-destructive long range inspection of multi-hole steel floral pipes. Furthermore, the propagation velocity of the investigated L(0,2) mode at 30 kHz decreases linearly with the increase of the number of grouting holes in a steel floral pipe. It is also noticeable that the effect of the grouting holes cumulates along with the increase in the number of grouting holes and subsequent increase in reflection times of longitudinal guided waves in the steel floral pipe. The application potential of the low frequency longitudinal guided wave technique for the inspection of embedded steel floral pipes is discussed.

76 FR 67473 - Stainless Steel Butt-Weld Pipe Fittings from Italy, Malaysia, and The Philippines; Institution of...

Science.gov (United States)

2011-11-01

... Concerning the Antidumping Duty Orders on Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and... stainless steel butt-weld pipe fittings from Italy, Malaysia, and the Philippines would be likely to lead to... antidumping duty orders on imports of stainless steel butt-weld pipe fittings from Italy, Malaysia, and the...

The residual stress distribution in welded pipe inner surface of stainless steel from the nuclear power plant in Ringhals

International Nuclear Information System (INIS)

Larsson, L.E.

1984-06-01

The axial residual stress distribution on the inner surface of welded pipes of stainless steel SS 2333 (AISI 304) have been measured using the X-ray diffraction technique. Four halves of two pipes with the outer diameter of 114 mm and wall thickness of 10 mm were investigated. The result on the pipe inner surface shows compressive stresses in the weld metal and tensile stresses within a region between 8-23 mm with a maximum of 180MPa at a distance of 17 mm from the weld centerline. The maximum axial and circumferential residual stresses on the pipe outer surface are of the magnitude of 100 MPa. By cutting the pipes into two halves these stresses are relaxed by about 35 MPa. (author)

Assessment of Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using Advanced Low-Frequency Ultrasonic Methods

Energy Technology Data Exchange (ETDEWEB)

Anderson, Michael T.; Crawford, Susan L.; Cumblidge, Stephen E.; Denslow, Kayte M.; Diaz, Aaron A.; Doctor, Steven R.

2007-03-01

Studies conducted at the Pacific Northwest National Laboratory in Richland, Washington, have focused on assessing the effectiveness and reliability of novel approaches to nondestructive examination (NDE) for inspecting coarse-grained, cast stainless steel reactor components. The primary objective of this work is to provide information to the U.S. Nuclear Regulatory Commission on the effectiveness and reliability of advanced NDE methods as related to the inservice inspection of safety-related components in pressurized water reactors (PWRs). This report provides progress, recent developments, and results from an assessment of low frequency ultrasonic testing (UT) for detection of inside surface-breaking cracks in cast stainless steel reactor piping weldments as applied from the outside surface of the components. Vintage centrifugally cast stainless steel piping segments were examined to assess the capability of low-frequency UT to adequately penetrate challenging microstructures and determine acoustic propagation limitations or conditions that may interfere with reliable flaw detection. In addition, welded specimens containing mechanical and thermal fatigue cracks were examined. The specimens were fabricated using vintage centrifugally cast and statically cast stainless steel materials, which are typical of configurations installed in PWR primary coolant circuits. Ultrasonic studies on the vintage centrifugally cast stainless steel piping segments were conducted with a 400-kHz synthetic aperture focusing technique and phased array technology applied at 500 kHz, 750 kHz, and 1.0 MHz. Flaw detection and characterization on the welded specimens was performed with the phased array method operating at the frequencies stated above. This report documents the methodologies used and provides results from laboratory studies to assess baseline material noise, crack detection, and length-sizing capability for low-frequency UT in cast stainless steel piping.

Fracture studies on stainless steel straight pipes under earthquake-type cyclic loading

International Nuclear Information System (INIS)

Raghava, G.; Vishnuvardhan, S.; Gandhi, P.; Vaze, K.K.

2014-01-01

In order to study the crack growth and cyclic fracture behaviour, which are required for realistic assessment of Leak Before Break (LBB) applicability, experimental investigations were carried out on straight pipes under quasi-crystal loading. Totally 13 pipes were tested; three were stainless steel welded (SSW) using conventional shielded metal arc welding (SMAW) technique and the remaining specimens were Narrow Gap Welded (NGW). The fracture tests were carried out under load control, displacement control and combination of the two; the pipes were subjected to different amplitudes of load or load-line displacement (LLD), which were decided based on the response of the pipes under monotonic loading. Cyclic tearing and crack growth studies on eight straight pipes of the same material reported earlier in published literature are also considered for studying the results and understanding the behaviour. Under load control, with almost equal load amplitude, the NGW pipe exhibited improved life in comparison with SMAW pipe when both are subjected to cyclic loading. The crack growth and tearing instability behaviour of the pipes were studied. The same were found to be different for load control, displacement control and combined control tests. Based in the load-controlled experimental results, material specific plot between cyclic load amplitude (as a percentage of maximum load carrying capacity of a specimen under monotonic fracture) and number of cycles to failure was obtained. The results indicate that the piping components subjected to quasi-cyclic loading may fail in very less number of cycles even when the load amplitude is sufficiently below the monotonic fracture/collapse load. These studies will be helpful in designing nuclear power plant (NPP) piping components subjected to earthquake-type cyclic loading. (author)

Application of laser cladding method to small-diameter stainless steel pipes in actual nuclear plant

International Nuclear Information System (INIS)

Atago, Y.; Yamadera, M.; Tsuji, H.; Shiraiwa, T.; Kanno, M.

1995-01-01

Recently, to prevent stress corrosion cracking (SCC) the material of stainless steel (Type 304), a laser cladding method which produces a highly corrosion-resisting coating (cladding) to be formed on the surface of the material was developed. This is applicable to a long distance and narrow space, because of the good accessibility of the YAG (Yttrium-Aluminum Garnet) laser beam that can be transmitted through an optical fiber. In this method, a paste mixed metallic powder and heating resistive organic solvent is firstly placed on the inner surface of a small pipe and then a YAG laser beam transmitted through an optical fiber is irradiated to the paste, which will be melted and formed a clad subsequently, which is excellent in corrosion resistance. Finally, it can be achieved further resistance against the SCC due to the clad layer formed thus on the surface of the material. Recently, this Laser Cladding method was practically and successfully applied to the actual BWR Nuclear Power Plant in Japan. This report introduces the laser cladding technique, the equipments developed for practical application in the field

77 FR 42697 - Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines: Continuation...

Science.gov (United States)

2012-07-20

...] Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines: Continuation of... from Italy, Malaysia, and the Philippines.\\2\\ \\1\\ See Antidumping Duty Orders: Stainless Steel Butt...), titled Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines (Investigation...

Effect of dynamic monotonic and cyclic loading on fracture behavior for Japanese carbon steel pipe STS410

Energy Technology Data Exchange (ETDEWEB)

Kinoshita, Kanji; Murayama, Kouichi; Ogata, Hiroyuki [and others

1997-04-01

The fracture behavior for Japanese carbon steel pipe STS410 was examined under dynamic monotonic and cyclic loading through a research program of International Piping Integrity Research Group (EPIRG-2), in order to evaluate the strength of pipe during the seismic event The tensile test and the fracture toughness test were conducted for base metal and TIG weld metal. Three base metal pipe specimens, 1,500mm in length and 6-inch diameter sch.120, were employed for a quasi-static monotonic, a dynamic monotonic and a dynamic cyclic loading pipe fracture tests. One weld joint pipe specimen was also employed for a dynamic cyclic loading test In the dynamic cyclic loading test, the displacement was controlled as applying the fully reversed load (R=-1). The pipe specimens with a circumferential through-wall crack were subjected four point bending load at 300C in air. Japanese STS410 carbon steel pipe material was found to have high toughness under dynamic loading condition through the CT fracture toughness test. As the results of pipe fracture tests, the maximum moment to pipe fracture under dynamic monotonic and cyclic loading condition, could be estimated by plastic collapse criterion and the effect of dynamic monotonic loading and cyclic loading was a little on the maximum moment to pipe fracture of the STS410 carbon steel pipe. The STS410 carbon steel pipe seemed to be less sensitive to dynamic and cyclic loading effects than the A106Gr.B carbon steel pipe evaluated in IPIRG-1 program.

Effect of dynamic monotonic and cyclic loading on fracture behavior for Japanese carbon steel pipe STS410

International Nuclear Information System (INIS)

Kinoshita, Kanji; Murayama, Kouichi; Ogata, Hiroyuki

1997-01-01

The fracture behavior for Japanese carbon steel pipe STS410 was examined under dynamic monotonic and cyclic loading through a research program of International Piping Integrity Research Group (EPIRG-2), in order to evaluate the strength of pipe during the seismic event The tensile test and the fracture toughness test were conducted for base metal and TIG weld metal. Three base metal pipe specimens, 1,500mm in length and 6-inch diameter sch.120, were employed for a quasi-static monotonic, a dynamic monotonic and a dynamic cyclic loading pipe fracture tests. One weld joint pipe specimen was also employed for a dynamic cyclic loading test In the dynamic cyclic loading test, the displacement was controlled as applying the fully reversed load (R=-1). The pipe specimens with a circumferential through-wall crack were subjected four point bending load at 300C in air. Japanese STS410 carbon steel pipe material was found to have high toughness under dynamic loading condition through the CT fracture toughness test. As the results of pipe fracture tests, the maximum moment to pipe fracture under dynamic monotonic and cyclic loading condition, could be estimated by plastic collapse criterion and the effect of dynamic monotonic loading and cyclic loading was a little on the maximum moment to pipe fracture of the STS410 carbon steel pipe. The STS410 carbon steel pipe seemed to be less sensitive to dynamic and cyclic loading effects than the A106Gr.B carbon steel pipe evaluated in IPIRG-1 program

The Study on Environmental Fatigue Behavior of Low Alloy Steel and Stainless Steel Pipes Using the Simplified Plant Transients

International Nuclear Information System (INIS)

Yoo, One; Song, M. S.; Kim, I. Y.; Park, S. H.; Lee, B. S.

2010-01-01

Nuclear components categorized as ASME Code Class 1 shall be evaluated for the fatigue and satisfy the fatigue acceptance criteria, CUF(cumulative usage factor) < 1 in accordance with ASME Code. However, recent studies have shown the fatigue evaluation procedure may not give conservative results when the components operate in the water environment. NRC issued Regulatory Guide 1.207 which enforces the new fatigue evaluation method or Fen(environmental fatigue correction factor) method to nuclear plants to be newly constructed. This paper describes the characteristics of the behavior of low alloy and austenitic stainless steel straight pipe related to environmental fatigue, which are obtained by using the method suggested by Regulatory Guide 1.207 and simplified plant transients

75 FR 76025 - Stainless Steel Butt-Weld Pipe Fittings From Japan, Korea, and Taiwan

Science.gov (United States)

2010-12-07

... INTERNATIONAL TRADE COMMISSION [Investigation Nos. 731-TA-376 and 563-564 (Third Review)] Stainless Steel Butt-Weld Pipe Fittings From Japan, Korea, and Taiwan AGENCY: United States International... steel butt-weld pipe fittings from Japan, Korea, and Taiwan would be likely to lead to continuation or...

Effects of the inner mould material on the aluminium–316L stainless steel explosive clad pipe

International Nuclear Information System (INIS)

Guo, Xunzhong; Tao, Jie; Wang, Wentao; Li, Huaguan; Wang, Chen

2013-01-01

Highlights: ► Different mould materials were adopted to evaluate the effect of the constraint on the clad quality. ► The interface characteristics of clad pipe were analyzed for the different clad pipe. ► The clad pipes possess excellent bonding quality. - Abstract: The clad pipe played an important part in the pipeline system of the nuclear power industry. To prepare the clad pipe with even macrosize and excellent bonding quality, in this work, different mould materials were adopted to evaluate the effect of the constraint on the clad quality of the bimetal pipe prepared by explosive cladding. The experiment results indicated that, the dimension uniformity and bonding interface of clad pipe were poor by using low melting point alloy as mould material; the local bulge or the cracking of the clad pipe existed when the SiC powder was utilized. When the steel mould was adopted, the outer diameter of the clad pipe was uniform from head to tail. In addition, the metallurgical bonding was formed. Furthermore, the results of shear test, bending test and flattening test showed that the bonding quality was excellent. Therefore, the Al–316L SS clad pipe could endure the second plastic forming

An Investigation of Aging Behaviour in Microalloyed Steel (X70) UOE Pipe

Science.gov (United States)

Wiskel, J. B.; Ma, J.; Ivey, D. G.; Henein, H.

Aging of microalloyed steel pipe can occur at relatively low temperatures associated with the pipe coating process and/or during long term storage or use. The aging phenomenon is primarily attributed to C diffusion to dislocations and subsequent pinning of these dislocations. Important factors in the aging process include time, temperature, chemical composition and plastic deformation (arising from the pipe forming process). The work presented in this paper uses a Box-Behnken experimental design to determine the effect of time, temperature, location in the UOE pipe (90° or 180° to the weld), position through the pipe wall thickness (ID, CL or OD) and the steel's C/Nb ratio (0.60, 1.25 and 1.80) on the change in yield strength of three (uncoated) X70 UOE pipes. Quantitative microstructure analysis is undertaken to determine the grain size and microconstituent fractions of the as-received pipe material. Quadratic equations and response surface(s) correlating the significant aging variables with changes in the longitudinal yield stress of the pipe are developed. Both through thickness position and the C/Nb ratio, followed by aging temperature, had the largest effect on the change in longitudinal yield strength.

75 FR 69125 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China

Science.gov (United States)

2010-11-10

... with material injury by reason of imports from China of certain seamless carbon and alloy steel standard, line, and pressure pipe (``seamless SLP pipe''), provided for in subheadings 7304.19.10, 7304.19... Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China Determination On the basis of...

78 FR 31574 - Welded Stainless Steel Pressure Pipe From Malaysia, Thailand, and Vietnam; Institution of...

Science.gov (United States)

2013-05-24

... INTERNATIONAL TRADE COMMISSION [Investigation Nos. 731-TA-1210-1212 (Preliminary)] Welded Stainless Steel Pressure Pipe From Malaysia, Thailand, and Vietnam; Institution of Antidumping Duty..., by reason of imports from Malaysia, Thailand, and Vietnam of welded stainless steel pressure pipe...

Investigation on Mechanical Properties of Austenitic Stainless-Steel Pipes Welded by TIG Method

Directory of Open Access Journals (Sweden)

Mushtaq Albdiry

2017-11-01

Full Text Available This paper investigates the mechanical properties of austenitic stainless steel (type 204 pipes welded by Tungsten Inert Gas (TIG welding process. Testing of hardness (HRC, tensile strength and bending strength was performed for the steel pipes welded at two different welding temperatures (700 °C and 900 °C with and without using the weld filler wire. The microstructure of the welding regions was examined by using an optical microscopy. The properties showed that the steel pipes welded by 900 °C with using the weld filler obtained the highest tensile strength and bending strength versus these welded by 700 °C without the use of the weld filler. This is attributed to the weld filler heated and melt at sufficient temperature (900 °C and compensate losing in the Ni metal occurred in the base steel metal during the welding process.

Experimental analysis on elasto-platic behaviour of T-branched stainless steel pipe

International Nuclear Information System (INIS)

Citti, P.; Nerli, G.; Reale, S.; Rissone, P.

1979-01-01

Paper relates on results of a research, still in progress at Laboratories of Istituto di Ingegneria Meccanica of Florence University with close cooperation of CNEN Casaccia Laboratories, on incremental collapse phenomena with progressively increasing deflections and plastic fatigue phenomena in stainless steel piping components subjected to variable repeated loads. The reference is to emergency and faulted load contitions as they are defined in ASME III Code. The models are made by stainless steel pipe and simulate some primary circuit piping components. Namely models are not-symmetrical T-branched pipes fixed at their flanged ends and loaded in two sections by variable repeated loads. Tests are carried out to determine: plastic collapse load; strain hardening behaviour; shackedown load conditions. A numerical model is also developed to describe the incremental collapse phenomena. (orig.)

Pipe-CUI-profiler: a portable nucleonic system for detecting corrosion under insulation (CUI) of steel pipes

International Nuclear Information System (INIS)

Jaafar Abdullah; Rasif Mohd Zain; Roslan Yahya

2003-01-01

Corrosion under insulation (CUI) on the external wall of steel pipes is a common problem in many types of industrial plants. This is mainly due to the presence of moisture or water in the insulation materials. A portable nucleonic system that can be used to detect CUI without the need to remove the insulation materials, has been developed. The system is based on dual-beam gamma-ray absorption technique. It is designed to inspect pipes of internal diameter 50, 65, 80, 90, 100 and 150 mm. Pipeline of these sizes with aluminium or thin steel sheathing, containing fibre-glass or calcium silicate insulation to thicknesses of 25, 40 and 50 mm can be inspected. The system has proven to be a safe, fast and effective method of inspecting insulated pipes. This paper describes the new nucleonic system that has been developed. This paper describes the basic principle of the system and outlines its performance. (Author)

77 FR 39735 - Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines

Science.gov (United States)

2012-07-05

... revocation of the antidumping duty orders on stainless steel butt-weld pipe fittings From Italy, Malaysia... INTERNATIONAL TRADE COMMISSION [Investigation Nos. 731-TA-865-867 (Second Review)] Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines Determination On the basis of the...

78 FR 72863 - Circular Welded Carbon-Quality Steel Pipe From the People's Republic of China: Continuation of...

Science.gov (United States)

2013-12-04

...-Quality Steel Pipe From the People's Republic of China: Continuation of Antidumping Duty Order AGENCY... circular welded carbon-quality steel pipe (``circular welded pipe'') from the People's Republic of China...\\ See Initiation of Five-Year (``Sunset'') Review, 78 FR 33063 (June 3, 2013). \\2\\ See Circular Welded...

77 FR 14002 - Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines: Final Results...

Science.gov (United States)

2012-03-08

...] Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines: Final Results of the... Duty Orders on Stainless Steel Butt-Weld Pipe Fittings from Italy, Malaysia, and the Philippines'' from... Commerce (the Department) initiated sunset reviews of the antidumping duty orders on stainless steel butt...

Fatigue crack growth in austenitic stainless steel piping

International Nuclear Information System (INIS)

Bethmont, M.; Cheissoux, J.L.; Lebey, J.

1981-04-01

The study presented in this paper is being carried out with a view to substantiating the calculations of the fatigue crack growth in pipes made of 316 L stainless steel. The results obtained may be applied to P.W.R. primary piping. It is divided into two parts. First, fatigue tests (cyclic pressure) are carried out under hot and cold conditions with straight pipes machined with notches of various dimensions. The crack propagation and the fatigue crack growth rate are measured here. Second, calculations are made in order to interpret experimental results. From elastic calculations the stress intensity factor is assessed to predict the crack growth rate. The results obtained until now and presented in this paper relate to longitudinal notches

Steel fiber reinforced concrete pipes: part 1: technological analysis of the mechanical behavior

Directory of Open Access Journals (Sweden)

A. D. de Figueiredo

Full Text Available This paper is the first part of an extensive work focusing the technological development of steel fiber reinforced concrete pipes (FRCP. Here is presented and discussed the experimental campaign focusing the test procedure and the mechanical behavior obtained for each of the dosages of fiber used. In the second part ("Steel fiber reinforced concrete pipes. Part 2: Numerical model to simulate the crushing test", the aspects of FRCP numerical modeling are presented and analyzed using the same experimental results in order to be validated. This study was carried out trying to reduce some uncertainties related to FRCP performance and provide a better condition to the use of these components. In this respect, an experimental study was carried out using sewage concrete pipes in full scale as specimens. The diameter of the specimens was 600 mm, and they had a length of 2500 mm. The pipes were reinforced with traditional bars and different contents of steel fibers in order to compare their performance through the crushing test. Two test procedures were used in that sense. In the 1st Series, the diameter displacement was monitored by the use of two LVDTs positioned at both extremities of the pipes. In the 2nd Series, just one LVDT is positioned at the spigot. The results shown a more rigidity response of the pipe during tests when the displacements were measured at the enlarged section of the socket. The fiber reinforcement was very effective, especially when low level of displacement was imposed to the FRCP. At this condition, the steel fibers showed an equivalent performance to superior class pipes made with traditional reinforced. The fiber content of 40 kg/m3 provided a hardening behavior for the FRCP, and could be considered as equivalent to the critical volume in this condition.

On the impact bending test technique for high-strength pipe steels

Science.gov (United States)

Arsenkin, A. M.; Odesskii, P. D.; Shabalov, I. P.; Likhachev, M. V.

2015-10-01

It is shown that the impact toughness (KCV-40 = 250 J/cm2) accepted for pipe steels of strength class K65 (σy ≥ 550 MPa) intended for large-diameter gas line pipes is ineffective to classify steels in fracture strength. The results obtained upon testing of specimens with a fatigue crack and additional sharp lateral grooves seem to be more effective. In energy consumption, a macrorelief with splits is found to be intermediate between ductile fracture and crystalline brittle fracture. A split formation mechanism is considered and a scheme is proposed for split formation.

Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

International Nuclear Information System (INIS)

Zhang Weiguo; Gao Fengqin; Zhou Hongyi

1992-03-01

The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

International Nuclear Information System (INIS)

Zhang Weiguo; Gao Fengqin; Zhou Hongyi

1993-01-01

The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy-800, Inconel-600, Inconel-690 and 321 SS which are used for steam generator in PWR. The effects of material metallurgy, shot-peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC are investigated

Use of Nitrocarburizing for Strengthening Threaded Joints of Drill Pipes from Medium-Carbon Alloy Steels

Science.gov (United States)

Priymak, E. Yu.; Stepanchukova, A. V.; Yakovleva, I. L.; Tereshchenko, N. A.

2015-05-01

Nitrocarburizing is tested at the Drill Equipment Plant for reinforcing threaded joints of drill pipes for units with retrievable core receiver (RCR). The effect of the nitrocarburizing on the mechanical properties of steels of different alloying systems is considered. Steels for the production of threaded joints of drill pipes are recommended.

Effects of dynamic coupling between freestanding steel containment and attached piping

International Nuclear Information System (INIS)

Kennedy, R.P.; Kincaid, R.H.; Short, S.A.

1981-01-01

This paper presents an accurate, practical method of converting uncoupled response time history results obtained from an uncoupled structure model into coupled response time histories using a post-processor routine. The method is rigorous and only requires the modal properties of the uncoupled structure model, the modal properties of the uncoupled attached equipment model, and the uncoupled time histories of the attachment points on the structure. Coupled response spectra or time histories for use as input to an uncoupled equipment model are obtained. Comparisons of coupled versus uncoupled analysis results are presented for representative piping systems attached to a typical BWR Mark III steel containment subjected to vibration from safety relief valve discharge with a fundamental frequency of 12 Hz. It is shown that the coupled response spectra at piping attachment points are reduced by a factor between 2 and 5 from the amplified uncoupled spectra at each significant piping modal frequency above 20 Hz for representative major piping systems attached to the unstiffened portion of the steel shell. Responses at lower frequencies are not generally reduced and may increase by coupling effects for the input loading and shell model studied. Peak accerations are generally significantly reduced while peak displacements may be decreased or increased. Rules are presented for estimating the coupling effects between freestanding steel shells and attached equipment. (orig./HP)

Qualification of diesel generator exhaust carbon steel piping to intermitted elevated temperatures

International Nuclear Information System (INIS)

Ratiu, M.D.; Moisidis, N.T.

1996-01-01

The diesel generator exhaust piping, usually made up of carbon steel piping (e.g., ASME SA-106, SA-53), is subjected to successive short time exposures at elevated temperatures up to 1,000 F (538 C). A typical design of this piping, without consideration for creep-fatigue cumulative damage, is at least incomplete, if not inappropriate. Also, a design for creep-fatigue, usually employed for long-term exposure to elevated temperatures, would be too conservative and will impose replacement of the carbon steel piping with heat-resistant CrMo alloy piping. The existing ASME standard procedures do not explicitly provide acceptance criteria for the design qualification to withstand these intermittent exposures to elevated temperatures. The serviceability qualification proposed is based on the evaluation of equivalent full temperature cycles which are presumed/expected to be experienced by the exhaust piping during the design operating life of the diesel engine. The proposed serviceability analysis consists of: (a) determination of the permissible stress at elevated temperatures, and (b) estimation of creep-fatigue damage for the total expected cycles of elevated temperature exposures following the procedure provided in ASME Code Cases N-253-6 and N-47-28

Pipe restraints for nuclear power plants

International Nuclear Information System (INIS)

Keever, R.E.; Broman, R.; Shevekov, S.

1976-01-01

A pipe restraint for nuclear power plants in which a support member is anchored on supporting surface is described. Formed in the support member is a semicylindrical wall. Seated on the semicylindrical wall is a ring-shaped pipe restrainer that has an inner cylindrical wall. The inner cylindrical wall of the pipe restrainer encircles the pressurized pipe. In a modification of the pipe restraint, an arched-shaped pipe restrainer is disposed to overlie a pressurized pipe. The ends of the arch-shaped pipe restrainer are fixed to support members, which are anchored in concrete or to a supporting surface. A strap depends from the arch-shaped pipe restrainer. The pressurized pipe is supported by the depending strap

Report of examination of the ruptured pipe at the Hamaoka Nuclear Power Station Unit-1

International Nuclear Information System (INIS)

2001-12-01

In order to investigate root cause of the pipe rupture, which took place at the Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company on November 7, 2001, a task force was established within the Nuclear and Industrial Safety Agency (NISA) and initiated a detailed investigation of the ruptured pipe. The Japan Atomic Energy Research Institute (JAERI) was asked from the Ministry of Education, Culture, Sports, Science and Technology (MEXT) in response to the request from NISA to cooperate as an independent neutral organization with NISA and perform an examination of the ruptured pipe independently from Chubu Electric Power Company. JAERI accepted the request by considering the fact that JAERI is an integrated research institution for nuclear research and development, a prime research institution for nuclear safety research, a research institution with experience of root-cause investigation of various nuclear incidents and accidents of domestic as well as overseas, and a research institution provided with advanced examination facilities necessary for examination of the ruptured pipe. The JAERI examination group was formed at the Tokai Research Establishment and conducted detailed and thorough examination of the pieces taken from the ruptured pipe primarily in the Reactor Fuel Examination Facility (RFEF) with the use of tools such as scanning electron microscopes and other equipments. Purpose of examination was to provide technical information in order to identify causes of the pipe rupture through examination of the pieces taken from the ruptured region of the pipe. The following findings and conclusion were made as the result of the present examination. (1) Wall thickness of the pipe was significantly reduced in the ruptured region. (2) Dimple pattern resulting from ductile fracture by shearing was observed in the fracture surfaces of nearly all of the pieces and no indication of fatigue crack growth was found. (3) Microstructure showed a typical carbon

Ultrasound propagation in steel piping at electric power plant using clamp-on ultrasonic pulse doppler velocity-profile flowmeter

International Nuclear Information System (INIS)

Tezuka, Kenichi; Mori, Michitsugu; Wada, Sanehiro; Aritomi, Masanori; Kikura, Hiroshige

2008-01-01

Venturi nozzles are widely used to measure the flow rates of reactor feedwater. This flow rate of nuclear reactor feedwater is an important factor in the operation of nuclear power reactors. Some other types of flowmeters have been proposed to improve measurement accuracy. The ultrasonic pulse Doppler velocity-profile flowmeter is expected to be a candidate method because it can measure the flow profiles across the pipe cross sections. For the accurate estimation of the flow velocity, the incidence angle of ultrasonic entering the fluid should be carefully estimated by the theoretical approach. However, the evaluation of the ultrasound propagation is not straightforward for the several reasons such as temperature gradient in the wedge or mode conversion at the interface between the wedge and pipe. In recent years, the simulation code for ultrasound propagation has come into use in the nuclear field for nondestructive testing. This article analyzes and discusses ultrasound propagation in steel piping and water, using the 3D-FEM simulation code and the Kirchhoff method, as it relates to the flow profile measurements in power plants with the ultrasonic pulse Doppler velocity-profile flowmeter. (author)

Materials Integrity Analysis for Application of Hyper Duplex Stainless Steels to Korean Nuclear Power Plants

International Nuclear Information System (INIS)

Chang, Hyun Young; Park, Heung Bae; Park, Yong Soo; Kim, Soon Tae; Kim, Young Sik; Kim, Kwang Tae; Jhang, Yoon Young

2010-01-01

Hyper duplex stainless steels have been developed in Korea for the purpose of application to the seawater system of Korean nuclear power plants. This system supplies seawater to cooling water heat exchanger tubes, related pipes and chlorine injection system. In normal operation, seawater is supplied to heat exchanger through the exit of circulating water pump headers, and the heat exchanged sea water is extracted to the discharge pipes in circulating water system connected to the circulating water discharge lines. The high flow velocity of some part of seawater system in nuclear power plants accelerates damages of components. Therefore, high strength and high corrosion resistant steels need to be applied for this environment. Hyper duplex stainless steel (27Cr-7.0Ni-2.5Mo-3.2W-0.35N) has been newly developed in Korea and is being improved for applying to nuclear power plants. In this study, the physical and mechanical properties and corrosion resistance of newly developed materials are quantitatively evaluated in comparative to commercial stainless steels in other countries. The properties of weld and HAZ (heat affected zone) are analyzed and the best compositions are suggested. The optimum conditions in welding process are derived for ensuring the volume fraction of ferrite(α) and austenite(γ) in HAZ and controlling weld cracks. For applying these materials to the seawater heat exchanger, CCT and CPT in weldments are measured. As a result of all experiments, it was found that the newly developed hyper duplex stainless steel WREMBA has higher corrosion resistance and mechanical properties than those of super austenitic stainless steels including welded area. It is expected to be a promising material for seawater systems of Korean nuclear power plants

76 FR 72173 - Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab...

Science.gov (United States)

2011-11-22

...-552-810] Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab... Steel Pipe from India, Oman, the United Arab Emirates, and Vietnam, dated October 26, 2011 (hereinafter... pipe from India, the Sultanate of Oman (``Oman''), the United Arab Emirates (``the UAE''), and the...

Seismic analysis of nuclear piping system

International Nuclear Information System (INIS)

Shrivastava, S.K.; Pillai, K.R.V.; Nandakumar, S.

1975-01-01

To illustrate seismic analysis of nuclear power plant piping, a simple piping system running between two floors of the reactor building is assumed. Reactor building floor response is derived from time-history method. El Centre earthquake (1940) accelerogram is used for time-history analysis. The piping system is analysed as multimass lumped system. Behaviour of the pipe during the said earthquake is discussed. (author)

The influence of gouge defects on failure pressure of steel pipes

International Nuclear Information System (INIS)

Alang, N A; Razak, N A; Zulfadli, M R

2013-01-01

Failure pressure of API X42 steel pipes with gouge defects was estimated through a nonlinear finite element (FE) analysis. The effect of gouge length on failure pressure of different pipe diameters was investigated. Stress modified critical strain (SMCS) model was applied as in predicting the failure of the pipe. The model uses strain based criteria to predict the failure. For validation of the model, the FE results were compared to experimental data in literature showing overall good agreement. The results show that the gouge length has significant influence on failure pressure. A smaller pipe diameter gives highest value of failure pressure

Experimental electro-thermal method for nondestructively testing welds in stainless steel pipes

International Nuclear Information System (INIS)

Green, D.R.

1979-01-01

Welds in austenitic stainless steel pipes are notoriously difficult to nondestructively examine using conventional ultrasonic and eddy current methods. Survace irregularities and microscopic variations in magnetic permeability cause false eddy current signal variations. Ultrasonic methods have been developed which use computer processing of the data to overcome some of the problems. Electro-thermal nondestructive testing shows promise for detecting flaws that are difficult to detect using other NDT methods. Results of a project completed to develop and demonstrate the potential of an electro-thermal method for nondestructively testing stainless steel pipe welds are presented. Electro-thermal NDT uses a brief pulse of electrical current injected into the pipe. Defects at any depth within the weld cause small differences in surface electrical current distribution. These cause short-lived transient temperature differences on the pipe's surface that are mapped using an infrared scanning camera. Localized microstructural differences and normal surface roughness in the welds have little effect on the surface temperatures

76 FR 66899 - Certain Circular Welded Non-Alloy Steel Pipe From Brazil, Mexico, the Republic of Korea, and...

Science.gov (United States)

2011-10-28

... construction, shipbuilding, trucking, farm equipment, and related industries. Unfinished conduit pipe is also..., shipbuilding, trucking, farm-equipment, and related industries. Unfinished conduit pipe is also included in the... Steel Tube Works Co., Ltd 11.63 Pusan Steel Pipe Co., Ltd 4.91 All Others 6.37 Taiwan: Kao Hsing Chang...

Nuclear class 1 piping stress analysis

International Nuclear Information System (INIS)

Lucas, J.C.R.; Maneschy, J.E.; Mariano, L.A.; Tamura, M.

1981-01-01

A nuclear class 1 piping stress analysis, according to the ASME code, is presented. The TRHEAT computer code has been used to determine the piping wall thermal gradient. The Nupipe computer code was employed for the piping stress analysis. Computer results were compared with the allowable criteria from the ASME code. (Author) [pt

Development of Wall Thinning Distinction Method using the Multi-inspecting UT Data of Carbon Steel Piping

Energy Technology Data Exchange (ETDEWEB)

Hwang, Kyeong Mo; Yun, Hun; Lee, Chan Kyoo [KEPCO E and C, Yongin (Korea, Republic of)

2012-05-15

To manage the wall thinning of carbon steel piping in nuclear power plants, the utility of Korea has performed thickness inspection for some quantity of pipe components during refueling outages and determined whether repair or replacement after evaluating UT (Ultrasonic Test) data. When the existing UT data evaluation methods, such as Band, Blanket, PTP (Point to Point) Methods, are applied to a certain pipe component, unnecessary re-inspecting situations may be generated even though the component does not thinned. In those cases, economical loss caused by repeated inspection and problems of maintaining the pipe integrity followed by decreasing of newly inspected components may be generated. EPRI (Electric Power Research Institute) in USA has suggested several statistical methods, TPM (Total Point Method), LSS (Least Square Slope) Method, etc. to distinguish whether multiple inspecting components have thinned or not. This paper presents the analysis results for multiple inspecting components over three times based on both NAM (Near Area of Minimum) Method developed by KEPCO-E and C and the other methods suggested by EPRI.

X-ray diffraction study of microstructural changes during fatigue damage initiation in steel pipes

Energy Technology Data Exchange (ETDEWEB)

Pinheiro, B., E-mail: bianca@lts.coppe.ufrj.br [Laboratory of Mechanics of Lille (LML), UMR CNRS 8107, University of Lille 1, Boulevard Paul Langevin, Cite Scientifique, 59655 Villeneuve d' Ascq (France); Lesage, J. [Laboratory of Mechanics of Lille (LML), UMR CNRS 8107, University of Lille 1, Boulevard Paul Langevin, Cite Scientifique, 59655 Villeneuve d' Ascq (France); Pasqualino, I. [Subsea Technology Laboratory (LTS), Ocean Engineering Department, COPPE/Federal University of Rio de Janeiro, PO Box 68508, Cidade Universitaria, CEP 21945-970, Rio de Janeiro/RJ (Brazil); Benseddiq, N. [Laboratory of Mechanics of Lille (LML), UMR CNRS 8107, University of Lille 1, Boulevard Paul Langevin, Cite Scientifique, 59655 Villeneuve d' Ascq (France); Bemporad, E. [Interdepartmental Laboratory of Electron Microscopy (LIME), University of Rome TRE, Via Della Vasca Navale 79, 00146 Rome (Italy)

2012-01-15

Highlights: Black-Right-Pointing-Pointer In this work we study the fatigue damage evolution in an API 5L X60 steel. Black-Right-Pointing-Pointer Microstructural changes and residual stresses are evaluated during fatigue tests. Black-Right-Pointing-Pointer Microdeformations and macro residual stresses are estimated by X-ray diffraction. Black-Right-Pointing-Pointer Results are discussed in view of an indicator of fatigue damage initiation. Black-Right-Pointing-Pointer This indicator could allow the prediction of residual life before macrocracking. - Abstract: Steel pipes used in the oil and gas industry undergo the action of cyclic loads that can cause their failure by fatigue. A consistent evaluation of the fatigue damage during the initiation phase should fundamentally be based on a nanoscale approach, i.e., at the scale of the dislocation network, in order to take into account the micromechanisms of fatigue damage that precede macrocrack initiation and propagation until the final fracture. In this work, microstructural changes related to fatigue damage initiation are investigated in the API 5L X60 grade steel, used in pipe manufacturing. Microdeformations and macro residual stress are evaluated using X-ray diffraction in real time during alternating bending fatigue tests performed on samples cut off from an X60 steel pipe. The aim of this ongoing work is to provide ground for further development of an indicator of fatigue damage initiation in X60 steel. This damage indicator could allow a good residual life prediction of steel pipes previously submitted to fatigue loading, before macroscopic cracking, and help to increase the reliability of these structures.

An overview of environmental degradation of materials in nuclear power plant piping systems

International Nuclear Information System (INIS)

Shack, W.J.

1987-08-01

Piping in light water reactor (LWR) power systems is affected by several types of environmental degradation: intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping in boiling water reactors (BWRs) has required research, inspection, and mitigation programs that will ultimately cost several billion dollars; erosion-corrosion of carbon steel piping has been observed frequently in the secondary systems of both BWRs and pressurized water reactors (PWRs); the effect of the BWR environment can greatly diminish the design margin inherent in the ASME Section III fatigue design curves for carbon steel piping; and cast stainless steels are subject to embrittlement after extended thermal aging at reactor operating temperatures. These problems are being addressed by wide-ranging research programs in this country and abroad. The purpose of this review is to highlight some of the accomplishments of these programs and to note some of the remaining unanswered questions

Piping engineering for nuclear power plant

International Nuclear Information System (INIS)

Curto, N.; Schmidt, H.; Muller, R.

1988-01-01

In order to develop piping engineering, an adequate dimensioning and correct selection of materials must be secured. A correct selection of materials together with calculations and stress analysis must be carried out with a view to minimizing or avoiding possible failures or damages in piping assembling, which could be caused by internal pressure, weight, temperature, oscillation, etc. The piping project for a nuclear power plant is divided into the following three phases. Phase I: Basic piping design. Phase II: Final piping design. Phase III: Detail engineering. (Author)

A simplified leak-before-break evaluation procedure for austenitic and ferritic steel piping

Energy Technology Data Exchange (ETDEWEB)

Gamble, R.M.; Zahoor, A.; Ghassemi, B. [NOVETECH Corp., Rockville, MD (United States)

1994-10-01

A simplified procedure has been defined for computing the allowable circumferential throughwall crack length as a function of applied loads in piping. This procedure has been defined to enable leak-before-break (LBB) evaluations to be performed without complex and time consuming analyses. The development of the LBB evaluation procedure is similar to that now used in Section 11 of the ASME Code for evaluation of part-throughwall flaws found in piping. The LBB evaluation procedure was bench marked using experimental data obtained from pipes having circumferential throughwall flaws. Comparisons of the experimental and predicted load carrying capacities indicate that the method has a conservative bias, such that for at least 97% of the experiments the experimental load is equal to or greater than 90% of the predicted load. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austenitic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials.

Equipment for inspection of austenitic stainless steel pipe welds

International Nuclear Information System (INIS)

Boehmer, W.D.; Horn, J.E.

1979-01-01

A computer controlled ultrasonic scanning system and a data acquisition and analysis system have been developed to perform the inservice inspection of welds in stainless steel sodium piping in the Fast Flux Test Facility. The scanning equipment consists of a six axis motion mechanism and control system which allows full articulation of an ultrasonic transducer as it follows the circumferential pipe welds. The data acquisition and analysis system consists of high speed ultrasonic waveform digitizing equipment, dedicated processors to perform on-line analysis, and data storage and display equipment

Nuclear piping and pipe support design and operability relating to loadings and small bore piping

International Nuclear Information System (INIS)

Stout, D.H.; Tubbs, J.M.; Callaway, W.O.; Tang, H.T.; Van Duyne, D.A.

1994-01-01

The present nuclear piping system design practices for loadings, multiple support design and small bore piping evaluation are overly conservative. The paper discusses the results developed for realistic definitions of loadings and loading combinations with methodology for combining loads under various conditions for supports and multiple support design. The paper also discusses a simplified method developed for performing deadweight and thermal evaluations of small bore piping systems. Although the simplified method is oriented towards the qualification of piping in older plants, this approach is applicable to plants designed to any edition of the ASME Section III or B31.1 piping codes

Application of tearing modulus stability concepts to nuclear piping. Final report

International Nuclear Information System (INIS)

Cotter, K.H.; Chang, H.Y.; Zahoor, A.

1982-02-01

The recently developed tearing modulus stability concept was successfully applied to several boiling water reactor (BWR) and pressurized water reactor (PWR) piping systems. Circumferentially oriented through-the-thickness cracks were postulated at numerous locations in each system. For each location, the simplified tearing stability methods developed in USNRC Report NUREG/CR-0838 were used to determine crack stability. The J-T diagram was used to present the results of the computations. The piping systems considered included Type 304 stainless steel as well as A106 carbon steel materials. These systems were analyzed using the piping analysis computer code MINK

Development of support system for nuclear power plant piping

International Nuclear Information System (INIS)

Horino, Satoshi

1987-01-01

Ishikawajima-Harima Heavy Industries Co., Ltd. has advanced the development of Integrated Nuclear Plant Piping System (INUPPS) for nuclear power plants since 1980, and continued its improvement up to now. This time as its component, a piping support system (PISUP) has been developed. The piping support system deals with the structures such as piping supports and the stands for maintenance and inspection, and as for standard supporting structures, it builds up automatically the structures including the selection of optimum members by utilizing the standard patterns in cooperation with the piping design system including piping stress analysis. As for the supporting structures deviating from the standard, by amending a part of the standard patterns in dialogue from, structures can be built up. By using the data produced in this way, this system draws up consistently a design book, production management data and so on. From the viewpoint of safety, particular consideration is given to the aseismatic capability of nuclear power plants, and piping is fundamentally designed regidly to avoid resonance. It is necessary to make piping supports so as to have sufficient strength and rigidity. The features of the design of piping supports for nuclear power plant, the basic concept of piping support system, the constitution of the software and hardware, the standard patterns and the structural patterns of piping support system and so on are described. (Kako, I.)

Piping support load data base for nuclear plants

International Nuclear Information System (INIS)

Childress, G.G.

1991-01-01

Nuclear Station Modifications are continuous through the life of a Nuclear Power Plant. The NSM often impacts an existing piping system and its supports. Prior to implementation of the NSM, the modified piping system is qualified and the qualification documented. This manual review process is tedious and an obvious bottleneck to engineering productivity. Collectively, over 100,000 piping supports exist at Duke Power Company's Nuclear Stations. Engineering support must maintain proper documentation of all data for each support. Duke Power Company has designed and developed a mainframe based system that: directly uses Support Load Summary data generated by a piping analysis computer program; streamlines the pipe support evaluation process; easily retrieves As-Built and NSM information for any pipe support from an NSM or AS-BUILT data base; and generated documentation for easy traceability of data to the information source. This paper discusses the design considerations for development of Support Loads Database System (SLDB) and reviews the program functionality through the user menus

75 FR 27987 - Certain Welded Stainless Steel Pipes From the Republic of Korea: Final Results of Antidumping...

Science.gov (United States)

2010-05-19

... Republic of Korea (Korea). This review covers one producer/exporter of the subject merchandise to the... Steel Pipes From the Republic of Korea: Final Results of Antidumping Duty Administrative Review AGENCY... WSSP from Korea. See Certain Welded Stainless Steel Pipes from the Republic of Korea: Preliminary...

76 FR 68208 - Circular Welded Carbon-Quality Steel Pipe From India, Oman, United Arab Emirates, and Vietnam...

Science.gov (United States)

2011-11-03

... (Preliminary)] Circular Welded Carbon-Quality Steel Pipe From India, Oman, United Arab Emirates, and Vietnam... carbon-quality steel pipe from India, Oman, United Arab Emirates, and Vietnam, provided for in... Governments of India, Oman, United Arab Emirates, and Vietnam. Unless the Department of Commerce extends the...

Fracture assessment of Savannah River Reactor carbon steel piping

International Nuclear Information System (INIS)

Mertz, G.E.; Stoner, K.J.; Caskey, G.R.; Begley, J.A.

1991-01-01

The Savannah River Site (SRS) production reactors have been in operation since the mid-1950's. One postulated failure mechanism for the reactor piping is brittle fracture of the original A285 and A53 carbon steel piping. Material testing of archival piping determined (1) the static and dynamic tensile properties; (2) Charpy impact toughness; and (3) the static and dynamic compact tension fracture toughness properties. The nil-ductility transition temperature (NDTT), determined by Charpy impact test, is above the minimum operating temperature for some of the piping materials. A fracture assessment was performed to demonstrate that potential flaws are stable under upset loading conditions and minimum operating temperatures. A review of potential degradation mechanisms and plant operating history identified weld defects as the most likely crack initiation site for brittle fracture. Piping weld defects, as characterized by radiographic and metallographic examination, and low fracture toughness material properties were postulated at high stress locations in the piping. Normal operating loads, upset loads, and residual stresses were assumed to act on the postulated flaws. Calculated allowable flaw lengths exceed the size of observed weld defects, indicating adequate margins of safety against brittle fracture. Thus, a detailed fracture assessment was able to demonstrate that the piping systems will not fail by brittle fracture, even though the NDTT for some of the piping is above the minimum system operating temperature

Laser cutting of thick steel plates and simulated steel components using a 30 kW fiber laser

International Nuclear Information System (INIS)

Tamura, Koji; Ishigami, Ryoya; Yamagishi, Ryuichiro

2016-01-01

Laser cutting of thick steel plates and simulated steel components using a 30 kW fiber laser was studied for application to nuclear decommissioning. Successful cutting of carbon steel and stainless steel plates up to 300 mm in thickness was demonstrated, as was that of thick steel components such as simulated reactor vessel walls, a large pipe, and a gate valve. The results indicate that laser cutting applied to nuclear decommissioning is a promising technology. (author)

Piping equipment; Materiel petrole

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This 'blue bible' of the perfect piping-man appeals to end-users of industrial facilities of the petroleum and chemical industries (purchase services, standardization, new works, maintenance) but also to pipe-makers and hollow-ware makers. It describes the characteristics of materials (carbon steels, stainless steels, alloyed steels, special alloys) and the dimensions of pipe elements: pipes, welding fittings, flanges, sealing products, forged steel fittings, forged steel valves, cast steel valves, ASTM standards, industrial valves. (J.S.)

Heat Transfer Modeling of an Annular On-Line Spray Water Cooling Process for Electric-Resistance-Welded Steel Pipe.

Science.gov (United States)

Chen, Zejun; Han, Huiquan; Ren, Wei; Huang, Guangjie

2015-01-01

On-line spray water cooling (OSWC) of electric-resistance-welded (ERW) steel pipes can replace the conventional off-line heat treatment process and become an important and critical procedure. The OSWC process improves production efficiency, decreases costs, and enhances the mechanical properties of ERW steel pipe, especially the impact properties of the weld joint. In this paper, an annular OSWC process is investigated based on an experimental simulation platform that can obtain precise real-time measurements of the temperature of the pipe, the water pressure and flux, etc. The effects of the modes of annular spray water cooling and related cooling parameters on the mechanical properties of the pipe are investigated. The temperature evolutions of the inner and outer walls of the pipe are measured during the spray water cooling process, and the uniformity of mechanical properties along the circumferential and longitudinal directions is investigated. A heat transfer coefficient model of spray water cooling is developed based on measured temperature data in conjunction with simulation using the finite element method. Industrial tests prove the validity of the heat transfer model of a steel pipe undergoing spray water cooling. The research results can provide a basis for the industrial application of the OSWC process in the production of ERW steel pipes.

Heat Transfer Modeling of an Annular On-Line Spray Water Cooling Process for Electric-Resistance-Welded Steel Pipe

Science.gov (United States)

Chen, Zejun; Han, Huiquan; Ren, Wei; Huang, Guangjie

2015-01-01

On-line spray water cooling (OSWC) of electric-resistance-welded (ERW) steel pipes can replace the conventional off-line heat treatment process and become an important and critical procedure. The OSWC process improves production efficiency, decreases costs, and enhances the mechanical properties of ERW steel pipe, especially the impact properties of the weld joint. In this paper, an annular OSWC process is investigated based on an experimental simulation platform that can obtain precise real-time measurements of the temperature of the pipe, the water pressure and flux, etc. The effects of the modes of annular spray water cooling and related cooling parameters on the mechanical properties of the pipe are investigated. The temperature evolutions of the inner and outer walls of the pipe are measured during the spray water cooling process, and the uniformity of mechanical properties along the circumferential and longitudinal directions is investigated. A heat transfer coefficient model of spray water cooling is developed based on measured temperature data in conjunction with simulation using the finite element method. Industrial tests prove the validity of the heat transfer model of a steel pipe undergoing spray water cooling. The research results can provide a basis for the industrial application of the OSWC process in the production of ERW steel pipes. PMID:26201073

AWWA C303-17 concrete pressure pipe, bar-wrapped, steel-cylinder type

CERN Document Server

2017-01-01

This standard describes the manufacture of concrete pressure pipe, reinforced with a steel cylinder that is helically wrapped with mild steel bar reinforcement, in sizes ranging from 10 in. through 72 in. (250 mm through 1,830 mm), inclusive, and for working pressures up to 400 psi (2,760 kPa).

77 FR 73015 - Circular Welded Non-Alloy Steel Pipe From the Republic of Korea: Preliminary Results of...

Science.gov (United States)

2012-12-07

... Steel Pipe From the Republic of Korea: Preliminary Results of Antidumping Duty Administrative Review... antidumping duty order on circular welded non-alloy steel pipe (CWP) from the Republic of Korea (Korea). The... preliminarily found that one respondent has made sales of the subject merchandise at prices below normal value...

Piping systems, containment pre-stressing and steel ventilation chimney

International Nuclear Information System (INIS)

Stuessi, U.

1996-01-01

Units 5 and 6 of NPP Kozloduy have been designed initially for seismic levels which are considered too low today. In the frame of an IAEA Coordinated Research Programme, a Swiss team has been commissioned by Natsionalna Elektricheska Kompania, Sofia, to analyse the relevant piping system, the containment prestressing and the steel ventilation chimney and to recommend upgrade measures for adequate seismic capacity where applicable. Seismic input had been specified by and agreed upon earlier by IAEA experts. The necessary investigations have been performed in 1995 and discussed with internationally recognized experts. The main results may be summarized as follows: Upgrades are necessary at different piping sy ports (additional snubbers or viscous dampers). These fixes can be done easily at low cost. The containment prestressing tendons are adequately designed for the specified load combinations. However, unfavourable construction features endanger the reliability. It is therefore strongly recommended to replace the tendons stepwise and to upgrade the existing monitoring system. Finally, the steel ventilation chimney may not withstand a seismic event, however the containment and diesel generator building will not be destroyed at possible impact by the chimney. On the other hand the roof of the main building has to be reinforced partially. It is recommended to continue the project for 1996 and 1997 to implement the upgrade measures mentioned above, to analyse the remaining piping systems and to consolidate all results obtained by different research groups of the IAEA programme with respect to piping systems including components and tanks

78 FR 21105 - Circular Welded Carbon Steel Pipes and Tubes From Thailand: Preliminary Results of Antidumping...

Science.gov (United States)

2013-04-09

.../exporters of the subject merchandise, Saha Thai Steel Pipe (Public) Company, Ltd. (Saha Thai), and Pacific Pipe Company Limited (Pacific Pipe). The period of review (POR) is March 1, 2011, through February 29...) Company, Ltd 0.00 Pacific Pipe Company Limited * * No shipments or sales subject to this review. The firm...

CT image reconstruction of steel pipe section from few projections using the method of rotating polar-coordinate

International Nuclear Information System (INIS)

Peng Shuaijun; Wu Zhifang

2008-01-01

Fast online inspection in steel pipe production is a big challenge. Radiographic CT imaging technology, a high performance non-destructive testing method, is quite appropriate for inspection and quality control of steel pipes. The method of rotating polar-coordinate is used to reconstruct the steel pipe section from few projections with the purpose of inspecting it online. It reduces the projection number needed and the data collection time, and accelerates the reconstruction algorithm and saves the inspection time evidently. The results of simulation experiment and actual experiment indicate that the image quality and reconstruction time of rotating polar-coordinate method meet the requirements of inspecting the steel tube section online basically. The study is of some theoretical significance and the method is expected to be widely used in practice. (authors)

Survey of heat-pipe application under nuclear environment

International Nuclear Information System (INIS)

Tsuyuzaki, Noriyoshi; Saito, Takashi; Okamoto, Yoshizo; Hishida, Makoto; Negishi, Kanji.

1986-11-01

Heat pipes today are employed in a wide variety of special heat transfer applications including nuclear reactor. In this nuclear technology area in Japan, A headway speed of the heat pipe application technique is not so high because of safety confirmation and investigation under each developing step. Especially, the outline of space craft is a tendency to increase the size. Therefore, the power supply is also tendency to increase the outlet power and keep the long life. Under SP-100 project, the development of nuclear power supply system which power is 1400 - 1600 KW thermal and 100 KW electric power is steadily in progress. Many heat pipes are adopted for thermionic conversion and coolant system in order to construct more safety and light weight system for the project. This paper describes the survey of the heat pipe applications under the present and future condition for nuclear environment. (author)

Non-metallic structural wrap systems for pipe

International Nuclear Information System (INIS)

Walker, R.H.; Wesley Rowley, C.

2001-01-01

The use of thermoplastics and reinforcing fiber has been a long-term application of non-metallic material for structural applications. With the advent of specialized epoxies and carbon reinforcing fiber, structural strength approaching and surpassing steel has been used in a wide variety of applications, including nuclear power plants. One of those applications is a NSWS for pipe and other structural members. The NSWS is system of integrating epoxies with reinforcing fiber in a wrapped geometrical configuration. This paper specifically addresses the repair of degraded pipe in heat removal systems used in nuclear power plants, which is typically caused by corrosion, erosion, or abrasion. Loss of structural material leads to leaks, which can be arrested by a NSWS for the pipe. The technical aspects of using thermoplastics to structurally improve degraded pipe in nuclear power plants has been addressed in the ASME B and PV Code Case N-589. Using the fundamentals described in that Code Case, this paper shows how this technology can be extended to pipe repair from the outside. This NSWS has already been used extensively in non-nuclear applications and in one nuclear application. The cost to apply this NSWS is typically substantially less than replacing the pipe and may be technically superior to replacing the pipe. (author)

Effects of toughness anisotropy and combined tension, torsion, and bending loads on fracture behavior of ferritic nuclear pipe

International Nuclear Information System (INIS)

Mohan, R.; Marschall, C.; Krishnaswamy, P.; Brust, F.; Ghadiali, N.; Wilkowski, G.

1995-04-01

This topical report summarizes the work on angled crack growth and combined loading effects performed within the Nuclear Regulatory Commission's research program entitled open-quotes Short Cracks in Piping and Piping Weldsclose quotes. The major impetus for this work stemmed from the observation that initial circumferential cracks in carbon steel pipes exhibited angular crack growth. This failure mode was little understood, and the effect of angled crack growth from an initially circumferential crack raised questions of how pipes under combined loading with torsional stresses would behave. There were three major conclusions from this work. The first was that virtually all ferritic nuclear pipes will have toughness anisotropy. The second was that the ratio of the normalized crack driving force (as a function of angle) to the normalized toughness (also as a function of the angle of crack growth) showed that there was an equal likelihood of cracks growing at any angle between 25 and 65 degrees. This agreed with the scatter of crack growth angles observed in pipe tests. Third, for combined loads with torsional stresses, an effective moment allows pure bending analyses to be used up to crack initiation. Crack opening area under combined loads could also be determined in this mariner

Effects of toughness anisotropy and combined tension, torsion, and bending loads on fracture behavior of ferritic nuclear pipe

Energy Technology Data Exchange (ETDEWEB)

Mohan, R.; Marschall, C.; Krishnaswamy, P.; Brust, F.; Ghadiali, N.; Wilkowski, G. [Battelle, Columbus, OH (United States)

1995-04-01

This topical report summarizes the work on angled crack growth and combined loading effects performed within the Nuclear Regulatory Commission`s research program entitled {open_quotes}Short Cracks in Piping and Piping Welds{close_quotes}. The major impetus for this work stemmed from the observation that initial circumferential cracks in carbon steel pipes exhibited angular crack growth. This failure mode was little understood, and the effect of angled crack growth from an initially circumferential crack raised questions of how pipes under combined loading with torsional stresses would behave. There were three major conclusions from this work. The first was that virtually all ferritic nuclear pipes will have toughness anisotropy. The second was that the ratio of the normalized crack driving force (as a function of angle) to the normalized toughness (also as a function of the angle of crack growth) showed that there was an equal likelihood of cracks growing at any angle between 25 and 65 degrees. This agreed with the scatter of crack growth angles observed in pipe tests. Third, for combined loads with torsional stresses, an effective moment allows pure bending analyses to be used up to crack initiation. Crack opening area under combined loads could also be determined in this mariner.

Effects of toughness anisotropy and combined tension, torsion, and bending loads on fracture behavior of ferritic nuclear pipe

Energy Technology Data Exchange (ETDEWEB)

Mohan, R.; Marshall, C.; Ghadiali, N.; Wilkowski, G. [Battelle, Columbus, OH (United States)

1997-04-01

This paper summarizes work on angled through-wall-crack initiation and combined loading effects on ferritic nuclear pipe performed as part of the Nuclear Regulatory Commission`s research program entitled {open_quotes}Short Cracks In Piping an Piping Welds{close_quotes}. The reader is referred to Reference 1 for details of the experiments and analyses conducted as part of this program. The major impetus for this work stemmed from the observation that initially circumferentially oriented cracks in carbon steel pipes exhibited a high tendency to grow at a different angle when the cracked pipes were subjected to bending or bending plus pressure loads. This failure mode was little understood, and the effect of angled crack grown from an initially circumferential crack raised questions about how cracks in a piping system subjected to combined loading with torsional stresses would behave. There were three major efforts undertaken in this study. The first involved a literature review to assess the causes of toughness anisotropy in ferritic pipes and to develop strength and toughness data as a function of angle from the circumferential plane. The second effort was an attempt to develop a screening criterion based on toughness anisotropy and to compare this screening criterion with experimental pipe fracture data. The third and more significant effort involved finite element analyses to examine why cracks grow at an angle and what is the effect of combined loads with torsional stresses on a circumferentially cracked pipe. These three efforts are summarized.

77 FR 15718 - Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab...

Science.gov (United States)

2012-03-16

...-811] Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab... Oman (Oman), the United Arab Emirates (UAE), and the Socialist Republic of Vietnam (Vietnam). See Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab Emirates, and...

77 FR 73674 - Circular Welded Carbon-Quality Steel Pipe From India, Oman, The United Arab Emirates, and Vietnam

Science.gov (United States)

2012-12-11

...)] Circular Welded Carbon-Quality Steel Pipe From India, Oman, The United Arab Emirates, and Vietnam... welded carbon-quality steel pipe from India, Oman, the United Arab Emirates, and Vietnam, provided for in... from India, Oman, the United Arab Emirates, and Vietnam were subsidized and/or dumped within the...

76 FR 78313 - Circular Welded Carbon-Quality Steel Pipe From India, Oman, the United Arab Emirates, and Vietnam

Science.gov (United States)

2011-12-16

... (Preliminary)] Circular Welded Carbon-Quality Steel Pipe From India, Oman, the United Arab Emirates, and... India, Oman, the United Arab Emirates, and Vietnam of circular welded carbon- quality steel pipe... the Governments of India, Oman, the United Arab Emirates, and Vietnam.\\2\\ \\1\\ The record is defined in...

77 FR 37711 - Circular Welded Carbon-Quality Steel Pipe From India, Oman, the United Arab Emirates, and Vietnam...

Science.gov (United States)

2012-06-22

...)] Circular Welded Carbon-Quality Steel Pipe From India, Oman, the United Arab Emirates, and Vietnam...-fair-value imports from India, Oman, the United Arab Emirates, and Vietnam of circular welded carbon... respect to circular welded carbon-quality steel pipe from Oman and the United Arab Emirates being sold in...

75 FR 44763 - Certain Circular Welded Non-Alloy Steel Pipe From Mexico; Extension of Time Limit for Preliminary...

Science.gov (United States)

2010-07-29

...-Alloy Steel Pipe From Mexico; Extension of Time Limit for Preliminary Results of Antidumping Duty... review of the antidumping duty order on certain circular welded non- alloy steel pipe from Mexico. We... preliminary results of this review within the original time frame because we require additional time with...

Loading rate effects on strength and fracture toughness of pipe steels used in Task 1 of the IPIRG program

International Nuclear Information System (INIS)

Marschall, C.W.; Landow, M.P.; Wilkowski, G.M.

1993-10-01

Material characterization tests were conducted on laboratory specimens machined from pipes to determine the effect of dynamic loading (i.e., rates comparable to those for high amplitude seismic events) on tensile properties and fracture resistance at 288 C (550 F). Specimens were fabricated from seven different pipes, including carbon steels and stainless steels (both base metal and weld metal), which were to be subjected to full-scale pipe tests in IPIRG Task 1.0. For the stainless steels tested at 288 C (550 F), tensile strength was unchanged, while yield strength and fracture resistance were increased. The increase in fracture resistance was modest for the wrought base metals and substantial for the weld metal and the cast base metal. The carbon steels tested were sensitive to dynamic strain aging, and hence the strength and toughness was affected by both temperature and strain rate effects. The carbon steel base metal and welds exhibited ultimate tensile strength values at 288 C (550 F) that were greater than at room temperature. Furthermore, the ultimate tensile strength at 288 C (550 F) was lowered significantly by increased strain rate and, in the carbon steel base metals, increased strain rate also lowered the fracture resistance, substantially in the base metal of one pipe. In comparing these results to the IPIRG pipe test results to date, it was found that the trends of these tests agree well with the Subtask 1.2 quasi-static and dynamic pipe fracture experiments. Loads measured in the Subtask 1.1 pipe experiments were, however, somewhat higher than would have been expected by the trends observed in the laboratory tests

APPLICATION OF STEEL PIPE PILE LOADING TESTS TO DESIGN VERIFICATION OF FOUNDATION OF THE TOKYO GATE BRIDGE

Science.gov (United States)

Saitou, Yutaka; Kikuchi, Yoshiaki; Kusakabe, Osamu; Kiyomiya, Osamu; Yoneyama, Haruo; Kawakami, Taiji

Steel sheet pipe pile foundations with large diameter steel pipe sheet pile were used for the foundation of the main pier of the Tokyo Gateway bridge. However, as for the large diameter steel pipe pile, the bearing mechanism including a pile tip plugging effect is still unclear due to lack of the practical examinations even though loading tests are performed on Trans-Tokyo Bay Highway. In the light of the foregoing problems, static pile loading tests both vertical and horizontal directions, a dynamic loading test, and cone penetration tests we re conducted for determining proper design parameters of the ground for the foundations. Design parameters were determined rationally based on the tests results. Rational design verification was obtained from this research.

Effect of Structure Factor on High-Temperature Ductility of Pipe Steels

Science.gov (United States)

Kolbasnikov, N. G.; Matveev, M. A.; Mishnev, P. A.

2016-05-01

Effects of various factors such as the grain size, the morphology of nonmetallic inclusions, and joint microalloying with boron and titanium on the high-temperature ductility of pipe steels are studied. Physical modeling of the conditions of cooling of the skin of a continuous-cast preform in the zone of secondary cooling in a Gleeble facility is performed. Technical recommendations are given for raising the hot ductility of steels under industrial conditions.

Effect of Pipe Flattening in API X65 Linepipe Steels Having Bainite vs. Ferrite/Pearlite Microstructures

Directory of Open Access Journals (Sweden)

Singon Kang

2018-05-01

Full Text Available The influence of microstructure on pipe flattening response was assessed using two different commercially produced U-ing, O-ing, and expansion (UOE pipes from API X65 steels having either a bainitic microstructure (steel B or a ferrite/pearlite microstructure (steel FP. A four-point bending apparatus and distinctive procedure were used to minimize strain localization during flattening. The flattened specimens were sectioned at different positions through the thickness, and tensile tested in both the longitudinal (LD and transverse directions (TD to assess the through-thickness variation in properties. Yield strength (YS distributions in the LD show V-shaped profiles through thickness in both steels, whereas the YS in the TD nearest the outside diameter (OD surface is reduced. These variations in YS are due to the Bauschinger effect associated with the compressive flattening pre-strain. The uniform elongation (UE of steel FP is almost independent of specimen position through the thickness, but for steel B there is a substantial reduction of the UE at both the inside and outside diameter positions and this reduction is greater in the LD. This work confirms that flattened pipe mechanical properties exhibit an important dependence on their microstructure type and it is postulated that the flattening procedure also influences the mechanical properties.

75 FR 4529 - Circular Welded Carbon Steel Pipes and Tubes From Thailand: Final Results of Antidumping Duty New...

Science.gov (United States)

2010-01-28

... Compliance Analyst, Office 6, Verification of the Sales Response of Pacific Pipe Public Company, Limited in... Pipe Public Company, Limited, dated August 24, 2009 (Bona Fides Preliminary Memorandum). The Department... Circular Welded Carbon Steel Pipes and Tubes from Thailand: Pacific Pipe Public Company, Limited, dated...

Residual stress measurement in 304 stainless steel weld overlay pipes

International Nuclear Information System (INIS)

Yen, H.J.; Lin, M.C.C.; Chen, L.J.

1996-01-01

Welding overlay repair (WOR) is commonly employed to rebuild piping systems suffering from intergranular stress corrosion cracking (IGSCC). To understand the effects of this repair, it is necessary to investigate the distribution of residual stresses in the welding pipe. The overlay welding technique must induce compressive residual stress at the inner surface of the welded pipe to prevent IGSCC. To understand the bulk residual stress distribution, the stress profile as a function of location within wall is examined. In this study the full destructive residual stress measurement technique -- a cutting and sectioning method -- is used to determine the residual stress distribution. The sample is type 304 stainless steel weld overlay pipe with an outside diameter of 267 mm. A pipe segment is cut from the circular pipe; then a thin layer is removed axially from the inner to the outer surfaces until further sectioning is impractical. The total residual stress is calculated by adding the stress relieved by cutting the section away to the stress relieved by axially sectioning. The axial and hoop residual stresses are compressive at the inner surface of the weld overlay pipe. Compressive stress exists not only at the surface but is also distributed over most of the pipe's cross section. On the one hand, the maximum compressive hoop residual stress appears at the pipe's inner surface. The thermal-mechanical induced crack closure from significant compressive residual stress is discussed. This crack closure can thus prevent IGSCC very effectively

Evaluation of LBB margin of nuclear piping systems

International Nuclear Information System (INIS)

Hwang, Il Soon; Kim, Ji Hyeon; Oh, Yeong Jin; Lim, Jun; Kim, In Seob; Kim, Yong Seon; Lee, Joo Seok

1999-04-01

Most of previous elastic-plastic fracture studies for LBB assessment of low alloy steel piping have been focused on base metals and weld metals. In contract, the heat affected zone of welded pipe has not been studied in detail primarily because the size of heat affected zone in welded pipe os too small to make specimens for mechanical properties measurement. When structural members are joined by welding, the base metal is heated to its melting point and then cooled rapidly. As a result of this very severe thermal cycle, mechanical properties in the heat affected zone can be degraded by grain coarsening, the precipitation and the segregation of trace impurities. In this study, a thermal and microstructural analysis is performed, and mechanical properties are measured for the weld heat affected zone of SA106Gr.C low allowed piping steel. In addition, inter critical annealing treatment. in two-phase (alpha+gamma) region was performed to investigate the possibilities of improving the toughness and reducing dynamic strain aging (DSA) susceptibility for giving allowable LBB safety margins. From the results, intercritical annealing is shown to give a smaller ductility loss due to DSA than the case of as-received material. Furthermore, the intercritical annealing was able to increase the impact toughness by a factor of 1.5 compared to the as-received material

Evaluation of LBB margin of nuclear piping systems

Energy Technology Data Exchange (ETDEWEB)

Hwang, Il Soon; Kim, Ji Hyeon; Oh, Yeong Jin; Lim, Jun [Seoul Nationl Univ., Seoul (Korea, Republic of); Kim, In Seob; Kim, Yong Seon; Lee, Joo Seok [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-04-15

Most of previous elastic-plastic fracture studies for LBB assessment of low alloy steel piping have been focused on base metals and weld metals. In contract, the heat affected zone of welded pipe has not been studied in detail primarily because the size of heat affected zone in welded pipe os too small to make specimens for mechanical properties measurement. When structural members are joined by welding, the base metal is heated to its melting point and then cooled rapidly. As a result of this very severe thermal cycle, mechanical properties in the heat affected zone can be degraded by grain coarsening, the precipitation and the segregation of trace impurities. In this study, a thermal and microstructural analysis is performed, and mechanical properties are measured for the weld heat affected zone of SA106Gr.C low allowed piping steel. In addition, inter critical annealing treatment. in two-phase (alpha+gamma) region was performed to investigate the possibilities of improving the toughness and reducing dynamic strain aging (DSA) susceptibility for giving allowable LBB safety margins. From the results, intercritical annealing is shown to give a smaller ductility loss due to DSA than the case of as-received material. Furthermore, the intercritical annealing was able to increase the impact toughness by a factor of 1.5 compared to the as-received material.

Evaluation of weld defects in stainless steel 316L pipe using guided wave

Energy Technology Data Exchange (ETDEWEB)

Lee, Joon Hyun [School of Mechanical Engineering, Pusan National University, Busan (Korea, Republic of); Lee, Jin Kyung [Dept. of Mechanical Engineering, Dongeui University, Busan (Korea, Republic of)

2015-02-15

Stainless steel is a popular structural materials for liquid-hydrogen storage containers and piping components for transporting high-temperature fluids because of its superior material properties such as high strength and high corrosion resistance at elevated temperatures. In general, tungsten inert gas (TIG) arc welding is used for bonding stainless steel. However, it is often reported that the thermal fatigue cracks or initial defects in stainless steel after welding decreases the reliability of the material. The objective of this paper is to clarify the characteristics of ultrasonic guided wave propagation in relation to a change in the initial crack length in the welding zone of stainless steel. For this purpose, three specimens with different artificial defects of 5 mm, 10 mm, and 20 mm in stainless steel welds were prepared. By considering the thickness of s stainless steel pipe, special attention was given to both the L(0,1) mode and L(0,2) mode in this study. It was clearly found that the L(0,2) mode was more sensitive to defects than the L(0,1) mode. Based on the results of the L(0,1) and L(0,2) mode analyses, the magnitude ratio of the two modes was more effective than studying each mode when evaluating defects near the welded zone of stainless steel because of its linear relationship with the length of the artificial defect.

Effect of vacuum arc melting/casting parameters on shrinkage cavity/piping of austenitic stainless steel ingot

International Nuclear Information System (INIS)

Kamran, J.; Feroz, M.; Sarwar, M.

2009-01-01

Shrinkage cavity/piping at the end of the solidified ingot of steels is one of the most common casting problem in 316L austenitic stainless steel ingot, when consumable electrode is melted and cast in a water-cooled copper mould by vacuum arc re-melting furnace. In present study an effort has been made to reduce the size of shrinkage cavity/ piping by establishing the optimum value of hot topping process parameters at the end of the melting process. It is concluded that the shrinkage cavity/piping at the top of the solidified ingot can be reduced to minimum by adjusting the process parameters particularly the melting current density. (author)

A study on ultrasonic inspection of long steel pipes using lamb waves

International Nuclear Information System (INIS)

Park, Moon Ho

1996-02-01

An ultrasonic inspection technique with use of Lamb waves was evaluated to detect and determine the exact location of flaws present in long steel pipes. Since multiple modes of Lamb waves are generated in the inspected pipes due to their dispersive characteristics, selection of a specific Lamb wave mode is very important for inspection of flaws. Experimental studies of flaw detectability with use of each Lamb wave mode, namely, A 0 , S 0 , A 1 , and S 1 mode and their ultrasonic attenuation characteristics were conducted. Experimental results showed that A 0 mode is the most effective for detection and exact determination of the location of flaws. A lucite wedge containing water column that generates the A 0 Lamb wave mode was developed and used in the present inspection study. It was found that the ultrasonic beam divergence after its wrapping around once the inspected pipe interferes with exact determination of the location of flaws and that maximum reflection signals are obtained when the transducer is located axially offset from the straight line with the position of the flaw. The present study showed feasibility of ultrasonic inspection with use of Lamb waves for detection of flaws in several meters long insulated or inaccessible steel pipes

Application of tearing modulus stability concepts to nuclear piping. Final report. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Cotter, K.H.; Chang, H.Y.; Zahoor, A.

1982-02-01

The recently developed tearing modulus stability concept was successfully applied to several boiling water reactor (BWR) and pressurized water reactor (PWR) piping systems. Circumferentially oriented through-the-thickness cracks were postulated at numerous locations in each system. For each location, the simplified tearing stability methods developed in USNRC Report NUREG/CR-0838 were used to determine crack stability. The J-T diagram was used to present the results of the computations. The piping systems considered included Type 304 stainless steel as well as A106 carbon steel materials. These systems were analyzed using the piping analysis computer code MINK.

Corrosion resistance of stainless steel pipes in soil

Energy Technology Data Exchange (ETDEWEB)

Sjoegren, L.; Camitz, G. [Swerea KIMAB AB, Box 55970, SE-102 16 Stockholm (Sweden); Peultier, J.; Jacques, S.; Baudu, V.; Barrau, F.; Chareyre, B. [Industeel and ArcelorMittal R and D, 56 rue Clemenceau, BP19, FR-71201 le Creusot, Cedex (France); Bergquist, A. [Outokumpu Stainless AB, P.O. Box 74, SE-774 22 Avesta (Sweden); Pourbaix, A.; Carpentiers, P. [Belgian Centre for Corrosion Study, Avenue des Petits-Champs 4A, BE 1410 Waterloo (Belgium)

2011-04-15

To be able to give safe recommendations concerning the choice of suitable stainless steel grades for pipelines to be buried in various soil environments, a large research programme, including field exposures of test specimens buried in soil in Sweden and in France, has been performed. Resistance against external corrosion of austenitic, super austenitic, lean duplex, duplex and super duplex steel grades in soil has been investigated by laboratory tests and field exposures. The grades included have been screened according to their critical pitting-corrosion temperature and according to their time-to-re-passivation after the passive layer has been destroyed locally by scratching. The field exposures programme, being the core of the investigation, uses large specimens: 2 m pipes and plates, of different grades. The exposure has been performed to reveal effects of aeration cells, deposits or confined areas, welds and burial depth. Additionally, investigations of the tendency of stainless steel to corrode under the influence of alternating current (AC) have been performed, both in the laboratory and in the field. Recommendations for use of stainless steels under different soil conditions are given based on experimental results and on operating experiences of existing stainless steel pipelines in soil. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Fracture analysis procedure for cast austenitic stainless steel pipe with an axial crack

International Nuclear Information System (INIS)

Kamaya, Masayuki

2012-01-01

Since the ductility of cast austenitic stainless steel pipes decreases due to thermal aging embrittlement after long term operation, not only plastic collapse failure but also unstable ductile crack propagation (elastic-plastic failure) should be taken into account for the structural integrity assessment of cracked pipes. In the fitness-for-service code of the Japan Society of Mechanical Engineers (JSME), Z-factor is used to incorporate the reduction in failure load due to elastic-plastic failure. However, the JSME code does not provide the Z-factor for axial cracks. In this study, Z-factor for axial cracks in aged cast austenitic stainless steel pipes was derived. Then, a comparison was made for the elastic-plastic failure load obtained from different analysis procedures. It was shown that the obtained Z-factor could derive reasonable elastic-plastic failure loads, although the failure loads were more conservative than those obtained by the two-parameter method. (author)

Evaluation of residual stresses for the multipass welds of 316L stainless steel pipe

International Nuclear Information System (INIS)

Kim, S. H.; Joo, Y. S.; Lee, J. H.

2003-01-01

It is necessary to evaluate the influence of the residual stress and distortion in the design and fabrication of welded structure and the sound welded structure can be maintained by this consideration. Multipass welds of the 316L stainless steel have been widely employed in the pipes of Liquid Metal Reactor. In this study, the residual stresses in the 316L stainless steel pipe welds were calculated by the finite element method using ANSYS code. Also, the residual stresses both on the surface and in the interior of the thickness were measured by HRPD(High Resolution Powder Diffractometer) instrumented in HANARO Reactor. The residual stresses were measured for each 18 points in small(t/d=0.075) and large pipe specimens (t/d=0.034). The experimental and calculated results were compared and the characteristics of the distribution of the residual stress discussed

76 FR 67673 - Welded ASTM A-312 Stainless Steel Pipe From South Korea and Taiwan: Final Results of Expedited...

Science.gov (United States)

2011-11-02

... DEPARTMENT OF COMMERCE International Trade Administration [A-580-810, A-583-815] Welded ASTM A-312... the antidumping duty orders on welded ASTM A-312 stainless steel pipe from South Korea and Taiwan... duty orders on welded ASTM A-312 stainless steel pipe from South Korea and Taiwan pursuant to section...

Damage mechanism of piping welded joints made from austenitic Steel for the type RBMK reactor

International Nuclear Information System (INIS)

Karzov, G.; Timofeev, B.; Gorbakony, A.; Petrov, V.; Chernaenko, T.

1999-01-01

In the process of operation of RBMK reactors the damages were taking place on welded piping, produced from austenitic stainless steel of the type 08X18H10T. The inspection of damaged sections in piping has shown that in most cases crack-like defects are of corrosion and mechanical character. The paper considers in details the reasons of damages appearance and their development for this type of welded joints of downcomers 325xl6 mm, which were fabricated from austenitic stainless steel using TlG and MAW welding methods. (author)

Characterization of bond line discontinuities in a high-Mn TWIP steel pipe welded by HF-ERW

Energy Technology Data Exchange (ETDEWEB)

Park, Gitae; Kim, Bongyoon; Kang, Yongjoon [Division of Materials Science and Engineering, Hanyang University, 222, Wangsimni-ro, Seongdong-gu, Seoul 04763 (Korea, Republic of); Kang, Heewoong [RD Team, Husteel, 131 Bugokgongdan-ro, Songak-eup, Dangjin-si, Chungnam 31721 (Korea, Republic of); Lee, Changhee, E-mail: chlee@hanyang.ac.kr [Division of Materials Science and Engineering, Hanyang University, 222, Wangsimni-ro, Seongdong-gu, Seoul 04763 (Korea, Republic of)

2016-08-15

In this work, the microstructure and defects in a high-frequency electrical resistance welded (HF-ERW) pipe of high-Mn twinning-induced plasticity (TWIP) steel were characterized. The microstructure of the base metal and the bond line were examined using both optical microscopy and scanning electron microscopy. The features of the bond line were similar to those of conventional steel. Simultaneously, the circumferential ductility was evaluated via a flaring test. It was concluded that the deterioration of the circumferential ductility in a high-Mn TWIP steel pipe was caused by irregular shaped oxide defects and a penetrator that had been formed during welding. Specifically, the penetrator, which is composed of MnO and Mn{sub 2}SiO{sub 4}, was found to be the most influential on the circumferential ductility of the welded pipe. The penetrator was analyzed using both an electron probe micro analyzer and transmission electron microscopy, and the formation sequence of the penetrator was evaluated. - Highlights: •This study focused on applying the HF-ERW process to the seam welding of expandable pipe using TWIP steels. •For improvement of the circumferential ductility, deterioration factors were characterized. •Penetrator which would mainly deteriorate the circumferential ductility consisted of round MnO and Mn{sub 2}SiO{sub 4}. •Metallurgical evidence of existing theory regarding the mechanism of defect formation during the HF-ERW was characterized.

Pipe rupture test results; 6 in. pipe whip test under BWR LOCA conditions

International Nuclear Information System (INIS)

Kurihara, Ryoichi; Yano, Toshikazu; Ueda, Shuzo; Isozaki, Toshikuni; Miyazaki, Noriyuki; Kato, Rokuro; Miyazono, Shohachiro

1983-02-01

A series of pipe rupture tests has been performed in JAERI to demonstrate the safety of the primary coolant circuits in the event of pipe rupture, in nuclear power plants. The present report summarizes the results of 6 in. pipe whip tests (RUN 5605, 5606), under BWR LOCA conditions (285 0 C, 6.8 MPa), which were performed in August, 1981. The test pipe is made of Type 304 stainless steel and its outer diameter is 6 in. and its thickness is 11.1 mm. The restraints are made of Type 304 stainless steel and its diameter is 16.0 mm. Two restraints were set on the restraint support with clearance of 100 mm. Overhang length was varied as the parameter in these tests and was 300 mm or 700 mm. The following results are obtained. (1) The deformations of a pipe and restraints are limited effectively by shorter overhang length of 300. However, they become larger when the overhang length is 700 mm, and the pipe deforms especially at the setting point of restraints. (2) Velocity at the free end of pipe becomes about 30 m/sec just after the break. However, velocity at the setting point of restraint becomes about only 4 m/sec just after the break. (3) It seems from the comparison between the 4 in. tests and 6 in. tests that the maximum restraint force of 6 in. tests is about two times as large as that of 4 in. tests. (author)

Pipe support optimization in nuclear power plants

International Nuclear Information System (INIS)

Cleveland, A.B.; Kalyanam, N.

1984-01-01

A typical 1000 MWe nuclear power plant consists of 80,000 to 100,000 feet of piping which must be designed to withstand earthquake shock. For the required ground motion, seismic response spectra are developed for safety-related structures. These curves are used in the dynamic analysis of piping systems with pipe-stress analysis computer codes. To satisfy applicable Code requirements, the piping systems also require analysis for weight, thermal and possibly other lasting conditions. Bechtel Power Corporation has developed a design program called SLAM (Support Location Algorithm) for optimizing pipe support locations and types (rigid, spring, snubber, axial, lateral, etc.) while satisfying userspecified parameters such as locations, load combinations, stress and load allowables, pipe displacement and cost. This paper describes SLAM, its features, applications and benefits

Microbiological corrosion of ASTM SA105 carbon steel pipe for industrial fire water usage

Science.gov (United States)

Chidambaram, S.; Ashok, K.; Karthik, V.; Venkatakrishnan, P. G.

2018-02-01

The large number of metallic systems developed for last few decades against both general uniform corrosion and localized corrosion. Among all microbiological induced corrosion (MIC) is attractive, multidisciplinary and complex in nature. Many chemical processing industries utilizes fresh water for fire service to nullify major/minor fire. One such fire water service line pipe attacked by micro-organisms leads to leakage which is industrially important from safety point of view. Also large numbers of leakage reported in similar fire water service of nearby food processing plant, paper & pulp plant, steel plant, electricity board etc…In present investigation one such industrial fire water service line failure analysis of carbon steel line pipe was analyzed to determine the cause of failure. The water sample subjected to various chemical and bacterial analyses. Turbidity, pH, calcium hardness, free chlorine, oxidation reduction potential, fungi, yeasts, sulphide reducing bacteria (SRB) and total bacteria (TB) were measured on water sample analysis. The corrosion rate was measured on steel samples and corrosion coupon measurements were installed in fire water for validating non flow assisted localized corrosion. The sulphide reducing bacteria (SRB) presents in fire water causes a localized micro biological corrosion attack of line pipe.

Research and design of hanger and support series of nuclear safety class process piping

International Nuclear Information System (INIS)

Mao Chengzhang; Shi Jiemin

1995-12-01

Hangers and supports of nuclear safety class piping are an important part of primary system piping in a nuclear power plant. They will directly affect the reliability of operation, the period at construction and the investment for a nuclear power plant. It is an absolutely necessary job for Pakistan Chashma Nuclear Power Plant Project to research and design a series of piping supports in accordance with ASME-III NF. It is also an important designing for developing nuclear power plant later in China. After working over two years, a series of piping supports of nuclear safety class which have 57 types and more than 2460 specifications have been designed. This series is perfect, and can satisfy the requirements of piping final designing for nuclear power plant. This series of hangers and supports is mainly used in the process piping of nuclear safety class 1,2,3. They can also be used in other piping of nuclear safety class and piping with aseismic requirement of non-nuclear safety class

Determination of Burst Pressure of API Steel Pipes using Stress Modified Critical Strain Model

International Nuclear Information System (INIS)

Alang, N A; Razak, N A; Sulaiman, A S

2012-01-01

This paper presents a technique which can be used to determine the burst pressure of defective steel pipes using non-linear finite element (FE) analysis. The technique uses stress modified critical strain (SMCS) failure criterion to study the effect of gouge defects on maximum working pressure of API X65 steel pipes. The procedures in determining the model parameters using 3-D, homogeneous isotropic elastic-plastic material model with large deformation finite element analyses from notched tensile bars were systematically discussed. The relationship between burst pressure and gouge depth was proposed. The burst pressure estimated then was compared to experimental data from the literature for validation showing overall good agreements.

Fracture toughness evaluations of TP304 stainless steel pipes

International Nuclear Information System (INIS)

Rudland, D.L.; Brust, F.W.; Wilkowski, G.M.

1997-02-01

In the IPIRG-1 program, the J-R curve calculated for a 16-inch nominal diameter, Schedule 100 TP304 stainless steel (DP2-A8) surface-cracked pipe experiment (Experiment 1.3-3) was considerably lower than the quasi-static, monotonic J-R curve calculated from a C(T) specimen (A8-12a). The results from several related investigations conducted to determine the cause of the observed toughness difference are: (1) chemical analyses on sections of Pipe DP2-A8 from several surface-cracked pipe and material property specimen fracture surfaces indicate that there are two distinct heats of material within Pipe DP2-A8 that differ in chemical composition; (2) SEN(T) specimen experimental results indicate that the toughness of a surface-cracked specimen is highly dependent on the depth of the initial crack, in addition, the J-R curves from the SEN(T) specimens closely match the J-R curve from the surface-cracked pipe experiment; (3) C(T) experimental results suggest that there is a large difference in the quasi-static, monotonic toughness between the two heats of DP2-A8, as well as a toughness degradation in the lower toughness heat of material (DP2-A8II) when loaded with a dynamic, cyclic (R = -0.3) loading history

77 FR 41967 - Certain Circular Welded Carbon Steel Pipes and Tubes From India, Thailand, and Turkey; Certain...

Science.gov (United States)

2012-07-17

... Determination of Sales at Less Than Fair Value: Certain Circular Welded Non-Alloy Steel Pipe From Korea, 57 FR... pipe tubing used for farming and support members for reconstruction or load bearing purposes in the...

Instability predictions for circumferentially cracked Type-304 stainless-steel pipes under dynamic loading. Final report

International Nuclear Information System (INIS)

Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

1982-04-01

This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models

Nuclear piping system damping data studies

International Nuclear Information System (INIS)

Ware, A.G.; Arendts, J.G.

1985-01-01

A programm has been conducted at the Idaho National Engineering Laboratory to study structural damping data for nuclear piping systems and to evaluate if changes in allowable damping values for structural seismic analyses are justified. The existing pipe damping data base was examined, from which a conclusion was made that there were several sets of data to support higher allowable values. The parameters which most influence pipe damping were identified and an analytical investigation demonstrated that increased damping would reduce the required number of seismic supports. A series of tests on several laboratory piping systems was used to determine the effect of various parameters such as types of supports, amplitude of vibration, frequency, insulation, and pressure on damping. A multiple regression analysis was used to statistically assess the influence of the various parameters on damping, and an international pipe damping data bank has been formed. (orig.)

Microstructural characterization of primary coolant pipe steel

International Nuclear Information System (INIS)

Miller, M.K.; Bentley, J.

1986-01-01

Atom probe field-ion microscopy, analytical electron microscopy, and optical microscopy have been used to investigate the changes that occur in the microstructure of cast CF 8 primary coolant pipe stainless steel after long term thermal aging. The cast duplex microstructure consisted of austenite with 15% delta-ferrite. Investigation of the aged material revealed that the ferrite spinodally decomposed into a fine scaled network of α and α'. A fine G-phase precipitate was also observed in the ferrite. The observed degradation in mechanical properties is probably a consequence of the spinodal decomposition in the ferrite

High-energy air shock study in steel and grout pipes

International Nuclear Information System (INIS)

Glenn, H.D.; Kratz, H.R.; Keough, D.D.; Duganne, D.A.; Ruffner, D.J.; Swift, R.P.; Baum, D.

1979-01-01

Voitenko compressors are used to generate 43 mm/μs air shocks in both a steel and a grout outlet pipe containing ambient atmospheric air. Fiber-optic ports provide diaphragm burst times, time-of-arrival (TOA) data, and velocities for the shock front along the 20-mm-ID exit pipes. Pressure profiles are obtained at higher enthalpy shock propagation than ever before and at many locations along the exit pipes. Numerous other electronic sensors and postshot observations are described, as well as experimental results. The primary objectives of the experiments are as follows: (1) provide a data base for normalization/improvement of existing finite-difference codes that describe high-energy air shocks and gas propagation; (2) obtain quantitative results on the relative attenuation effects of two very different wall materials for high-energy air shocks and gas flows. The extensive experimental results satisfy both objectives

Application of heat pipes in nuclear reactors for passive heat removal

Energy Technology Data Exchange (ETDEWEB)

Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

Logistics Outsourcing and The Role of Logistics Service Providers: A Study About Steel Pipe Production Enterprises in Turkey

OpenAIRE

Yıldız, Mehmet Selami; Turan, İlker

2015-01-01

Specialization requirement led the outsourcing preferred intensely by firms. Evidently “logistics activities” constitute an important part in the outsourced business activities. In this study, interviews were conducted with the managers from Turkey's steel pipe production enterprises who have knowledge of logistics management. Fourteen steel pipe enterprises were included in the study. The collected data were analyzed by using quantitative methods and data were obtained by interview and face-...

Application of leak-before-break to primary loop piping to eliminate pipe whip restraints in a Spanish nuclear power plant

International Nuclear Information System (INIS)

Rodriguez, M.; Esteban, A.

1990-01-01

The Spanish plant described in this study is a 982 MWe PWR with a three-loop primary circuit of piping made from centrifugally-cast stainless steel SA351 CF8A. The licensee requested from Consejo de Seguridad Nuclear (CSN) an exemption from the general design criterion, GDC-4, so as to avoid the need to postulate a guillotine rupture of the primary loop piping. The request was based on the generic work performed for a US PWR plant group in order to have such an exemption. As the piping material in the Spanish plant is different from that in the plants included in the generic work, CSN performed a review of the applicability of the generic results to the Spanish plant. Also, aspects such as fatigue evaluation, net section collapse, crack growth and leak detection, specifically analyzed for the Spanish plant, were reviewed. CSN found that fracture toughness test results from generic work are applicable to the Spanish plant; sufficient margin exists against unstable crack extension, and adequate leak detection capability exists with the leakage detection systems available in the plant. Exemption from GDC-4 was approved and CSN authorized the licensee to remove protection devices against dynamic loads from guillotine breaks in the primary coolant loops. (author)

A Study on the Characteristics of Corrosion in Cold Worked Flexible STS 304 Stainless Steel Pipes

International Nuclear Information System (INIS)

Kim, In Soo; Kim, Sung Jin

1993-01-01

Effects of cold working on the corrosion resistance of austenitic STS 304 stainless steel pipes were investigated using anodic polarization method, EDX analysis and SEM technique. Corrosion products had a lots of S and Cl - ion. Generally, corrosion patterns as a result of STS 304 stainless steel to concrete environment were proceeded in the order of the pitting to intergranular corrosion. In the case of the flexible pipes were covered tightly with other polymer materials, crevice corrosion occurred to a much greater extent on austenitic than on martensitic region

Analysis of Defective Pipings in Nuclear Power Plants and Applications of Guided Ultrasonic Wave Techniques

International Nuclear Information System (INIS)

Koo, Dae Seo; Cheong, Yong Moo; Jung, Hyun Kyu; Park, Chi Seung; Park, Jae Suck; Choi, H. R.; Jung, S. S.

2006-07-01

In order to apply the guided ultrasonic techniques to the pipes in nuclear power plants, the cases of defective pipes of nuclear power plants, were investigated. It was confirmed that geometric factors of pipes, such as location, shape, and allowable space were impertinent for the application of guided ultrasonic techniques to pipes of nuclear power plants. The quality of pipes, supports, signals analysis of weldment/defects, acquisition of accurate defects signals also make difficult to apply the guided ultrasonic techniques to pipes of nuclear power plants. Thus, a piping mock-up representing the pipes in the nuclear power plants were designed and fabricated. The artificial flaws will be fabricated on the piping mock-up. The signals of guided ultrasonic waves from the artificial flaws will be analyzed. The guided ultrasonic techniques will be applied to the inspection of pipes of nuclear power plants according to the basis of signals analysis of artificial flaws in the piping mock-up

Elastic-plastic fracture analysis of carbon steel piping using the latest CEGB R6 approach

International Nuclear Information System (INIS)

Kanno, S.; Hasegawa, K.; Shimizu, T.; Kobayashi, H.

1991-01-01

The elastic-plastic fracture of carbon steel piping having various pipe diameters and circumferential crack angles and subjected to a bending moment is analyzed using the latest United Kingdom Central Electricity Generating Board R6 approach. The elastic-plastic fracture criterion must be applied instead of the plastic collapse criterion with increase of the pipe diameter and the crack angle. A simplified elastic-plastic fracture analysis procedure based on the R6 approach is proposed. (author)

Steel structures for nuclear facilities

International Nuclear Information System (INIS)

1993-01-01

In the guide the requirements concerning design and fabrication of steel structures for nuclear facilities and documents to be submitted to the Finnish Centre for Radiation and Nuclear Safety (STUK) are presented. Furthermore, regulations concerning inspection of steel structures during construction of nuclear facilities and during their operation are set forth

Applications of the essay at slow deformation velocity in pipes of stainless steel AISI-304

International Nuclear Information System (INIS)

Zamora R, L.; Mora R, T. De la

2004-01-01

Nowadays is carried out research related with the degradation mechanisms of structures, systems and/or components in the nuclear power plants, since many of the involved processes are those responsible for the dependability of these, of the integrity of the components and of the aspects of safety. The purpose of this work, was to determine the grade of susceptibility to the corrosion of a pipe of Austenitic stainless steel AISI 304, in a solution of Na CI (3.5%) to the temperatures of 60 and 90 C, in two different thermal treatments - 1. - Sensitive 650 C by 4 hours and cooled in water. 2. Solubilized to 1050 C by 1 hour and cooled in water

Recent studies on the welding of austenitic stainless steel piping for BWR service

International Nuclear Information System (INIS)

Childs, W.J.

1986-01-01

The incidence of intergranular stress corrosion cracking (IGSCC) in stainless steel piping in BWR power plants has led to the development of various countermeasures. Replacement of the susceptible Type 304 stainless steel with Type 316 nuclear grade stainless steel has been done by a number of plants. In order to minimize radiation exposure to welding personnel, automatic GTA welding has been used wherever possible when we make the field welds. Studies have shown that the residual stresses in the welded butt joints are affected by the welding process, weld joint design and welding procedures. A new weld joint design has been developed which minimizes the volume of deposited metal while providing adequate access for welding. It also minimizes axial and radial shrinkage and the resulting residual stresses. Other countermeasures, which have been used, include stress modifications such as induction heating stress improvement (IHSI) and last pass heat sink welding (LPHSW). It has been shown that these remedies must be process adjusted to account for the welding process employed. In some cases where UT cracking indication have been detected or where through wall cracking has occurred, weld surfacing has been used to extend life. A further approach to preventing IGSCC in the weld HAZ has been through improvement of the water chemistry by injecting hydrogen to reduce the oxygen level and by keeping the impurity level low

Development of nonlinear dynamic analysis program for nuclear piping systems

International Nuclear Information System (INIS)

Kamichika, Ryoichi; Izawa, Masahiro; Yamadera, Masao

1980-01-01

In the design for nuclear power piping, pipe-whip protection shall be considered in order to keep the function of safety related system even when postulated piping rupture occurs. This guideline was shown in U.S. Regulatory Guide 1.46 for the first time and has been applied in Japanese nuclear power plants. In order to analyze the dynamic behavior followed by pipe rupture, nonlinear analysis is required for the piping system including restraints which play the role of an energy absorber. REAPPS (Rupture Effective Analysis of Piping Systems) has been developed for this purpose. This program can be applied to general piping systems having branches etc. Pre- and post- processors are prepared in this program in order to easily input the data for the piping engineer and show the results optically by use of a graphic display respectively. The piping designer can easily solve many problems in his daily work by use of this program. This paper describes about the theoretical background and functions of this program and shows some examples. (author)

Defect detection of wall thinning defect in pipes using lock-in photo-infrared thermography technique

Energy Technology Data Exchange (ETDEWEB)

Jang, Su Ok; Park, Jong Hyun; Choi, Tae Ho; Jung, Hyun Chul; Kim, Kyoung Suk [Chosun Univ., Gwangju (Korea, Republic of)

2008-07-01

Piping in the Nuclear Power plants (NPP) are mostly consisted of carbon steel pipe. The wall thinning defect is mainly occurred by the affect of the Flow Accelerated Corrosion (FAC) of fluid which flows in carbon steel pipes. This type of defect becomes the cause of damage or destruction of piping. Therefore, it is very important to measure defect which is existed not only on the welding partbut also on the whole field of pipe. Over the years, Infrared Thermography (IRT) has been used as a non destructive testing methods of the various kinds of materials. This technique has many merits and applied to the industrial field but has limitation to the materials. Therefore, this method was combined with lock-in technique. So IRT detection resolution has been progressively improved using lock-in technique. In this paper, the quantitative analysis results of the location and the size of wall thinning defect that is artificially processed inside the carbon steel pipe by using IRT are obtained.

Impact of inservice inspection on the reliability of nuclear piping

International Nuclear Information System (INIS)

Woo, H.H.

1983-12-01

The reliability of nuclear piping is a function of piping quality as fabricated, service loadings and environments, plus programs of continuing inspection during operation. This report presents the results of a study of the impact of inservice inspection (ISI) programs on the reliability of specific nuclear piping systems that have actually failed in service. Two major factors are considered in the ISI programs: one is the capability of detecting flaws; the other is the frequency of performing ISI. A probabilistic fracture mechanics model issued to estimate the reliability of two nuclear piping lines over the plant life as functions of the ISI programs. Examples chosen for the study are the PWR feedwater steam generator nozzle cracking incident and the BWR recirculation reactor vessel nozzle safe-end cracking incident

Influence of heterogeneity of austenitic steel 08x18H10T on the integrity of important installations for the nuclear safety

International Nuclear Information System (INIS)

Dominguez, H.; Menendez, C.M.; Sendoya, F.; Herrera, V.; Rodriguez, R.

1993-01-01

The results of the analysis of failure due to holes occurred in austenitic steel pipes assembled in the channeling system of the special building and in the cooling system of the recharge pond of Juragua nuclear power plant are shown in this work

Ductile fracture of circumferentially cracked type-304 stainless steel pipes in tension

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Norris, D.M.

1984-11-01

Circumferentially cracked pipes subjected to tensile load were analyzed for finite length and constant depth part-through cracks located at the inside of the pipe wall. The analysis postulated loads sufficient to cause net-section yielding of the flawed section. It was demonstrated that a propensity for predominantly radial growth exists for part-through cracks loaded in tension. This result is similar to the result for bend loading, except that bend loading causes more favorable conditions for wall breakthrough than tension loading. Numerical results were developed for 4-in. and 24-in-dia pipes. Safety margins for displacement controlled loads were described by a safety assessment diagram. This diagram defines a curve delineating leak from fracture in a space of nondimensional crack length and crack depth. 4-india schedule 80 Type-304 stainless steel pipes with length to radius ratio (L/R) of up to 100 exhibited leak-before-break behavior.

Ductile fracture of circumferentially cracked type-304 stainless steel pipes in tension

International Nuclear Information System (INIS)

Zahoor, A.; Norris, D.M.

1984-01-01

Circumferentially cracked pipes subjected to tensile load were analyzed for finite length and constant depth part-through cracks located at the inside of the pipe wall. The analysis postulated loads sufficient to cause net-section yielding of the flawed section. It was demonstrated that a propensity for predominantly radial growth exists for part-through cracks loaded in tension. This result is similar to the result for bend loading, except that bend loading causes more favorable conditions for wall breakthrough than tension loading. Numerical results were developed for 4-in. and 24-in-dia pipes. Safety margins for displacement controlled loads were described by a safety assessment diagram. This diagram defines a curve delineating leak from fracture in a space of nondimensional crack length and crack depth. 4-india schedule 80 Type-304 stainless steel pipes with length to radius ratio (L/R) of up to 100 exhibited leak-before-break behavior

Response of buried pipes to missile impact

International Nuclear Information System (INIS)

Vardanega, C.; Cremonini, M.G.; Mirone, M.; Luciani, A.

1989-01-01

This paper presents the methodology and results of the analyses carried out to determine an effective layout and the dynamic response of safety related cooling water pipes, buried in backfill, for the Alto Lazio Nuclear Power Plant in Italy, subjected to missile impact loading at the backfill surface. The pipes are composed of a steel plate encased in two layers of high-quality reinforced concrete. The methodology comprises three steps. The first step is the definition of the 'free-field' dynamic response of the backfill soil, not considering the presence of the pipes, through a dynamic finite element direct integration analysis utilizing an axisymmetric model. The second step is the pipe-soil interaction analysis, which is conducted by utilizing the soil displacement and stress time-histories obtained in the previous steps. Soil stress time-histories, combined with the geostatic and other operational stresses (such as those due to temperature and pressure), are used to obtain the actions in the pipe walls due to ring type deformation. For the third step, the analysis of the beam type response, a lumped parameter model is developed which accounts for the soil stiffness, the pipe characteristics and the position of the pipe with respect to the impact area. In addition, the effect of the presence of large concrete structures, such as tunnels, between the ground surface and the pipe is evaluated. The results of the structural analyses lead to defining the required steel thickness and also allow the choice of appropriate embedment depth and layout of redundant lines. The final results of the analysis is not only the strength verification of the pipe section, but also the definition of an effective layout of the lines in terms of position, depth, steel thickness and joint design. (orig.)

Corrosion by concentrated sulfuric acid in carbon steel pipes and tanks: state of the art

Energy Technology Data Exchange (ETDEWEB)

Panossian, Zehbour; Almeida, Neusvaldo Lira de; Sousa, Raquel Maria Ferreira de [Instituto de Pesquisas Tecnologicas (IPT), Sao Paulo, SP (Brazil); Pimenta, Gutemberg de Souza [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil). Centro de Pesquisas e Desenvolvimento (CENPES); Marques, Leandro Bordalo Schmidt [PETROBRAS Engenharia, Rio de Janeiro, RJ (Brazil)

2009-07-01

PETROBRAS, allied to the policy of reduction of emission of pollutants, has been adjusting the processes of the new refineries to obtain products with lower sulfur content. Thus, the sulfur dioxide, extracted from the process gases of a new refinery to be built in the Northeast, will be used to produce sulfuric acid with concentration between (94-96) %. This acid will be stored in carbon steel tanks and transported through a buried 8-km carbon steel pipe from the refinery to a pier, where it will be loaded onto ships and sent to the consumer markets. Therefore, the corrosion resistance of carbon steel by concentrated acid will become a great concern for the mentioned storage and transportation. When the carbon steel comes into contact with concentrated sulfuric acid, there is an immediate acid attack with the formation of hydrogen gas and ferrous ions which, in turn, forms a protective layer of FeSO{sub 4} on the metallic surface. The durability of the tanks and pipes made of carbon steel will depend on the preservation of this protective layer. This work presents a review of the carbon steel corrosion in concentrated sulfuric acid and discusses the preventive methods against this corrosion, including anodic protection. (author)

Instability predictions for circumferentially cracked Type-304 stainless-steel pipes under dynamic loading. Final report. [BWR

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

1982-04-01

This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models.

Reactor Primary Coolant System Pipe Rupture Study. Progress report No. 32, July--December 1974

International Nuclear Information System (INIS)

1975-03-01

The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase I), analytical and experimental efforts (Phase II) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue studies focused on Elastic/Plastic ASME Code Design Rules, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, and (c) studies directed at quantifying weld sensitization in T-304 stainless steel. (auth)

Round Robin Analyses on Stress Intensity Factors of Inner Surface Cracks in Welded Stainless Steel Pipes

Directory of Open Access Journals (Sweden)

Chang-Gi Han

2016-12-01

Full Text Available Austenitic stainless steels (ASSs are widely used for nuclear pipes as they exhibit a good combination of mechanical properties and corrosion resistance. However, high tensile residual stresses may occur in ASS welds because postweld heat treatment is not generally conducted in order to avoid sensitization, which causes a stress corrosion crack. In this study, round robin analyses on stress intensity factors (SIFs were carried out to examine the appropriateness of structural integrity assessment methods for ASS pipe welds with two types of circumferential cracks. Typical stress profiles were generated from finite element analyses by considering residual stresses and normal operating conditions. Then, SIFs of cracked ASS pipes were determined by analytical equations represented in fitness-for-service assessment codes as well as reference finite element analyses. The discrepancies of estimated SIFs among round robin participants were confirmed due to different assessment procedures and relevant considerations, as well as the mistakes of participants. The effects of uncertainty factors on SIFs were deducted from sensitivity analyses and, based on the similarity and conservatism compared with detailed finite element analysis results, the R6 code, taking into account the applied internal pressure and combination of stress components, was recommended as the optimum procedure for SIF estimation.

Round robin analysis on stress intensity factor of inner surface cracks in welded stainless steel pipes

Energy Technology Data Exchange (ETDEWEB)

Han, Chang Gi; Chang, Yoon Suk [Dept. of Nuclear Engineering, College of Engineering, Kyung Hee University, Yongin (Korea, Republic of); Kim, Jong Sung [Dept. of Mechanical Engineering, Sunchon National University, Sunchon (Korea, Republic of); Kim, Maan Won [Central Research Institute, Korea Hydro and Nuclear Power Company, Daejeon (Korea, Republic of)

2016-12-15

Austenitic stainless steels (ASSs) are widely used for nuclear pipes as they exhibit a good combination of mechanical properties and corrosion resistance. However, high tensile residual stresses may occur in ASS welds because postweld heat treatment is not generally conducted in order to avoid sensitization, which causes a stress corrosion crack. In this study, round robin analyses on stress intensity factors (SIFs) were carried out to examine the appropriateness of structural integrity assessment methods for ASS pipe welds with two types of circumferential cracks. Typical stress profiles were generated from finite element analyses by considering residual stresses and normal operating conditions. Then, SIFs of cracked ASS pipes were determined by analytical equations represented in fitness-for-service assessment codes as well as reference finite element analyses. The discrepancies of estimated SIFs among round robin participants were confirmed due to different assessment procedures and relevant considerations, as well as the mistakes of participants. The effects of uncertainty factors on SIFs were deducted from sensitivity analyses and, based on the similarity and conservatism compared with detailed finite element analysis results, the R6 code, taking into account the applied internal pressure and combination of stress components, was recommended as the optimum procedure for SIF estimation.

Nuclear Car Wash sensitivity in varying thicknesses of wood and steel cargo

International Nuclear Information System (INIS)

Church, J; Slaughter, D; Asztalos, S; Biltoft, P; Descalle, M; Hall, J; Manatt, D; Mauger, J; Norman, E; Petersen, D; Prussin, S

2006-01-01

The influence of incident neutron attenuation on signal strengths in the Nuclear Car Wash has been observed experimentally for both wood and steel-pipe mock cargos. Measured decay curves are presented for β-delayed high-energy γ-rays and thermalized neutrons following neutron-induced fission of HEU through varying irradiation lengths. Error rates are extracted for delayed-γ and delayed-n signals integrated to 30 seconds, assuming Gaussian distributions for the active background. The extrapolation to a field system of 1 mA deuterium current and to a 5 kg sample size is discussed

An appraisal of procedures used to give the criterion for instability of a through-wall circumferential crack in a stainless steel piping system

International Nuclear Information System (INIS)

Smith, E.

1989-01-01

Against the background of the problem of intergranular stress corrosion cracking of 304 stainless steel in Boiling Water Reactor piping systems, this paper presents a critical appraisal of procedures that are currently used to give the criterion for instability of a through-wall circumferential crack in a stainless steel piping system. Particular attention is focussed on a simple procedure developed by Cotter, Chang and Zahoor, which has been applied to specific piping systems, the objective being to underpin its viability. The considerations are applicable to not only Boiling Water Reactor piping systems, but to other piping systems where pipe failure due to circumferential cracking is a potential problem. (author)

Situation of secondary system piping wearing in overseas nuclear power plants

International Nuclear Information System (INIS)

Chiba, Goro

2005-01-01

In consideration of secondary system piping rupture accident at Mihama Nuclear Power Station Unit 3 of Kansai Electric Power Company in August 2004, the management system of secondary pipe wall thickness of Japan and foreign countries were investigated. Moreover, the tendency of the secondary piping thinning events on overseas which the Institute of Nuclear Safety System, Inc. (INSS) obtained was analyzed in order to verify the validity of the Japanese management system. Consequently, it was shown that in the U.S., the fault phenomenon of secondary system piping was reported continuously, and there were also many cases of both degradation and penetration of pipe wall. (author)

Heat pipe nuclear reactor for space power

Science.gov (United States)

Koening, D. R.

1976-01-01

A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

Survey on application of probabilistic fracture mechanics approach to nuclear piping

International Nuclear Information System (INIS)

Kashima, Koichi

1987-01-01

Probabilistic fracture mechanics (PFM) approach is newly developed as one of the tools to evaluate the structural integrity of nuclear components. This report describes the current status of PFM studies for pressure vessel and piping system in light water reactors and focuses on the investigations of the piping failure probability which have been undertaken by USNRC. USNRC reevaluates the double-ended guillotine break (DEGB) of rector coolant piping as a design basis event for nuclear power plant by using the PFM approach. For PWR piping systems designed by Westinghouse, two causes of pipe break are considered: pipe failure due to the crack growth and pipe failure indirectly caused by failure of component supports due to an earthquake. PFM approach shows that the probability of DEGB from either cause is very low and that the effect of earthquake on pipe failure can be neglected. (author)

Damping values for nuclear power plant piping during seismic events and fluid-induced transients

International Nuclear Information System (INIS)

Ware, A.G.

1986-01-01

For several years the Idaho National Engineering Laboratory (INEL) has been assisting the United States Nuclear Regulatory Commission (USNRC) in efforts to establish best-estimate damping values for use in the dynamic analysis of nuclear power plant piping systems. Data from a number of piping vibration tests conducted at facilities worldwide (including the INEL) have been collected, evaluated, reported, and placed in a nuclear piping data bank at the INEL. These data are being used to justify changes in allowable damping values for use in nuclear piping design, thus making piping systems safer, less costly, and easier to inspect and maintain

Residual-stresses in austenitic stainless-steel primary coolant pipes and welds of pressurized-water reactors

International Nuclear Information System (INIS)

Faure, F.; Leggatt, R.H.

1996-01-01

Surface and through thickness residual stress measurements were performed on an aged cast austenitic-ferritic stainless steel pipe and on an orbital TIG weld representative of those of primary coolant pipes in pressurized water reactors. An abrasive-jet hole drilling method and a block removal and layering method were used. Surface stresses and through thickness stress profiles are strongly dependent upon heat treatments, machining and welding operations. In the aged cast stainless steel pipe, stresses ranged between -250 and +175 MPa. On and near the orbital TIG weld, the outside surface of the weld was in tension both in the axial and hoop directions, with maximum values reaching 420 MPa in the weld. On the inside surface, the hoop stresses were compressive, reaching -300 MPa. However, the stresses in the axial direction at the root of the weld were tensile within 4 mm depth from the inside surface, locally reaching 280 MPa. (author)

77 FR 10773 - Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines; Scheduling of...

Science.gov (United States)

2012-02-23

... Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines; Scheduling of Expedited Five-Year... orders on stainless steel butt-weld pipe fittings from Italy, Malaysia, and the Philippines would be... certificate of service. Determination.--The Commission has determined to exercise its authority to extend the...

77 FR 32531 - Circular Welded Carbon-Quality Steel Pipe From the Sultanate of Oman: Preliminary Determination...

Science.gov (United States)

2012-06-01

...-Quality Steel Pipe From the Sultanate of Oman: Preliminary Determination of Sales at Less Than Fair Value... pipe) from the Sultanate of Oman (Oman) is being, or is likely to be, sold in the United States at less... from India, Oman, the United Arab Emirates (UAE), and the Socialist Republic of Vietnam (Vietnam) on...

RESEARCH OF INFLUENCE OF THE HIGH-SPEED THERMAL PROCESSING REGIMES ON STRUCTURE AND MECHANICAL PROPERTIES OF PIPE STEEL 32G2

Directory of Open Access Journals (Sweden)

A. I. Gordienko

2012-01-01

Full Text Available Researches on influence of high-speed heating temperature, regimes of cooling and temperature of abatement on structure and mechanical properties of pipe steel 32G2 are carried out. Recommendations on the regimes of high-speed thermal processing of steel 32G2 which can be used at manufacturing of seamless pipes are given.

Electrolytic etching of fine stainless-steel pipes patterned by laser-scan lithography

Science.gov (United States)

Takahashi, Hiroshi; Sagara, Tomoya; Horiuchi, Toshiyuki

2017-07-01

Recently, it is required to develop a method for fabricating cylindrical micro-components in the field of measurement and medical engineering. Here, electrolytic etching of fine stainless-steel pipes patterned by laser-scan lithography was researched. The pipe diameter was 100 μm. At first, a pipe coated with 3-7 μm thick positive resist (tok, PMER P LA-900) was exposed to a violet laser beam with a wavelength of 408 nm (Neoark,TC20-4030-45). The laser beam was reshaped in a circle by placing a pinhole, and irradiated on the pipe by reducing the size in 1/20 using a reduction projection optics. Linearly arrayed 22 slit patterns with a width of 25 μm and a length of 175 μm were delineated in every 90-degree circumferential direction. That is, 88 slits in total were delineated at an exposure speed of 110 μm/s. In the axial direction, patterns were delineated at intervals of 90 μm. Following the pattern delineation, the pipe masked by the resist patterns was electrolytically etched. The pipe was used as an anode and an aluminum cylinder was set as a cathode around the pipe. As the electrolyte, aqueous solution of NaCl and NH4Cl was used. After etching the pipe, the resist was removed by ultrasonic cleaning in acetone. Although feasibility for fabricating multi-slit pipes was demonstrated, sizes of the etched slits were enlarged being caused by the undercut, and the shapes were partially deformed, and all the pipes were snapped at the chuck side.

Continuous wavelet transform analysis and modal location analysis acoustic emission source location for nuclear piping crack growth monitoring

Energy Technology Data Exchange (ETDEWEB)

Mohd, Shukri [Nondestructive Testing Group, Industrial Technology Division, Malaysian Nuclear Agency, 43000, Bangi, Selangor (Malaysia); Holford, Karen M.; Pullin, Rhys [Cardiff School of Engineering, Cardiff University, Queen' s Buildings, The Parade, CARDIFF CF24 3AA (United Kingdom)

2014-02-12

Source location is an important feature of acoustic emission (AE) damage monitoring in nuclear piping. The ability to accurately locate sources can assist in source characterisation and early warning of failure. This paper describe the development of a novelAE source location technique termed 'Wavelet Transform analysis and Modal Location (WTML)' based on Lamb wave theory and time-frequency analysis that can be used for global monitoring of plate like steel structures. Source location was performed on a steel pipe of 1500 mm long and 220 mm outer diameter with nominal thickness of 5 mm under a planar location test setup using H-N sources. The accuracy of the new technique was compared with other AE source location methods such as the time of arrival (TOA) techniqueand DeltaTlocation. Theresults of the study show that the WTML method produces more accurate location resultscompared with TOA and triple point filtering location methods. The accuracy of the WTML approach is comparable with the deltaT location method but requires no initial acoustic calibration of the structure.

Continuous wavelet transform analysis and modal location analysis acoustic emission source location for nuclear piping crack growth monitoring

International Nuclear Information System (INIS)

Shukri Mohd

2013-01-01

Full-text: Source location is an important feature of acoustic emission (AE) damage monitoring in nuclear piping. The ability to accurately locate sources can assist in source characterisation and early warning of failure. This paper describe the development of a novelAE source location technique termed Wavelet Transform analysis and Modal Location (WTML) based on Lamb wave theory and time-frequency analysis that can be used for global monitoring of plate like steel structures. Source location was performed on a steel pipe of 1500 mm long and 220 mm outer diameter with nominal thickness of 5 mm under a planar location test setup using H-N sources. The accuracy of the new technique was compared with other AE source location methods such as the time of arrival (TOA) technique and DeltaTlocation. The results of the study show that the WTML method produces more accurate location results compared with TOA and triple point filtering location methods. The accuracy of the WTML approach is comparable with the deltaT location method but requires no initial acoustic calibration of the structure. (author)

Continuous wavelet transform analysis and modal location analysis acoustic emission source location for nuclear piping crack growth monitoring

International Nuclear Information System (INIS)

Mohd, Shukri; Holford, Karen M.; Pullin, Rhys

2014-01-01

Source location is an important feature of acoustic emission (AE) damage monitoring in nuclear piping. The ability to accurately locate sources can assist in source characterisation and early warning of failure. This paper describe the development of a novelAE source location technique termed 'Wavelet Transform analysis and Modal Location (WTML)' based on Lamb wave theory and time-frequency analysis that can be used for global monitoring of plate like steel structures. Source location was performed on a steel pipe of 1500 mm long and 220 mm outer diameter with nominal thickness of 5 mm under a planar location test setup using H-N sources. The accuracy of the new technique was compared with other AE source location methods such as the time of arrival (TOA) techniqueand DeltaTlocation. Theresults of the study show that the WTML method produces more accurate location resultscompared with TOA and triple point filtering location methods. The accuracy of the WTML approach is comparable with the deltaT location method but requires no initial acoustic calibration of the structure

Thermal aging evaluation of cast austenitic stainless steel pipe

International Nuclear Information System (INIS)

Song, T. H.; Jung, I. S.

2002-01-01

24 years have been passed since Kori Unit 1 began its commercial operation, and 19 years have been passed since Kori Unit 2 began its commercial operation. As the end point of design life become closer, plant life extension and periodic safety assessment is paid more and more attention to by utility company. In this paper, the methodologies and results of cast austenitic stainless steel pipe thermal aging evaluations of both units have been presented in association with aging time of 10, 20, and 30 years and operating temperature, respectively. Life extension cases respectively. As a result of this, at the operating temperature of 280 .deg. C, thermal aging was not a problem as long as Charpy V-notch room temperature minimum impact energy is concerned. However, more than 300 .deg. C and 30 years of operating condition, we should perform detailed fracture mechanics analysis with CMTR of NPP pipe

Inspection indications, stress corrosion cracks and repair of process piping in nuclear materials production reactors

International Nuclear Information System (INIS)

Louthan, M.R. Jr.; West, S.L.; Nelson, D.Z.

1991-01-01

Ultrasonic inspection of Schedule 40 Type 304 stainless steel piping in the process water system of the Savannah River Site reactors has provided indications of discontinuities in less than 10% of the weld heat affected zones. Pipe sections containing significant indications are replaced with Type 304L components. Post removal metallurgical evaluation showed that the indications resulted from stress corrosion cracking in weld heat-affected zones and that the overall weld quality was excellent. The evaluation also revealed weld fusion zone discontinuities such as incomplete penetration, incomplete fusion, inclusions, underfill at weld roots and hot cracks. Service induced extension of these discontinuities was generally not significant although stress corrosion cracking in one weld fusion zone was noted. One set of UT indications was caused by metallurgical discontinuities at the fusion boundary of an extra weld. This extra weld, not apparent on the outer pipe surface, was slightly overlapping and approximately parallel to the weld being inspected. This extra weld was made during a pipe repair, probably associated with initial construction processes. The two nearly parallel welds made accurate assessment of the UT signal difficult. The implications of these observations to the inspection and repair of process water systems of nuclear reactors is discussed

Contributions of the ORNL piping program to nuclear piping design codes and standards

International Nuclear Information System (INIS)

Moore, S.E.

1975-11-01

The ORNL Piping Program was conceived and established to develop basic information on the structural behavior of nuclear power plant piping components and to prepare this information in forms suitable for use in design analysis and codes and standards. One of the objectives was to develop and qualify stress indices and flexibility factors for direct use in Code-prescribed design analysis methods. Progress in this area is described

Some electron beam welding equipments for the nuclear industry

International Nuclear Information System (INIS)

Helm, H.; Rodier, R.; Sayegh, G.

1978-01-01

Results of various electron beam welding equipment developed for the nuclear industry obtained from a 100 kW electron beam machine to weld thick plates made of stainless steel and reactor steel, and from some equipment with local vacuum to weld pipes onto a pipe wall. (orig.) [de

Fatigue check of nuclear safety class 1 reactor coolant pipe

International Nuclear Information System (INIS)

Wang Qing; Fang Yonggang; Chu Qibao; Xu Yu; Li Hailong

2015-01-01

Fatigue and thermal ratcheting analyses of nuclear safety Class 1 reactor coolant pipe in a nuclear power plant were independently carried out in this paper. The software used for calculation is ROCOCO, which is based on RCC-M code. The difference of nuclear safety Class 1 pipe fatigue evaluation between RCC-M code and ASME code was compared. The main aspects of comparison include the calculation scoping of fatigue design, the calculation method of primary plus secondary stress intensity, the elastic-plastic correction coefficient calculation, and the dynamic load combination method etc. By correcting inconsistent algorithm of ASME code within ROCOCO, the fatigue usage factor and thermal ratcheting design margin of 65 mm and 55 mm wall thickness of the pipe were obtained. The results show that the minimum wall thickness of the pipe must exceed 55 mm and the design value of the thermal ratcheting of 55 mm wall thickness reaches 95% of the allowable value. (authors)

Basic concepts about application of dual vibration absorbers to seismic design of nuclear piping systems

International Nuclear Information System (INIS)

Hara, F.; Seto, K.

1987-01-01

The design value of damping for nuclear piping systems is a vital parameter in ensuring safety in nuclear plants during large earthquakes. Many experiments and on-site tests have been undertaken in nuclear-industry developed countries to determine rational design values. However damping value in nuclear piping systems is so strongly influenced by many piping parameters that it shows a tremendous dispersion in its experimental values. A new trend has recently appeared in designing nuclear pipings, where they attempt to use a device to absorb vibration energy induced by seismic excitation. A typical device is an energy absorbing device, made of a special material having a high capacity of plasticity, which is installed between the piping and the support. This paper deals with the basic study of application of dual vibration absorbers to nuclear piping systems to accomplish high damping value and reduce consequently seismic response at resonance frequencies of a piping system, showing their effectiveness from not only numerical calculation but also experimental evaluation of the vibration responses in a 3D model piping system equipped with dual two vibration absorbers

76 FR 77770 - Certain Circular Welded Non-Alloy Steel Pipe From Mexico: Final Results of Antidumping Duty...

Science.gov (United States)

2011-12-14

... previous review or the original less-than-fair-value (LTFV) investigation, the cash deposit rate will... Determination of Sales at Less Than Fair Value: Circular Welded Non-Alloy Steel Pipe From Mexico, 57 FR 42953... structural pipe tubing used for framing and support members for reconstruction or load-bearing purposes in...

75 FR 69052 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From the People's...

Science.gov (United States)

2010-11-10

... threatened with material injury by reason of imports of seamless pipe from the PRC. According to section 736... that determination is based on the threat of material injury and is not accompanied by a finding that... Alloy Steel Standard, Line, and Pressure Pipe From the People's Republic of China: Amended Final...

Contamination of stainless steel process piping with radioactive cobalt colloids and methods for decontamination

International Nuclear Information System (INIS)

Adeleye, S.A.; White, D.A.

1992-01-01

Significant deposits of activity can occur on the internal surface of pipework carrying aqueous radioactive liquid. This paper describes experimental work to contaminate stainless steel piping with colloidal particles of Co 60 and considers decontamination methods. The effects on steel contamination of varying cobalt concentration and aqueous liquor pH were investigated. Deposition increased with increasing cobalt concentration and the ''plate-out'' increased markedly with increasing liquid pH. Low deposition occurred at pH ∼ 2 increasing by orders of magnitude at pH ∼ 10. Steel type had an effect on activity picked up. It was shown that liquid turbulence, on the surface, reduced activity deposition. Since the extent of contamination to be removed depends on deposition or ''plate-out'' kinetics, the factors affecting the rate of activity deposition are considered. Specimens of steel piping were treated by contacting with acid, electroetching or abrasion with emery cloth. Surface treatment was shown to delay deposition, in certain instances, but did not have much effect on overall equilibrium level of surface activity. The surface activity could be reduced by treatment with dilute nitric acid: however significant activity remained. Most of the remaining activity could be removed by treatment with nitric acid in an ultrasonic bath. (Author)

76 FR 67146 - Stainless Steel Butt-Weld Pipe Fittings From Italy; Extension of Time Limit for Preliminary...

Science.gov (United States)

2011-10-31

... DEPARTMENT OF COMMERCE International Trade Administration [A-475-828] Stainless Steel Butt-Weld... antidumping duty order on stainless steel butt-weld pipe fittings from Italy in the Federal Register. See... preliminary results of this review within the original time frame because it needs to obtain additional...

Drop Weight Impact Behavior of Al-Si-Cu Alloy Foam-Filled Thin-Walled Steel Pipe Fabricated by Friction Stir Back Extrusion

Science.gov (United States)

Hangai, Yoshihiko; Nakano, Yukiko; Utsunomiya, Takao; Kuwazuru, Osamu; Yoshikawa, Nobuhiro

2017-02-01

In this study, Al-Si-Cu alloy ADC12 foam-filled thin-walled stainless steel pipes, which exhibit metal bonding between the ADC12 foam and steel pipe, were fabricated by friction stir back extrusion. Drop weight impact tests were conducted to investigate the deformation behavior and mechanical properties of the foam-filled pipes during dynamic compression tests, which were compared with the results of static compression tests. From x-ray computed tomography observation, it was confirmed that the fabricated foam-filled pipes had almost uniform porosity and pore size distributions. It was found that no scattering of the fragments of collapsed ADC12 foam occurred for the foam-filled pipes owing to the existence of the pipe surrounding the ADC12 foam. Preventing the scattering of the ADC12 foam decreases the drop in stress during dynamic compression tests and therefore improves the energy absorption properties of the foam.

Nuclear Power Plants Secondary Circuit Piping Wall-Thinning Management in China

International Nuclear Information System (INIS)

Zhong Zhimin; Li Jinsong; Zheng Hui

2012-01-01

Research and field feedbacks showed that nuclear power plants secondary circuit steam and water piping are more sensitive than that of fuel plant to the attack of flow-accelerated corrosion (FAC). FAC, Liquid droplet impingement or cavitation erosion will cause secondary circuit piping local wall-thinning in NPPs. Without effective management, the wall-thinning in those high energy piping will cause leakage or pipe rupture during nuclear power plant operation, more seriously cause unplanned shut down, injured and fatality, or heavy economic losses. This paper briefly introduces the history, development and state of the art of secondary circuit piping wall-thinning management in China NPPs. Then, the effectiveness of inspection grid size selecting was analyzed in detail based on field feedbacks. EPRI recommendatory inspection grid, JSME code recommendatory grid and plant specific inspection grid were compared and the detection probabilities of local wall-thinning were estimated. Then, the development and application of NPPs Secondary Circuit Piping Wall Thickness Management Information System, developed, operated and maintained by our team, was briefly introduced and the statistical analysis results of 11 PWR units were shared. It was conclude that the long term, systemic, effective wall-thinning management strategy of high energy piping was very important to the safety and economic operation of NPPs. Furthermore, take into account the actual situation of China nuclear power plants, some advice and suggestion on developing effective nuclear power plant secondary circuit steam and water piping wall-thinning management system are put forward from code development, design and manufacture, operation management, pipeline and locations selection, inspection method selection and application, thickness measurement result evaluation, residual life predication and decision making, feedbacks usage, personnel training and etc. (author)

Assessment of cracked pipes in primary piping systems of PWR nuclear reactors

International Nuclear Information System (INIS)

Jong, Rudolf Peter de

2004-01-01

Pipes related to the Primary System of Pressurized Water Reactors (PWR) are manufactured from high toughness austenitic and low alloy ferritic steels, which are resistant to the unstable growth of defects. A crack in a piping system should cause a leakage in a considerable rate allowing its identification, before its growth could cause a catastrophic rupture of the piping. This is the LBB (Leak Before Break) concept. An essential step in applying the LBB concept consists in the analysis of the stability of a postulated through wall crack in a specific piping system. The methods for the assessment of flawed components fabricated from ductile materials require the use of Elasto-Plastic Fracture Mechanics (EPFM). Considering that the use of numerical methods to apply the concepts of EPFM may be expensive and time consuming, the existence of the so called simplified methods for the assessment of flaws in piping are still considered of great relevance. In this work, some of the simplified methods, normalized procedures and criteria for the assessment of the ductile behavior of flawed components available in literature are described and evaluated. Aspects related to the selection of the material properties necessary for the application of these methods are also discussed. In a next .step, the methods are applied to determine the instability load in some piping configurations under bending and containing circumferential through wall cracks. Geometry and material variations are considered. The instability loads, obtained for these piping as the result of the application of the selected methods, are analyzed and compared among them and with some experimental results obtained from literature. The predictions done with the methods demonstrated that they provide consistent results, with good level of accuracy with regard to the determination of maximum loads. These methods are also applied to a specific Study Case. The obtained results are then analyzed in order to give

An assessment of seismic margins in nuclear plant piping

International Nuclear Information System (INIS)

Chen, W.P.; Jaquay, K.R.; Chokshi, N.C.; Terao, D.

1995-01-01

Interim results of an ongoing program to assist the U.S. Nuclear Regulatory Commission (NRC) in developing regulatory positions on the seismic analyses of piping and overall safety margins of piping systems are reported. Results of reviews of previous seismic testing, primarily the Electric Power Research Institute (EPRI)/NRC Piping and Fitting Dynamic Reliability Program, and assessments of the ASME Code, Section III, piping seismic design criteria as revised by the 1994 Addenda are reported. Major issues are identified herein only. Technical details are to be provided elsewhere. (author). 4 refs., 2 figs

High-cycle fatigue properties of small-bore socket-welded pipe joint

International Nuclear Information System (INIS)

Maekawa, Akira; Noda, Michiyasu; Suzuki, Michiaki

2009-01-01

Piping and equipment in nuclear power plants are structures including many welded joints. Reliability of welded joints is one of high-priority issues to improve the safety of nuclear power plants. However, occurrence of fatigue failures in small-bore socket-welded pipe joints by high-cycle vibrations is still reported. In this study, fatigue experiments on a socket-welded joint of austenitic stainless steel pipe was conducted under excitation conditions similar to those in actual plants to investigate vibration characteristics and fatigue strength. It was found that the natural frequency of pipe with socket-welded joint gradually decreased as fatigue damage developed, according to the Miner rule for fatigue life evaluation. The results indicate that the fatigue life of the welded pipe joint could be estimated by monitoring the decreasing ratio of the natural frequency of the pipe. The evaluation of decreasing ratio of the natural frequency in addition to fatigue damage evaluation by the Miner rule could enhance the accuracy of fatigue life evaluation. (author)

Failure pressure of straight pipe with wall thinning under internal pressure

International Nuclear Information System (INIS)

Kamaya, Masayuki; Suzuki, Tomohisa; Meshii, Toshiyuki

2008-01-01

The failure pressure of pipe with wall thinning was investigated by using three-dimensional elastic-plastic finite element analyses (FEA). With careful modeling of the pipe and flaw geometry in addition to a proper stress-strain relation of the material, FEA could estimate the precise burst pressure obtained by the tests. FEA was conducted by assuming three kinds of materials: line pipe steel, carbon steel, and stainless steel. The failure pressure obtained using line pipe steel was the lowest under the same flaw size condition, when the failure pressure was normalized by the value of unflawed pipe defined using the flow stress. On the other hand, when the failure pressure was normalized by the results of FEA obtained for unflawed pipe under various flaw and pipe configurations, the failure pressures of carbon steel and line pipe steel were almost the same and lower than that of stainless steel. This suggests that the existing assessment criteria developed for line pipe steel can be applied to make a conservative assessment of carbon steel and stainless steel

76 FR 3612 - Circular Welded Carbon Steel Pipes and Tubes From Taiwan; Extension of Time Limit for Preliminary...

Science.gov (United States)

2011-01-20

... DEPARTMENT OF COMMERCE International Trade Administration [A-583-008] Circular Welded Carbon Steel... Steel Pipes and Tubes From Taiwan: Notice of Partial Rescission of Antidumping Duty Administrative... complete the preliminary results of this review within the original time frame because we require...

Manufacture of piping components for nuclear power plants

International Nuclear Information System (INIS)

Bartecek, R.

1983-01-01

Hammer forging of hollow forging ingots, extrusion and elestroslag remelting may be used for the manufacture of large pipes. Technologies have been developed for the manufacture of elbows based on various types of forming. These procedures mainly include the hydraulic pressing of elbows from tubes and the pressing of symmetrical halves of elbows with subsequent welding. The hammer forging of valves, cross pieces, etc., has been replaced by forging and pressing. In order to prevent failures from occurring in the pipes during operation of nuclear power plants, pipes are being made of larger forgings, which reduces the number of welds. This improves the quality of the pipes, reduces production and assembly costs and is metal-saving. (E.S.)

An assessment of composite repair system in offshore platform for corroded circumferential welds in super duplex steel pipe

Directory of Open Access Journals (Sweden)

Silvio de Barros

2018-04-01

Full Text Available The main aim of this study is to assess the effectiveness of a composite repair system in severely corroded circumferential welds in super duplex stainless steel pipes as a preventive measure against the premature corrosion damage at the welds. Artificial defects were fabricated on the super duplex steel tube in order to reproduce the localized corrosion damage defects found in real welded joints. Three kinds of through thickness defects were considered: 25%, 50% and 96% of the perimeter of the pipe. The performance of the repaired pipe was assessed by hydrostatic tests as per ISO 24817 standard. The results showed that the composite repair system can sustain the designed failure pressure even for the pipe damaged with through-wall defect up to 96% of the perimeter of the pipe. Hence, the composite repair system can be used as a preliminary tool to protect the unexpected or premature failure at the welds and maintain an adequate level of mechanical strength for a given operating pressure. This composite repair system can assure that the pipe will not leak until a planned maintenance of the line. Nevertheless, further work is still desirable to improve the confidence in the long-term performance of bonded composite

77 FR 64483 - Circular Welded Carbon-Quality Steel Pipe from the Socialist Republic of Vietnam: Notice of Final...

Science.gov (United States)

2012-10-22

...-0649, respectively. SUPPLEMENTARY INFORMATION: Background On June 1, 2012, the Department published in... Pipe from the Socialist Republic of Vietnam;'' ``Verification of the Sales Response of Midwest Air... Steel Joint Stock Company.... Sun Steel Joint Stock 4.57 Company. Huu Lien Asia Corporation........ Huu...

Nuclear power plant pressure vessels. Control of piping

International Nuclear Information System (INIS)

2000-01-01

The guide presents requirements for the pipework of nuclear facilities in Finland. According to the section 117 of the Finnish Nuclear Energy Degree (161/88), the Radiation and Nuclear Safety Authority of Finland (STUK) controls the pressure vessels of nuclear facilities in accordance with the Nuclear Energy Act (990/87) and, to the extent applicable in accordance with the Act of Pressure Vessels (98/73) and the rules and regulations issued by the virtue of these. In addition STUK is an inspecting authority of pressure vessels of nuclear facilities in accordance with the Pressure Vessel Degree (549/1973). According to the section of the Pressure Vessel Degree, a pressure vessel is a steam boiler, pressure container, pipework of other such appliance in which the pressure is above or may come to exceed the atmospheric pressure. Guide YVL 3.0 describes in general terms how STUK controls pressure vessels. STUK controls Safety Class 1, 2 and 3 piping as well as Class EYT (non-nuclear) and their support structures in accordance with this guide and applies the provisions of the Decision of the Ministry of Trade and Industry on piping (71/1975) issued by virtue of the Pressure Vessel Decree

Instability predictions for circumferentially cracked Type-304 stainless steel pipes under dynamic loading. Volume 2. Appendixes. Final report

International Nuclear Information System (INIS)

Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

1982-04-01

This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models

Reactor primary coolant system pipe rupture study. Progress report No. 33, January--June 1975

International Nuclear Information System (INIS)

1975-10-01

The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase 1), analytical and experimental efforts (Phase 2) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue crack growth rate studies focused on LWR primary piping materials in a simulated BWR primary coolant environment, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, (c) studies directed at quantifying weld sensitization in Type 304 stainless steel, (d) support studies to characterize the electrochemical potential behavior of a typical BWR primary water environment and (e) special tests related to simulation of fracture surfaces characteristic of IGSCC field failures

75 FR 1335 - Circular Welded Carbon Steel Pipes and Tubes from Taiwan; Extension of Time Limit for Preliminary...

Science.gov (United States)

2010-01-11

... DEPARTMENT OF COMMERCE International Trade Administration [A-583-008] Circular Welded Carbon Steel... review of the antidumping duty order on circular welded carbon steel pipes and tubes from Taiwan.\\1\\ On... review within the original time frame because we require additional time to obtain information from the...

Nuclear power plant steam pipes repairing with Tirant 3 Robot system

Energy Technology Data Exchange (ETDEWEB)

Soto, M.; Curiel, M. [Logistica y Acondicionamientos Industriales SAU, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain); Lazaro, F. [Revestimientos Anticorrosivos Industriales, S. L. U., Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain); Arnaldos, A., E-mail: m.soto@lainsa.co [TITANIA Servicios Tecnologicos SL, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain)

2010-10-15

The metallization arc spray process is based on the projection of molten metal, supplied by means of different stainless alloys wire, over a surface of carbon steel usually, with the object of serving as protection against erosion-corrosion, increasing resistance to abrasion and detrition. A typical application functions covering the steam pipes inner surface in coal-fired power station and nuclear power plants. The results of this process are spectacular in terms of protection against corrosion and abrasion, but its application has conditioning factors, such as: Severe application conditions for workers. Due to the worker's postural position (usually kneeling) in 32 diameter pipes and working with fireproof clothing and masks with outdoor air supplying, due to fumes, sparks and molten metal particles, radiological contamination, confined space, poor lighting, ... Coating uniformity. As metallization is a manual process, the carried out measurements show small variations in the thickness of the coating, always within the tolerance limits established by the applicable regulations and quality assurance. An increase in the uniformity of the projected coating, increase the resistance and give a better surface protection. For all these reasons, Lainsa has developed the Tirant 3 robot, a worldwide innovative system, for metallization of steam pipes inner surface. Tirant 3 robot is tele operated from outside of the pipe, so that human intervention is reduced to the operations of robot positioning and change of metallization wire. As it is an independent system of the human factor, metallization process performance is significantly increased by reducing rest periods due only to the robot maintenance. Likewise, Tirant 3 system permits to increase resulting coating uniformity and thus its resistance, keeping selected parameters constant (forward speed, rotation speed and inner surface distance) depending on required type and thickness of wire. (Author)

Development of nuclear grade stainless steels at KCSSL

International Nuclear Information System (INIS)

Balachandran, G.; Dhere, M.; Mahadik, A.; Hinge, N.M.; Balasubramanian, V.

2011-01-01

Kalyani Carpenter Special Steels Ltd is an alloy steel plant, where a variety of alloy steel grades are produced for automotive, defence, nuclear and aerospace applications. The plant has developed expertise in processing of several alloy steel grades of superior quality that meets stringent specifications. Primary steel is processed through a combination of electric arc furnace, ladle furnace and vacuum degassing where stringent control over dephosphorisation, desulphurization, deoxidation is effected to get a refined high quality steel. The molten steel is cast through continuous casting of slabs or ingot casting. In grades specific to nuclear application, the primary cast products are further subjected to electroslag remelting to achieve further freedom from inclusions and to achieve a favourable solidification grain structure, which ultimately improve the hot workability of the alloy steel. Appropriate choice of slag and operating parameters are needed for realising the required ingot quality. The present study would examine the processing and quality aspects of some important grades of steels used in nuclear industry namely ferritic 9Cr-1Mo steel, martensitic stainless steels 403, 410, precipitation hardenable 17-4 PH stainless steel and austenitic 321, 316LN stainless steel, which were made and supplied for applications to Indian nuclear industry. The expertise developed in processing the steels in terms of melting, heat treatment and their relationship to structural features and mechanical properties would be highlighted. (author)

Secondary pipe rupture at Mihama unit 3

International Nuclear Information System (INIS)

Hajime Ito; Takehiko Sera

2005-01-01

The secondary system pipe rupture occurred on August 9, 2004, while Mihama unit 3 was operating at the rated thermal power. The rupture took place on the condensate line-A piping between the No.4 LP heater and the deaerator, downstream of an orifice used for measuring the condensate flux. The pipe is made of carbon steel, and normally has 558.8 mm diameter and 10 mm thickness. The pipe wall had thinned to 0.4 mm at the point of minimum thickness. It is estimated that the disturbed flow of water downstream of the orifice caused erosion/corrosion and developed wall thinning, leading to a rupture at the thinnest section under internal pressure, about 1MPa. Observation of the pipe internal surface revealed a scale-like pattern typical in this kind of phenomenon. Eleven workers who were preparing for an annual outage that was to start from August 14 suffered burn injuries, of who five died. Since around 1975, we, Kansai Electric, have been checking pipe wall thickness while focusing on the thinning of carbon steel piping in the secondary system. Summarizing the results from such investigation and reviewing the latest technical knowledge including operating experience from overseas utilities, we compiled the pipe thickness management guideline for PWR secondary pipes, 1990. The pipe section that ruptured at the Mihama unit 3 should have been included within the inspection scopes according to the guideline but was not registered on the inspection list. It had not been corrected for almost thirty years. As the result, this pipe section had not been inspected even once since the beginning of the plant operation, 1976. It seems that the quality assurance and maintenance management had not functioned well regarding the secondary system piping management, although we were responsible for the safety of nuclear power plants as licensee. We will review the secondary system inspection procedure and also improve the pipe thickness management guideline. And also, we would replace

Plastic fracture mechanics prediction of fracture instability in a circumferentially cracked pipe in bending--2. Experimental verification on a Type 304 stainless steel pipe

International Nuclear Information System (INIS)

Wilkowski, G.M.; Zahoor, A.; Kanninen, M.F.

1980-01-01

The possibility of a pipe fracture emanating from a stress corrosion crack in the heat-affected zones of girth-welds in Type 304 stainless steel pipes was investigated. The J-resistance curve--tearing modulus parameter for the prediction of crack initiation, stable growth and fracture instability--was employed. In the actual experiment, the onset of fracture instability occurred beyond maximum load at an average stable crack growth of 16 to 19 mm (0.63 to 0.75-in.) at each tip. 6 refs

Defect Depth Measurement of Straight Pipe Specimen Using Shearography

International Nuclear Information System (INIS)

Chang, Ho Seob; Kim, Kyung Suk

2012-01-01

In the nuclear industry, wall thinning defect of straight pipe occur the enormous loss in life evaluation and safety evaluation. To use non-destructive technique, we measure deformation, vibration, defect evaluation. But, this techniques are a weak that is the measurement of the wide area is difficult and the time is caught long. In the secondary side of nuclear power plants mostly used steel pipe, artificiality wall thinning defect make in the side and different thickness make to the each other, wall thinning defect part of deformation measure by using shearography. In addition, optical measurement through deformation, vibration, defect evaluation evaluate pipe and thickness defects of pressure vessel is to evaluate quantitatively. By shearography interferometry to measure the pipe's internal wall thinning defect and the variation of pressure use the proposed technique, the quantitative defect is to evaluate the thickness of the surplus. The amount of deformation use thickness of surplus prediction of the actual thickness defect and approximately 7 percent error by ensure reliability. According to pressure the amount of deformation and the thickness of the surplus through DB construction, nuclear power plant pipe use wall thinning part soundness evaluation. In this study, pressure vessel of thickness defect measure proposed nuclear pipe of wall thinning defect prediction and integrity assessment technology development. As a basic research defected theory and experiment, pressure vessel of advanced stability and soundness and maintainability is expected to contribute foundation establishment

Automatic welding machine for piping

International Nuclear Information System (INIS)

Yoshida, Kazuhiro; Koyama, Takaichi; Iizuka, Tomio; Ito, Yoshitoshi; Takami, Katsumi.

1978-01-01

A remotely controlled automatic special welding machine for piping was developed. This machine is utilized for long distance pipe lines, chemical plants, thermal power generating plants and nuclear power plants effectively from the viewpoint of good quality control, reduction of labor and good controllability. The function of this welding machine is to inspect the shape and dimensions of edge preparation before welding work by the sense of touch, to detect the temperature of melt pool, inspect the bead form by the sense of touch, and check the welding state by ITV during welding work, and to grind the bead surface and inspect the weld metal by ultrasonic test automatically after welding work. The construction of this welding system, the main specification of the apparatus, the welding procedure in detail, the electrical source of this welding machine, the cooling system, the structure and handling of guide ring, the central control system and the operating characteristics are explained. The working procedure and the effect by using this welding machine, and the application to nuclear power plants and the other industrial field are outlined. The HIDIC 08 is used as the controlling computer. This welding machine is useful for welding SUS piping as well as carbon steel piping. (Nakai, Y.)

Studying the effect of elastic-plastic strain and hydrogen sulphide on the magnetic behaviour of pipe steels as applied to their testing

Directory of Open Access Journals (Sweden)

Povolotskaya Anna

2018-01-01

Full Text Available The paper reports results of magnetic measurements made on samples of the 12GB pipe steel (strength group X42SS designed for producing pipes to be used in media with high hydrogen sulphide content, both in the initial state and after exposure to hydrogen sulphide, for 96, 192 and 384 hours under uniaxial elastic-plastic tension. At the stage of elastic deformation there is a unique correlation between the coercive force measured on a minor hysteresis loop in weak fields and tensile stress, which enables this parameter to be used for the evaluation of elastic stresses in pipes made of the 12 GB pipe steel under different conditions, including a hydrogen sulphide containing medium. The effect of the value of preliminary plastic strain, viewed as the initial stress-strain state, on the magnetic behaviour of X70 pipe steels under elastic tension and compression is studied. Plastic strain history affects the magnetic behaviour of the material during subsequent elastic deformation since plastic strain induces various residual stresses, and this necessitates taking into account the initial stress-strain state of products when developing magnetic techniques for the determination of their stress-strain parameters during operation.

Steels for nuclear power. I

International Nuclear Information System (INIS)

Bohusova, O.; Brumovsky, M.; Cukr, B.; Hatle, Z.; Protiva, K.; Stefec, R.; Urban, A.; Zidek, M.

1976-01-01

The principles are listed of nuclear reactor operation and the reactors are classified by neutron energy, fuel and moderator designs, purpose and type of moderator. The trend and the development of light-water reactor applications are described. The fundamental operating parameters of the WWER type reactors are indicated. The effect is discussed of neutron radiation on reactor structural materials. The characteristics are described of steel corrosion due to the contact of the steel with steam or sodium in the primary coolant circuit. The reasons for stress corrosion are given and the effects of radiation on corrosion are listed. The requirements and criteria are given for the choice of low-alloy steel for the manufacture of pressure vessels, volume compensators, steam generators, cooling conduits and containment. A survey is given of most frequently used steels for pressure vessels and of the mechanical and structural properties thereof. The basic requirements for the properties of steel used in the primary coolant circuit are as follows: sufficient strength in operating temperature, toughness, good weldability, resistance to corrosion and low brittleness following neutron irradiation. The materials are listed used for the components of light-water and breeder reactors. The production of corrosion-resistant steels is discussed with a view to raw materials, technology, steel-making processes, melting processes, induction furnace steel-making, and to selected special problems of the chemical composition of steels. The effects are mainly discussed of lead, bismuth and tin as well as of some other elements on hot working of high-alloy steels and on their structure. The problems of corrosion-resistant steel welding and of pressure vessel cladding are summed up. Also discussed is the question of the concept and safeguards of the safety of nuclear installation operation and a list is presented of most commonly used nondestructive materials testing methods. The current

Application of numerical analysis technique to make up for pipe wall thinning prediction program

International Nuclear Information System (INIS)

Hwang, Kyeong Mo; Jin, Tae Eun; Park, Won; Oh, Dong Hoon

2009-01-01

Flow Accelerated Corrosion (FAC) leads to wall thinning of steel piping exposed to flowing water or wet steam. Experience has shown that FAC damage to piping at fossil and nuclear plants can lead to costly outages and repairs and can affect plant reliability and safety. CHEWORKS have been utilized in domestic nuclear plants as a predictive tool to assist FAC engineers in planning inspections and evaluating the inspection data to prevent piping failures caused by FAC. However, CHECWORKS may be occasionally left out local susceptible portions owing to predicting FAC damage by pipeline group after constructing a database for all secondary side piping in nuclear plants. This paper describes the methodologies that can complement CHECWORKS and the verifications of the CHECWORKS prediction results in terms of numerical analysis. FAC susceptible locations based on CHECWORKS for the two pipeline groups of a nuclear plant was compared with those of numerical analysis based on FLUENT.

A Hydrogen Ignition Mechanism for Explosions in Nuclear Facility Piping Systems

Energy Technology Data Exchange (ETDEWEB)

Leishear, Robert A.

2013-09-18

Hydrogen explosions may occur simultaneously with water hammer accidents in nuclear facilities, and a theoretical mechanism to relate water hammer to hydrogen deflagrations and explosions is presented herein. Hydrogen and oxygen generation due to the radiolysis of water is a recognized hazard in pipe systems used in the nuclear industry, where the accumulation of hydrogen and oxygen at high points in the pipe system is expected, and explosive conditions may occur. Pipe ruptures in nuclear reactor cooling systems were attributed to hydrogen explosions inside pipelines, i.e., Hamaoka, Nuclear Power Station in Japan, and Brunsbuettel in Germany. Prior to these accidents, an ignition source for hydrogen was not clearly demonstrated, but these accidents demonstrated that a mechanism was, in fact, available to initiate combustion and explosion. A new theory to identify an ignition source and explosion cause is presented here, and further research is recommended to fully understand this explosion mechanism.

Radioactive recontamination on mechanically polished piping at Shimane-1 Nuclear Power Plant

International Nuclear Information System (INIS)

Umeda, K.; Komoto, I.; Imamura, K.; Kataoka, I.; Uchida, S.

1998-01-01

In a series of preventive maintenance tasks for an aging plant, recirculation pipes of Shimane-1 NPP have been replaced by newly fabricated type 316 NG stainless steel pipes. Suppression of shutdown dose rate caused by 60 Co recontamination on the newly replaced piping was one of the major concerns in the recirculation pipe replacement. In order to suppress the shutdown dose rate, control of the 60 Co deposition rate coefficient as well as 60 Co radioactivity in the reactor water are essential. The deposition rate coefficient depends on surface roughness. The coefficient is suppressed by reduction of the effective surface area of pipes through mechanical polishing. Then the inner surface of the pipes was polished mechanically to reduce roughness prior to application in the plant. After measuring and evaluating radioactive recontamination, it was estimated that deposited amounts of radioactive corrosion products on the pipe inner surface would reach the saturated value in a few years, and would not exceed the level before replacement unless water chemistry is degraded. (author)

Validation and Application of Computed Radiography (CR) Tangential Technique for Wall Thickness Measurement of 10 Inch Carbon Steel Pipe

International Nuclear Information System (INIS)

Norhazleena Azaman; Khairul Anuar Mohd Salleh; Amry Amin Abas; Arshad Yassin; Sukhri Ahmad

2016-01-01

Oil and gas industry requires Non Destructive Testing (NDT) to ensure each components, in-service and critical, are fit-for-purpose. Pipes that are used to transfer oil or gas are amongst the critical component that needs to be well maintained and inspected. Typical pipe discontinuities that may lead to unintended incidents are erosion, corrosion, dent, welding defects, etc. Wall thickness assessment, with Radiography Testing (RT) is normally used to inspect such discontinuities and can be performed with two approaches; (a) center line beam tangential technique (b) offset from the centre pipe tangential technique. The latter is a method of choice for this work because of the pipe dimension and limited radiation safe distance at site. Two successful validation approaches (simulation and experimental) were performed to determine the probability of successfulness before the actual RT work with tangential technique is carried out. The pipe was a 10 inch diameter in-service wrapped carbon steel. A 9 Ci Ir-192 and white Imaging Plate (IP) were used as a gamma radiation source and to record the radiographic image. Result of this work suggest that RT with tangential technique for 10 inch wrapped in-service carbon steel pipe can be successfully performed. (author)

Nuclear-piping-repair planning today needs skill, organization

International Nuclear Information System (INIS)

O'Keefe, W.

1986-01-01

Nuclear power plant piping continues to experience failures and imminent threat of failure, despite a high level of care in design, analysis, fabrication, or installation. Continual inspection and surveillance and letter-by-letter following of procedures are not completely effective remedies, either. Both short-time-frame accidents and slowly progressing insidious complaints have caused loss of capacity, availability, and even confidence that the unit will work at close-to-expected performance. The fixes for nuclear-piping complaints cover a wide span, from mere carrying out of well-known repair procedures on either small scale or large, all the way to highly engineered solutions to a problem, with months of study and analysis followed by weighing of alternative methods. With some of the problems, little special planning is necessary. The repair is understood, and the time it needs is well within the envelope of a scheduled outage. Radiation exposure of personnel will not exceed expected moderate limits. And if the repair is a repeat performance of a recent similar one, little can go wrong. The planning for many other repairs, however, is so essential that even a minor failing in it will bring a debacle, with over-run, losses in revenue, and senseless expenditure of man-rems. Look at two types of planning for nuclear piping repair, as revealed at a recent American Welding Society conference on maintenance welding in nuclear power plants

Nuclear power plant piping damping parametric effects

International Nuclear Information System (INIS)

Ware, A.G.

1983-01-01

The present NRC guidelines for structural damping to be used in the dynamic stress analyses of nuclear power plant piping systems are generally considered to be overly conservative. As a result, plant designers have in many instances used a considerable number of seismic supports to keep stresses calculated by large scale piping computer codes below the allowable limits. In response to this problem, the NRC and EG and G Idaho are engaged in programs to evaluate piping system damping, in order to provide more realistic and less conservative values to be used in seismic analyses. To generate revised guidelines, solidly based on technical data, new experimental data need to be generated and assessed, and the parameters which influence piping system damping need to be quantitatively identified. This paper presents the current state-of-the-art knowledge in the United States on parameters which influence piping system damping. Examples of inconsistencies in the data and areas of uncertainty are explained. A discussion of programs by EG and G Idaho and other organizations to evaluate various effects is included, and both short and long range goals of the program are outlined

Numerical simulation of micro-crack occurring in pipe made of stainless steel

Science.gov (United States)

Wotzka, Daria

2017-10-01

Research works carried out regard to studies aiming at determination of the effect of cumulative duty operation on the development of micro-cracks in pipelines for transport of chemical substances. This paper presents results of computer simulations of a pipeline made of stainless steel. The model was investigated using the COMSOL Multiphysics environment. The object under study was divided into sub areas and then discretized according to the FEM method. The physico-chemical parameters of individual areas were defined based on measurement data. The main aim of research works was the modeling of acoustic emission wave, which is emitted in the vicinity of the tip of micro-crack as a result of its development. In order to solve the task, heterogeneity in the structure of the material, referred to damage/micro-crack, causing local stresses was assumed. The local stresses give rise to elastic waves, which propagate in the material in all directions. When the emission waves reach the boundaries of the pipe they are then transferred into acoustic waves and propagate in the surround air, until their natural attenuation. The numerical model takes into account the effect of high pressure (3.6 MPa) and negative temperature (-100°C) of the gas, transported inside the pipe. The influence of changes of these values in the range of ± 20% on the obtained results was investigated. The main contribution of the works is the multiphysical simulation model of transportation pipe made of steel, coupling structural mechanics, thermal conductivity and acoustic waves.

Numerical simulation of micro-crack occurring in pipe made of stainless steel

Directory of Open Access Journals (Sweden)

Wotzka Daria

2017-01-01

Full Text Available Research works carried out regard to studies aiming at determination of the effect of cumulative duty operation on the development of micro-cracks in pipelines for transport of chemical substances. This paper presents results of computer simulations of a pipeline made of stainless steel. The model was investigated using the COMSOL Multiphysics environment. The object under study was divided into sub areas and then discretized according to the FEM method. The physico-chemical parameters of individual areas were defined based on measurement data. The main aim of research works was the modeling of acoustic emission wave, which is emitted in the vicinity of the tip of micro-crack as a result of its development. In order to solve the task, heterogeneity in the structure of the material, referred to damage/micro-crack, causing local stresses was assumed. The local stresses give rise to elastic waves, which propagate in the material in all directions. When the emission waves reach the boundaries of the pipe they are then transferred into acoustic waves and propagate in the surround air, until their natural attenuation. The numerical model takes into account the effect of high pressure (3.6 MPa and negative temperature (-100̊C of the gas, transported inside the pipe. The influence of changes of these values in the range of ± 20% on the obtained results was investigated. The main contribution of the works is the multiphysical simulation model of transportation pipe made of steel, coupling structural mechanics, thermal conductivity and acoustic waves.

Instability predictions for circumferentially cracked Type-304 stainless steel pipes under dynamic loading. Volume 2. Appendixes. Final report. [BWR

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

1982-04-01

This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models.

Ratcheting study in pressurized piping components under cyclic loading at room temperature

International Nuclear Information System (INIS)

Ravi Kiran, A.; Agrawal, M.K.; Reddy, G.R.; Vaze, K.K.; Ghosh, A.K.; Kushwaha, H.S.

2006-07-01

The nuclear power plant piping components and systems are often subjected to reversing cyclic loading conditions due to various process transients, seismic and other events. Earlier the design of piping subjected to seismic excitation was based on the principle of plastic collapse. It is believed that during such events, fatigue-ratcheting is likely mode of failure of piping components. The 1995 ASME Boiler and Pressure Vessel code, Section-III, has incorporated the reverse dynamic loading and ratcheting into the code. Experimental and analytical studies are carried out to understand this failure mechanism. The biaxial ratcheting characteristics of SA 333, Gr. 6 steel and SS 304 stainless steel at room temperature are investigated in the present work. Experiments are carried out on straight pipes subjected to internal pressure and cyclic bending load applied in a three point and four point bend test configurations. A shake table test is also carried out on a pressurized elbow by applying sinusoidal base excitation. Analytical simulation of ratcheting in the piping elements is carried out. Chaboche nonlinear kinematic hardening model is used for ratcheting simulation. (author)

Design specifications for ASME B and PV Code Section III nuclear class 1 piping

International Nuclear Information System (INIS)

Richardson, J.A.

1978-01-01

ASME B and PV Code Section III code regulations for nuclear piping requires that a comprehensive Design Specification be developed for ensuring that the design and installation of the piping meets all code requirements. The intent of this paper is to describe the code requirements, discuss the implementation of these requirements in a typical Class 1 piping design specification, and to report on recent piping failures in operating light water nuclear power plants in the US. (author)

Safety catching device for pipes in missile shielding cylinders of nuclear power plants

International Nuclear Information System (INIS)

Hering, S.; Doll, B.

1976-01-01

The safety catching device consists of a steel wire passed in U-shape around the pipe to be caught and supported by two anchor ties embedded in the concrete of the missile shielding cylinder. This flexible catching device is to cause the energy released in case of a pipe rupture to be absorbed and no dangerous bending shesses to be transferred to the walls of the missile shielding cylinder. (UWI) [de

Acoustic emission reviling and danger level evaluation of stress corrosion cracking in stainless steel pipes

International Nuclear Information System (INIS)

Muravin, Gregory; Muravin, Boris; Lezvinsky, Luidmila

2000-01-01

Breakdowns and catastrophic damage occurring during the operation of nuclear power stations pipelines cause substantial economic and social loss annually throughout the world. Stress corrosion, vibration, fatigue, erosion, water shock, dynamic load, construction defects/errors are the main causes of pipes failures. For these reasons and in view of the age of nuclear power station pipes, there is an increased interest in finding means to prevent potential pipe failures. Nevertheless, statistical data of pipe failures continues to show significant numbers of accidents mainly due to stress corrosion cracking (about 65-80% of total number). To this end, a complex of investigations was carried out for the reliable AE diagnosis of pipes undergone stress corrosion cracking. These include: finding AE indications (fingerprints) of flaws developing in the metal in original condition as well as in metal subjected to stress corrosion; preparing AE criteria for evaluating the danger level of defects. (author)

Method of vertically and horizontally cutting steel pipe piles and removing them based on the development of a steel pipe pile vertically cutting machine; Kokanko tatehoko setsudanki no kaihatsu ni yoru kochi chubu no juo setsudan tekkyo koho

Energy Technology Data Exchange (ETDEWEB)

Watanabe, S.; Takeshita, A.; Kobayashi, K.

1997-07-25

A machine for vertically cutting steel pipe piles has newly been developed for the purpose of removing the end portions the shore protection steel pipe piles which interfere with the shield tunneling work in the Ohokagawa River tunneling section on the Minato Mirai 21 Line. This paper reports the development of the machine for vertically cutting steel pipe piles, and a method of cutting the shield tunneling work hindering piles under the ground by using this machine. The obstacle-constituting portions of the piles are removed by destroying the copings, excavating the interior of the piles to make the same hollow so that a cutting machine can be inserted, and cutting the piles vertically and horizontally. The basic structure of the cutting machine comprises a lower cutting unit for making forward and backward and upward and downward movements of a cutter, and an upper movable unit for controlling the rotation of the cutting unit. The cutting of a pile is done by projecting the cutter by a cylinder the base of which is joined to a cutter driver, and then moving the rotating cutter upward. The amounts of movements of these parts are detected by sensors, and an arbitrary range of the underground portion of a pile can be cut by a remote control operation. 10 figs., 1 tab.

Operating Experience Insights into Pipe Failures for Electro-Hydraulic Control and Instrument Air Systems in Nuclear Power Plant. A Topical Report from the Component Operational Experience, Degradation and Ageing Programme

International Nuclear Information System (INIS)

2015-01-01

Structural integrity of piping systems is important for plant safety and operability. In recognition of this, information on degradation and failure of piping components and systems is collected and evaluated by regulatory agencies, international organisations (e.g. OECD/NEA and IAEA) and industry organisations worldwide to provide systematic feedback for example to reactor regulation and research and development programmes associated with non-destructive examination (NDE) technology, in-service inspection (ISI) programmes, leak-before-break evaluations, risk-informed ISI, and probabilistic safety assessment (PSA) applications involving passive component reliability. Several OECD member countries have agreed to establish the OECD/NEA 'Component Operational Experience, Degradation and Ageing Programme' (CODAP) to encourage multilateral co-operation in the collection and analysis of data relating to degradation and failure of metallic piping and non-piping metallic passive components in commercial nuclear power plants. The scope of the data collection includes service-induced wall thinning, part through-wall cracks, through-wall cracks with and without active leakage, and instances of significant degradation of metallic passive components, including piping pressure boundary integrity. The OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) acts as an umbrella committee of the Project. CODAP is the continuation of the 2002-2011 'OECD/NEA Pipe Failure Data Exchange Project' (OPDE) and the Stress Corrosion Cracking Working Group of the 2006-2010 'OECD/NEA Stress Corrosion Cracking and Cable Ageing Project' (SCAP). OPDE was formally launched in May 2002. Upon completion of the third term (May 2011), the OPDE project was officially closed to be succeeded by CODAP. SCAP was enabled by a voluntary contribution from Japan. It was formally launched in June 2006 and officially closed with an international workshop held in Tokyo in May

The First Assembly Line of Large-longitudinally-welded Steel Pipe in China Went into Operation

Institute of Scientific and Technical Information of China (English)

Li Bing

2002-01-01

@@ On July 27, the first assembly line to produce JCOE large diameter Longitudinally-submerged-arc-welded steel pipe in China, Which is the key homemade equipment project of "West-East Gas Transmission"project, was put into production. Chen Gen, vice general manager of CNPC; Xie Zhiqiang and Liu Haisheng, assistant chief manager of CNPC; Shi Xingquan, vice president of PetroChina; and the president of Itochu-Marubeni Steel & iron Co., Ltd.of Japan; attended the opening ceremony and cut the ribbon.

Nuclear Power Plant Steam Pipes repairing with Tirant 3R Robot System

International Nuclear Information System (INIS)

Ruiz-Martinez, Jose-Tomas; Soto-Tomas, Marcelo; Curiel-Nieva, Marceliano; Monzo-Blasco, Enrique; Pineda-Rodriguez, Salvador; Vaquer-Perez, Juan-Ignacio

2012-09-01

The metallization arc spray process is based on the projection of molten metal, supplied by means of different stainless alloys wire, over a surface of carbon steel usually, with the object of serving as protection against flow assisted corrosion (FAC), increasing resistance to abrasion and deteriorations. A typical application functions covering the steam pipes inner surface in Coal-fired power station and Nuclear Power Plants. The results of this process are spectacular in terms of protection against flow assisted corrosion and abrasion, but its application has conditioning factors, such as: Severe application conditions for workers. Due to the worker's postural position (usually kneeling) in 32' diameter pipes and working with fireproof clothing and masks with outdoor air supplying, due to fumes, sparks and molten metal particles, radiological contamination, confined space, poor lighting... Coating uniformity. As metallization is a manual process, the carried out measurements show small variations in the thickness of the coating, always within the tolerance limits established by the applicable regulations and Quality Assurance. An increase in the uniformity of the projected coating, increase the resistance and give a better surface protection. For all these reasons, Lainsa has developed the TIRANT 3 R system, a worldwide innovative system, for metallization of steam pipes inner surface. TIRANT 3 R system is tele-operated from outside of the pipe, so that human intervention is reduced to the operations of robot positioning and change of metallization wire. As it is an independent system of the human factor, metallization process performance is significantly increased by reducing rest periods due only to the robot maintenance. Likewise, TIRANT 3 R system permits to increase resulting coating uniformity and thus its resistance, keeping selected parameters constant (forward speed, rotation speed and inner surface distance) depending on required type and

Development of engineering program for integrity evaluation of pipes with local wall thinned defects

International Nuclear Information System (INIS)

Park, Chi Yong; Lee, Sung Ho; Kim, Tae Ryong; Park, Sang Kyu

2008-01-01

Integrity evaluation of pipes with local wall thinning by erosion and corrosion is increasingly important in maintenance of wall thinned carbon steel pipes in nuclear power plants. Though a few program for integrity assessment of wall thinned pipes have been developed in domestic nuclear field, however those are limited to straight pipes and methodology proposed in ASME Sec.XI Code Case N-597. Recently, the engineering program for integrity evaluation of pipes with all kinds of local wall defects such as straight, elbow, reducer and branch pipes was developed successfully. The program was designated as PiTEP (Pipe Thinning Evaluation Program), which name was registered as a trademark in the Korea Intellectual Property Office. A developed program is carried out by sequential step of four integrity evaluation methodologies, which are composed of construction code, code case N-597, its engineering method and two developed owner evaluation method. As PiTEP program will be performed through GUI (Graphic User Interface) with user's familiarity, it would be conveniently used by plant engineers with only measured thickness data, basic operation conditions and pipe data

Development of stainless steels for nuclear power plant - Advanced nuclear materials development -

International Nuclear Information System (INIS)

Hong, Jun Hwa; Ryu, Woo Seog; Chi, Se Hwan; Lee, Bong Sang; Oh, Yong Jun; Byun, Thak Sang; Oh, Jong Myung

1994-07-01

This report reviews the status of R and D and the material specifications of nuclear components in order to develop the stainless steels for nuclear applications, and the technology of computer-assisted alloy design is developed to establish the thermodynamic data of Fe-Cr-Ni-Mo-Si-C-N system which is the basic stainless steel systems. High strength and corrosion resistant stainless steels, 316LN and super clean 347, are developed, and the manufacturing processes and heat treatment conditions are determined. In addition, a martensitic steel is produced as a model alloy for turbine blade, and characterized. The material properties showed a good performance for nuclear applications. (Author)

Mechanism of selective corrosion in electrical resistance seam welded carbon steel pipe

Energy Technology Data Exchange (ETDEWEB)

Lopez Fajardo, Pedro; Godinez Salcedo, Jesus; Gonzalez Velasquez, Jorge L. [Instituto Politecnico Nacional, Mexico D.F., (Mexico). Escuela Superior de Ingenieria Quimica e Industrias Extractivas. Dept. de Ingenieria Metalurgica

2009-07-01

In this investigation the studies of the mechanism of selective corrosion in electrical resistance welded (ERW) carbon steel pipe was started. Metallographic characterizations and evaluations for inclusions were performed. The susceptibility of ERW pipe to selective corrosion in sea water (NACE 1D182, with O{sub 2} or CO{sub 2} + H{sub 2}S) was studied by the stepped potential Potentiostatic electrochemical test method in samples of 1 cm{sup 3} (ASTM G5) internal surface of the pipe (metal base-weld). The tests were looking for means for predicting the susceptibility of ERW pipe to selective corrosion, prior to placing the pipeline in service. Manganese sulfide inclusions are observed deformed by the welding process and they are close to the weld centerline. A slight decarburization at the weld line is observed, and a distinct out bent fiber pattern remains despite the post-weld seam annealing. The microstructure of the weld region consists of primarily polygonal ferrite grains mixed with small islands of pearlite. It is possible to observe the differences of sizes of grain of the present phases in the different zones. Finally, scanning electron microscopic observation revealed that the corrosion initiates with the dissolution of MnS inclusions and with small crack between the base metal and ZAC. (author)

Steels and welding nuclear

International Nuclear Information System (INIS)

Sessa, M.; Milella, P.P.

1987-01-01

This ENEA Data-Base regards mechanical properties, chemical composition and heat treatments of nuclear pressure vessel materials: type A533-B, A302-B, A508 steel plates and forgings, submerged arc welds and HAZ before and after nuclear irradiation. Irradiation experiments were generally performed in high flux material test reactors. Data were collected from international available literature about water nuclear reactors pressure vessel materials embrittlement

Application of tearing instability analysis for complex crack geometries in nuclear piping

International Nuclear Information System (INIS)

Pan, J.; Wilkowski, G.

1984-01-01

The analysis of the experimental data of 304 stainless steel pipes using Zahoor and Kanninen's estimation scheme has shown that the J resistance curve of a circumferentially cracked pipe with a simulated internal surface crack around the remaining net section is much lower than the J resistance curve of pipes with a idealized through-wall crack (without a simulated internal surface crack). The implications of the low J at initiation and tearing modulus on the stability analysis of typical BWR piping systems are discussed on the condition that an internal circumferential surface crack is assumed to occur along with a circumferential through-wall crack due to stress corrosion. The results presented here show that the margin of safety is reduced and in some cases instability is predicted due to the low J resistance curve and tearing modulus

Effect of ferrite on the precipitation of σ phase in cast austenitic stainless steel used for primary coolant pipes of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Wang, Yongqiang; Li, Na, E-mail: wangyongqiang1124@163.com [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology, Beijing (China)

2017-11-15

The effect of ferrite phase on the precipitation of σ phase in a Z3CN20.09M cast austenitic stainless steel (CASS) used for primary coolant pipes of pressurized water reactor (PWR) nuclear power plants was investigated by using isothermal heat-treatment, optical microscopy (OM), transmission electron microscopy (TEM) and electron probe microanalysis (EPMA) techniques. The influence of different morphologies and volume fractions of ferrite in the σ phase formation mechanism was discussed. The amount of σ phase precipitated in all specimens with different microstructures increased with increasing of aging time, however, the precipitation rate is significant different. The formation of σ phase in specimens with the coarsest ferrite and the dispersively smallest ferrite is slowest. The lowest level Cr content in ferrite and fewest α/γ interfaces in specimen are the main reasons for the slowest σ precipitation due to they are unfavorable for the kinetics and thermodynamics of phase transformation respectively. By contraries, the fastest formation of σ phase takes place in specimens with narrow and long ferrite due to the most α/γ interfaces and higher Cr content in ferrite which are beneficial for preferential nucleation and formation thermodynamics of σ phase. (author)

75 FR 78216 - Certain Circular Welded Non-Alloy Steel Pipe From Mexico: Preliminary Results of Antidumping Duty...

Science.gov (United States)

2010-12-15

... Non-Alloy Steel Pipe from Brazil, the Republic of Korea (Korea), Mexico, and Venezuela and Amendment..., Director, Office 7 to Michael Walsh, Director, AD/CVD Revenue Policy & Programs, U.S. Customs and Border...

Viscoelastic behavior and durability of steel wire - reinforced polyethylene pipes under a high internal pressure

NARCIS (Netherlands)

Ivanov, S.; Anoshkin, A.N.; Zuyko, V.Yu

2011-01-01

The strength tests of steel-wire-reinforced polyethylene pipe specimens showed that, under a constant internal pressure exceeding 80% of their short-term ultimate pressure, the fracture of the specimens occurred in less than 24 hours. At pressures slightly lower than this level, some specimens did

Evaluation of J-R curve testing of nuclear piping materials using the direct current potential drop technique

International Nuclear Information System (INIS)

Hackett, E.M.; Kirk, M.T.; Hays, R.A.

1986-08-01

A method is described for developing J-R curves for nuclear piping materials using the DC Potential Drop (DCPD) technique. Experimental calibration curves were developed for both three point bend and compact specimen geometries using ASTM A106 steel, a type 304 stainless steel and a high strength aluminum alloy. These curves were fit with a power law expression over the range of crack extension encountered during J-R curve tests (0.6 a/W to 0.8 a/W). The calibration curves were insensitive to both material and sidegrooving and depended solely on specimen geometry and lead attachment points. Crack initiation in J-R curve tests using DCPD was determined by a deviation from a linear region on a plot of COD vs. DCPD. The validity of this criterion for ASTM A106 steel was determined by a series of multispecimen tests that bracketed the initiation region. A statistical differential slope procedure for determination of the crack initiation point is presented and discussed. J-R curve tests were performed on ASTM A106 steel and type 304 stainless steel using both the elastic compliance and DCPD techniques to assess R-curve comparability. J-R curves determined using the two approaches were found to be in good agreement for ASTM A106 steel. The applicability of the DCPD technique to type 304 stainless steel and high rate loading of ferromagnetic materials is discussed. 15 refs., 33 figs

77 FR 18266 - Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines; Revised...

Science.gov (United States)

2012-03-27

... INTERNATIONAL TRADE COMMISSION [Investigation Nos. 731-TA-865-867 (Second Review)] Stainless Steel Butt-Weld Pipe Fittings From Italy, Malaysia, and the Philippines; Revised Schedule for the Subject Reviews AGENCY: United States International Trade Commission. ACTION: Notice. DATES: Effective Date: March...

Erosion/corrosion-induced pipe wall thinning in US Nuclear Power Plants

International Nuclear Information System (INIS)

Wu, P.C.

1989-04-01

Erosion/corrosion in single-phase piping systems was not clearly recognized as a potential safety issue before the pipe rupture incident at the Surry Power Station in December 1986. This incident reminded the nuclear industry and the regulators that neither the US Nuclear Regulatory Commission (NRC) nor Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code require utilities to monitor erosion/corrosion in the secondary systems of nuclear power plants. This report provides a brief review of the erosion/corrosion phenomenon and its major occurrence in nuclear power plants. In addition, efforts by the NRC, the industry, and the ASME Section XI Committee to address this issue are described. Finally, results of the survey and plant audits conducted by the NRC to assess the extent of erosion/corrosion-induced piping degradation and the status of program implementation regarding erosion/corrosion monitoring are discussed. This report will support a staff recommendation for an additional regulatory requirement concerning erosion/corrosion monitoring. 21 refs., 3 tabs

Residual stress distribution in carbon steel pipe welded joint measured by neutron diffraction

International Nuclear Information System (INIS)

Hayashi, Makoto; Ishiwata, Masayuki; Morii, Yukio; Minakawa, Nobuaki

2000-01-01

In order to estimate crack growth behavior of fatigue and stress corrosion cracking in pipes, the residual stress distribution near the pipe weld region has to be measured through the wall thickness. Since the penetration depth of neutron is deep enough to pass through the thick pipe wall, the neutron diffraction technique for the residual stress measurement is effective for this purpose. At the first step the residual stress distribution near the weld region in a butt-welded carbon steel pipe was measured by the neutron diffraction. Significant stresses extended only to a distance of 30 mm from the center of the weld. The major tensile stresses occurred in the hoop direction in the fusion and heat affected zones of the weldment, and they attained a level greater than 200 MPa through the thickness. While the axial residual stress at the inside surface was 50 MPa, the stress at the outside surface was -100 MPa. The comparison of residual stress distributions measured by the neutron diffraction, the X-ray diffraction and the strain gauge method reveals that the neutron diffraction is the most effective for measuring the residual stress inside the structural components. (author)

Novel manufacturing method by using stainless steel pipes expanded into aluminium profiles for the ITER Neutral Beam cryopumps

Energy Technology Data Exchange (ETDEWEB)

Dremel, Matthias, E-mail: matthias.dremel@iter.org; Boissin, Jean-Claude; Déléage, Vincent; Quinn, Eamonn; Pearce, Robert

2015-10-15

This paper describes the novel engineering and manufacturing solution of stainless steel pipe expansion into aluminium extrusion profiles for use at cryogenic temperatures up to 400 K. This fabrication method will be used for the thermal radiation shields and the cryopanels of the ITER Neutral Beam cryopumps. The use of stainless steel pipes expanded into aluminium extrusion profiles is a solution that combines standard stainless steel welding procedures for the manifolds of the cooling circuits with extended aluminium structures taking advantage of the high thermal conductivity of aluminium. The cryogenic cooling circuits of the pump are a first confinement barrier in the ITER vacuum vessel and the risk of a leakage needs to be minimized as far as possible. The expansion method avoids the need of joints of dissimilar materials in the primary confinement barrier. The fabrication method and results of the prototyping of full scaled components for the ITER Neutral Beam cryopumps are outlined in this paper.

Failure rates in piping manufactured to different standards

International Nuclear Information System (INIS)

Barnes, R.W.; Cooper, G.D.

1995-11-01

Most non-nuclear process piping systems in Canada and the United States are constructed to the requirements of the piping codes of the American Society of Mechanical Engineers (ASME B31.1 and B31.3). Section III of the ASME Boiler and Pressure Vessel Code, has additional requirements for piping that are expected to provide further assurance of pressure boundary integrity. This project attempted to determine if the additional requirements of Section III were beneficial in preventing failure of the pressure boundary. The approach taken in the study was to determine the causes of failure of non-nuclear piping subjected to service similar to that experienced by piping in CANDU nuclear power plants. The study examined information on carbon steel piping systems filled with water/steam which operate up to a maximum temperature of 600 F and a maximum pressure of 1600 psi. The failure mechanisms were identified and analysed to determine whether application of the requirements of Section III would have prevented the failure. Through a process of interviews and literature search, 186 failures were identified and assembled into a reference database. Many of the records were incomplete; therefore, the reference database was trimmed to include a subset of 65 failure points supported by complete data. This subset formed the basis for this study. The results from the study of other databases assembled for similar purposes were reviewed and compared to the conclusions reached in this study. These reviews confirmed the conclusions reached in this study. (author). 48 refs., 20 tabs

Application of ultrasonic NDT technique for butt fusion joints of plastic pipes

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Gyung; Lim, Hyung Taik; Choi, Jeong Guen; Lee, Jae Myung [ANSCO, Daejeon (Korea, Republic of)

2016-05-15

The long term-durability and the soundness of the plastic pipe have been sufficiently verified through use of the plastic pipe during the past several decades. Recently, as the pipe material have been constantly developed, the application of the plastic pipe is expanded to various industrial fields, such as an increase of a use pressure, the bigger diameter of the pipe, etc. In a nuclear power plant where a carbon steel pipe and a stainless steel pipe are mainly used as a safety class III buried pipe, safety in an operation is seriously threatened by abrasion, heat deterioration or the like, frequently generated in a metal pipe. Therefore, in order to provide an alternate to this problem there is a rising interest on using high-density polyethylene (HDPE) pipes which are known to provide much enhanced corrosion, abrasion and impact resistant properties. With polyethylene piping, one of the issue that is being looked into is the integrity of butt-fusion joint. At the present time the Referencing Code and Standard for the NDT technology of for butt fusion joint of the plastic pipe have not yet established. An optimum inspection parameters were determined according to the thickness of the HDPE pipe. It was also confirmed that most of the detection results of two techniques have matched with each other. In the PAUT, it is easy to distinguish signals with from the flaws made by the thin plate and the void. Also the resolving power of PAUT on the detection in the depth direction has been demonstrated to be satisfactory.

Leak-before-break behaviour of nuclear piping systems

International Nuclear Information System (INIS)

Bartholome, G.; Wellein, R.

1992-01-01

The general concept for break preclusion of nuclear piping systems in the FRG consists of two main prerequisites: Basic safety; independent redundancies. The leak-before-break behaviour is open of these redundancies and will be verified by fracture mechanics. The following items have to be evaluated: The growth of detected and postulated defects must be negligible in one life time of the plant; the growth behaviour beyond design (i.e. multiple load collectives are taken into account) leads to a stable leak; This leakage of the piping must be detected by an adequate leak detection system long before the critical defect size is reached. The fracture mechanics calculations concerning growth and instability of the relevant defects and corresponding leakage areas are described in more detail. The leak-before-break behaviour is shown for two examples of nuclear piping systems in pressurized water reactors: main coolant line of SIEMENS-PWR 1300 MW (ferritic material, diameter 800 mm); surge line of Russian WWER 440 (austenitic material, diameter 250 mm). The main results are given taking into account the relevant leak detection possibilities. (author). 9 refs, 9 figs

Variation behavior of residual stress distribution by manufacturing processes in welded pipes of austenitic stainless steel

International Nuclear Information System (INIS)

Ihara, Ryohei; Hashimoto, Tadafumi; Mochizuki, Masahito

2012-01-01

Stress corrosion cracking (SCC) has been observed near heat affected zone (HAZ) of primary loop recirculation pipes made of low-carbon austenitic stainless steel type 316L in the nuclear power plants. For the non-sensitization material, residual stress is the important factor of SCC, and it is generated by machining and welding. In the actual plants, welding is conducted after machining as manufacturing processes of welded pipes. It could be considered that residual stress generated by machining is varied by welding as a posterior process. This paper presents residual stress variation due to manufacturing processes of pipes using X-ray diffraction method. Residual stress distribution due to welding after machining had a local maximum stress in HAZ. Moreover, this value was higher than residual stress generated by welding or machining. Vickers hardness also had a local maximum hardness in HAZ. In order to clarify hardness variation, crystal orientation analysis with EBSD method was performed. Recovery and recrystallization were occurred by welding heat near the weld metal. These lead hardness decrease. The local maximum region showed no microstructure evolution. In this region, machined layer was remained. Therefore, the local maximum hardness was generated at machined layer. The local maximum stress was caused by the superposition effect of residual stress distributions due to machining and welding. Moreover, these local maximum residual stress and hardness are exceeded critical value of SCC initiation. In order to clarify the effect of residual stress on SCC initiation, evaluation including manufacturing processes is important. (author)

Corrosion resistance of ERW (Electric Resistance Welded) seam welds as compared to metal base in API 5L steel pipes

Energy Technology Data Exchange (ETDEWEB)

Gonzalez Velasquez, Jorge L.; Godinez Salcedo, Jesus G.; Lopez Fajardo, Pedro [Instituto Politecnico Nacional (IPN), Mexico D.F. (Mexico). Escuela Superior de Ingenieria Quimica e Industrias Extractivas (ESIQIE). Dept. de Ingenieria Metalurgica

2009-07-01

The corrosion resistance of ERW seam welds and the base metal in API 5L X70 steel pipes was evaluated by Tafel tests. The procedure was according to ASTM G3 standard. The study was completed with metallographic and chemical characterization of the tested zones, that is, the welded zone and the base metal away of the weld. All tests were made on the internal surface of the pipe in order to assess the internal corrosion of an in-service pipeline made of the API 5L X70 steel. The test solution was acid brine prepared according to NACE Publications 1D182 and 1D196. The results showed that the ERW seam weld corrodes as much as three times faster than the base material. This behavior is attributed to a more heterogeneous microstructure with higher internal energy in the ERW seam weld zone, as compared to the base metal, which is basically a ferrite pearlite microstructure in a normalized condition. This result also indicates that pipeline segments made of ERW steel pipe where the seam weld is located near or at the bottom of the pipe are prone to a highly localized attack that may form channels of metal loss if there is water accumulation at the bottom of the pipeline. (author)

[Effect of chloramines disinfection for biofilm formation control on copper and stainless steel pipe materials].

Science.gov (United States)

Zhou, Ling-ling; Zhang, Yong-ji; Li, Xing; Li, Gui-bai

2008-12-01

Two rotating annular bioreactors (RABs) with copper and stainless steel pipe materials were adopted in the study, the effects of these two pipe materials and chloramines disinfection on biofilms formation in drinking water distribution system were evaluated. The maximum viable bacterial number in biofilm of copper and stainless steel reached 5.5 x 10(3) CFU/cm2 and 2.5 x 10(5) CFU/cm2 at 18th and 21st day without chloramines, and the viable bacterial number at the apparent steady state was 1.0 x 10(3) CFU/cm2 and 1.3 x 10(5) CFU/cm2 respectively. It was obvious that the biomass on copper materials was lower than that of the stainless steel. The maximum viable bacterial on copper and stainless steel under chloramines was 5.0 x 10(2) CFU/cm2 and 5.0 x 10(4) CFU/cm2, which was one order of magnitude lower than that of without chloramines, and its number was 10 CFU/cm2 and 3.5 x 10(4) CFU/cm2 at the steady state. These results illustrated that chloramines had apparent ability in controlling biomass when the biofilm was on steady states, especially for copper material. There was exponential relationship between biomass in biofilm and residue chloramines, which meant less biomass with more chloramines, synergistic effects were observed between chloramines and copper materials on biomass in biofilms inactivation.

Creep properties in similar weld joint of a thick-walled P92 steel pipe

Czech Academy of Sciences Publication Activity Database

Sklenička, Václav; Kuchařová, Květa; Svobodová, M.; Kvapilová, Marie; Král, Petr; Horváth, P.

2016-01-01

Roč. 119, č. 1 (2016), s. 1-12 ISSN 1044-5803 R&D Projects: GA ČR(CZ) GA16-09518S; GA MPO FR-TI4/406 Institutional support: RVO:68081723 Keywords : 9–12%Cr steels * Creep testing * High temperature creep * Thick-walled pipe * Welding Subject RIV: JG - Metallurgy Impact factor: 2.714, year: 2016

Development of seamless forged pipe and fitting for BWR recirculation loop piping with improved resistance to intergranular stress corrosion cracking

International Nuclear Information System (INIS)

Ohnishi, Keizo; Tsukada, Hisashi; Kobayashi, Masayoshi; Iwadate, Tadao; Ono, Shinichi

1981-01-01

As a primary remedy for IGSCC of primary loop piping, especially Recirculation Loop Piping of BWR, extra low carbon stainless steel with high nitrogen content has become to be used. While, in order to make In-service Inspection easier and complete, new design of piping which decrease both number and total length of weld line has been considered. Japan Steel Works has developed the research on large size seamless forged pipe and fitting made from high nitrogen extra low carbon 316 stainless steel. This paper describes the key points of manufacturing technology as well as the material properties, especially strength and intergranular-corrosion and intergranular- stress-corrosion-cracking-resistivities of these forged pipe and fitting. (author)

PPOOLEX experiments with two parallel blowdown pipes

Energy Technology Data Exchange (ETDEWEB)

Laine, J.; Puustinen, M.; Raesaenen, A. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

2011-01-15

This report summarizes the results of the experiments with two transparent blowdown pipes carried out with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through either one or two vertical transparent blowdown pipes to the condensation pool. Five experiments with one pipe and six with two parallel pipes were carried out. The main purpose of the experiments was to study loads caused by chugging (rapid condensation) while steam is discharged into the condensation pool filled with sub-cooled water. The PPOOLEX test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. In the experiments the initial temperature of the condensation pool water varied from 12 deg. C to 55 deg. C, the steam flow rate from 40 g/s to 1 300 g/s and the temperature of incoming steam from 120 deg. C to 185 deg. C. In the experiments with only one transparent blowdown pipe chugging phenomenon didn't occur as intensified as in the preceding experiments carried out with a DN200 stainless steel pipe. With the steel blowdown pipe even 10 times higher pressure pulses were registered inside the pipe. Meanwhile, loads registered in the pool didn't indicate significant differences between the steel and polycarbonate pipe experiments. In the experiments with two transparent blowdown pipes, the steamwater interface moved almost synchronously up and down inside both pipes. Chugging was stronger than in the one pipe experiments and even two times higher loads were measured inside the pipes. The loads at the blowdown pipe outlet were approximately the same as in the one pipe cases. Other registered loads around the pool were about 50-100 % higher than with one pipe. The experiments with two parallel blowdown pipes gave contradictory results compared to the earlier studies dealing with chugging loads in case of multiple pipes. Contributing

Development of pipe welding, cutting and inspection tools for the ITER blanket

International Nuclear Information System (INIS)

Oka, Kiyoshi; Ito, Akira; Taguchi, Kou; Takiguchi, Yuji; Takahashi, Hiroyuki; Tada, Eisuke

1999-07-01

In D-T burning reactors such as International Thermonuclear Experimental Reactor (ITER), an internal access welding/cutting of blanket cooling pipe with bend sections is inevitably required because of spatial constraint due to nuclear shield and available port opening space. For this purpose, internal access pipe welding/cutting/inspection tools for manifolds and branch pipes are being developed according to the agreement of the ITER R and D task (T329). A design concept of welding/cutting processing head with a flexible optical fiber has been developed and the basic feasibility studies on welding, cutting and rewelding are performed using stainless steel plate (SS316L). In the same way, a design concept of inspection head with a non-destructive inspection probe (including a leak-testing probe) has been developed and the basic characteristic tests are performed using welded stainless steel pipes. In this report, the details of welding/cutting/inspection heads for manifolds and branch pipes are described, together with the basic experiment results relating to the welding/cutting and inspection. In addition, details of a composite type optical fiber, which can transmit both the high-power YAG laser and visible rays, is described. (author)

Ductile fracture mechanics methodology for complex cracks in nuclear piping

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1988-02-01

Limit load and J-integral estimation solutions are developed for circumferentially complex-cracked pipes in bending. The limit load solution is developed using thick-walled cylinder analysis which included the effects of flaw depth accurately. J-integral estimation solutions are developed that are suitable for a wide range of loading from linear elastic, elastic-plastic to net-section yielding of the flawed section. Mode I stress intensity factor solution is developed from experimental compliance data. Two types of J solutions are developed. First, J solutions for determining the J-resistance curve from single load-displacement record are presented. Next, elastic-plastic J solution in the format of EPRI J estimation scheme is presented. The latter solution was used to predict the load carrying capacity of complex-cracked pipes made of Type-304 stainless steel, Inconel 600, and A106 GrB materials. Predictions were compared against pipe tests to demonstrate the accuracy of the limit load and J estimation solutions.

Ductile fracture mechanics methodology for complex cracks in nuclear piping

International Nuclear Information System (INIS)

Zahoor, A.

1988-01-01

Limit load and J-integral estimation solutions are developed for circumferentially complex-cracked pipes in bending. The limit load solution is developed using thick-walled cylinder analysis which included the effects of flaw depth accurately. J-integral estimation solutions are developed that are suitable for a wide range of loading from linear elastic, elastic-plastic to net-section yielding of the flawed section. Mode I stress intensity factor solution is developed from experimental compliance data. Two types of J solutions are developed. First, J solutions for determining the J-resistance curve from single load-displacement record are presented. Next, elastic-plastic J solution in the format of EPRI J estimation scheme is presented. The latter solution was used to predict the load carrying capacity of complex-cracked pipes made of Type-304 stainless steel, Inconel 600, and A106 GrB materials. Predictions were compared against pipe tests to demonstrate the accuracy of the limit load and J estimation solutions. (orig.)

Nuclear energy and the steel industry

International Nuclear Information System (INIS)

Barnes, R.S.

1977-01-01

Fossil fuels represent a large part of the cost of iron and steel making and their increasing cost has stimulated investigation of methods to reduce the use of fossil fuels in the steel industry. Various iron and steel making routes have been studied by the European Nuclear Steelmaking Club (ENSEC) and others to determine to what extent they could use energy derived from a nuclear reactor to reduce the amount of fossil fuel consumed. The most promising concept is a High-Temperature Gas-Cooled Nuclear Reactor heating helium to a temperature sufficient to steam reform hydrocarbons into reducing gases for the direct reduction of iron ores. It is proposed that the reactor/reformer complex should be separate from the direct-reduction plant/steelworks and should provide reducing gas by pipeline, not only to a number of steel works but to other industrial users. The composition of suitable reducing gases and the methods of producing them from various feedstocks are discussed. Highly industrialised countries with large steel and chemical industries have shown greatest interest in the concept, but those countries with large iron-ore reserves and growing direct capacity should consider the future value of the High-Temperature Gas-Cooled Reactor as a means of extending the life of their gas reserves. (author)

Electron Cloud in Steel Beam Pipe vs Titanium Nitride Coated and Amorphous Carbon Coated Beam Pipes in Fermilab's Main Injector

Energy Technology Data Exchange (ETDEWEB)

Backfish, Michael

2013-04-01

This paper documents the use of four retarding field analyzers (RFAs) to measure electron cloud signals created in Fermilab’s Main Injector during 120 GeV operations. The first data set was taken from September 11, 2009 to July 4, 2010. This data set is used to compare two different types of beam pipe that were installed in the accelerator. Two RFAs were installed in a normal steel beam pipe like the rest of the Main Injector while another two were installed in a one meter section of beam pipe that was coated on the inside with titanium nitride (TiN). A second data run started on August 23, 2010 and ended on January 10, 2011 when Main Injector beam intensities were reduced thus eliminating the electron cloud. This second run uses the same RFA setup but the TiN coated beam pipe was replaced by a one meter section coated with amorphous carbon (aC). This section of beam pipe was provided by CERN in an effort to better understand how an aC coating will perform over time in an accelerator. The research consists of three basic parts: (a) continuously monitoring the conditioning of the three different types of beam pipe over both time and absorbed electrons (b) measurement of the characteristics of the surrounding magnetic fields in the Main Injector in order to better relate actual data observed in the Main Injector with that of simulations (c) measurement of the energy spectrum of the electron cloud signals using retarding field analyzers in all three types of beam pipe.

Nuclear power plant piping prefabrication and assembly

International Nuclear Information System (INIS)

Schmidt, H.

1990-01-01

The piping design for nuclear power plants projects reveals, at the beginning, a modification through the application of new fabrication techniques for prefabrication and assembly. This report presents a fabrication methodology which aims to minimize the fabrication and assembly costs as well as to improve and assure quality. (Author) [es

Seismic ratchet-fatigue failure of piping systems

International Nuclear Information System (INIS)

Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

1986-01-01

Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-in. and a pressurized 6-in. diameter carbon steel nuclear pipe systems subjected to high level shaking have been accomplished. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occurred in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate very well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules may be appropriate to cover the ratchet-fatigue failure mode

Thinned pipe management program of Korean NPPs

International Nuclear Information System (INIS)

Lee, S.H.; Kim, T.R.; Jeon, S.C.; Hwang, K.M.

2003-01-01

Wall thinning of carbon steel pipe components due to Flow-Accelerated Corrosion (FAC) is one of the most serious threats to the integrity of steam cycle systems in Nuclear Power Plants (NPP). If the thickness of a pipe component is reduced below the critical level, it cannot sustain stress and consequently results in leakage or rupture. In order to minimize the possibility of excessive wall thinning, Thinned Pipe Management Program (TPMP) has been set up and being implemented to all Korean NPPs. Important elements of the TPMP include the prediction of the FAC rate for each component based on model analysis, prioritization of pipe components for inspection, thickness measurement, calculation of wear and wear rate for each component. Additionally, decision making associated with replacement or continuous service for thinned pipe components and establishment of long-term strategic management plan based on diagnosis of plant condition regarding overall wall thinning also are essential part of the TPMP. From pre-service inspection data, it has been found that initial thickness is varies, which influences wear and wear rate calculations. (author)

Prediction on corrosion rate of pipe in nuclear power system based on optimized grey theory

International Nuclear Information System (INIS)

Chen Yonghong; Zhang Dafa; Chen Dengke; Jiang Wei

2007-01-01

For the prediction of corrosion rate of pipe in nuclear power system, the pre- diction error from the grey theory is greater, so a new method, optimized grey theory was presented in the paper. A comparison among predicted results from present and other methods was carried out, and it is seem that optimized grey theory is correct and effective for the prediction of corrosion rate of pipe in nuclear power system, and it provides a fundamental basis for the maintenance of pipe in nuclear power system. (authors)

A mechanism for corrosion product deposition on the carbon steel piping in the residual heat removal system of BWRs

International Nuclear Information System (INIS)

Aizawa, Motohiro; Chiba, Yoshinori; Hosokawa, Hideyuki; Ohsumi, Katsumi; Uchida, Shunsuke; Ishizawa, Noboru

2002-01-01

The dose rate of the residual heat removal (RHR) piping has been considered to be caused by accumulation of insoluble (crud) radioactive corrosion products on carbon steel surfaces. Soft shutdown procedures (i.e., plant shutdown with moderate coolant temperature reduction rate) used to be applied to reduce crud radioactivity release from the fuel surface, but these are no longer used because of the need for shorter plant shutdown times. In order to apply other suitable countermeasures to reduce RHR dose rate, assessment of plant data, experiments on deposition of crud and ion species on carbon steel, and mass balance evaluation of radioactive corrosion products based on plant and laboratory data were carried out and the following findings were made. (1) Deposits of ion species on carbon steel surfaces of the RHR piping was much more numerous than for crud. (2) Ion species accumulation behavior on RHR piping, which is temperature dependent, can be evaluated with the calculation model used for the dehydration reaction of corrosion products generated during the wet lay-up period. (3) Deposition amounts could be reduced to 1/2.5 when the starting RHR system operation temperature was lowered from 155degC to 120degC. (author)

Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method

International Nuclear Information System (INIS)

Lin, Y.C.; Xie, Y.J.; Wang, X.H.

2004-01-01

Failure analysis of in-service nuclear piping containing defects is an important subject in the nuclear power plants. Considering the uncertainties in various internal operating loadings and external forces, including earthquake and wind, flaw sizes, material fracture toughness and flow stress, this paper presents a probabilistic assessment methodology for in-service nuclear piping containing defects, which is especially designed for programming. A general sampling computation method of the stress intensity factor (SIF), in the form of the relationship between the SIF and the axial force, bending moment and torsion, is adopted in the probabilistic assessment methodology. This relationship has been successfully used in developing the software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), based on a well-known engineering safety assessment procedure R6. A numerical example is given to show the application of the SAPP-2003 software. The failure probabilities of each defect and the whole piping can be obtained by this software

Study to define NDE research for inspection of stainless steels

International Nuclear Information System (INIS)

Reinhart, E.R.

1978-08-01

After the boiling water reactor (BWR) stress corrosion cracking incidents on 4- and 10-inch stainless steel piping, the Electric Power Research Institute (EPRI) organized a round-robin ultrasonic examination of piping removed from service (TPS-75-609). Five inspection teams participated in this program, using both a standard procedure and the individual team procedure. The original intent was to section the piping after the program to evaluate the effectiveness of state-of-the-art ultrasonics in finding stress corrosion cracking. The sectioning was delayed, however, to allow research and development (R and D) groups time to perform basic measurements aimed at determining optimum search unit and instrument characteristics for the ultrasonic examination of stainless steel piping and to study the applicability of various advanced inspection methods. This additional effort was funded as part of an EPRI technical planning study (TPS-75-620), A Study to Define NDE Research for Inspection of Stainless Steels. Inspection methods evaluated in this study included (1) processing of manual scan data using a miniature programmable calculator (Aerojet Nuclear); (2) investigation into the performance characteristics of three experimental ultrasonic transducers (Battelle-Columbus Laboratories); (3) analysis of fundamental ultrasonic response data from intergranular stress corrosion cracks in stainless steels (Southwest Research Institute); and (4) a feasibility study of advanced signal processing and pattern recognition for analyzing flaws in stainless steel piping (Ultrasonics International). The results of the studies compiled in the report have indicated the direction for future research and development and have formed the basis for the recently initiated EPRI Research Project 892, Ultrasonic System Optimization

An experience with in-service fabrication and inspection of austenitic stainless steel piping in high temperature sodium system

Energy Technology Data Exchange (ETDEWEB)

Ravi, S., E-mail: sravi@igcar.gov.in; Laha, K.; Sakthy, S.; Mathew, M.D.; Bhaduri, A.K.

2015-04-01

Highlights: • Procedure for changing 304L SS pipe to 316L SS in sodium loop has been established. • Hot leg made of 304L SS was isolated from existing cold leg made of 316LN SS. • Innovative welding was used in joining the new 316L SS pipe with existing 316LN SS. • The old components of 304L SS piping have been integrated with the new piping. - Abstract: A creep testing facility along with dynamic sodium loop was installed at Indira Gandhi Centre for Atomic Research, Kalpakkam, India to assess the creep behavior of fast reactor structural materials in flowing sodium. Type 304L austenitic stainless steel was used in the low cross section piping of hot-leg whereas 316LN austenitic stainless steel in the high cross section cold-leg of the sodium loop. The intended service life of the sodium loop was 10 years. The loop has performed successfully in the stipulated time period. To enhance its life time, it has been decided to replace the 304L piping with 316L piping in the hot-leg. There were more than 300 welding joints involved in the integration of cold-leg with the new 316L hot-leg. Continuous argon gas flow was maintained in the loop during welding to avoid contamination of sodium residue with air. Several innovative welding procedures have been adopted for joining the new hot-leg with the existing cold-leg in the presence of sodium residue adopting TIG welding technique. The joints were inspected for 100% X-ray radiography and qualified by performing tensile tests. The components used in the discarded hot-leg were retrieved, cleaned and integrated in the renovated loop. A method of cleaning component of sodium residue has been established. This paper highlights the in-service fabrication and inspection of the renovation.

Characterization of thermal aging of duplex stainless steel by SQUID

International Nuclear Information System (INIS)

Isobe, Y.; Kamimura, A.; Aoki, K.; Nakayasu, F.

1995-01-01

Thermal aging is a growing concern for long-term-aged duplex stainless steel piping in nuclear power plants. Superconducting QUantum Interference Device (SQUID) was used for the detection of thermal aging of SUS329 rolled duplex stainless steel and SCS16 cast duplex stainless steel. It was found that the SQUID output signal pattern in the presence of AC magnetic field applied to the specimen was sensitive to the changes in electromagnetic properties due to thermal aging

76 FR 5205 - Carbon Steel Butt-Weld Pipe Fittings from Brazil, China, Japan, Taiwan, and Thailand

Science.gov (United States)

2011-01-28

...)] Carbon Steel Butt-Weld Pipe Fittings from Brazil, China, Japan, Taiwan, and Thailand AGENCY: United... Thailand. SUMMARY: The Commission hereby gives notice of the scheduling of expedited reviews pursuant to..., Taiwan, and Thailand would be likely to lead to continuation or recurrence of material injury within a...

75 FR 60814 - Carbon Steel Butt-Weld Pipe Fittings From Brazil, China, Japan, Taiwan, and Thailand

Science.gov (United States)

2010-10-01

...)] Carbon Steel Butt-Weld Pipe Fittings From Brazil, China, Japan, Taiwan, and Thailand AGENCY: United... Thailand. SUMMARY: The Commission hereby gives notice that it has instituted reviews pursuant to section... Thailand would be likely to lead to continuation or recurrence of material injury. Pursuant to section 751...

77 FR 64473 - Circular Welded Carbon-Quality Steel Pipe From the Sultanate of Oman: Final Affirmative...

Science.gov (United States)

2012-10-22

...-Quality Steel Pipe From the Sultanate of Oman: Final Affirmative Countervailing Duty Determination AGENCY... Sultanate of Oman (``Oman''). DATES: Effective Date: October 22, 2012. FOR FURTHER INFORMATION CONTACT... Sultanate of Oman (``GSO'') on April 5, April 20, and May 10, 2012. We received the GSO's responses...

Application of Leak Before Break concept in 316LN austenitic steel pipes welded using 316L

International Nuclear Information System (INIS)

Cunto, Gabriel Giannini de

2017-01-01

This work presents a study of application of the Leak Before Break (LBB) concept, usually applied in nuclear power plants, in a pipe made from steel AISI type 316LN welded a coated electrode AISI type 316L. LBB concept is a criterion based on fracture mechanics analysis to show that a crack leak, present in a pipe, can be detected by leak detection systems, before this crack reaches a critical size that results in pipe fail. In the studied pipe, tensile tests and Ramberg-Osgood analyses were performed, as well as fracture toughness tests for obtaining the material resistance curve J-R. The tests were performed considering the base metal, weld and heat affected zone (HAZ), at the same operating temperatures of a nuclear power plant. For the mechanical properties found in these tests, load limit analyses were performed in order to determine the size of a crack which could cause a detectable leakage and the critical crack size, considering failure by plastic collapse. For the critical crack size found in the weld, which is the region that presented the lowest toughness, Integral J and tearing modulus T analyses were performed, considering failure by tearing instability. Results show a well-defined behavior between the base metal, HAZ and weld zones, where the base metal has a high toughness behavior, the weld has a low toughness behavior and the HAZ showed intermediate mechanical properties between the base metal and the weld. Using the PICEP software, the leak rate curves versus crack size and also the critical crack size were determined by considering load limit analysis. It was observed that after a certain crack size, the leak rate in base metal is much higher than for the HAZ and the weld, considering the same crack length. This occurs because in the base metal crack, it is expected that the crack grows in a more rounded form due to its higher toughness. The lowest critical crack size was found for the base metal presenting circumferential cracks. For the

Microbiologically influenced corrosion of stainless steel in a nuclear waste facility

International Nuclear Information System (INIS)

Jenkins, C.F.; Doman, D.L.

1992-01-01

Corrosion in stainless steel cooling water piping in a nuclear waste processing facility occurred during an extended system lay-up. The failure characteristics indicated microbiologically influenced corrosion (MIC). The corrosion occurred at welds as pinhole penetrations in the surfaces, which opened into large subsurface void formations. Corrosive attack started in the heat-affected zones of the assembly welds, usually adjacent to fusion lines. Stepwise grinding, polishing, and etching in the affected areas revealed that voids generally grew in the wrought material as uniform, general corrosion. Tunneling (wormholing) erosion was also present. Selective attack occurred within the two-phase weld filler zone. The result was a void wall that was rough and porous-appearing, a consequence of preferential attack on the austenite. The three-dimensional spongy surface was studied optically and with the scanning electron microscope

Development of reliability-based load and resistance factor design methods for piping

International Nuclear Information System (INIS)

Ayyub, Bilal M.; Hill, Ralph S. III; Balkey, Kenneth R.

2003-01-01

Current American Society of Mechanical Engineers (ASME) nuclear codes and standards rely primarily on deterministic and mechanistic approaches to design. The American Institute of Steel Construction and the American Concrete Institute, among other organizations, have incorporated probabilistic methodologies into their design codes. ASME nuclear codes and standards could benefit from developing a probabilistic, reliability-based, design methodology. This paper provides a plan to develop the technical basis for reliability-based, load and resistance factor design of ASME Section III, Class 2/3 piping for primary loading, i.e., pressure, deadweight and seismic. The plan provides a proof of concept in that LRFD can be used in the design of piping, and could achieve consistent reliability levels. Also, the results from future projects in this area could form the basis for code cases, and additional research for piping secondary loads. (author)

Application of cyclic J-integral to low cycle fatigue crack growth of Japanese carbon steel pipe

Energy Technology Data Exchange (ETDEWEB)

Miura, N.; Fujioka, T.; Kashima, K. [and others

1997-04-01

Piping for LWR power plants is required to satisfy the LBB concept for postulated (not actual) defects. With this in mind, research has so far been conducted on the fatigue crack growth under cyclic loading, and on the ductile crack growth under excessive loading. It is important, however, for the evaluation of the piping structural integrity under seismic loading condition, to understand the fracture behavior under dynamic and cyclic loading conditions, that accompanies large-scale yielding. CRIEPI together with Hitachi have started a collaborative research program on dynamic and/or cyclic fracture of Japanese carbon steel (STS410) pipes in 1991. Fundamental tensile property tests were conducted to examine the effect of strain rate on tensile properties. Cracked pipe fracture tests under some loading conditions were also performed to investigate the effect of dynamic and/or cyclic loading on fracture behavior. Based on the analytical considerations for the above tests, the method to evaluate the failure life for a cracked pipe under cyclic loading was developed and verified. Cyclic J-integral was introduced to predict cyclic crack growth up to failure. This report presents the results of tensile property tests, cracked pipe fracture tests, and failure life analysis. The proposed method was applied to the cracked pipe fracture tests. The effect of dynamic and/or cyclic loading on pipe fracture was also investigated.

Heat treatments in a conventional steel to reproduce the microstructure of a nuclear grade steel

International Nuclear Information System (INIS)

Rosalio G, M.

2014-01-01

The ferritic steels used in the manufacture of pressurized vessels of Boiling Water Reactors (BWR) suffer degradation in their mechanical properties due to damage caused by the neutron fluxes of high energy bigger to a Mega electron volt (E> 1 MeV) generated in the reactor core. The materials with which the pressurized vessels of nuclear reactors cooled by light water are built correspond to low alloy ferritic steels. The effect of neutron irradiation on these steels is manifested as an increase in hardness, mechanical strength, with the consequent decrease in ductility, fracture toughness and an increase in temperature of ductile-brittle transition. The life of a BWR is 40 years, its design must be considered sufficient margin of safety because pressure forces experienced during operation, maintenance and testing of postulated accident conditions. It is necessary that under these conditions the vessel to behave ductile and likely to propagate a fracture is minimized. The vessels of light water nuclear reactors have a bainite microstructure. Specifically, the reactor vessels of the nuclear power plant of Laguna Verde (Veracruz, Mexico) are made of a steel Astm A-533, Grade B Class 1. At present they are carrying out some welding tests for the construction of a model of a BWR, however, to use nuclear grade steel such as Astm A-533 to carry out some of the welding tests, is very expensive; perform these in a conventional material provides basic information. Although the microstructure present in the conventional material does not correspond exactly to the degree of nuclear material, it can take of reference. Therefore, it is proposed to conduct a pilot study to establish the thermal treatment that reproduces the microstructure of nuclear grade steel, in conventional steel. The resulting properties of the conventional steel samples will be compared to a JRQ steel, that is a steel Astm A-533, Grade B Class 1, provided by IAEA. (Author)

Study on the residual stress relaxation in girth-welded steel pipes under bending load using diffraction methods

International Nuclear Information System (INIS)

Hempel, Nico; Nitschke-Pagel, Thomas; Dilger, Klaus

2017-01-01

This research is dedicated to the experimental investigation of the residual stress relaxation in girth-welded pipes due to quasi-static bending loads. Ferritic-pearlitic steel pipes are welded with two passes, resulting in a characteristic residual stress state with high tensile residual stresses at the weld root. Also, four-point bending is applied to generate axial load stress causing changes in the residual stress state. These are determined both on the outer and inner surfaces of the pipes, as well as in the pipe wall, using X-ray and neutron diffraction. Focusing on the effect of tensile load stress, it is revealed that not only the tensile residual stresses are reduced due to exceeding the yield stress, but also the compressive residual stresses for equilibrium reasons. Furthermore, residual stress relaxation occurs both parallel and perpendicular to the applied load stress.

Future directions for ferritic/martensitic steels for nuclear applications

International Nuclear Information System (INIS)

Klueh, R.L.; Swindeman, R.W.

2000-01-01

High-chromium (7-12% Cr) ferritic/martensitic steels are being considered for nuclear applications for both fission and fusion reactors. Conventional 9-12Cr Cr-Mo steels were the first candidates for these applications. For fusion reactors, reduced-activation steels were developed that were patterned on the conventional steels but with molybdenum replaced by tungsten and niobium replaced by tantalum. Both the conventional and reduced-activation steels are considered to have an upper operating temperature limit of about 550degC. For improved reactor efficiency, higher operating temperatures are required. For ferritic/martensitic steels that could meet such requirements, oxide dispersion-strengthened (ODS) steels are being considered. In this paper, the ferritic/martensitic steels that are candidate steels for nuclear applications will be reviewed, the prospect for ODS steel development and the development of steels produced by conventional processes will be discussed. (author)

Damping test results for straight sections of 3-inch and 8-inch unpressurized pipes

International Nuclear Information System (INIS)

Ware, A.G.; Thinnes, G.L.

1984-04-01

EG and G Idaho is assisting the Nuclear Regulatory Commission and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on unpressurized 3-in. and 8-in. Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 33-ft straight sections of piping were supported at the ends. Additionally, intermediate supports comprising spring, rod, and constant-force hangers, as well as a sway brace and snubbers, were used. Excitation was provided by low-force-level hammer impacts, a hydraulic shaker, and a 50-ton overhead crane for snapback testing. Data was recorded using acceleration, strain, and displacement time histories. This report presents test results showing the effect of stress level and type of supports on structural damping in piping

76 FR 19788 - Carbon Steel Butt-Weld Pipe Fittings From Brazil, China, Japan, Taiwan, and Thailand

Science.gov (United States)

2011-04-08

...)] Carbon Steel Butt-Weld Pipe Fittings From Brazil, China, Japan, Taiwan, and Thailand Determinations On... fittings from Brazil, China, Japan, Taiwan, and Thailand would be likely to lead to continuation or recurrence of material injury to an industry in the United States within a reasonably foreseeable time. \\1...

75 FR 26273 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China

Science.gov (United States)

2010-05-11

...)] Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China AGENCY: United States... materially injured or threatened with material injury, or the establishment of an industry in the United States is materially retarded, by reason of subsidized and less-than-fair-value imports from China of...

75 FR 973 - Certain Welded Stainless Steel Pipes From the Republic of Korea: Preliminary Results of...

Science.gov (United States)

2010-01-07

... to welded austenitic stainless steel pipes. The HTSUS subheadings are provided for convenience and... with sections 772(d)(1) and (2) of the Act, we also deducted, where applicable, those selling expenses associated with economic activities occurring in the United States, including U.S. direct selling expenses (i...

Development of advanced low alloy steel for nuclear RPV

Energy Technology Data Exchange (ETDEWEB)

Lee, H. C.; Shin, K. S.; Lee, S. H.; Lee, B. J. [Seoul National Univ., Seoul (Korea)

2000-04-01

Low carbon low alloy steels are used in nuclear power plants as pressure vessel, steam generator, etc. Nuclear pressure vessel material requires good combination of strength/ toughness, good weldability and high resistance to neutron irradiation and corrosion fatigue. For SA508III steels, most widely used in the production of nuclear power plant, attaining toughness is more difficult than strength. When taking into account the loss of toughness due to neutron irradiation, attaining as low transition temperature as possible prior to operation is a critical task in the production of nuclear pressure vessels. In the present study, we investigated detrimental microstructural features of SA508III steels to toughness, then alloy design directions to achieve improved mechanical properties were devised. The next step of alloy design was determined based on phase equilibrium thermodynamics and obtained results. Low carbon low alloy steels having low transition temperatures with enough strength and hardenability were developed. Microstructure and mechanical properties of HAZ of SA508III steels and alloy designed steels were investigated. 22 refs., 147 figs., 38 tabs. (Author)

Calculation of dynamic hydraulic forces in nuclear plant piping systems

International Nuclear Information System (INIS)

Choi, D.K.

1982-01-01

A computer code was developed as one of the tools needed for analysis of piping dynamic loading on nuclear power plant high energy piping systems, including reactor safety and relief value upstream and discharge piping systems. The code calculates the transient hydraulic data and dynamic forces within the one-dimensional system, caused by a pipe rupture or sudden value motion, using a fixed space and varying time grid-method of characteristics. Subcooled, superheated, homogeneous two-phase and transition flow regimes are considered. A non-equilibrium effect is also considered in computing the fluid specific volume and fluid local sonic velocity in the two-phase mixture. Various hydraulic components such as a spring loaded or power operated value, enlarger, orifice, pressurized tank, multiple pipe junction (tee), etc. are considered as boundary conditions. Comparisons of calculated results with available experimental data shows a good agreement. (Author)

Development of new Z-factors for the evaluation of the circumferential surface crack in nuclear pipes

International Nuclear Information System (INIS)

Choi, Y.H.; Chung, Y.K.; Park, Y.W.; Lee, J.B.

1997-01-01

The purpose of this study is to develop new Z-factors to evaluate the behavior of a circumferential surface crack in nuclear pipe. Z-factor is a load multiplier used in the Z-factor method, which is one of the ASME Code Sec. XI's recommendations for the estimation of a surface crack in nuclear pipe. It has been reported that the load carrying capacities predicted from the current ASME Code Z-factors, are not well in agreement with the experimental results for nuclear pipes with a surface crack. In this study, new Z-factors for ferritic base metal, ferritic submerged arc welding (SAW) weld metal, austenitic base metal, and austenitic SAW weld metal are obtained by use of the surface crack for thin pipe (SC.TNP) method based on GE/EPRI method. The desirability of both the SC.TNP method and the new Z-factors is examined using the results from 48 pipe fracture experiments for nuclear pipes with a circumferential surface crack. The results show that the SC.TNP method is good for describing the circumferential surface crack behavior and the new Z-factors are well in agreement with the measured Z-factors for both ferritic and austenitic pipes. (orig.)

Mitigation of inside surface residual stress of type 304 stainless steel pipe welds by inside water cooling method

International Nuclear Information System (INIS)

Sasaki, R.

1980-01-01

The weld residual stress distributions, macro- and microstructures of heat affected zone and IGSCC susceptibility of Type 304 stainless steel pipe welds by natural and inside water cooling methods have been investigated. The residual stresses of pipe welds by the natural cooling method are high tensile on both the inside and the outside surface. While the residual stresses on the inside surface of pipe welds by the inside water cooling method are compressive in both axial and circumferential directions for each pipe size from 2 to 24 inch diameter. The sensitized zones of welds by the inside water cooling method are closer to the fusion line, much narrower and milder than those by the natural cooling method. According to the constant extension rate test results for specimens taken from the inside surface of pipe welds, the inside water cooled welds are more resistant to IGSCC than naturally cooled ones

Ring thermal shield piping modification at Pickering Nuclear Generating Station 'A' Unit 1

International Nuclear Information System (INIS)

Brown, R.; Cobanoglu, M.M.

1995-01-01

Each of the four Pickering Nuclear Generating Station A (PNGSA) CANDU units was constructed with its reactor and dump tank surrounded by a concrete Calandria Vault (CV). The Ring Thermal Shield (RTS) system at PNGSA units is a water cooled structure with internal cooling channels with the purpose of attenuating excessive heat flux from the calandria shell to the end shield rings and adjoining concrete (Figure 1). In newer CANDU units the reactor calandria vessel is surrounded by a large water filled shield tank which eliminates the requirement for the RTS system. The RTS structures are situated in the space between the calandria and the vault walls. Each RTS is assembled from eight flat sided carbon steel segments, tilted towards the calandria and supported from the end shield rings. Cooling water to the RTS is supplied by carbon steel cooling pipes with a portion of the pipe run embedded in the vault walls. Flow through each RTS is divided into two independent circuits, having an inlet and an outlet cooling line. There are four locations of RTS inlet and outlet cooling lines. The inlet lines are located at the bottom and the outlet lines at the top of the RTS. The 'L' shaped section of RTS inlet and outlet cooling lines, from the RTS waterbox to the start of embedded portion at the concrete wall, had become defective due to corrosion induced by excessive Moisture levels in the calandria vaults. An on-line leak sealing capability was developed and placed in service in all four PNGSA units. However, a leak found during the 1994 Unit 1 outage was too large,to seal with the current capability, forcing Ontario Hydro (OH) to develop a method to replace the corroded pipes. The repair project was subject to some lofty performance targets. All tools had to be able to withstand dose rates of up to 3000 Rem/hour. These tools, along with procedures and personnel had to successfully repair the RTS system within 6 months otherwise a costly outage extension would result. This

Evolution of stainless steels in nuclear industry

International Nuclear Information System (INIS)

Tavassoli, Farhad

2010-01-01

Starting with the stainless steels used in the conventional industry, their adoption and successive evolutions in the nuclear industry, from one generation of nuclear reactors to another, is presented. Specific examples for several steels are given, covering fabrication procedures, qualification methods, property databases and design allowable stresses, to show how the ever-increasing demands for better performance and reliability, in particular under neutron irradiation, have been met. Particular attention is paid to the austenitic stainless steels types 304L, 316L, 316L(N), 316L(N)-IG, titanium stabilized grade 321, precipitation strengthened alloy 800, conventional and low activation ferritic/martensitic steels and their oxygen dispersion strengthening (ODS) derivatives. For each material, the evolution of the associated filler metal and welding techniques are also presented. (author)

X-ray diffraction study of microstructural changes during fatigue damage initiation in pipe steels: Role of the initial dislocation structure

International Nuclear Information System (INIS)

Pinheiro, B.; Lesage, J.; Pasqualino, I.; Bemporad, E.; Benseddiq, N.

2013-01-01

The present work is the second part of an ongoing study of microstructural changes during fatigue damage initiation in pipe steels [B. Pinheiro et al., Mat. Sci. Eng., A 532 (2012) 158–166]. Microdeformations and residual stresses (macrostresses) are evaluated by X-ray diffraction during alternating bending fatigue tests on samples taken from an API 5L X60 grade steel pipe. Microdeformations are evaluated from the full width at half maximum (FWHM) of the diffraction peak and residual stresses are estimated from the peak shift. Here, to understand the role of the initial dislocation structure, annealed samples are considered. As previously found for as-machined samples, the evolution of microdeformations shows three regular successive stages, but now with an increase during the first stage. The amplitude of each stage is accentuated with increasing stress amplitude, while its duration is reduced. Residual stresses show a similar trend, with stages of the same durations than those observed for FWHM, respectively. Additionally, changes in density and distribution of dislocations are observed by transmission electron microscopy combined with the technique of focused ion beam. The results are very encouraging for the development of a future indicator of fatigue damage initiation for pipe steels based on microstructural changes measured by X-ray diffraction

X-ray diffraction study of microstructural changes during fatigue damage initiation in pipe steels: Role of the initial dislocation structure

Energy Technology Data Exchange (ETDEWEB)

Pinheiro, B., E-mail: bianca@lts.coppe.ufrj.br [Laboratory of Mechanics of Lille (LML), UMR CNRS 8107, University Lille 1, Boulevard Paul Langevin, Cité Scientifique, 59655 Villeneuve d' Ascq (France); Lesage, J. [Laboratory of Mechanics of Lille (LML), UMR CNRS 8107, University Lille 1, Boulevard Paul Langevin, Cité Scientifique, 59655 Villeneuve d' Ascq (France); Pasqualino, I. [Subsea Technology Laboratory (LTS), Ocean Engineering Department, COPPE/Federal University of Rio de Janeiro, PO Box 68508, Cidade Universitária, CEP 21945-970, Rio de Janeiro/RJ (Brazil); Bemporad, E. [University of Rome “ROMA TRE”, Mechanical and Industrial Eng. Dept., Via Vasca Navale 79, 00146 Rome (Italy); Benseddiq, N. [Laboratory of Mechanics of Lille (LML), UMR CNRS 8107, University Lille 1, Boulevard Paul Langevin, Cité Scientifique, 59655 Villeneuve d' Ascq (France)

2013-09-15

The present work is the second part of an ongoing study of microstructural changes during fatigue damage initiation in pipe steels [B. Pinheiro et al., Mat. Sci. Eng., A 532 (2012) 158–166]. Microdeformations and residual stresses (macrostresses) are evaluated by X-ray diffraction during alternating bending fatigue tests on samples taken from an API 5L X60 grade steel pipe. Microdeformations are evaluated from the full width at half maximum (FWHM) of the diffraction peak and residual stresses are estimated from the peak shift. Here, to understand the role of the initial dislocation structure, annealed samples are considered. As previously found for as-machined samples, the evolution of microdeformations shows three regular successive stages, but now with an increase during the first stage. The amplitude of each stage is accentuated with increasing stress amplitude, while its duration is reduced. Residual stresses show a similar trend, with stages of the same durations than those observed for FWHM, respectively. Additionally, changes in density and distribution of dislocations are observed by transmission electron microscopy combined with the technique of focused ion beam. The results are very encouraging for the development of a future indicator of fatigue damage initiation for pipe steels based on microstructural changes measured by X-ray diffraction.

Stress-assisted, microbial-induced corrosion of stainless steel primary piping and other aging issues at the Omega West Reactor

International Nuclear Information System (INIS)

Andrade, A.

1995-01-01

After the discovery of cooling system leak of about 284 liters per twenty-four (24) hour period, an investigation determined that the 76.2-cm diameter, 33.5-m long stainless-steel (304) OWR delay line was losing water at the same nominal rate. An excavation effort revealed that a circumferential crack, approximately 0.0025 cm in width, extended around the bottom half of the delay line. In addition, other evidence of what appeared to be microcracking and pitting that originated at random nucleated sites around the pipe were also found. Results of destructive analysis and nondestructive testing allowed Los Alamos staff to conclude that the direct cause for the main crack and other pitting resulted from stress-assisted, microbial-induced corrosion of the stainless steel primary piping. The results also indicated that microbial action from bacteria that are normally present in earth can be extremely harmful to stainless- steel piping under certain conditions. Other potential problems that could have also eventually led to a permanent shutdown of the OWR were discussed. These problems, although never encountered nor associated with the current shutdown, were identified in aging studies and are associated with: (1) the water-cooled, bismuth gamma-ray shield and, (2) the aluminum thermal column head seal that prevents reactor vessel water from entering into the graphite-filled thermal column

Using an equation based on flow stress to estimate structural integrity of annealed Type 304 stainless steel plate and pipes containing surface defects

International Nuclear Information System (INIS)

Reuter, W.G.; Place, T.A.

1981-01-01

An accurate assessment of the influence of defects on structural component integrity is needed. Generally accepted analytical techniques are not available for the very ductile materials used in many nuclear reactor components. Some results are presented from a test programme to obtain data by which to evaluate proposed models. Plate and pipe specimens containing surface flaws were fabricated from annealed Type 304 stainless steel and tested at room temperature. An evaluation of an empirical equation based on flow stress is presented. In essentially all instances the flow stress is not a constant but varies as a function of the size of the surface flaw. (author)

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

Energy Technology Data Exchange (ETDEWEB)

Ehud Greenspan

2008-09-30

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

Failure probability assessment of wall-thinned nuclear pipes using probabilistic fracture mechanics

International Nuclear Information System (INIS)

Lee, Sang-Min; Chang, Yoon-Suk; Choi, Jae-Boong; Kim, Young-Jin

2006-01-01

The integrity of nuclear piping system has to be maintained during operation. In order to maintain the integrity, reliable assessment procedures including fracture mechanics analysis, etc., are required. Up to now, this has been performed using conventional deterministic approaches even though there are many uncertainties to hinder a rational evaluation. In this respect, probabilistic approaches are considered as an appropriate method for piping system evaluation. The objectives of this paper are to estimate the failure probabilities of wall-thinned pipes in nuclear secondary systems and to propose limited operating conditions under different types of loadings. To do this, a probabilistic assessment program using reliability index and simulation techniques was developed and applied to evaluate failure probabilities of wall-thinned pipes subjected to internal pressure, bending moment and combined loading of them. The sensitivity analysis results as well as prototypal integrity assessment results showed a promising applicability of the probabilistic assessment program, necessity of practical evaluation reflecting combined loading condition and operation considering limited condition

Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 5. Summary - Piping Review Committee conclusions and recommendations

International Nuclear Information System (INIS)

1985-04-01

This document summarizes a comprehensive review of NRC requirements for Nuclear Piping by the US NRC Piping Review Committee. Four topical areas, addressed in greater detail in Volumes 1 through 4 of this report, are included: (1) Stress Corrosion Cracking in Piping of Boiling Water Reactor Plants; (2) Evaluation of Seismic Design; (3) Evaluation of Potential for Pipe Breaks; and (4) Evaluation of Other Dynamic Loads and Load Combinations. This volume summarizes the major issues, reviews the interfaces, and presents the Committee's conclusions and recommendations for updating NRC requirements on these issues. This report also suggests research or other work that may be required to respond to issues not amenable to resolution at this time

Summary and accomplishments of the ORNL program for nuclear piping design criteria

International Nuclear Information System (INIS)

Greenstreet, W.L.

1975-11-01

The ORNL Piping Program was defined and established to develop basic information on the structure behavior of nuclear power plant piping components and to prepare this information in forms suitable for use in design codes and standards. Charts are presented showing the percentage completion of the various program tasks

Safety evaluation of socket weld integrity in nuclear piping

International Nuclear Information System (INIS)

Choi, Y.H.; Kim, H.J.; Choi, S.Y.; Kim, Y.J.; Kim, Y.J.

2004-01-01

The purposes of this paper are to evaluate the integrity of socket weld in nuclear piping and prepare the technical basis for a new guideline on radiographic testing (RT) for the socket weld. Recently, the integrity of the socket weld is regarded as a safety concern in nuclear power plants because lots of failures and leaks have been reported in the socket weld. The root causes of the socket weld failure are known as unanticipated loadings such as vibration or thermal fatigue and improper weld joint during construction. The ASME Code sec. III requires 1/16 inch gap between the pipe and fitting in the socket weld. Many failure cases, however, showed that the gap requirement was not satisfied. The Code also requires magnetic particle examination (MT) or liquid penetration examination (PT) on the socket weld, but not radiographic examination (RT). It means that it is not easy to examine the 1/16 inch gap in the socket weld by using the NDE methods currently required in the Code. In this paper, the effects of the requirements in the ASME Code sec. III on the socket weld integrity were evaluated by using finite element method. The crack behavior in the socket weld was also investigated under vibration event in nuclear power plants. The results showed that the socket weld was very susceptible to the vibration if the requirements in ASME Code were not satisfied. The constraint between the pipe and fitting due to the contact significantly affects the integrity of the socket weld. This paper also suggests a new guideline on the RT for the socket weld during construction stage in nuclear power plants. (orig.)

The installation welding of pressure water reactor coolant piping

International Nuclear Information System (INIS)

Deng Feng

2010-01-01

Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

Assessment of Pipe Wall Loss Using Guided Wave Testing

International Nuclear Information System (INIS)

Joo, Kyung Mun; Jin, Seuk Hong; Moon, Yong Sig

2010-01-01

Flow accelerated corrosion(FAC) of carbon steel pipes in nuclear power plants has been known as one of the major degradation mechanisms. It could have bad influence on the plant reliability and safety. Also detection of FAC is a significant cost to the nuclear power plant because of the need to remove and replace insulation. Recently, the interest of the guided wave testing(GWT) has grown because it allows long range inspection without removing insulation of the pipe except at the probe position. If GWT can be applied to detection of FAC damages, it will can significantly reduce the cost for the inspection of the pipes. The objective of this study was to determine the capability of GWT to identify location of FAC damages. In this paper, three kinds of techniques were used to measure the amplitude ratio between the first and the second welds at the elbow area of mock-ups that contain real FAC damages. As a result, optimal inspection technique and minimum detectability to detect FAC damages drew a conclusion

76 FR 49437 - Certain Circular Welded Non-Alloy Steel Pipe From Mexico: Preliminary Results of Antidumping Duty...

Science.gov (United States)

2011-08-10

... Non-Alloy Steel Pipe from Brazil, the Republic of Korea (Korea), Mexico, and Venezuela and Amendment... 23954 (May 6, 2003) (Assessment Policy Notice). Because ``as entered'' liquidation instructions do not... assessment'' regulation on May 6, 2003. See Assessment Policy Notice. This clarification will apply to POR...

Modeling of hot tensile and short-term creep strength for LWR piping materials under severe accident conditions

International Nuclear Information System (INIS)

Harada, Y.; Maruyama, Y.; Chino, E.; Shibazaki, H.; Kudo, T.; Hidaka, A.; Hashimoto, K.; Sugimoto, J.

2000-01-01

The analytical study on severe accident shows the possibility of the reactor coolant system (RCS) piping failure before reactor pressure vessel failure under the high primary pressure sequence at pressurized water reactors. The establishment of the high-temperature strength model of the realistic RCS piping materials is important in order to predict precisely the accident progression and to evaluate the piping behavior with small uncertainties. Based on material testing, the 0.2% proof stress and the ultimate tensile strength above 800degC were given by the equations of second degree as a function of the reciprocal absolute temperature considering the strength increase due to fine precipitates for the piping materials. The piping materials include type 316 stainless steel, type 316 stainless steel of nuclear grade, CF8M cast duplex stainless steel and STS410 carbon steel. Also the short-term creep rupture time and the minimum creep rate at high-temperature were given by the modified Norton's Law as a function of stress and temperature considering the effect of the precipitation formation and resolution on the creep strength. The present modified Norton's Law gives better results than the conventional Larson-Miller method. Correlating the creep data (the applied stress versus the minimum creep rate) with the tensile data (the 0.2% proof stress or the ultimate tensile strength versus the strain rate), it was found that the dynamic recrystallization significantly occurred at high-temperature. (author)

Leak-before-break analysis of thermally aged nuclear pipe under different bending moments

Energy Technology Data Exchange (ETDEWEB)

Lv, Xuming; Li, Shilei; Zhang, Hailong; Wang, Yanli; Wang, Xitao [University of Science and Technology Beijing, Beijing (China); Wang, Zhaoxi [CPI Nuclear Power Institute, Beijing (China); Xue, Fei [Suzhou Nuclear Power Research Institute, Suzhou (China)

2015-10-15

Cast duplex stainless steels are susceptible to thermal aging during long-term service at temperatures ranging from 280°C to 450°C. To analyze the effect of thermal aging on leak-before-break (LBB) behavior, three-dimensional finite element analysis models were built for circumferentially cracked pipes. Based on the elastic–plastic fracture mechanics theory, the detectable leakage crack length calculation and J-integral stability assessment diagram approach were carried out under different bending moments. The LBB curves and LBB assessment diagrams for unaged and thermally aged pipes were constructed. The results show that the detectable leakage crack length for thermally aged pipes increases with increasing bending moments, whereas the critical crack length decreases. The ligament instability line and critical crack length line for thermally aged pipes move downward and to the left, respectively, and unsafe LBB assessment results will be produced if thermal aging is not considered. If the applied bending moment is increased, the degree of safety decreases in the LBB assessment.

Comparisons of ASME-code fatigue-evaluation methods for nuclear Class 1 piping with Class 2 or 3 piping

International Nuclear Information System (INIS)

Rodabaugh, E.C.

1983-06-01

The fatigue evaluation procedure used in the ASME Boiler and Pressure Vessel Code, Sect. III, Nuclear Power Plant Components, for Class 1 piping is different from the procedure used for Class 2 or 3 piping. The basis for each procedure is described, and correlations between the two procedures are presented. Conditions under which either procedure or both may be unconservative are noted. Potential changes in the Class 2 or 3 piping procedure to explicitly cover all loadings are discussed. However, the report is intended to be informative, and while the contents of the report may guide future Code changes, specific recommendations are not given herein

PWR composite materials use. A particular case of safety-related service water pipes

International Nuclear Information System (INIS)

Pays, M.F.; Le Courtois, T.

1997-11-01

This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during 'lifetime'); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author)

PWR composite materials use. A particular case of safety-related service water pipes

Energy Technology Data Exchange (ETDEWEB)

Pays, M.F.; Le Courtois, T

1997-11-01

This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during `lifetime`); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author) 2 refs.

78 FR 72863 - Circular Welded Carbon Quality Steel Pipe From the People's Republic of China: Continuation of...

Science.gov (United States)

2013-12-04

... Quality Steel Pipe From the People's Republic of China: Continuation of Countervailing Duty Order AGENCY...) from the People's Republic of China (PRC) would likely lead to continuation or recurrence of net... reasonably foreseeable time.\\3\\ \\1\\ See Initiation of Five-Year (``Sunset'') Review, 78 FR 33063 (June 3...

Experimental verification on limit load estimation method for pipes with an arbitrary shaped circumferential surface flaw

International Nuclear Information System (INIS)

Li, Yinsheng; Hasegawa, Kunio; Miura, Naoki; Hoshino, Katsuaki

2010-01-01

When a flaw is detected in stainless steel pipes during in-service inspection, the limit load criterion given in the codes such as JSME Rules on Fitness-for-Service for Nuclear Power Plants or ASME Boiler and Pressure Vessel Code Section XI can be applied to evaluate the integrity of the pipe. However, in these codes, the limit load criterion is only provided for pipes containing a flaw with uniform depth, although many flaws with complicated shape such as stress corrosion cracking have been actually detected in pipes. In order to evaluate the integrity of the flawed pipes for general case, a limit load estimation method has been proposed by authors considering a circumferential surface flaw with arbitrary shape. The plastic collapse bending moment and corresponding stress are obtained by dividing the surface flaw into several segmented sub-flaws. In this paper, the proposed method was verified by comparing with experimental results. Four-point bending experiments were carried out for full scale stainless steel pipes with a symmetrical or non-symmetrical circumferential flaw. Estimated failure bending moments by the proposed method were found to be in good agreement with the experimental results, and the proposed method was confirmed to be effective for evaluating bending failure of pipes with flaw. (author)

Advanced concepts, analysis approaches and criteria for nuclear piping system design

International Nuclear Information System (INIS)

Tang, H.T.; Tagart, S.W. Jr.; Tang, Y.K.

1992-01-01

Recent research in piping system design and analysis has resulted in advancements on damping values, independent support motion (ISM), static coefficient method, simplified inelastic method and ASME code criteria changes. In the support area, passive type of supports such as energy-absorbing device and gap stopper have been developed. These advancements provide bases for improved and cost-effective design of future nuclear piping systems. (author)

Estimation of residual stress distribution for pressurizer nozzle of Kori nuclear power plant considering safe end

Energy Technology Data Exchange (ETDEWEB)

Song, Tae Kwang; Bae, Hong Yeol; Chun, Yun Bae; Oh, Chang Young; Kim, Yun Jae [Korea University, Seoul (Korea, Republic of); Lee, Kyoung Soo; Park, Chi Yong [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2008-08-15

In nuclear power plants, ferritic low alloy steel nozzle was connected with austenitic stainless steel piping system through alloy 82/182 butt weld. Accurate estimation of residual stress for weldment is important in the sense that alloy 82/182 is susceptible to stress corrosion cracking. There are many results which predict residual stress distribution for alloy 82/182 weld between nozzle and pipe. However, nozzle and piping system usually connected through safe end which has short length. In this paper, residual stress distribution for pressurizer nozzle of Kori nuclear power plant was predicted using FE analysis, which considered safe end. As a result, existing residual stress profile was redistributed and residual stress of inner surface was decreased specially. It means that safe end should be considered to reduce conservatism when estimating the piping system.

Damping test results for straight sections of 3-inch and 8-inch unpressurized pipes. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Ware, A.G.; Thinnes, G.L.

1984-04-01

EG and G Idaho is assisting the Nuclear Regulatory Commission and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on unpressurized 3-in. and 8-in. Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 33-ft straight sections of piping were supported at the ends. Additionally, intermediate supports comprising spring, rod, and constant-force hangers, as well as a sway brace and snubbers, were used. Excitation was provided by low-force-level hammer impacts, a hydraulic shaker, and a 50-ton overhead crane for snapback testing. Data was recorded using acceleration, strain, and displacement time histories. This report presents test results showing the effect of stress level and type of supports on structural damping in piping.

Application of high strength steel to nuclear reactor containment vessel

International Nuclear Information System (INIS)

Susukida, H.; Sato, M.; Takano, G.; Uebayashi, T.; Yoshida, K.

1976-01-01

Nuclear reactor containment vessels are becoming larger in size with the increase in the power generating capacity of nuclear power plants. For example, a containment vessel for a PWR power plant with an output of 1,000 MWe becomes an extremely large one if it is made of the conventional JIS SGV 49 (ASTM A 516 Gr. 70) steel plates less than 38 mm in thickness. In order to design the steel containment vessel within the conventional dimensional range, therefore, it is necessary to use a high strength steel having a higher tensile strength than SGV 49 steel, good weldability and a higher fracture toughness and moreover, possessing satisfactory properties without undergoing post-weld heat treatment. The authors conducted a series of verification tests on high strength steel developed by modifying the ASTM A 543 Grade B Class 1 steel with a view to adopting it as a material for the nuclear reactor containment vessels. As the result of evaluation of the test results from various angles, we confirmed that the high strength steel is quite suitable for the manufacture of nuclear reactor containment vessels. (auth.)

Program to justify life extension of older nuclear piping systems

International Nuclear Information System (INIS)

Burr, T.K.; Dwight, J.E. Jr.; Morton, D.K.

1991-01-01

The Idaho National Engineering Laboratory (INEL) has a history of more than 40 years devoted to the operation of nuclear reactors designed for research and experiments. The Advanced Test Reactor (ATR) is one such operating reactor whose mission requires continued operation for an additional 25 years or more. Since the ATR is approaching its design life of twenty years, life extension evaluations have been initiated. Of particular importance are the associated high temperature, high pressure loop piping system supporting in--reactor experiments. Failure of this piping could challenge core safety margins. Since regulatory rules for nuclear power plant life extension are only in the formulation stage, the current technical guidance on this subject provided by the Department of Energy (DOE) or the commercial nuclear industry is incomplete. In the interim, order to assure continued safe operation of this piping beyond its initial design life, a program has been developed to provide the necessary technical justification for life extension. This paper describes a program that establishes Section 11 of the ASME Boiler and Pressure Vessel Code as the governing criteria document, retains B31.1 as the Code of record for Section 11 activities, specifies additional inservice inspection requirements more strict than Section 11, and relies heavily on flaw detection and fracture mechanics evaluations. 18 refs., 2 figs

EBSD applications in the steel and nuclear industries

International Nuclear Information System (INIS)

Nave, M.D.

2005-01-01

EBSD has established itself as an invaluable tool for materials science problem-solving in the steel and nuclear industries. In the steel industry, it increases our understanding of the deformation and recrystallization processes that influence the formability of steel sheets. It is also used to improve welding procedures and identify phases that accelerate corrosion. In the nuclear industry, EBSD plays a central role in extending the life of fuel cladding materials by shedding new light on the mechanisms of hydride formation. It is also used in efforts to improve the processing of material used for the storage of nuclear waste. This presentation provides an overview of EBSD applications within these two industries, emphasizing the broad applicability and practical usefulness of the technique. (author)

75 FR 20342 - Certain Circular Welded Non-Alloy Steel Pipe From Mexico: Final Results of Antidumping Duty...

Science.gov (United States)

2010-04-19

... covered in this review, but was covered in a previous review or the original less-than-fair-value (LTFV... investigation. See Final Determination of Sales at Less Than Fair Value: Circular Welded Non- Alloy Steel Pipe... for framing and support members for reconstruction or load-bearing purposes in the construction...

78 FR 5170 - Circular Welded Carbon Quality Steel Pipe From the People's Republic of China: Rescission of...

Science.gov (United States)

2013-01-24

... International Trade Co., Ltd., Weifang East Steel Pipe Co., Ltd., WISCO & CRM Wuhan Material & Trade., Wuxi... relevant entries during this review period. Failure to comply with this requirement could result in the.... Failure to comply with the regulations and terms of an APO is a violation which is subject to sanction...

75 FR 69050 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From the People's...

Science.gov (United States)

2010-11-10

... amended (``the Act''), that an industry in the United States is threatened with material injury by reason... case is based on the threat of material injury and is not accompanied by a finding that injury would... Alloy Steel Standard, Line, and Pressure Pipe From the People's Republic of China: Amended Final...

Titanium Loop Heat Pipes for Space Nuclear Radiators, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — This Small Business Innovation Research Phase I project will develop titanium Loop Heat Pipes (LHPs) that can be used in low-mass space nuclear radiators, such as...

76 FR 78615 - Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab...

Science.gov (United States)

2011-12-19

...-810] Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab..., the Sultanate of Oman (``Oman''), The United Arab Emirates (``the UAE''), and the Socialist Republic... Oman, the United Arab Emirates, and the Socialist Republic of Vietnam: Initiation of Countervailing...

78 FR 34342 - Certain Circular Welded Non-Alloy Steel Pipe From Mexico: Final Results and Partial Rescission of...

Science.gov (United States)

2013-06-07

... Welded Non-Alloy Steel Pipe from Brazil, the Republic of Korea (Korea), Mexico, and Venezuela and... Proceedings: Assessment of Antidumping Duties, 68 FR 23954 (May 6, 2003) (reseller policy). This clarification... antidumping duties in accordance with the reseller policy. Cash Deposit Requirements The following cash...

Self-contained pipe cutting shear. Innovative technology summary report

International Nuclear Information System (INIS)

1998-11-01

The US Department of Energy (DO) is in the process of decontaminating and decommissioning (D and D) many of its nuclear facilities throughout the country. Facilities have to be dismantled and demolition waste must be sized into manageable pieces for handling and disposal. Typically, the facilities undergoing D and D are contaminated, either chemically, radiologically, or both. In its D and D work, the DOE was in need of a tool capable of cutting steel and stainless steel pipe up to 6.4 cm in diameter. The self-contained pipe cutting shear was developed by Lukas Hydraulic GmbH and Co. KG to cut pipes up to 6.4 cm (2.5 in.) in diameter. This tool is a portable, hand-held hydraulic shear that is powered by a built-in rechargeable battery or a portable auxiliary rechargeable battery. Adding to its portability, it contains no hydraulic fluid lines or electrical cords, making it useful in congested areas or in areas with no power. Both curved and straight blades can be attached, making it adaptable to a variety of conditions. This tool is easy to set up, operates quietly, and cuts through pipes quickly. It is especially useful on contaminated pipes, as it crimps the ends while cutting and produces no residual cuttings. This shear is a valuable alternative to baseline technologies such as portable band saws, electric hacksaws, and other hydraulic shears. Costs using the innovative shear for cutting 2.5 cm (1-in.) pipe, for example, are comparable to costs using a conventional shear, approximately 80% of portable bandsaw costs and half of electric hacksaw costs

The development of design method of nuclear piping system supported by elasto-plastic support structures (part 2)

International Nuclear Information System (INIS)

Endo, R.; Murota, M.; Kawabata, J-I.; Hirose, J.; Nekomoto, Y.; Takayama, Y.; Kobayashi, H.

1995-01-01

The conventional seismic design method of nuclear piping system is very conservative because of the accumulation of various safety factors in the design process, and nuclear piping systems are thought to have a large safety margin. Considering this situations, research program was promoted to furthermore rationalize nuclear power plants by reducing the amount of support structures and reducing the piping's seismic response through vibration energy absorption resulting from the elasto-plastic behavior of piping support structures. The research had the following three stages. In the first stage, we selected conventional piping support structures in light-water reactors that exhibited elasto-plastic behavior, and studied the effect of displacement and the vibration frequency on the stiffness and on the energy absorption by testing these models. In the second stage, vibration tests were performed using piping models with support structures on shaking tables. The piping vibration characteristics were clarified by sinusoidal sweep tests and the piping response characteristics by seismic wave vibration tests when the support structures were in an elasto-plastic condition. In the third stage, a general method was developed to evaluate the characteristics of a variety of support structures in the tests. A simplified analysis method was also developed to evaluate the piping seismic response using the piping model test result. To expand the results mentioned above, we also established a new seismic design method of piping systems that allowed support structures to have elasto-plastic behavior. This paper reports the newly developed seismic design method based on the results of experiments conducted under the joint research program of Japanese electric power companies (The Japan Atomic Power Co., Hokkaido EPC, Tohoku EPC, Tokyo EPC, Chubu EPC, Hokuriku EPC, Kansai EPC, Chugoku EPC, Shikoku EPC, Kyushu EPC) and nuclear plant makers (Hitachi Ltd., Toshiba Co., MHI Ltd., HEC Ltd

Study on quality control measures of static casting main pipe in PWR nuclear power plant

International Nuclear Information System (INIS)

Jiang Zhenbiao; Li Guanying; Liu Zhicheng

2013-01-01

This study analyzes the main reasons which impact the quality of primary pipe static casting elbows in PWR-M310 nuclear power plant. The quality control measures are developed from the election and inspection of material, improving sand production and casting process, improving lean management of personnel. The static casting defects of primary pipe elbows for Fuqing Unit 1 and 2 were down to less than 50% of the former project. The quality of static casting for the primary pipe elbows was significantly improved. Moreover, the implementation saves human resources and financing to repair casting defects, and also helps to win the delivery schedule. The quality control measures are good reference for improving primary pipe casting process. This study provides valuable experience for further study of improving the quality of static casting for the primary pipe of PWR nuclear power plant. (authors)

Low Frequency Phased Array Application for Crack Detection in Cast Austenitic Piping

International Nuclear Information System (INIS)

Anderson, Michael T.; Cumblidge, Stephen E.; Doctor, Steven R.

2006-01-01

As part of a multi-year program funded by the United States Nuclear Regulatory Commission (US NRC) to address nondestructive examination (NDE) reliability of inservice inspection (ISI) programs, studies conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, have focused on assessing novel NDE approaches for the inspection of coarse-grained, cast stainless steel reactor components. The primary objective of this work is to provide information to the US NRC on the utility, effectiveness and reliability of ultrasonic testing (UT) as related to the ISI of primary piping components in US commercial nuclear power plants. This paper describes progress, recent developments and results from an assessment of a portion of the work relating to the ultrasonic low frequency phased array inspection technique. Westinghouse Owner's Group (WOG) cast stainless steel pipe segments with thermal and mechanical fatigue cracks, PNNL samples containing thermal fatigue cracks and several blank vintage specimens having very coarse grains that are representative of early centrifugally cast piping installed in PWRs, were used for assessing the inspection method. The phased array approach was implemented using an R/D Tech Tomoscan III system operating at 1.0 MHz and 500 kHz, providing composite volumetric images of the samples. Several dual, transmit-receive, custom designed low-frequency arrays were employed in laboratory trials. Results from laboratory studies for assessing detection, localization and length sizing effectiveness are discussed.

Effects of Induction Heat Bending and Heat Treatment on the Boric Acid Corrosion of Low Alloy Steel Pipe for Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Ki-Tae; Kim, Young-Sik [Andong National University, Gyeongbuk (Korea, Republic of); Chang, Hyun-Young; Park, Heung-Bae [KEPCO EandC, Gyeongbuk (Korea, Republic of); Sung, Gi-Ho; Shin, Min-Chul [Sungil SIM Co. Ltd, Busan (Korea, Republic of)

2016-11-15

In many plants, including nuclear power plants, pipelines are composed of numerous fittings such as elbows. When plants use these fittings, welding points need to be increased, and the number of inspections also then increases. As an alternative to welding, the pipe bending process forms bent pipe by applying strain at low or high temperatures. This work investigates how heat treatment affects on the boric acid corrosion of ASME SA335 Gr. P22 caused by the induction heat bending process. Microstructure analysis and immersion corrosion tests were performed. It was shown that every area of the induction heat bent pipe exhibited a high corrosion rate in the boric acid corrosion test. This behavior was due to the enrichment of phosphorous in the ferrite phase, which occurred during the induction heat bending process. This caused the ferrite phase to act as a corrosion initiation site. However, when re-heat treatment was applied after the bending process, it enhanced corrosion resistance. It was proved that this resistance was closely related to the degree of the phosphorus segregation in the ferrite phase.

Investigations on penetration control for automated pipe welding system

International Nuclear Information System (INIS)

Fujiki, Daisuke; Sato, Akihiro; Funamoto, Takao; Matsumoto, Toshimi; Kobayashi, Masahiro

1995-01-01

We have been investigating process conditions forming sound root bead by orbital welding technique for nuclear power stations. Specimens used were stainless steel (SUS304) pipes (318.5 mm outside diameter and 15.4 mm thickness), and pulsed gas tungsten-arc (GTA) welder was adopted. We have found process conditions to form sound root bead by changing both heat input conditions and joint designs. It is found that reducing volume of molten metal is necessary to form sound root bead. And it is also found that changing joint designs is effective to reduce volume of molten metal. By selecting proper joint designs, we could form sound root bead in constant heat input conditions in every position of pipe. (author)

Numerical simulation and experimental verification of microstructure evolution in large forged pipe used for AP1000 nuclear power plants

International Nuclear Information System (INIS)

Wang, Shenglong; Yang, Bin; Zhang, Mingxian; Wu, Huanchun; Peng, Jintao; Gao, Yang

2016-01-01

Highlights: • Establish systematically the database of 316LN stainless steel for Deform-3D. • Simulate the microstructure evolution during forging of AP1000 primary coolant pipe. • Carry out full-scale forging experiment for verification in engineering practice. • Get desirable grain size in simulation and experiment. • The variation trends of grain sizes in simulation and experiment are consistent. - Abstract: AP1000 primary coolant pipe is a large special-shaped forged pipe made of 316LN stainless steel. Due to the non-uniform temperature and deformation during its forging, coarse and fine grains usually coexist in the forged pipe, resulting in the heterogeneous microstructure and anisotropic performance. To investigate the microstructure evolution during the entire forging process, in the present research, the database of the 316LN stainless steel was established and a numerical simulation was performed. The results indicate that the middle body section of the forged pipe has an extremely uniform average grain size with the value smaller than 30 μm. The grain sizes in the ends of body sections were ranged from 30 μm to 60 μm. Boss sections have relatively homogeneous microstructure with the average grain size 30 μm to 44 μm. Furthermore, a full-scale hot forging was carried out for verification. Comparison of theoretical and experimental results showed good agreement and hence demonstrated the capabilities of the numerical simulation presented here. It is noteworthy that all grains in the workpiece were confirmed less than 180 μm, which meets the designer’s demands.

76 FR 72164 - Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab...

Science.gov (United States)

2011-11-22

...-811] Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab... Emirates, and Vietnam), or Angelica Mendoza (Oman), AD/CVD Operations, Office 7, Import Administration... Sultanate of Oman (Oman), the United Arab Emirates (UAE), and the Socialist Republic of Vietnam (Vietnam...

Effect of Cr content on the FAC of pipe material at 150 .deg. C

International Nuclear Information System (INIS)

Park, Tae Jun; Kim, Hong Pyo

2013-01-01

Flow accelerated corrosion (FAC) of the carbon steel piping in nuclear power plants (NPPs) has been major issue in nuclear industry. During the FAC, a protective oxide layer on carbon steel dissolves into flowing water leading to a thinning of the oxide layer and accelerating corrosion of base material. As a result, severe failures may occur in the piping and equipment of NPPs. Effect of alloying elements on FAC of pipe materials was studied with rotating cylinder FAC test facility at 150 .deg. C and at flow velocity of 4m/s. The facility is equipped with on line monitoring of pH, conductivity, dissolved oxygen(DO) and temperature. Test solution was the demineralized water, and DO concentration was less than 1 ppb. Surface appearance of A 106 Gr. B which is used widely in secondary pipe in NPPs showed orange peel appearance, typical appearance of FAC. The materials with Cr content higher than 0.17wt.% showed pit. The pit is thought to early degradation mode of FAC. The corrosion product within the pit was enriched with Cr, Mo, Cu, Ni and S. But S was not detected in SA336 F22V with 2.25wt.% Cr. The enrichment of Cr and Mo seemed to be related with low, solubility of Cr and Mo compared to Fe. Measured FAC rate was compared with Ducreaux's relationship and showed slightly lower FAC rate than Ducreaux's relationship

Fatigue evaluation of socket welded piping in nuclear power plant

International Nuclear Information System (INIS)

Vecchio, R.S.

1996-01-01

Fatigue failures in piping systems occur, almost without exception, at the welded connections. In nuclear power plant systems, such failures occur predominantly at the socket welds of small diameter piping ad fillet attachment welds under high-cycle vibratory conditions. Nearly all socket weld fatigue failures are identified by leaks which, though not high in volume, generally are costly due to attendant radiological contamination. Such fatigue cracking was recently identified in the 3/4 in. diameter recirculation and relief piping socket welds from the reactor coolant system (RCS) charging pumps at a nuclear power plant. Consequently, a fatigue evaluation was performed to determine the cause of cracking and provide an acceptable repair. Socket weld fatigue life was evaluated using S-N type fatigue life curves for welded structures developed by AASHTO and the assessment of an effective cyclic stress range adjacent to each socket weld. Based on the calculated effective tress ranges and assignment of the socket weld details to the appropriate AASHTO S-N curves, the socket weld fatigue lives were calculated and found to be in excellent agreement with the accumulated cyclic life to-date

An evaluation of detection ability of ultrasonic testing with a large aperture transducer for axial cracks in cast stainless steel pipe welds

International Nuclear Information System (INIS)

Nishikawa, Yoshito; Ishida, Hitoshi; Kurozumi, Yasuo

2013-01-01

Ultrasonic testing is difficult to apply to cast stainless steel which is the material of the main coolant pipes in pressurized water reactors, because of the large attenuation and scattering of ultrasonic waves caused by its macro structure. In this study, ultrasonic testing for progression of axial fatigue cracks of a welded area in the test piece of cast stainless steel pipe was performed using double big-size ultrasonic probes which were formerly developed in INSS. It was found that detection of defects that were over 6% of the target depth for the specimen thickness of 69mm is possible, and detection of defects with over 10% of the target depth is possible for all test conditions. (author)

A study on prediction of metal loss by flow-accelerated corrosion in the CANDU NPP secondary piping systems

International Nuclear Information System (INIS)

Shim, S. H.; Song, J. S.; Yoon, K. B.; Hwang, K. M.; Jin, T. E.; Lee, S. H.; Kim, W. S.

2001-01-01

Flow-Accelerated Corrosion(FAC) is a phenomenon that results in metal loss from piping, vessels, and equipment made of carbon steel. FAC occurs only under certain conditions of flow, chemistry, geometry, and material. Unfortunately, those conditions are in much of the high-energy piping in nuclear and fossil-fueled power plants. Also, for domestic NPP secondary pipings whose operating time become longer, more evidences of FAC have been reported. The authors are studying on FAC management using CHECWORKS, computer code developed by EPRI. This paper is on the prediction results of metal loss by FAC in the one of CANDU type NPP secondary piping systems

Estimation of leak rate through circumferential cracks in pipes in nuclear power plants

Directory of Open Access Journals (Sweden)

Jai Hak Park

2015-04-01

Full Text Available The leak before break (LBB concept is widely used in designing pipe lines in nuclear power plants. According to the concept, the amount of leaking liquid from a pipe should be more than the minimum detectable leak rate of a leak detection system before catastrophic failure occurs. Therefore, accurate estimation of the leak rate is important to evaluate the validity of the LBB concept in pipe line design. In this paper, a program was developed to estimate the leak rate through circumferential cracks in pipes in nuclear power plants using the Henry–Fauske flow model and modified Henry–Fauske flow model. By using the developed program, the leak rate was calculated for a circumferential crack in a sample pipe, and the effect of the flow model on the leak rate was examined. Treating the crack morphology parameters as random variables, the statistical behavior of the leak rate was also examined. As a result, it was found that the crack morphology parameters have a strong effect on the leak rate and the statistical behavior of the leak rate can be simulated using normally distributed crack morphology parameters.

Data book of examination of the ruptured pipe at the Hamaoka Nuclear Power Station Unit-1

International Nuclear Information System (INIS)

2002-03-01

In order to investigate root cause of the pipe rupture, which took place at the Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company on November 7, 2001, a task force was established within the Nuclear and Industrial Safety Agency (NISA) and initiated a detailed investigation of the ruptured pipe. The Japan Atomic Energy Research Institute (JAERI) was asked from the Ministry of Education, Culture, Sports, Science and Technology (MEXT) in response to the request from NISA to cooperate as an independent neutral organization with NISA and perform an examination of the ruptured pipe independently from Chubu Electric Power Company. JAERI accepted the request by considering the fact that JAERI is an integrated research institution for nuclear research and development, a prime research institution for nuclear safety research, a research institution with experience of root-cause investigation of various nuclear incidents and accidents of domestic as well as overseas, and a research institution provided with advanced examination facilities necessary for examination of the ruptured pipe. The JAERI examination group was formed at the Tokai Research Establishment and conducted detailed and thorough examination of the pieces taken from the ruptured pipe primarily in the Reactor Fuel Examination Facility (RFEF) with the use of tools such as scanning electron microscopes and other equipments. Purpose of examination was to provide technical information in order to identify causes of the pipe rupture through examination of the pieces taken from the ruptured region of the pipe. The result of the present examination has already been reported to NISA and has also been published as the JAERI-Tech report No.2001-94. This report is a data book containing the detailed data obtained by the present examination. (author)

Sodium heat pipe module test for the SAFE-30 reactor prototype

International Nuclear Information System (INIS)

Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

2001-01-01

Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

Structural damping results from vibration tests of straight piping sections

International Nuclear Information System (INIS)

Ware, A.G.; Thinnes, G.L.

1984-01-01

EG and G Idaho is assisting the USNRC and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on 76-mm and 203-mm (3-in. amd 8-in.) Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 10-m (33-ft) straight sections of piping were rigidly supported at the ends. Spring, rod, and constant force hangers, as well as a sway brace and snubbers were included as intermediate supports. Excitation was provided by low-force level hammer inpacts, a hydraulic shaker, and a 445-kN (50-ton) overhead crane. Data was recorded using acceleration, strain, and displacement time histories. This paper presents results from the testing showing the effect of stress level and type of supports on structural damping in piping

Manufacture of mold of polymeric composite water pipe reinforced charcoal

Science.gov (United States)

Zulfikar; Misdawati; Idris, M.; Nasution, F. K.; Harahap, U. N.; Simanjuntak, R. K.; Jufrizal; Pranoto, S.

2018-03-01

In general, household wastewater pipelines currently use thermoplastic pipes of Polyvinyl Chloride (PVC). This material is known to be not high heat resistant, contains hazardous chemicals (toxins), relatively inhospitable, and relatively more expensive. Therefore, researchers make innovations utilizing natural materials in the form of wood charcoal as the basic material of making the water pipe. Making this pipe requires a simple mold design that can be worked in the scale of household and intermediate industries. This research aims to produce water pipe mold with simple design, easy to do, and making time relatively short. Some considerations for molding materials are weight of mold, ease of raw material, strong, sturdy, and able to cast. Pipe molds are grouped into 4 (four) main parts, including: outer diameter pipe molding, pipe inside diameter, pipe holder, and pipe alignment control. Some materials have been tested as raw materials for outer diameter of pipes, such as wood, iron / steel, cement, and thermoset. The best results are obtained on thermoset material, where the process of disassembling is easier and the resulting mold weight is relatively lighter. For the inside diameter of the pipe is used stainless steel, because in addition to be resistant to chemical processes that occur, in this part of the mold must hold the press load due to shrinkage of raw materials of the pipe during the process of hardening (polymerization). Therefore, it needs high pressure resistant material and does not blend with the raw material of the pipe. The base of the mold is made of stainless steel material because it must be resistant to corrosion due to chemical processes. As for the adjustment of the pipe is made of ST 37 carbon steel, because its function is only as a regulator of the alignment of the pipe structure.

Accumulation and preparation of nondestructive inspection data for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

In recent years, flaws due to stress corrosion cracking (SCC) in stainless steel piping and nickel based alloy welds were detected at nuclear power plants in Japan. Weld Overlay (WOL) has been developed as a repairing method for piping items without removing flaws. Since the inspection techniques for WOL pipes and nickel based alloy welds are not verified enough in Japan, Japan Nuclear Energy Safety Organization (JNES) has been carrying out a six year research project entitled 'Accumulation and Preparation of Nondestructive Inspection Data for Nuclear Power Plants' regarding nondestructive inspection since FY2007. In this research project, detection and sizing capability of SCC by Ultrasonic Testing (UT) are evaluated using mockup tests. In addition, the results of this project and past nondestructive inspection data performed by JNES projects are gathered, and inputted into the database of NDT information. In FY2012, followings were conducted. 1) Analysis for UT measurement results of nickel based alloy weld simulating safe end of reactor vessel outlet nozzle. 2) Analysis for UT measurement results of cast stainless steel piping. 3) Development of interface of UT simulation. 4) Development of nondestructive testing guideline. (author)

Structural integrity evaluation of nuclear piping cracket

International Nuclear Information System (INIS)

Cadiz Deleito, J.C.

1985-01-01

The methodology to evaluation of cracks in nuclear piping is exposed. Linear elastic fracture mechanic is used to prediction of growing crack and the net section collapse theory compared with acceptation criteria of both ASME III and ASME XI code. A case allowable under ASME XI criteria is analysed under ASME III requirements. Consideration must be given to local phenomenon in crack area and local stress evaluated and compared with ASME III acceptation criteria. (author)

Using Low-Frequency Phased Arrays to Detect Cracks in Cast Austenitic Piping Components

International Nuclear Information System (INIS)

Anderson, Michael T.; Cumblidge, Stephen E.; Doctor, Steven R.

2005-01-01

As part of a multi-year program funded by the United States Nuclear Regulatory Commission (US NRC) to address NDE reliability of inservice inspection (ISI) programs, recent studies conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, have focused on assessing novel NDE approaches for the inspection of coarse-grained, cast stainless steel reactor components. The primary objective of this work is to provide information to the US NRC on the utility, effectiveness and reliability of ultrasonic testing (UT) and eddy current testing (ET) inspection techniques as related to the ISI of primary piping components in pressurized water reactors (PWRs). This paper describes progress, recent developments and early results from an assessment of a portion of this work relating to the ultrasonic low frequency phased array inspection technique. Westinghouse Owner's Group (WOG) cast stainless steel pipe segments with thermal and mechanical fatigue cracks, PNNL samples containing thermal fatigue cracks and several blank vintage specimens having very coarse grains that are representative of early centrifugally cast piping installed in PWRs, are being used for assessing the inspection method. The phased array approach was implemented using an R/D Tech Tomoscan III system operating at 1.0 MHz and 500 kHz, providing composite volumetric images of the samples. Several dual, transmit-receive, custom designed low-frequency arrays are employed in laboratory trials. Results from laboratory studies for assessing detection of thermal and mechanical fatigue cracks in cast stainless steel piping welds are discussed

Seismic ratchet-fatigue failure of piping systems

International Nuclear Information System (INIS)

Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

1987-01-01

Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-inch and a pressurized 6-inch diameter carbon steel nuclear pipe systems subjected to high-level shaking have been accomplished. The high-level shaking loads needed to cause failure were much higher than ASME Code rules would permit with present design limits. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occured in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate reasonably well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules to reduce unneeded conservatisms and to cover the ratchet-fatigue failure mode may be appropriate

Pile load test on large diameter steel pipe piles in Timan-Pechora, Russia

Energy Technology Data Exchange (ETDEWEB)

McKeown, S. [Golder Associates Inc., Houston, TX (United States); Tart, B. [Golder Associates Inc., Anchorage, AK (United States); Swartz, R. [Paragon Engineering Services Inc., Houston, TX (United States)

1994-12-31

Pile load testing conducted in May and June of 1993 at the Polar Lights Ardalin project in Arkangelsk province, Russia, was documented. Pile load testing was conducted to determine the ultimate and allowable pile loads for varying pile lengths and ground temperature conditions and to provide creep test data for deformation under constant load. The piles consisted of 20 inch diameter steel pipe piles driven open ended through prebored holes into the permafrost soils. Ultimate pile capacities, adfreeze bond, and creep properties observed were discussed. 10 figs., 4 tabs.

Safety catching device for pipe lines in missile shielding cylinders of nuclear power plants

International Nuclear Information System (INIS)

Hering, S.; Doll, B.

1975-01-01

The safety catching device for pipes in the missile shielding cylinders consists of a flexible steel cable surrounding the pipe in a distance in U-shape. The arrester cable - which works as a spring and is freely movable in all directions - is attached to the cylinder wall. For this, the ends of the cable are primarily fastened to anchor boxes which are then inserted in a stay tube with the same axis as the cable ends. The anchor boxes are fastened to the outer wall of the missile shielding cylinder by anchor bolts and holding plates. (DG/AK) [de

Rupture hardware minimization in pressurized water reactor piping

International Nuclear Information System (INIS)

Mukherjee, S.K.; Ski, J.J.; Chexal, V.; Norris, D.M.; Goldstein, N.A.; Beaudoin, B.F.; Quinones, D.F.; Server, W.L.

1989-01-01

For much of the high-energy piping in light reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but also improves the overall safety and integrity of the plant since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied a Beaver Valley Power Station- Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferrutic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in. (152-mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel line as small as 3-in. (76-mm) diameter (outside containment) can qualify for pipe rupture hardware elemination

Steel for nuclear applications

International Nuclear Information System (INIS)

Zorev, N.N.; Astafiev, A.A.; Loboda, A.S.

1978-01-01

A steel contains, in percent by weight, the following constituents: carbon from 0.13 to 0.18, silicon from 0.17 to 0.37, manganese from 0.30 to 0.60, chromium from 1.7 to 2.4, nickel from 1.0 to 1.5, molybdenum from 0.5 to 0.7, vanadium from 0.05 to 0.12, aluminium from 0.01 to 0.035, nitrogen from 0.05 to 0.012, copper from 0.11 to 0.20, arsenic from 0.0035 to 0.0055, iron and impurities, the balance. This steel is preferable for use in the manufacture of nuclear reactors. 1 table

Application of Lab View Software for Thinning Measurement of Steel Pipe Sample by Using Gamma Ray Techniques

International Nuclear Information System (INIS)

Wibisono; Sugiharto

2004-01-01

The computer program, namely scaling pipe has been constructed to support the work of radiation monitoring either utilizing of sealed or unsealed source as well. The paper describes the performance of that computer program which is able to display numerical data coming from rate-meter to become graphical data and its ability to store data in form of software file with text format. The stored file can be performed by other software therefore the data can be calculated and analyzed. This computer program has been applied to measure the thickness of steel pipe with diameter of 3 inch which is machined with the different of 0.1 mm for each 50 mm length of the pipe. In this research the gamma radiation originated from 200 mCi of 241 Am source has been passed through the center of the pipe and the dose rate at the opposite of the source is measured. The attenuation coefficient calculated from that measurement is 0.0944 mm -1 with the deviation factor of 44 cps/mm. (author)

Parametric studies for stress corrosion in Type 304 stainless steel pipe

International Nuclear Information System (INIS)

Horn, R.M.

1984-01-01

Stress corrosion tests were conducted in the General Electric Pipe Test Laboratory using 4-inch diameter welded pipe to evaluate the role of stress, oxygen level, cyclic loading rate, temperature, and material composition on the intergranular stress corrosion cracking (IGSCC) behavior of welded Type-304 stainless steel in high temperature, high purity water. The role of applied stress was evaluated in environments containing either 0.2 ppm or 8 ppm oxygen. The tests established that applied stress is the dominant variable among those studied. An increase in applied axial stress from 116 MPa (16.9 ksi) to 254 MPa (36.9 ksi) produced up to a 30 old decrease in lifetime. The change in oxygen level from 0.2 to 8 ppm produced up to a factor of four decrease in lifetime. The role of cyclic loading rate, investigated with only limited tests, was found to accelerate failure at high applied stresses. Finally one test was conducted at 232 0 C with no effect on pipe lifetime. The effects of the above parameters were defined using one heat of material. To compare the results with that of other susceptible heats, additional tests were conducted using material taken from an archive heat that had cracked in the field and from a second heat with lower carbon content that had not cracked in the field. The archive heat exhibited lifetimes that were consistent with the other test results. The low carbon material did not fail demonstrating its much reduced cracking tendency

Failure mode and fracture behavior evaluation of pipes with local wall thinning subjected to bending load

International Nuclear Information System (INIS)

Ahn, Seok Hwan; Nam, Ki Woo; Kim, Seon Jin; Kim, Jin Hwan; Kim, Hyun Soo; Do, Jae Yoon

2003-01-01

Fracture behaviors of pipes with local wall thinning are very important for the integrity of nuclear power plant. In pipes of energy plants, sometimes, the local wall thinning may result from severe Erosion-Corrosion (E/C) damage. However, the effects of local wall thinning on strength and fracture behaviors of piping system were not well studied. In this paper, the monotonic bending tests were performed of full-scale carbon steel pipes with local wall thinning. A monotonic bending load was applied to straight pipe specimens by four-point loading at ambient temperature without internal pressure. From the tests, fracture behaviors and fracture strength of locally thinned pipe were manifested systematically. The observed failure modes were divided into four types; ovalization, crack initiation/growth after ovalization, local buckling and crack initiating/growth after local buckling. Also, the strength and the allowable limit of piping system with local wall thinning were evaluated

Random cyclic stress-strain responses of a stainless steel pipe-weld metal. I. A statistical investigation

International Nuclear Information System (INIS)

Zhao, Y.X.; Wang, J.N.

2000-01-01

For pt.II see ibid., vol.199, p.315-26, 2000. This paper pays a special attention to the issue that there is a significant scatter of the stress-strain responses of a nuclear engineering material, 1Cr18Ni9Ti stainless steel pipe-weld metal. Statistical investigation is made to the cyclic stress amplitudes of this material. Three considerations are given. They consist of the total fit, the consistency with fatigue physics and the safety in practice of the seven commonly used statistical distributions, namely Weibull (two- and three-parameter), normal, lognormal, extreme minimum value, extreme maximum value and exponential. Results reveal that the data follow meanwhile the seven distributions but the local effects of the distributions yield a significant difference. Any of the normal, lognormal, extreme minimum value and extreme maximum value distributions might be an appropriate assumed distribution for characterizing the data. The normal and extreme minimum models are excellent. Other distributions do not fit the data as they violate two or three of the mentioned considerations. (orig.)

Evaluation of piping fracture analysis method by benchmark study, 1

International Nuclear Information System (INIS)

Takahashi, Yukio; Kashima, Koichi; Kuwabara, Kazuo

1987-01-01

Importance of strength evaluation methods for cracked piping is growing with the progress of the rationalization of the nuclear piping system based on the leak-before-break concept. As an analytical tool, finite element method is principally used. To obtain the reliable solutions by the finite element programs, it is important to grasp the influences of various factors on the solutions. In this study, benchmark analysis is carried out for a stainless steel pipe with a circumferential through-wall crack subjected to four-point bending loading. Eight solutions obtained by using five finite element programs are compared with each other. Good agreement is obtained between the solutions on the deformation characteristics as well as fracture mechanics parameters. It is found through this study that the influence of the difference in the solution technique is generally small. (author)

Supply chain of steel industries for the nuclear power plant construction in Indonesia

International Nuclear Information System (INIS)

Dharu Dewi; Sahala M Lumbanraja

2017-01-01

Nuclear Power Plant (NPP) Construction needs steel materials for the manufacturing of heavy components and civil work construction. National industries is expected to supply steel components especially for non nuclear component needs. Supply chain of steel industries is required to know the potency of steel industries from upstream to downstream industries which can support the NPP construction sustainability. The type of steel needed in the NPP construction consist of structure steel, rebar, steel plate, etc. The aim of the study is to identify supply chain of steel industries from upstream industries to downstream industries so that they can supply steel needs in the NPP construction. The methodology used are literature review and industries survey by purposive sampling test which sent questionnaires and carrying out technical visits to the potential industries to supply steel components for NPP construction. From the analysis of the questionnaires and survey, it has been obtained that the Indonesian steel industries capable of supplying steel for construction materials of non-nuclear parts are PT. Krakatau Steel, PT. Gunung Steel Group (PT Gunung Garuda and PT. Gunung Raja Paksi), PT. Cilegon Fabricators and PT. Ometraco Arya Samanta. While steel materials for primary components with nuclear grade, such as steel materials for reactor vessels and pressure vessels, the Indonesian steel industry has not been able to supply them. Therefore, the Indonesian steel industries must improve its capability, both in raw material processing and fabrication capability in order to meet the requirements of specifications, codes and standards of nuclear grade. (author)

Variations of fracture toughness and stress-strain curve of cold worked stainless steel and their influence on failure strength of cracked pipe

International Nuclear Information System (INIS)

Kamaya, Masayuki

2016-01-01

In order to assess failure probability of cracked components, it is important to know the variations of the material properties and their influence on the failure load assessment. In this study, variations of the fracture toughness and stress-strain curve were investigated for cold worked stainless steel. The variations of the 0.2% proof and ultimate strengths obtained using 8 specimens of 20% cold worked stainless steel (CW20) were 77 MPa and 81 MPa, respectively. The respective variations were decreased to 13 and 21 MPa for 40% cold worked material (CW40). Namely, the variation in the tensile strength was decreased by hardening. The COVs (coefficients of variation) of fracture toughness were 7.3% and 16.7% for CW20 and CW40, respectively. Namely, the variation in the fracture toughness was increased by hardening. Then, in order to investigate the influence of the variations in the material properties on failure load of a cracked pipe, flaw assessments were performed for a cracked pipe subjected to a global bending load. Using the obtained material properties led to variation in the failure load. The variation in the failure load of the cracked pipe caused by the variation in the stress-strain curve was less than 1.5% for the COV. The variation in the failure load caused by fracture toughness variation was relatively large for CW40, although it was less than 2.0% for the maximum case. It was concluded that the hardening induced by cold working does not cause significant variation in the failure load of cracked stainless steel pipe. (author)

Evaluation of flawed-pipe experiments: Final report

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Gamble, R.M.

1986-11-01

The purpose of this work was to perform elastic plastic fracture mechanics evaluations of experimental data that have become available from the NRC Degraded Pipe Program, Phase II (DPII) and other NRC and EPRI sponsored programs. These evaluations were used to assess flaw evaluation procedures for austenitic and ferritic steel piping. The results also have application to leak before break fracture mechanics analysis. An improved relationship was developed for computing the J-Integral for pipes containing throughwall flaws and loaded in pure bending. The results from several DPII experiments were compared to predictions based on new J estimation scheme solutions for circumferential, finite length part-throughwall flaws in pipes with bending loading. Comparisons of experimental maximum loads with those predicted using procedures in Paragraph IWB-3640, Section XI of the ASME Code indicate that the Code flaw evaluation procedures and allowables for austenitic steel pipe are appropriate and conservative. However, the comparisons also indicate that the base metal Code allowable loads may be about 15 to 20% high for small diameter piping (less than 8-inch diameter) at allowable a/t larger than about 0.5. The work further indicates that there is justification for reducing the conservatism in IWB-3640 allowable flaw sizes and loads for austenitic steel pipe with submerged or shielded metal arc welds.

76 FR 78614 - Welded ASTM A-312 Stainless Steel Pipe From South Korea and Taiwan: Continuation of Antidumping...

Science.gov (United States)

2011-12-19

... DEPARTMENT OF COMMERCE International Trade Administration [A-580-810, A-583-815] Welded ASTM A-312... revocation of the antidumping duty orders on welded ASTM A-312 stainless steel pipe from South Korea (Korea... December 30, 1992, the Department published the antidumping duty orders on welded ASTM A-312 stainless...

Development of automatic pipe welder for nuclear power plant

International Nuclear Information System (INIS)

Iwamoto, Taro; Ando, Shimon; Omae, Tsutomu; Ito, Yoshitoshi; Araya, Takeshi.

1978-01-01

Numerous pipings are installed in nuclear power plants, and of course, the reliability of these pipings are very important to preserve the safety of the plants. These pipings undergo periodic inspection yearly, and when some defects are found or some reconstructions to superior systems are made, field welding in the plants is required. When the places to be welded are in containment vessels, the works must be carried out in radiation environment. In order to maintain the highest quality of welding and to reduce the radiation exposure of workers, many skilled workers are required. This automatic pipe welder was developed to solve these problems, aiming at carrying out welding works by remote control at the safe places outside containment vessels. Especially in order to obtain the highest quality of welding, it was not perfectly automated, but the man-machine system so as to enable to utilize the delicate sense of workers was adopted. The visual and contact detecting systems to monitor welding works, remote control system, computer control, light, small and easily installed welding head, grinding and supersonic flow detecting equipments, the power source of transistor switching type, air cooling equipment, and the function for setting welding conditions according to algorithm were added to the welding machine. The outline and main components of this automatic pipe welder are explained. (Kako, I.)

Investigation of the ductile fracture properties of Type 304 stainless steel plate, welds, and 4-inch pipe

International Nuclear Information System (INIS)

Vassilaros, M.G.; Hays, R.A.; Gudas, J.P.

1985-01-01

J-integral fracture toughness tests were performed on welded 304 stainless steel 2-inch plate and 4-inch diameter pipe. The 2-inch plate was welded using a hot-wire automatic gas tungsten arc process. The tests were performed at 550 0 F, 300 0 F and room temperature. The results of the J-integral tests indicate that the Jsub(Ic) of the base plate ranged from 4400 to 6100 in lbs/in 2 at 550 0 F. The Jsub(Ic) values for the tests performed at 300 0 F and room temperature were beyond the measurement capacity of the specimens and appear to indicate that Jsub(Ic) was greater than 8000 in lb/in 2 . The J-integral tests performed on the weld metal specimens indicate that the Jsub(Ic) values ranged from 930 to 2150 in lbs/in 2 at 550 0 F. The Jsub(Ic) values of the weld metal specimens tested at 300 0 F and room temperature were 2300 and 3000 in lbs/in 2 respectively. One HAZ specimen was tested at 550 0 F and found to have a Jsub(Ic) value of 2980 in lbs/in 2 which indicates that the HAZ is an average of the base metal and weld metal toughness. These test results indicate that there is a significant reduction in the initiation fracture toughness as a result of welding. The second phase of this task dealt with the fracture toughness testing of 4-inch diameter 304 stainless steel pipes containing a gas tungsten arc weld. The pipes were tested at 550 0 F in four point bending. Three tests were performed, two with a through wall flaw growing circumferentially and the third pipe had a part through radial flaw in combination with the circumferential flaw. These tests were performed using unloading compliance and d.c. potential drop crack length estimate methods. The results of these tests indicate that the presence of a complex crack (radial and circumferential) reduces in the initiation toughness and the tearing modulus of the pipe material compared to a pipe with only a circumferentially growing crack. (orig.)

Underground nuclear power station using self-regulating heat-pipe controlled reactors

International Nuclear Information System (INIS)

Hampel, V.E.

1989-01-01

The author presents a nuclear reactor for generating electricity disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor

Underground nuclear power station using self-regulating heat-pipe controlled reactors

Science.gov (United States)

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Nuclear power supply (Japan Nuclear Safety Institute)

International Nuclear Information System (INIS)

Kameyama, Masashi

2013-01-01

After experienced nuclear disaster occurred on March 11, 2011, role of nuclear power in future energy share in Japan became uncertain because most public seemed to prefer nuclear power phase out to energy security or costs. Whether nuclear power plants were safe shutdown or operational, technologies were requisite for maintaining their equipment by refurbishment, partly replacement or pressure proof function recovery works, all of which were basically performed by welding. Nuclear power plants consisted of tanks, piping and pumps, and considered as giant welded structures welding was mostly used. Reactor pressure vessel subject to high temperature and high pressure was around 200mm thick and made of low-alloy steels (A533B), stainless steels (308, 316) and nickel base alloys (Alloy 600, 690). Kinds of welding at site were mostly shielded-metal arc welding and TIG welding, and sometimes laser welding. Radiation effects on welding of materials were limited although radiation protection was needed for welding works under radiation environment. New welding technologies had been applied after their technical validation by experiments applicable to required regulation standards. Latest developed welding technologies were seal welding to prevent SCC propagation and temper-bead welding for cladding after removal of cracks. Detailed procedures of repair welding of Alloy 600 at the reactor outlet pipe at Oi Nuclear Power Plants unit 3 due to PWSCC were described as an example of crack removal and water jet peening, and then overlay by temper-bead welding using Alloy 600 and clad welding using Alloy 690. (T. Tanaka)

Bending of pipes with inconel cladding; Curvamento de tubos revestidos com inconel

Energy Technology Data Exchange (ETDEWEB)

Nachpitz, Leonardo; Menezes, Carlos Eduardo B; Vieira, Carlos R. Tavares [PROTUBO, Macae, RJ (Brazil)

2008-07-01

The pipes used in offshore equipment, such as wet Christmas trees, sub sea manifolds and rigid platform risers, as well as some pipes for refineries and ships, must have mechanical resistance to high pressure and also be resistant to corrosion from acids. Some special materials, such as stainless steel, duplex and super duplex steel are used to resolve this problem, but the cost is very high. Besides the problem of cost, these materials have other drawbacks, such as the difficulty of welding them, a technology mastered in few countries. As a better alternative, the use has been increasing of carbon steel and API pipes coated internally with inconel by welding deposition. This groundbreaking technology, of proven efficiency, has a far superior cost-benefit relation. Carbon steel and API pipes, besides having better mechanical resistance to high pressure and corrosion resistance, can be fabricated with technology mastered worldwide. Nickel alloys, such as inconel, are highly resistant to corrosion and temperature, and in these aspects are better than stainless steels. The pipes for transportation equipment and for refining hydrocarbons, as mentioned above, require various turns and special geometries, which generally are solved by the use of bends and spools made by high-frequency induction. This technology, already well established for various carbon and stainless steels, was developed to work with pipes coated internally with inconel (inconel cladding). Therefore, our work describes the process of fabricating bends from API steel pipes with inconel cladding, demonstrating the efficacy of this technology along with its quality gains and cost reduction. (author)

Estimation of Leak Rate Through Cracks in Bimaterial Pipes in Nuclear Power Plants

Directory of Open Access Journals (Sweden)

Jai Hak Park

2016-10-01

Full Text Available The accurate estimation of leak rate through cracks is crucial in applying the leak before break (LBB concept to pipeline design in nuclear power plants. Because of its importance, several programs were developed based on the several proposed flow models, and used in nuclear power industries. As the flow models were developed for a homogeneous pipe material, however, some difficulties were encountered in estimating leak rates for bimaterial pipes. In this paper, a flow model is proposed to estimate leak rate in bimaterial pipes based on the modified Henry–Fauske flow model. In the new flow model, different crack morphology parameters can be considered in two parts of a flow path. In addition, based on the proposed flow model, a program was developed to estimate leak rate for a crack with linearly varying cross-sectional area. Using the program, leak rates were calculated for through-thickness cracks with constant or linearly varying cross-sectional areas in a bimaterial pipe. The leak rate results were then compared and discussed in comparison with the results for a homogeneous pipe. The effects of the crack morphology parameters and the variation in cross-sectional area on the leak rate were examined and discussed.

The characteristic investigation on narrow-gap TIG weld joint of heavy wall austenitic stainless steel pipe

International Nuclear Information System (INIS)

Shim, Deog Nam; Jung, In Cheol

2003-01-01

Although Gas Tungsten Arc Welding (GTAW or TIG welding) is considered as high quality and precision welding process, it also has demerit of low melting rate. Narrow-gap TIG welding which has narrow joint width reduces the groove volume remarkably, so it could be shorten the welding time and decrease the overall shrinkage in heavy wall pipe welding. Generally narrow-gap TIG welding is used as orbital welding process, it is important to select the optimum conditions for the automatic control welding. This paper looks at the application and metallurgical properties on narrow-gap TIG welding joint of heavy wall large austenitic stainless steel pipe to determine the deposition efficiency, the resultant shrinkage and fracture toughness. The fracture toughness depends slightly on the welding heat input

Fire protection in Angra-2 nuclear power plant. The use of fire protection collars on plastic piping systems

International Nuclear Information System (INIS)

Oliveira Segabinaze, R. de

1994-01-01

The object of this paper is to briefly the use of fire protection collars on plastic piping systems passing through wall and floor penetration. The fire protection collars consist of a stainless steel housing, in which the leading edges of two pivoting plates are in constant pressure contact with the pipe. In case of fire these plates react on the softened pipe with a guillotine action, thereby stopping the flow; within the housing a foam material expands to fill the space when subject to the heat of the fire. The piping project has to be modified to permit the fixing of the collars to walls and floor penetrations. (author). 2 refs, 9 figs

Steel castings of valves for nuclear power station

International Nuclear Information System (INIS)

Yamasaki, Yutaka

1975-01-01

The manufacturing of the steel castings of valves for nuclear power plants is reported. The report is divided in six parts. The first part describes the reliability of the steel castings of valves for nuclear power plants. Particular attention must be paid to larger diameter and lower pressure rating for the valves in nuclear power plants than those in thermal power plants. The second part describes the characteristics of steel casting quality, defects and their cause. The defects that may be produced in steel castings are as follows: (a) cavities caused by the insufficient supply of molten steel, (b) sand bites caused by the mold destruction due to thermal shock, and (c) pinholes caused by the gas absorption of molten steel. The third part describes the clarification of quality level and the measures quality project. Gaseous defects and the indications detected by magnetic powder test are attributed to electric furnace steel making. In particular, the method to minimize gas content is important. The fourth part describes the quality control of manufacturing processes. In practice, thirteen semi-automatic testers using gamma radiation are employed. A full automatic inspection plant having capacity of 20,000 radiographs per month is under design. The fifth part describes a quality warrant system. A check sheet system concerning quality and safety is employed in all work shops. The reliability of all testers and measuring instruments as well as the skill of workmen are examined periodically. The seventh part deals with future problems. The manufacturing plan must be controlled so that non-destructive inspection becomes the main means for quality control. (Iwakiri, K.)

Investigations into the fatigue behaviour of nuclear grades of austenitic stainless steel

International Nuclear Information System (INIS)

Mann, J.

2015-01-01

Full text of publication follows. Fatigue is an important problem within the nuclear industry due to the complex combination of thermal and mechanical loading that components experience during the operation of a nuclear reactor. Austenitic stainless steels are widely used within nuclear reactors for a number of applications including piping systems and pressure vessels. A number of studies have shown that austenitic stainless steel components operating within a light water reactor (LWR) environment may experience a significant reduction in fatigue life under certain circumstances, however the precise mechanisms responsible for the reduction are still not fully understood. The effects of environment are included in some fatigue assessment methods, however these are generally considered to be over-conservative and predicted fatigue lifetimes are not reflected well by service experience. This project aims to enhance the understanding of fatigue in both air and LWR environments through the synergistic use of a wide range of different microscopy techniques. It is expected that a better understanding of each of the different stages of fatigue will lead to more accurate fatigue predictions that ultimately result in better and safer lifetime predictions. This paper focuses on introducing the background behind the project, highlighting the current methods for assessing fatigue lifetimes and the motivations for the current research. The results of various initial microscopic investigations are presented, with a focus on a number of novel applications using laser scanning confocal microscopy to perform large scale analyses of fatigue fracture surfaces and test specimen gauge length surfaces. The use of surface replicas in conjunction with laser scanning confocal microscopy is discussed along with its potential applications for the assessment of fatigue damage in in-service components. Initial finite element modelling of crack growth within fatigue test specimens is discussed

Accelerated development of advanced steels for nuclear applications

International Nuclear Information System (INIS)

Ghoniem, N.; Zinkle, S.

2009-01-01

Significant progress has been achieved in the operational performance and radiation resistance of ferritic-martensitic steels during the past few decades. Conventional high temperature steels, such as HT-9 and 2 1/4 Cr-1Mo have evolved into super Oxide Dispersion Strengthened (ODS) steels through successive optimization to meet strict performance and radiation-resistance constraints. Such evolution was possible through a combination of experimentation, modeling and empirical information. Further development and optimization of structural steels in nuclear applications will require full utilization of the available array of sophisticated experimental techniques and multiscale computational modeling, in addition to empirical data. We present here a systematic approach to the process of optimum steel development, by linking material fabrication to thermo-mechanical properties through a physical understanding of microstructure evolution. The optimization process is based on the application of design constraints (e.g. low activation, low DBTT, low swelling, creep resistance, and weldability) to describe the required microstructures, which in turn, can be controlled through material processing techniques. Prospects for future design of optimum structural steels in nuclear applications by utilization of the hierarchy of multiscale experimental and computational strategies will be described. (author)

Corrosion of Spiral Rib Aluminized Pipe

Science.gov (United States)

2012-08-01

Large diameter, corrugated steel pipes are a common sight in the culverts that run alongside many Florida roads. Spiral-ribbed aluminized pipe (SRAP) has been widely specified by the Florida Department of Transportation (FDOT) for runoff drainage. Th...

Effect of prestrain on ductility and toughness in high strength line pipe steels

Energy Technology Data Exchange (ETDEWEB)

Shinohara, Y.; Besson, J. [Paristech, Evry (France). Centre des Materiaux, Mines Paris; Madi, Y. [Ecole d' Ingenieurs, Sceaux (France). Ermess EPF; Paristech, Evry (France). Centre des Materiaux, Mines Paris

2009-07-01

The anisotropic plasticity, ductility and toughness of an X100 steel pipeline was investigated both before and after a series of prestraining experiments. The aim of the study was to determine the effect of prestraining on ductility and toughness in high strength pipe steels. Results of the study showed that primary void growth and coalescence was dependent on initial plastic anisotropy and not dependent on tensile prestrain. Secondary void nucleation and growth was not influenced by either the initial plastic anisotropy or by prestraining. Scanning electron microscopy (SEM) studies showed that the main damage mechanism was the void growth of primary dimples. Dimples in the prestrained materials were larger than those observed in materials that had not been prestrained. However, the effect on prestrain on dimple size was limited. Results showed both plastic and rupture anisotropies. It was concluded that prestraining induces a decrease in ductility, but has a significant impact on toughness. 4 refs., 2 tabs., 12 figs.

Gel structure of the corrosion layer on cladding pipes of nuclear fuels

Czech Academy of Sciences Publication Activity Database

Medek, Jiří; Weishauptová, Zuzana

2009-01-01

Roč. 393, č. 2 (2009), s. 306-310 ISSN 0022-3115 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : cladding pipes of nuclear fuels * corrosion layer * zirconium alloys Subject RIV: JF - Nuclear Energetics Impact factor: 1.933, year: 2009

MODEL TESTS AND 3D ELASTIC FINITE ELEMENT ANALYSIS FOR STEEL PIPE PILES WITH WINGS IN STALLED IN SOIL CEMENT COLUMN

Science.gov (United States)

Tamai, Toshiyuki; Teramoto, Shuntarou; Kimura, Makoto

Steel pipe piles with wings installed in soil cement column is a composite foundation of pile consisting of soil improvement with cement and steel pipe with wings. This type of pile shows higher vertical bearing capacity when compared to steel pipe piles that are installed without soil cement. It is thought the wings contribute to higher bearing capacity of this type of piles. The wings are also thought to play the role of structural unification of pile foundations and load transfer. In this study, model test and 3D elastic finite element analysis was carried out in order to elucidate the effect of wings on the structural unification of pile foundation and the load transfer mechanism. Firstly, the model test was carried out in order to grasp the influence of pile with and without wings, the shape of wings of the pile and the unconfined compression strength of the soil cement on the structural unification of the pile foundation. The numerical analysis of the model test was then carried out on the intermediate part of the pile foundation with wings and mathematical model developed. Finally load tran sfer mechanism was checked for the entire length of the pile through this mathematical model and the load sharing ratio of the wings and stress distribution occurring in the soil cement clarified. In addition, the effect of the wing interval on the structural unification of the pile foundation and load transfer was also checked and clarified.

Pipe rupture hardware minimization in pressurized water reactor system

International Nuclear Information System (INIS)

Mukherjee, S.K.; Szyslowski, J.J.; Chexal, V.; Norris, D.M.; Goldstein, N.A.; Beaudoin, B.; Quinones, D.; Server, W.

1987-01-01

For much of the high energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but the overall safety and integrity of the plant are improved since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station - Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in (152 mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel lines as small as 3-in (76 mm) diameter (outside containment) can qualify for pipe rupture hardware elimination

Development of the ultrasonic technique for examination of centrifugally-cast stainless steel in pressure piping

International Nuclear Information System (INIS)

Jurenka, H.J.

1983-01-01

Centrifugally - cast stainless steel (CCSS) are used to manufacture a large variety of components in the nuclear industry. Weldments are also made to join these parts to carbon steel items. These welds are usually made of stainless steel or inconel alloys. CCSS is sophisticated material and justification for its use in critical components is safety and reliability. These steels, as any other materials, need to be inspected to assess their structural integrity. Ultrasonic testing is one of the possible techniques. In some cases it is the only one of the feasible methods for this examination. This mainly concerns components in the primary and auxiliary circuits of nuclear plants. For a long time it has been recognized that CCSS items can be extremely difficult to inspect using ultrasonics. Many attempts in various research laboratories were conducted to improve the testing technique

Energy-saving by the optimization of the operation conditions in the vinylcloride lining steel pipe baking furnace. Enbirainingu kokan yakitsukero no sogyo joken saitekika ni yoru sho energy

Energy Technology Data Exchange (ETDEWEB)

Kinoshita, Y. (Kawasaki Steel Corp., Tokyo (Japan))

1991-02-04

By the investigation of the operation conditions in the small experimental furnace, the improvement of the heat pattern of the steel pipes, and the remodeling and arrangement of the furnace, the progress of the yield and the productivity, and the reduction of the fuel unit were expected. In baking of a PVC pipe with a steel pipe, defects of PVC pipes as the blister of the inner surface occurred often. Accordingly the conditions when the blister occurred were investigated in the experimental furnace and optimized operation conditions when it does not occur were established. Next the flow rate of the hot air and the temperature distribution in the actual furnace were examined and the mechanism of occurrence of the defects of the blister of the inner surface was investigated. As a result, it was proved that the deviation of the hot air flux and the intrusion of the outer air cause it, and the places where the outer air intruded were closed by steel plates. After the improvements the rate of the defects of PVC pipes of the blister of the inner surface was decreased from 5% to 0.5%. Further, by the optimization of the operation conditions the fuel unit decreased and 31% of the rate of the energy-saving was achieved. The reduction of 10,414,000 yen a year was achieved. 29 figs., 4 tabs.

Pipe support

International Nuclear Information System (INIS)

Pollono, L.P.

1979-01-01

A pipe support for high temperature, thin-walled piping runs such as those used in nuclear systems is described. A section of the pipe to be suppported is encircled by a tubular inner member comprised of two walls with an annular space therebetween. Compacted load-bearing thermal insulation is encapsulated within the annular space, and the inner member is clamped to the pipe by a constant clamping force split-ring clamp. The clamp may be connected to pipe hangers which provide desired support for the pipe

Comparison of ICEPEL predictions with single elbow flexible piping system experiment

International Nuclear Information System (INIS)

A-Moneim, M.T.; Chang, Y.W.

1978-01-01

The ICEPEL Code for coupled hydrodynamic-structural response analysis of piping systems is used to analyze an experiment on the response of flexible piping systems to internal pressure pulses. The piping system consisted of two flexible Nickel-200 pipes connected in series through a 90 0 thick-walled stainless steel elbow. A tailored pressure pulse generated by a calibrated pulse gun is stabilized in a long thick-walled stainless steel pipe leading to the flexible piping system which ended with a heavy blind flange. The analytical results of pressure and circumferential strain histories are discussed and compared against the experimental data obtained by Stanford Research Institute

A numerical study on the conjugate natural convection in a circular pipe containing water

Energy Technology Data Exchange (ETDEWEB)

Choi, Myoung-Young; Choi, Hyoung-Gwon [Seoul National University of Science & Technology, Seoul (Korea, Republic of)

2017-07-15

In this paper, the effect of material property of pipe on the conjugate natural convection in a circular pipe containing water was investigated by solving the unsteady incompressible Navier-Stokes equations coupled with energy equations of the water and pipe. Natural convection and conduction of water inside the pipe was coupled with the conduction of the pipe whose bottom was subject to uniform heat source. From the present grid resolution and time-step independent solutions, it has been confirmed that the water temperature inside a PVC pipe was higher than that inside a steel pipe due to the smaller heat capacity of PVC and that the streamline patterns of the two cases were found to be opposite because the thermal diffusivity of steel (PVC) is larger (smaller) than that of water such that steel (PVC) pipe is heated faster (slower) than water. Furthermore, a quantitative comparison of heat flux to water was performed by examining the distributions of the heat flux along the inside walls of steel/PVC. The average temperature of water inside steel was found to be higher than that inside PVC at the initial stage of heating. On the other hand, PVC provided a larger heat flux to water when it reached a steady value.

A fatigue initiation parameter for gas pipe steel submitted to hydrogen absorption

Energy Technology Data Exchange (ETDEWEB)

Capelle, J; Gilgert, J; Pluvinage, G [LaBPS - Ecole Nationale d' Ingenieurs de Metz et Universite Paul Verlaine Metz, Ile du Saulcy, 57045 Metz (France)

2010-01-15

Fatigue initiation resistance has been determined on API 5L X52 gas pipe steel. Tests have been performed on Roman Tile (RT) specimen and fatigue initiation was detected by acoustic emission. A comparison between specimens electrolytically charged with hydrogen and specimens without hydrogen absorption were made and it has been noted that fatigue initiation time is reduced of about 3 times when hydrogen embrittlement occurs. It has been proposed to use the concept of Notch Stress Intensity Factor as parameter to describe the fatigue initiation process. Due to the fact that hydrogen is localised in area with high hydrostatic pressure, definitions of local effective stress and distance have been modified when hydrogen is absorbed. This modification can be explained by existence of a ductile-brittle transition with hydrogen concentration. The fatigue initiation resistance curve allows that to determine a threshold for large number of cycles of fatigue non initiation. This parameter introduced in a Failure Assessment Diagram (FAD) provides supplementary information about defect nocivity in gas pipes: a non-critical defect can be detected as dormant or not dormant defect i.e., as non propagating defect. (author)

Electrical design of a 110-ft long muon pipe with automatic degaussing

International Nuclear Information System (INIS)

Visser, A.T.

1985-11-01

This memo describes a magnetized cylindrical pipe made from tape wound grain oriented low carbon steel rolls. Grain oriented steel yields much higher magnetic fields at low ampereturns than cast iron or other steel pipes. This is especially important when only a few windings are allowed in the inner bore. The power supply and operating cost are also much lower. The pipe has a high (approx.9 kG) remnant field, but is automatically degaussed upon shutdown of the DC excitation power supply. A remnant field detector senses whether degaussing was successful. The pipe is used in the muon beam line. Its magnetic field deflects unwanted halo muons. Tests need to be conducted with and without pipe field. It is therefore desirable that the pipe field automatically returns to zero when the DC excitation is shut off. This can be rather easily accomplished

A simple computational method for predicting magnetic field in the vicinity of a three-phase underground cable with a fluid-filled steel-pipe enclosure

International Nuclear Information System (INIS)

Xu, X.B.; Yang, X.M.

1994-01-01

This paper presents a simple computational method for predicting the magnetic field above ground, generated by an underground three-phase pipe-type cable. In the computation, an approximation is made to simplify the problem a Fourier series technique and an iterative procedure are employed to handle the nonlinear B-H characteristic of the steel pipe. To validate the computational method, measurements were made and the numerical results are compared with the measurement data. Also, data of magnetic fields generated by the pipe type cable are compared with those due to the cable in absence of the pipe. The advantages and disadvantages of this simple method are discussed

The conservatism of the net-section stress criterion for the failure of cracked stainless steel piping

International Nuclear Information System (INIS)

Smith, E.

1991-01-01

The failure of cracked stainless steel piping can be predicted by assuming that failure conforms to a net-section stress criterion, using as input an appropriate value for the critical net-section stress together with a knowledge of the anticipated loadings. The stresses at the cracked section are usually calculated via a purely elastic analysis based on the piping being uncracked. However because the piping is built-in at its ends into a larger component, this limits the amount of elastic follow-up and, consequently, use of the net-section stress approach in this manner can lead to conservative failure predictions. This paper quantifies the extent of this conservatism, and shows that it can be quite marked. There is an additional measure of conservatism due to the fact that unstable failure need not necessarily be associated with the onset of crack extension. A key parameter with regard to both these conservatisms is L EFF , a length parameter which is a measure of the degree of elastic follow-up in the system. (author)

Management of manufacture and installation of plant pipings by bar code system

International Nuclear Information System (INIS)

Suwa, Minoru

1995-01-01

As for the piping system of nuclear power plants, the number of parts is very large, and the mill sheet is attached to each part, therefore, it is necessary to manage them individually, and large man power is required. In order to resolve the delay of mechanization in the factory, bar code system was adopted on full scale. At the time of taking parts out from the store, bar code labels are stuck to all piping parts. By this means, all the processes of manufacture and inspection are managed with a computer, and it is useful for labor saving and the prevention of mistaken input. This system is centering around the system of the progress management for piping manufacture, and is operated by being coupled with respective systems of production design, order and inventory, mill sheet management and installation management. The management of production design, manufacture, inspection and installation is explained. There is the problem of sticking bar code labels again as the labels become dirty or parts pass through coating and pickling processes. The direct carving of bar codes on parts by laser marker was tried, and it was successful for stainless steel, but in carbon steel pipes, it was hard to read. It is desirable to develop the bar codes which endure until the end of plant life. (K.I.)

Nuclear power plant steam pipes repairing with TIRANT 3 robot system

International Nuclear Information System (INIS)

Soto Tomas, Marcelo; Curiel Nieva, Marceliano; Monzo Blasco, Enrique; Rodriguez, Salvador Pineda; Vaquer Perez, Juan I.

2011-01-01

A typical application functions covering the steam pipes inner surface in coal-fired power station and nuclear power plants. The results of this process are spectacular in terms of protection against corrosion and abrasion, but its application has conditioning factors, such as: Severe application conditions for workers. Due to the postural position (usually kneeling) in small diameter pipes and working with fireproof clothing and masks with outdoor air supplying, due to fumes, sparks and molten metal particles, radiological contamination, confined space, poor lighting... Coating uniformity. As metallization is a manual process, the carried out measurements show small variations in the thickness of the coating, always within the tolerance limits established by the applicable regulations and quality assurance. For all these reasons, Grupo Dominguis has developed the TIRANT 3 robot, a worldwide innovative system, for metallization of steam pipes inner surface. TIRANT 3 robot is teleoperated from outside of the pipe, so that human intervention is reduced to the operations of robot positioning and change of metallization wire. As it is an independent system of the human factor, metallization process performance is significantly increased by reducing rest periods due only to the robot maintenance. Likewise, TIRANT 3 system permits to increase resulting coating uniformity, and thus its resistance, keeping selected parameters constant depending on required type and thickness of wire. TIRANT 3 system has successfully worked in 2010 during the stops refueling of the Units I and II of Laguna Verde nuclear power plant in Mexico. (author)

The Analysis of the Field Application Methodology of Electromagnetic Ultrasonic Testing for Piping in Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Park, Chi Seung; Joo, Keum Jong; Choi, Jung Kweun; Um, Byung Kook; Park, Jea Suk [Korea Advanced Ispection Technology Co., Daejeon (Korea, Republic of)

2008-08-15

Nuclear plant piping is classified as the safety class and non-safety class piping in usual. Safety class piping has been examined in accordance with ASME Section XI and V during PSI/ISI using RT, UT, PT, ECT, etc and evaluated periodically for integrity. But failures in piping had reported at non-welded parts and non-safety class pipings as well as the safety class pipings. The existing NDT methods are suitable for the specific parts for instance weldments to inspect but difficult to examine all parts (total coverage) of pipe line and very expensive in cost and consume the time. And also inspection using those methods is difficult and limited for the parts which are complex configuration, embedded under ground and installed at high radiation area in nuclear power plants. In order to inspect all parts of long range piping systems and reduce the inspection time and cost, the electromagnetic ultrasonic inspection technology is suitable and effective. The electromagnetic ultrasonic method can cover more than 50 m apart from sensor at one time without moving the sensor and examined the parts which are in difficulties for accessibility, for example, high radiation area, insulated components and embedded under ground.

Team effort leads to versatile handling solution for pipe manufacturer

Energy Technology Data Exchange (ETDEWEB)

Anon.

2010-09-15

This article discussed the development of a new pipe-handling system that resulted in increased efficiencies in plant-to-yard transport for a custom steel pipe manufacturer. In the previous system, loaders would move finished pipe to the yard for storage. However, for transport loading, the pipe would have to be brought back indoors because only the inside cranes could handle loading the pipe without damaging the special outer coating on the pipe. In the new pipe-handling system, the loader is replaced with a Sennebogen 850 M rubber-tired material handler, which was developed for the steel recycling industry. The generator that comes on the material handler is retrofitted to power a purpose-built pipe-handler attachment. The machine's higher lifting reach allows for higher stacking, effectively increasing the capacity of the yard. The new pipe-handling machine allows trucks to be loaded right in the yard, eliminating the need to double-handle the pipe. 1 fig.

Comparison of fracture toughness values from an IPIRG-1 large-scale pipe system test and C(T) specimens on wrought TP304 stainless steel

International Nuclear Information System (INIS)

Olson, R.J.; Scott, P.; Marschall, C.W.; Wilkowski, G.M.

1994-01-01

Within the First International Piping Integrity Research Group (IPIRG-1) program, pipe system experiments involving dynamic loading with intentionally circumferentially cracked pipe were conducted. The pipe system was fabricated from 406-mm (16-inch) diameter Schedule 100 pipe, and the experiments were conducted at a pressure of 15.5 MPa (2,250 psi) and 288 C (550 F). The loads consisted of pressure, dead-weight, thermal expansion, inertia, and dynamic anchor motion. Significant instrumentation was used to allow the material fracture resistance to be calculated from these large-scale experiments. Three independent analyses were used to calculate the toughness directly from one of these pipe experiments. A comparison of the toughness values from the stainless steel base metal pipe experiment to standard quasi-static and dynamic C(T) specimen tests showed the pipe toughness value was significantly lower than that obtained from C(T) specimens. It is hypothesized that the cyclic loading from inertial stresses in this pipe system experiment caused local degradation of the material toughness. Such effects are not considered in current LBB or pipe flaw evaluation criteria

Reliability Data for Piping Components in Nordic Nuclear Power Plants 'R-Book'. Project Phase 1. Rev 1

International Nuclear Information System (INIS)

Lydell, Bengt; Olsson, Anders

2008-01-01

This report constitutes a planning document for a new RandD project to develop a piping component reliability parameter handbook for use in probabilistic safety assessment (PSA) and related activities. The Swedish acronym for this handbook is 'R-Book.' The objective of the project is to utilize the OECD Nuclear Energy Agency 'OECD Pipe Failure Data Exchange Project' (OPDE) database to derive piping component failure rates and rupture probabilities for input to internal flooding probabilistic safety assessment, high-energy line break' (HELB) analysis, risk-informed in-service inspection (RI-ISI) program development, and other activities related to PSA. This new RandD project is funded by member organizations of the Nordic PSA Group (NPSAG) - Forsmark AB, OKG AB, Ringhals AB, and the Swedish Nuclear Power Inspectorate (SKI). The history behind the current effort to produce a handbook of piping reliability parameters goes back to 1994 when SKI funded a 5-year RandD project to explore the viability of establishing an international database on the service experience with piping system components in commercial nuclear power plants. An underlying objective behind this 5-year program was to investigate the different options and possibilities for deriving pipe failure rates and rupture probabilities directly from service experience data as an alternative to probabilistic fracture mechanics. The RandD project culminated in an international piping reliability seminar held in the fall of 1997 in Sigtuna (Sweden) and a pilot project to demonstrate an application of the pipe failure database to the estimation of loss-of-coolant-accident (LOCA) frequency (SKI Report 98:30). A particularly important outcome of the 5-year project was a decision by SKI to transfer the pipe failure database including the lessons learned to an international cooperative effort under the auspices of the OECD Nuclear Energy Agency. Following on information exchange and planning meetings that were