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Sample records for austenitic piping components

  1. LOW-FREQUENCY PHASED-ARRAY METHODS FOR CRACK DETECTION IN CAST AUSTENITIC PIPING COMPONENTS

    International Nuclear Information System (INIS)

    Studies at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, are being conducted to evaluate nondestructive examination (NDE) approaches for inspecting coarse-grained, austenitic stainless steel reactor components. The work provides information to the United States Nuclear Regulatory Commission (NRC) on the utility, effectiveness, limitations, and reliability of advanced inspection techniques for application on safety-related components in commercial nuclear power plants. This paper describes results from recent assessments using a low-frequency phased-array methodology for detecting cracks in cast austenitic piping welds. Piping specimens that contain thermal and mechanical fatigue cracks located adjacent to welds were examined. The specimens have surface geometrical conditions and weld features that simulate portions of primary piping systems in many U.S. pressurized water reactors (PWRs). In addition, segments of vintage centrifugally cast piping were examined to assess inherent acoustic noise and scattering due to grain structures and determine consistency of ultrasonic (UT) responses from varied circumferential locations. The phased-array UT methods were applied from the outside surface of the specimens using automated scanning devices and water coupling, and employed a modified instrument operating between 500 kHz and 1.0 MHz. Composite volumetric images of the specimens were generated. Results from laboratory studies for assessing crack detection and sizing effectiveness are discussed, including acoustic parameters observed in centrifugally cast piping base materials

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

    International Nuclear Information System (INIS)

    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

  3. Improvements in Low-Frequency, Ultrasonic Phased-Array Evaluation for Thick Section Cast Austenitic Stainless Steel Piping Components

    International Nuclear Information System (INIS)

    Research is being conducted for the U.S. Nuclear Regulatory Commission (NRC) at the Pacific Northwest National Laboratory (PNNL) to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light water reactor (LWR) components. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in coarse-grained steel components. This particular study focused on the evaluation of custom-designed, low-frequency (500 kHz) phased-array (PA) probes for examining welds in thick-section cast austenitic stainless steel (CASS) piping. In addition, research was conducted to observe ultrasonic sound field propagation effects from known coarse-grained microstructures found in parent CASS material. The study was conducted on a variety of thick-wall, coarse-grained CASS specimens that were previously inspected by an older generation 500-kHz PA-UT probe and acquisition instrument configuration. This comparative study describes the impact of the new PA probe design on flaw detection and sizing in a low signal-to-noise environment. The set of Pressurized Water Reactor Owners Group (PWROG) CASS specimens examined in this study are greater than 50.8-mm (2.0-in.) thick with documented flaws and microstructures. These specimens are on loan to PNNL from the Electric Power Research Institute (EPRI) NDE Center in Charlotte, North Carolina. The flaws contained within these specimens are thermal fatigue cracks (TFC) or mechanical fatigue cracks (MFC) and range from 13% to 42% in through-wall extent. In addition, ultrasonic signal continuity was evaluated on two CASS parent material ring sections by examining the edge-of-pipe response (corner geometry) for regions of signal loss.

  4. Ultrasonic testing of austenitic and austenitic-ferritic welds made possible by NDT-based optimization of welding techniques, illustrated for some circumferential pipe welds at NPP components, and their qualification

    International Nuclear Information System (INIS)

    The paper summarizes practical results of the making and testing of austenitic and austenitic-ferritic welds in compliance with the KTA codes. The subsequent evaluation is based on mechanized ultrasonic testing for longitudinal flaws in pipe welds made by narrow-gap TIG welding. It is shown that NDT of austenitic as well as austenitic-ferritic root welds produced after optimization of welding techniques can be done with shear wave transducers, and yields satisfactory results. (orig./CB)

  5. 76 FR 43981 - Circular Welded Austenitic Stainless Pressure Pipe From the People's Republic of China: Final...

    Science.gov (United States)

    2011-07-22

    ... International Trade Administration Circular Welded Austenitic Stainless Pressure Pipe From the People's Republic... antidumping duty order on circular welded austenitic stainless pressure pipe from the People's Republic of..., 2010. \\1\\ See Circular Welded Austenitic Stainless Pressure Pipe From the People's Republic of...

  6. Austenitic steel piping testing exercises in PISC

    International Nuclear Information System (INIS)

    In this paper capability and reliability studies of NDT procedures for the inspection of wrought and cast stainless steel piping used in nuclear power plants will be presented. The capability study was designed to identify procedures that have the potential to detect and size defects and to discriminate between flawed and unflawed material. The reliability study was undertaken to quantify on real and realistic flaws in-service inspection performance (detection and false call capability) under realistic field conditions. Furthermore parametric studes were performed to complement the capability and reliability studies by evaluating the effect of important material and flaw variables.The specimens used in these studies were cast-to-cast, cast-to-wrought, and wrought-to-wrought pipework welds. The evaluation methods used to quantify the inspection performance were selected to be as comparable as possible to the PISC II methods. These were adapted to allow also the evaluation of the effect of false calls. During the PISC II screening exercise for the cast-to-cast stainless steel round robin test and other piping round robin studies, it was indeed found that false call probabilities were large and could not be ignored in the evaluation of the inspection performance. The matrix of samples has also been designed to allow the implementation of specific statistical analysis procedures for the evaluation of results such as for example the relative operative characteristics analysis. (orig.)

  7. Mechanized ultrasonic inspection of austenitic pipe systems; Mechanisierte Ultraschallpruefung von austenitischen Rohrleitungen

    Energy Technology Data Exchange (ETDEWEB)

    Dressler, K.; Luecking, J.; Medenbach, S. [ABB ZAQ GmbH, Essen (Germany)

    1999-08-01

    The contribution explains the system of standard testing methods elaborated by ABB ZAQ GmbH for inspection of austenitic plant components. The inspection tasks explained in greater detail are basic materials testing (straight pipes, bends, and pipe specials), and inspection of welds and dissimilar welds. The techniques discussed in detail are those for detection and sizing of defects. (orig./CB) [Deutsch] Das Ziel dieses Beitrages ist die Vorstellung der von der ABB ZAQ GmbH eingesetzten Standardprueftechniken fuer die Pruefung austenitischer Anlagenkomponenten. Im einzelnen wird die Grundwerkstoffpruefung (Rohre, Boegen, Formstuecke), die Schweissnahtpruefung und die Mischnahtpruefung angesprochen. Es werden dabei die Techniken fuer `Detection` und `Sizing` differenziert betrachtet und erlaeutert. (orig.)

  8. Ultrasonic flaw detection of austenitic stainless steel longitudinally welded pipe and tubing

    International Nuclear Information System (INIS)

    Recently there has been the trend that the welded austenitic stainless steel pipe and tubing are used in the nuclear industry in place of the seamless pipe and tubing. For most of the nuclear components, the pipe and tubing must be examined with ultrasonic method by the demands of ASME SA-655. But the ultrasonic flaw detection of the austenitic welds is generally difficult because of scattering and deflection of the acoustic beam due to the coarse grained and elastically anisotropic preferred oriented structure of dendrite. For the thinner welds of pipe and tubing, the attenuation and deflection of the beam do not make the serious problems, however, the deterioration of the signal-to-noise ratio by the coherent structural noise from the welds may still disturb the flaw detection. The longitudinal wave can be employed to suppress the structural noise. This depends upon its longer wave length. In the automatic examination of the pipe and tubing, however, the customary shear wave angle beam technique must be applied to use the skipped beam

  9. 75 FR 70908 - Circular Welded Austenitic Stainless Pressure Pipe From the People's Republic of China: Extension...

    Science.gov (United States)

    2010-11-19

    ... International Trade Administration Circular Welded Austenitic Stainless Pressure Pipe From the People's Republic... of the antidumping duty order on circular welded austenitic stainless pressure pipe from the People's... for Revocation in Part, 75 FR 22107 (April 27, 2010). The period of review (``POR'') is September...

  10. Technical basis for the extension of ASME Code Case N-494 for assessment of austenitic piping

    International Nuclear Information System (INIS)

    In 1990, the ASME Boiler and Pressure Vessel Code for Nuclear Components approved Code Case N-494 as an alternative procedure for evaluating flaws in light water reactor (LWR) ferritic piping. The approach is an alternate to Appendix H of the ASME Code and allows the user to remove some unnecessary conservatism in the existing procedure by allowing the use of pipe specific material properties. The Code case is an implementation of the methodology of the deformation plasticity failure assessment diagram (DPFAD). The key ingredient in the application of DPFAD is that the material stress-strain curve must be in the format of a simple power law hardening stress-strain curve such as the Ramberg-Osgood (R-O) model. Ferritic materials can be accurately fit by the R-O model and, therefore, it was natural to use the DPFAD methodology for the assessment of LWR ferritic piping. An extension of Code Case N-494 to austenitic piping required a modification of the existing DPFAD methodology. The modified DPFAD approach, coined piecewise failure assessment diagram (PWFAD), extended an approximate engineering approach proposed by Ainsworth in order to consider materials whose stress-strain behavior cannot be fit to the R-O model. The Code Case N-494 approach was revised using the PWFAD procedure in the same manner as in the development of the original N-494 approach for ferritic materials

  11. Evaluation of thermal stratification and primary water environment effects on fatigue life of austenitic piping

    International Nuclear Information System (INIS)

    During the last two decades, lots of efforts have been devoted to resolve thermal stratification phenomenon and primary water environment issues. While several effective methods were proposed especially in related to thermally stratified flow analyses and corrosive material resistance experiments, however, lack of details on specific stress and fatigue evaluation make it difficult to quantify structural behaviors. In the present work, effects of the thermal stratification and primary water are numerically examined from a structural integrity point of view. First, a representative austenitic nuclear piping is selected and its stress components at critical locations are calculated in use of four stratified temperature inputs and eight transient conditions. Subsequently, both metal and environmental fatigue usage factors of the piping are determined by manipulating the stress components in accordance with NUREG/CR-5704 as well as ASME B and PV Codes. Key findings from the fatigue evaluation with applicability of pipe and three-dimensional solid finite elements are fully discussed and a recommendation for realistic evaluation is suggested

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 (KRMS) 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 (KRMS) evaluated using second scheme gives better fatigue crack growth life prediction compared to that of first scheme. (author)

  15. A simplified LBB evaluation procedure for austenitic and ferritic steel piping

    International Nuclear Information System (INIS)

    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

  16. 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.

  17. Stress corrosion cracking susceptibility of various austenitic stainless steel pipe welds in high temperature oxygenated water

    International Nuclear Information System (INIS)

    Stress corrosion cracking (SCC) susceptibility of various austenitic stainless steel pipe welds has been studied by means of constant load tensile tests and pipe tests in 2880C water containing 26 ppm dissolved oxygen. The results obtained are summarized as follows: (1) SCC susceptibility of SUS 304 pipe welds is comparatively low under the condition of as-welded. It becomes, however, high remarkably by grinder operation and/or low temperature sensitization heat treatment. The distribution of time of failure on SUS 304 pipe welds can be expressed as a log-normal or Weibull distribution. (2) SUS 304L, 304NG, 316NG, and 347 stainless steel pipe welds have a good SCC resistance and sensitization resistance. Furthermore, the life estimation on alternate pipe welds was conducted statistically. (author)

  18. Application of advanced austenitic alloys to fossil power system components

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.

    1996-06-01

    Most power and recovery boilers operating in the US produce steam at temperatures below 565{degrees}C (1050{degrees}F) and pressures below 24 MPa (3500 psi). For these operating conditions, carbon steels and low alloy steels may be used for the construction of most of the boiler components. Austenitic stainless steels often are used for superheater/reheater tubing when these components are expected to experience temperatures above 565{degrees}C (1050{degrees}F) or when the environment is too corrosive for low alloys steels. The austenitic stainless steels typically used are the 304H, 321H, and 347H grades. New ferritic steels such as T91 and T92 are now being introduced to replace austenitic: stainless steels in aging fossil power plants. Generally, these high-strength ferritic steels are more expensive to fabricate than austenitic stainless steels because the ferritic steels have more stringent heat treating requirements. Now, annealing requirements are being considered for the stabilized grades of austenitic stainless steels when they receive more than 5% cold work, and these requirements would increase significantly the cost of fabrication of boiler components where bending strains often exceed 15%. It has been shown, however, that advanced stainless steels developed at ORNL greatly benefit from cold work, and these steels could provide an alternative to either conventional stainless steels or high-strength ferritic steels. The purpose of the activities reported here is to examine the potential of advanced stainless steels for construction of tubular components in power boilers. The work is being carried out with collaboration of a commercial boiler manufacturer.

  19. Residual Stresses Due to Circumferential Girth Welding of Austenitic Stainless Steel Pipes

    Science.gov (United States)

    Tarak, Farzan

    Welding, as a joining method in fabrication of engineering products and structural elements, has a direct influence on thermo-mechanical behavior of components in numerous structural applications. Since these thermo-mechanical behaviors have a major role in the life of welding components, predicting thermo-mechanical effects of welding is a major factor in designing of welding components. One of the major of these effects is generation of residual stresses due to welding. These residual stresses are not the causes of failure in the components solely, but they will add to external loads and stresses in operating time. Since, experimental methods are time consuming and expensive, computational simulation of welding process is an effective method to calculate these residual stresses. This investigation focuses on the evaluation of residual stresses and distortions due to circumferential girth welding of austenitic stainless steel pipes using the commercial finite element software ESI Visual-Environment and SYSWELDRTM to simulate welding process. Of particular importance is the comparison of results from three different types of mechanics models: 1) Axisymmetric, 2) Shell, and 3) Full 3-D.

  20. Evaluation of residual stress distribution in austenitic stainless steel pipe butt-welded joint

    International Nuclear Information System (INIS)

    This paper reports measured and estimated results of residual stress distributions of butt-welded austenitic stainless steel pipe in order to improve estimation accuracy of welding residual stress. Neutron diffraction and strain gauge method were employed for the measurement of the welding residual stress and its detailed distributions on inner and outer surface of the pipe as well as the distributions within the pipe wall were obtained. Finite element method was employed for the estimation. Transient and residual stresses in 3D butt-welded joint model were computed by employing Iterative Substructure Method and also commercial FEM code ABAQUS for a reference. The measured and estimated distributions presented typical characteristic of straight butt-welded pipe which had decreasing trend along the axial direction and bending type distributions through wall of the pipe. Both results were compared and the accuracy of measurement and estimation was discussed. (author)

  1. Fatigue crack growth in austenitic stainless steel piping

    International Nuclear Information System (INIS)

    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

  2. Heat sink welding of austenitic stainless steel pipes to control distortion and residual stress

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, H.; Albert, S.K.; Bhaduri, A.K. [Materials Technology Div., Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2007-07-01

    Construction of India's Prototype Fast Breeder Reactor (PFBR) involves extensive welding of austenitic stainless steels pipes of different dimensions. Due to high thermal expansion coefficient and poor thermal conductivity of this class of steels, welding can result in significant distortion of these pipes. Attempts to arrest this distortion can lead to high levels of residual stresses in the welded parts. Heat sink welding is one of the techniques often employed to minimize distortion and residual stress in austenitic stainless steel pipe welding. This technique has also been employed to repair welding of the piping of the Boiling Water Reactors (BWRs) subjected to radiation induced intergranular stress corrosion cracking (IGSCC). In the present study, a comparison of the distortion in two pipe welds, one made with heat sink welding and another a normal welds. Pipes of dimensions 350{phi} x 250(L) x 8(t) mm was fabricated from 316LN plates of dimensions 1100 x 250 x 8 mm by bending and long seam (L-seam) welding by SMAW process. Two fit ups with a root gap of 2 mm, land height of 1mm and a groove angle of 70 were prepared using these pipes for circumferential seam (C-seam) welding. Dimensions at predetermined points in the fit up were made before and after welding to check the variation in radius, circumference and and ovality of the pipes. Root pass for both the pipe fit up were carried out using conventional GTAW process with 1.6 mm AWS ER 16-8-2 as consumables. Welding of one of the pipe fit ups were completed using conventions GTAW process while the other was completed using heat sink welding. For second and subsequent layers of welding using this process, water was sprayed at the root side of the joint while welding was in progress. Flow rate of the water was {proportional_to}6 1/minute. Welding parameters employed were same as those used for the other pipe weld. Results of the dimensional measurements showed that there is no circumferential shrinkage in

  3. Eddy current testing of longitudinal welds in austenitic steel pipes

    International Nuclear Information System (INIS)

    The existing steel-iron test sheet 1914 'Nondestructive Testing of Fusion Welded Joints in Tubes of Stainless Steels' (SEP 1914) had to be revised. The physical correlation between test frequency, phase position, amplitude and defect size for different pipe dimensions were pointed out and discussed on the basis of numerous measurement series. The influence of coil-specific parameters, filtration and electromagnetic magnetization on the eddy current signal were clarified. Additionally, the transferability of the amplitude and phase behaviour of simulated defects on naturally occurring defects was investigated in order to guarantee that the determined test parameters for the revision of existing test regulations are practice-oriented. As the most important result, it is stated that the definition of the test sensitivity in the steel-iron test sheet 1914 is too insensitive for a great number of pipe dimensions. Therefore, the existing steel-iron test sheet 1914 was revised with regard to the definition of the test sensitivity on the basis of the required detection sensitivity of a 20% internal groove. (orig./HP)

  4. Crack resistance of austenitic pipes with circumferential through-wall cracks

    International Nuclear Information System (INIS)

    For monotonously increasing load the correct evaluation of the crack resistance properties of a structure is essential for safety analyses. Considerable attention has been given to the through-wall case, since this is generally believed to be the controlling case with regard to complete pipe failure. The maximum load conditions for circumferential crack growth in pipes under displacement-controlled loadings has been determined. The need for crack resistance curves, measured on circumferentially through-wall cracked straight pipes of austenitic stainless steel 316L under bending, is emphasized by the limitation in the data range on small specimens and by the differences in the procedures. To answer open questions and to improve calculational methods a joint fracture mechanics program is being performed by Electricite de France, Novatome and Siemens-Interatom. The working program contains experimental and theoretical investigations on the applicability of small-specimen data to real structures. 10 refs., 10 figs., 4 tabs

  5. Reliability examination of the radiography of austenitic pipe welds

    International Nuclear Information System (INIS)

    The radiographic examination of pipe welds with stress corrosion cracks is a sophisticated testing task. Enhanced reliability is achieved primarily through better seam preparation, welding technique, and post-welding treatment. The statistical methods ROC and POD are very suitable for quantitative measurement of the reliability of NDE, with POD requiring a broader statistical basis. Digital image evaluation requires object-related filtering (longitudinal cracks) or profile generation in order to represent the information describing the crack as an optical optimization. Training and on-screen training is required, and testing experts ought to have good experience with defect types possible. Experienced evaluators will be able to detect with high reliability cracks located half through wall. (orig./MM)

  6. Seismic analysis and testing of piping components

    International Nuclear Information System (INIS)

    There is general concern that overconservatism in the treatment of infrequent dynamic loads in the design of nuclear power plant piping may compromise safety when considering the effects of frequent thermal loading. To address these concerns, the 1985, the Electric Power Research Institute (EPRI), in conjunction with the U.S. Nuclear Regulatory Commission (NRC), initiated the Piping and Fitting Dynamic Reliability (PFDR) Program. The ultimate objective of this program is to introduce new, improved, realistic and defensible ASME Code design rules which take advantage of the inherent dynamic margin in piping. The basis for the proposed rule changes will be an extensive testing program, together with supporting dynamic analyses, which is focused on behavior of typical piping components when subjected to dynamic loading introduced through hydraulically operated sleds. Seismic time history inputs, typical of those occurring at actual nuclear power plants but scaled up in amplitude, have been applied to the piping components. The amplitude of input has varied from 15 to 25 times the input required to achieve ASME Code Level D Service Limits using linear response spectrum analysis with 2% damping. At these level inputs, cyclic peak-to-peak strains of 3.6%, equivalent damping of 34% and dynamic moments of twice the static limit moment are developed in the components. Cumulative strains as high as 32% have been observed for components with high mean stress (e.g. σ/sub hoop/ = 1.0S/sub m/). The through-wall cracks that developed as a result of fatigue ratcheting were detected and their growth monitored before leakage occurred

  7. Investigation of Residual Stress Distributions of Induction Heating Bended Austenitic Stainless Steel (316 Series) Piping

    International Nuclear Information System (INIS)

    The induction heating bending process, which has been recently applied to nuclear piping, can generate residual stresses due to thermomechanical mechanism during the process. This residual stress is one of the crack driving forces that have important effects on crack initiation and propagation. However, previous studies have focused only on geometric shape variations such as the change in thickness and ovality. Moreover, very few studies are available on the effects of process variables on residual stresses. This study investigated the effects of process variables on the residual stress distributions of induction heating bended austenitic stainless steel (316 series) piping using parametric finite element analysis. The results indicated that the heat generation rate and feed velocity have significant effects on the residual stresses whereas the moment and bending angle have insignificant effects

  8. Comparison between shear wave ultrasonic examination and radiographic examination of some austenitic stainless steel pipe welds

    International Nuclear Information System (INIS)

    This paper compares the results of the ultrasonic and radiographic examinations of five austenitic pipe to pipe welds, which contained different types of intentional weld defects. Both the ultrasonic and radiographic examinations were made under laboratory conditions, and thus the results cannot directly be generalized to workshop examinations. The ultrasonic examination was performed using a conventional shear wave angle beam technique due to the geometric conditions of the welds. Longitudinal wave angle beam probes were not used in this work. Comparison between the results of the ultrasonic and radiographic examinations leads to the conclusion that neither method gives quite satisfactory results. Both methods missed some defects and the correlation between the results of the methods was not very good. In the ultrasonic examination the most difficult problem was the evaluation of the defects. (author)

  9. Survey report on unmanned site welding of austenitic stainless steel pipe and its ultrasonic examination

    International Nuclear Information System (INIS)

    In the field welding of austenitic stainless steel pipings and its non-destructive test in complicated and narrow places, reliable welding method and non-destructive testing method are required, and also, it is desirable to mechanize them (unmanned operation). In this study, the present state of the automatic welding of austenitic stainless steel pipings and ultrasonic flaw detection was investigated through the literatures in Japan and foreign countries. As the result, it was clarified that energetic research has been made recently to mechanize the welding, and though many points are left for future research and development, it is promising. In the ultrasonic flaw detection, many technical problems concerning the detectability of flaws remain at present, but is is expected to become feasible by future systematic research and development. The design of weld joints, the welding method and the remote automatic control of welding operation must be appropriate for guaranteeing the quality of welding, and these points were surveyed. The problems in the ultrasonic flaw detection are the attenuation of ultrasonic waves, the conditions of probes, the mode of wave motion and frequency, and the welding suitable to the ultrasonic flaw detection. (Kako, I.)

  10. Improvement of austenitic pipings with welds-corrosion resistance, non-destructive testing and loadability

    International Nuclear Information System (INIS)

    Pipes made from austenitic materials are employed extensively in power plant technology and in the chemical industry. An important aspect is the integrity of the piping, proof of which is quite complex especially due to connecting welds. In the case of an inappropriate application of energy when welding, precipitations develop. For example, M23C6 type precipitations, which weaken the corrosion resistance considerably and give rise to residual stresses which interfere with the load stresses. Simultaneously wrinkles and notches can occur in the area of the root of the weld seam, which in general is medium contacting, and this increases stress and corrosion susceptibility and reduces the loadability as well as the critical crack size. This is especially disadvantageous, as due to the development of the microstructure with non-destructive testing measures, it is very unlikely that small cracks can be detected. In the contribution these influencing variables are covered quantitatively and the optimization possibilities are shown using experimental and numerical simulations. On real pipes of the dimension DN 200 the success of the measures undertaken dependant on the crack sizes, especially the loadability, is verified experimentally. (author)

  11. Measurement of residual stresses in weld overlay pipe component

    International Nuclear Information System (INIS)

    Weld overlay is one of the improvement methods for austenitic stainless steel suffered intergranular stress corrosion cracking (IGSCC). The weld overlay itself increases the wall thickness and change the stress field from inner to outer surface of pipes. To insure the improvement of the stress field on the pipe after weld overlay, the stress distribution of 10 inches pipes weld overlay mock up specimens were insured in this study. The strain-gage hole-drilling method and FEM analysis were applied to measure and calculate the residual stress along the pipe wall thickness. The results show that it behaves good agreement on the residual stress from the measurement and calculation. The weld overlay also induces a compressive stress field on the inner wall, whereas the tensile stress field is created on the outer wall surface. The mechanical grinding on the weld overlay surface will affect the residual stress distribution about 0.04 inches in depth. (author)

  12. Report on Japanese performance demonstration examinations for depth sizing of SCC in austenitic stainless steel pipes

    International Nuclear Information System (INIS)

    The PD Center of Central Research Institute of Electric Power Industry (CRIEPI) commenced Performance Demonstration examinations for flaw depth sizing of austenitic stainless steel pipes in March 2006. As of January 2013, 37 examination sessions have been completed and 44 candidates have passed the examination. The total number of tests administered including re-tests and re-certification was 89. It was noted that depth sizing using a phased array plus the manual UT technique was the procedure most used by successful PD applicants. A major reason for failure is 'the overestimation caused by the lack of skill in distinguishing the base metal-to-weld metal interface echo and the SCC tip echo'. Moreover, the possibility of psychological bias is also incontrovertible because candidates are under pressure not to make a -4.4 mm critical mistake. (author)

  13. Crack growth of intergranular stress corrosion cracks in austenitic stainless steel pipes of boiling water reactors

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking (IGSCC) of boiling water reactor (BWR) piping is considered from the crack growth rate point of view. Crack growth rate of sensitized austenitic stainless steel welds is dependent on the degree of sensitization of the material and the severity of the environment as well as the stress state. In evaluation of actual crack growth rate there are three major sources of uncertainty: knowledge of actual crack size and shape, actual stress distribution in he area of the crack and the degree of sensitization. In the report the crack growth calculations used in the USA and in Sweden are presented. Finally, the crack growth rate predictions based on mechanistic modelling of IGSCC and some needs of further research in Finland are considered

  14. Experimental and numerical assessment of thermal fatigue in 316 austenitic steel pipes

    OpenAIRE

    PAFFUMI Elena; Nilsson, Karl-Fredrik; Szaraz, Zoltan

    2013-01-01

    This paper presents an experimental and numerical investigation of thermal fatigue of 316L steel pipe components with 14 mm wall thickness heated by induction to 300–550 °C on the outer surface and cyclically cooled internally with room temperature water. The damage is initiated as network of surface cracks where some cracks become dominant. At 550 °C the pipe fails after typically 50,000 cycles whereas at 300 °C the deepest cracks have only penetrated half the thickness after 250,000 cycles....

  15. Reliability of piping system components. Volume 4: The pipe failure event database

    International Nuclear Information System (INIS)

    Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A 'data driven and systems oriented' analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failure. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today's PSAs to allow for aging analysis and effective, on-line risk management. 42 refs, 25 figs

  16. Comprehensive residual stress distributions in a range of plate and pipe components

    International Nuclear Information System (INIS)

    A comprehensive review of through thickness transverse residual stress distributions in a range of as-welded and mechanically bent components made up of a range of steels has been carried out, and simplified generic transverse residual stress profiles for a plate and pipe components have been proposed. The geometries consisted of welded pipe butt joints, T-plate joints, tubular T-joints, tubular Y-joints and a pipe on plate joints as well as cold bent tubes and pipes. The collected data covered a range of engineering steels including ferritic, austenitic, C-Mn and Cr-Mo steels. Measured residual stress data, normalised with respect to the parent material yield stress, has shown a good linear correlation versus the normalised depth of the region containing the residual stress resulting from the welding or cold-bending process. The proposed simplified generic residual stress profiles based on the mean statistical linear fit of all the data provides a reasonably conservative prediction of the stress intensity factors. Whereas the profiles for the assessment procedures are fixed and case specific, the simple bilinear profiles for the residual stresses obtained by shifting the mean and bending stress from the mean regression line have been proposed and validated

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

    International Nuclear Information System (INIS)

    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)

  18. Dynamic tests of cracked pipe components

    International Nuclear Information System (INIS)

    Dynamic tests were conducted involving notched sections of 4-in. (10-cm) stainless steel and Inconel-600 pipe. The specimen was a four-point bending beam with end masses sized to give an elastic first-mode frequency near that of typical field installed piping systems (15 Hz). Specimens were loaded using sinewave excitation at this first mode natural frequency. Specimen response was compared to predictions from an elastic-plastic dynamic analysis previously developed on this program. In addition, specimen loads at failure were compared to those predicted from a net section collapse failure criterion. The results confirmed that the elasticplastic dynamic analysis adequately predicted the dynamic response of flawed pipes under seismic-type excitation. Furthermore, net section collapse does not occur under dynamic loading conditions which simulate natural frequencies of asinstalled light water reactor piping systems. Finally, a net section collapse criterion yields conservative estimates of the load capacity of flawed pipe sections provided crack growth is properly accounted for

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

    International Nuclear Information System (INIS)

    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.

  20. Stress corrosion on austenitic stainless steels components after sodium draining

    International Nuclear Information System (INIS)

    The damage study performed on 316 pipes of a loop after two leakages allows to conclude that a stress corrosion process in sodium hydroxide environment has induced trans-crystaline cracks. The research of conditions inducing such a phenomenon is developed, including parametric tests under uniaxial load and some tests on pipe with welded joints. In aqueous sodium hydroxide, two corrosion processes have been revealed: a general oxidization increasing with environment aeration and a transcrystalline cracking appearing for stresses of the order of yield strength. Other conditions such a temperature (upper than 1000C) and time exposures (some tens of hours) are necessary. Cautions in order to limit introduction of wet air into drained loop and a choice of appropriate preheating conditions when restarting the installation must permit to avoid such a type of incident

  1. Reliability of piping system components. Volume 1: Piping reliability - A resource document for PSA applications

    International Nuclear Information System (INIS)

    SKI has undertaken a multi-year research project to establish a comprehensive passive component failure database, validate failure rate parameter estimates and establish a model framework for integrating passive component failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure events in the nuclear and chemical industries. This phase 2 report gives a graphical presentation of piping system operating experience, and compares key failure mechanisms in commercial nuclear power plants and chemical process industry. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A data-driven-and-systems-oriented analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failures. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today's PSAs to allow for aging analysis and effective, on-line risk management. 111 refs, 36 figs, 20 tabs

  2. ASME code and ratcheting in piping components. Final technical report

    International Nuclear Information System (INIS)

    The main objective of this research is to develop an analysis program which can accurately simulate ratcheting in piping components subjected to seismic or other cyclic loads. Ratcheting is defined as the accumulation of deformation in structures and materials with cycles. This phenomenon has been demonstrated to cause failure to piping components (known as ratcheting-fatigue failure) and is yet to be understood clearly. The design and analysis methods in the ASME Boiler and Pressure Vessel Code for ratcheting of piping components are not well accepted by the practicing engineering community. This research project attempts to understand the ratcheting-fatigue failure mechanisms and improve analysis methods for ratcheting predictions. In the first step a state-of-the-art testing facility is developed for quasi-static cyclic and seismic testing of straight and elbow piping components. A systematic testing program to study ratcheting is developed. Some tests have already been performed and the rest will be completed by summer'99. Significant progress has been made in the area of constitutive modeling. A number of sophisticated constitutive models have been evaluated in terms of their simulations for a broad class of ratcheting responses. From the knowledge gained from this evaluation study two improved models are developed. These models are demonstrated to have promise in simulating ratcheting responses in piping components. Hence, implementation of these improved models in widely used finite element programs, ANSYS and/or ABAQUS, is in progress. Upon achieving improved finite element programs for simulation of ratcheting, the ASME Code provisions for ratcheting of piping components will be reviewed and more rational methods will be suggested. Also, simplified analysis methods will be developed for operability studies of piping components and systems. Some of the future works will be performed under the auspices of the Center for Nuclear Power Plant Structures

  3. Numeric and experimental studies on surface cracking of disks and pipes of ferritic and austenitic steels

    International Nuclear Information System (INIS)

    For safeguarding the transmission chain from the two-dimensional sample to real three-dimensional structural components, the behavior of surface cracks in disks and pipes was investigated in an experimental and a theoretic-numerical way. The objective was a quantitative failure analysis which covers essential phases resulting from operation and breakdown conditions, in fact from the first propagation of the incipient crack under fatigue loading to stabile crack growth under monotonously increasing load up to wall breakthrough and a possible subsequent instability. The investigations of the growth of fatigue cracks show that the growth of partial through-cracks in disks and pipes is generally overestimated in analytic calculations, if the constants of the crack propagation determined for CT samples are used for these calculations. The results presented here show that the J-integral concept is suited to give a good description of the behavior of cracks in constructional components under shear fracture conditions as to the quality and quantity, if the influence of the multi-axiality of the state of stress on the resistance against crack propagation is appropriately considered. The transferability of samples to constructional components seems to be possible for purely mechanical strain even with a superposed amount of bending for a crack propagation which is not too great. (orig.)

  4. Cracking in stabilized austenitic stainless steel piping of German boiling water reactors - characteristic features and root cause

    International Nuclear Information System (INIS)

    Cracks have been found in the welds of piping systems made from stabilized austenitic stainless steels in German boiling water reactors (BWR). In the course of the intensive failure analysis metallographic examinations, microstructural investigations by electron microscopy, corrosion experiments and welding tests have been performed. The results show that cracking under the given medium conditions is due to intergranular stress corrosion cracking (IGSCC) in those parts of the heat affected zone (HAZ) which are overheated during welding and where solution of titanium carbides and subsequent precipitation of chromium carbides and depletion of chromium along the affected grain boundaries could occur. (orig.)

  5. Technical Letter Report Assessment of Ultrasonic Phased Array Inspection Method for Welds in Cast Austenitic Stainless Steel Pressurizer Surge Line Piping JCN N6398, Task 1B

    Energy Technology Data Exchange (ETDEWEB)

    Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Mathews, Royce; Moran, Traci L.; Anderson, Michael T.

    2009-07-28

    Research is being conducted for the U.S. Nuclear Regulatory Commission (NRC) at the Pacific Northwest National Laboratory (PNNL) to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light water reactor components. The scope of this research encompasses primary system pressure boundary materials including cast austenitic stainless steels (CASS); dissimilar metal welds; piping with corrosion-resistant cladding; weld overlays, inlays and onlays; and far-side examinations of austenitic piping welds. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in coarse-grained steel components. In this effort, PNNL supports cooperation with Commissariat à l’Energie Atomique (CEA) to assess reliable inspection of CASS materials. The NRC Project Manager has established a cooperative effort with the Institut de Radioprotection et de Surete Nucleaire (IRSN). CEA, under funding from IRSN, are supporting collaborative efforts with the NRC and PNNL. Regarding its work on the NDE of materials, CEA is providing its modeling software (CIVA) in exchange for PNNL offering expertise and data related to phased-array detection and sizing, acoustic attenuation, and back scattering on CASS materials. This collaboration benefits the NRC because CEA performs research and development on CASS for Électricité de France (EdF). This technical letter report provides a summary of a technical evaluation aimed at assessing the capabilities of phased-array (PA) ultrasonic testing (UT) methods as applied to the inspection of welds in CASS pressurizer (PZR) surge line nuclear reactor piping. A set of thermal fatigue cracks (TFCs) was implanted into three CASS PZR surge-line specimens (pipe-to-elbow welds) that were fabricated using vintage CASS materials formed in the 1970s, and flaw responses from these cracks were used to evaluate detection and sizing

  6. Case study of qualification of remote automated ultrasonic inspection of austenitic pipe weld

    International Nuclear Information System (INIS)

    This paper describes a recent specific example of how Nuclear Electric has demonstrated the capability of a remote automated ultrasonic inspection on its Advanced Gas Cooled Reactor plant. Following the discovery of a small leak in a 300 mm diameter austenitic steam tube there was an urgent requirement to develop a capability for inspecting all similar welds and to provide a demonstration of capability of the inspection. The first part of the paper describes the inspection requirements, the ultrasonic techniques used and the procedure and equipment which was developed. The second part describes how the capability was demonstrated: all the evidence to support the capability was embodied in a technical justification which was prepared under contract by the AEA Technology Inspection Validation Centre (WC). The technical justification included a combination of theoretical evidence and practical evidence from open and blind test block trials together with experience of inspection of other similar components. Within a period of less than two months from initial request, an inspection capability had been developed and justified and was then successfully deployed on-site

  7. A repair process for an heterogenous welded joint between a nuclear reactor component tube and a pipe

    International Nuclear Information System (INIS)

    The repairing process involves cutting a tubular section of the tube and the pipe, which includes the welded joint, and preparing an austenitic stainless steel tubular section for substitution; the section is then narrow-joint welded with the low-alloy steel tube, and finally welded to the austenitic stainless steel pipe. Application to repairing a welded joint between a PWR pressurizer tube and the expansion pipe of the pressurizer. (authors). 7 refs., 3 figs

  8. Prevision of in-service aging of molded austenitic-ferritic stainless steels components

    International Nuclear Information System (INIS)

    After having recalled the service conditions of the nuclear PWR boilers, the austenitic-ferritic molded stainless steels and their uses in the primary coolant circuit are described. The main consequences of the thermal aging on the rupture mechanisms and the mechanical properties are recalled too. Then are described the laboratory studies carried out in France and abroad which have allowed the development of an extensive knowledge of the aging reaction kinetics and then of embrittlement anticipation formulae. Measures and sampling carried out on down-rated components or even on in service components are used to verify the quality of the in-service aging anticipation. At last are identified the subjects on which it will be important to advance to improve our knowledge of the behaviour of the austenitic-ferritic stainless steels components. (O.M.)

  9. Numerical simulation of residual stress in piping components at Framatome-ANP

    International Nuclear Information System (INIS)

    Numerous manufacturing processes induce residual stresses and distortions in piping components and associated welds: quenching of cast pipings, machining and welding. In Pressurized Water Reactors, most of the components have a large thickness for sustaining pressure and distortions are a minor source of concern. This is not the case for residual stresses which may have a strong influence on several type of damage such as fatigue, corrosion, brittle fracture. In low toughness components, residual stress fields may contribute to ductile tearing initiation. These potential damages are mitigated after welding by stress relief heat treatment, which is applied in a systematic manner to ferritic components of the primary system in nuclear reactors. This treatment is not applied on austenitic piping for which the heat treatment temperature is limited due to the risk of sensitization and residual stresses are difficult to eliminate completely. Since on site measurements are costly and difficult to perform, numerical simulation appears to be an attractive tool for estimating residual stress distributions. Framatome-ANP is working on modelling manufacturing processes with that purpose in mind. This paper presents three kinds of applications illustrating efforts on welding, quenching and machining simulation. First a comparison is shown between computations and measurements of residual stress induced by welding of a dissimilar weld metal junction. Then numerical simulations of quenching of a cast stainless steel nozzle are presented. Finally quenching followed by machining and grinding of this cast component are considered in a full simulation of the manufacturing process. Computed distortions and residual stresses are compared with experimental measurements at different stages of the manufacturing process. (authors)

  10. Reactor Materials Program -- weldment component toughness of SRS PWS piping materials. [Process Water System

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.

    1993-02-01

    The mechanical properties of austenitic stainless steel materials from the reactor systems in the unirradiated (baseline) and the irradiated conditions have been developed previously for structural and fracture analyses of the pressure boundary of the SRS reactor Process Water System (PWS) components. Individual mechanical specimen test results were compiled into three separate weldment components or regions, namely, the base, weld, and weld heat-affected-zone (HAZ), for two orientations (L-C and C-L) with respect to the pipe axis of the source materials and for two test temperatures of 25 and 125[degrees]C. Twelve separate categories were thus defined to assess the effect of test conditions on the mechanical properties and to facilitate selection of properties for structural and fracture analyses. The testing results show high fracture toughness of the materials and support the demonstration of PWS pressure boundary structural integrity under all conditions of reactor operation. The fracture toughness of a fourth weldment component, namely, the weld fusion line region, has been measured to evaluate the potential for a region of low toughness in the interface between the Type 308 stainless steel weld metal and the Type 304 stainless steel pipe. The testing details and results of the weld fusion line toughness are contained in this report.

  11. Leak before break behaviour of austenitic and ferritic pipes containing circumferential defects

    Energy Technology Data Exchange (ETDEWEB)

    Stadtmueller, W.; Sturm, D.

    1997-04-01

    Several research projects carried out at MPA Stuttgart to investigate the Leak-before-Break (LBB) behavior of safety relevant pressure bearing components are summarized. Results presented relate to pipes containing circumferential defects subjected to internal pressure and external bending loading. An overview of the experimentally determined results for ferritic components is presented. For components containing postulated or actual defects, the dependence of the critical loading limit on the defect size is shown in the form of LBB curves. These are determined experimentally and/or by calculation for through-wall slits, and represent the boundary curve between leakage and massive fracture. For surface defects and a given bending moment and internal pressure, no fracture will occur if the length at leakage remains smaller than the critical defect length given by the LBB curve for through-wall defects. The predictive capability of engineering calculational methods are presented by way of example. The investigation programs currently underway, testing techniques, and initial results are outlined.

  12. Technique for ultrasonic testing of austenitic steel weldments of NPP components

    International Nuclear Information System (INIS)

    Special literature on ultrasonic testing of weldments of austenitic steel is analysed. Technique for ultrasonic testing of the ring and longitudinal butt welded joints of NPP components without reinforcing bead removal is described. Special converter design and fabrication practice are described. Results of experimental check of the developed testing technology and its application during NNPs' mounting and operation are presented. Results of ultrasonic and X-ray testing are compared

  13. Detection of material degradation and imperfections of thickwalled austenitic components with electromagnetic testing

    International Nuclear Information System (INIS)

    The C-Scan technique is applied on: 1.) in eddy current tests of tensile specimens with plastic deformations in order to identify martensitic structural changes; 2.) in eddy current tests of so-called 'hourglass' test pieces with low-frequency component loads in low cycle fatigue tests; 3.) in the development of far field eddy current tests for crack detection in the basic material of austenitic pipelines

  14. A repair process for an heterogenous welded joint between a nuclear reactor component tube and a pipe

    International Nuclear Information System (INIS)

    The repairing process involves cutting a tubular section of the tube (made of low alloy steel) and the pipe (made of austenitic stainless steel), which includes the welded joint, and preparing an heterogenous tubular section for substitution (a first section, made of ferritic steel, is butt welded to a second section, made of austenitic stainless steel); the tubular section is then narrow-joint welded with the low-alloy steel tube, and finally welded to the austenitic stainless steel pipe. Application to repairing a welded joint between a pressurizer tube and an expansion pipe connected to the primary circuit. (author). 5 refs., 4 figs

  15. Residual stress and microstructure evolution by manufacturing processes for welded pipe joint in austenitic stainless steel type 316L

    International Nuclear Information System (INIS)

    Stress corrosion cracking (SCC) has been observed near the heat affected zone (HAZ) of welded pipe joint made of austenitic stainless steel type 316L, even though sensitization is not observed. Therefore, It can be considered that the effect of residual stress on SCC is more important. In the joining process of pipes, butt-welding is conducted after machining. Residual stress is generated by both processes. In case of welding after machining, it can be considered that residual stress due to machining is changed by welding thermal cycle. In this study, residual stress and microstructure evolution due to manufacturing processes is investigated. Change of residual stress distribution caused by processing history is examined by X-ray diffraction method. Residual stress distribution has a local maximum stress in the middle temperature range of the HAZ caused by processing history. Hardness measurement result also has a local maximum hardness in the same range of the HAZ. By using FE-SEM/EBSD, it is clarified that microstructure shows recovery in the high temperature range of HAZ. Therefore, residual stress distribution is determined by microstructure evolution and superposition effect of processing history. In summary, not only any part of manufacturing processes such as welding or machining but also treating all processes as processing history of pipes are important to evaluate SCC. (author)

  16. Three dimensional analysis of piping components using BARC finite element based damage mechanics code MADAM

    International Nuclear Information System (INIS)

    This work has been carried out at State Institute for Material Testing (MPA), University of Stuttgart, Germany as part of the research project named Transferabililty of specimen data to component level under Indo-German Bilateral project (IND-98/329) during the period 5 th August, 2000 to 30 th December, 2000. In this project, we have used Gurson-Tvergaard-Needleman's model for predicting the fracture behaviour of real life pipes and elbows made of two different materials (one German austenitic steel and other Indian ferritic steel). The inhouse damage mechanics MADAM has been used for all the calculations. The results have been compared with the experimental results in order to establish the method and the Gurson parameters. The Gurson parameters have been determined by a hybrid methodoly of metallographic analysis, numerical analysis of notched tensile tests and compact tension (C(T)) tests and by comparison with experimental results. Analysis has also be done for determining the multiaxiality parameter q existing in the crack plane of these components for both stationary crack and running crack. The parameter q has been studied for transferability of J-R curve from specimen to component level. The Gurson parameters have then been used to analyse a straight pipe with 122 deg circumferential throughwall crack under internal pressure of 16 Mpa and increasing bending moment for the German steel. For SA333 Gr.6 steel, the components tested are straight pipes and elbows with throughwall circumferential cracks of different crack angles under four point bending load. This report has been divided into three sections. Section-I deals with numerical analysis of ductile fracture for the German austenitic steel, i.e., DIN X6CrNiNb 18 10. Section-II deals with numerical analysis of ductile fracture for the Indian PHT material, i.e., SA333 Gr.6 carbon steel. Section-III deals with evaluation of stress multiaxiality quotient q for all the cracked geometries of importance at

  17. Mechanized radiation testing of austenitic pipe welds. Testing of media filled pipes and determination of the flaw depth by tomosynthesis

    International Nuclear Information System (INIS)

    A compact detection system was built for multi-angle inspection of pipes, consisting of a high-sensitivity radiometric line scanner and an ultrasonic manipulator. Improved flaw imaging quality is achieved with this system as compared to film radiography. Measurements have been carried out on site in a nuclear power plant and in a laboratory. Better flaw imaging quality was also achieved in the testing of water-filled pipes. Non-linear tomosynthesis was applied for processing and interpretation of measured data. The system delivers considerably better images of planary materials inhomogeneitites, (such as cracks and lack-of-bond defects). (orig./CB)

  18. Seismic design of ITER component cooling water system-1 piping

    International Nuclear Information System (INIS)

    The successful performance of ITER machine very much depends upon the effective removal of heat from the in-vessel components and other auxiliary systems during Tokamak operation. This objective will be accomplished by the design of an effective Cooling Water System (CWS). The optimized piping layout design is an important element in CWS design and is one of the major design challenges owing to the factors of large thermal expansion and seismic accelerations; considering safety, accessibility and maintainability aspects. An important sub-system of ITER CWS, Component Cooling Water System-1 (CCWS-1) has very large diameter of pipes up to DN1600 with many intersections to fulfill the process flow requirements of clients for heat removal. Pipe intersection is the weakest link in the layout due to high stress intensification factor. CCWS-1 piping up to secondary confinement isolation valves as well as in-between these isolation valves need to survive a Seismic Level-2 (SL-2) earthquake during the Tokamak operation period to ensure structural stability of the system in the Safe Shutdown Earthquake (SSE) event. This paper presents the design, qualification and optimization of layout of ITER CCWS-1 loop to withstand SSE event combined with sustained and thermal loads as per the load combinations defined by ITER and allowable limits as per ASME B31.3. This paper also highlights the Modal and Response Spectrum Analyses done to find out the natural frequency and system behavior during the seismic event. (author)

  19. Residual stress distribution in austenitic stainless steel pipe butt-welded joint measured by neutron diffraction technique

    International Nuclear Information System (INIS)

    Residual stress is inevitable consequence of welding or manufacturing process, which might greatly affect propagation of high-cycle fatigue or SCC crack. In order to evaluate damages due to the crack, it is required to estimate residual stress and to reflect them to the evaluation process as well. The magnitude and distribution of residual stress greatly depend on the individual process of welding or manufacturing, while the accuracy of prediction or measurement is still insufficient. This paper reports the result of residual stress measurement of butt-welded pipe made of austenitic stainless steel. It also intended to improve prediction and measurement techniques concerning to residual stress. The measurement was conducted by neutron diffraction technique employing the diffractometer for residual stress analysis developed by Japan Atomic Energy Agency. The measured results showed typical characteristics of butt-welded pipe both in decline of stress along axial direction and in radial distribution of bending due to axial stress. The measured result agreed qualitatively with the result predicted by the finite element analysis. A quantitative comparison between measured result and analysis showed a shift of the measured stress toward higher tensile. The measured result was also compared with the results by X-ray diffraction and strain-gauge methods to grasp the distinctive results of the methods. (author)

  20. 49 CFR 195.101 - Qualifying metallic components other than pipe.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Qualifying metallic components other than pipe... components other than pipe. Notwithstanding any requirement of the subpart which incorporates by reference an edition of a document listed in § 195.3, a metallic component other than pipe manufactured in...

  1. ADAPTION OF NONSTANDARD PIPING COMPONENTS INTO PRESENT DAY SEISMIC CODES

    Energy Technology Data Exchange (ETDEWEB)

    D. T. Clark; M. J. Russell; R. E. Spears; S. R. Jensen

    2009-07-01

    With spiraling energy demand and flat energy supply, there is a need to extend the life of older nuclear reactors. This sometimes requires that existing systems be evaluated to present day seismic codes. Older reactors built in the 1960s and early 1970s often used fabricated piping components that were code compliant during their initial construction time period, but are outside the standard parameters of present-day piping codes. There are several approaches available to the analyst in evaluating these non-standard components to modern codes. The simplest approach is to use the flexibility factors and stress indices for similar standard components with the assumption that the non-standard component’s flexibility factors and stress indices will be very similar. This approach can require significant engineering judgment. A more rational approach available in Section III of the ASME Boiler and Pressure Vessel Code, which is the subject of this paper, involves calculation of flexibility factors using finite element analysis of the non-standard component. Such analysis allows modeling of geometric and material nonlinearities. Flexibility factors based on these analyses are sensitive to the load magnitudes used in their calculation, load magnitudes that need to be consistent with those produced by the linear system analyses where the flexibility factors are applied. This can lead to iteration, since the magnitude of the loads produced by the linear system analysis depend on the magnitude of the flexibility factors. After the loading applied to the nonstandard component finite element model has been matched to loads produced by the associated linear system model, the component finite element model can then be used to evaluate the performance of the component under the loads with the nonlinear analysis provisions of the Code, should the load levels lead to calculated stresses in excess of Allowable stresses. This paper details the application of component-level finite

  2. Application of the results of pipe stress analyses into fracture mechanics defect analyses for welds of nuclear piping components

    International Nuclear Information System (INIS)

    For the fracture mechanical assessment of postulated or detected crack-like defects in welds of piping systems it is necessary to know the stresses in the un-cracked component normal to the crack plane. Results of piping stress analyses may be used if these are evaluated for the locations of the welds in the piping system. Using stress enhancing factors (stress indices, stress factors) the needed stress components are calculated from the component specific sectional loads (forces and moments). For this procedure the tabulated stress enhancing factors, given in the standards (ASME Code, German KTA regulations) for determination and limitation of the effective stresses, are not always and immediately adequate for the calculation of the stress component normal to the crack plane. The contribution shows fundamental possibilities and validity limits for adoption of the results of piping system analyses for the fracture mechanical evaluation of axial and circumferential defects in welded joints, with special emphasis on typical piping system components (straight pipe, elbow, pipe fitting, T-joint). The lecture is supposed to contribute to the standardization of a code compliant and task-related use of the piping system analysis results for fracture mechanical failure assessment

  3. Transient cooling of electronic components by flat heat pipes

    International Nuclear Information System (INIS)

    This paper presents a theoretical investigation of a Flat Heat Pipe (spreader) designed for the cooling of multiple electronic components in transient state. This model is a transient model, coupling 3D thermal model with a 2D hydrodynamic one through the mass flux of evaporation-condensation, which occurs in a mass conservation equation. The model makes it possible to obtain the FHP wall transient temperatures, the transient pressures, velocities and temperatures in both liquid and vapor phases. A comparison of the behaviour of the FHP and an equivalent solid plate submitted to a transient thermal cycle shows that the FHP enhanced the electronic components cooling for the long thermal cycle duration when the solid plate is more efficient for the very short transient thermal cycles. The FHP provides also a very low thermal resistance, which helps to minimise the temperature gradient and then the hot spots and overheating. - Highlights: → In this study we model a Flat Heat Pipe (FHP) designed for the cooling of electronic components. → This model is a transient model, coupling 3D thermal and hydrodynamic models. → The model makes it possible to obtain the FHP wall transient temperatures. → Transient pressures, velocities and temperatures in liquid and vapor phases are also obtained. → We conclude that FHP is more efficient for long than very short thermal cycles.

  4. NDE Assessments of Cast Stainless Steel Reactor Piping Components

    International Nuclear Information System (INIS)

    Studies conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, have focused on developing and evaluating the effectiveness and reliability of 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 United States Nuclear Regulatory Commission (US NRC) on the utility, effectiveness and reliability of ultrasonic testing (UT) and eddy current testing (ET) inspection techniques as related to the in-service inspection of primary piping components in pressurized water reactors (PWRs). This paper describes recent developments and results from assessments of three different NDE approaches including an ultrasonic phased array inspection methodology, an eddy current testing technique and a low-frequency ultrasonic inspection methodology coupled with a synthetic aperture focusing technique (SAFT). Westinghouse Owner's Group (WOG) cast stainless steel pipe segments with thermal and mechanical fatigue cracks located close to the weld roots, were used for assessing the inspection methods. ET studies were conducted on the inner diameter (ID) surface of piping specimens while the ultrasonic inspection methods were performed from the outer diameter (OD) surface of the specimens. The ET technique employed a ZETEC MIZ-27SI Eddy Current instrument and a ZETEC Z0000857-1 cross point spot probe with an operating frequency of 250 kHz. On some samples where noise levels were high, degaussing of the sample resulted in significant improvements. The phased array approach was implemented using an RD Tech Tomoscan III system operating at 1 MHz and composite volumetric images of the samples were generated. The low-frequency ultrasonic method employs a zone-focused, multi-incident angle; inspection protocol (operating at 250-450 kHz) coupled with a synthetic aperture focusing technique (SAFT) for improved signal-to-noise and advanced imaging capabilities

  5. Detection and sizing of stress corrosion cracks in austenitic components using ultrasonic testing and synthetic aperture focusing technique

    Energy Technology Data Exchange (ETDEWEB)

    Dugan, Sandra; Wagner, Sabine [Stuttgart Univ. (Germany). Materialpruefungsanstalt; Dillhoefer, Alexander [NDT Global GmbH and Co.KG, Stutensee (Germany); Rieder, Hans; Spies, Martin [Fraunhofer-Institut fuer Zerstoerungsfreie Pruefverfahren (IZFP), Saarbruecken (Germany)

    2015-05-01

    Flaw detection and sizing using NDT techniques is an important factor for reliably assessing the integrity of components. In the case of dissimilar metal welds and austenitic stainless steel welds, the grain structure of the weld in combination with the elastic anisotropy of the material will present major challenges for UT. A study on austenitic base metal test blocks with artificially grown IGSCCs has shown that the Synthetic Aperture Focusing Technique (SAFT) can improve the signal-to-noise ratio, particularly for crack tip signals. In welded test blocks, the influence of the inhomogeneous, anisotropic weld has to be considered.

  6. Detection and sizing of stress corrosion cracks in austenitic components using ultrasonic testing and synthetic aperture focusing technique

    International Nuclear Information System (INIS)

    Flaw detection and sizing using NDT techniques is an important factor for reliably assessing the integrity of components. In the case of dissimilar metal welds and austenitic stainless steel welds, the grain structure of the weld in combination with the elastic anisotropy of the material will present major challenges for UT. A study on austenitic base metal test blocks with artificially grown IGSCCs has shown that the Synthetic Aperture Focusing Technique (SAFT) can improve the signal-to-noise ratio, particularly for crack tip signals. In welded test blocks, the influence of the inhomogeneous, anisotropic weld has to be considered.

  7. Technical Letter Report, An Evaluation of Ultrasonic Phased Array Testing for Reactor Piping System Components Containing Dissimilar Metal Welds, JCN N6398, Task 2A

    Energy Technology Data Exchange (ETDEWEB)

    Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Anderson, Michael T.

    2009-11-30

    Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light-water reactor components. The scope of this research encom¬passes primary system pressure boundary materials including dissimilar metal welds (DMWs), cast austenitic stainless steels (CASS), piping with corrosion-resistant cladding, weld overlays, inlays and onlays, and far-side examinations of austenitic piping welds. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in steel components that challenge standard and/or conventional inspection methodologies. This interim technical letter report provides a summary of a technical evaluation aimed at assessing the capabilities of phased-array (PA) ultrasonic testing (UT) methods as applied to the inspection of small-bore DMW components that exist in the reactor coolant systems (RCS) of pressurized water reactors (PWRs). Operating experience and events such as the circumferential cracking in the reactor vessel nozzle-to-RCS hot leg pipe at V.C. Summer nuclear power station, identified in 2000, show that in PWRs where primary coolant water (or steam) are present under normal operation, Alloy 82/182 materials are susceptible to pressurized water stress corrosion cracking. The extent and number of occurrences of DMW cracking in nuclear power plants (domestically and internationally) indicate the necessity for reliable and effective inspection techniques. The work described herein was performed to provide insights for evaluating the utility of advanced NDE approaches for the inspection of DMW components such as a pressurizer surge nozzle DMW, a shutdown cooling pipe DMW, and a ferritic (low-alloy carbon steel)-to-CASS pipe DMW configuration.

  8. Experience with one-layer high-strength ferrite and austenite bellous of pipe joint compensators in 20 ata hot steam pipelines

    International Nuclear Information System (INIS)

    Numerous signs of damage have occured on one-layer high-strength ferrite and austenite bellows of pipe joint compensators in 20 ata superheated-steam pipelines. From a precise analysis of the damage cases, it was found that in ferrite material, the high creep-alternating stress lead to damage, whereas in the austenite material, above all the difficult workability and the slight ductility at operational temperature were determining factors. An attempt was made to prolong the lifetime of the compensators equipped with ferrite bellows by increasing the bellow wall thickness from 3 to 4 mm. These new compensators have so far achieved an operational time of about 10,000 hours with 50 drives without visible deformations or damage. (orig./LH)

  9. Background of SIFs and Stress Indices for Moment Loadings of Piping Components

    International Nuclear Information System (INIS)

    This report provides background information, references, and equations for twenty-four piping components (thirteen component SIFs and eleven component stress indices) that justify the values or expressions for the SIFs and indices

  10. Reactor Materials Program -- weldment component toughness of SRS PWS piping materials. Task number: 89-023-1

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.

    1993-02-01

    The mechanical properties of austenitic stainless steel materials from the reactor systems in the unirradiated (baseline) and the irradiated conditions have been developed previously for structural and fracture analyses of the pressure boundary of the SRS reactor Process Water System (PWS) components. Individual mechanical specimen test results were compiled into three separate weldment components or regions, namely, the base, weld, and weld heat-affected-zone (HAZ), for two orientations (L-C and C-L) with respect to the pipe axis of the source materials and for two test temperatures of 25 and 125{degrees}C. Twelve separate categories were thus defined to assess the effect of test conditions on the mechanical properties and to facilitate selection of properties for structural and fracture analyses. The testing results show high fracture toughness of the materials and support the demonstration of PWS pressure boundary structural integrity under all conditions of reactor operation. The fracture toughness of a fourth weldment component, namely, the weld fusion line region, has been measured to evaluate the potential for a region of low toughness in the interface between the Type 308 stainless steel weld metal and the Type 304 stainless steel pipe. The testing details and results of the weld fusion line toughness are contained in this report.

  11. Probabilistic structural integrity assessment based on uncertainty of weld residual stress at the piping butt-welds of nuclear reactor components

    International Nuclear Information System (INIS)

    Stress corrosion cracking (SCC), which affects the structural integrity of reactor component, has been observed at some piping joints made by austenitic stainless steel in BWR plants. It is well known that the SCC behavior is significantly scattered depending upon the various conditions such as materials, piping geometry, crack growth rate, weld residual stress, and so on. Since probabilistic fracture mechanics (PFM) analysis method treats such scatter and uncertainties in the structural integrity evaluation, it is, therefore, useful to apply the PFM analysis to the evaluation of the piping integrity. In JAEA, the PFM analysis code of PASCAL-SP for aged piping has been developed based on Monte Carlo method as described in our previous paper. Among the conditions related to SCC behavior, weld residual stress near the welded joint is one of the most important factors to assess the structural integrity of piping because the tensile residual stress becomes a driving force of a SCC. Welding conditions such as heat input, welding speed and piping geometry affect weld residual stress distribution at the welded joint of piping. Effect of the welding conditions on the weld residual stress distribution has not yet been evaluated quantitatively. Hence, in this study, an effect of uncertainty of welding conditions, such as scatters of heat input and welding speed during welding, on weld residual stress at the piping butt-welds was evaluated using the simulation method by varying the welding conditions. Probabilistic fracture mechanics analysis using PASCAL-SP was also performed to evaluate the effect of uncertainty of weld residual stress on the break probability of piping. It was clarified that the break probability increased with increasing the uncertainties of residual stress. (author)

  12. Trend of large size, integral type component materials for light water reactor piping

    International Nuclear Information System (INIS)

    As the substitute energy for oil, nuclear power generation is given second look, because it has been technically established, and the fuel price has been stable. The reliability of nuclear power generation and the structural soundness of its equipment have been demanded more strictly. The rationalization of in-service inspection and other measures to improve the operational efficiency are strongly sought. For these purposes, the manufacture of large, one-body components has been promoted for pressure vessels and others in light water reactors. In the piping members occupying large weight quantitatively and qualitatively in nuclear power stations, this tendency has been similarly seen, and forged steel components are materialized. The basic concept of making large, one-body components for piping system is to make seamless pipes and fittings, to elongate pipes, and to make pipes and fittings or two or more fittings into one-body. The typical examples of making large, one-body piping components are shown, and the planning, manufacture and various properties of large, one-body piping components are reported. The manufacturing techniques for piping components can satisfy these demands. (Kako, I.)

  13. Experiments and calculations to leak openings and leak rates on typical piping components and systems

    International Nuclear Information System (INIS)

    Calculations of leak opening and leak rate for through cracks in piping components have been performed. The analyses are pre- or mostly post-calculations to experiments performed at the HDR facility under PWR operating conditions. Piping components under consideration are small diameter straight pipes with circumferential cracks, pipe bends with longitudinal or circumferential cracks and pipe branches with weldment cracks. The component are loaded by internal pressure and opening as well as closing bending moment. The finite element method and two-phase flow leak rate programs are used for the calculations. Results of the analyses are presented as J-integral values, crack opening displacements and areas and leak rates as well as comparisons to the experimental results. 6 refs., 16 figs., 2 tabs

  14. Testing Of RSG-GAS Secondary Cooling Component

    International Nuclear Information System (INIS)

    The aim of this activity is to improve the mechanical testing technology knowledge of the research reactor cooling pipe material, through a series of testing for the components especially for the mechanical properties of austenitic steel pipe used in RSG-GAS secondary cooling pipe. Scopes of these testing activities are chemical composition analysis, tensile testing, and hardness testing

  15. Probabilistic failure assessment of PWR nuclear power plant piping components against erosion-corrosion

    International Nuclear Information System (INIS)

    Erosion-corrosion (EC) is one of the important and complex degradation mechanisms in the nuclear power plant piping systems. Depending on the nature of piping material, piping geometry and operating conditions, different components of piping system are susceptible to EC to different degrees. Due to variations in operating conditions and inherent uncertainty associated with the prediction models, EC is to be treated as a random phenomenon. The effect of randomness should be considered in the design of piping components. In this paper, an attempt is made to apply system reliability concept to determine the reliability of an elbow against EC at different times. The application of system reliability concept helps in taking into account: there are a number of sections within a given piping component that are vulnerable to undergo EC - reflecting the complexity of EC mechanism, and, the safety margins of these sections within a component, connected in series, are positively correlated. The usefulness of the model developed in estimation of reliability of elbows at different times is demonstrated through two example problems. A flowchart that can be used for reliability-based design of piping components against EC (in conjunction with ASME design procedure) is also presented. (author)

  16. Concept for the monitoring of weld seams on pressure pipes made of austenitic steel using ultrasonic processes

    International Nuclear Information System (INIS)

    A concept for the repeat testing of weld seams in pipelines made of Austenitic steel using ultrasonic processes is developed and tested. Even with varying sound attenuation in the material and on the occurence of 'ghost' indications, it ensures certain indication of faults. Further, the depth of cracks can be assessed, for example by the height of echos, even for large extents. (orig.)

  17. Plastic fracture toughness of austenitic welding connection for Ver-1000 nuclear reactor piping of 300-350 mm diameter

    International Nuclear Information System (INIS)

    The outside welding technology for circular welds in a pearlitic tube using austenitic welding wire materials is developed and applied in manufacturing pipelines of CPP and ECC. Mechanical properties and fracture toughness of austenitic welded joints in pearlitic tubes are determined to substantiate by calculation the practicality of the leakage prior to failure concept. The work is accomplished on experimental tube manufactured by hand arc welding. When manufactured the tube is cut into 5 rings. From the rings the tensile specimens are cut for testing at 20 and 350 deg C as well as Charpy V-notch impact specimens and compact specimens ST-1T. It is shown that the materials of the experimental tube meet the standard requirements. Only axial specimens cut across the weld are not in conformity with the requirements for specific elongation

  18. Evaluation of material integrity on electricity generator water steam cycles component (Main Steam Pipe)

    International Nuclear Information System (INIS)

    The evaluation of material integrity on electricity generator component has been done. That component was main steam pipe of Unit II Suralaya Coal Fired Power Plant. evaluation was done by replication technique. The damage was found are two porosity's, from two point samples of six points sample population. Based on cavity evaluation in steels, which proposed by Neubauer and Wedel that porosity's still at class A damage. For class A damage, its means no remedial action would be required until next major scheduled maintenance outage. That porosity's was grouped on isolated cavities and not need ti repair that main steam pipe component less than three year after replication test

  19. Development of Low-Cost Austenitic Stainless Gas-Turbine and Diesel Engine Components with Enhanced High-Temperature Reliability

    Energy Technology Data Exchange (ETDEWEB)

    Maziasz, P.J.; Swindeman, R.W.; Browning, P.F. (Solar Turbines, Inc.); Frary, M.E. (Caterpillar, Inc.); Pollard, M.J.; Siebenaler, C.W.; McGreevy, T.E.

    2004-06-01

    In July of 1999, a Cooperative Research and Development Agreement (CRADA) was undertaken between Oak Ridge National Laboratory (ORNL) and Solar Turbines, Inc. and Caterpillar, Inc. (Caterpillar Technical Center) to evaluate commercial cast stainless steels for gas turbine engine and diesel engine exhaust component applications relative to the materials currently being used. If appropriate, the goal was to develop cast stainless steels with improved performance and reliability rather than switch to more costly cast Ni-based superalloys for upgraded performance. The gas-turbine components considered for the Mercury-50 engine were the combustor housing and end-cover, and the center-frame hot-plate, both made from commercial CF8C cast austenitic stainless steel (Fe-l9Cr-12Ni-Nb,C), which is generally limited to use at below 650 C. The advanced diesel engine components considered for truck applications (C10, C12, 3300 and 3400) were the exhaust manifold and turbocharger housing made from commercial high SiMo ductile cast iron with uses limited to 700-750 C or below. Shortly after the start of the CRADA, the turbine materials emphasis changed to wrought 347H stainless steel (hot-plate) and after some initial baseline tensile and creep testing, it was confirmed that this material was typical of those comprising the abundant database; and by 2000, the emphasis of the CRADA was primarily on diesel engine materials. For the diesel applications, commercial SiMo cast iron and standard cast CN12 austenitic stainless steel (Fe-25Cr-13Ni-Nb,C,N,S) baseline materials were obtained commercially. Tensile and creep testing from room temperature to 900 C showed the CN12 austenitic stainless steel to have far superior strength compared to SiMo cast iron above 550 C, together with outstanding oxidation resistance. However, aging at 850 C reduced room-temperature ductility of the standard CN12, and creep-rupture resistance at 850 C was less than expected, which triggered a focused

  20. Simulation and control of the cyclic damage of the NPP pipe made of austenite steel after long-term service

    International Nuclear Information System (INIS)

    Simulation of the cyclic damage of austenite steel 12Kh18N10T and simultaneous control and study of the metal structure using nondestructive methods were performed. The metal was studied at the initial stage and at different stages of aging according to the low-cycle fatigue mechanism. The local magnetic method thus used is highly informative and the results obtained correspond to the character of changes in the cyclic strength properties of steel 12Kh18N10T. The structure study revealed that structure change attributed to plastic deformation at different stages of cyclic damage correlates with the magnetic signal

  1. Development of a high temperature austenitic stainless steel for Stirling engine components

    International Nuclear Information System (INIS)

    An alloy, designed NASAUT 4G-A1, was developed which exhibited an excellent balance of oxidation resistance and high temperature strength while maintaining an austenitic matrix necessary for hydrogen compatibility. This alloy, having the composition 15Cr-15Mn-2Mo-1Nb-1Si-1.5C-bal. Fe in wt%, was microstructurally characterized and shown to contain a fine M/sub 23/C/sub 6/ precipitated phase. Subsequent heat treatments were shown to substantially modify this microstructure resulting in improved mechanical properties. Yield, creep and low cycle fatigue strengths were found to be superior to the best iron base alloy thus far identified as a potential heater head candidate material, XF-818

  2. Ratchetting failure of the piping components subjected to seismic loading- experimental and numerical studies

    International Nuclear Information System (INIS)

    Strain accumulation induced by cyclic loading, i.e., ratchetting is important in designing structural components. It can reduce the fatigue life or can cause failure of piping components or systems subjected to seismic or other cyclic loads. The 1995 ASME B and PV code, Section III; has been modified to incorporate reverse dynamic loading and ratchetting. In the present investigation ANSYS software package, which incorporates Chaboche kinematic hardening model, was used to study the ratchetting. The basic features of Chaboche model and the determination of the parameters for the model have been discussed in this paper. Two sets of experimental data namely viz. (a) three point and four point bending test on straight pipe (b) shake table test on pipe elbow system, performed by BARC were used for validating ANSYS results. ANSYS over predicts ratchetting compared to the experimental values. (author)

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

    International Nuclear Information System (INIS)

    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)

  4. Assessment of thermal aging embrittlement of cast austenitic stainless steel components in the Babcock and Wilcox -designed PWR reactor internals

    International Nuclear Information System (INIS)

    The currently operating Babcock and Wilcox (BW) designed pressurized water reactors (PWRs) were constructed during the late sixties and seventies. Some of the reactor internals components were fabricated from cast austenitic stainless steel (CASS). The selection of CASS for the internals components was made to expedite the construction schedule by reducing machining and allowing production in large quantities. Since then, test data have shown that some CASS materials are susceptible to thermal aging embrittlement at PWR operating temperatures and its effect on functionality is of concern. Recently, the US nuclear power industry has developed inspection and evaluation guidelines (MRP-227, Rev.0) for managing aging degradation in PWR reactor internals for both the current and extended license periods. The MRP-227, Rev.0 guidelines recommend additional inspections for certain internals components including CASS components in BW PWRs due to thermal aging embrittlement concerns. The thermal aging embrittlement susceptibility for CASS can be assessed by the casting method and ferrite content if sufficient information in the original fabrication records is available. AREVA NP has performed a fabrication records search to identify several CASS components in the BW PWR internals and reviewed the archived fabrication records. A database has been assembled as a result of this records search. Based on the fabrication records, the ferrite content is determined using Hull's equivalent factors. Grade CF8 castings (without molybdenum) have been found to not be susceptible to thermal aging embrittlement. However, thermal aging embrittlement is a potential concern for Grade CF3M castings (containing 2 to 3% molybdenum). As a result of this assessment, several CASS components in the BW PWRs are concluded to not be susceptible to thermal aging embrittlement. The findings provide the basis for the removal of these CASS components from the additional inspection requirements in MRP-227

  5. Service life determination for a fatigue-limited Class 1 piping component

    International Nuclear Information System (INIS)

    The design fatigue life assessment of an ASME Class 1 piping component typically has a significant safety margin. Methods containing varying degrees of detail may be used to develop the component's fatigue usage factor. This paper presents the technical bases used to establish a more realistic fatigue life for a piping component which was nearing its calculated design fatigue limit. The original design basis methodology was superseded by a more detailed inelastic evaluation to demonstrate structural integrity and determine the revised component fatigue life. An effective design cycle curve was developed to assess future fatigue damage, and for tracking ongoing fatigue accumulation. In addition, the existing transient tracking procedure was updated to take into account the actual transient severity for all future fatigue-significant events

  6. Pipe and cable guide for moving components in rotation

    International Nuclear Information System (INIS)

    This device is for guiding the cables and tubes of the appliances arranged in the core cover. It is characterized by a vertical guiding mast integral at its bottom end with a rotating component with an upper plate, a lower plate and at least one intermediate plate. These plates are horizontal and comprise around their periphery the regularly distributed facilities for securing the cables. The intermediate plate or plates include means for limiting its rotation around the mast so that the angle of rotation between the top plate and the bottom plate is evenly distributed between the intermediate plates

  7. Non-Newtonian Liquid Flow through Small Diameter Piping Components: CFD Analysis

    Science.gov (United States)

    Bandyopadhyay, Tarun Kanti; Das, Sudip Kumar

    2016-05-01

    Computational Fluid Dynamics (CFD) analysis have been carried out to evaluate the frictional pressure drop across the horizontal pipeline and different piping components, like elbows, orifices, gate and globe valves for non-Newtonian liquid through 0.0127 m pipe line. The mesh generation is done using GAMBIT 6.3 and FLUENT 6.3 is used for CFD analysis. The CFD results are verified with our earlier published experimental data. The CFD results show the very good agreement with the experimental values.

  8. Proceedings of the specialists' meeting on reliability of the ultrasonic inspection of austenitic materials

    International Nuclear Information System (INIS)

    The contributions of this meeting addressed several topics: the fundamentals of ultrasonic examination of austenitic materials (effect of anisotropy on propagation, improvement of ultrasonic testing to thick bimetallic welds, aspects of the ultrasonic testing of austenitic steel structures, utilization of a Fisher linear discriminant function in intergranular stress corrosion cracking or IGSCC detection, case of coarse grain austenitic welds, efforts of the Argonne National Laboratory), instruments and methods (longitudinal wave ultrasonic inspection, Grass echo suppression technique during the ultrasonic inspection of fuel cladding tubes, inspections of fillet and butt welds, improvement by signal averaging techniques, multiple bearing angle crack detector for cladded pipes examinations, flow-to-grain echo enhancement by split-spectrum processing, ultrasonic imaging techniques, ultrasonic inspection of pipe weldments for IGSCC), industrial practice (ultrasonic testing techniques for fabrication and in-service inspection, experiences in ultrasonic examination of austenitic steel components, experience and practice on nuclear piping in Spain, detection of underclad defects, sizing of cracks perpendicular to stainless overlay), and reliability (survey of ultrasonic testing in austenitic weld material, examination of electron beam welds, factors affecting the reliability of ultrasonic examination, detectability of IGSCC, ultrasonic inspection reliability for primary piping systems)

  9. Service Life Of Main Piping Component Due To Low Thermal Stresses.Fatigue

    International Nuclear Information System (INIS)

    The paper deals with estimating the service life of the power station Main piping component and describing the repair process for extending of its service life. After a long period of service, several circular fatigue cracks have been discovered at the bottom of the Main piping component chamber. Finite element analyses of transient thermal stresses, caused by power station startup, are carried out in the paper. The calculation results show good agreement between the theoretical locations of the maximum stresses and the actual locations of the cracks. There is a good agreement between theoretical evaluation and actual service life, as well. The possibility of machining out the cracks in order to prevent their growing is examined here. The machining enables us to extend the power station component's life service

  10. Thermal fatigue screening criteria for identifying susceptible piping components in CANDU stations

    Energy Technology Data Exchange (ETDEWEB)

    Schefski, C.; Chen, Q.; Pentecost, S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2011-07-01

    In December 1987, a fatigue failure in a non-isolable section of a safety injection line at the Farley-2 plant prompted the U.S. Nuclear Regulatory Commission (NRC) to issue Bulletin 88-08 requiring that U.S. utilities review all non-isolable branch lines to determine if they are susceptible to thermal fatigue. The thermal fatigue incident at Farley-2 was caused by stresses in the pipe wall resulting from large-scale temperature fluctuations. Shortly after the Farley-2 event, several other incidents with through-wall cracks due to thermal fatigue had occurred in plant subsystems and piping configurations similar to the Farley-2 safety injection line. Thermal fatigue cracks have also occurred in piping configurations with different geometries, such as drain, residual heat removal, and shutdown cooling suction lines in various pressurized water reactors (PWR) and boiling water reactors (BWR). Thermal fatigue, caused by local thermal stratification phenomena, has received significant attention in the PWR and BWR communities in the past two decades. Although CANDU stations have experienced relatively few thermal fatigue failures; the impact of this known fatigue mechanism for CANDU designs has not been rigorously assessed. Screening and evaluation methodology, which has been developed by Electric Power Research Institute (EPRI) to identify locations susceptible to thermal cycling in PWR systems, has recently been modified under a CANDU Owners Group (COG) project for application in CANDU piping systems. This paper describes a new software tool for evaluating locations susceptible to thermal fatigue in CANDU piping systems in an effort to avoid failures that lead to costly plant shutdowns. The software, combined with engineering judgement, will assist CANDU station staff to focus their inspections on key components, therefore reducing dose, time and cost during outages. Computational Fluid Dynamics (CFD) was used to form the basis for expanding the range of validity in

  11. CORROSION ISSUES ASSOCIATED WITH AUSTENITIC STAINLESS STEEL COMPONENTS USED IN NUCLEAR MATERIALS EXTRACTION AND SEPARATION PROCESSES

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J.; Louthan, M.; Sindelar, R.

    2012-12-17

    This paper illustrated the magnitude of the systems, structures and components used at the Savannah River Site for nuclear materials extraction and separation processes. Corrosion issues, including stress corrosion cracking, pitting, crevice corrosion and other corrosion induced degradation processes are discussed and corrosion mitigation strategies such as a chloride exclusion program and corrosion release testing are also discussed.

  12. Compilation of references, data sources and analysis methods for LMFBR primary piping system components

    International Nuclear Information System (INIS)

    A survey and review program for application of fracture mechanics methods in elevated temperature design and safety analysis has been initiated in December of 1976. This is the first of a series of reports, the aim of which is to provide a critical review of the theories of fracture and the application of fracture mechanics methods to life prediction, reliability and safety analysis of piping components in nuclear plants undergoing sub-creep and elevated temperature service conditions

  13. Compilation of references, data sources and analysis methods for LMFBR primary piping system components

    Energy Technology Data Exchange (ETDEWEB)

    Reich, M.; Esztergar, E.P.; Ellison, E.G.; Erdogan, F.; Gray, T.G.F.; Wells, C.W.

    1977-03-01

    A survey and review program for application of fracture mechanics methods in elevated temperature design and safety analysis has been initiated in December of 1976. This is the first of a series of reports, the aim of which is to provide a critical review of the theories of fracture and the application of fracture mechanics methods to life prediction, reliability and safety analysis of piping components in nuclear plants undergoing sub-creep and elevated temperature service conditions.

  14. Microstructural Characteristic of Dissimilar Welded Components (AISI 430 Ferritic-AISI 304 Austenitic Stainless Steels) by CO2 Laser Beam Welding (LBW)

    OpenAIRE

    Caligulu, Ugur; Dikbas, Halil; Taskin, Mustafa

    2012-01-01

    In this study, microstructural characteristic of dissimilar welded components (AISI 430 ferritic-AISI 304 austenitic stainless steels) by CO2 laser beam welding (LBW) was investigated. Laser beam welding experiments were carried out under argon and helium atmospheres at 2000 and 2500 W heat inputs and 100-200-300 cm/min. welding speeds. The microstructures of the welded joints and the heat affected zones (HAZ) were examined by optical microscopy, SEM, EDS and XRD analysis. The tensile strengt...

  15. Review of environmental effects on fatigue crack growth of austenitic stainless steels

    International Nuclear Information System (INIS)

    Fatigue and environmentally assisted cracking of piping, pressure vessel cladding, and core components in light water reactors are potential concerns to the nuclear industry and regulatory agencies. The degradation processes include intergranular stress corrosion cracking of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or stress corrosion cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Crack growth data for wrought and cast austenitic SSs in simulated BWR water, developed at Argonne National Laboratory under US Nuclear Regulatory Commission sponsorship over the past 10 years, have been compiled into a data base along with similar data obtained from the open literature. The data were analyzed to develop corrosion-fatigue curves for austenitic SSs in aqueous environments corresponding to normal BWR water chemistries, for BWRs that add hydrogen to the feedwater, and for pressurized water reactor primary-system-coolant chemistry

  16. A sensibility study of the piping spans following the weight variations of the in-line components

    International Nuclear Information System (INIS)

    Piping systems in nuclear power plants are composed of piping and in-line components such as flanges, valves, insulations, etc. Piping analyses including dead weight and seismic analyses are based on the weights and locations of these piping components. Frequently, due to various material changes, site, and construction problems, piping design changes are inevitable after piping drawings were issued for construction. These design changes could lead to serious impact on cost and schedule. As such, site engineers must react quickly based on experiences to make the correct decision to mitigate the problems with construction costs and schedule. This paper provides simple equations for rapid span-checks to determine if the changes are deemed acceptable instead of undertaking time-consuming detailed analyses which take long time to complete. These simple span-check equations are derived from the sensitivity analysis of piping spans through variation of masses, sizes, and locations of in-line components. The criteria used in deriving the simple span-check equations are based on the acceptable deflections and stresses of various pipe sizes used in APR1400 (Korea's Advanced Power Reactor 1400 MWe class). The results show that the simple span-check equations derived herein can be a useful tool for site engineers to respond to frequent design change requests. (authors)

  17. Progress report on a NDT round robin on austenitic circumferential pipe welds; Fortschrittsbericht ueber einen ZfP-Ringversuch an austenitischen Rohrleitungs-Rundschweissnaehten

    Energy Technology Data Exchange (ETDEWEB)

    Brast, G. [Preussische Elektrizitaets-AG (Preussenelektra), Hannover (Germany); Maier, H.J.; Knoch, P.; Mletzko, U. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt

    1998-11-01

    The objective of the project is establish on the basis of Round Robin tests the current state of efficiency of various, defined testing methods, so that required or achievable optimizations can be defined and made. The project work up to date encompasses mon-destructive examinations of 15 austenitic welds with nominal widths DN 150/200/250 and wall thicknesses from 8 to 18 mm. Except for one test piece, (elbow/elbow), the joining welds are straight pipe to elbow welds. The results of the Round Robin tests show that the NDE detection limits for the fault examined (intercrystalline stress corrosion cracking) are in the range assumed so far, i.e. from about 20 to 25% of the wall thickness to be examined. The defect detection rates of the ultrasonic test methods applied are approx. 70% and thus are about equal in achievement with comparable international Round Robin tests (PISC; ASME/PDI, ENIQ, etc.). Clearly better are the fault detection rates of radiography. Evaluation of the individual results indicates the detection limits can be improved, by 1. reducing the misalignment of edges, 2. grinding of welds, 3. avoiding sharp notches at the root, 4. producing coaxial surfaces. (orig./CB) [Deutsch] Ein Ziel des Vorhabens ist es, mit Ringversuchen den derzeitigen Stand der Leistungsfaehigkeit einzelner Pruefverfahren und -techniken zu erkennen, um moeglicherweise notwendige Optimierungen vornehmen zu koennen. Das Vorhaben umfasst bis jetzt zerstoerungsfreie Pruefungen an 15 austenitischen Naehten mit Nennweiten DN 150/200/250 und Wandstaerken zwischen 8 und 18 mm. Mit einer Ausnahme (Bogen/Bogen) handelt es sich um Verbindungen Geradrohr/Bogen. Die Ergebnisse des Ringversuches weisen darauf hin, dass die Nachweisgrenzen der ZfP fuer den vorliegenden Fehlertyp (Interkristalline Spannungsrisskorrosion) in der bisher schon angenommenen Groessenordnung von ca. 20-25% der geprueften Wanddicke liegen. Die Fehlerauffind-Raten der US-Pruefung liegen mit ca. 70% im Rahmen

  18. A ferric-austenitic CrNiMoN steel alloy to be used as material to manufacture welded components

    International Nuclear Information System (INIS)

    A chromium-nickel-molybdenum-nitrogen steel alloy (ferritic-austenite) is used to manufacture welded articles which without thermal treatment are resistant to pitting corrosion, intergranular corrosion (Monypenny-Stauss test) or boiling in 65% nitric acid with subsequent cross-breaking test. (IHOE)

  19. Application of PHADEC method for the decontamination of radioactive steam piping components of Caorso plant

    International Nuclear Information System (INIS)

    Highlights: • Application of PHADEC chemical off-line methodology. • Decontamination of radioactive steam piping components of Caorso turbine building. • Experimental characterization of metallic components, e.g., by SEM analysis. • Measure of the efficiency of treatment by means of the reduction of activity and vs. the treatment time. • Minimization of secondary waste produced during decontamination activity of Caorso BWR plant. - Abstract: The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue. In this paper the attention was focused on the PHADEC (PHosphoric Acid DEContamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc. The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing. Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters. This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e., taken along the main steam piping line. Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the

  20. Report on round robin activities on the calculation of crack opening behaviour and leak rates for small bore piping components

    International Nuclear Information System (INIS)

    Results of a bench mark test on crack opening and leak rate calculation are presented. The bench mark test is based on two experiments performed in phase III of the german HDR-Safety-Program which was sponsored by the German Federal Minister of Research and Technology (BMFT). The pipes considered in these experiments were a straight pipe with 80 mm diameter and a circumferential through wall crack as well as a pipe branch with a crack in the weldment between nozzle and main pipe. Both test pieces were made of austenitic steel and loaded by internal pressure and varying bending moment. The round robin was initiated by Principal Working Group No.3 (PWG-3) of the committee on the Safety of Nuclear Installations (CSNI), witch is part of OECD (Organization for Economic Cooperation and Development)'s Nuclear Energy Agency (NEA). Scientists of five institutions in four countries (Canada, USA, Czech Republic and Germany) participated in the bench mark test. For the evolution of the crack opening either analytical methods, estimation schemes or the finite element method were used, while leak rates were calculated by means of two-phase flow models. The compilation of the results shows very large scatter bands in general, with deviations equally large between the calculations of the different participants and the calculation and the measurements. To identify reasons for this scatter - probably originating from differences between the methods used and from uncertainties in the experiment - in detail, further evaluations were made meanwhile. These are described in chapter 9, witch is added to the draft report of June 1993 of the first phase. (authors). 17 refs., 51 figs., 14 tabs

  1. Impact of the amount of working fluid in loop heat pipe to remove waste heat from electronic component

    Directory of Open Access Journals (Sweden)

    Smitka Martin

    2014-03-01

    Full Text Available One of the options on how to remove waste heat from electronic components is using loop heat pipe. The loop heat pipe (LHP is a two-phase device with high effective thermal conductivity that utilizes change phase to transport heat. It was invented in Russia in the early 1980’s. The main parts of LHP are an evaporator, a condenser, a compensation chamber and a vapor and liquid lines. Only the evaporator and part of the compensation chamber are equipped with a wick structure. Inside loop heat pipe is working fluid. As a working fluid can be used distilled water, acetone, ammonia, methanol etc. Amount of filling is important for the operation and performance of LHP. This work deals with the design of loop heat pipe and impact of filling ratio of working fluid to remove waste heat from insulated gate bipolar transistor (IGBT.

  2. Impact of the amount of working fluid in loop heat pipe to remove waste heat from electronic component

    Science.gov (United States)

    Smitka, Martin; Kolková, Z.; Nemec, Patrik; Malcho, M.

    2014-03-01

    One of the options on how to remove waste heat from electronic components is using loop heat pipe. The loop heat pipe (LHP) is a two-phase device with high effective thermal conductivity that utilizes change phase to transport heat. It was invented in Russia in the early 1980's. The main parts of LHP are an evaporator, a condenser, a compensation chamber and a vapor and liquid lines. Only the evaporator and part of the compensation chamber are equipped with a wick structure. Inside loop heat pipe is working fluid. As a working fluid can be used distilled water, acetone, ammonia, methanol etc. Amount of filling is important for the operation and performance of LHP. This work deals with the design of loop heat pipe and impact of filling ratio of working fluid to remove waste heat from insulated gate bipolar transistor (IGBT).

  3. Progress in the reliable inspection of cast stainless steel reactor piping components

    International Nuclear Information System (INIS)

    Studies conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, have focused on assessing the effectiveness and reliability of 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 United States Nuclear Regulatory Commission (US NRC) on the utility, effectiveness and reliability of ultrasonic testing (UT) and eddy current testing (ET) inspection techniques as related to the inservice ultrasonic inspection of primary piping components in pressurized water reactors (PWRs). This paper describes progress, recent developments and results from assessments of three different NDE approaches including ultrasonic phased array inspection techniques, eddy current testing for surface-breaking flaws, and a low-frequency ultrasonic inspection methodology coupled with a synthetic aperture focusing technique (SAFT). 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 spool pieces were used for assessing the inspection methods. Eddy current studies were conducted on the inner diameter (ID) surface of piping specimens while the ultrasonic inspection methods were applied from the outer diameter (OD) surface of the specimens. The eddy current technique employed a Zetec MIZ-27SI Eddy Current instrument and a Zetec Z0000857-1 cross point spot probe with an operating frequency of 250 kHz. In order to reduce noise effects, degaussing of a subset of the samples resulted in noticeable improvements. The phased array approach was implemented using an R/D Tech Tomoscan III system operating at 1 MHz, providing composite volumetric images of the samples. The low-frequency ultrasonic method employs a zone-focused, multi-incident angle inspection protocol (operating at 250-500 kHz) coupled with SAFT for improved signal-to-noise and

  4. Progress in the Reliable Inspection of Cast Stainless Steel Reactor Piping Components

    International Nuclear Information System (INIS)

    Studies conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, have focused on assessing the effectiveness and reliability of 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 United States Nuclear Regulatory Commission (US NRC) on the utility, effectiveness and reliability of ultrasonic testing (UT) and eddy current testing (ET) inspection techniques as related to the inservice ultrasonic inspection of primary piping components in pressurized water reactors (PWRs). This paper describes progress, recent developments and results from assessments of three different NDE approaches including ultrasonic phased array inspection techniques, eddy current testing for surface-breaking flaws, and a low-frequency ultrasonic inspection methodology coupled with a synthetic aperture focusing technique (SAFT). 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 spool pieces were used for assessing the inspection methods. Eddy current studies were conducted on the inner diameter (ID) surface of piping specimens while the ultrasonic inspection methods were applied from the outer diameter (OD) surface of the specimens. The eddy current technique employed a Zetec MIZ-27SI Eddy Current instrument and a Zetec Z0000857-1 cross point spot probe with an operating frequency of 250 kHz. In order to reduce noise effects, degaussing of a subset of the samples resulted in noticeable improvements. The phased array approach was implemented using an R/D Tech Tomoscan III system operating at 1 MHz, providing composite volumetric images of the samples. The low-frequency ultrasonic method employs a zone-focused, multi-incident angle inspection protocol (operating at 250-500 kHz) coupled with SAFT for improved signal-to-noise and

  5. Early detection of damage in austenitic materials under thermocyclic stress

    International Nuclear Information System (INIS)

    Austenitic pipe systems in nuclear power plants are subject to thermomechanical stress. Temperature gradients resulting from alternating cold and warm fluids in the pipelines lead to stress gradients, especially in the surge line, lines of the volume control and aftercooling system, and spray lines. The thermomechanical stress induces microstructural changes and fatigue of the pipeline material. The project focused on the development of new, nondestructive sensor systems and concepts for detecting and interprating fatigue-induced microstructural changes prior to crack initiation. In a joint research project of IZFP (Fraunhofer Institute of Non-Destructive Test Methods) and the WKK (Materials Science Department) of TU Kaiserslautern University, thermo-elastic/plastic load states of an austenitic pipeline steel were investigated systematically in fatigue tests. Conventional and nondestructive in-situ characterisation of the fatigue characteristics were applied to find an early detection concept for selective component monitoring in the context of pro-active ageing management.

  6. Statistical laws of IGSCC of welded joints in the austenitic Du-300 piping of the multiple forced circulation circuit of the RBMK reactors

    International Nuclear Information System (INIS)

    The process of intergranular corrosion stress cracking is analyzed for welded joints of Du-300 austenitic pipelines of multiple forced circulation circuit of the RBMK reactor. Statistical estimates of parameters determining the service life of pipelines are obtained. The calculation procedure is developed for residual lifetime of Du-300 pipeline welded joints with taking into account the scatter in the data on critical degree of sensitization and crack growth rates

  7. Imaging and size analysis of stress corrosion cracks in austenitic components using the synthetic aperture focus technique

    International Nuclear Information System (INIS)

    Riss formation and growth by intercrystalline stress corrosion cracking occurs especially in nickel alloys in case of mixed steels and also in the heat-affected zone in some austenitic Cr-Ni steels. In view of the strong branching of these cracks, amplitude-based ultrasonic methods of measurement may fail. The contribution describes the detection and size analysis of stress corrosion cracks. The synthetic aperture focus technique (SAFT) was used to improve the signal-noise ratio of the ultrasonic inspection data, especially for crack tip identification. several test bodies with intercrystalline stress corrosion cracks with depths ranging from 2.5 mm to 16 mm were analyzed successfully by a combination of conventional techniques for acquisition of B-scan data, followed by SAFT processing.

  8. Reliability Data for Piping Components in Nordic Nuclear Power Plants 'R-Book'. Project Phase 1. Rev 1

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, Bengt (Scandpower Risk Management Inc., Houston, TX (US)); Olsson, Anders (Relcon Scandpower AB, Stockholm (SE))

    2008-01-15

    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

  9. Reliability Data for Piping Components in Nordic Nuclear Power Plants 'R-Book'. Project Phase 1. Rev 1

    International Nuclear Information System (INIS)

    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

  10. Evidence of cracks in austenitic pipe weldings with a radiometric inspection system; Nachweis von Rissen in austenitischen Rohrleitungsnaehten mit einem radiometrischen Pruefsystem

    Energy Technology Data Exchange (ETDEWEB)

    Maier, H.J.; Wuensch, W. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt

    1999-08-01

    The paper reports the development of a radiometric prototype device and its application to inspection of austenitic weldings with intercrystalline crack defects. The device initially was intended to be used for supplemental inspection for clarification of contradictory or unclear testing results, but the results obtained justify to consider the possibility of using it as an independent, full-scope testing instrument. (orig./CB) [Deutsch] Berichtet wird ueber die Entwicklung eines Prototypes eines Radiometrie-Geraetes zur Pruefung von austenitischen Schweissnaehten mit interkristalliner Rissbildung, zunaechst als Entscheidungshilfe bei unklaren bzw. sich widersprechenden Pruefresultaten. Zwischenzeitlich wird auch daran gedacht, ein solches Geraet fuer eine vollstaendige Pruefung weiter zu entwickeln. (orig./DGE)

  11. Reliability of piping system components. Framework for estimating failure parameters from service data

    Energy Technology Data Exchange (ETDEWEB)

    Nyman, R. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hegedus, D.; Tomic, B. [ENCONET Consulting GesmbH, Vienna (Austria); Lydell, B. [RSA Technologies, Vista, CA (United States)

    1997-12-01

    This report summarizes results and insights from the final phase of a R and D project on piping reliability sponsored by the Swedish Nuclear Power Inspectorate (SKI). The technical scope includes the development of an analysis framework for estimating piping reliability parameters from service data. The R and D has produced a large database on the operating experience with piping systems in commercial nuclear power plants worldwide. It covers the period 1970 to the present. The scope of the work emphasized pipe failures (i.e., flaws/cracks, leaks and ruptures) in light water reactors (LWRs). Pipe failures are rare events. A data reduction format was developed to ensure that homogenous data sets are prepared from scarce service data. This data reduction format distinguishes between reliability attributes and reliability influence factors. The quantitative results of the analysis of service data are in the form of conditional probabilities of pipe rupture given failures (flaws/cracks, leaks or ruptures) and frequencies of pipe failures. Finally, the R and D by SKI produced an analysis framework in support of practical applications of service data in PSA. This, multi-purpose framework, termed `PFCA`-Pipe Failure Cause and Attribute- defines minimum requirements on piping reliability analysis. The application of service data should reflect the requirements of an application. Together with raw data summaries, this analysis framework enables the development of a prior and a posterior pipe rupture probability distribution. The framework supports LOCA frequency estimation, steam line break frequency estimation, as well as the development of strategies for optimized in-service inspection strategies. 63 refs, 30 tabs, 22 figs.

  12. Reliability of piping system components. Framework for estimating failure parameters from service data

    International Nuclear Information System (INIS)

    This report summarizes results and insights from the final phase of a R and D project on piping reliability sponsored by the Swedish Nuclear Power Inspectorate (SKI). The technical scope includes the development of an analysis framework for estimating piping reliability parameters from service data. The R and D has produced a large database on the operating experience with piping systems in commercial nuclear power plants worldwide. It covers the period 1970 to the present. The scope of the work emphasized pipe failures (i.e., flaws/cracks, leaks and ruptures) in light water reactors (LWRs). Pipe failures are rare events. A data reduction format was developed to ensure that homogenous data sets are prepared from scarce service data. This data reduction format distinguishes between reliability attributes and reliability influence factors. The quantitative results of the analysis of service data are in the form of conditional probabilities of pipe rupture given failures (flaws/cracks, leaks or ruptures) and frequencies of pipe failures. Finally, the R and D by SKI produced an analysis framework in support of practical applications of service data in PSA. This, multi-purpose framework, termed 'PFCA'-Pipe Failure Cause and Attribute- defines minimum requirements on piping reliability analysis. The application of service data should reflect the requirements of an application. Together with raw data summaries, this analysis framework enables the development of a prior and a posterior pipe rupture probability distribution. The framework supports LOCA frequency estimation, steam line break frequency estimation, as well as the development of strategies for optimized in-service inspection strategies

  13. Qualification methodologies for mechanical components, I and C, piping using on-site testing

    International Nuclear Information System (INIS)

    A qualification procedure shall confirm that the equipment is capable of meeting, throughout its design operational life, the requirements for performing safety functions while subject to the environmental conditions prevailing at the time of need. (The Safety of Nuclear Power Plant: Design IAEA Safety Guide). IEC 780 gives the following definition: Qualification is the generation and maintenance of evidence to ensure that the equipment will operate on demand to meet the system performance requirements. When it should be used on-site testing? Seismic design and qualification for nuclear power plants IAEA Safety Guide is requiring periodic safety review. (Maintenance throughout the design operational life.) Inspection and Testing for Acceptance IAEA Q4 Safety Guide requires that in some circumstances final acceptance of a supplied item is only possible after it has been installed. Testing methodology includes: low impedance tests (mechanical excitation impact, soil blast). Mechanical excitation: small exciters (electro-dynamic shakers or servo-hydraulic actuators) are used. Impulse technique is used for structural frequency response testing. The flowing examples are presented: in-situ dynamic test on reactor tank, in-situ dynamic test on cabinets; in-situ soil explosion test on NPP. Qualification by in-situ testing is more or less always a qualification by combination of tests and analyses. In situ testing can be performed only with low level excitation and then very special attention should be paid to non linear behaviour of structures and components. When dealing with components with bolted connections the behaviour can be very different from low and high level excitation both due to changing of restraint conditions and energy dissipating mechanisms. Piping restraints allowing thermal expansion can have a very different behaviour from low to high level excitation

  14. Assessment and management of ageing of major nuclear power plant components important to safety. Primary piping in PWRs

    International Nuclear Information System (INIS)

    guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. This report addresses the primary piping in PWRs including main coolant piping, surge and spray lines, Class 1 piping in attached systems, and small diameter piping that cannot be isolated from the primary coolant system. Maintaining the structural integrity of this piping throughout NPP service life in spite of several ageing mechanisms is essential for plant safety

  15. Characterization of cooling systems based on heat pipe principle to control operation temperature of high-tech electronic components

    OpenAIRE

    Dobre, Tănase; Stoica, Anicuţa; Iavorschi, Gustav; Pârvulescu, Oana Cristina

    2010-01-01

    Abstract The use of cooling systems based on heat pipe principle to control operation temperature of electronic components is very efficient. They have an excellent miniaturizing capacity and this fact creates adaptability for more practical situations. Starting from the observation that these cooling systems are not precisely characterized from the thermal efficiency point of view, the present paper proposes a methodology of data acquisition for their thermal characterization. An ...

  16. Evaluation of thermal embrittlement susceptibility in cast austenitic stainless steel using artificial neural network

    International Nuclear Information System (INIS)

    Cast austenitic stainless steel is used for several components, such as primary coolant piping, elbow, pump casing and valve bodies in light water reactors. These components are subject to thermal aging at the reactor operating temperature. Thermal aging results in spinodal decomposition of the delta-ferrite leading to increased strength and decreased toughness. This study shows that ferrite content can be predicted by use of the artificial neural network. The neural network has trained learning data of chemical components and ferrite contents using backpropagation learning process. The predicted results of the ferrite content using trained neural network are in good agreement with experimental ones

  17. Qualification methodologies for mechanical component, I and C, piping using test lab

    International Nuclear Information System (INIS)

    There are many methods of verifying the intensity of a structure, a function, a vibration characteristics, etc. The seismic test which verifies the function during the earthquake of a components simple substance (seismic test which checks durability according to components types). How to verify the analysis technique by the scale model and to check the intensity of plant operating conditions by the scale model results. The model of the same size as the actual plant is created and there is a method of verifying intensity and the function directly. A seismic test is restrained by the frequency of an evaluation objective, and the capability of actuator equipment, and is carried out. Moreover, otherwise, restrictions are the size of a table, actuation power, environment, etc. Here, further examples are introduced, such as evaluation by the examination that combined analysis, experimental test use and analysis, and the experimental test, and the method of proving only by test, and have the seismic check method by test learned in this lecture. Typical examples are explained. Based on the seismic test result carried out with experimental research equipment, how to verify that the required function to components, such as a structure of reactor internals, is maintained at the time of an earthquake is explained. In this case, differences of the simulation environment of the model in. a test, earthquake conditions simulated by shaker table of test conditions and actual plant conditions are important for the evaluation method determination. In nuclear equipment, there is what is required to achieve the static function to hold pressure boundary to the high temperature inside apparatus piping - high-pressure flow, and dynamic functions, such as insertion of a valve, a pump, and a control rod. Moreover, in order to maintain and carry out the safe stop of the safe operation, there is I and C for controlling - supervising these components. In order for this functional maintenance

  18. Development of ultrasonic testing techniques for welding parts of austenitic stainless. Part 2. Development of ultrasonic testing system and numerical analysis of wave propagation for field pipes

    International Nuclear Information System (INIS)

    An automatic ultrasonic testing system is developed for circumferential seam-weld inspection of pipes, in which not only advanced phased array and TOFD methods but also other conventional methods are applicable. Various specimens with slit or stress corrosion crack are used to verify the system. The verification shows (1) the detectability of the system is 100% and (2) the error between actual height and the measured is smaller than 4.4 mm, namely the system has sufficient precision. However, due to the effect of scattering at the boundaries of large crystal grains, receiving diffracted waves from crack tips is difficult. FEM analysis of wave propagation considering crystal grains is performed to investigate this effect. Numerical results show that attenuation coefficient increases and the central frequency of back echo decreases with increment in grain size, which is in good agreement with experimental ones. (author)

  19. Probabilistic procedure to evaluate integrity of degraded pipes under internal pressure and bending moment

    International Nuclear Information System (INIS)

    The determination of critical crack sizes or permissible/allowable loading levels in pipes with degraded pipe sections (circumferential cracks) for the assurance of component integrity is usually based on deterministic approaches. Therefore along with numerical calculational methods (finite element (FE) analyses) limit load calculations, such as e.g. the 'Plastic limit load concept' and the 'Flow stress concept' as well as fracture mechanics approximation methods as e.g. the R-curve method or the 'Ductile fracture handbook' and the R6-Method are currently used for practical application. Numerous experimental tests on both ferritic and austenitic pipes with different pipe dimensions were investigated at MPA Stuttgart. The geometries of the pipes were comparable to actual piping systems in Nuclear Power Plants, both BWR as well as PWR. Through wall cracks and part wall through cracks on the inside surface of the pipes were considered. The results of these tests were used to determine the flow stresses used within the limit load calculations. Therefore the deterministic concepts assessing the integrity of degraded pipes are available A new post-calculation of the above mentioned tests was performed using probabilistic approaches to assure the component integrity of degraded piping systems. As a result the calculated probability of failure was compared to experimental behaviour during the pipe test. Different reliability techniques were used for the verification of the probabilistic approaches. (author)

  20. Fabrication of mechanical components and piping design for Brazilian nuclear reactors

    International Nuclear Information System (INIS)

    The supply of Brazilian equipment and piping design for Angra 2 (and Angra 3 in some cases) have reached an advanced status in spite of the continuous outside difficulties which affect these nuclear power plants. The achieved quality is similar to the quality achieved in foreign countries and the nationalization program foreseen in 1975 is being largely surpassed. In this paper the actual situation is presented as well as the future perspectives. (Author)

  1. Mechanized radiation testing of austenitic pipe welds. Testing of media filled pipes and determination of the flaw depth by tomosynthesis; Mechanisierte Durchstrahlungspruefung von Rundschweissnaehten. Pruefung mediengefuellter Rohrleitungen und Tiefenlagenbestimmung durch Tomosynthese

    Energy Technology Data Exchange (ETDEWEB)

    Ewert, U.; Redmer, B. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany); Mueller, J. [COMPRA GmbH, Frechen (Germany); Trobitz, M. [Kernkraftwerke Gundremmingen Betriebsgesellschaft mbH, Gundremmingen (Germany); Baranov, V.A. [Institute for Introscopy, Tomsk (Russian Federation)

    1999-08-01

    A compact detection system was built for multi-angle inspection of pipes, consisting of a high-sensitivity radiometric line scanner and an ultrasonic manipulator. Improved flaw imaging quality is achieved with this system as compared to film radiography. Measurements have been carried out on site in a nuclear power plant and in a laboratory. Better flaw imaging quality was also achieved in the testing of water-filled pipes. Non-linear tomosynthesis was applied for processing and interpretation of measured data. The system delivers considerably better images of planary materials inhomogeneitites, (such as cracks and lack-of-bond defects). (orig./CB) [Deutsch] Eine hoch empfindliche radiometrische Zeilenkamera wurde mit einem Ultraschall-Manipulator zu einem Gesamtsystem aufgebaut und fuer Mehrwinkel-Inspektionen von Rohrleitungen angewandt. Bei der Inspektion von Rundschweissnaehten an Rohren mit ca. 8... 20 mm Wanddicke wurde eine Verbesserung der Bildqualitaet im Vergleich zur Filmradiographie erreicht. Diese Messungen wurden in einem Kernkraftwerk unter Vor-Ort-Bedingungen sowie im Labor ausgefuehrt. Ein signifikantes Ansteigen der Bildqualitaet wurde auch bei der Pruefung von wassergefuellten Rohren erzielt. Methoden der nicht-linearen Tomosynthese wurden fuer die Verarbeitung und Interpretation der gemessenen Projektionsdaten genutzt. Das entwickelte System gestattet eine erhebliche Verbesserung der Anzeige von planaren Materialinhomogenitaeten (z.B. Risse und Bindefehler). (orig./DGE)

  2. Authors's reply to `Generation of surface degraded layer on austenitic stainless steel piping exposed to flowing sodium in a loop: inter comparison of long term exposure data', by S. Rajendran Pillai

    Science.gov (United States)

    Ganesan, Vaidehi; Ganesan, V.; Borgstedt, H. U.

    2004-09-01

    This is an elaborate author's reply to a comment `Generation of surface degraded layer on austenitic steel piping exposed to flowing sodium in a loop: inter comparison of long term exposure data' by S. Rajendran Pillai appearing in this proceedings. The basic misunderstanding as seen in the above comment about the mass loss due to sodium exposure, which is reflected throughout the above comment, has been explained in detail in this reply for better understanding of the phenomenon. It is precisely mentioned and understood that Thorley and Tyzack model deals with complete mass loss and not mere degradation. The total mass loss corresponds to mass loss due to wall thinning and that due to degraded layer formation. Though Thorley and Tyzack model is the most pioneering model in the field of sodium corrosion, the inadequacies of this model for materials without molybdenum such as SS 304 with very long exposure in sodium is clearly brought out in this paper. This model has been successfully applied to calculate life of clad tubes, which have relatively short stay in reactor core. Yoshida models are highlighted and compared with our experimental results. Yoshida models are not valid below certain durations owing to the empirical nature of such expressions. Thorley and Tyzack model can be used for SS 316 LN as this alloy contains molybdenum and nitrogen both of which imparts corrosion resistance in sodium. What is required is that one needs to establish the extent to which this model can be applied for materials exposed to high temperatures and very long durations. The details are discussed in this reply.

  3. Improvements in 500-kHz Ultrasonic Phased-Array Probe Designs for Evaluation of Thick Section Cast Austenitic Stainless Steel Piping Welds

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, Susan L.; Cinson, Anthony D.; Moran, Traci L.; Anderson, Michael T.; Diaz, Aaron A.

    2011-02-01

    PNNL has been studying and performing confirmatory research on the inspection of piping welds in coarse-grained steels for over 30 years. More recent efforts have been the application of low frequency phased array technology to this difficult to inspect material. The evolution of 500 kHz PA probes and the associated electronics and scanning protocol are documented in this report. The basis for the probe comparisons are responses from one mechanical fatigue crack and two thermal fatigue cracks in large-bore cast mockup specimens on loan from the Electric Power Research Institution. One of the most significant improvements was seen in the use of piezo-composite elements in the later two probes instead of the piezo-ceramic material used in the prototype array. This allowed a reduction in system gain of 30 dB and greatly reduced electronic noise. The latest probe had as much as a 5 dB increase in signal to noise, adding to its flaw discrimination capability. The system electronics for the latest probe were fully optimized for a 500 kHz center frequency, however significant improvements were not observed in the center frequency of the flaw responses. With improved scanner capabilities, smaller step sizes were used, allowing both line and raster data improvements to be made with the latest probe. The small step sizes produce high resolution images that improve flaw discrimination and, along with the increased signal-to-noise ratio inherent in the latest probe design, enhanced detection of the upper regions of the flaw make depth sizing more plausible. Finally, the physical sizes of the probes were progressively decreased allowing better access to the area of interest on specimens with weld crowns, and the latest probe was designed with non-integral wedges providing flexibility in focusing on different specimen geometries.

  4. Bench-KJ: benchmark on analytical calculation of fracture mechanics parameters KI and J cracked piping components

    International Nuclear Information System (INIS)

    For many design and ageing considerations fracture mechanics is needed to evaluate cracked components. The major parameters used are K and J. For that, the different codes (RSE-M appendix 5, RCC-MRx appendix A16, R6 rule, ASME B and PV Code Section XI, API, VERLIFE, Russian Code..) propose compendia of stress intensity factors, and for some of them compendia of limit loads for usual situations, in terms of component geometry, type of defect and loading conditions. The benchmark bench-KJ, proposed in the frame of the OECD/IAGE Group, aims to compare these different estimation schemes by comparison to reference analyses done by Finite Element Method, for representative cases (pipes and elbows, mechanical or/and thermal loadings, different type and size of cracks). The objective is to have a global comparison of the procedures but also of all independent elements as stress intensity factor or reference stress. The benchmark will cover simple cases with basic mechanical loads like pressure and bending up to complex load combinations and complex geometries (cylinders and elbows) including cladding or welds: these cases are classified into 6 tasks. Twenty-nine partners are involved in this benchmark. This paper gives a short overview of the different tasks of the benchmark and presents the analysis of the results for the four first tasks, devoted on the elastic stress intensity factor calculation (task 1) and J calculation in cracked pipes (tasks 2 and 3). (authors)

  5. ADIMEW: Fracture assessment and testing of an aged dissimilar metal weld pipe assembly

    International Nuclear Information System (INIS)

    ADIMEW (Assessment of Aged Piping Dissimilar Metal Weld Integrity) was a three-year collaborative research programme carried out under the EC 5th Framework Programme. The objective of the study was to advance the understanding of the behaviour and safety assessment of defects in dissimilar metal welds between pipes representative of those found in nuclear power plant. ADIMEW studied and compared different methods for predicting the behaviour of defects located near the fusion boundaries of dissimilar metal welds typically used to join sections of austenitic and ferritic piping operating at high temperature. Assessment of such defects is complicated by issues that include: severe mis-match of yield strength of the constituent parent and weld metals, strong gradients of material properties, the presence of welding residual stresses and mixed mode loading of the defect. The study includes the measurement of material properties and residual stresses, predictive engineering analysis and validation by means of a large-scale test. The particular component studied was a 453mm diameter pipe that joins a section of type A508 Class 3 ferritic pipe to a section of type 316L austenitic pipe by means of a type 308 austenitic weld with type 308/309L buttering laid on the ferritic pipe. A circumferential, surface-breaking defect was cut using electro discharge machining into the 308L/309L weld buttering layer parallel to the fusion line. The test pipe was subjected to four-point bending to promote ductile tearing of the defect. This paper presents the results of TWI contributions to ADIMEW including: fracture toughness testing, residual stress measurements and assessments of the ADIMEW test using elastic-plastic, cracked body, finite element analysis. (orig.)

  6. Biased insert for installing data transmission components in downhole drilling pipe

    Science.gov (United States)

    Hall, David R.; Briscoe, Michael A.; Garner, Kory K.; Wilde, Tyson J.

    2007-04-10

    An apparatus for installing data transmission hardware in downhole tools includes an insert insertable into the box end or pin end of drill tool, such as a section of drill pipe. The insert typically includes a mount portion and a slide portion. A data transmission element is mounted in the slide portion of the insert. A biasing element is installed between the mount portion and the slide portion and is configured to create a bias between the slide portion and the mount portion. This biasing element is configured to compensate for varying tolerances encountered in different types of downhole tools. In selected embodiments, the biasing element is an elastomeric material, a spring, compressed gas, or a combination thereof.

  7. Probabilistic assessment of crack initiation in a piping under thermal-fatigue loading

    International Nuclear Information System (INIS)

    The European research project THERFAT concerned with an evaluation of the thermal fatigue phenomenon in mixing tees of austenite piping. Computational results obtained for conservatively selected boundary conditions revealed high stresses in a mixing tee and, correspondingly, a possibility of fatigue damage. In this paper, statistical methods are applied to properly account for the scatter and uncertainties in the input data, leading to a probabilistic assessment of the component lifetime. Results are compared with previous predictions based on a deterministic approach. (orig.)

  8. Application of the results of pipe stress analyses into fracture mechanics defect analyses for welds of nuclear piping components; Uebernahme der Ergebnisse von Rohrsystemanalysen (Spannungsanalysen) fuer bruchmechanische Fehlerbewertungen fuer Schweissnaehte an Rohrleitungsbauteilen in kerntechnischen Anlagen

    Energy Technology Data Exchange (ETDEWEB)

    Dittmar, S.; Neubrech, G.E.; Wernicke, R. [TUeV Nord SysTec GmbH und Co.KG (Germany); Rieck, D. [IGN Ingenieurgesellschaft Nord mbH und Co.KG (Germany)

    2008-07-01

    For the fracture mechanical assessment of postulated or detected crack-like defects in welds of piping systems it is necessary to know the stresses in the un-cracked component normal to the crack plane. Results of piping stress analyses may be used if these are evaluated for the locations of the welds in the piping system. Using stress enhancing factors (stress indices, stress factors) the needed stress components are calculated from the component specific sectional loads (forces and moments). For this procedure the tabulated stress enhancing factors, given in the standards (ASME Code, German KTA regulations) for determination and limitation of the effective stresses, are not always and immediately adequate for the calculation of the stress component normal to the crack plane. The contribution shows fundamental possibilities and validity limits for adoption of the results of piping system analyses for the fracture mechanical evaluation of axial and circumferential defects in welded joints, with special emphasis on typical piping system components (straight pipe, elbow, pipe fitting, T-joint). The lecture is supposed to contribute to the standardization of a code compliant and task-related use of the piping system analysis results for fracture mechanical failure assessment. [German] Fuer die bruchmechanische Bewertung von postulierten oder bei der wiederkehrenden zerstoerungsfreien Pruefung detektierten rissartigen Fehlern in Schweissnaehten von Rohrsystemen werden die Spannungen in der ungerissenen Bauteilwand senkrecht zur Rissebene benoetigt. Hierfuer koennen die Ergebnisse von Rohrsystemanalysen (Spannungsanalysen) genutzt werden, wenn sie fuer die Orte der Schweissnaehte im Rohrsystem ausgewertet werden. Mit Hilfe von Spannungserhoehungsfaktoren (Spannungsindizes, Spannungsbeiwerten) werden aus den komponentenweise berechneten Schnittlasten (Kraefte und Momente) die benoetigten Spannungskomponenten berechnet. Dabei sind jedoch die in den Regelwerken (ASME

  9. Load bearing capacity of degraded nuclear piping

    International Nuclear Information System (INIS)

    Integrity assessment of piping components with postulated cracks is important for safe and reliable operation of power plants. While various equations and methods are available for prediction of the load bearing capacity of pipes and elbows, it is very important to choose the correct equation and method whose predictions are consistent, safe but not too conservative with respect to the experimental results. Towards this goal, a comprehensive Component Integrity Assessment Program was initiated under a joint MPA-BARC collaborative program where a large number of austenitic and ferritic pipes and elbows of nominal diameter of 50-400 mm with various crack configurations and sizes were tested. These test results along with results of previous tests were analysed with various available limit load equations present and also with the R6 method. Based on the comparison of these test results and predictions, the correct equation and method are recommended to reliably predict the load bearing capacity of flawed pipes and elbows reliably. (authors)

  10. Pressure-dependent fragilities for piping components: Pilot study on Davis-Besse Nuclear Power Station

    International Nuclear Information System (INIS)

    The capacities of four, low-pressure fluid systems to withstand pressures and temperatures above the design levels were established for the Davis-Besse Nuclear Power Station. The results will be used in evaluating the probability of plant damage from Interfacing System Loss of Coolant Accidents (ISLOCA) as part of the probabilistic risk assessment of the Davis-Besse nuclear power station undertaken by EG ampersand G Idaho, Inc. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The probabilities of failure, as a function of internal pressure, are evaluated as well as the variabilities associated with them. Leak rates or leak areas are estimated for the controlling modes of failure. The pressure capacities for the pipes and vessels are evaluated using limit-state analyses for the various failure modes considered. The capacities are dependent on several factors, including the material properties, modeling assumptions, and the postulated failure criteria. The failure modes for gasketed-flange connections, valves, and pumps do not lend themselves to evaluation by conventional structural mechanics techniques and evaluation must rely primarily on the results from ongoing gasket research test programs and available vendor information and test data. 21 refs., 7 figs., 52 tabs

  11. Evaluation of material properties considering thermal embrittlement for accelerated aged CF-8M and CF-8A cast austenitic stainless steel

    International Nuclear Information System (INIS)

    Cast austenitic stainless steel have been widely used for primary coolant piping in light water reactors. This material is subject to thermal embrittlement at reactor operating temperature. CF-8M and CF-8A cast austenitic stainless steel is used for several components, such as primary coolant piping, elbow, pump casing, and valve bodies in light water reactors. Thermal embrittlement results in spinodal decomposition of delta-ferrite leading to decreased fracture toughness. In this study, the specimens were prepared using an accelerated aging method. The measurement of ferrite content, Charpy impact test and J-R test were performed to verify the predicting equation for aged material properties. In case of above 25% ferrite content, predicted result of J-R curve might be non-conservative

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

    International Nuclear Information System (INIS)

    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

  13. The use of risk based methods for establishing ISI-priorities for piping components at Oskarshamn 1 nuclear power station

    International Nuclear Information System (INIS)

    Quantitative risk evaluations are an efficient way to guide inspection priorities for primary piping components. The results of the pilot study have shown that: it is important to include a model for leak rate evaluation and detection in order to obtain realistic estimations of risk; the highest risk contributions are those components which have IGSCC and vibration fatigue as damage mechanisms at the same time; if no qualified inspection has been done before, it is important for an effective risk reduction to perform a qualified inspection as soon as possible. The current ISI-selection procedure for piping components: is efficient to select all high risk locations; also selects many low risk locations; is a very efficient starting point for making quantitative RBI-analyses. This is because it gives information of which damage mechanisms that are present and where in the plant they occur. It is possible to apply both the ASME/WOG- and the EPRI-procedure for RBI for Swedish nuclear power plants. However, if quantitative risk levels are needed, only the ASME/WOG-procedure seems to be a realistic alternative. With a quantitative RBI-analysis for the OKG-1 reactor, it is possible to combine a reduced number of inspection sites with a reduction of overall risk. This is possible due to: a shorter inspection interval is suggested for some of the high risk locations in system 315 which have both vibrations and IGSCC; a more effective inspection technique is suggested for system 354. Many low risk locations are suggested not to be included in the new ISI-selection. This means that the radiation exposure to plant personnel can be reduced and resources can be redirected to other safety related issues. The present pilot study with quantitative risk evaluations can be used to: optimise the selection of inspection locations; optimise the inspection interval; give quantitative information of the changes in risk and costs due to plant modifications, for example when: a qualified

  14. The use of risk based methods for establishing ISI-priorities for piping components at Oskarshamn 1 nuclear power station

    International Nuclear Information System (INIS)

    Quantitative risk evaluations are an efficient way to guide inspection priorities for primary piping components. The results of the pilot study have shown that: it is important to include a model for leak rate evaluation and detection in order to obtain realistic estimations of risk; the highest risk contributions are those components which have IGSCC and vibration fatigue as damage mechanisms at the same time; if no qualified inspection has been done before, it is important for an effective risk reduction to perform a qualified inspection as soon as possible. The current ISI-selection procedure for piping components: is efficient to select all high risk locations; also selects many low risk locations; is a very efficient starting point for making quantitative RBI-analyses. This is because it gives information of which damage mechanisms that are present and where in the plant they occur. It is possible to apply both the ASME/WOG- and the EPRI-procedure for RBI for Swedish nuclear power plants. However, if quantitative risk levels are needed, only the ASME/WOG-procedure seems to be a realistic alternative. With a quantitative RBI-analysis for the O1-reactor, it is possible to combine a reduced number of inspection sites with a reduction of overall risk. This is possible due to: a shorter inspection interval is suggested for some of the high risk locations in system 315 which have both vibrations and IGSCC; a more effective inspection technique is suggested for system 354. Many low risk locations are suggested not to be included in the new ISI-selection. This means that the radiation exposure to plant personnel can be reduced and resources can be redirected to other safety related issues. The present pilot study with quantitative risk evaluations can be used to: optimise the selection of inspection locations; optimise the inspection interval; give quantitative information of the changes in risk and costs due to plant modifications, for example when: a qualified

  15. Applicability of Equivalent Static Method to seismic response of piping and other components

    International Nuclear Information System (INIS)

    The Equivalent Static Method (ESM) is a simple and cost effective approach in the design of systems and components subjected to seismic loads. However, its applicability is restricted to systems which can be represented by a ''simple model.'' In this paper the restriction to a simple model is examined using the example of a propped cantilever, for which some codes or standards explicitly state that ESM is not applicable. By comparing ESM results for the propped cantilever with those for a regular (un-propped) cantilever, it is found that ESM can conditionally be applied to the propped cantilever configuration

  16. Applicability of equivalent static method to seismic response of piping and other components

    International Nuclear Information System (INIS)

    The Equivalent Static Method (ESM) is a simple and cost effective approach in the design of systems and components subjected to seismic loads. However, its applicability is restricted to systems which can be represented by a simple model. In this paper the restriction to a simple model is examined using the example of a propped cantilever, for which some codes or standards explicitly state that ESM is not applicable. By comparing ESM results for the propped cantilever with those for a regular (un-propped) cantilever, it is found that ESM can conditionally be applied to the propped cantilever configuration

  17. Cold crack formation during the welding of thick-walled structural components. The part of hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Cerjak, H.; Schmidt, J.

    1987-01-01

    Outward appearance of cold cracks - characteristics of cold crack formation - cold cracks under austenitic or nickel-base cladding - investigations on pipe-end bottoms of steam boilers in relation to the forging technology, the welding process and the annealing state of the component - the influence of preheating before welding - behaviour of hydrogen - comparison of tests on samples from original components - influence of segregations on cold crack formation - remedial measures.

  18. Ice plugging of pipes using liquid nitrogen

    International Nuclear Information System (INIS)

    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

  19. Creep Behavior at 1273 K (1000 °C) in Nb-Bearing Austenitic Heat-Resistant Cast Steels Developed for Exhaust Component Applications

    Science.gov (United States)

    Zhang, Yinhui; Li, Mei; Godlewski, Larry A.; Zindel, Jacob W.; Feng, Qiang

    2016-05-01

    A series of Nb-bearing austenitic heat-resistant cast steels with variations of N/C ratios were investigated, and the morphological change of Nb(C,N) from faceted blocks, mixed flake-blocks to "Chinese-script" was observed as N/C ratios decreased. The creep behavior of these alloys was studied at 1273 K (1000 °C), and the longest creep life and lowest creep rate occurred in model alloys with script Nb(C,N). Residual δ-ferrites and (Cr,Fe)23C6 were adverse to creep properties. This work indicates that the control of N/C ratio is required for the as-cast microstructural strengthening.

  20. Creep Behavior at 1273 K (1000 °C) in Nb-Bearing Austenitic Heat-Resistant Cast Steels Developed for Exhaust Component Applications

    Science.gov (United States)

    Zhang, Yinhui; Li, Mei; Godlewski, Larry A.; Zindel, Jacob W.; Feng, Qiang

    2016-07-01

    A series of Nb-bearing austenitic heat-resistant cast steels with variations of N/C ratios were investigated, and the morphological change of Nb(C,N) from faceted blocks, mixed flake-blocks to "Chinese-script" was observed as N/C ratios decreased. The creep behavior of these alloys was studied at 1273 K (1000 °C), and the longest creep life and lowest creep rate occurred in model alloys with script Nb(C,N). Residual δ-ferrites and (Cr,Fe)23C6 were adverse to creep properties. This work indicates that the control of N/C ratio is required for the as-cast microstructural strengthening.

  1. Quantitative evaluation of ultrasonic wave propagation in inhomogeneous anisotropic austenitic welds using 3D ray tracing method. Numerical and experimental validation

    International Nuclear Information System (INIS)

    Austenitic welds and dissimilar welds are extensively used in primary circuit pipes and pressure vessels in nuclear power plants, chemical industries and fossil fuelled power plants because of their high fracture toughness, resistance to corrosion and creep at elevated temperatures. However, cracks may initiate in these weld materials during fabrication process or stress operations in service. Thus, it is very important to evaluate the structural integrity of these materials using highly reliable non-destructive testing (NDT) methods. Ultrasonic non-destructive inspection of austenitic welds and dissimilar weld components is complicated because of anisotropic columnar grain structure leading to beam splitting and beam deflection. Simulation tools play an important role in developing advanced reliable ultrasonic testing (UT) techniques and optimizing experimental parameters for inspection of austenitic welds and dissimilar weld components. The main aim of the thesis is to develop a 3D ray tracing model for quantitative evaluation of ultrasonic wave propagation in an inhomogeneous anisotropic austenitic weld material. Inhomogenity in the anisotropic weld material is represented by discretizing into several homogeneous layers. According to ray tracing model, ultrasonic ray paths are traced during its energy propagation through various discretized layers of the material and at each interface the problem of reflection and transmission is solved. The influence of anisotropy on ultrasonic reflection and transmission behaviour in an anisotropic austenitic weld material are quantitatively analyzed in three dimensions. The ultrasonic beam directivity in columnar grained austenitic steel material is determined three dimensionally using Lamb's reciprocity theorem. The developed ray tracing model evaluates the transducer excited ultrasonic fields accurately by taking into account the directivity of the transducer, divergence of the ray bundle, density of rays and phase

  2. Failure of austenitic stainless steel tubes during steam generator operation

    OpenAIRE

    M. Głowacka; J. Łabanowski; S. Topolska

    2012-01-01

    Purpose: of this study is to analyze the causes of premature failure of steam generator coil made of austenitic stainless steel. Special attention is paid to corrosion damage processes within the welded joints.Design/methodology/approach: Examinations were conducted several segments of the coil made of seamless cold-formed pipes Ø 23x2.3 mm, of austenitic stainless steel grade X6CrNiTi18-10 according to EN 10088-1:2007. The working time of the device was 6 months. The reason for the withdrawa...

  3. Mathematical modelling and numerical simulation of two-phase multi-component flows of CO₂ mixtures in pipes

    OpenAIRE

    Morin, Alexandre

    2012-01-01

    In this thesis, the modelling of one-dimensional two-phase flows is studied, as well as the associated numerical methods. The background for this study is the need for numerical tools to simulate fast transients in pressurised carbon dioxide pipelines, amongst other things the crack arrest problem. This is a coupled mechanical and fluid-dynamical problem, where the pressurised gas causes a crack to propagate along the pipe, while being depressurised to the atmosphere. The crack stops when the...

  4. Reliability of piping system components. Volume 2: PSA LOCA data base. Review of methods for LOCA evaluation since the WASH-1400

    Energy Technology Data Exchange (ETDEWEB)

    Nyman, R.; Erixon, S. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tomic, B. [ENCONET Consulting GmbH, Vienna (Austria); Lydell, B. [RSA Technologies, Visat, CA (United States)

    1996-09-01

    The Swedish Nuclear Power Inspectorate has undertaken a project to establish a comprehensive passive components database, validate failure rate parameter estimates and model framework for enhancement of integrating passive components failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure. Approx. 2300 failure events allowed for data exploration in Phase 2 to develop a sound basis for PSA treatment of piping system failure. In addition, a comprehensive review of the current consideration of LOCA in PSA and of all available literature in this area was undertaken. This report is devoted to identification of treatment of LOCA in PSAs. The report contains a detailed review of many programs and dozens of specific PSA studies for different reactor types. This collection and analysis of information together with information for the relational database was used to develop a matrix approach on contribution to LOCA events from different components which are part of the reactor coolant system pressure boundary. The overall conclusion of the work is that although there are some further developments in this area, there is still no significant enhancement of ways how LOCA are considered in PSAs as compared to the mid 70s, only selected studies attempted to address LOCAs in a more comprehensive way. Later phases of this project are expected to contribute to enhancement of treatment of LOCA events in PSA studies. 54 refs, 25 tabs.

  5. Reliability of piping system components. Volume 2: PSA LOCA data base. Review of methods for LOCA evaluation since the WASH-1400

    International Nuclear Information System (INIS)

    The Swedish Nuclear Power Inspectorate has undertaken a project to establish a comprehensive passive components database, validate failure rate parameter estimates and model framework for enhancement of integrating passive components failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure. Approx. 2300 failure events allowed for data exploration in Phase 2 to develop a sound basis for PSA treatment of piping system failure. In addition, a comprehensive review of the current consideration of LOCA in PSA and of all available literature in this area was undertaken. This report is devoted to identification of treatment of LOCA in PSAs. The report contains a detailed review of many programs and dozens of specific PSA studies for different reactor types. This collection and analysis of information together with information for the relational database was used to develop a matrix approach on contribution to LOCA events from different components which are part of the reactor coolant system pressure boundary. The overall conclusion of the work is that although there are some further developments in this area, there is still no significant enhancement of ways how LOCA are considered in PSAs as compared to the mid 70s, only selected studies attempted to address LOCAs in a more comprehensive way. Later phases of this project are expected to contribute to enhancement of treatment of LOCA events in PSA studies. 54 refs, 25 tabs

  6. Investigation of special-purpose processors for real-time synthetic-aperture-focusing techniques for nondestructive evaluation of nuclear-reactor vessels and piping components

    International Nuclear Information System (INIS)

    Progress in the development of a special purpose system for use in a real-time in-service inspection system for reactor vessels and piping components is described in this report. An analysis of the synthetic aperture processing algorithm is presented and new methods of speedup are described. A number of special purpose processor architectures are presented and two of the more promising ones are described in detail and are compared and evaluated. Proposed specifications for an initial field inspection system are presented. A brief description of the capabilities of a laboratory prototype processor (to be fabricated) is given

  7. Carbon Concentration of Austenite

    Directory of Open Access Journals (Sweden)

    Z. Ławrynowicz

    2007-07-01

    Full Text Available The investigation was carried out to examine the influence of temperature and times of austempering process on the maximum extend towhich the bainite reaction can proceed and the carbon content in retained austenite. It should be noted that a small percentage change in theaustenite carbon content can have a significant effect on the subsequent austempering reaction changing the volume fraction of the phasespresent and hence, the resulting mechanical properties. Specimens were prepared from an unalloyed ductile cast iron, austenitised at 950oCfor 60 minutes and austempered by the conventional single-step austempering process at four temperatures between BS and MS, eg., 250,300, 350 and 400oC. The samples were austempered at these temperatures for 15, 30, 60, 120 and 240 minutes and finally quenched toambient temperature. Volume fractions of retained austenite and carbon concentration in the residual austenite have been observed byusing X-ray diffraction. Additionally, carbon concentration in the residual austenite was calculated using volume fraction data of austeniteand a model developed by Bhadeshia based on the McLellan and Dunn quasi-chemical thermodynamic model. The comparison ofexperimental data with the T0, T0' and Ae3' phase boundaries suggests the likely mechanism of bainite reaction in cast iron is displacive rather than diffusional. The carbon concentration in retained austenite demonstrates that at the end of bainite reaction the microstructure must consist of not only ausferrite but additionally precipitated carbides.

  8. In-service inspection - a vital role in monitoring and health assessment of nuclear pressure vessels, piping and components at Tarapur Atomic Power Station - 1 and 2

    International Nuclear Information System (INIS)

    Tarapur Atomic Power Station is a twin unit Boiling Water Reactors (BWRs) built in 1960's and presently operating at 160MWe. TAPS has completed 36 years of successful commercial operation and is continuing to provide safe, economic and reliable power supply. The design life of Tarapur nuclear reactors is 40 Effective Full Power Years (EFPY). So far TAPS has completed about 20EFPY for each reactor. In order to estimate the healthiness of nuclear components, a comprehensive study was made by the station in consultation with design group of Nuclear Power Corporation of India Ltd.. In-Service Inspection (ISI) substantially enhances confidence in component performance. Consolidated inspection early in service life provides greater assurance of component's integrity. Periodic in-service inspection provides vital information in the form of flaw characterization for assessment of structural integrity. This paper describes various degradation mechanisms (SCC, IGSCC, TGSCC, EC, FAC etc.,) identified for critical components, their method of detection, methodologies followed for In-Service inspection and developmental activities to assess the integrity of nuclear reactor vessels, piping and components for continued service. Also a comprehensive examination carried out on Structures, Systems and Components (SSCs) as part of plant ageing management programme is also discussed. (author)

  9. Characterization of microstructure and texture across dissimilar super duplex/austenitic stainless steel weldment joint by austenitic filler metal

    Energy Technology Data Exchange (ETDEWEB)

    Eghlimi, Abbas, E-mail: a.eghlimi@ma.iut.ac.ir [Department of Materials Engineering, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Shamanian, Morteza [Department of Materials Engineering, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Eskandarian, Masoomeh [Department of Materials Engineering, Shiraz University, Shiraz 71348-51154 (Iran, Islamic Republic of); Zabolian, Azam [Department of Natural Resources, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Szpunar, Jerzy A. [Department of Mechanical Engineering, University of Saskatchewan, Saskatoon, SK S7N 5A9 (Canada)

    2015-08-15

    The evolution of microstructure and texture across an as-welded dissimilar UNS S32750 super duplex/UNS S30403 austenitic stainless steel joint welded by UNS S30986 (AWS A5.9 ER309LMo) austenitic stainless steel filler metal using gas tungsten arc welding process was evaluated by optical micrography and EBSD techniques. Due to their fabrication through rolling process, both parent metals had texture components resulted from deformation and recrystallization. The weld metal showed the highest amount of residual strain and had large austenite grain colonies of similar orientations with little amounts of skeletal ferrite, both oriented preferentially in the < 001 > direction with cub-on-cube orientation relationship. While the super duplex stainless steel's heat affected zone contained higher ferrite than its parent metal, an excessive grain growth was observed at the austenitic stainless steel's counterpart. At both heat affected zones, austenite underwent some recrystallization and formed twin boundaries which led to an increase in the fraction of high angle boundaries as compared with the respective base metals. These regions showed the least amount of residual strain and highest amount of recrystallized austenite grains. Due to the static recrystallization, the fraction of low degree of fit (Σ) coincident site lattice boundaries, especially Σ3 boundaries, was increased in the austenitic stainless steel heat affected zone, while the formation of subgrains in the ferrite phase increased the content of < 5° low angle boundaries at that of the super duplex stainless steel. - Graphical abstract: Display Omitted - Highlights: • Extensive grain growth in the HAZ of austenitic stainless steel was observed. • Intensification of < 100 > orientated grains was observed adjacent to both fusion lines. • Annealing twins with Σ3 CSL boundaries were formed in the austenite of both HAZ. • Cub-on-cube OR was observed between austenite and ferrite in the weld

  10. Characterization of microstructure and texture across dissimilar super duplex/austenitic stainless steel weldment joint by austenitic filler metal

    International Nuclear Information System (INIS)

    The evolution of microstructure and texture across an as-welded dissimilar UNS S32750 super duplex/UNS S30403 austenitic stainless steel joint welded by UNS S30986 (AWS A5.9 ER309LMo) austenitic stainless steel filler metal using gas tungsten arc welding process was evaluated by optical micrography and EBSD techniques. Due to their fabrication through rolling process, both parent metals had texture components resulted from deformation and recrystallization. The weld metal showed the highest amount of residual strain and had large austenite grain colonies of similar orientations with little amounts of skeletal ferrite, both oriented preferentially in the < 001 > direction with cub-on-cube orientation relationship. While the super duplex stainless steel's heat affected zone contained higher ferrite than its parent metal, an excessive grain growth was observed at the austenitic stainless steel's counterpart. At both heat affected zones, austenite underwent some recrystallization and formed twin boundaries which led to an increase in the fraction of high angle boundaries as compared with the respective base metals. These regions showed the least amount of residual strain and highest amount of recrystallized austenite grains. Due to the static recrystallization, the fraction of low degree of fit (Σ) coincident site lattice boundaries, especially Σ3 boundaries, was increased in the austenitic stainless steel heat affected zone, while the formation of subgrains in the ferrite phase increased the content of < 5° low angle boundaries at that of the super duplex stainless steel. - Graphical abstract: Display Omitted - Highlights: • Extensive grain growth in the HAZ of austenitic stainless steel was observed. • Intensification of < 100 > orientated grains was observed adjacent to both fusion lines. • Annealing twins with Σ3 CSL boundaries were formed in the austenite of both HAZ. • Cub-on-cube OR was observed between austenite and ferrite in the weld

  11. Residual stress determination in a dissimilar weld overlay pipe by neutron diffraction

    International Nuclear Information System (INIS)

    Highlights: → Determined residual stress distribution in a dissimilar weld overlay pipe. → Consists of a ferritic (SA508), austenitic (F316L) steels, Alloy 182 consumable. → Measured significant compression (-600 MPa) near the inner wall of overlay. → Validate integrity of the inner wall for the pressurized nozzle nuclear structure. - Abstract: Residual stresses were determined through the thickness of a dissimilar weld overlay pipe using neutron diffraction. The specimen has a complex joining structure consisting of a ferritic steel (SA508), austenitic steel (F316L), Ni-based consumable (Alloy 182), and overlay of Ni-base superalloy (Alloy 52M). It simulates pressurized nozzle components, which have been a critical issue under the severe crack condition of nuclear power reactors. Two neutron diffractometers with different spatial resolutions have been utilized on the identical specimen for comparison. The macroscopic 'stress-free' lattice spacing (do) was also obtained from both using a 2-mm width comb-like coupon. The results show significant changes in residual stresses from tension (300-400 MPa) to compression (-600 MPa) through the thickness of the dissimilar weld overlay pipe specimen.

  12. Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Energy Technology

    2002-08-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life ({var_epsilon}-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of fatigue crack initiation in austenitic stainless steels in LWR coolant environments. The existing fatigue {var_epsilon}-N data have been evaluated to establish the effects of key material, loading, and environmental parameters (such as steel type, strain range, strain rate, temperature, dissolved-oxygen level in water, and flow rate) on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. The influence of reactor environments on the mechanism of fatigue crack initiation in these steels is also discussed.

  13. Fracture mechanical analyses concerning the load-bearing capacity of dissimilar welds in piping with circumferential cracks

    International Nuclear Information System (INIS)

    In power plants numerous dissimilar welds exist between ferritic and austenitic components. The evaluation of the influence of real or postulated flaws to the load bearing and failure behaviour in the area of dissimilar welded joints, e. g. in piping of the primary circuit of nuclear power plants, is an essential part of the safety assessment of these plants. The fracture assessment of dissimilar welds containing flaws is therefore an important issue of structural integrity considerations. During the past decades several assessment procedures for the evaluation of flaws in welds have been developed mainly based on methods originally developed for homogenous materials. The main difficulties in applying these methods are that they are restricted to two different materials and the very limited experimental evaluation especially by component tests. To improve the general understanding of the influence of flaws in piping systems with dissimilar welds and to have a more reliable basis for the assessment of the fracture behaviour, full scale tests on pipes containing a weld between ferritic and austenitic steel have been performed. The test results of these tests are summarised and used for the evaluation of analytical and numerical assessment procedures. (orig.)

  14. Long term creep properties and microstructural evolution of ferritic and austenitic grades for USC power plants

    Energy Technology Data Exchange (ETDEWEB)

    Caminada, S.; Cumino, G. [Tenaris, Dalmine (BG) (Italy); Cipolla, L.; Di Gianfrancesco, A. [Centro Sviluppo Materiali SpA, Material and Product Directorate, Rome (Italy); Minami, Y.; Ono, T. [TenarisNKKt, R and D, Kawasaki, Kanagawa (Japan)

    2007-07-01

    The steam parameters in the new high efficiency fossil fuel power plants are continuously increasing, requiring new advanced materials with enhanced creep strength able to operate on the most severe temperature and pressure conditions. Tenaris focused on the development of ferritic-martensitic and austenitic grades for tubes and pipes applications. The product development in TenarisDalmine for the ferritic-martensitic grades has been focused on: low alloyed ASTM Grade 23 as substitute of Grade 22 for components operating at relatively low temperatures, containing 1.5% W and with quite good weldability and creep properties up to 580 C and a competitive cost; high alloyed ASTM Grade 92, an improved version of the well known Grade 91 for the superheaters, headers and other parts of the boiler operating at temperatures up to 620 C: its tempered martensitic structure offers very high creep strength and long term stability. The product development in TenarisNKKt R and D on austenitic grades has been focused on: TEMPALOY AA-1 as improved version of 18Cr8NiNbTi with the 3%Cu, showing high creep and corrosion properties, TEMPALOY A-3: a 20Cr-15Ni-Nb-N showing good creep behaviour and corrosion properties better than AA-1 due to the higher Cr content. This paper describes the Tenaris products, the process routes and the main characteristics of these steels, including the effect of shot blasting on steam oxidation properties of the austenitic grades, as well as, the R and D activities in the field of alloy design, creep tests, data assessment, microstructural analysis and damage modelling, conducted with the support of the Centro Sviluppo Materiali. (orig.)

  15. Tritium in austenitic stainless steel vessels

    International Nuclear Information System (INIS)

    Austenitic stainless steels are normally recommended for components of hydrogen-handling equipment in applications where high in-service reliability is required. The literature leading to this recommendation is reviewed, and it is shown that AISI Type 316L stainless is particularly suitable for use in tritium-handling and storage systems. When made of this steel, the storage vessels will be extremely resistant to any degradation from tritium in both routine and accident conditions. (author)

  16. Results of VGB research work with respect to operation of BWR pipes made of austenitic SS; Ergebnisse des VGB-Forschungsvorhabens zur Absicherung des Betriebsverhaltens austenitischer Staehle in SWR-Rohrleitungen

    Energy Technology Data Exchange (ETDEWEB)

    Kilian, R. [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Bruemmer, G. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany)

    1998-11-01

    The VGB research project was to examine and characterize various, operation-induced impacts on the crack formation in stabilized austenitic steels, caused by intercrystalline stress corrosion cracking as a result of sensitization after chromium depletion at the grain boundaries. The results of this project as well as available operating experience show that the measures taken so far for the future operation of the German BWR plants, for avoiding in these plants intercrystalline stress corrosion cracking, correspond to the state of the art and achieve the wanted purpose. These measures are: use of optimized material W-No. 1.4550 with reduced carbon contents; use of optimized welding techniques for reducing the heat input and the welding shrinkage (cold deformation.); optimized preparation of welding work in order to avoid shape defects during welding (eg. edge misalignment, defective mash welds); reduction of tensile stresses occurring during welding; compliance with the recent VGB water chemistry code. (orig./CB) [Deutsch] Das VGB-Forschungsvorhaben sollte verschieden gelagerte Einfluesse auf die Rissbildung im Betrieb von stabilisierten austenitischen Staehlen, verursacht durch interkristalline Spannungsrisskorrosion infolge Sensbilisierung durch Chromverarmung an den Korngrenzen, systematisch erfassen. Aus den Forschungsergebnissen dieses VGB-Programms sowie den bisher vorliegenden Betriebserfahrungen ist festzuhalten, dass die bisher durchgefuehrten Massnahmen fuer den zukuenftigen Betrieb der deutschen SWR-Anlagen zur Vermeidung von interkristalliner Spannungsrisskorrosion zielgerichtet waren und dem heutigen Wissensstand entsprechen. Diese Massnahmen sind: 1. Einsatz von optimiertem Werkstoff W.-Nr. 1.4550 mit abgesenktem Kohlenstoffgehalt; 2. Einsatz von optimierten Schweissverfahren zur Verminderung der Waermeeinbringung und zur Verringerung des Schweissschrumpfes (Kaltverformung.); 3. Durchfuehrung einer optimierten Schweissnahtvorbereitung zur Vermeidung

  17. Fracture behavior evaluations for ferritic steel piping with circumferential double flaws on the inner surface

    International Nuclear Information System (INIS)

    Methods for assessing the structural integrity of nuclear components having some flaws are provided in the Rules on Fitness-for-Service for Nuclear Power Plants of the JSME code (JSME FFS code). Although the JSME FFS code provides such methods for piping with a single flaw, it does not describe any method for fracture assessment of piping with multiple flaws including flaw coalescence criteria. Some investigations on the fracture behavior of mainly austenitic stainless steel piping with multiple flaws, whose fracture mode is plastic collapse, have recently been reported and fracture assessment methods have been proposed. In the present study, fracture tests and analyses of carbon steel piping with a single and two circumferential flaw(s) on the inner surface were conducted to investigate a method for fracture assessment of ferritic steel piping with multiple flaws. It was found that fracture assessment based on the twice elastic slope method and the plastic collapse mechanism gave inadequate results for a large single flaw. Including this case, two kinds of elastic–plastic fracture assessment method, one using the Z-factor in the JSME FFS code and the other by ductile instability analysis, gave conservative estimates of fracture strength even when the structural factor SF was not considered (i.e. SF = 1)

  18. A one-dimensional mathematical model of multi-component fluid flow in pipes and its application to rapid decompression in dry natural gas mixtures

    International Nuclear Information System (INIS)

    The paper presents a one-dimensional transient mathematical model of compressible thermal multi-component gas mixture flow in a shock tube. The set of mass, momentum and enthalpy conservation equations is solved for the gas phase. Thermo-physical properties of multi-component natural gas mixture are calculated by solving the Equation of State (EOS) in the form of the Soave-Redlich-Kwong (SRK-EOS) model. The proposed mathematical model is validated against the experiments where the decompression wave speed in dry natural gases was measured at low temperatures and shows a good agreement with the experimental data at high and low initial pressure. The effect of the initial temperature on rapid decompression process is investigated numerically using the proposed model. Numerical results show that the proposed model simulates the decompression in natural gases much better and accurate than other models, and shows a great potential because it can be extended on the case of gas–liquid two-phase flow in a shock tube. Highlights: ► 1D transient mathematical model of thermal multi-component gas mixture pipe flows is developed. ► The model is validated on the experiments on RGD in dry natural gases. ► Numerical analysis on RGD in dry natural gas mixtures is performed. ► Predictions fit to the experiments much better than any other models

  19. Failure of austenitic stainless steel tubes during steam generator operation

    Directory of Open Access Journals (Sweden)

    M. Głowacka

    2012-12-01

    Full Text Available Purpose: of this study is to analyze the causes of premature failure of steam generator coil made of austenitic stainless steel. Special attention is paid to corrosion damage processes within the welded joints.Design/methodology/approach: Examinations were conducted several segments of the coil made of seamless cold-formed pipes Ø 23x2.3 mm, of austenitic stainless steel grade X6CrNiTi18-10 according to EN 10088-1:2007. The working time of the device was 6 months. The reason for the withdrawal of the generator from the operation was leaks in the coil tube caused by corrosion damage. The metallographic investigations were performed with the use of light microscope and scanning electron microscope equipped with the EDX analysis attachment.Findings: Examinations of coil tubes indicated severe corrosion damages as pitting corrosion, stress corrosion cracking, and intergranular corrosion within base material and welded joints. Causes of corrosion was defined as wrong choice of austenitic steel grade, improper welding technology, lack of quality control of water supply and lack of surface treatment of stainless steel pipes.Research limitations/implications: It was not known the quality of water supply of steam generator and this was the reason for some problems in the identification of corrosion processes.Practical implications: Based on the obtained research results and literature studies some recommendations were formulated in order to avoid failures in the application of austenitic steels in the steam generators. These recommendations relate to the selection of materials, processing technology and working environment.Originality/value: Article clearly shows that attempts to increase the life time of evaporator tubes and steam coils by replacing non-alloy or low alloy structural steel by austenitic steel, without regard to restrictions on its use, in practice often fail.

  20. Finite element thermal analysis of the fusion welding of a P92 steel pipe

    OpenAIRE

    Yaghi, A. H.; Tanner, D. W. J.; Hyde, T.H.; A. A. Becker; Sun, W.

    2012-01-01

    Fusion welding is common in steel pipeline construction in fossil-fuel power generation plants. Steel pipes in service carry steam at high temperature and pressure, undergoing creep during years of service; their integrity is critical for the safe operation of a plant. The high-grade martensitic P92 steel is suitable for plant pipes for its enhanced creep strength. P92 steel pipes are usually joined together with a similar weld metal. Martensitic pipes are sometimes joined to austenitic steel...

  1. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    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)

  2. Ultrasonic testing of austenitic stainless steel welds

    International Nuclear Information System (INIS)

    Ultrasonic testing of austenitic stainless steel welds has been considered difficult because of the high noise level and remarkable attenuation of ultrasonic waves. To improve flaw detectability in this kind of steel, various inspection techniques have been studied. A series of tests indicated: (1) The longitudinal angle beam transducers newly developed during this study can detect 4.8 mm dia. side drilled holes in dissimilar metal welds (refraction angle: 550 from SUS side, 450 from CS side) and in cast stainless steel welds (refraction angle: 450, inspection frequency: 1 MHz). (2) Cracks more than 5% t in depth in the heat affected zones of fine-grain stainless steel pipe welds can be detected by the 450 shear wave angle beam method (inspection frequency: 2 MHz). (3) The pattern recognition method using frequency analysis technology was presumed useful for discriminating crack signals from spurious echoes. (author)

  3. Expanded austenite in nitrided layers deposited on austenitic and super austenitic stainless steel grades

    International Nuclear Information System (INIS)

    In this work nitrided layers deposited on austenitic and super austenitic stainless steels were analyzed through optical microscopy and X-rays diffraction analysis (XRD). It was observed that the formation of N supersaturated phase, called expanded austenite, has promoted significant increment of hardness (> 1000HV). XRD results have indicated the anomalous displacement of the diffracted peaks, in comparison with the normal austenite. This behavior, combined with peaks broadening, it was analyzed in different nitriding temperatures which results showed good agreement with the literature. (author)

  4. Austenitic stainless steels with cryogenic resistance

    International Nuclear Information System (INIS)

    The most used austenitic stainless steels are alloyed with chromium and nickel and have a reduced carbon content, usually lower than 0.1 % what ensures corresponding properties for processing by plastic deformation at welding, corrosion resistance in aggressive environment and toughness at low temperatures. Steels of this kind alloyed with manganese are also used to reduce the nickel content. By alloying with manganese which is a gammageneous element one ensures the stability of austenites. Being cheaper these steels may be used extensively for components and equipment used in cryogenics field. The best results were obtained with steels of second group, AMnNi, in which the designed chemical composition was achieved, i.e. the partial replacement of nickel by manganese ensured the toughness at cryogenic temperatures. If these steels are supplementary alloyed, their strength properties may increase to the detriment of plasticity and toughness, although the cryogenic character is preserved

  5. Narrow gap HST welding process and its application to candidate pipe material for 700 C USC boiler component

    Energy Technology Data Exchange (ETDEWEB)

    Bao, Gang; Sato, Takashi; Fukuda, Yuji [Babcock-Hitachi K.K., Hiroshima (Japan). Kure Research Lab.; Mitsuhata, Koichi [Babcock-Hitachi K.K., Hiroshima (Japan). Kure Div.

    2008-07-01

    Increasing steam temperature and pressure conditions of 700 C USC (Ultra Super Critical) power plants under consideration require the adoption of Ni-based alloys. One of the most crucial issues for the application of 700 C USC power plants is the establishment of welding technology for the thick-walled components. This paper reports the research results on the practicability of candidate material for the thickwalled components. The weld test was conducted on Ni-based Alloy617 (52Ni-22Cr- 13Co-9Mo-Ti-Al) by using the narrow gap HST (Hot wire Switching TIG) welding process developed by Babcock-Hitachi K.K with the matching filler wire of Alloy617. The weldability and strength properties of weld joint were examined. The sound weld joint was achieved. The advantages of narrow gap HST welding process for the thick-walled components of Ni-based alloy were discussed from the viewpoints of weld metal chemical composition and creep rupture strength. Due to the good shielding effect, the melting loss of alloy elements in the weld consumable during the narrow gap HST welding procedure was suppressed successfully. The narrow gap HST weld joint showed comparable strength with the parent metal. (orig.)

  6. Applicability of fatigue life reduction factor in design analyses of PWR Primary components considering effects of reactor coolant environment

    International Nuclear Information System (INIS)

    This paper investigates applicability of the USNRC Regulatory Guide (RG) 1.207 for new reactors for evaluating fatigue analyses incorporating the life reduction of metal components due to the effects of the light water reactor environment. Sample fatigue evaluations for critical parts of the reactor coolant system (RCS) component and piping are conducted using the environmental factor method as described in RG 1.207. Results of the fatigue evaluations are presented and discussed on the application of the regulatory guide. It if found that the fatigue design of the rector pressure vessel will be able to meet the requirements of Regulatory Guide 1.207, but the surge line piping can not. In order to comply with the regulatory guide on the environmental fatigue for new reactors, design analysis methodologies need be improved for the austenitic stainless steel piping in particular

  7. The PISC parametric study on the effect of cast austenitic steel texture on the capability of ultrasonic examination

    International Nuclear Information System (INIS)

    Within the framework of Action 4 (Austenitic Steel Testing) of PISC III (Programme for the Inspection of Steel Components), a parametric study was carried out on a set of centrifugally cast stainless steel samples, representative of the main coolant piping of pressurized water nuclear reactors. The samples are obtained from different manufacturers, and feature various grain textures and dimensions. Artificial and realistic flaws were used to assess the detection and sizing capability of ultrasonic examination techniques. The paper analyzes the data as a function of the metal structure and of the main parameters of the testing techniques, which include TRL contact probes and immersion focusing transducers. Guidelines are deduced as to the selection of inspection techniques, in relation with the metallurgical texture of each specimen. In addition, the influence of the presence of a weld across the wave path is evaluated, as well as the similarity between the responses obtained from crack-like machined reflectors and mechanical fatigue cracks

  8. Fracture mechanics environmental degradation assessment of nuclear pressure vessel and piping materials

    International Nuclear Information System (INIS)

    A new set of fracture mechanics stress corrosion crack growth data is presented for transgranular cracking of low alloy steels used in various nuclear components as well as intergranular cracking of stabilized austenitic stainless steels used primarily for nuclear piping. The essential observations are as follows: In low alloy steels, fast stress corrosion crack growth rates between 10-9 and 10-8 m/s may be observed down to 400 ppb dissolved oxygen at water conductivities of 0.5 μS/cm in refreshed autoclaves. However, there is an indication, substantiated by only few data, that at lower conductivities and/or lower oxygen concentrations the crack growth rates in ferritic steels come down to about 3 x 10-11 m/s. This crack velocity has also been found to be typical for stabilized austenitic stainless steels exposed to faulted simulated BWR water. Crack growth at this rate is observed in mill annealed (fine grained) as well as in sensitized or coarse grained or cold worked stabilized austenitic stainless steels. On average, the crack growth rates measured in the laboratory by using fracture mechanics methods correlate well with stress corrosion service experience of the steels discussed here

  9. Residual stress studies of austenitic and ferritic steels

    International Nuclear Information System (INIS)

    Residual studies have been made on austenitic and ferritic steels of the types used as structural materials. The residual stress results presented here will include residual stress measurements in the heat-affected zone on butt welded Type 304 stainless steel pipes, and the stresses induced in Type 304 austenitic stainless steel and Type A508 ferritic steel by several surface preparations. Such surface preparation procedures as machining and grinding can induce large directionality effects in the residual stresses determined by X-ray techniques and some typical data will be presented. A brief description is given of the mobile X-ray residual stress apparatus used to obtain most of the data in these studies. (author)

  10. Aging and service wear of hydraulic and mechanical snubbers used on safety-related piping and components of nuclear power plants. Phase I study

    International Nuclear Information System (INIS)

    This report presents an overview of hydraulic and mechanical snubbers used on nuclear piping systems and components, based on information from the literature and other sources. The functions and functional requirements of snubbers are discussed. The real versus perceived need for snubbers is reviewed, based primarily on studies conducted by a Pressure Vessel Research Committee. Tests conducted to qualify snubbers, to accept them on a case-by-case basis, and to establish their fitness for continued operation are reviewed. This report had two primary purposes. The first was to assess the effects of various aging mechanisms on snubber operation. The second was to determine the efficacy of existing tests in determining the effects of aging and degradation mechanisms. These tests include breakaway force, drag force, velocity/ acceleration range for activation in tension or compression, release rates within specified tension/compression limits, and restricted thermal movement. The snubber operating experience was reviewed using licensee event reports and other historical data for a period of more than 10 years. Data were statistically analyzed using arbitrary snubber populations. Value-impact was considered in terms of exposure to a radioactive environment for examination/ testing and the influence of lost snubber function and subsequent testing program expansion on the costs and operation of a nuclear power plant. The implications of the observed trends were assessed; recommendations include modifying or improving examination and testing procedures to enhance snubber reliability. Optimization of snubber populations by selective removal of unnecessary snubbers was also considered. (author)

  11. Fracture mechanical analyses concerning the load-bearing capacity of dissimilar welds in piping with circumferential cracks; Bruchmechanische Untersuchungen zum Tragvermoegen von Rohrleitungen mit Umfangsrissen in Mischschweissnaehten

    Energy Technology Data Exchange (ETDEWEB)

    Schuett, T.; Hoffmann, M.; Schuler, X.; Stumpfrock, L. [Materialpruefungsanstalt (MPA), Univ. Stuttgart, Stuttgart (Germany)

    2007-07-01

    In power plants numerous dissimilar welds exist between ferritic and austenitic components. The evaluation of the influence of real or postulated flaws to the load bearing and failure behaviour in the area of dissimilar welded joints, e. g. in piping of the primary circuit of nuclear power plants, is an essential part of the safety assessment of these plants. The fracture assessment of dissimilar welds containing flaws is therefore an important issue of structural integrity considerations. During the past decades several assessment procedures for the evaluation of flaws in welds have been developed mainly based on methods originally developed for homogenous materials. The main difficulties in applying these methods are that they are restricted to two different materials and the very limited experimental evaluation especially by component tests. To improve the general understanding of the influence of flaws in piping systems with dissimilar welds and to have a more reliable basis for the assessment of the fracture behaviour, full scale tests on pipes containing a weld between ferritic and austenitic steel have been performed. The test results of these tests are summarised and used for the evaluation of analytical and numerical assessment procedures. (orig.)

  12. MICROSTRUCTURAL CHARACTERIZATION OF PRIMARY COOLANT PIPE STEEL

    OpenAIRE

    Miller, M; 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 long term thermal aging. The cast duplex microstructure consisted of austenite with 15% δ-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. T...

  13. Experimental data on seismic response of piping

    International Nuclear Information System (INIS)

    Displacement-controlled, sinusoidal vibration tests were performed on one-inch to four-inch diameter pipe segments. Straight pipe segments; straight pipe with a local reduced wall thickness; double spans; pipe segments with back-to-back flanges, socket welds, and elbows were tested at the Berkeley Nuclear Laboratories. Test results demonstrate that, within the limits of the test parameters, pipe spans will not collapse at extreme levels of dynamic loading. Extreme, in this case, means input levels as much as 16 times the Section III Level D allowable of 3 Sm or 2Sy. Material ductility limits the response to levels such that the fatigue damage from one seismic event of ten equivalent maximum stress cycles should be insignificant if the pressure is limited to the design pressure of 2/3 Sy. The cyclic life of austenitic pipe is superior to that of carbon steel. The fundamental frequency of the pipe span has a significant effect on pipe response. The lower the frequency, the greater the acceleration response at the same input acceleration. The lowest frequency tested was 5 Hz. Dramatic levels of material dynamic strain hardening were noted

  14. Thermal fatigue of pipes induced by fluid temperature change (18). Applicability of crack growth law based on continuum fracture mechanics criterion to small fatigue cracks

    International Nuclear Information System (INIS)

    High cycle thermal fatigue failure of pipes induced by fluid temperature change is one of the interdisciplinary issues to be concerned for long term structural reliability of high temperature components in energy systems. In order to explore advanced life assessment methods to prevent the failure, fatigue crack propagation tests were earned out in an austenitic stainless steel. Special attention was paid to the applicability of continuum fracture mechanics treatment to small or short cracks. It was shown exponentially that the crack propagation analysis based on continuum fracture mechanics was almost successfully applied, to the small fatigue cracks of which size was comparable to a few times of material grain size. (author)

  15. Parameters influencing the transgranular stress corrosion cracking behaviour of austenitic stainless steels in systems conveying reactor coolant

    International Nuclear Information System (INIS)

    During replacement of an auxiliary system in the German PWR KKS (NPP Stade) a damage was detected in a valve housing and in the connected piping both made from stabilised austenitic stainless steel. During operation stagnant conditions are present in this area. Based on the failure analysis chloride induced stress corrosion cracking (SCC) was found as the dominating root cause. In the open literature many cases of corrosion observed in the water/steam interface in valve components as well as in adjacent portions of auxiliary circuits made of un-stabilized stainless steels are mentioned. A common feature of the reported cases is that transgranular cracking was found. Extensive laboratory investigations revealed that non-stabilised austenitic stainless steels are also sensitive to transgranular cracking in boric acid solutions particularly in concentrated solutions. Often these solutions are contaminated with chlorides and/or oxygen is present. Taking into account the literature data the question could arise whether the above mentioned cracking may be also caused by boric acid attack. Thus, for stabilised stainless steels laboratory exposure tests at 80 C in saturated aerated boric acid solution and at 300 C in (at 100 C) saturated, oxygen free boric acid solution have been performed. Double-U-bend specimens and wedge loaded 1T-CT specimens made of Ti- and Nb-stabilised austenitic stainless steels were used. The results revealed no evidence of crack initiation and crack growth. Based on the laboratory results and the literature data an attempt is undertaken to separate parameters influencing chloride induced SCC from the effect of boric acid. (authors)

  16. Predictions for fatigue crack growth life of cracked pipes and pipe welds using RMS SIF approach and experimental validation

    International Nuclear Information System (INIS)

    The objective of the present study is to understand the fatigue crack growth behavior in austenitic stainless 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) specimens machined from the actual pipe/pipe weld. Analyses have been carried out to predict the fatigue crack growth life of the austenitic stainless steel pipes/pipes 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 (KRMS) at deepest and surface points. Crack growth and the crack shape with loading cycles have been evaluated. 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 (KRMS) evaluated using second scheme gives better fatigue crack growth life prediction compared to that of first scheme. Fatigue crack growth in pipe weld (Gas Tungsten Arc Welding) can be predicted well using Paris constants of base material but prediction is non-conservative for pipe weld (Shielded Metal Arc Welding). Further, predictions using fatigue crack growth rate curve of ASME produces conservative results for pipe and GTAW pipe welds and comparable results for SMAW pipe welds. - Highlights: → Predicting fatigue crack growth of Austenitic Stainless Steel pipes and pipe welds. → Use of RMS-SIF and local SIF at maximum

  17. Effect of re-austenitization on the transformation texture inheritance

    Science.gov (United States)

    Kaijalainen, A.; Suikkanen, P.; Porter, D. A.

    2015-04-01

    Bainitic-martensitic microstructures produced by direct quenching austenite subjected to different degrees of pancaking have been re-austenitized and quenched to fully martensitic structures in order to investigate the effect of prior texture on the final martensite texture. Three different prior austenite pancaking states varying from convex-like to highly pancaked were investigated using an ultrahigh-strength strip steel hot rolled with various finish rolling temperatures followed by direct quenching. Microstructures were characterized using FESEM and transformation texture analysed using FESEM-EBSD at the strip surface, quarter- thickness and mid-thickness positions. The results show that an increase in rolling reduction below the non-recrystallization temperature increases the intensities of ∼{554}α and ∼{112}α texture components in the ferrite along the strip mid-thickness and of the ∼{112}α component at the surface. The re-austenitization of the materials at 910°C for 30 min led to an inheritance of the same components from the parent specimens, but also increased the intensity of {001}α, {110}α and {011}α components.

  18. Austenite decomposition in carbon steel under dynamic deformation conditions

    Directory of Open Access Journals (Sweden)

    A. Nowotnik

    2007-01-01

    Full Text Available Purpose: The main purpose of this paper was to estimate the effect of the dynamic conditions resulting fromdeformation process on the austenite decomposition into ferrite and pearlite (A→F+P in the commercial carbon steel.Design/methodology/approach: In the paper flow stress curves and microstructure of deformed steel within therange of discontinuous (austenite to pearlite and austenite to ferrite transformation at different strain rates andcooling rates were presented. The microstructure of hot deformed samples was tested by means of an opticaland electron microscopy.Findings: It was shown that the flow localization during hot deformation and preferred growth of the pearlitecolonies at shear bands was very limited. The most characteristic feature of the microstructure observed for hotdeformed samples was the development of carbides that nucleated along elongated ferrite grains.Research limitations/implications: In spite of intense strain hardening due to deformation and phasetransformation overlapping, microstructural observation of deformed samples did not reveal significant flowlocalization effects or heterogeneous distribution of the eutectoid components. Therefore, complementary testsshould be carried out on the steel with higher strain above the 0.5 value.Originality/value: There was no data referred to particular features of the dynamic processes, such as dynamicrecrystallization and recovery, dynamic precipitation, that can occur during austenite decomposition into ferrite,and especially during discontinuous transformation of austenite to pearlite.

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

    International Nuclear Information System (INIS)

    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

  20. Proceedings of a specialist meeting on the ultrasonic inspection of reactor components

    International Nuclear Information System (INIS)

    Beside synthesis of two conferences on nondestructive testing and on inspection, the contributions of this conference are reporting experimental observations and research works on ultrasonic techniques, methods, procedures (pre-service or in-service) and equipment for the inspection of nuclear reactor components (pressure vessels, tubing and piping), generally in stainless steel (often austenitic or ferritic) material or in zirconium alloy. Some contributions are also dealing with the relationship between material microstructure and ultrasonic inspection method and equipment, or with the detection and sizing precision of flaws (cracks)

  1. EVALUATION OF PIPE CUTTING TECHNOLOGIES IN SHIPBUILDING

    OpenAIRE

    Kafali, Mustafa; Ozkok, Murat; Cebi, Selcuk

    2014-01-01

    Pipes are the most significant ones of the components which constitutes the vessel body. Pipes are fabricated in piping plant at shipyard and exposed to some processes such as cutting, bending, hydrostatic tests, galvanizing and so on. Cutting operation is also vital process among the other ones since it is very crucial that the cutting surfaces are flat and the right angles. In shipyards, there are various pipe cutting methods such as plasma, oxygen, metal saw, band saw and abrasive cutting ...

  2. The mechanical stability of retained austenite in low-alloyed TRIP steel under shear loading

    International Nuclear Information System (INIS)

    The microstructure evolution during shear loading of a low-alloyed TRIP steel with different amounts of the metastable austenite phase and its equivalent DP grade has been studied by in-situ high-energy X-ray diffraction. A detailed powder diffraction analysis has been performed to probe the austenite-to-martensite transformation by characterizing simultaneously the evolution of the austenite phase fraction and its carbon concentration, the load partitioning between the austenite and the ferritic matrix and the texture evolution of the constituent phases. Our results show that for shear deformation the TRIP effect extends over a significantly wider deformation range than for simple uniaxial loading. A clear increase in average carbon content during the mechanically-induced transformation indicates that austenite grains with a low carbon concentration are least stable during shear loading. The observed texture evolution indicates that under shear loading the orientation dependence of the austenite stability is relatively weak, while it has previously been found that under tensile load the {110}〈001〉 component transforms preferentially. The mechanical stability of retained austenite in TRIP steel is found to be a complex interplay between the interstitial carbon concentration in the austenite, the grain orientation and the load partitioning

  3. Through-Thickness Measurements of Residual Stresses in an Overlay Dissimilar Weld Pipe using Neutron Diffraction

    International Nuclear Information System (INIS)

    The distribution of residual stresses in dissimilar material joints has been extensively studied because of the wide applications of the dissimilar welds in many inevitable complex design structures. Especially the cracking of dissimilar welding has been a long standing issue of importance in many components of the power generation industries such as nuclear power plant, boiling pressure system, and steam generators. In particular, several failure analysis and direct observations have shown that critical fractures have frequently occurred in one side of the dissimilar welded parts. For example, the heat-affected zone on the ferrite steel side is known to critical in many dissimilar welding pipes when ferrite (low carbon steel) and austenite (stainless) steels are joined. The main cause of the residual stresses can be attributed to the mismatch in the coefficient of thermal expansion between the dissimilar metals (ferrite and austenite). Additional cladding over circumferential welds is known to reinforce the mechanical property due to the beneficial compressive residual stress imposed on the weld and heat-affected zone. However, science-based quantitative measurement of the through thickness residual stress distribution is very limited in literature. The deep penetration capability of neutrons into most metallic materials makes neutron diffraction a powerful tool to investigate and map the residual stresses of materials throughout the thickness and across the weld. Furthermore, the unique volume averaged bulk characteristic of materials and mapping capability in three dimensions are suitable for the engineering purpose. Thus, the neutron-diffraction measurement method has been selected as the most useful method for the study of the residual stresses in various dissimilar metal welded structures. The purpose of this study is to measure the distribution of the residual stresses in a complex dissimilar joining with overlay in the weld pipe. Specifically, we measured

  4. Austenite formation during intercritical annealing

    Directory of Open Access Journals (Sweden)

    A. Lis

    2008-07-01

    Full Text Available Purpose: of this paper is the effect of the soft annealing of initial microstructure of the 6Mn16 steel on the kinetics of the austenite formation during next intercritical annealing.Design/methodology/approach: Analytical TEM point analysis with EDAX system attached to Philips CM20 was used to evaluate the concentration of Mn, Ni and Cr in the microstructure constituents of the multiphase steel and mainly Bainite- Martensite islands.Findings: The increase in soft annealing time from 1-60 hours at 625°C increases Mn partitioning between ferrite and cementite and new formed austenite and decreases the rate of the austenite formation during next intercritical annealing in the (α+γ temperature range at 700 and 750°C. The general equations for carbide dissolution and austenite formation in intercritical temperature range were established.Research limitations/implications: The final multiphase microstructure can be optimised by changing the time / temperature parameters of the intercritical heating in the (α+γ temperature range.Originality/value: The knowledge of partitioning of alloying elements mainly Mn during soft annealing and intercritical heating is very important to optimise the processing technology of intercritical annealing for a given amount of the austenite.

  5. Heat pipes

    CERN Document Server

    Dunn, Peter D

    1994-01-01

    It is approximately 10 years since the Third Edition of Heat Pipes was published and the text is now established as the standard work on the subject. This new edition has been extensively updated, with revisions to most chapters. The introduction of new working fluids and extended life test data have been taken into account in chapter 3. A number of new types of heat pipes have become popular, and others have proved less effective. This is reflected in the contents of chapter 5. Heat pipes are employed in a wide range of applications, including electronics cooling, diecasting and injection mo

  6. A study of austenitization of SG iron

    Indian Academy of Sciences (India)

    Uma Batra; Pankaj Tandon; Kulbir Kaur

    2000-10-01

    Austenitization process of three SG irons with varying compositions and as cast matrix microstructure has been studied at three austenitization temperatures of 850, 900 and 950C for different time periods. Microstructure, hardness and X-ray diffraction have been used to reveal the nature of dependence of the process on austenitization temperature, time and as cast structure. The optimum austenitization time is maximum for ferritic and minimum for pearlitic matrix.

  7. Detection by ultrasonic waves of discontinuities in cast steels and weldings of austenitic stainless steels

    International Nuclear Information System (INIS)

    The study of discontinuities in cast iron and austenitic weldings by means of ultrasound is extremely difficult when materials present rough structures. By virtue of the necessity existing in nuclear power plants and others, of verify to the integrity of cast pieces an austenitic weldings, such a verification is being studying at global level. Materials with a rough grain structure, frequently present in industry, are: bombs to moving fluids, pipe fittings, elbows and austenitic weldings. This problem, traditionally, has been studied varying the frequency in gropers. A new approach will be presented here, based in the use of the high sensibility of the equipment and piezoelectric tablet, as well as the maximum withdrawal of perturbation zone in X-ray tube (Author)

  8. Investigation on ductile fracture behavior of 3-inch diameter Type 304 stainless steel pipe with a circumferential through-wall crack

    International Nuclear Information System (INIS)

    At Japan Atomic Energy Research Institute (JAERI), a pipe fracture test program has been conducted as a part of the Leak-Before-Break (LBB) verification research in LWR pressure boundary pipings. In this program, fracture behavior and fracture criteria of the circumferentially cracked pipe have been investigated, using austenitic stainless steel pipes and carbon steel pipes. This report presents a four-point bending test results of 3-inch diameter Type 304 austenitic stainless steel pipes with circumferential through-wall crack at room temperature. Pipe fracture data were obtained from the test in regard to load-loadline displacement, crack extension, crack opening area, and so on. Discussions are performed on the effect of pipe ovalization ratio at maximum load, the application of the net-section collapse criterion, and the effect of initial crack angle, wall-thickness etc. on J-R curve. Furthermore, the crack opening area was estimated by assuming a simple crack model. (author)

  9. Stress corrosion cracking of austenitic stainless steel in glycerol solution and chloride solution at elevated temperature

    International Nuclear Information System (INIS)

    Stress Corrosion Cracking (SCC) is an environmentally assisted failure caused by exposure to a corrodant while under a sustained tensile stress. SCC is most often rapid, unpredictable and catastrophic. Failure can occur in as little as a few hours or take years to happen. Most alloys are susceptible to SCC in one or more environments requiring careful consideration of alloy type in component design. In aqueous chloride environments austenitic stainless steels and many nickel based alloys are known to perform poorly. One of products Oleo chemical is glycerol solution. Glycerol solution contains chloride with concentration 50 ppm - 150 ppm. Austenitic stainless steel is usually used in distillation construction tank and pipe line of glycerol. Material AISI 304 will be failure in this glycerol solution with this concentration in 5 years. In production process, concentration of chloride in glycerol becomes more than 150 ppm at temperature 150 degree Celsius. The reason is that the experiment I conducted in high chloride with concentration such as 6000 ppm, 9000 ppm, and 12000 ppm. The stress corrosion cracking of the austenitic stainless steels of types AISI 304, 316 and 316L in glycerol solution at elevated temperature 150 degree Celsius is investigated as a function variation of chloride concentration, namely 50, 6000, 9000 and 12000 ppm using a constant load method with two kinds of initial tensile stress as 50 % and 70 % yield strength. The experiment uses a spring loaded fixture type and is based on ASTM G49 for experiment method, and E292 for geometry of specimen. Pitting corrosion occurs on the surface specimen until the stress level reaches the ultimate strength. Pitting corrosion attack and depletion occur on the surface as initiation of SCC failure as the stress reaches the ultimate strength. Failure has occurred in catastrophic brittle fracture type of transgranular. AISI 304 was more susceptible for all conditions. In chloride solution with concentration of

  10. Examination of overlay repaired BWR pipe joints

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking (IGSCC) in a large number of austenitic stainless steel girth welds in boiling water reactor (BWR) piping has prompted the development of the weld overlay for repair (WOR) as a short-term remedy. It is necessary to examine the deposited overlay weld material for adequate definition of its condition and to monitor the overlaid IGSCC to determine if it grows past the bounds assumed in the design of the repair. This paper reports on NDE techniques evaluated using weld overlaid pipe samples containing known defects, overlaid samples removed from BWR service, and overlaid weld joints in plant. These samples included overlays containing fabrication defects and overlaid pipes containing deep and shallow laboratory- and service-induced IGSCC

  11. Remote controlled in-pipe manipulators for dye-penetrant inspection and grinding of weld roots inside of pipes

    International Nuclear Information System (INIS)

    Technical plants which have to satisfy stringent safety criteria must be continuously kept in line with the state of art. This applies in particular to nuclear power plants. The quality of piping in nuclear power plants has been improved quite considerably in recent years. By virtue of the very high quality requirements fulfilled in the manufacture of medium-carrying and pressure-retaining piping, one of the focal aspects of in-service inspections is the medium wetted inside of the piping. A remote controlled pipe crawler has been developed to allow to perform dye penetrant testing of weld roots inside piping (ID ≥ 150 mm). The light crawler has been designed such that it can be inserted into the piping via valves (gate valves, check valves,...) with their internals removed. Once in the piping, all crawler movements are remotely controlled (horizontal and vertical pipes incl. the elbows). If indications are found these discontinuities are ground according to a qualified procedure using a special grinding head attached to the crawler with complete extraction of all grinding residues. The in-pipe grinding is a special qualified three (3) step performance that ensures no residual tensile stress (less than 50 N/mm2) in the finish machined austenitic material surface. The in-pipe inspection system, qualified according to both the specifications of the German Nuclear Safety Standards Commission (KTA) and the American Society of Mechanical Engineers (ASME), has already been used successfully in nuclear power plants on many occasions. (author)

  12. Heat Pipes

    Science.gov (United States)

    1996-01-01

    Heat Pipes were originally developed by NASA and the Los Alamos Scientific Laboratory during the 1960s to dissipate excessive heat build- up in critical areas of spacecraft and maintain even temperatures of satellites. Heat pipes are tubular devices where a working fluid alternately evaporates and condenses, transferring heat from one region of the tube to another. KONA Corporation refined and applied the same technology to solve complex heating requirements of hot runner systems in injection molds. KONA Hot Runner Systems are used throughout the plastics industry for products ranging in size from tiny medical devices to large single cavity automobile bumpers and instrument panels.

  13. Effect of sodium environment on the creep-rupture and low-cycle fatigue behavior of austenitic stainless steels

    International Nuclear Information System (INIS)

    Austenitic stainless steels used for in-core structural components, piping, valves, and the intermediate heat exchanger in Liquid-Metal Fast-Breeder Reactors (LMFBRs) are subjected to sodium at elevated temperatures and to complex stress conditions. As a result, the materials can undergo compositional and microstructural changes as well as mechanical deformation by creep and cyclic fatigue processes. In the present paper, information is presented on the creep-rupture and low-cycle fatigue behavior of Types 304 and 316 stainless steel in the solution-annealed condition and after long-term exposure to flowing sodium. The nonmetallic impurity-element concentrations in the sodium were controlled at levels similar to those in EBR-II primary sodium. Strain-time relationships developed from the experimental creep data were used to generate isochronous stress-creep strain curves as functions of sodium-exposure time and temperature. The low-cycle fatigue data were used to obtain relationships between plastic strain range and cycles-to-failure based on the Coffin-Manson formalism and a damage-rate approach developed at ANL. An analysis of the cyclic stress-strain behavior of the materials showed that the strain-hardening rates for the sodium-exposed steels were larger than those for the annealed material. However, the sodium-exposed specimens showed significant softening, as evidenced by the lower stress at half the fatigue life. Microstructural information obtained from the different specimens suggests that crack initiation is more difficult in the long-term sodium-exposed specimens when compared with the solution-annealed material. Based on the expected carbon concentrations in LMFBR primary system sodium, moderate carburization of the austenitic stainless steels will not degrade the mechanical properties to a significant extent, and therefore, will not limit the performance of out-of-core components. (author)

  14. Effect of sodium environment on the creep-rupture and low-cycle fatigue behavior of austenitic stainless steels

    International Nuclear Information System (INIS)

    Austenitic stainless steels used for in-core structural components, piping, valves, and the intermediate heat exchanger in Liquid-Metal Fast-Breeder Reactors (LMFBRs) are subjected to sodium at elevated temperatures and to complex stress conditions. As a result, the materials can undergo compositional and microstructural changes as well as mechanical deformation by creep and cyclic fatigue processes. Information is presented on the creep-rupture and low-cycle fatigue behavior of Types 304 and 316 stainless steel in the solution-annealed condition and after long-term exposure to flowing sodium. The nonmetallic impurity-element concentrations in the sodium were controlled at levels similar to those in EBR-II primary sodium. Strain-time relationships developed from the experimental creep data were used to generate isochronous stress-creep strain curves as functions of sodium-exposure time and temperature. The low-cycle fatigue data were used to obtain relationships between plastic strain range and cycles-to-failure based on the Coffin-Manson formalism and a damage-rate approach developed at ANL. An analysis of the cyclic stress-strain behavior of the materials showed that the strain-hardening rates for the sodium-exposed steels were larger than those for the annealed material. However, the sodium-exposed specimens showed significant softening, as evidenced by the lower stress at half the fatigue life. Microstructural information obtained from the different specimens suggests that crack initiation is more difficult in the long-term sodium-exposed specimens when compared with the solution-annealed material. Based on the expected carbon concentrations in LMFBR primary system sodium, moderate carburization of the austenitic stainless steels will not degrade the mechanical properties to a significant extent, and therefore, will not limit the performance of out-of-core components

  15. An approach to prior austenite reconstruction

    International Nuclear Information System (INIS)

    One area of interest in Friction Stir Welding (FSW) of steels is to understand microstructural evolution during the process. Most of the deformation occurs in the austenite temperature range. Quantitative microstructural measurements of prior austenite microstructure are needed in order to understand evolution of the microstructure. Considering the fact that room temperature microstructure in ferritic steels contains very little to no retained austenite, prior austenite microstructure needs to be recovered from the room temperature ferrite. In this paper, an approach based on Electron Backscattered Diffraction (EBSD) is introduced to detect Bain zones. Bain zone detection is used to reconstruct prior austenite grain structure. Additionally, a separate approach based on phase transformation orientation relationships is introduced in order to recover prior austenite orientation. - Highlights: ►This approach provides a tool to reconstruct large-scale austenite microstructures. ► It recovers prior austenite orientation without relying on retained austenite. ► It utilizes EBSD data from the room temperature microstructure. ► Higher number of active variants leads to more accurate reconstructions. ► At least two variants are needed in order to recover prior austenite orientation.

  16. Ultrasonic thickness measurement criteria in thinned pipe management program

    International Nuclear Information System (INIS)

    Credibility of thickness data is very important in the thinned pipe management program. This report presents following criteria; thickness measurement for each pipe component type, wear and wear rate calculation, and remaining service life assessment of thinned pipe component. And, the necessary items should be contained in the inspection report are presented

  17. Pipe whip analysis using the Tedel code

    International Nuclear Information System (INIS)

    In view of their abundance, piping systems are one of the main components in power industries and in particular in nuclear power plants. They must be designed for normal as well as faulted conditions, for safety requirements. For example, in case of a sudden break, the pipe whip must be studied in order to determine if the free pipe may damage neighbouring structures like other pipes, concrete containments, etc... The prediction of the dynamic behaviour of the free pipe requires accounting for several non-linearities. For this purpose, a beam type finite element program (TEDEL) has been used. The aim of this paper is to outline the main features of this program, when applied to pipe whip analysis. An example of application to a real case (the behaviour of PWR primary piping under LOCA conditions) will also be presented. (author)

  18. Ultrasonic inspection of austenitic welds

    International Nuclear Information System (INIS)

    The metallurgical structure of austenitic welds is described and contrasted with that found in ferritic welds. It is shown that this structure imparts a marked elastic anisotropy in the ultrasonic propagation parameters. Measurements of variations in the apparent attenuation of sound and deviations in the beam direction are described. The measurements are interpreted in terms of the measured velocity anisotropy. Two applications of the fundamental work are described. In the first it is shown how, by using short pulse compression wave probes, and with major modification of the welding procedure, a stainless steel fillet weld in an AGR boiler can be inspected. In the second application, alternative designs of a transition butt weld have been compared for ease of ultrasonic inspection. The effects of two different welding processes on such an inspection are described. Finally, the paper examines the prospects for future development of inspection and defect-sizing techniques for austenitic welds. (author)

  19. MODELING OF AUSTENITE GRAIN SIZE IN LOW-ALLOY STEEL WELD METAL

    Institute of Scientific and Technical Information of China (English)

    A.G.Huang; Y.S.Wang; Z.Y.Li; J.G.Xiong; Q.Hu

    2004-01-01

    The size of austenite grain hassignificant effects on components and proportions of various ferrites in low-alloy steel weld metal.Therefore,it is important to determine the size of austenite grain in the weld metal.In this paper,a model based upon the carbon diffusion rate is developed for computing austenite grain size in low-alloy steel weld metal during continuous cooling.The model takes into account the effects of the weld thermal cycles,inclusion particles and various alloy elements on the austenite grain growth.The calculating results agree reasonably with those reported experimental observations.The model demonstrates a significant promise to understand the weld microstructure and properties based on the welding science.

  20. Developments in austenitic steels containing manganese

    International Nuclear Information System (INIS)

    Two broad categories of austenitic steels are considered in this review: (a) alloys based on the Fe-Mn-C system, typified by austenitic wear resistant (Hadfield) steels and (b) alloys based on the Fe-Mn-Cr system, typified by austenitic corrosion resistant steels. Advances made in recent years in understanding and improving the relevant properties and manufacturing methods of these steels are critically appraised. The development of austenitic manganese bearing high technology steels for fusion reactor and other non-magnetic applications, as well as those that can be used in cryogenic structures, is also considered. (author)

  1. ABSTRACTS WELDEL PIPE AND TUBE

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    article mainly analysis the weld character and the crack producing mechanism. Subject Terms: low alloy steel high strength brittleness crack weld seam HeHong,JiangXiuhan,LiLing (27) Technical Development of the Inner pipeline Corrosion Inspection Abroad This article introduces the technical development of the pipeline inner corrosion inspection abroad, and the advanced leak magnetic intelligent acrawl unit and the principle of ultrasonic intelligent acrawl machine to inspect the pipeline corrosion as well as the equipment of this intelligent acrawl machine. It points out that the gap of the pipeline corrosion inspection in China and the problems must be solved now. Subject Terms:long distance pipeline corrosion inspection ultrasonic leak magnetic intelligent acrawl machine Li Ye(32) The Development of the Large Sized Cold Bend Square Pipe Through reforming of the equipment and improving of the process of the cold bend steel unit, Han Kou Steel Rolling Plant of Wuhan Steel Group can produce large sized cold bend square pipe with the size of 300X300mm. This article introduces the establishment of the reforming project and the process of the reforming practice in details. Subject Terms:cold bend square pipe development standing roller width reform Yinguoyao,Huchengzhou,Zhangbinjie(35) The Application of Automation in Pipelines NorthWest Petroleum Bureau The automation system of Ku-Shan pipeline realizes the objective to transport oil safely. steadily and efficiently. The article introduces the components of the automation system in second-control center and station-control system. Through the examples of the control of pressure-reducing and non-heated oil transportation with the aid of additive, it also presents that automation plays an important role in the respects of managing production, reducing personnels and increasing profit. Subject Terms:long-distance transportation pipeline automatic control management Li Hongliang,Qian Wenxian,Qian Xiaolin,Qian Lezhong(37

  2. Early detection of micro-structural changes due to fatigue of non-corrosive austenitic stainless steels

    International Nuclear Information System (INIS)

    In view of life extension efforts of nuclear power plants, many investigations are in progress in order to assess the structural integrity of different components. In many cases, this involves unexpected loads, which were not taken into account during design of components, e.g. temperature cycling arising from unforeseen stratification flow conditions. Under certain power plant transients (start-up/shut-down, hot stand-by, thermal stratification) at critical locations of piping and nozzles, material degradation caused by accumulated cyclic plastic strain takes place. However, materials subjected to cyclic loading exhibit changes in microstructure already before macroscopic crack initiation begins, this period covers a considerable part of fatigue life. Existing methods for in-service inspection are mainly specialised for crack detection. Advanced non-destructive testing methods for monitoring of material degradation are sensitive to any micro-structural changes in the material leading to a degradation of the mechanical properties. Therefore, these indirect methods require a careful interpretation of the measured signal in terms of micro-structural evolutions due to ageing. During cyclic loading of austenitic stainless steel, microstructural changes occur, which affect both the mechanical and the physical properties. Typical features are the rearrangement of dislocations and, in some cases, a deformation-induced martensitic phase transformation. In our investigation martensite formation was used as an indication for material degradation due to fatigue. Knowledge about mechanisms and influencing parameters of the martensitic transformation process is essential for the application in a lifetime monitoring system. The investigations showed that for a given austenitic stainless steel the deformation-induced martensite depends on the applied strain amplitude, the cycle number (usage factor, lifetime) and the temperature. It was demonstrated that the volume fraction of

  3. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...... contact with outside envelope of heat pipes and collectors are in contact with liquid metal secondary cooling system that transfers waste heat to radiator....

  4. Finite element simulation of residual stresses induced by the dissimilar welding of a P92 steel pipe with weld metal IN625

    International Nuclear Information System (INIS)

    Residual stresses induced by the fusion arc-welding of steel pipe joints in power generation plants are a concern to the industry. Residual stresses are induced by the process of welding due to the extreme nature of thermal cycles during the process. Welding is essential in the construction of high-grade steel pipelines, used as a conduit for steam at high temperature and pressure. The integrity and endurance of the welded pipes are necessary for the safe operation in power plants, which may be compromised by the presence of residual stresses. The finite element (FE) method is an effective tool for the prediction of residual stresses in such components, as long as the material behaviour can be accurately modelled. This paper reports the FE simulation of residual stresses, due to the arc-welding of a P92 steel pipe mainly using a nickel-based alloy (IN625) as a dissimilar weld material. The structural analysis part of the FE method of determining the residual stress field in the welded pipe is described and the results presented and discussed. Two user-defined subroutines have been used in the FE structural analysis to simulate the way the different phases of steel evolve during welding, including their differing plastic and hardening behaviour, derived from uniaxial tensile material testing carried out over a wide range of temperature. Thermal-expansion, including the effects of solid-state phase transformations in P92, has also been numerically modelled in the two subroutines, one of which prescribes two phases of P92 steel (tempered martensite and austenite) while the other assumes three phases (tempered martensite, austenite and untempered martensite). -- Highlights: • FE simulation of residual stresses due to welding P92 steel pipes with IN625. • FE simulation of P92 phases: austenite as well as tempered and untempered martensite. • Effect of transformation plasticity and annealing on residual stresses is negligible. • Very high tensile residual

  5. The application of the internal friction damping nondestructive evaluation technique for detecting incipient cracking in critical components, by-pass lines and piping systems in boiling water reactors

    International Nuclear Information System (INIS)

    A technical feasibility program of utilizing an internal friction damping (IFD) nondestructive evaluation (NDE) technique under laboratory and field conditions for piping systems has been initiated. The applicability of the IFD-NDE technique for four inch (10.1 cm) stainless steel by-pass lines and ten inch (25.4 cm) feedwater lines has been established. The overall objectives of the program include: 1. application of the IFD-NDE technique for four inch (10.1 cm) stainless steel lines that simulate by-pass lines, 2. application of the IFD-NDE technique to full size feedwater lines in an on-line boiling water reactor, 3. evaluate the feasibility of utilizing the IFD-NDE technique as an incipient crack detection method for laboratory stress corrosion cracking tests. (orig.)

  6. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  7. Researches upon the cavitation erosion behaviour of austenite steels

    Science.gov (United States)

    Bordeasu, I.; Popoviciu, M. O.; Mitelea, I.; Salcianu, L. C.; Bordeasu, D.; Duma, S. T.; Iosif, A.

    2016-02-01

    Paper analyzes the cavitation erosion behavior of two stainless steels with 100% austenitic structure but differing by the chemical composition and the values of mechanical properties. The research is based on the MDE(t) and MDER(t) characteristic curves. We studied supplementary the aspect of the eroded areas by other to different means: observations with performing optical microscopes and roughness measurements. The tests were done in the T2 vibratory facility in the Cavitation Laboratory of the Timisoara Polytechnic University. The principal purpose of the study is the identification of the elements influencing significantly the cavitation erosion resistance. It was established the effect of the principal chemical components (determining the proportion of the structural components in conformity the Schaffler diagram) upon the cavitation erosion resistance. The results of the researches present the influence of the proportion of unstable austenite upon cavitation erosion resistance. The stainless steel with the great proportion of unstable austenite has the best behavior. The obtained conclusion are important for the metallurgists which realizes the stainless steels used for manufacturing the runners of hydraulic machineries (turbines and pumps) with increased resistance to cavitation attack.

  8. Manual ultrasonic inspection of austenitic and dissimilar welds

    International Nuclear Information System (INIS)

    In the third phase of the Programme for Inspection of Steel Components (PISCIII) several round robin tests with austenitic test samples and assemblies containing dissimilar welds were carried out. Based on the destructive analysis the performance of different inspection procedures and participating teams was evaluated. By using low recording level and a combination of conventional shear wave angle beam probes and mode conversion technique the team of VTT Manufacturing Technology detected all the flaws considered in the evaluation of the results of round robin tests. The majority of flaws in the austenitic test samples were stress corrosion cracks (IGSCC). The sizes (heights) of the flaws were mainly overestimated slightly by the team but the sizing capability clearly fulfils the requirements of the latest ASME Code Appendix VIII. In the nozzle and dissimilar weld action the team detected all rejectable flaws in the assembly that was estimated to be the most difficult assembly of this action. (orig.)

  9. Fatigue crack propagation in welded joint of austenitic steel for nuclear power engineering

    International Nuclear Information System (INIS)

    The crack propagation characteristics were obtained for Cr-Ni type austenitic steel 08Kh18N10T under variable stress in the individual zones of a welded joint on a pipe. Measurements of the threshold deviation of the stress intensity factor, ΔKp, showed that the root zone of the pipe welded joint was the weakest point as concerns crack propagation. The threshold values obtained for the filler metal on the pipe outer surface were considerably greater than those for the root zone of the welded joint and slightly greater than those for the base material and for the transition between the joint and the base material. The measured propagation response showed that the rate of fatigue crack propagation was for the base material higher by up to one order for low ΔK than for the filler joint and the root zone of the joint. (J.B.). 5 figs., 3 tabs., 6 refs

  10. Expanded austenite, crystallography and residual stress

    DEFF Research Database (Denmark)

    Christiansen, Thomas; Hummelshøj, Thomas Strabo; Somers, Marcel A. J.

    2010-01-01

    The identity of expanded austenite as developing during low temperature nitriding and/or carburising of austenitic stainless steel has been under debate since the very first observation of this phase. In the present article, recent results obtained with (a) homogeneous samples of various uniform ...

  11. Expanded austenite; crystallography and residual stress

    DEFF Research Database (Denmark)

    Christiansen, Thomas; Hummelshøj, Thomas Strabo; Somers, Marcel A. J.

    2009-01-01

    The identity of expanded austenite as developing during low temperature nitriding and/or carburizing of austenitic stainless steel has been under debate since the very first observation of this phase. In the present article recent results obtained with i) homogeneous samples of various uniform co...

  12. Expanded austenite; crystallography and residual stress

    DEFF Research Database (Denmark)

    Christiansen, Thomas; Hummelshøj, Thomas Strabo; Somers, Marcel A. J.

    The identity of expanded austenite as developing during low temperature nitriding and/or carburizing of austenitic stainless steel has been under debate since the very first observation of this phase. In the present article recent results obtained with i) homogeneous samples of various uniform co...

  13. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in dire...

  14. Heat pipes

    CERN Document Server

    Dunn, Peter D

    1982-01-01

    A comprehensive, up-to-date coverage of the theory, design and manufacture of heat pipes and their applications. This latest edition has been thoroughly revised, up-dated and expanded to give an in-depth coverage of the new developments in the field. Significant new material has been added to all the chapters and the applications section has been totally rewritten to ensure that topical and important applications are appropriately emphasised. The bibliography has been considerably enlarged to incorporate much valuable new information. Thus readers of the previous edition, which has established

  15. Pipe whip restraints - protection for safety related equipment of WWER nuclear power plants

    International Nuclear Information System (INIS)

    The paper concerns the problem of enhancing the protection of WWER NPP equipment against the effect of a high energy piping break which results in a pipe whip. A pipe whip restraint has been designed in order to protect nearby safety related systems and components. The pipe whip restraint properties have been optimized using results of iterative non-linear dynamic analyses of the piping system response to forces due to fluid streaming out of the broken pipe. (author)

  16. Development of dynamic vibration absorber for nuclear piping system

    International Nuclear Information System (INIS)

    The dynamic vibration absorber was newly developed for the piping system. The higher damping ratio was accomplished and the seismic response of the piping system was consequently reduced. In this dynamic vibration absorber, a stainless mesh spring is used and can be modeled as a complex spring element. From the results of the component test using the straight piping and the three dimensional piping model test (8 inch in diameter and 18 m in length, this dynamic vibration absorber is confirmed to be effective to suppress the vibration for the piping system of wide frequency range. The application method of the dynamic vibration absorber to the three dimensional piping system is also described

  17. Resolution of concerns in auxiliary feedwater piping

    International Nuclear Information System (INIS)

    Auxiliary feedwater piping systems at pressurized water reactor (PWR) nuclear power plants have experienced unanticipated operating conditions during plant operation. These unanticipated conditions have included plant events involving backleakage through check valves, temperatures in portions of the auxiliary feedwater piping system that exceed design conditions, and the occurrence of unanticipated severe fluid transients. The impact of these events has had an adverse effect at some nuclear stations on plant operation, installed plant components and hardware, and design basis calculations. Beaver Valley Unit 2, a three loop pressurized water reactor nuclear plant, has observed anomalies with the auxiliary feedwater system since the unit went operational in 1987. The consequences of these anomalies and plant events have been addressed and resolved for Beaver Valley Unit 2 by performing engineering and construction activities. These activities included pipe stress, pipe support and pipe rupture analysis, the monitoring of auxiliary feedwater system temperature and pressure, and the modification to plant piping, supports, valves, structures and operating procedures

  18. Lightweight Heat Pipes Made from Magnesium

    Science.gov (United States)

    Rosenfeld, John N.; Zarembo, Sergei N.; Eastman, G. Yale

    2010-01-01

    Magnesium has shown promise as a lighter-weight alternative to the aluminum alloys now used to make the main structural components of axially grooved heat pipes that contain ammonia as the working fluid. Magnesium heat-pipe structures can be fabricated by conventional processes that include extrusion, machining, welding, and bending. The thermal performances of magnesium heat pipes are the same as those of equal-sized aluminum heat pipes. However, by virtue of the lower mass density of magnesium, the magnesium heat pipes weigh 35 percent less. Conceived for use aboard spacecraft, magnesium heat pipes could also be attractive as heat-transfer devices in terrestrial applications in which minimization of weight is sought: examples include radio-communication equipment and laptop computers.

  19. Flexibility of trunnion piping elbows

    International Nuclear Information System (INIS)

    Flexibility factors and stress indices for piping component such as straight pipe, elbows, butt-welding tees, branch connections, and butt-welding reducers are contained in the code, but many of the less common piping components, like the trunnion elbow, do not have flexibility factors or stress indices defined. The purpose of this paper is to identify the in-plane and out-of-plane flexibility factors in accordance with code procedures for welded trunnions attached to the tangent centerlines of long radius elbows. This work utilized the finite element method as applicable to plates and shells for calculating the relative rotations of the trunnion elbow-ends for in-plane and out-of-plane elbow moment loadings. These rotations are used to derive the corresponding in-plane and out-of-plane flexibility factors. (orig./GL)

  20. A Review of Buried Piping Management in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Over the past several years, instances of buried piping leaks have occurred in safety-related and nonsafety related piping at nuclear power plants. Buried piping systems are used for fire suppression, radiation waste treatment, or component cooling. This piping may be either concrete or metal. For example, nuclear power plants require an external heat sink, such as a lake or river, in order to maximize thermal cycle efficiencies and provide an ultimate safety heat sink. Typically, the piping between these heat sinks and the plant secondary cooling loop is known as raw water piping. Degradation of raw water piping affects the plant's ability to remove excess heat in case of an accident. Access to these pipes could be extremely limited. In this paper, various issues and activities related to buried piping are discussed

  1. STUDY OF ELECTROMAGNETIC STIRRING REFINING MICRO- STRUCTURES OF PIPE-LINE STEEL SAW DEPOSITS

    Institute of Scientific and Technical Information of China (English)

    Y. Zhang; B.N. Qian; X.M. Guo

    2002-01-01

    The effects of electromagnetic stirring on the microstructures of pipe-line steel SAWdeposited metal were investigated. The results showed that electromagnetic stirringincreased the number density of inclusions with 0.2-0.6μm in diameter and promotedthe formation and refining of acicular ferrite within austenite grains. The low tem-perature toughness of deposited metal was improved.

  2. 49 CFR 192.193 - Valve installation in plastic pipe.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Valve installation in plastic pipe. 192.193 Section 192.193 Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND... Components § 192.193 Valve installation in plastic pipe. Each valve installed in plastic pipe must...

  3. High Mn austenitic stainless steel

    Science.gov (United States)

    Yamamoto, Yukinori [Oak Ridge, TN; Santella, Michael L [Knoxville, TN; Brady, Michael P [Oak Ridge, TN; Maziasz, Philip J [Oak Ridge, TN; Liu, Chain-tsuan [Knoxville, TN

    2010-07-13

    An austenitic stainless steel alloy includes, in weight percent: >4 to 15 Mn; 8 to 15 Ni; 14 to 16 Cr; 2.4 to 3 Al; 0.4 to 1 total of at least one of Nb and Ta; 0.05 to 0.2 C; 0.01 to 0.02 B; no more than 0.3 of combined Ti+V; up to 3 Mo; up to 3 Co; up to 1W; up to 3 Cu; up to 1 Si; up to 0.05 P; up to 1 total of at least one of Y, La, Ce, Hf, and Zr; less than 0.05 N; and base Fe, wherein the weight percent Fe is greater than the weight percent Ni, and wherein the alloy forms an external continuous scale including alumina, nanometer scale sized particles distributed throughout the microstructure, the particles including at least one of NbC and TaC, and a stable essentially single phase FCC austenitic matrix microstructure that is essentially delta-ferrite-free and essentially BCC-phase-free.

  4. Specialist meeting on leak before break in reactor piping and vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bartholome, G.; Bazant, E.; Wellein, R. [Siemens KWU, Stuttgart (Germany)] [and others

    1997-04-01

    A series of research projects sponsored by the Federal Minister for Education, Science, Research and Technology, Bonn are summarized and compared to utility, manufacturer, and vendor tests. The purpose of the evaluation was to experimentally verify Leak-before-Break behavior, confirm the postulation of fracture preclusion for piping (straight pipe, bends and branches), and quantify the safety margin against massive failure. The results are applicable to safety assessment of ferritic and austenitic piping in primary and secondary nuclear power plant circuits. Moreover, because of the wide range of the test parameters, they are also important for the design and assessment of piping in other technical plant. The test results provide justification for ruling out catastrophic fractures, even on pipes of dimensions corresponding to those of a main coolant pipe of a pressurized water reactor plant on the basis of a mechanical deterministic safety analysis in correspondence with the Basis Safety Concept (Principle of Fracture Exclusion).

  5. Equipment and piping for nuclear power plants, test and research reactors, and nuclear installations

    International Nuclear Information System (INIS)

    The standard provides rules for testing the following welding operations: welding of ferritic pearlitic steels, welding of corrosion resistant austenitic steels, welding of components made of corrosion resistant austenitic steels and ferritic pearlitic steels, build-up welding of groove faces, and build-up welding of corrosion protecting layers

  6. Equipment and piping for nuclear power plants, test and research reactors, and nuclear installations

    International Nuclear Information System (INIS)

    The standard applies to the following welding operations: welding of ferritic pearlitic steels, welding of corrosion resistant austenitic steels, welding of components made of corrosion resistant austenitic steels and ferritic pearlitic steels, build-up welding of groove faces, and build-up welding of corrosion protecting layers

  7. Experimental determination of the crack driving force J for circumferencially cracked piping elbows

    International Nuclear Information System (INIS)

    In order to study the nocivity of defects on the mechanical behaviour of piping systems, experiments on circumferencially cracked piping components have been conducted. We aimed to compute a value of the so-called ''Jic integral'', which characterizes the capacity to initiate of a pre-existing crack. Using experimental results only, we tried to get rid of the usual formulae hypothesis, based on semi-computed results. Mechanical through-walled defects have been machined in order to represent crack from 600 to 1500 center angle. Bending moment to curvature curves have been outlined and initiation of propagation was determined using electric potential drop technic. Ten in-plan closing tests on austenitic elbows were conducted: 5 in an opening way, 5 in a closing way. J integral value was determined by integration of experimental results. Direct integrations were performed as well as calculations using scaling functions, fitted on the test results. The crack driving force Jic at initiation was slightly greater than other values found in other publications

  8. Plastics pipe couplings

    International Nuclear Information System (INIS)

    A method is described of making a pipe coupling of the type comprising a plastics socket and a resilient annular sealing member secured in the mouth thereof, in which the material of at least one component of the coupling is subjected to irradiation with high energy radiation whereby the material is caused to undergo cross-linking. As examples, the coupling may comprise a polyethylene or plasticised PVC socket the material of which is subjected to irradiation, and the sealing member may be moulded from a thermoplastic elastomer which is subjected to irradiation. (U.K.)

  9. Conquering service water pipe corrosion

    International Nuclear Information System (INIS)

    Damage to the components of Hope Creek's service water system from corrosion was so severe that a six-year US$37 million project to replace 2850 feet of pipe was begun in 1988. Due for completion in 1994, the bulk of the work having already been done, the project offers lessons for existing plants and for future designs. (Author)

  10. Improvements in or relating to pipe joints

    International Nuclear Information System (INIS)

    Pipe joints are described that are particularly suitable for liquid metal cooled nuclear reactors. The object is to provide a joint capable of accommodating movements resulting from differential expansion of the reactor components. Full constructional details are given. (UK)

  11. Influence of surface texture on the galling characteristics of lean duplex and austenitic stainless steels

    DEFF Research Database (Denmark)

    Wadman, Boel; Eriksen, J.; Olsson, M.;

    2010-01-01

    Two simulative test methods were used to study galling in sheet forming of two types of stainless steel sheet: austenitic (EN 1.4301) and lean duplex LDX 2101 (EN 1.4162) in different surface conditions. The pin-on-disc test was used to analyse the galling resistance of different combinations of...... industrial tool used for high volume production of pump components, to compare forming of LDX 2101 and austenitic stainless steel with equal thickness. The forming forces, the geometry and the strains in the sheet material were compared for the same component. It was found that LDX steels can be formed to...... high strain levels in tools normally applied for forming of austenitic steels, but tool adaptations are needed to comply with the higher strength and springback of the material....

  12. Ultrasonic Characterization of Cast Austenitic Stainless Steel Microstructure: Discrimination between Equiaxed- and Columnar-Grain Material – An Interim Study

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep; Good, Morris S.; Diaz, Aaron A.; Anderson, Michael T.; Watson, Bruce E.; Peters, Timothy J.; Dixit, Mukul; Bond, Leonard J.

    2009-10-27

    Ultrasonic nondestructive evaluation (NDE) and inspection of cast austenitic stainless steel (CASS) components used in the nuclear power industry is neither as effective nor reliable as is needed due to detrimental effects upon the interrogating ultrasonic beam and interference from ultrasonic backscatter. The root cause is the coarse-grain microstructure inherent to this class of materials. Some ultrasonic techniques perform better for particular microstructural classifications and this has led to the hypothesis that an ultrasonic inspection can be optimized for a particular microstructural class, if a technique exists to reliably classify the microstructure for feedback to the inspection. This document summarizes scoping experiments of in-situ ultrasonic methods for classification and/or characterization of the material microstructures in CASS components from the outside surface of a pipe. The focus of this study was to evaluate ultrasonic methods and provide an interim report that documents results and technical progress. An initial set of experiments were performed to test the hypothesis that in-service characterization of cast austenitic stainless steel (CASS) is feasible, and that, if reliably performed, such data would provide real-time feedback to optimize in-service inspections in the field. With this objective in mind, measurements for the experiment were restricted to techniques that should be robust if carried forward to eventual field implementation. Two parameters were investigated for their ability to discriminate between different microstructures in CASS components. The first parameter was a time-of-flight ratio of a normal incidence shear wave to that of a normal incidence longitudinal wave (TOFRSL). The ratio removed dependency on component thickness which may not be accurately reported in the field. The second parameter was longitudinal wave attenuation. The selected CASS specimens provided five equiaxed-grain material samples and five columnar

  13. Mathematical Model of the Processoof Pearlite Austenitization

    Directory of Open Access Journals (Sweden)

    Olejarczyk-Wożeńska I.

    2014-10-01

    Full Text Available The paper presents a mathematical model of the pearlite - austenite transformation. The description of this process uses the diffusion mechanism which takes place between the plates of ferrite and cementite (pearlite as well as austenite. The process of austenite growth was described by means of a system of differential equations solved with the use of the finite difference method. The developed model was implemented in the environment of Delphi 4. The proprietary program allows for the calculation of the rate and time of the transformation at an assumed temperature as well as to determine the TTT diagram for the assigned temperature range.

  14. High-energy X-ray diffraction study on the temperature-dependent mechanical stability of retained austenite in low-alloyed TRIP steels

    International Nuclear Information System (INIS)

    The stability of the retained austenite has been studied in situ in low-alloyed transformation-induced-plasticity (TRIP) steels using high-energy X-ray diffraction during tensile tests at variable temperatures down to 153 K. A detailed powder diffraction analysis has been performed to probe the austenite-to-martensite transformation by characterizing the evolution of the phase fraction, load partitioning and texture of the constituent phases simultaneously. Our results show that at lower temperatures the mechanically induced austenite transformation is significantly enhanced and extends over a wider deformation range, resulting in a higher elongation at fracture. Low carbon content grains transform first, leading to an initial increase in average carbon concentration of the remaining austenite. Later the carbon content saturates while the austenite still continues to transform. In the elastic regime the probed {h k l} planes develop different strains reflecting the elastic anisotropy of the constituent phases. The observed texture evolution indicates that the austenite grains oriented with the {2 0 0} plane along the loading direction are transformed preferentially as they show the highest resolved shear stress. For increasing degrees of plastic deformation the combined preferential transformation and grain rotation results in the standard deformation texture for austenite with the {1 1 1} component along the loading direction. The mechanical stability of retained austenite in TRIP steel is found to be a complex interplay between carbon concentration in the austenite, grain orientation, load partitioning and temperature.

  15. Shield For Flexible Pipe

    Science.gov (United States)

    Ponton, Michael K.; Williford, Clifford B.; Lagen, Nicholas T.

    1995-01-01

    Cylindrical shield designed to fit around flexible pipe to protect nearby workers from injury and equipment from damage if pipe ruptures. Designed as pressure-relief device. Absorbs impact of debris ejected radially from broken flexible pipe. Also redirects flow of pressurized fluid escaping from broken pipe onto flow path allowing for relief of pressure while minimizing potential for harm.

  16. Ultrasonic inspection of austenitic welds

    International Nuclear Information System (INIS)

    The ultrasonic examination of austenitic stainless steel weld metal has always been regarded as a difficult proposition because of the large and variable ultrasonic attenuations and back scattering obtained from apparently similar weld deposits. The work to be described shows how the existence of a fibre texture within each weld deposit (as a result of epitaxial growth through successive weld beads) produces a systematic variation in the ultrasonic attenuation coefficient and the velocity of sound, depending upon the angle between the ultrasonic beam and the fibre axis. Development work has shown that it is possible to adjust the welding parameters to ensure that the crystallographic texture within each weld is compatible with improved ultrasonic transmission. The application of the results to the inspection of a specific weld in type 316 weld metal is described

  17. Fundamentals of piping design

    CERN Document Server

    Smith, Peter

    2013-01-01

    Written for the piping engineer and designer in the field, this two-part series helps to fill a void in piping literature,since the Rip Weaver books of the '90s were taken out of print at the advent of the Computer Aid Design(CAD) era. Technology may have changed, however the fundamentals of piping rules still apply in the digitalrepresentation of process piping systems. The Fundamentals of Piping Design is an introduction to the designof piping systems, various processes and the layout of pipe work connecting the major items of equipment forthe new hire, the engineering student and the vetera

  18. Comparisons of simplified method JS and finite element for cracked pipes under thermal and mechanical loading

    International Nuclear Information System (INIS)

    In order to estimate the value of the parameter J for cracks in pipes, a complete set of finite element computations was conducted in France by CEA. The axisymetrical crack is located in the inner wall of the pipe. The crack depth is one quarter or one eighth of the thickness. The pipe, in austenitic steel or in ferritic steel, is subjected to internal pressure, axial load and linear thermal gradient through the thickness. The finite element results of J were compared to those predicted by simplified method Js, introduced in the A 16 guide developed by CEA for fast breeder reactor. (orig.)

  19. Seismic proving test of ultimate piping strength (current status of preliminary tests)

    International Nuclear Information System (INIS)

    In 1998 Fiscal Year, the 6 year program of piping tests was initiated with the following objectives: i) to clarify the elasto-plastic response and ultimate strength of nuclear piping, ii) to ascertain the seismic safety margin of the current seismic design code for piping, and iii) to assess new allowable stress rules. In order to resolve extensive technical issues before proceeding on to the seismic proving test of a large-scale piping system, a series of preliminary tests of materials, piping components and simplified piping systems is intended. In this paper, the current status of the material tests and the piping component tests is reported. (author)

  20. Piping and structural interaction

    OpenAIRE

    Tveit, Siv-Anni

    2015-01-01

    Structural and piping stress analyses are generally conducted separately in most of the situations. A pipe stress analysis is performed based on loads caused by pressure, temperature variations, weight of the pipe (with contents) and blast. The pipes are analysed for several load cases separated into different design types like sustained, expansion and occasional. Established practice in piping design and calculation does not specifically account for structural flexibility. It is assumed that...

  1. Cast alumina forming austenitic stainless steels

    Science.gov (United States)

    Muralidharan, Govindarajan; Yamamoto, Yukinori; Brady, Michael P

    2013-04-30

    An austenitic stainless steel alloy consisting essentially of, in terms of weight percent ranges 0.15-0.5C; 8-37Ni; 10-25Cr; 2.5-5Al; greater than 0.6, up to 2.5 total of at least one element selected from the group consisting of Nb and Ta; up to 3Mo; up to 3Co; up to 1W; up to 3Cu; up to 15Mn; up to 2Si; up to 0.15B; up to 0.05P; up to 1 total of at least one element selected from the group consisting of Y, La, Ce, Hf, and Zr; alumina, and a stable essentially single phase FCC austenitic matrix microstructure, the austenitic matrix being essentially delta-ferrite free and essentially BCC-phase-free. A method of making austenitic stainless steel alloys is also disclosed.

  2. Considerations on ultrasonic testing of austenitic steel weld joints

    International Nuclear Information System (INIS)

    Starting from concrete examples, the Working Group describes the difficulties encountered when ultrasonic testing of welds is carried out on austenitic alloys. It indicates particularly the technique used for the detection of defects such as lack of fusion and cracks and also where inspection has to be carried out on welds between dissimilar metals or between strongly attenuated parent metals. It concludes on the necessity of carrying out a case study for each testing problem encountered, taking into account the testability of a component from the stage of manufacturing

  3. Microbial deterioration of materials - simulation, case histories and countermeasures for metallic materials: Pitting corrosion caused by microbiological activity at austenitic stainless pipes used for river water. Mikrobielle Werkstoffzerstoerung - Simulation, Schadensfaelle und Gegenmassnahmen fuer metallische Werkstoffe: Lochfrass an Flusswasser-Rohrleitungen aus hochlegierten austenitischen Staehlen durch mikrobielle Aktivitaet

    Energy Technology Data Exchange (ETDEWEB)

    Korkhaus, J. (BASF AG, Ludwigshafen am Rhein (Germany). Abt. Werkstofftechnik); Titz, J.T. (BASF AG, Ludwigshafen am Rhein (Germany). Abt. Werkstofftechnik); Wagner, G.H. (BASF AG, Ludwigshafen am Rhein (Germany). Abt. Werkstofftechnik)

    1994-02-01

    Rhine River water used for cooling purposes increasingly causes pitting at circumferential welds in type 321 and 616Ti stainless steel piping. Type 321 is attacked even at ambient temperature, type 316Ti at 35 to 55 C. At these temperatures both materials did not show pitting in the past. Failure analysis and corrosion tests, both in the laboratory and in a pilot plant (river water circuit), showed that the open circuit potential of stainless steels is shifted into a region of instable pitting resistance by a microbiological effect. Thus, pit nuclei are able to grow even at circumferential joints welded without any visible annealing colours. The actual concept to select cooling water pipe materials, therefore, has to be reconsidered. (orig.)

  4. Nuclear components

    International Nuclear Information System (INIS)

    The main features of the EPR concerning the fabrication of the reactor are: -) the size of the components, -) the modification of the design compared with classical PWR, and -) an intensive use of forging (in particular the cold and hot legs of the primary circuit are forged). This series of slides overviews the fabrication of the components for the EPR by highlighting the differences with the previous generation of reactors. 4 types of components are reviewed: the reactor vessel and internals, steam generators, primary circuit pipes, and primary coolant pumps. (A.C.)

  5. Insulated pipe clamp design

    International Nuclear Information System (INIS)

    Thin wall large diameter piping for breeder reactor plants can be subjected to significant thermal shocks during reactor scrams and other upset events. On the Fast Flux Test Facility, the addition of thick clamps directly on the piping was undesired because the differential metal temperatures between the pipe wall and the clamp could have significantly reduced the pipe thermal fatigue life cycle capabilities. Accordingly, an insulated pipe clamp design concept was developed. The design considerations and methods along with the development tests are presented. Special considerations to guard against adverse cracking of the insulation material, to maintain the clamp-pipe stiffness desired during a seismic event, to minimize clamp restraint on the pipe during normal pipe heatup, and to resist clamp rotation or spinning on the pipe are emphasized

  6. Piping Stress Analysis

    International Nuclear Information System (INIS)

    Piping stress analysis on Primary Sampling System, Reactor Cooling System, and Feedwater System for AP600 have been performed. Piping stress analysis is one of the requirements in the design of piping system. Piping stress is occurred due to static and dynamic loads during service. Analysis was carried out. Using PS+CAEPIPE software based on the individual and combination loads with assumption that failure could be happened during normal, upset, emergency and faulted condition as describe in ASME III/ANSI B31.1. With performing the piping stress analysis, the layout (proper pipe routing) of the piping system can be design with the requirements of piping stress and pipe supports in mind I.e sufficient flexibility for thermal expansion, etc to commensurate with the i tended service such as temperatures, pressure, seismic and anticipated loading

  7. Flexible ocean upwelling pipe

    Science.gov (United States)

    Person, Abraham

    1980-01-01

    In an ocean thermal energy conversion facility, a cold water riser pipe is releasably supported at its upper end by the hull of the floating facility. The pipe is substantially vertical and has its lower end far below the hull above the ocean floor. The pipe is defined essentially entirely of a material which has a modulus of elasticity substantially less than that of steel, e.g., high density polyethylene, so that the pipe is flexible and compliant to rather than resistant to applied bending moments. The position of the lower end of the pipe relative to the hull is stabilized by a weight suspended below the lower end of the pipe on a flexible line. The pipe, apart from the weight, is positively buoyant. If support of the upper end of the pipe is released, the pipe sinks to the ocean floor, but is not damaged as the length of the line between the pipe and the weight is sufficient to allow the buoyant pipe to come to a stop within the line length after the weight contacts the ocean floor, and thereafter to float submerged above the ocean floor while moored to the ocean floor by the weight. The upper end of the pipe, while supported by the hull, communicates to a sump in the hull in which the water level is maintained below the ambient water level. The sump volume is sufficient to keep the pipe full during heaving of the hull, thereby preventing collapse of the pipe.

  8. Probabilistic Risk Assessment: Piping Fragility due to Earthquake Fault Mechanisms

    OpenAIRE

    Bu Seog Ju; WooYoung Jung; Myung-Hyun Noh

    2015-01-01

    A lifeline system, serving as an energy-supply system, is an essential component of urban infrastructure. In a hospital, for example, the piping system supplies elements essential for hospital operations, such as water and fire-suppression foam. Such nonstructural components, especially piping systems and their subcomponents, must remain operational and functional during earthquake-induced fires. But the behavior of piping systems as subjected to seismic ground motions is very complex, owing ...

  9. The Impact Of the Welded Joints Made Of X8CrNiTi18–10 Stainless Steel on the Reliability Estimation of Pipes

    OpenAIRE

    Raimondas Skindaras; Jonas Bendikas; Vigantas Kumšlytis

    2011-01-01

    The chrome-nickel stainless steels of austenitic class applied in chemistry and energy industry are often used in the production of exceptional structures employed in an environment aggressive and dangerous for human life. Therefore, it is particularly significant for durability and reliability requirements. The article explores cracks that appeared in a tube made of X8CrNiTi18–10 austenitic steel. The examined pipe has worked for 90 000 hours under high temperature and pressure in an aggress...

  10. A review on nickel-free nitrogen containing austenitic stainless steels for biomedical applications.

    Science.gov (United States)

    Talha, Mohd; Behera, C K; Sinha, O P

    2013-10-01

    The field of biomaterials has become a vital area, as these materials can enhance the quality and longevity of human life. Metallic materials are often used as biomaterials to replace structural components of the human body. Stainless steels, cobalt-chromium alloys, commercially pure titanium and its alloys are typical metallic biomaterials that are being used for implant devices. Stainless steels have been widely used as biomaterials because of their very low cost as compared to other metallic materials, good mechanical and corrosion resistant properties and adequate biocompatibility. However, the adverse effects of nickel ions being released into the human body have promoted the development of "nickel-free nitrogen containing austenitic stainless steels" for medical applications. Nitrogen not only replaces nickel for austenitic structure stability but also much improves steel properties. Here we review the harmful effects associated with nickel and emphatically the advantages of nitrogen in stainless steel, as well as the development of nickel-free nitrogen containing stainless steels for medical applications. By combining the benefits of stable austenitic structure, high strength, better corrosion and wear resistance and superior biocompatibility in comparison to the currently used austenitic stainless steel (e.g. 316L), the newly developed nickel-free high nitrogen austenitic stainless steel is a reliable substitute for the conventionally used medical stainless steels. PMID:23910251

  11. Theoretical and experimental investigations on integrity assessment of pipes and elbows

    OpenAIRE

    Chattopadhyay, Jayanta

    2004-01-01

    Integrity assessment of piping components is very essential for safe and reliable operation of both conventional and nuclear power plants. It is especially important for nuclear power plants because the leak-before-break (LBB) concept, which involves detailed integrity assessment of piping components, is now widely used to design the primary heat transport (PHT) system piping. There are various issues in the integrity assessment of piping components that are unresolved or not fully resolve...

  12. Early detection of micro-structural changes due to fatigue of non-corrosive austenitic stainless steels; Frueherkennung von mikrostrukturellen Aenderungen bei Ermuedung in nichtrostenden austenitischen Staehlen

    Energy Technology Data Exchange (ETDEWEB)

    Kalkhof, D.; Niffenegger, M.; Grosse, M

    2003-03-01

    In view of life extension efforts of nuclear power plants, many investigations are in progress in order to assess the structural integrity of different components. In many cases, this involves unexpected loads, which were not taken into account during design of components, e.g. temperature cycling arising from unforeseen stratification flow conditions. Under certain power plant transients (start-up/shut-down, hot stand-by, thermal stratification) at critical locations of piping and nozzles, material degradation caused by accumulated cyclic plastic strain takes place. However, materials subjected to cyclic loading exhibit changes in microstructure already before macroscopic crack initiation begins, this period covers a considerable part of fatigue life. Existing methods for in-service inspection are mainly specialised for crack detection. Advanced non-destructive testing methods for monitoring of material degradation are sensitive to any micro-structural changes in the material leading to a degradation of the mechanical properties. Therefore, these indirect methods require a careful interpretation of the measured signal in terms of micro-structural evolutions due to ageing. During cyclic loading of austenitic stainless steel, microstructural changes occur, which affect both the mechanical and the physical properties. Typical features are the rearrangement of dislocations and, in some cases, a deformation-induced martensitic phase transformation. In our investigation martensite formation was used as an indication for material degradation due to fatigue. Knowledge about mechanisms and influencing parameters of the martensitic transformation process is essential for the application in a lifetime monitoring system. The investigations showed that for a given austenitic stainless steel the deformation-induced martensite depends on the applied strain amplitude, the cycle number (usage factor, lifetime) and the temperature. It was demonstrated that the volume fraction of

  13. Experimental benchmark for piping system dynamic-response analyses

    International Nuclear Information System (INIS)

    This paper describes the scope and status of a piping system dynamics test program. A 0.20 m(8 in.) nominal diameter test piping specimen is designed to be representative of main heat transport system piping of LMFBR plants. Particular attention is given to representing piping restraints. Applied loadings consider component-induced vibration as well as seismic excitation. The principal objective of the program is to provide a benchmark for verification of piping design methods by correlation of predicted and measured responses. Pre-test analysis results and correlation methods are discussed

  14. Determination of the collapse load of circumferentially cracked pipes

    International Nuclear Information System (INIS)

    Pipes related to the Primary System of PWR reactors are manufactured from high toughness austenitic and ferritic steels, which are resistant to the unstable growth of defects. A flaw in a piping system has to cause a leakage in a considerable rate, before any growth of the flaw causes a catastrophic rupture of the piping. This is the concept of LBB (Leak-Before-Break). If a conservative analysis can demonstrate that this leakage can be detected and repaired before a sudden rupture of the pipe occurs, the regulatory commission can exclude from the design basis, the postulation of a DEGB (Double-Ended Guillotine Break) hypothetic and the considerations to its associated dynamic effects. As a consequence, the protections against dynamic effects can be desconsidered, bringing an immediate economic benefice. In terms of security, the removal of the protections, give also more benefits considering the existence of more space available at the installation for inspection activities. A fundamental stage in the LBB methodology consists in the analysis of the stability of a postulated throughwall flaw in a specific piping system. In this work, the methods DPFAD (Deformation Plasticity Failure Assessment Diagram), J-T Analysis and DFM (Ductile Fracture Method) are described and applied on the determination of the instability load in some piping configurations submitted to bending containing circumferential throughwall flaws, and where geometry and material variations are considered. The instability loads obtained by these methods are compared among them and compared with some experimental results. (author)

  15. Integrity assessment for safety relevant nuclear piping

    International Nuclear Information System (INIS)

    The failure behavior of pipes and piping components (e.g. straight pipe, pipe bend, T-joint) with and without cracks under different loading conditions has been investigated in numerous experimental and analytical/numerical research projects. The results of these projects were used to adjust and to verify different methodologies and procedures to calculate the failure loads, the respective critical crack sizes as well as the leak area and the leak rates. On the basis of the actual material characteristics, the actual as-built configurations and design of the piping systems, the knowledge of possible failure mechanism, concepts for the assessment of the integrity of the systems could be developed for the different actual as well as for postulated loading conditions. Based on the integrity assessment the leak before break behavior and break preclusion of safety relevant nuclear piping can be demonstrated. Examples are presented of the German assessment procedure for a main feed water line of a PWR as well as for the Indian assessment procedure for the primary heat transport system piping of a PHWR. - Highlights: • Description of the German Basis Safety Concept and Integrity Concept is presented. • Application is demonstrated for the main feed water line of a PWR plant. • Leak-before-break qualification of nuclear power plants in India. • Practical application to PHT system piping of a pressurized heavy water reactor. • Evaluation of leakage size crack and stability analysis verified by experiment

  16. Aspects and mechanisms of austenitic stainless steel corrosion in case of sodium leaks under mineral wool insulation

    International Nuclear Information System (INIS)

    Sodium pipe rupture tests representative of Fast Reactors accidents have been carried out on austenitic stainless steel surfaces. These tests improve the authors knowledge of small sodium leakage propagation in mineral wool insulation. Furthermore, they explain the new and unexpected aspects of the crevice corrosion phenomenon which has been observed on austenitic stainless steel external pipe surfaces. Experimental results show that the corrosion is limited to a peripheral annular zone, which extends out in concentric waves. The diameter of this corrosion zone is practically constant. Furthermore, the tests show that sodium does not expand directly on the pipe surface. The sodium sprays through the mineral wool insulation, where chemical reactions between silica fibers, occluded oxygen and water vapor occur at the same time. Simultaneously, there is a diffusion phenomenon of liquid sodium droplets on the mineral wool fibers. The study allows to prove the electrochemical nature of the corrosion. The excess liquid sodium, spraying as droplets on the pipe surface, induces an anodic dissolution mechanism by differential aeration. This phenomenon explains the random microscopic and macroscopic aspects of material removal

  17. PC-PRAISE, BWR Piping Reliability Analysis

    International Nuclear Information System (INIS)

    1 - Description of program or function: PC-PRAISE is a probabilistic fracture mechanics computer code developed for IBM or IBM compatible personal computers to estimate probabilities of leak and break in nuclear power plant cooling piping. 2 - Method of solution: PC-PRAISE considers the initiation and/or growth of crack-like defects in piping weldments. The initiation analyses are based on the results of laboratory studies and field observations in austenitic piping material operating under boiling water reactor conditions. The considerable scatter in such results is quantified and incorporated into a probabilistic model. The crack growth analysis is based on (deterministic) fracture mechanics principles, in which some of the inputs (such as initial crack size) are considered to be random variables. Monte Carlo simulation, with stratified sampling on initial crack size, is used to generate weldment reliability results. 3 - Restrictions on the complexity of the problem: There is essentially no limitation with PC-PRAISE but for large number of replications used in the Monte Carlo simulation scheme, computation time may become prohibitive

  18. Predictions of failure for some of the international pipe tests using the R6 method

    International Nuclear Information System (INIS)

    The use of state-of-the-art analysis methods to predict the behavior of large scale fracture experiments has identified some significant discrepancies. The same methods are used for assessing the integrity of structures and may be giving rise to similar discrepancies which need to be examined. This paper focuses on the application to piping of the assessment method developed in the United Kingdom known as R6. The examples chosen correspond to some specific piping experiments covering cases where failure is governed by both toughness as well as plastic instability. Most of the tests chosen were conducted in the Degraded Piping Program Phase 2 of the US Nuclear Regulatory Commission, in which about 70 tests were performed. Pipes with surface cracks as well as pipes with circumferential cracks were tested under three types of loading conditions; bending moment alone, pressure alone and combined bending and pressure. Two types of material were used in the program, carbon steel and austenitic steel, some of which contained welds. Results have also been obtained on liquid metal fast breeder reactor piping in a collaborative program between Interatom (FRG) and EDF and Novatome, a Division of Framatome. The data concerned initiation and instability behavior of circumferentially through-cracked pipes. These tests were made on large austenitic pipes with material properties quite different from those of the usual light water reactor piping. Cracks were located either in the base material, in the weld metal or in the heat-affected-zone. The pipes were subjected to bending only. In this paper the results of the R6 assessments are compared with the experimental data in terms of initiation and maximum load. A statistical treatment has been used to assess the significance of the discrepancy between the theoretical and experimental results

  19. Whip restraint for a steam pipe rupture event on a nuclear power plant / Alfred Cornelius Pieters

    OpenAIRE

    Pieters, Alfred Cornelius

    2013-01-01

    One of the requirements of a safe nuclear power plant design is the postulation of the dynamic effects of a steam pipe rupture. The dynamic effects are the discharging fluid and pipe whip on structures, systems or components. A pipe rupture can be caused in the steam pipe system where a defect such as a crack exists. Multiple factors contribute to the initiation of pipe cracks during the plant’s life. Cracks may start microscopically small and over time, with the assistance...

  20. Pipe whip analysis using the TEDEL code

    International Nuclear Information System (INIS)

    In view of their abundance, piping systems are one of the main components in power industries and in particular in nuclear power plants. They must be designed for normal as well as faulted conditions, for safety requirements. The prediction of the dynamic behaviour of the free pipe requires accounting for several nonlinearities. For this purpose, a beam type finite element program (TEDEL) has been used. The aim of this paper is to enlight the main features of this program, when applied to pipe whip analysis. An example of application to a real case will also be presented

  1. Thinned pipe management program of Korean NPPs

    International Nuclear Information System (INIS)

    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)

  2. Thermal stresses in pipes

    OpenAIRE

    Al-Zaharnah, Iyad

    2002-01-01

    This study presents results about thermal stresses in externally heated pipes that are subjected to different flow types: laminar flow, turbulent flow, and pulsating flow. The effect o f flow Reynolds number on thermal stresses in the pipe is studied. To investigate the influence o f fluid and solid properties on the resulting thermal stresses in pipes, two solids namely; steel and cooper and three fluids namely; water, coolanol-25, and mercury are used in the study. Pipes with different diam...

  3. Applications of the essay at slow deformation velocity in pipes of stainless steel AISI-304; Aplicaciones del ensayo a velocidad de deformacion lenta en tuberias de acero inoxidable AISI-304

    Energy Technology Data Exchange (ETDEWEB)

    Zamora R, L.; Mora R, T. De la [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2004-07-01

    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.

  4. Short cracks in piping and piping welds

    International Nuclear Information System (INIS)

    This program started on March 23, 1990, and has a duration of 4 years. The objective of the program is to develop and verify analyses by using existing and new experimental data for circumferentially cracked pipes, so modifications and improvements can be made to LBB and in-service flaw evaluation criteria. There are 7 technical tasks dealing, in general, with circumferentially cracked straight pipe under quasi-static loading. The tasks are as follows: short through wall cracked (TWC) pipe evaluations, short surface-cracked pipe evaluations, bi-metallic cracked pipe evaluations, dynamic strain aging and crack jump evaluations, anisotropic fracture evaluations, crack-opening-area evaluations, and NRCPIPE code improvements. There is also a separate task to develop international cooperation, interact with Section 11 of the ASME code, and perform program management functions

  5. Pipe-to-pipe impact program

    International Nuclear Information System (INIS)

    This report documents the tests and analyses performed as part of the Pipe-to-Pipe Impact (PTPI) Program at the Pacific Northwest Laboratory. This work was performed to assist the NRC in making licensing decisions regarding pipe-to-pipe impact events following postulated breaks in high energy fluid system piping. The report scope encompasses work conducted from the program's start through the completion of the initial hot oil tests. The test equipment, procedures, and results are described, as are analytic studies of failure potential and data correlation. Because the PTPI Program is only partially completed, the total significance of the current test results cannot yet be accurately assessed. Therefore, although trends in the data are discussed, final conclusions and recommendations will be possible only after the completion of the program, which is scheduled to end in FY 1984

  6. Influence of Martensite Fraction on the Stabilization of Austenite in Austenitic-Martensitic Stainless Steels

    Science.gov (United States)

    Huang, Qiuliang; De Cooman, Bruno C.; Biermann, Horst; Mola, Javad

    2016-05-01

    The influence of martensite fraction ( f α') on the stabilization of austenite was studied by quench interruption below M s temperature of an Fe-13Cr-0.31C (mass pct) stainless steel. The interval between the quench interruption temperature and the secondary martensite start temperature, denoted as θ, was used to quantify the extent of austenite stabilization. In experiments with and without a reheating step subsequent to quench interruption, the variation of θ with f α' showed a transition after transformation of almost half of the austenite. This trend was observed regardless of the solution annealing temperature which influenced the martensite start temperature. The transition in θ was ascribed to a change in the type of martensite nucleation sites from austenite grain and twin boundaries at low f α' to the faults near austenite-martensite (A-M) boundaries at high f α'. At low temperatures, the local carbon enrichment of such boundaries was responsible for the enhanced stabilization at high f α'. At high temperatures, relevant to the quenching and partitioning processing, on the other hand, the pronounced stabilization at high f α' was attributed to the uniform partitioning of the carbon stored at A-M boundaries into the austenite. Reduction in the fault density of austenite served as an auxiliary stabilization mechanism at high temperatures.

  7. Manufacture and Erection of SFR Components: Feedback from PFBR Experience

    International Nuclear Information System (INIS)

    Unique Features of SFR Components: • Large diameter thin walled shell and slender structures calling for stringent tolerances posing challenges in manufacturing, handling and erection. • Single side welds are unavoidable at some difficult locations. • In-service inspection is difficult. • Residual stresses should be minimum calling for robust heat treatment strategy. • Minimum number of materials to be used from reliability point of view (but not preferred from economic considerations). • Mainly austenitic stainless steels calling for careful considerations for welding without significant weld repairs and distortions. • Reactor assembly components decide the project time schedule (large manufacturing, assembly and erection time). • Leak tightness is very important in view of resulting sodium leaks. • Limited experience on manufacturing and erection of components. • Design and manufacturing codes still evolvingPFBR Reactor Assembly – Major Lessons: • Grid plate Large number of sleeves, posing difficulty in assembly, hard facing of large diameter plates and heavy flange construction. • Roof slab Large box type structure with many penetrations – complicated manufacturing process, time consuming and difficulty to overcome lamellar tearing problems. • Inclined Fuel Transfer Machine Complex manufacturing processes leading to large time and extensive qualification tests. • Increase of number of primary pipes – essential for enhancing safety. • Integration of components manufactured by different industries took unduly long time

  8. Explosive Surface Hardening of Austenitic Stainless Steel

    Science.gov (United States)

    Kovacs-Coskun, T.

    2016-04-01

    In this study, the effects of explosion hardening on the microstructure and the hardness of austenitic stainless steel have been studied. The optimum explosion hardening technology of austenitic stainless steel was researched. In case of the explosive hardening used new idea mean indirect hardening setup. Austenitic stainless steels have high plasticity and can be easily cold formed. However, during cold processing the hardening phenomena always occurs. Upon the explosion impact, the deformation mechanism indicates a plastic deformation and this deformation induces a phase transformation (martensite). The explosion hardening enhances the mechanical properties of the material, includes the wear resistance and hardness. In case of indirect hardening as function of the setup parameters specifically the flayer plate position the hardening increased differently. It was find a relationship between the explosion hardening setup and the hardening level.

  9. The Mossbauer spectroscopy studies of retained austenite

    Directory of Open Access Journals (Sweden)

    J. Frackowiak

    2007-10-01

    Full Text Available Purpose: of this paper: This paper completes the knowledge concerning the mechanisms of destabilization and properties of retained austenite. Investigations were performed on 120MnCrMoV8-6-4-2 steel, which was designed in 1998, in Phase Transformations Research Group of Department of Physical and Powder Metallurgy at the Faculty of Metals Engineering and Industrial Computer Science at AGH University of Science and Technology in Krakow.Design/methodology/approach: The samples of investigated steel were austenitized at the temperature of 900ºC and hardened in oil. Next, three from four samples were tempered. Tempering consisted of heating the samples up to chosen temperatures with a heating rate of 0.05ºC/s and, after reaching desired temperature, fast cooling. CEMS technique was applied for Mössbauer studies.Findings: Stabilized by heating up to 80ºC retained austenite, in the result of mechanical destabilization, transforms into low-temperature tempered martensite, with the structure of low bainite (into the structural constituent in which ε carbide exists.Research limitations/implications: The influence of the temperature, up to which the samples were heated during tempering, on the mechanical stability of retained austenite and on the products of its transformation, was determined.Practical implications: Changes occuring in retained austenite during tempering of steel of high hardenability (hardness, developed for potential applications on tools of enhanced wear resistance, were described.Originality/value: Mössbauer spectroscopy was applied not only for qantitative analysis of retained austenite, but also to analyze the values of quadrupole splitting and isomeric shift, what resulted in significant conclusions concerning the changes in its chemical composition, microstructure, and the level of stresses being present in it.

  10. Influence of Silicon on Swelling and Microstructure in Russian Austenitic Stainless Steels Irradiated to High Neutron Doses

    Energy Technology Data Exchange (ETDEWEB)

    Porollo, S.I.; Shulepin, S.V.; Konobeev, Y.V.; Garner, F. [Pacific Northwest National Laboratory, P.o. Box 999, Richland WA, AK 99352 (United States)

    2007-07-01

    Full text of publication follows: For some applications in fusion devices austenitic stainless steels are still considered to be candidates for use as structural components, but high neutron exposures must be endured by the steels. Operational experience of fast reactors in Western Europe, USA and Japan provides evidence of the possible use of austenitic steels up to {approx} 150 dpa. Studies aimed at improvement of existing Russian austenitic steels are being carried out in Russia. For improvement of irradiation resistance of Russian steels it is necessary to understand the basic mechanisms responsible for deterioration of steel properties. This understanding can be achieved by continuing detailed investigations of the microstructure of cladding steels after irradiation to high doses. By investigating the evolution of radiation-induced microstructure in neutron irradiated steels of different chemical composition one can study the effect of chemical variations on steel properties. Silicon is one of the most important chemical elements that strongly influence the behavior of austenitic steel properties under irradiation. In this paper results are presented of investigations of the effect of silicon additions on void swelling and microstructure of base austenitic stainless steel EI-847 (0.06C-16Cr-15Ni- 3Mo-Nb) irradiated as fuel pin cladding of both regular and experimental assemblies in the BOR-60, BN-350 and BN-600 fast reactors to neutron doses up to 49 dpa. The possible mechanisms of silicon's effect on void swelling in austenitic stainless steels are presented and analyzed. (authors)

  11. Integrity evaluation of ice plugged pipes applied on short jacket

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yeong Don; Son, Geum Su [Korea Hydro and Nuclear Power Co., Ltd, Ulsan (Korea, Republic of); Yun, Doo Ri; Ha, Byeong Guk; Hwang, Sang Moon; Kang, Beom Soo [Busan National University, Busan (Korea, Republic of)

    2002-04-01

    In special industrial fields such as nuclear power plants and chemical plants, it is often necessary to repair system components without plant shutdown or drainage of system having many piping structures which may have hazardous or expensive fluid. A temporary ice plugging method for blocking internal flow is considered as a useful method in that case. According to the pipe freezing guideline of the nuclear power plant, the length of a freezing jacket must be longer than twice of the pipe diameter. However, for applying the ice plugging to short pipes which do not have enough freezing length because of geometrical configuration, it is inevitable to use shorter jacket less than twice of the pipe diameter. In this study, the integrity evaluation for short pipes in the nuclear power plant is conducted by an experiment and the finite element analysis. From the results, the ice plugging process in short pipes can be safely carried out without any plastic deformation and fracture.

  12. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  13. Recycle of radiologically contaminated austenitic stainless steels

    International Nuclear Information System (INIS)

    The United States Department of Energy owns large quantities of radiologically contaminated austenitic stainless steel which could by recycled for reuse if appropriate release standards were in place. Unfortunately, current policy places the formulation of a release standard for USA industry years, if not decades, away. The Westinghouse Savannah River Company, Idaho National Engineering Laboratory and various university and industrial partners are participating in initiative to recycle previously contaminated austenitic stainless steels into containers for the storage and disposal of radioactive wastes. This paper describes laboratory scale experiments which demonstrated the decontamination and remelt of stainless steel which had been contaminated with radionuclides

  14. Modeling of austenite to ferrite transformation

    Indian Academy of Sciences (India)

    Mohsen Kazeminezhad

    2012-06-01

    In this research, an algorithm based on the -state Potts model is presented for modeling the austenite to ferrite transformation. In the algorithm, it is possible to exactly track boundary migration of the phase formed during transformation. In the algorithm, effects of changes in chemical free energy, strain free energy and interfacial energies of austenite–austenite, ferrite–ferrite and austenite–ferrite during transformation are considered. From the algorithm, the kinetics of transformation and mean ferrite grain size for different cooling rates are calculated. It is found that there is a good agreement between the calculated and experimental results.

  15. Nanoscopic strength analysis of work-hardened low carbon austenitic stainless steel, 316SS

    International Nuclear Information System (INIS)

    Stress corrosion cracking (SCC) occurs in shrouds and piping of low carbon austenitic stainless steels at nuclear power plants. A work-hardened layer, where the transgranular SCC initiates, is considered to be one of the probable cause for this occurrence. In order to clarify the microstructural characteristics of work-hardened layer at the surface of shrouds or piping, the strengthen analysis of low carbon austenitic stainless steel, 316SS, rolled at the reduction in area, RA, of 10, 20, 30, 40 and 50% at room temperature were conducted on a nanoscopic scale, using an ultra-microhardness tester, TEM and SEM. TEM and SEM observation showed that the microstructural parameters are the dislocation cell size, dcel, coarse slip spacing, lcsl, and austenitic grain size, dγ. Referring 10dcel and 10lcsl, Vickers hardness, HV, corresponding to macro strength was expressed as Hυ=Hυ*bas + Hυ*sol + Hυ*dis + Hυ*cel + Hυ*csl. Hυ*bas (=100) is the base hardness, Hυ*sol is the solid solution strengthening hardness, Hυ*dis is the dislocation strengthening hardness in the dislocation cell, and Hυ*cel and Hυ*csl are the fine grain strengthening hardness due to the dislocation cell and coarse slip. Hυ*sol was about 50, independently of RA. Hυ*dis was zero at RA 30%. Hυ*cel and Hυ*csl increased with increasing in RA and were kept constant at about 50 and 120 at RA=20 and 30%, respectively. It was suggested from these results that all dislocations introduced by rolling might be dissipated for the creation of dislocation cells and coarse slips at RA 30%. (author)

  16. Nanoscopic strength analysis of work-hardened L-grade austenitic stainless steel, 316(NG)

    International Nuclear Information System (INIS)

    Stress corrosion cracking (SCC) occurs in shrouds and piping of L-grade austenitic stainless steels at nuclear power plants. A work-hardened layer, where the transgranular SCC initiates, is considered to be one of the probable cause for this occurrence. In order to clarify the microstructural characteristics of work-hardened layer at the surface of shrouds or piping, the strengthen analysis of L-grade austenitic stainless steel, 316(NG), rolled at the reduction in area, RA, of 10, 20, 30, 40 and 50% at room temperature were conducted on a nanoscopic scale, using an ultra-microhardness tester, TEM and SEM. TEM and SEM observation showed that the microstructural parameters are the dislocation cell size, dcel, coarse slip spacing, lcsl, and austenitic grain size, dγ. Referring 10dcel and 10lcsl, Vickers hardness, Hυ, corresponding to macro strength was expressed as Hυ = Hυ*bas + Hυ*sol + Hυ*dis + Hυ*cel + Hυ*csl. Hυ*bas(=100) is the base hardness, Hυ*sol is the solid solution strengthening hardness, Hυ*dis is the dislocation strengthening hardness in the dislocation cell, and Hυ*cel and Hυ*csl are the fine grain strengthening hardness due to the dislocation cell and coarse slip. Hυ*sol was about 50, independently of RA. Hυ*dis was zero at RA30%. Hυ*cel and Hυ*csl increased with increasing in RA and were kept constant at about 50 and 120 at RA=20 and 30%, respectively. It was suggested from these results that all dislocations introduced by rolling might be dissipated for the creation of dislocation cells and coarse slips at RA30%. (author)

  17. Reactor process water (PW) piping inspections, 1984--1990

    International Nuclear Information System (INIS)

    In July 1983, the NRC ordered the shutdown of five boiling water reactors (BWR's) because of concerns about reliability of ultrasonic examination for detecting intergranular stress corrosion cracking (IGSCC). These concerns arose because of leaking piping at Niagara Mohawk's Nine Mile Point which was attributed to IGSCC. The leaks were detected shortly after completion of ultrasonic examinations of the piping. At that time, the Dupont plant manager at Savannah River (SR) directed that investigations be performed to determine if similar problems could exist in SR reactors. Investigation determined that all conditions believed necessary for the initiation and propagation of IGSCC in austenitic stainless steel exist in SR reactor process water (PW) systems. Sensitized, high carbon, austenitic stainless steel, a high purity water system with high levels of dissolved oxygen, and the residual stresses associated with welding during construction combine to provide the necessary conditions. A periodic UT inspection program is now in place to monitor the condition of the reactor PW piping systems. The program is patterned after NRC NUREG 0313, i.e., welds are placed in categories based on their history. Welds in upgraded or replacement piping are examined on a standard schedule (at least every five years) while welds with evidence of IGSCC, evaluated as acceptable for service, are inspected at every extended outage (15 to 18 months). This includes all welds in PW systems three inches in diameter and above. Welds are replaced when MSCC exceeds the replacement criteria of more than twenty percent of pipe circumference of fifty percent of through-wall depth. In the future, we intend to perform flow sizing with automated UT techniques in addition to manual sizing to provide more information for comparison with future examinations

  18. Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Shack, W. J.

    2008-01-21

    In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heat-affected-zone materials, that were irradiated to fluence levels as high as {approx} 2x 10{sup 21} n/cm{sup 2} (E > 1 MeV) ({approx} 3 dpa) in a light water reactor at 288-300 C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated.

  19. HPFRCC - Extruded Pipes

    DEFF Research Database (Denmark)

    Stang, Henrik; Pedersen, Carsten

    1996-01-01

    The present paper gives an overview of the research onHigh Performance Fiber Reinforced Cementitious Composite -- HPFRCC --pipes recently carried out at Department of Structural Engineering, Technical University of Denmark. The project combines material development, processing technique development......-w$ relationship is presented. Structural development involved definition of a new type of semi-flexiblecement based pipe, i.e. a cement based pipe characterized by the fact that the soil-pipe interaction related to pipe deformation is an importantcontribution to the in-situ load carrying capacity of the pipe...... itself. The structural modeling of the pipe was done making direct use of the $\\sigma-w$ material characterization. The processing technique developed is a novel type of extrusion combiningease of material mixing and few requirements for material pre-processingwith a high degree of accuracy and stability...

  20. HPFRCC - Extruded Pipes

    DEFF Research Database (Denmark)

    Stang, Henrik; Pedersen, Carsten

    The present paper gives an overview of the research onHigh Performance Fiber Reinforced Cementitious Composite -- HPFRCC --pipes recently carried out at Department of Structural Engineering, Technical University of Denmark. The project combines material development, processing technique development......-w$ relationship is presented. Structural development involved definition of a new type of semi-flexiblecement based pipe, i.e. a cement based pipe characterized by the fact that the soil-pipe interaction related to pipe deformation is an importantcontribution to the in-situ load carrying capacity of the pipe...... itself. The structural modeling of the pipe was done making direct use of the $\\sigma-w$ material characterization. The processing technique developed is a novel type of extrusion combiningease of material mixing and few requirements for material pre-processingwith a high degree of accuracy and stability...

  1. Automation in local stress evaluation for pipe support and pipe attachment

    International Nuclear Information System (INIS)

    This paper discusses the methodology and benefit of using a special purpose finite element computer program to compute local stress at the pipe and support attachment. This program may be viewed as a simplified and improved method to perform detailed stress evaluation on piping components, pipe support attachments and pipe-attachment welds. The piping component may be circular run pipe, elbow, or a square tubular steel, and the support attachment may be a circular pipe, rectangular tube, or rectangular solid lug. This approach addresses those situations where the limitations in the Welding Research Council (WRC) Bulletin No. 107 and ASME code cases are exceeded. This paper presents a detailed evaluation of component and attachment stress intensity as compared to various code cases, bulletins, and finite element analyses. This paper also discusses the convergent solution of various mesh sizes in the auto-generated finite element models. Finally, this paper summarizes the overall effectiveness of employing this procedure to achieve automation and realistic results

  2. The influence of fabricating conditions and stability of austenite on forming behaviour of austenitic stainless steels

    International Nuclear Information System (INIS)

    The object of the investigation is the effect of various conditions of cold rolling austenitic stainless steels on the mechanical and technological properties and on the behaviour during forming with requirements in stretching and deep drawing. Fabricating 3 coils of various stability of austenite the degree of cold forming between the annealing processes is varied by cold rolling from the thickness of hot rolled coil to final thickness without or with one or two intermediate annealings. The most important results for cold forming sheets are: most favourable stretch forming behaviour is gained with instable austenitic steels, becomes better with increasing sheet thickness most favourable deep drawing behaviour is gained with highest degrees of cold rolling before final annealing, is undependent from the stability of austenite. Favourable is cold rolling to the highest degree before intermediate annealing, whilst the deformation before final annealing is of greater importance. According to the results conditions can be given for cold rolling to get best forming behaviour. (orig.)

  3. A powder metallurgy austenitic stainless steel for application at very low temperatures

    CERN Document Server

    Sgobba, Stefano; Liimatainen, J; Kumpula, M

    2000-01-01

    The Large Hadron Collider to be built at CERN will require 1232 superconducting dipole magnets operating at 1.9 K. By virtue of their mechanical properties, weldability and improved austenite stability, nitrogen enriched austenitic stainless steels have been chosen as the material for several of the structural components of these magnets. Powder Metallurgy (PM) could represent an attractive production technique for components of complex shape for which dimension tolerances, dimensional stability, weldability are key issues during fabrication, and mechanical properties, ductility and leak tightness have to be guaranteed during operation. PM Hot Isostatic Pressed test plates and prototype components of 316LN-type grade have been produced by Santasalo Powdermet Oy. They have been fully characterized and mechanically tested down to 4.2 K at CERN. The fine grained structure, the absence of residual stresses, the full isotropy of mechanical properties associated to the low level of Prior Particle Boundaries oxides ...

  4. Magnetic State of Deformed Austenite Before and After Martensite Nucleation in Austenitic Stainless Steels

    Institute of Scientific and Technical Information of China (English)

    GennadiiVSnizhnoi; MariyaSRasshchupkyna’

    2012-01-01

    The effect of the increase in the paramagnetic susceptibility of austenite up to the true value of the deformation-induced martensite transition point es has been experimentally established in steels X6CrNiTil8-10 (correspon& ing to AISI 321 steels). At this point nucleation and accumulation of martensite with the increase in the extent of de- formation but at a constant magnetic state of austenite takes place.

  5. High-temperature materials for nuclear power plant piping

    International Nuclear Information System (INIS)

    The authors discuss the properties and problems of austenitic high-temperature steels or Ni alloys used as materials for pipelines with high operating temperatures in nuclear power plants, e.g. sodium-cooled fast breeders (5500C) and high-temperature reactors (7500C or 9500C). Sturcture and properties (mechanical and technical) of materials are described, e.g. cyclic strength, fatigue life, fracture mechanics, corrosion. Unresolved problems, e.g. multiaxial leads on pipe geometries and accumulation of defects at very high temperatures, are discussed. (orig.)

  6. Austenitic stainless steels for cryogenic service

    Energy Technology Data Exchange (ETDEWEB)

    Dalder, E.N.C.; Juhas, M.C.

    1985-09-19

    Presently available information on austenitic Fe-Cr-Ni stainless steel plate, welds, and castings for service below 77 K are reviewed with the intent (1) of developing systematic relationships between mechanical properties, composition, microstructure, and processing, and (2) of assessing the adequacy of these data bases in the design, fabrication, and operation of engineering systems at 4 K.

  7. Corrosion of plasma nitrided austenitic stainless steels

    International Nuclear Information System (INIS)

    The corrosion behaviour of plasma nitrided austenitic stainless steel grades AISI 304, 316 and 321 was studied at various temperatures. Certain plasma nitriding cycles included a post-oxidation treatment. The corrosion rates were measured using linear polarisation technique. Results showed that corrosion rate increased with the plasma nitriding temperature. Minimum deterioration occurred at 653K. (author). 2 tabs., 4 figs., 10 refs

  8. Bainite orientation in plastically deformed austenite

    OpenAIRE

    Klobčar, Damjan; Shirzadi, A. A.; Abreu, H.; Pocock, L.; Withers, P.J.; Bhadeshia, Harshad Kumar Dharamshi Hansraj

    2015-01-01

    Experiments have been conducted to see whether specific crystallographic variants of bainite form in polycrystalline steel when transformation occurs from plastically deformed austenite which is otherwise free from externally applied stress. It is demonstrated by studying both overall and microtexture that there is no perceptible variant selection as bainite forms. Indeed, the texture is found to weaken on transformation.

  9. Austenitic stainless steels for cryogenic service

    International Nuclear Information System (INIS)

    Presently available information on austenitic Fe-Cr-Ni stainless steel plate, welds, and castings for service below 77 K are reviewed with the intent (1) of developing systematic relationships between mechanical properties, composition, microstructure, and processing, and (2) of assessing the adequacy of these data bases in the design, fabrication, and operation of engineering systems at 4 K

  10. Evaluation of aging of cast stainless steel components

    International Nuclear Information System (INIS)

    Cast stainless steel is used extensively in nuclear reactors for primary-pressure-boundary components such as primary coolant pipes, elbows, valves, pumps, and safe ends. These components are, however, susceptible to thermal aging embrittlement in light water reactors because of the segregation of Cr atoms from Fe and Ni by spinodal decomposition in ferrite and the precipitation of Cr-rich carbides on ferrite/austenite boundaries. A recent advance in understanding the aging kinetics is presented. Aging kinetics are strongly influenced by the synergistic effects of other metallurgical reactions that occur in parallel with spinodal decomposition, i.e., clustering of Ni, Mo, and Si solute atoms and the nucleation and growth of G-phase precipitates in the ferrite phase. A number of methods are outlined for estimating aging embrittlement under end-of-life of life-extension conditions, depending on several factors such as degree of permissible conservatism, availability of component archive material, and methods of estimating and verifying the activation energy of aging. 33 refs., 6 figs., 3 tabs

  11. Expanded austenite in nitrided layers deposited on austenitic and super austenitic stainless steel grades; Analise da austenita expandida em camadas nitretadas em acos inoxidaveis austeniticos e superaustenitico

    Energy Technology Data Exchange (ETDEWEB)

    Casteletti, L.C.; Fernandes, F.A.P.; Heck, S.C. [Universidade de Sao Paulo (EESC/USP), Sao Carlos, SP (Brazil). Escola de Engenharia. Dept. de Engenharia de Materais, Aeronautica e Automobilistica; Oliveira, A.M. [Instituto de Educacao, Ciencia e Tecnologia do Maranhao (IFMA), Sao Luis, MA (Brazil); Gallego, J., E-mail: gallego@dem.feis.unesp.b [UNESP, Ilha Solteira, SP (Brazil). Dept. Engenharia Mecanica

    2010-07-01

    In this work nitrided layers deposited on austenitic and super austenitic stainless steels were analyzed through optical microscopy and X-rays diffraction analysis (XRD). It was observed that the formation of N supersaturated phase, called expanded austenite, has promoted significant increment of hardness (> 1000HV). XRD results have indicated the anomalous displacement of the diffracted peaks, in comparison with the normal austenite. This behavior, combined with peaks broadening, it was analyzed in different nitriding temperatures which results showed good agreement with the literature. (author)

  12. Aspects and mechanisms of austenitic stainless steel corrosion in case of sodium leaks under mineral wool insulation

    International Nuclear Information System (INIS)

    Sodium pipe rupture tests representative of Fast Reactors Accidents have been carried out on austenitic stainless steel surfaces. These tests improve our knowledge of small sodium leakage propagation in mineral wool insulation. They explain the new and unexpected aspects of the crevice corrosion phenomenon which has been observed on austenitic stainless steel pipe surfaces. Experimental results show that corrosion is limited to a peripheral annular zone, which extends out in concentric waves. The diameter of this corrosion zone is practically constant. Tests show that sodium does not expand directly on the pipe surface. Sodium sprays through mineral wool insulation, where chemical reaction between silica fibers, occluded oxygen and water vapor occur at the same time. Simultaneously, there is a diffusion phenomenon of liquid Na droplets on the mineral wool fibers. The study allows to prove the electrochemical nature of the corrosion. The excess liquid Na, spraying as droplets induces an anodic dissolution mechanism by differential aeration. This phenomenon explains the random microscopic and macroscopic aspects of material removal. (authors). 1 ref., 16 figs

  13. Protective coating of austenitic steel using robotized GMAW temper-bead technique; Rechargement d'inox austenitique en MAG temperbead robotise

    Energy Technology Data Exchange (ETDEWEB)

    Carpreau, J.M. [Electricite de France (EDF/R and D), Recherche et Developpement, 92 - Chatou (France); Dainelli, P. [Institut de Soudure, 57 - Yutz (France)

    2009-07-15

    This paper summarises experimental results obtained in a study of GMAW temper-bead on low alloyed steel with austenitic consumables. Temper-bead on low alloyed steel with austenitic consumables is mainly used for repairing operations of heavy components such as vessel reactor of nuclear power plants. Experimental work aims at showing the performance of GMAW compared to GTAW and the consequences of GMAW temper-bead on 2OMND5 heat affected zones. (authors)

  14. Analysis of the optimization of the secondary hot piping for a sodium fast reactor

    International Nuclear Information System (INIS)

    Mod. 9Cr-1Mo steel (T91) is a candidate material for Sodium Fast Reactor (SFR) components and in particular for secondary hot piping. As compared to austenitic stainless steels used in the past reactors, 9Cr-1Mo steel's good conductivity and low thermal expansion let the possibility to reduce the size of the loops and thus to gain on the costs. In order to validate this choice, it is necessary, firstly to verify that this alloy can resist the planned environmental and operating conditions, secondly to check its supply, fabrication and welding possibilities and finally to ensure that the existing design codes cover mechanical design rules. A large R and D program on mod. 9Cr-1Mo steel has been undertaken in France, in order to characterize the behavior of this material and of its welded junctions in operating conditions representative of SFR. In this program, a numerical analysis on secondary hot piping design has been carried out using a stainless steel 316L(N) (used in the previous SFRs Phenix and Super Phenix) and a mod. 9Cr-1Mo steel. The aim of this study was to optimize the secondary hot piping by minimizing the size of the loop and by comparing both candidate materials. This analysis deals with the secondary piping considered for the European Fast Reactor (EFR) and the design has been made for realistic operating conditions of EFR for a period of 60 years. The analysis is based on the creep-fatigue damage and the application of the RCC-MR rules. The results show that the use of mod. 9Cr-1Mo steel has generally an advantage for moderate temperature (below 525 deg. C). On the contrary, when the temperature is more important, stainless steel 316L(N) presents lower damage than 9Cr steel. Indeed, thanks to advantageous thermal properties of mod 9Cr-1Mo steel, the stress state due to mechanical and thermal loading for this material is 20 to 30% lower than this of 316L(N) stainless steel. But at high temperatures this benefit is too low to compensate for the lower

  15. Evaluation of austenitic alloys abrasive wear of FeMnAlC system; Avaliacao de desgaste abrasivo de ligas austeniticas do sistema FeMnAlC

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Allan Ribeiro de; Acselrad, Oscar [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Metalurgica e de Materiais. Lab. de Processamento Termomecanico e Engenharia Microestrutural]. E-mail: allariba@metalmat.ufrj.br

    2003-07-01

    Alloys of the FeMnAlC system have been studied as an alternative to stainless steels applications. Such alloys, when solubilized, are non-magnetic and present an austenitic structure that can be modified by thermal treatments. In this way, a large spectrum of mechanical and physical properties can be obtained. They are oxidation-resistant alloys, and by 15 hours aging at 550 deg C mechanical strength can be as high as conventional structural alloy steels. Information concerning the performance of these alloys under wear conditions are still limited. The possibility of application in components exposed to cavitation or abrasive loads, such as pipes, pumps and drilling systems is still a subject for fundamental research, such as the one that is now reported. Samples of a FeMnAlC alloy have been submitted to different thermal processing, leading to microstructures that have been characterized by optical, transmission and atomic force microscopy and by X-ray diffraction. They were subsequently subjected to a micro-abrasion test in which the abrasive wear resistance could be determined. The results have been used to differentiate the performance of different microstructures and allowed also a comparative analysis with the performance of an AISI M2 tool steel. (author)

  16. Long time high temperature strain gage measurements on pipes and dissimilar welds

    International Nuclear Information System (INIS)

    In the dissimilar weld program, a pipe consisting of several ferritic (X20 CrMoV 121) and Austenitic (X10 NiCrAlTi 32 20) pipe sections with dissimilar welds is exposed to combined internal pressure temperature loads. In long tern experiments lasting 20,000 hours, both capacitative high temperature strain gauges and resistance HT DMS are used to measure strain. There is then a check of the long term stability of the measuring system. The experiments are carried out for proving the integrity of the water/steam circuit of an HTR. (DG)

  17. A fatigue analysis including environmental effects for a pipe system in a Swedish BWR

    International Nuclear Information System (INIS)

    A BWR feed water piping system (austenitic steel) has been analyzed with two different fatigue curves and environmental factors. Original fatigue curve from ASME is compared to a new fatigue curve; ANL. The influence of environmental correction factors (Fen) is studied further for the piping system. It is noted that the results apply for this particular system, and general conclusions should be cautiously drawn. Typical for this system is that all dominant loads are within the low-cycle regime. This implies that the change of fatigue curve only leads to limited increases in usage factors. Larger changes can occur if larger number of cycles is within the high-cycle regime

  18. Heat Pipe Planets

    Science.gov (United States)

    Moore, William B.; Simon, Justin I.; Webb, A. Alexander G.

    2014-01-01

    When volcanism dominates heat transport, a terrestrial body enters a heat-pipe mode, in which hot magma moves through the lithosphere in narrow channels. Even at high heat flow, a heat-pipe planet develops a thick, cold, downwards-advecting lithosphere dominated by (ultra-)mafic flows and contractional deformation at the surface. Heat-pipes are an important feature of terrestrial planets at high heat flow, as illustrated by Io. Evidence for their operation early in Earth's history suggests that all terrestrial bodies should experience an episode of heat-pipe cooling early in their histories.

  19. Heat Pipe Materials Compatibility

    Science.gov (United States)

    Eninger, J. E.; Fleischman, G. L.; Luedke, E. E.

    1976-01-01

    An experimental program to evaluate noncondensable gas generation in ammonia heat pipes was completed. A total of 37 heat pipes made of aluminum, stainless steel and combinations of these materials were processed by various techniques, operated at different temperatures and tested at low temperature to quantitatively determine gas generation rates. In order of increasing stability are aluminum/stainless combination, all aluminum and all stainless heat pipes. One interesting result is the identification of intentionally introduced water in the ammonia during a reflux step as a means of surface passivation to reduce gas generation in stainless-steel/aluminum heat pipes.

  20. Introduction to Heat Pipes

    Science.gov (United States)

    Ku, Jentung

    2015-01-01

    This is the presentation file for the short course Introduction to Heat Pipes, to be conducted at the 2015 Thermal Fluids and Analysis Workshop, August 3-7, 2015, Silver Spring, Maryland. NCTS 21070-15. Course Description: This course will present operating principles of the heat pipe with emphases on the underlying physical processes and requirements of pressure and energy balance. Performance characterizations and design considerations of the heat pipe will be highlighted. Guidelines for thermal engineers in the selection of heat pipes as part of the spacecraft thermal control system, testing methodology, and analytical modeling will also be discussed.

  1. Analyses of magnetic field in spiral steel pipe

    International Nuclear Information System (INIS)

    In order to confirm the feasibility of identifying the girth welds using the magnetic field in spiral pipelines, the distributions of the magnetic field in spiral steel pipes with different sizes and different magnetizations were analyzed using the equivalent magnetic charge method, and were verified experimentally. The magnetic field inside spiral steel pipes is generally uniform with very small magnetic sudden changes at the spiral welds, whereas the magnetic field near the pipe ends has very big local changes. The size of spiral pipes, including its wall thickness, length, diameter, and the lift-off, has various influences on the local magnetic sudden changes at the spiral welds (LMASW) and the magnetic incremental near the pipe ends (MINPE), whereas the difference between LMASW and MINPE is always quite considerable. The bigger the radial magnetization component is, the bigger the difference between LMASW and MINPE is. When the radial magnetization component is small, changes of the circumferential and axial magnetization components can reduce this difference. Since the magnetizations of each pipe are seldom identical, the magnetic field inside each pipe is usually quite different. Thus there will be a big local magnetic sudden change at the girth weld inside the spiral pipeline, and this sudden change is much stronger than LMASW. Therefore, we can still consider identifying the girth welds using the magnetic field in spiral pipelines to improve the positioning accuracy of the in-pipe detector. - Highlights: • An analyzing method of the magnetic field in spiral steel pipe is proposed. • Magnetic field in spiral steel pipe was analyzed and verified experimentally. • Magnetic sudden change near pipe end is much bigger than that near spiral weld. • We can identify girth weld using this sudden change to locate in-pipe detector

  2. Development and Implementation of an Effective and Permanent Solution for Resolving Flow-Assisted Corrosion Problems in High-Pressure Turbine Cross-Under Piping at the Krsko Nuclear Power Plant

    International Nuclear Information System (INIS)

    Under certain conditions, the flow-accelerated corrosion process can lead to catastrophic thinning of the walls in low-alloy steel pipes exposed to flowing water or wet steam. The rate of metal loss depends on complex interactions between various factors such as water chemistry, pipe material composition and fluid dynamics. PCI Energy Services - a WEC Welding and Machining LLC / Westinghouse Electric company, developed a highly effective, reliable welding solution for permanently resolving flow-assisted corrosion problems in turbine cross-under piping systems by applying a protective layer of austenitic stainless steel on the inner surface of these components. This technology was successfully implemented at a number of nuclear power plants in the US, and most recently at the Krsko plant in Slovenia. Our paper describes this life-extension process, the engineering analysis performed by PCI/Westinghouse in order to support and validate it, and the steps leading to its successful implementation at Krsko with a multi-cultural team consisting of American, Belgian, Slovenian and Croatian engineers and technicians.(author)

  3. Application of narrow groove welding process to nuclear pipes

    International Nuclear Information System (INIS)

    Experiments on narrow groove welding with a single string bead deposition per layer were performed using automatic orbital TIG welding equipment, and the narrow groove shape and welding conditions were optimized for stainless steel and carbon steel pipes. The characteristics of narrow groove weld joints of these materials were investigated in the areas of metallurgical structure and mechanical properties. The process of one bead per layer was found to produce a good homogeneous weld and the total weld has the same micro structure between two regular fusion lines and, therefore, uniform mechanical properties. Based on these test results, the narrow groove welding process was applied to butt weld joints for austenitic stainless steel pipe with a large diameter. (author)

  4. Production of Austenitic Steel for the LHC Superconducting Dipole Magnets

    CERN Document Server

    Bertinelli, F; Komori, T; Peiro, G; Rossi, L

    2006-01-01

    The austenitic-steel collars are an important component of the LHC dipole magnets, operating at cryogenic temperature under high mechanical stress. The required steel, known as YUS 130S, has been specifically developed for this application by Nippon Steel Corporation (NSC), who was awarded a CERN contract in 1999 for the supply of 11 500 tonnes. In 2005 - after six years of work - the contract is being successfully completed, with final production being ensured since October 2003 by Nippon Steel & Sumikin Stainless Steel Corporation (NSSC). The paper describes the steel properties, its manufacturing and quality control process, organization of production, logistics and contract follow-up. Extensive statistics have been collected relating to mechanical, physical and technological parameters. Specific attention is dedicated to measurements of magnetic permeability performed at cryogenic temperatures by CERN, the equipment used and statistical results. Reference is also made to the resulting precision of the...

  5. Formability analysis of austenitic stainless steel-304 under warm conditions

    Science.gov (United States)

    Lade, Jayahari; Singh, Swadesh Kumar; Banoth, Balu Naik; Gupta, Amit Kumar

    2013-12-01

    A warm deep drawing process of austenitic stainless steel-304 (ASS-304) of circular blanks with coupled ther mal analysis is studied in this article. 65 mm blanks were deep drawn at different temperatures and thickness distribution is experimentally measured after cutting the drawn component into two halves. The process is simulated using explicit fin ite element code LS-DYNA. A Barlat 3 parameter model is used in the simulation, as the material is anisotropic up to 30 0°C. Material properties for the simulation are determined at different temperatures using a 5 T UTM coupled with a furn ace. In this analysis constant punch speed and variable blank holder force (BHF) is applied to draw cups without wrinkle.

  6. Reeling of tight fit pipe

    OpenAIRE

    Focke, E.S.

    2007-01-01

    If it would be possible to install Tight Fit Pipe by means of reeling, it would be an attractive new option for the exploitation of offshore oil and gas fields containing corrosive hydrocarbons. Tight Fit Pipe is a mechanically bonded double walled pipe where a corrosion resistant alloy liner pipe is mechanically fitted inside a carbon steel outer pipe through a thermo-hydraulic manufacturing process. Reeling is a fast method of offshore pipeline installation where a pipe is spooled on a reel...

  7. Gas explosions in process pipes

    OpenAIRE

    Kristoffersen, Kjetil

    2004-01-01

    In this thesis, gas explosions inside pipes are considered. Laboratory experiments and numerical simulations are the basis of the thesis. The target of the work was to study gas explosions in pipes and to develop numer- ical models that could predict accidental gas explosions inside pipes. Experiments were performed in circular steel and plexiglass pipes. The steel pipes have an inner diameter of 22.3 mm and lengths of 1, 2, 5 and 11 m. The plexiglass pipe has an inner diame...

  8. Technology of processing furnaces for refining and petrochemistry. Criteria for the choice of materials for pipes of bundles

    Energy Technology Data Exchange (ETDEWEB)

    Pingeot, M. (ENSEEG, Grenoble (France))

    1981-12-01

    The present state of technology is examined for the determination of steel types as a function of service conditions: temperature, pressure and corrosion. Austenitic structure, grain size, carbon content and additional elements are studied for creep resistance at a temperature of 1000 and a pressure up to 300 atmospheres. Influence of hydrogen, sulfhydric acid, polythionic acids, naphtenic acids and carburation on internal corrosion of pipes is examined and also oxidation and attack by fuel oil ashes of the external surfaces for different types of steels. Precautions to be taken for welding of chromium mobybdenum steels and austenitic stainless steels are indicated.

  9. Thermal fatigue cracking of austenitic stainless steels

    International Nuclear Information System (INIS)

    This report deals with the thermal fatigue cracking of austenitic stainless steels as AISI 316 LN and 304 L. Such damage has been clearly observed for some components used in Fast Breeder reactors (FBR) and Pressure Water Reactor (PWR). In order to investigate thermal fatigue, quasi-structural specimen have been used. In this frame, facilities enforcing temperature variations similar to those found under the operation conditions have been progressively developed. As for components, loading results from impeded dilatation. In the SPLASH facility, the purpose was to establish accurate crack initiation conditions in order to check the relevance of the usual component design methodology. The tested specimen is continuously heated by the passage of an electrical DC current, and submitted to cyclic thermal down shock (up to 1000 deg C/s) by means of periodical spraying of water on two opposite specimen faces. The number of cycles to crack initiation Ni is deduced from periodic examinations of the quenched surfaces, by means of optical microscopy. It is considered that initiation occurs when at least one 50μm to 150□m long crack is observed. Additional SPLASH tests were performed for N >> Ni, with a view to investigate the evolution of a surface multiple cracking network with the number of cycles N. The CYTHIA test was mainly developed for the purpose of assessing crack growth dynamics of one isolated crack in thermal fatigue conditions. Specimens consist of thick walled tubes with a 1 mm circular groove is spark-machined at the specimen centre. During the test, the external wall of the tube is periodically heated by using a HF induction coil (1 MHz), while its internal wall is permanently cooled by flowing water. Total crack growth is derived from post-mortem examinations, whereby the thermal fatigue final rupture surface is oxidized at the end of the test. The specimen is broken afterwards under mechanical fatigue at room temperature. All the tests confirm that usual

  10. Characterization of microstructure and texture across dissimilar super duplex/austenitic stainless steel weldment joint by super duplex filler metal

    Energy Technology Data Exchange (ETDEWEB)

    Eghlimi, Abbas, E-mail: a.eghlimi@ma.iut.ac.ir [Department of Materials Engineering, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Shamanian, Morteza [Department of Materials Engineering, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Eskandarian, Masoomeh [Department of Materials Engineering, Shiraz University, Shiraz 71348-51154 (Iran, Islamic Republic of); Zabolian, Azam [Department of Natural Resources, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Szpunar, Jerzy A. [Department of Mechanical Engineering, University of Saskatchewan, Saskatoon SK S7N 5A9 (Canada)

    2015-08-15

    In the present paper, microstructural changes across an as-welded dissimilar austenitic/duplex stainless steel couple welded by a super duplex stainless steel filler metal using gas tungsten arc welding process is characterized with optical microscopy and electron back-scattered diffraction techniques. Accordingly, variations of microstructure, texture, and grain boundary character distribution of base metals, heat affected zones, and weld metal were investigated. The results showed that the weld metal, which was composed of Widmanstätten austenite side-plates and allotriomorphic grain boundary austenite morphologies, had the weakest texture and was dominated by low angle boundaries. The welding process increased the ferrite content but decreased the texture intensity at the heat affected zone of the super duplex stainless steel base metal. In addition, through partial ferritization, it changed the morphology of elongated grains of the rolled microstructure to twinned partially transformed austenite plateaus scattered between ferrite textured colonies. However, the texture of the austenitic stainless steel heat affected zone was strengthened via encouraging recrystallization and formation of annealing twins. At both interfaces, an increase in the special character coincident site lattice boundaries of the primary phase as well as a strong texture with <100> orientation, mainly of Goss component, was observed. - Graphical abstract: Display Omitted - Highlights: • Weld metal showed local orientation at microscale but random texture at macroscale. • Intensification of <100> orientated grains was observed adjacent to the fusion lines. • The austenite texture was weaker than that of the ferrite in all duplex regions. • Welding caused twinned partially transformed austenites to form at SDSS HAZ. • At both interfaces, the ratio of special CSL boundaries of the primary phase increased.

  11. Characterization of microstructure and texture across dissimilar super duplex/austenitic stainless steel weldment joint by super duplex filler metal

    International Nuclear Information System (INIS)

    In the present paper, microstructural changes across an as-welded dissimilar austenitic/duplex stainless steel couple welded by a super duplex stainless steel filler metal using gas tungsten arc welding process is characterized with optical microscopy and electron back-scattered diffraction techniques. Accordingly, variations of microstructure, texture, and grain boundary character distribution of base metals, heat affected zones, and weld metal were investigated. The results showed that the weld metal, which was composed of Widmanstätten austenite side-plates and allotriomorphic grain boundary austenite morphologies, had the weakest texture and was dominated by low angle boundaries. The welding process increased the ferrite content but decreased the texture intensity at the heat affected zone of the super duplex stainless steel base metal. In addition, through partial ferritization, it changed the morphology of elongated grains of the rolled microstructure to twinned partially transformed austenite plateaus scattered between ferrite textured colonies. However, the texture of the austenitic stainless steel heat affected zone was strengthened via encouraging recrystallization and formation of annealing twins. At both interfaces, an increase in the special character coincident site lattice boundaries of the primary phase as well as a strong texture with <100> orientation, mainly of Goss component, was observed. - Graphical abstract: Display Omitted - Highlights: • Weld metal showed local orientation at microscale but random texture at macroscale. • Intensification of <100> orientated grains was observed adjacent to the fusion lines. • The austenite texture was weaker than that of the ferrite in all duplex regions. • Welding caused twinned partially transformed austenites to form at SDSS HAZ. • At both interfaces, the ratio of special CSL boundaries of the primary phase increased

  12. Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and altered mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs

  13. Modeling of Ni Diffusion Induced Austenite Formation in Ferritic Stainless Steel Interconnects

    OpenAIRE

    Chen, Ming; Molin, Sebastian; Zhang, L; Ta, Na; Hendriksen, Peter Vang; Kiebach, Wolff-Ragnar; Y. Du

    2015-01-01

    Ferritic stainless steel interconnect plates are widely used in planar solid oxide fuel cell (SOFC) or electrolysis cell (SOEC) stacks. During stack production and operation, nickel from the Ni/YSZ fuel electrode or from the Ni contact component diffuses into the IC plate, causing transformation of the ferritic phase into an austenitic phase in the interface region. This is accompanied with changes in volume and in mechanical and corrosion properties of the IC plates. In this work, kinetic mo...

  14. Piping research program plan

    International Nuclear Information System (INIS)

    This document presents the piping research program plan for the Structural and Seismic Engineering Branch and the Materials Engineering Branch of the Division of Engineering, Office of Nuclear Regulatory Research. The plan describes the research to be performed in the areas of piping design criteria, environmentally assisted cracking, pipe fracture, and leak detection and leak rate estimation. The piping research program addresses the regulatory issues regarding piping design and piping integrity facing the NRC today and in the foreseeable future. The plan discusses the regulatory issues and needs for the research, the objectives, key aspects, and schedule for each research project, or group of projects focussing of a specific topic, and, finally, the integration of the research areas into the regulatory process is described. The plan presents a snap-shot of the piping research program as it exists today. However, the program plan will change as the regulatory issues and needs change. Consequently, this document will be revised on a bi-annual basis to reflect the changes in the piping research program. (author)

  15. These Pipes Are "Happening"

    Science.gov (United States)

    Skophammer, Karen

    2010-01-01

    The author is blessed with having the water pipes for the school system in her office. In this article, the author describes how the breaking of the pipes had led to a very worthwhile art experience for her students. They practiced contour and shaded drawing techniques, reviewed patterns and color theory, and used their reasoning skills--all while…

  16. Transients in pipes

    International Nuclear Information System (INIS)

    The motion of a fluid in a pipe is commonly modeled utilizing the one space dimension conservation laws of mass and momentum. The development of shocks and spikes utilizing the uniform sampling method is studied. The effects of temperature variations and friction are compared for gas pipes. (Author)

  17. Heat pipe technology issues

    International Nuclear Information System (INIS)

    Critical high temperature, high power applications in space nuclear power designs are near the current state of the art of heat pipe technology in terms of power density, operating temperature, and lifetime. Recent heat pipe development work at Los Alamos National Laboratory has involved performance testing of typical space reactor heat pipe designs to power levels in excess of 19 kW/cm2 axially and 300 W/cm2 radially at temperatures in the 1400 to 1500 K range. Operation at conditions in the 10 kW/cm2 range has been sustained for periods of up to 1000 hours without evidence of performance degradation. The effective length for heat transport in these heat pipes was from 1.0 to 1.5 M. Materials used were molybdenum alloys with lithium employed as the heat pipe operating fluid. Shorter, somewhat lower power, molybdenum heat pipes have been life tested at Los Alamos for periods of greater than 25,000 hours at 1700 K with lithium and 20,000 hours at 15000K with sodium. These life test demonstrations and the attendant performance limit investigations provide an experimental basis for heat pipe application in space reactor design and represent the current state-of-the-art of high temperature heat pipe technology

  18. The evaluation of erosion-corrosion problems in Taiwan PWR carbon steel piping

    International Nuclear Information System (INIS)

    Taiwan PWR Nuclear Power Plant Units 1 and 2 implemented the projects of Pipe Wall Thinning Measurement under the request of ROCAEC to prevent the events due to the piping erosion/corrosion. The purpose of this paper is to present the improvements in the evaluation method for the identification of the potential piping systems and components in PWR

  19. Seismic design technology for breeder reactor structures. Volume 4. Special topics in piping and equipment

    International Nuclear Information System (INIS)

    This volume is divided into five chapters: experimental verification of piping systems, analytical verification of piping restraint systems, seismic analysis techniques for piping systems with multisupport input, development of floor spectra from input response spectra, and seismic analysis procedures for in-core components

  20. Grain boundary strengthening in austenitic nitrogen steels

    International Nuclear Information System (INIS)

    The effect of nitrogen and carbon on the strengthening of the austenitic steel Cr18Ni16Mn10 by grain boundaries is studied. It is established in accordance with previous results that contrary to carbon nitrogen increases the coefficient k in the Hall-Petch equation markedly. Because of a pronounced planar slip induced by nitrogen and the absence of any noticeable segregation of nitrogen atoms at the grain boundaries, nitrogen austenite presents an excellent object for testing different existing models of grain boundary strengthening (pile-up, grain boundary dislocation sources, work hardening theories). Based on the analysis of available data and measurements of interaction between nitrogen (carbon) atoms and dislocations it is shown that the nitrogen effect can be attributed to a strong blocking of dislocation sources in grains adjacent to those where the slip started. (orig.)

  1. Reliability of welded austenitic stainless steel containing base metal delta ferrite

    Energy Technology Data Exchange (ETDEWEB)

    Shalaby, Hamdy M. [Kuwait Institute for Scientific Research (Kuwait)

    2004-07-01

    The paper presents the results of a failure case study carried out on welded 304L stainless steel (SS) pipeline of waste gas header (WGH). The environment inside the WGH was mainly wet steam with hydrocarbons, H{sub 2}S, oxygen, CO{sub 2}, organic acids, and organic chlorides. The outside pipe wall temperature was 91-97 deg C. The failure of the pipe was at the heat-affected zone (HAZ). The study was made on four welded pipeline samples, three of which were in service. The pipe samples were welded using three different techniques that included autogenous gas tungsten arc, shielded metal arc, and flux core arc. The investigation revealed that cracking at HAZ was due to base metal delta ferrite decay accompanied with sigma phase formation due to high heat input during welding. However, the morphology and orientation of the cracks suggested that stress-rupture and stress corrosion cracking had occurred. The presence of base metal delta ferrite made all used welding procedures un-successful. The study concluded that utilization of delta ferrite free austenitic SS should eliminate the problem. (author)

  2. Influence of damage rate on physical and mechanical properties and swelling of 18Cr-9Ni austenitic steel in the range of 3.10{sup -9} to 4.10{sup -8} dpa/sec

    Energy Technology Data Exchange (ETDEWEB)

    Shcherbakov, E.N.; Kozlov, A.V.; Yagovitin, R.I.; Evseev, M.V.; Kinev, E.A. [FSUE, Institute of Nuclear Materials, Zarechney, Sverdlovsk (Russian Federation); Isobe, Y.; Sagisaka, M. [Nuclear Fuel Industries, Osaka (Japan); Okita, T.; Sekimura, N. [University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, 113-8654 Tokyo (Japan); Garner, F. [Pacific Northwest National Laboratory, P.O. Box 999, Richland WA, AK 99352 (United States)

    2007-07-01

    Full text of publication follows: Whereas most data on radiation-induced changes in mechanical properties or dimensional stability needed for fusion - relevant dpa levels and dpa rates are generated at relatively high neutron flux in fast reactors, many fusion and fission components will operate at much lower dpa rates. Much less data are available from long-lived structural components operating at very low flux levels. In addition most published data were generated from relatively thin specimens ({approx}1-2 mm or less), while some actual fusion structural components can be on the order of 1-2 cm thick. In this study we have examined a 9 cm diameter pipe constructed from Fe-18Cr-9Ni steel analogous to AISI 304 that stayed outside the core of BN-600 for 22 years. The walls of the pipe were 2 cm thick and experienced temperatures in the range 370-375 deg. C. The walls were sectioned into 5 slices at a number of positions to yield doses in the range 1.5 to 22 dpa at 3 x 10{sup -9} to 4 x 10{sup -8} dpa/s. Changes in elastic moduli were studied using an ultrasonic technique and changes in electrical resistivity and mechanical properties of the 18Cr9Ni austenitic steel was examined. Swelling was measured both by immersion density and electron microscopy, reaching a maximum of {approx}3 %. Swelling appears to be accelerated somewhat at these lower dpa rates as observed in other recent studies. Tensile properties were also measured. Radiation-induced changes of electrical resistivity, Young's and shear moduli were observed but did not agree fully with predictions based on voids alone. Strong contributions from second phase precipitates were found to be contributing to changes in both physical and mechanical properties. (authors)

  3. The effect of retained austenite on steel fracture toughness

    International Nuclear Information System (INIS)

    This paper is an attempt of reviewing the outlooks about the favourable influence of retained austenite on fracture toughness of tool steels. The tests were performed on the samples made of the new 70HG2MF steel in which the fraction of retained austenite was changed by subquenching or by changing the austenitizing temperature. It was revealed that in the subquenched samples retained austenite affects strongly the increase of fracture toughness. On the other hand, however, in the samples austenitized at growing temperatures, the effect of this phase on fracture toughness is not so univocal since not only the volume fraction of retained austenite is subjected to changes but also the character of fractures and the grain size. (author)

  4. Pitting corrosion resistant austenite stainless steel

    Science.gov (United States)

    van Rooyen, D.; Bandy, R.

    A pitting corrosion resistant austenite stainless steel comprises 17 to 28 wt. % chromium, 15 to 26 wt. % nickel, 5 to 8 wt. % molybdenum, and 0.3 to 0.5 wt. % nitrogen, the balance being iron, unavoidable impurities, minor additions made in the normal course of melting and casting alloys of this type, and may optionally include up to 10 wt. % of manganese, up to 5 wt. % of silicon, and up to 0.08 wt. % of carbon.

  5. Corrosion of austenitic steel in sodium loops

    International Nuclear Information System (INIS)

    The possibility of predicting corrosion effects for austenitic steel exposed to liquid sodium with an analytical diffusion model is presented. The analytically predicated corrosion effects are compared with experimental measurements of corrosion effects achieved in an accurately controlled sodium loop. The diffusion model is described with figures showing disc sample weight loss and sodium flow guidance tube chromium and nickel profiles. Finally, the concentration profile in the fuel rod wall (diffusion model) is presented for iron, chromium and nickel

  6. Oscillating heat pipes

    CERN Document Server

    Ma, Hongbin

    2015-01-01

    This book presents the fundamental fluid flow and heat transfer principles occurring in oscillating heat pipes and also provides updated developments and recent innovations in research and applications of heat pipes. Starting with fundamental presentation of heat pipes, the focus is on oscillating motions and its heat transfer enhancement in a two-phase heat transfer system. The book covers thermodynamic analysis, interfacial phenomenon, thin film evaporation,  theoretical models of oscillating motion and heat transfer of single phase and two-phase flows, primary  factors affecting oscillating motions and heat transfer,  neutron imaging study of oscillating motions in an oscillating heat pipes, and nanofluid’s effect on the heat transfer performance in oscillating heat pipes.  The importance of thermally-excited oscillating motion combined with phase change heat transfer to a wide variety of applications is emphasized. This book is an essential resource and learning tool for senior undergraduate, gradua...

  7. Residual stress simulation in thin and thick-walled stainless steel pipe welds including pipe diameter effects

    International Nuclear Information System (INIS)

    In this paper, residual stresses in welded components are discussed and a brief review of weld simulation is presented. The general methodology of the FE analysis methods used for welded sections of steel pipes is explained. FE analyses are performed for two axisymmetric butt welds in stainless steel pipes having a 4-pass or a 36-pass weld in a pipe with a wall thickness of 7.1 or 40.0 mm, respectively. In addition, more FE models with inside radius to wall thickness ratio ranging from 1 to 100 have been analysed to investigate the effect of pipe diameter on residual stresses. Residual axial and hoop stresses are plotted for the considered range of pipe diameters for the two simulated pipe wall thicknesses and the differences are discussed

  8. Criterions of UT thickness measurement on thinned pipe management program

    International Nuclear Information System (INIS)

    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 piping 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. Since the mid-1990s, secondary side piping systems in Korean NPPs have experienced wall thinning, leakages and ruptures caused by FAC. Korea Electric Power Research Institute (KEPRI) and Korea Hydro and Nuclear Power Co., LTD. (KHNP) have conducted a study to develop the methodology for systematic pipe management and as a result, established the Korean thinned pipe management program (TPMP) which is being implemented to all Korean NPPs. TPMP consists of several technical elements such as 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. To effectively monitor and manage the thinning pipe components, NDE person as well as FAC engineer should understand the criterions of ultrasonic thickness measurement and there background. This paper describes the technical items of TPMP and the basis of thickness measurement criterions. This paper also shows the initial thickness variations which influence wear and wear rate calculations to obtain the reasonable integrity assessment results. (orig.)

  9. Electron beam welding of austenitic stainless steel

    International Nuclear Information System (INIS)

    Austenitic stainless steel is used for liquid metal-cooled fast breeder reactors with operating temperature of about 550 deg C, because its elevated temperature properties are excellent and the results of use are abundant. The welded joints in LMFBRs require high degree of safety, and the application of electron beam welding is studied to make welding joints of high quality. When the inelastic deformation in a certain limit is allowed as prescribed in the ASME Code, Case 1592, the elevated temperature properties of the welded joints of structures are particularly important. The materials tested were 10 mm thick plates of SUS 304, SUS 316 and SUS 321 steels, and 150 kV - 40 mA electron beam welder was employed. The method of welding was one side, one pass Uranami welding, and first, the appropriate welding conditions were decided. Elevated temperature tensile test was carried out on the parent materials and welded joints by electron beam welding and coated arc welding. Creep rupture test and elevated temperature fatigue test were also carried out. In EB-welded austenitic stainless steel, delta ferrite is scattered finely in austenite, and its quantity is very small and less than 1.5%. The tensile strength and 0.2% proof stress of EB-welded joints are almost same as those of parent materials. The creep rupture and fatigue properties of the joints are also close to those of parent materials. (Kako, I.)

  10. Precipitation effects in austenitic stainless weld metals

    International Nuclear Information System (INIS)

    Creep-rupture specimen of similar welded joints of X6CrNi 18 11 (1.4948/AISI 304) and X6CrNiMo 17 13 (1.4919/AISI 316) show only low elongation after longtime testing. The reason for this loss of ductility was found by metallographic investigations. The weld metal of five joints had ferrite contents from 3 to 7.5%, due to a primary ferritic solidification. During creep testing in the temperature range from 500 to 800deg C carbide precipitation takes place at the austenite-ferrite grain boundaries, because delta ferrite is not in a state of equilibrium at these temperatures. After carbon has been used up, the remaining delta ferrite changes into sigma phase, if its alloying element content is high enough. In the upper temperature range, coagulation of sigma phase is dominating. At these large particles grain boundary migration is hindered. During the grain boundary sliding sigma phase particles break and initiate creep cracks. In fully austenintic weld metal sigma precipitation starts at austenite - austenite grain boundaries. During creep testing, void formation starts at sigma particles. Growing of voids leads to grain separations. (orig.)

  11. Experiments in turbulent pipe flow

    Energy Technology Data Exchange (ETDEWEB)

    Torbergsen, Lars Even

    1998-12-31

    This thesis reports experimental results for the mean velocity and turbulence statistics in two straight pipe sections for bulk Reynolds numbers in the range 22000 to 75000. The flow was found consistent with a fully developed state. Detailed turbulence spectra were obtained for low and moderate turbulent Reynolds number. For the pipe centre line location at R{sub {lambda}} = 112, a narrow range in the streamwise power spectrum applied to the -5/3 inertial subrange. However this range was influenced both by turbulence production and viscous dissipation, and therefore did not reflect a true inertial range. The result indicates how the intermediate range between the production and dissipative scales can be misinterpreted as an inertial range for low and moderate R{sub {lambda}}. To examine the universal behaviour of the inertial range, the inertial scaling of the streamwise power spectrum is compared to the inertial scaling of the second order longitudinal velocity structure function, which relate directly by a Fourier transform. Increasing agreement between the Kolmogorov constant C{sub K} and the second order structure function scaling constant C{sub 2} was observed with increasing R{sub {lambda}}. The result indicates that a true inertial range requires several decades of separation between the energy containing and dissipative scales. A method for examining spectral anisotropy is reported and applied to turbulence spectra in fully developed pipe flow. It is found that the spectral redistribution from the streamwise to the two lateral spectra goes primarily to the circumferential component. Experimental results are reported for an axisymmetric contraction of a fully developed pipe flow. 67 refs., 75 figs., 9 tabs.

  12. Drill pipe downhole unthreading apparatus

    International Nuclear Information System (INIS)

    This paper describes an apparatus for unthreading a threaded connection in a drill string. It comprises: an elongate shaft; fluid powered means for moving the shaft in repeated movement between first and second positions; upper and lower mandrels supporting the shaft and exposed to joints making up the drill string, the mandrels joining together to permit rotation therebetween; upper and lower pipe gripping means cooperatively engaging pipe joints in the drill string wherein the upper pipe gripping means engages a pipe joint above a threaded connection in the pipe string and the lower pipe gripping means engages a pipe joint below the threaded connection in the pipe string; and means coupling the shaft to impart repeated movement through the upper and lower mandrels and pipe gripping means to the pipe joints so that the threaded connection in the pipe string is rotated to unthread

  13. Potential high fluence response of pressure vessel internals constructed from austenitic stainless steels

    International Nuclear Information System (INIS)

    Many of the in-core components in pressurized water reactors are constructed of austenitic stainless steels. The potential behavior of these components can be predicted using data on similar steels irradiated at much higher displacement rates in liquid-metal reactors or water-cooled mixed-spectrum reactors. Consideration of the differences between the pressurized water environment and that of the other reactors leads to the conclusion that significant amounts of void swelling, irradiation creep, and embrittlement will occur in some components, and that the level of damage per atomic displacement may be larger that the level of damage per atomic displacement may be larger in the pressurized water environment

  14. Influence of localized deformation on A-286 austenitic stainless steel stress corrosion cracking in PWR primary water

    International Nuclear Information System (INIS)

    Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels is known to be a critical issue for structural components of nuclear reactor cores. The deformation of irradiated austenitic stainless steels is extremely heterogeneous and localized in deformation bands that may play a significant role in IASCC. In this study, an original approach is proposed to determine the influence of localized deformation on austenitic stainless steels SCC in simulated PWR primary water. The approach consists in (i) performing low cycle fatigue tests on austenitic stainless steel A-286 strengthened by γ' precipitates Ni3(Ti,Al) in order to shear and dissolve the precipitates in intense slip bands, leading to a localization of the deformation within and in (ii) assessing the influence of these γ'-free localized deformation bands on A-286 SCC by means of comparative CERT tests performed on specimens with similar yield strength, containing or not γ'-free localized deformation bands. Results show that strain localization significantly promotes A-286 SCC in simulated PWR primary water at 320 and 360 C. Moreover, A-286 is a precipitation-hardening austenitic stainless steel used for applications in light water reactors. The second objective of this work is to gain insights into the influence of heat treatment and metallurgical structure on A-286 SCC susceptibility in PWR primary water. The results obtained demonstrate a strong correlation between yield strength and SCC susceptibility of A-286 in PWR primary water at 320 and 360 C. (author)

  15. Piping and fitting dynamic reliability program

    International Nuclear Information System (INIS)

    The Electric Power Research Institute (EPRI) in conjunction with the U.S. Nuclear Regulatory Commission (NRC) initiated the Piping and Fitting Dynamic Reliability (PFDR) Program. The ultimate objective of this program is to introduce new, improved, realistic and defensible ASME Code design rules which take advantage of the inherent dynamic margin in piping and which result in a more balanced piping design between infrequent dynamic loads and daily operating loads. The basis for the proposed changes in design criteria will be derived from an extensive testing program together with supporting analyses. The first of three testing programs is focusing on the behaviour of typical piping components when subjected to dynamic loading introduced through hydraulically operated sleds. A second testing program is investigating the behaviour of piping systems under dynamic loading and the third program is focusing on development of a laboratory type specimen which can be used to quantitatively evaluate low cycle fatigue in the presence of ratcheting. This paper represents a status report of results of the component tests to date

  16. Postservice examination of weld discontinuities in large diameter stainless steel piping

    International Nuclear Information System (INIS)

    Metallurgical examination of pipe-to-pipe weld regions in large-diameter pipe removed from nuclear reactor service has revealed the presence of small fusion zone (FZ) discontinuities. The pipe was fabricated in the 1950's from AISI Type 304 stainless steel and welded using Type 308 filler material. The objectives of the investigation were to characterize discontinuities in weld FZ, determine the mechanisms of formation, establish the tendency for in-service propagation, and to assess the impact of FZ discontinuities on mechanical properties and service life. To accomplish this, UT and Pt were performed in weld areas to identify discontinuities and their location. Coupons were cut for metallography, optical microscopy, and, as necessary, electron microscopy (scanning electron microscopy/energy-dispersive spectroscopy [SEM/EDS]). Discontinuities which were observed in the FZ included one case of IGSCC propagation, lack-of-penetration (LOP), lack-of-fusion (LOF), underfill at the weld root, microfissures, and inclusions. Some of the discontinuities, such as microfissures, were found to have been extended by a crevice corrosion mechanism, predominantly when in fully duplex (austenite/δ ferrite) microstructures. When discontinuities such as LOF bridged austenitic and duplex structures and were open to the ID, some cracks were extended by intergranular stress corrosion, especially in the austenitic HAZ, where some grain dropping occurred. This work showed that the weld discontinuities were generally present from the time of fabrication and were most frequent in near-ID surface regions. Only those discontinuities which were open to the ID surface showed any propagation, but they did not impact the service life or structural integrity of the piping. Therefore, the in-service inspection program should continue to concentrate on IGSCC in the weld HAZ regions

  17. Status of studies on the elaboration of the KTA guide concerning non-destructive testing of austenitic materials

    International Nuclear Information System (INIS)

    There are a great deal of procedure-specific difficulties in ultrasonic testing of austenites. The study commission for non-destructive testing has taken a lot of trouble in elaborating KTA regulation 3201, which deals with the determinations for ultrasonic testing of austenites. A technological regulation usually describes the status of such technology and can not possibly reflect the respective status of research. The fact that in non-destructive testing feasibility of a test is merely a matter of expenses applies in particular for testing of austenites. Considerations on the commensurableness of expenses must include the idea that maybe sufficient proof of the security of the component in question can be obtained in a totally different way. (orig.)

  18. Model of Primary Austenite Dendrite Structure in Hypoeutectic Cast Iron

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    The solidification of primary austenite in hypoeutectic gray cast iron was studied by stepped grinding and quantitative metallography. The dendrite structure of primary austenite can be described by three models: typical dendrite crystal model, metamorphic dendrite crystal model and network dendrite crystal model. The dendrite crystals formed according to 3rd model is much more than those formed according to other models in this experiment. The primary austenites are connected each other, and the primary stems of austenite could be regarded as secondary arms and vice versa.

  19. Crystallography of lath martensite and stabilization of retained austenite

    Energy Technology Data Exchange (ETDEWEB)

    Sarikaya. M.

    1982-10-01

    TEM was used to study the morphology and crystallography of lath martensite in low and medium carbon steels in the as-quenched and 200/sup 0/C tempered conditions. The steels have microduplex structures of dislocated lath martensite and continuous thin films of retained austenite at the lath interfaces. Stacks of laths form the packets which are derived from different (111) variants of the same austenite grain. The residual parent austenite enables microdiffraction experiments with small electron beam spot sizes for the orientation relationships (OR) between austenite and martensite. All three most commonly observed ORs, namely Kurdjumov-Sachs, Nishiyama-Wassermann, and Greninger-Troiano, operate within the same sample.

  20. Crystallography of lath martensite and stabilization of retained austenite

    International Nuclear Information System (INIS)

    TEM was used to study the morphology and crystallography of lath martensite in low and medium carbon steels in the as-quenched and 2000C tempered conditions. The steels have microduplex structures of dislocated lath martensite and continuous thin films of retained austenite at the lath interfaces. Stacks of laths form the packets which are derived from different [111] variants of the same austenite grain. The residual parent austenite enables microdiffraction experiments with small electron beam spot sizes for the orientation relationships (OR) between austenite and martensite. All three most commonly observed ORs, namely Kurdjumov-Sachs, Nishiyama-Wassermann, and Greninger-Troiano, operate within the same sample

  1. Section of CMS Beam Pipe Removed

    CERN Multimedia

    2013-01-01

    Seven components of the beam pipe located at the heart of the CMS detector were removed in recent weeks. The delicate operations were performed in several stages as the detector was opened. Video of the extraction of one section: http://youtu.be/arGuFgWM7u0

  2. Resolution of thermal striping issue downstream of a horizontal pipe elbow in stratified pipe flow

    International Nuclear Information System (INIS)

    A thermally stratified pipe flow produced by a thermal transient when passing through a horizontal elbow as a result of secondary flow gives rise to large thermal fluctuations on the inner curvature wall of the downstream piping. These fluctuations were measured in a specially instrumented horizontal pipe and elbow system on a test set-up using water in the Mixing Components Technology Facility (MCTF) at Argonne National Laboratory (ANL). This study is part of a larger program which is studying the influence of thermal buoyancy on general reactor component performance. This paper discusses the influence of pipe flow generated thermal oscillations on the thermal stresses induced in the pipe walls. The instrumentation was concentrated around the exit plane of the 900 sweep elbow, since prior tests had indicated that the largest thermal fluctuations would occur within about one hydraulic diameter downstream of the elbow exit. The thermocouples were located along the inner curvature of the piping and measured the near surface fluid temperature. The test matrix involved thermal downramps under turbulent flow conditions

  3. Heat pipe technology issues

    International Nuclear Information System (INIS)

    Recent heat pipe development work at Los Alamos National Laboratory as involved performance testing of typical space reactor heat pipe designs to power levels in excess of 19 kW/cm2 axially and 300 W/cm2 radially at temperatures in the 1400 to 15000K range. Operation at conditions in the 10 kW/cm2 range has been sustained for periods of up to 1000 hours without evidence of performance degradation. The effective length for heat transport in these heat pipes was from 1.0 to 1.5 M. Materials used were molybdenum alloys with lithium employed as the heat pipe operating fluid. Shorter, somewhat lower power density, molybdenum heat pipes have been life tested at Los Alamos for periods of greater than 25,000 hours at 17000K with lithium and 20,000 hours at 15000K with sodium. These life test demonstrations and the attendant performance limit investigations provide an experimental basis for heat pipe application in space reactor design and represent the current state-of-the-art of high temperature heat pipe technology. 9 refs., 11 figs

  4. Studies on ductile unstable fracture of piping materials in light water reactors, (2)

    International Nuclear Information System (INIS)

    A pipe fracture test program has been conducted in Japan Atomic Energy Research Institute (JAERI) using austenitic stainless steel pipes and carbon steel pipes to investigate fracture behavior of the cracked pipe and to demonstrate the validity of the ''Leak Before Break'' concept for nuclear pressure boundary piping. This report describes the results of the laboratory-scale tests on center-cracked-tension (CCT) specimens machined from the 12-inch diameter Type 304 Nuclear Grade stainless steel pipe under tensile loading condition at room temperature. The program approach includes two efforts. The first phase develops J-R curve data for CCT specimens to characterize crack growth resistance. Furthermore, the flow stress obtained from these tests was applied to predict the collapse load of stainless steel pipe with a through-wall circumferential crack, based on the net-section collapse criterion. In the second phase of this program, unstable fracture experiments were conducted using compliant disc spring device. J-based tearing instability criterion was compared with the test results, and the validity of this criterion was discussed. (author)

  5. Influence of Different Parameters on Heat Pipe Performance

    OpenAIRE

    Sharmishtha Singh Hada; Prof. P. K. Jain

    2015-01-01

    In electrical and electronic industry due to miniaturization of electronic components heat density increases which, in turns increases the heat flux inside it. Scientist and many researchers are doing lot of work in this field for thermal management of devices. Heat pipe is a device that is used in electronic circuit (micro and power electronics), spacecraft & electrical components for cooling purpose. It is based on the principle of evaporation and condensation of working fluid. Heat pipe...

  6. Experimenting with a "Pipe" Whistle

    Science.gov (United States)

    Stafford, Olga

    2012-01-01

    A simple pipe whistle can be made using pieces of PVC pipe. The whistle can be used to measure the resonant frequencies of open or closed pipes. A slightly modified version of the device can be used to also investigate the interesting dependence of the sound frequencies produced on the orifice-to-edge distance. The pipe whistle described here…

  7. Reeling of tight fit pipe

    NARCIS (Netherlands)

    Focke, E.S.

    2007-01-01

    If it would be possible to install Tight Fit Pipe by means of reeling, it would be an attractive new option for the exploitation of offshore oil and gas fields containing corrosive hydrocarbons. Tight Fit Pipe is a mechanically bonded double walled pipe where a corrosion resistant alloy liner pipe i

  8. Steam generator local water chemistry and SCC of austenitic steel

    International Nuclear Information System (INIS)

    The titanium stabilized austenitic steel similar to the type of 321 is sensitive to the stress corrosion cracking under horizontal steam generator operating condition. SCC was observed under crevice corrosion parameters and has resulted in the transgranular or intergranular cracking at the both, components primary collectors and heat exchange tubes. The crevice environment is characterized by aggressive impurities and 'non aggressive' compounds. Sulfates and chlorides as aggressive species and silicates and alumino-silicates as 'non aggressive' species on the other hand are present in significant amount in the crevice environment under operating condition. Local water chemistry parameters were evaluated with MULTEQ Code. As input data the measured operational values of local and bulk environments have been used. The determined parameters were compared with the results of thread hole environment analyses and tube surface investigations respectively. Results of the hideout return profiles measurement showed an increase of sulfate concentration by one order of magnitude. Increase of the chloride content was not been observed, its value remains at operation levels. Examination of surface layers showed the preferential accumulation of sulfates, silicates and alumino-silicates in the deposit at tube support plates and in thread holes comparing relative to free span surfaces. The content of species in the water and deposits and the crystallographic structure of deposits correspond to MULTEQ results. Rising displacement tests were carried out with 0.5T CT specimens at a temperature 275 degrees C in the model water environment which simulated the crevice conditions. The experimental values are presented for crack growth rate versus stress intensity factor. Corrosion damage of the titanium stabilized austenitic steel is likely to be determined by the presence of sulfates and chlorides and other aggressive agents, as Cu. It is supposed that other decisive factor is the

  9. Pipe drafting and design

    CERN Document Server

    Parisher, Roy A

    2011-01-01

    Pipe Drafting and Design, Third Edition provides step-by-step instructions to walk pipe designers, drafters, and students through the creation of piping arrangement and isometric drawings. It includes instructions for the proper drawing of symbols for fittings, flanges, valves, and mechanical equipment. More than 350 illustrations and photographs provide examples and visual instructions. A unique feature is the systematic arrangement of drawings that begins with the layout of the structural foundations of a facility and continues through to the development of a 3-D model. Advanced chapters

  10. Optimization of Pipe Networks

    DEFF Research Database (Denmark)

    Hansen, C. T.; Madsen, Kaj; Nielsen, Hans Bruun

    1991-01-01

    The paper treats a piping system, where the layout of the network is given but the diameters of the pipes should be chosen among a small number of different values. The cost of realizing the system should be minimized while keeping the energy heads at the nodes above some lower limits. A new...... algorithm using successive linear programming is presented. The performance of the algorithm is illustrated by optimizing a network with 201 pipes and 172 nodes. It is concluded that the new algorithm seems to be very efficient and stable, and that it always finds a solution with a cost near the best...

  11. Articulated pipes conveying fluid pulsating with high frequency

    DEFF Research Database (Denmark)

    Jensen, Jakob Søndergaard

    1999-01-01

    Stability and nonlinear dynamics of two articulated pipes conveying fluid with a high-frequency pulsating component is investigated. The non-autonomous model equations are converted into autonomous equations by approximating the fast excitation terms with slowly varying terms. The downward hanging...... pipe position will lose stability if the mean flow speed exceeds a certain critical value. Adding a pulsating component to the fluid flow is shown to stabilize the hanging position for high values of the ratio between fluid and pipe-mass, and to marginally destabilize this position for low ratios. An...

  12. Metastable structure of austenite base obtained by rapid solidification in a semi-solid state

    International Nuclear Information System (INIS)

    Research highlights: → The influence of cooling rate from semi-solid state was analyzed for X210Cr12 steel. → Unconventional microstructures with over 90% of austenite were created. → Cooling rate influenced the morphology of network among globular austenite grains. → Fine troostit nuclei surrounded globular austenite grains after slower cooling. - Abstract: Material processing in a semi-solid state with rapid solidification is an innovative technology, which enables us to produce complex-shaped semi products in one operation. Unconventional properties and microstructures can be obtained in this way. Material processing in a semi-solid state has been used for materials with lower melting temperatures, particularly for Al alloys. This paper concentrates on the development of new technologies for production of miniature thin-walled steel components with complicated shapes. Ledeburitic steel with 1.8% of carbon and 11% of chromium was chosen for this experimental study. This material was used to produce very small thin-walled semi products. From the initial structure consisting of primary and secondary carbides distributed in a ferrite matrix was obtained a microstructure with over 90% of metastable austenite after cooling from the semi-solid state. The main aim of this experimental program was to describe the effect of two different methods of heating to the semi-solid state. The first method used unique heating equipment, combining high frequency and resistance heating. The second method consisted of conventional heating in a furnace. The influence of the cooling rate on the development of the microstructure was investigated. If was found that both heating and cooling rates influence grain size and the size and the morphology of carbide network placed between the globular austenite grains. Structure analysis was performed with the help of light microscopy, laser scanning confocal microscopy and scanning electron microscopy. EDX analysis was applied to

  13. Relative merits of duplex and austenitic stainless steels for applications in the oil and gas industry

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, Elisabeth; Wegrelius, Lena; Pettersson, Rachel [Outokumpu Stainless AB, Avesta (Sweden)

    2012-07-01

    The broad range of available stainless steel grades means that these materials can fulfil a wide variety of requirements within the oil and gas industry. The duplex grades have the advantage of higher strength than standard austenitic grades, while the superaustenitic grades provide a cost-effective alternative to nickel-base alloys in a number of cases. The paper presents the results of various types of laboratory testing to rank the grades in terms of resistance to pitting, crevice corrosion and stress corrosion cracking. Results from field testing in actual or simulated service conditions are discussed and a number of application examples, including process piping flexible, heat exchangers and topside equipment are presented. (author)

  14. Study of ultrasound waves attenuation: application to the nondestructive control of austenitic stainless steel welds

    International Nuclear Information System (INIS)

    Ultrasonic propagation simulation in anisotropic and heterogeneous media is essential for nondestructive testing by ultrasounds of multipass austenitic stainless steel welds that are specific of piping in nuclear power stations. Scattering at grain boundaries leads to a strong attenuation as a function of grain orientation. Attenuation measurement is complex. The implemented technique allows taking into account the physical reality of the beams and the material anisotropy. Ultrasonic propagation through the samples is modeled with transmission coefficients calculated with any incidence on a triclinic material. This method results in an increase of the attenuation versus grain orientation. For the first time, measured attenuation coefficients are integrated into a simulation code that validated them by comparison with experience. (author)

  15. Stochastic aspects of evolution of creep damage in austenitic stainless steel

    International Nuclear Information System (INIS)

    A stochastic model for the creep damage evolution and associated scatter in austenitic stainless steel has been developed in terms of a discontinuous Markov process. The magnitude of damage has been described in the form of a probability distribution function whose evolution in time characterizes the nondeterministic nature of the damage accumulation process. The long-term creep behavior on samples obtained from different locations of a thick walled SS304 LN steel pipe are studied under an identical stress and temperature condition so as to observe the scatter in creep deformation and failure data. Also the occurrences of damage and its accumulation due to creep deformation were evaluated through microstructural assessment using light optical microscope and scanning electron microscope. The validity of the model has been established by repeat data of SS304 LN steel and 316 stainless steel .

  16. The PISC parametric study on the effect of the texture of cast austenitic stainless steel

    International Nuclear Information System (INIS)

    Within the framework of Action 4 (Austenitic Steel Testing) of PISC III, a parametric study was carried out on a set of centrifugally cast stainless steel samples, representative of the main coolant piping of pressurized water nuclear reactors. The samples are obtained from different manufacturers, and feature various grain textures and dimensions. Artificial and realistic flaws were used to assess the detection and sizing capability of ultrasonic examination techniques. The paper analyzes the data as a function of the metal structure and of the main parameter of the testing techniques, which include TRL contact probes and immersion focusing transducers. Guidelines are deduced as to the selection of inspection techniques, in relation with the metallurgical texture of each specimen. In addition, the influence of the presence of a weld across the wavepath is evaluated, as well as the similarity between the responses obtained from crack-like machined reflectors and mechanical fatigue cracks

  17. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 5: probabilistic fracture mechanics analysis. Final report

    International Nuclear Information System (INIS)

    The purpose of the portion of the Load Combination Program covered in this volume was to estimate the probability of a seismic induced loss-of-coolant accident (LOCA) in the primary piping of a commercial pressurized water reactor (PWR). Such results are useful in rationally assessing the need to design reactor primary piping systems for the simultaneous occurrence of these two potentially high stress events. The primary piping system at Zion I was selected for analysis. Attention was focussed on the girth butt welds in the hot leg, cold leg and cross-over leg, which are centrifugally cast austenitic stainless steel lines with nominal outside diameters of 32 - 37 inches

  18. Ultrasonic Examination of AN Austenitic Weld: Illustration of the Disturbances of the Ultrasonic Beam

    Science.gov (United States)

    Chassignole, B.; Dupond, O.; Doudet, L.; Duwig, V.; Etchegaray, N.

    2009-03-01

    The ultrasonic examination of the primary coolant piping of Pressurized Water Reactors (PWR) is an important challenge for the nuclear industry. Numerous studies were undertaken by EDF R&D for a few years to improve the NDT process on these applications and to help to their qualification. More particularly, a great deal was made on the examination of the austenitic stainless steel welds. Indeed, the anisotropic, heterogeneous and coarse granular structures of these welds lead to important disturbances of the ultrasonic propagation. This paper presents some examples of the experimental and numerical studies which allowed to highlight the different disturbances (beam deviation, division and attenuation). We pay more attention on spurious echoes which may appear when a plane defect is located in an austenitic weld. The experimental approach is based on tests on mock-ups containing artificial defects. On the other hand, a numerical approach with the finite element code ATHENA, developed by EDF, allows to explain the origin of the disturbances. We show the interest of this tool to carry out a close analysis of the ultrasonic mode conversions in the complex structure of the weld which produce the spurious echoes. Finally, the influence of the ultrasonic disturbances on performances in term of detection of the defects is discussed.

  19. Microstructures and mechanical properties of cast austenite stainless steels after long-term thermal aging at low temperature

    International Nuclear Information System (INIS)

    Highlights: ► The primary circuit piping materials from Ling Ao Nuclear Power Plant was thermally aged for as long as 20,000 h. ► G-phase precipitation was characterized by HRTEM. ► Hardness in ferrite and austenite, tensile properties and impact behaviors of the long-term aged materials were studied. ► The mechanism of thermal aging embrittlement was proposed. - Abstract: The cast austenite stainless steels were investigate in order to understand the microstructural evolution and mechanical properties in the long-term thermal aging at 400 °C for up to 20,000 h. Spinodal decomposition and G-phase precipitation in ferrite after long-term thermal aging lead to the degradation of mechanical properties. Ferrite hardness increases with aging time, but the austenite hardness does not change. Tensile strength is not strongly affected by aging time, but the plasticity has a significant decrease after long-term aging. Under impact with high strain rate, the ferrite phases deform by the way of deformation twinning. High stress concentration on the phase boundaries cause the phase boundary separating and the austenite’s tearing off

  20. Use of an IGSCC damage index for cost benefit evaluation of BWR pipe cracking remedies

    International Nuclear Information System (INIS)

    A number of alternative approaches for eliminating or significantly reducing intergranular stress corrosion cracking (IGSCC) in BWR austenitic stainless steel piping have been developed. These include alternate materials, modified environments, and remedies which reduce steady state stresses. The cost and potential benefits of various remedies along with the implications of making no changes to the piping or plant chemistry are essential for making remedy implementation decisions and contingency planning. Unfortunately, the potential for IGSCC cracking with and without implementation of remedies depends on a large number of factors. For the current plant condition these include the original piping material and material condition, weld joint stresses, length of service and startup cycles and primary water chemistry history. In this paper, a damage index based on an engineering model for predicting the initiation of IGSCC in sensitized austenitic materials is used to quantify the likelihood of cracking with and without pipe cracking remedies as a function of time. This model is based on a fundamental understanding of IGSCC and explicitly accounts for material, material condition, environment, stress, cycles, time and cyclic wave shapes. Combination of these estimates with cost estimates for remedies and repairs provides a cost-benefit comparison of alternatives to a BWR plant owner

  1. High temperature strength of simple and solute-modified 10Cr-30Mn austenitic steels

    International Nuclear Information System (INIS)

    In order to develop potential reduced activation manganese-stabilized austenitic steels for use in the first wall component of a fusion reactor, tensile and high temperature creep properties have been investigated for simple and solute-modified 10% Cr-30% Mn austenitic steels. The yield stress increased linearly with carbon concentration over the range from room temperature to 873 K. The creep-rupture strength at 873 K increased linearly with carbon concentration at short times, below 360 ks (100 h). The contribution of carbon to the increase in creep rupture strength decreased at high carbon concentration, above 0.2%, and at long times, above 3600 ks (1000 h). A solute-modified 10Cr-30Mn-2W-0.2Ti-0.008B-0.04P-0.15C(wt%) steel exhibited very high tensile and creep rupture strength that were superior to those of type 316 steel. (orig.)

  2. Miniature pipe crawler tractor

    International Nuclear Information System (INIS)

    A pipe crawler tractor may comprise a half tractor assembly having a first base drive wheel, a second base drive wheel, and a top drive wheel. The drive wheels are mounted in spaced-apart relation so that the top drive wheel is positioned between the first and second base drive wheels. The mounting arrangement is also such that the first and second base drive wheels contact the inside surface of the pipe at respective first and second positions and so that the top drive wheel contacts the inside surface of the pipe at a third position, the third position being substantially diametrically opposed to the first and second positions. A control system connected to the half tractor assembly controls the rotation of the first base wheel, the second base wheel, and the top drive wheel to move the half tractor assembly within the pipe

  3. Miniature pipe crawler tractor

    Science.gov (United States)

    McKay, Mark D.; Anderson, Matthew O.; Ferrante, Todd A.; Willis, W. David

    2000-01-01

    A pipe crawler tractor may comprise a half tractor assembly having a first base drive wheel, a second base drive wheel, and a top drive wheel. The drive wheels are mounted in spaced-apart relation so that the top drive wheel is positioned between the first and second base drive wheels. The mounting arrangement is also such that the first and second base drive wheels contact the inside surface of the pipe at respective first and second positions and so that the top drive wheel contacts the inside surface of the pipe at a third position, the third position being substantially diametrically opposed to the first and second positions. A control system connected to the half tractor assembly controls the rotation of the first base wheel, the second base wheel, and the top drive wheel to move the half tractor assembly within the pipe.

  4. Heat Pipe Systems

    Science.gov (United States)

    1993-01-01

    The heat pipe was developed to alternately cool and heat without using energy or any moving parts. It enables non-rotating spacecraft to maintain a constant temperature when the surface exposed to the Sun is excessively hot and the non Sun-facing side is very cold. Several organizations, such as Tropic-Kool Engineering Corporation, joined NASA in a subsequent program to refine and commercialize the technology. Heat pipes have been installed in fast food restaurants in areas where humid conditions cause materials to deteriorate quickly. Moisture removal was increased by 30 percent in a Clearwater, FL Burger King after heat pipes were installed. Relative humidity and power consumption were also reduced significantly. Similar results were recorded by Taco Bell, which now specifies heat pipe systems in new restaurants in the Southeast.

  5. Machined Titanium Heat-Pipe Wick Structure

    Science.gov (United States)

    Rosenfeld, John H.; Minnerly, Kenneth G.; Gernert, Nelson J.

    2009-01-01

    Wick structures fabricated by machining of titanium porous material are essential components of lightweight titanium/ water heat pipes of a type now being developed for operation at temperatures up to 530 K in high-radiation environments. In the fabrication of some prior heat pipes, wicks have been made by extruding axial grooves into aluminum unfortunately, titanium cannot be extruded. In the fabrication of some other prior heat pipes, wicks have been made by in-situ sintering of metal powders shaped by the use of forming mandrels that are subsequently removed, but in the specific application that gave rise to the present fabrication method, the required dimensions and shapes of the heat-pipe structures would make it very difficult if not impossible to remove the mandrels due to the length and the small diameter. In the present method, a wick is made from one or more sections that are fabricated separately and assembled outside the tube that constitutes the outer heat pipe wall. The starting wick material is a slab of porous titanium material. This material is machined in its original flat configuration to form axial grooves. In addition, interlocking features are machined at the mating ends of short wick sections that are to be assembled to make a full-length continuous wick structure. Once the sections have been thus assembled, the resulting full-length flat wick structure is rolled into a cylindrical shape and inserted in the heatpipe tube (see figure). This wick-structure fabrication method is not limited to titanium/water heat pipes: It could be extended to other heat pipe materials and working fluids in which the wicks could be made from materials that could be pre-formed into porous slabs.

  6. Silicon Heat Pipe Array

    Science.gov (United States)

    Yee, Karl Y.; Ganapathi, Gani B.; Sunada, Eric T.; Bae, Youngsam; Miller, Jennifer R.; Beinsford, Daniel F.

    2013-01-01

    Improved methods of heat dissipation are required for modern, high-power density electronic systems. As increased functionality is progressively compacted into decreasing volumes, this need will be exacerbated. High-performance chip power is predicted to increase monotonically and rapidly with time. Systems utilizing these chips are currently reliant upon decades of old cooling technology. Heat pipes offer a solution to this problem. Heat pipes are passive, self-contained, two-phase heat dissipation devices. Heat conducted into the device through a wick structure converts the working fluid into a vapor, which then releases the heat via condensation after being transported away from the heat source. Heat pipes have high thermal conductivities, are inexpensive, and have been utilized in previous space missions. However, the cylindrical geometry of commercial heat pipes is a poor fit to the planar geometries of microelectronic assemblies, the copper that commercial heat pipes are typically constructed of is a poor CTE (coefficient of thermal expansion) match to the semiconductor die utilized in these assemblies, and the functionality and reliability of heat pipes in general is strongly dependent on the orientation of the assembly with respect to the gravity vector. What is needed is a planar, semiconductor-based heat pipe array that can be used for cooling of generic MCM (multichip module) assemblies that can also function in all orientations. Such a structure would not only have applications in the cooling of space electronics, but would have commercial applications as well (e.g. cooling of microprocessors and high-power laser diodes). This technology is an improvement over existing heat pipe designs due to the finer porosity of the wick, which enhances capillary pumping pressure, resulting in greater effective thermal conductivity and performance in any orientation with respect to the gravity vector. In addition, it is constructed of silicon, and thus is better

  7. Heat pipe reactors for space power applications

    Science.gov (United States)

    Koenig, D. R.; Ranken, W. A.; Salmi, E. W.

    1977-01-01

    A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

  8. Lattice expansion of carbon-stabilized expanded austenite

    DEFF Research Database (Denmark)

    Hummelshøj, Thomas Strabo; Christiansen, Thomas; Somers, Marcel A. J.

    2010-01-01

    The lattice parameter of expanded austenite was determined as a function of the content of interstitially dissolved carbon in homogeneous, carburized thin stainless steel foils. For the first time this expansion of the face-centered cubic lattice is determined on unstrained austenite. It is found...

  9. Flow lines and microscopic elemental inhomogeneities in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Mosley, Jr, W C

    1982-01-01

    Flow lines in mechanically formed austenitic stainless steels are known to influence fracture behavior. Enhancement of flow lines by chemical etching is evidence of elemental inhomogeneity. This paper presents the results of electron microprobe analyses to determine the nature of flow lines in three austenitic stainless steels: 21Cr-6Ni-9Mn, 304L, and 19Ni-18Cr.

  10. X-ray fractography studies on austenitic stainless steels

    NARCIS (Netherlands)

    Rajanna, K.; Pathiraj, B.; Kolster, B.H.

    1996-01-01

    In this investigation, the fracture surfaces of SS 304 and SS 316 austenitic steels were analysed using the X-ray fractography technique. In both cases, a decrease in the austenite content was observed at the fracture surface as a result of deformation induced martensite, indicating a linear relatio

  11. Investigation of joining techniques for advanced austenitic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Lundin, C.D.; Qiao, C.Y.P.; Kikuchi, Y.; Shi, C.; Gill, T.P.S.

    1991-05-01

    Modified Alloys 316 and 800H, designed for high temperature service, have been developed at Oak Ridge National Laboratory. Assessment of the weldability of the advanced austenitic alloys has been conducted at the University of Tennessee. Four aspects of weldability of the advanced austenitic alloys were included in the investigation.

  12. Electropolishing and chemical passivation of austenitic steel

    Directory of Open Access Journals (Sweden)

    A. Baron

    2008-12-01

    Full Text Available Purpose: The aim of the paper is investigations a dependence between the parameters of the electrochemical treatment of austenitic steel and their electrochemical behavior in Tyrod solution.Design/methodology/approach: Specimens (rode 30 mm × ø1 mm were to give in to the surface treatment – mechanically polishing, electrolytic polishing and passivation with various parameter. Electrochemical investigations concerning the corrosion resistance of austenitic steel samples were carried out by means of the potentiodynamic and electrochemical impedance spectroscopy method.Findings: The analysis of the obtained results leads to the conclusion that chemical passivation affects also the chemical composition of the passive layer of steel and changes its resistance to corrosion. Electrolytic polishing improves corrosion resistance, as can be proved by the shift of the value of the corrosion potential and break-down potential of the passive layer and the initiation of pittings.Research limitations/implications: The obtained results are the basis for the optimization of anodic passivation parameters of the austenitic steel as a metallic biomaterial. The future research should be focused on selected more suitable parameters of the electrochemical impedance spectroscopy test to better describe process on the solid/ liquid interface.Practical implications: In result of the presented investigations it has been found that the best corrosion resistance can be achieved thanks to the application of electrolytic polishing of the steel in a special bath and chemical passivation in nitric (V acid with an addition of chromic (VI acid temperature t = 60°C for one hour.Originality/value: The enormous demand for metal implants has given rise to a search for cheap materials with a good biotolerance and resistance to corrosion. Most commonly used are steel implants assigned to remain in the organism for some limited time only. It was compare two electrochemical methods

  13. Grain growth in heat resisting austenitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Denisova, I.K.; Zakharov, V.N.; Karpova, N.M.; Farber, V.M.

    1985-01-01

    A study was made on kinetics of grain growth in steels of 37Kh12N8G8 type alloyed by V, Nb, Ti, Mo, W. It was concluded that the nature of carbide phase and kinetics of its dissolution in heat resisting austenitic steels dictate steel tendency to grain growth. At the same time decrease of diffusion mobility of atoms in steel matrix during its alloying by titanium aid tungsten results in sufficient decrease of the tendency to grain growth and variation in grain size.

  14. Embrittlement of austenitic stainless steel welds

    International Nuclear Information System (INIS)

    To prevent hot-cracking, austenitic stainless steel welds generally contain a small percent of delta ferrite. Although ferrite has been found to effectively prevent hot-cracking, it can lead to embrittlement of welds when exposed to elevated temperatures. The aging behavior of type-308 stainless steel weld has been examined over a range of temperatures 475--850 C for times up to 10,000 hrs. Upon aging, and depending on the temperature range, the unstable ferrite may undergo a variety of solid state transformations. These phase changes creep-rupture and Charpy impact properties

  15. Tribocorrosion wear of austenitic and martensitic steels

    Directory of Open Access Journals (Sweden)

    G. Rozing

    2016-07-01

    Full Text Available This paper explores the impact of tribocorrosion wear caused by an aggressive acidic media. Tests were conducted on samples made of stainless steel AISI 316L, 304L and 440C. Austenitic steels were tested in their nitrided state and martensitic in quenched and tempered and then induction hardened state. Electrochemical corrosion resistance testing and analysis of the microstructure and hardness in the cross section was carried out on samples of selected steels. To test the possibility of applying surface modification of selected materials in conditions of use, tests were conducted on samples/parts in a worm press for final pressing.

  16. Wear behavior of austenite containing plate steels

    Science.gov (United States)

    Hensley, Christina E.

    As a follow up to Wolfram's Master of Science thesis, samples from the prior work were further investigated. Samples from four steel alloys were selected for investigation, namely AR400F, 9260, Hadfield, and 301 Stainless steels. AR400F is martensitic while the Hadfield and 301 stainless steels are austenitic. The 9260 exhibited a variety of hardness levels and retained austenite contents, achieved by heat treatments, including quench and tempering (Q&T) and quench and partitioning (Q&P). Samples worn by three wear tests, namely Dry Sand/Rubber Wheel (DSRW), impeller tumbler impact abrasion, and Bond abrasion, were examined by optical profilometry. The wear behaviors observed in topography maps were compared to the same in scanning electron microscopy micrographs and both were used to characterize the wear surfaces. Optical profilometry showed that the scratching abrasion present on the wear surface transitioned to gouging abrasion as impact conditions increased (i.e. from DSRW to impeller to Bond abrasion). Optical profilometry roughness measurements were also compared to sample hardness as well as normalized volume loss (NVL) results for each of the three wear tests. The steels displayed a relationship between roughness measurements and observed wear rates for all three categories of wear testing. Nanoindentation was used to investigate local hardness changes adjacent to the wear surface. DSRW samples generally did not exhibit significant work hardening. The austenitic materials exhibited significant hardening under the high impact conditions of the Bond abrasion wear test. Hardening in the Q&P materials was less pronounced. The Q&T microstructures also demonstrated some hardening. Scratch testing was performed on samples at three different loads, as a more systematic approach to determining the scratching abrasion behavior. Wear rates and scratch hardness were calculated from scratch testing results. Certain similarities between wear behavior in scratch testing

  17. Verification of ultrasonic indications in austenitic overlay

    International Nuclear Information System (INIS)

    An austenitic overlay on steel 15Kh2MFA was tested by an ultrasonic probe. Overlays of this kind are used in nuclear reactor pressure vessels. The results of the ultrasonic method were compared with those of the metallographic method. Metallographic analysis showed that the majority of defects found by ultrasonic tests included clusters of cavities, welded-in slag and cracks, which mostly occurred between the weld beads. The experiments gave evidence that defects not smaller than 0.5 mm in size can be well detected at depths not exceeding 7 mm. (M.D.). 3 figs

  18. Correlation between magnetic field quality and mechanical components of the Large Hadron Collider main dipoles

    Energy Technology Data Exchange (ETDEWEB)

    Bellesia, B

    2006-12-15

    The 1234 superconducting dipoles of the Large Hadron Collider, working at a cryogenic temperature of 1.9 K, must guarantee a high quality magnetic field to steer the particles inside the beam pipe. Magnetic field measurements are a powerful way to detect assembly faults that could limit magnet performances. The aim of the thesis is the analysis of these measurements performed at room temperature during the production of the dipoles. In a large scale production the ideal situation is that all the magnets produced were identical. However all the components constituting a magnet are produced with certain tolerance and the assembly procedures are optimized during the production; due to these the reality drifts away from the ideal situation. We recollected geometrical data of the main components (superconducting cables, coil copper wedges and austenitic steel coil collars) and coupling them with adequate electro-magnetic models we reconstructed a multipolar field representation of the LHC dipoles defining their critical components and assembling procedures. This thesis is composed of 3 main parts: 1) influence of the geometry and of the assembling procedures of the dipoles on the quality of the magnetic field, 2) the use of measurement performed on the dipoles in the assembling step in order to solve production issues and to understand the behaviour of coils during the assembling step, and 3) a theoretical study of the uncertain harmonic components of the magnetic field in order to assess the dipole production.

  19. Correlation between magnetic field quality and mechanical components of the Large Hadron Collider main dipoles

    International Nuclear Information System (INIS)

    The 1234 superconducting dipoles of the Large Hadron Collider, working at a cryogenic temperature of 1.9 K, must guarantee a high quality magnetic field to steer the particles inside the beam pipe. Magnetic field measurements are a powerful way to detect assembly faults that could limit magnet performances. The aim of the thesis is the analysis of these measurements performed at room temperature during the production of the dipoles. In a large scale production the ideal situation is that all the magnets produced were identical. However all the components constituting a magnet are produced with certain tolerance and the assembly procedures are optimized during the production; due to these the reality drifts away from the ideal situation. We recollected geometrical data of the main components (superconducting cables, coil copper wedges and austenitic steel coil collars) and coupling them with adequate electro-magnetic models we reconstructed a multipolar field representation of the LHC dipoles defining their critical components and assembling procedures. This thesis is composed of 3 main parts: 1) influence of the geometry and of the assembling procedures of the dipoles on the quality of the magnetic field, 2) the use of measurement performed on the dipoles in the assembling step in order to solve production issues and to understand the behaviour of coils during the assembling step, and 3) a theoretical study of the uncertain harmonic components of the magnetic field in order to assess the dipole production

  20. Spatial image compounding applied to a phase coherence corrected UT-PA technique for inspecting nuclear components of coarse-grained structure

    Science.gov (United States)

    Brizuela, Jose; Katchadjian, Pablo; Garcia, Alejandro; Desimone, Carlos

    2016-02-01

    The aim of this work is to obtain a C-Scan view of an austenitic stainless steel weld from a nuclear use pipe. In order to obtain this result Sectorial Scans (S-Scan) from both sides of the weld are obtained by Ultrasonic Phase Array (UT-PA). Then, spatial image compounding is performed to generate a single image from the S-Scans acquired at the same circumferential position of the transducer. These joints have a coarse grain structure which significantly reduce the transmission of the ultrasonic wave due to attenuation characteristics and backscattered noise from microstructures inside the material. For this reason, phase coherence imaging technique has been also applied to reduce the structural noise and improve the image quality. To verify detected defects, and given the impossibility of cutting the component, gammagraphy were performed with Co60.

  1. Austenite Recrystallization and Controlled Rolling of Low Carbon Steels

    Institute of Scientific and Technical Information of China (English)

    DU Lin-xiu; ZHANG Zhong-ping; SHE Guang-fu; LIU Xiang-hua; WANG Guo-dong

    2006-01-01

    The dynamic recrystallization and static recrystallization in a low carbon steel were investigated through single-pass and double-pass experiments. The results indicate that as the deformation temperature increases and the strain rate decreases, the shape of the stress-strain curve is changed from dynamic recovery shape to dynamic recrystallization shape. The austenite could not recrystallize within a few seconds after deformation at temperature below 900 ℃. According to the change in microstructure during deformation, the controlled rolling of low carbon steel can be divided into four stages: dynamic recrystallization, dynamic recovery, strain-induced ferrite transformation, and rolling in two-phase region. According to the microstructure after deformation, the controlled rolling of low carbon steel can be divided into five regions: non-recrystallized austenite, partly-recrystallized austenite, fully-recrystallized austenite, austenite to ferrite transformation, and dual phase.

  2. Thermal control of electronic equipment by heat pipes; Controle thermique de composants electroniques par caloducs

    Energy Technology Data Exchange (ETDEWEB)

    Groll, M.; Schneider, M. [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme; Sartre, V.; Chaker Zaghdoudi, M.; Lallemand, M. [Institut National des Sciences Appliquees (INSA), 69 - Villeurbanne (France). Centre de Thermique de Lyon, Upresa CNRS

    1998-05-01

    In the frame of the BRITE-EURAM european programme (KHIEPCOOL project), a literature survey on the main beat pipe and micro heat pipe technologies developed for thermal control of electronic equipment has been carried out. The conventional heat pipes are cylindrical, flat or bellow tubes, using wicks or axial grooves as capillary structures. In the field of micro heat pipes, the component interconnection substrate. The best performances were achieved with Plesch`s axially grooved flat miniature heat pipe, which is able to transfer a heat flux of about 60 W.cm{sup -2}. Theoretical models have shown that the performance of micro heat pipe arrays increase with increasing tube diameter, decreasing tube length and increasing heat pipe density. The heat pipe technologies are classified and compared according to their geometry and location in the system. A list of about 150 references, classified according to their subjects, is presented. (authors) 160 refs.

  3. Effect of aging on mechanical properties of austenitic stainless steel castings and welds

    International Nuclear Information System (INIS)

    Study of the influence of long time aging on the properties of the cast austenitic steel and associated welds or cladding in the components of the primary loop of nuclear plants: embrittlement by precipitation of α'(chromium rich) in ferrite islands (mostly for castings); precipitation hardens the ferrite wich breaks by cleavage. The impact energy and Isub(IC) value are lowered by this phenomenon. Low cycle fatigue properties and fatigue crack growth rates are not modified by aging. Study of correlation between KCU impact toughness at the end of the life of a component, chemical composition and ferrite content

  4. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 2800C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 2800C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  5. Piping stress analysis for AP600 secondary system

    International Nuclear Information System (INIS)

    Piping stress analysis for AP600 secondary system has been done using software PS-CAEPIPE version mainframe and CAEPIPE version PC. The loading applied to the system are statical load consist of deadweight, pressure load and thermal expansion load. Standard used in this calculation is ASME/ANSI B31.1. A piping system consists of pipes and appropriate components, such as achors, valves, pumps, flanges, etc. The parameters to be evaluated are pipe stress (psi), pipe displacements (in) and component loading (lbs). The use of support in the optimal manner is to be considered to reach a favorable condition. The allowable stress for sustained loads (death-weight and pressure) is SH (15000 psi in these cases) and for thermal load is SA (22500 psi in these cases). The allowable pipe displacement within 0.125 inches for total load. Therefore, the allowable load of components depends on the component itself. Three piping analysis packages for secondary system of AP600 have been done, those are HDS-310 (turbine building) and VYS-210 (auxiliary building). These system contain pipes with the diameter of 1 inch, 8 inches, 10 inches and 16 inches. The design pressures are in the range of 50 to 550 psi and the design temperatures are in the range of 185 deg F. The result shows that for analysis without supports, only CDS-080 is acceptable. After locating a variable support in HDS-310 and 3 rigid supports in VYS-210, all system are acceptable with the maximum pipe stress of 6533 psi, maximum displacement for sustained load of 0.069 inches and for total load of 0.635 inches

  6. An assessment of seismic margins in nuclear plant piping

    Energy Technology Data Exchange (ETDEWEB)

    Chen, W.P.; Jaquay, K.R. [Energy Technology Engineering Center, Canoga Park, CA (United States); Chokshi, N.C.; Terao, D. [Nuclear Regulatory Commission, Washington DC (United States)

    1996-03-01

    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: (1) reviews of seismic testing of piping components performed as part of the Electric Power Research Institute (EPRI)/NRC Piping and Fitting Dynamic Reliability (PFDR) Program, and (2) assessments of safety margins inherent in the ASME Code, Section III, piping seismic design criteria as revised by the 1994 Addenda are reported. The reviews indicate that the margins inherent in the revised criteria may be less than acceptable and that modifications to these criteria may be required.

  7. An assessment of seismic margins in nuclear plant piping

    International Nuclear Information System (INIS)

    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: (1) reviews of seismic testing of piping components performed as part of the Electric Power Research Institute (EPRI)/NRC Piping and Fitting Dynamic Reliability (PFDR) Program, and (2) assessments of safety margins inherent in the ASME Code, Section III, piping seismic design criteria as revised by the 1994 Addenda are reported. The reviews indicate that the margins inherent in the revised criteria may be less than acceptable and that modifications to these criteria may be required

  8. Evaluation of thermal aging effect on primary pipe material in nuclear power plant by micro hardness test method

    International Nuclear Information System (INIS)

    The investigation was carried out on the changes in mechanical properties of the primary pipe material Z3CN20.09M after 10000 h aging at 400℃ by using micro- Vickers and impact testing machine. The results show that the impact energy of testing material decreases. However, the micro-Vickers hardness of ferrite phase and austenite phase which constitute the testing material increase and keep constant, respectively. The intrinsic relations were analyzed between the micro-Vickers hardness and the impact energy to make an attempt to present the micro-Vickers hardness measurement as a method applicable to evaluating the thermal aging of the primary pipe material. (authors)

  9. Fabrication and ageing of cast austenitic steels

    International Nuclear Information System (INIS)

    An investigation has been undertaken to determine the magnitude of any reduction in properties which may occur in cast duplex stainless steels and weldments during long term exposure to reactor operating conditions. Test panels were fabricated in CF3 stainless steel by a manual metal arc (MMA) process using 19.9.L (Type 308L) consumables. The mechanical properties and intergranular corrosion resistance of parent material and weldments were measured following accelerated ageing at 3750 and 4000C for up to 10,000 hours. Both the impact energy and J/sub R/ fracture toughness properties of the cast austenitic/ferritic stainless steel were reduced following aging at 4000C for 10,000 hours, whereas austenitic stainless steel MMA weld metals exhibited a reduction in J/sub R/ fracture toughness but no change in impact energy. Even in the unaged state, MMA weld metals were shown to have a much lower resistance to stable crack growth than the parent cast steel, and, following aging, there is a further reduction in the ductile tearing resistance of such weld metals. Therefore, in any assessment of the structural integrity of the reactor coolant pump bowl for a pressurized water reactor (PWR), the weld metal fracture properties during service are likely to be of considerable importance

  10. Austenite Formation from Martensite in a 13Cr6Ni2Mo Supermartensitic Stainless Steel

    NARCIS (Netherlands)

    Bojack, A.; Zhao, L.; Morris, P.F.; Sietsma, J.

    2016-01-01

    The influence of austenitization treatment of a 13Cr6Ni2Mo supermartensitic stainless steel (X2CrNiMoV13-5-2) on austenite formation during reheating and on the fraction of austenite retained after tempering treatment is measured and analyzed. The results show the formation of austenite in two stage

  11. Qualification of PHT piping of Indian 500 MW PHWR for LBB, using R-6 method

    International Nuclear Information System (INIS)

    This document discusses the qualification of straight pipe portion of the primary heat transport (PHT) piping of Indian 500 MWe pressurised heavy water reactor (PHWR) for leak before break (LBB). The evaluation is done using R-6 [1] method. The results presented here are: the safety margins which exist on straight pipe components of main PHT piping of 500 MWe, under leakage size crack (LSC) and design basis accident loads; the sensitivity of safety margins with respect to different analysis parameters and the qualification of PHT piping for LBB based on criterion given by NUREG-1061 [2] and TECDOC-774 [3]. (author)

  12. Development of high pressure two-phase choked flow analysis methodology in complex piping system

    International Nuclear Information System (INIS)

    Choked flow mechanism, characteristics of two-phase flow sound velocity and compressibility effects on flow through various piping system components are studied to develop analysis methodology for high pressure two-phase choked flow in complex piping system which allows choking flow rate evaluation and piping system design related analysis. Piping flow can be said choked if Mach number is equal to 1 and compressibility effects can be accounted through modified incompressible formula in momentum equation. Based on these findings, overall analysis system is developed to study thermal-hydraulic effects on steady-state piping system flow and future research items are presented. (Author)

  13. Determining Experimental Parameters for Thermal-Mechanical Forming Simulation considering Martensite Formation in Austenitic Stainless Steel

    Science.gov (United States)

    Schmid, Philipp; Liewald, Mathias

    2011-08-01

    The forming behavior of metastable austenitic stainless steel is mainly dominated by the temperature-dependent TRIP effect (transformation induced plasticity). Of course, the high dependency of material properties on the temperature level during forming means the temperature must be considered during the FE analysis. The strain-induced formation of α'-martensite from austenite can be represented by using finite element programs utilizing suitable models such as the Haensel-model. This paper discusses the determination of parameters for a completely thermal-mechanical forming simulation in LS-DYNA based on the material model of Haensel. The measurement of the martensite evolution in non-isothermal tensile tests was performed with metastable austenitic stainless steel EN 1.4301 at different rolling directions between 0° and 90 °. This allows an estimation of the influence of the rolling direction to the martensite formation. Of specific importance is the accuracy of the martensite content measured by magnetic induction methods (Feritscope). The observation of different factors, such as stress dependence of the magnetisation, blank thickness and numerous calibration curves discloses a substantial important influence on the parameter determination for the material models. The parameters obtained for use of Haensel model and temperature-dependent friction coefficients are used to simulate forming process of a real component and to validate its implementation in the commercial code LS-DYNA.

  14. Mechanical properties and damage behavior of non-magnetic high manganese austenitic steels

    International Nuclear Information System (INIS)

    Fe-Cr-Mn steels have been considered as materials of structural components for fusion reactor because of their low induced-radio-activity compared with SUS316 stainless steels. It has been expected to develop a non-magnetic steel with a high stability of the austenitic phase and a strong resistance to irradiation environments. For these objectives, a series of the Fe-Cr-Mn steels have been examined by tensile tests and simulation irradiation by electrons. The main alloying compositions of the steels developed are: C:0.02-0.2 wt%, Mn: 15 wt%, Cr: 15-16 wt%, N: 0.2 wt%. These steels were heat-treated at 1323 K for 1 h. The structure of the steels after the heat-treatment was austenite single phase. The yield stress of the steels was 350-450 MPa and the elongation were 55-60%. When the steels of high C and N was electron-irradiated at below 673 K, no voids were nucleated and only small dislocation loops were formed with high density. The austenite phase was also stable during irradiation below 673 K. Thus, newly developed high manganese steels have excellent mechanical proprieties and high irradiation resistance at relatively low temperature. (orig.)

  15. On-line model for control of hot rolling of austenitic steel strips

    International Nuclear Information System (INIS)

    The on-line model of strip rolling for austenitic steels is described in the paper. Three components are included in the model. The first is a new thermal model, which is based on an analytical solution of the Fourier equation. Thermophysical properties of the austenitic steels are introduced in the model. The numerical procedure, which designs a rolling schedule, is the second part of the model. The rolling velocities and the reductions in subsequent passes are determined by a solution of a set of non-linear equations, which compose continuity condition and energy balance for all stands. The problem is solved using optimisation techniques with constraints, that allows imposing of the technological limitations on the solution. The adaptive procedure is applied to adjust coefficient in the flow stress equation during the on-line work of the model. Microstructure evolution model is the third part of the system. This model is based on semi-empirical equations describing microstructural phenomena for austenitic steels. The microstructural model is included in the temperature calculations, but it does not take part in the control of the rolling process directly. Its task is to supply information regarding the microstructure and mechanical properties predicted for the current rolling technology, designed by the system. Description of all the developed models is given in the paper. Results of numerical experiments including calculations of rolling schedules are presented. (author)

  16. Development of a robust modeling tool for radiation-induced segregation in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Allen, Todd R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Busby, Jeremy T [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels in Light Water Reactor (LWR) components has been linked to changes in grain boundary composition due to irradiation induced segregation (RIS). This work developed a robust RIS modeling tool to account for thermodynamics and kinetics of the atom and defect transportation under combined thermal and radiation conditions. The diffusion flux equations were based on the Perks model formulated through the linear theory of the thermodynamics of irreversible processes. Both cross and non-cross phenomenological diffusion coefficients in the flux equations were considered and correlated to tracer diffusion coefficients through Manning’s relation. The preferential atomvacancy coupling was described by the mobility model, whereas the preferential atom-interstitial coupling was described by the interstitial binding model. The composition dependence of the thermodynamic factor was modeled using the CALPHAD approach. Detailed analysis on the diffusion fluxes near and at grain boundaries of irradiated austenitic stainless steels suggested the dominant diffusion mechanism for chromium and iron is via vacancy, while that for nickel can swing from the vacancy to the interstitial dominant mechanism. The diffusion flux in the vicinity of a grain boundary was found to be greatly influenced by the composition gradient formed from the transient state, leading to the oscillatory behavior of alloy compositions in this region. This work confirms that both vacancy and interstitial diffusion, and segregation itself, have important roles in determining the microchemistry of Fe, Cr, and Ni at irradiated grain boundaries in austenitic stainless steels.

  17. Self-consistent modeling of rolling textures in an austenitic-ferritic duplex steel

    International Nuclear Information System (INIS)

    Research highlights: → The selection of slip systems is linked to the grain-boundary-mediated activities. → In the duplex steel interactions between phases play a big role on the texture. → For austenite, a reliable prediction of texture is achieved at small deformations. → A model incorporating micro-scale shear banding in f.c.c. phases was developed. - Abstract: Rolling textures of the constituent phases in an austenitic-ferritic duplex stainless steel are measured by X-ray diffraction experiments, showing that the brass-type texture, typical of f.c.c. materials with low SFE, is developed in the austenitic phase, and the rotated-cube and brass-R textures are developed in the ferritic phase. On the basis of the experimental texture components and fibers at different reductions, rolling textures of the respective phases in the duplex steel are simulated using a self-consistent model. After considering various micromechanical interactions within the steel, a reliable prediction of the evolution of grain orientation distributions for the phases at small reductions is achieved. An attempt in modeling the brass-type texture for the f.c.c. metallic phase is also performed by incorporating the shear banding mechanism into the presented model.

  18. Leak-before-break behaviour of nuclear piping systems

    International Nuclear Information System (INIS)

    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

  19. Microstructure influence on fatigue behaviour of austenitic stainless steels with high molybdenum content; Influencia de la microestructura en el comportamiento a fatiga de aceros inoxidables austeniticos con alto contenido en molibdeno

    Energy Technology Data Exchange (ETDEWEB)

    Onoro, J.; Gamboa, R.; Ranninger, C.

    2006-07-01

    Austenitic stainless steels with molybdenum present high mechanical properties and corrosion resistance to aggressive environments. These steels have been used to tank and vessel components for high liquids as phosphoric, nitric and sulphuric acids. These materials with low carbon and nitrogen addition have been proposed candidates as structural materials for the international thermonuclear experimental reactor (ITER) in-vessel components. Molybdenum addition in austenitic stainless steel improves mechanical and corrosion properties, but with it can produce the presence of nitrogen microstructure modifications by presence or precipitation of second phases. This paper summarises the fatigue and corrosion fatigue behaviour of two 317LN stainless steels with different microstructure. Fully austenitic steel microstructure show better fatigue, corrosion fatigue resistance and better ductility than austenitic steel with delta ferrite microstructure, mainly at low stresses. (Author)

  20. Composite drill pipe

    Science.gov (United States)

    Leslie, James C.; Leslie, II, James C.; Heard, James; Truong, Liem , Josephson; Marvin , Neubert; Hans

    2008-12-02

    A composite pipe segment is formed to include tapered in wall thickness ends that are each defined by opposed frustoconical surfaces conformed for self centering receipt and intimate bonding contact within an annular space between corresponding surfaces of a coaxially nested set of metal end pieces. The distal peripheries of the nested end pieces are then welded to each other and the sandwiched and bonded portions are radially pinned. The composite segment may include imbedded conductive leads and the axial end portions of the end pieces are shaped to form a threaded joint with the next pipe assembly that includes a contact ring in one pipe assembly pierced by a pointed contact in the other to connect the corresponding leads across the joint.

  1. Pipe whip and impact

    International Nuclear Information System (INIS)

    Over the past few years changes in economic and safety considerations in nuclear power plants have resulted in a need to examine the problem of pipe whip in greater detail. Consequently, experimental programmes were set up in France, North America and Britain. Results from these tests combined with analytical work indicate that pipe whip followed by impact with surrounding pipework and structures may not be as serious as had been believed. Impact loads have been found to be much less (at least five times) than those predicted to the appropriate design regulations. Hence, the use of pipe whip restraints may have been overconservative. The use of fewer, better designed restraints, would result in greater accessibility of pipework, a reduced need for inspection of restraints, and a considerable financial saving. (author)

  2. Heat pipes and use of heat pipes in furnace exhaust

    Science.gov (United States)

    Polcyn, Adam D.

    2010-12-28

    An array of a plurality of heat pipe are mounted in spaced relationship to one another with the hot end of the heat pipes in a heated environment, e.g. the exhaust flue of a furnace, and the cold end outside the furnace. Heat conversion equipment is connected to the cold end of the heat pipes.

  3. Structural integrity of welded bi-metallic components (BIMET) - Task Group 5 'Analysis'. Prediction by EAM and FEA

    International Nuclear Information System (INIS)

    Investigations in the EU BIMET project focused on two pipe segments of ferritic and austenitic steel with a special weld in between which is characterised by a multiphase transition with strongly diverging characteristics (strength mis-matching). The BIMET project is described, and some of the findings of Task Group 5 'Analysis' are presented which are based on EAM and FEA

  4. Effect of the manufacturing process on the thermal aging of PWR duplex stainless steel components

    International Nuclear Information System (INIS)

    Some components of the primary loop of Pressurized Water Reactors (pump casings, some elbows, pipes, fittings and valves) are made of cast duplex stainless steels. The manufacturing process of these components has been carefully studied. The manufacturing process consists of a solidification stage followed by heat treatments to homogenize the material (by dissolving the embrittling phases precipitated at the end of the solidification) and set the ferrite content. Apart from the chemical composition, the main manufacturing parameters identified were: the solidification speed depending on foundry practice; the homogenizing heat treatment temperature; the homogenizing heat treatment holding time, the quenching rate after the homogenizing heat treatment. A program of simulation in laboratory of the effect of each of these parameters on the thermal aging has been initiated. This program, run on industrial products cast for studies, completes the aging surveillance programme on test ingots (cast at the same time as components). The metallurgical and mechanical characteristics of the materials (as-quenched and after aging up to 10 000 h at 350 deg C) have been studied. The main results of this parametric study are as follows: the solidification speed affects the morphology of the ferrite-austenite microstructure and the characteristics of the toughness transition curve; the homogenizing heat treatment temperature especially affect the ferrite content of the material and the chemical composition of each phase; the homogenizing heat treatment holding time and the quenching rate affects the beginning of the decomposition of the ferrite and consequently the whole kinetics of aging and embrittlement. (author)

  5. Modeling of Incubation Time for Austenite to Ferrite Phase Transformation

    Institute of Scientific and Technical Information of China (English)

    ZHOU Xiao-guang; LIU Zhen-yu; WU Di; WANG Wei; JIAO Si-hai

    2006-01-01

    On the basis of the classical nucleation theory, a new model of incubation time for austenite to ferrite transformation has been developed, in which the effect of deformation on austenite has been taken into consideration. To prove the precision of modeling, ferrite transformation starting temperature (Ar3) has been calculated using the Scheil′s additivity rule, and the Ar3 values were measured using a Gleeble 1500 thermomechanical simulator. The Ar3 values provided by the modeling method coincide with the measured ones, indicating that the model is precise in predicting the incubation time for austenite to ferrite transformation in hot deformed steels.

  6. PE 100 pipe systems

    CERN Document Server

    Brömstrup, Heiner

    2012-01-01

    English translation of the 3rd edition ""Rohrsysteme aus PE 100"". Because of the considerably increased performance, pipe and pipe systems made from 100 enlarge the range of applications in the sectors of gas and water supply, sewage disposal, industrial pipeline construction and in the reconstruction and redevelopment of defective pipelines (relining). This book applies in particular to engineers, technicians and foremen working in the fields of supply, disposal and industry. Subject matters of the book are all practice-relevant questions regarding the construction, operation and maintenance

  7. THERMION: Verification of a Thermionic Heat Pipe in Microgravity

    OpenAIRE

    Powell, George; Redd, Frank

    1991-01-01

    The Idaho National Engineering Laboratory (lNEL) is conducting intensive research in the design and development of a small excore heat-pipe-thermionic space nuclear reactor power system (SEHPTR). The SEHPTR spacecraft will be able to supply 40 kW of power in any given orbit. Figure 1-1 shows a conceptual diagram of the SEHPTR spacecraft. The key components in this reactor are the thermionic heat pipes. The heat pipes have two major functions: first, to convert heat energy into electrical ener...

  8. Heat pipe gas combustion system endurance test for Stirling engine

    Science.gov (United States)

    Mahrle, P.

    1990-12-01

    Stirling Thermal Motors, Inc. has been developing a general purpose Heat Pipe Gas Combustion (HPGC) system suitable for use with the STM4-120 Stirling engine. The HPGC consists of a parallel plate recuperative preheater, a finned heat pipe evaporator, and a film-cooled gas combustor. The principal component is the heat pipe evaporator which collects and distributes the liquid sodium over the heat transfer surfaces. The liquid sodium evaporates and flows to the condensers where it delivers its latent heat. Given here are the test results of the endurance tests run on a Gas Fired Stirling Engine (GFSE).

  9. Thermal fatigue due to stratification and thermal shock loading of piping

    Energy Technology Data Exchange (ETDEWEB)

    Schuler, X.; Herter, K.H. [Materials Testing Inst. (MPA) Univ. of Stuttgart, Stuttgart (Germany)

    2004-07-01

    Most of the fatigue relevant stresses in piping systems are caused by thermal loading. The difference between the density of the fluid caused by the temperature gradient from bottom to top of the pipe cross section combined with low flow rates can result in thermal stratification in the horizontal portions of a piping system. The hot and cold fluid levels of the stratified flow conditions are separated by an interface or mixing layer. On the other hand high flow rates can cause a temperature gradient in pipe longitudinal direction (jump of temperature) and result in a thermal shock loading on the inside pipe surface constant throughout the pipe cross section. These loading conditions impact the secondary stress and the fatigue usage analysis typically performed for piping components by equations in the technical codes. Thermal stratification in piping system causes a circumferentially varying temperature distribution in the pipe wall resulting in local through wall axial stresses and global bending stresses in the piping system. Maximum local thermal stress is found when a thin interface (mixing) layer occurs in the upper or lower parts of the pipe cross section. Maximum global thermal bending stress is found when a thin interface layer occurs in the middle of the pipe cross section. (orig.)

  10. Study of piping configurations. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Y.; Rafer, A.; Ahmed, H.

    1980-01-01

    A study of piping and elbow flexibility is performed using an analytical approach and piping analysis computer programs ADLPIPE and MARC. The study focuses on pipe loop configurations commonly used to accommodate thermal expansion in such applications as Liquid Metal Fast Breeder Reactors.

  11. Denting of coated and uncoated offshore steel pipes

    OpenAIRE

    Holm, Nicolay Line; Røshol, Eivind Torgunrud

    2015-01-01

    In the present thesis, numerical and experimental investigations of coated offshore pipelines were conducted. The objective was to investigate how the polypropylene coating affected the global response of the pipes. The present work was a continuation of previous theses done on X65 steel pipes without coating, and part of an ongoing research program between SIMLab (NTNU) and Statoil. The experimental work consisted of material testing and component tests. A tensile test of the X6...

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

    International Nuclear Information System (INIS)

    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

  13. 76 FR 17819 - Circular Welded Austenitic Stainless Pressure Pipe From the People's Republic of China...

    Science.gov (United States)

    2011-03-31

    ...-economy purchase price of steel. \\26\\ See Antidumping Duties; Countervailing Duties, 62 FR 27296, 27366...-Market Economy Wages, Duty Drawback; and Request for Comments, 71 FR 61716, 61717 (October 19, 2006... Carbon-Quality Steel Plate from Indonesia: Final Results of Expedited Sunset Review, 70 FR 45692...

  14. Fracture toughness investigation of the welded joints in austenitic piping Du-300 MFCC at Ignalina NPP

    International Nuclear Information System (INIS)

    A study is made into mechanical properties, stress intensity factors, J-R curves of welded joint materials for downcomers of pipelines and for distributing group collectors. The Du 300 of steel 08Kh18N10T, hand and automatic arc welds produced with Sv-04Kh19N11M3 welding wire as well as a heat affected zone (HAZ) metal of this joint are investigated at room and elevated (285 deg C) temperatures. It is shown that all materials studied have lower mechanical properties at 285 deg C in comparison with those at room temperature. The exceptions are represented by the constraint and the modulus of elasticity. Mechanical strength at room temperature is maximal for a well metal and minimal for a HAZ metal. At elevated temperature the strength is maximal for a HAZ metal and minimal for a base metal

  15. Evaluation of piping integrity in thinned main feedwater pipes

    International Nuclear Information System (INIS)

    Significant wall thinning due to flow accelerated corrosion(FAC) was recently reported in main feedwater pipes in 3 Korean pressurized water reactor (PWR) plants. The main feedwater pipes in one plant were repaired using overlay weld method at the outside of pipe, while those in 2 other plants were replaced with new pipes. In this study, the effect of the wall thinning in the main feedwater pipes on piping integrity was evaluated using finite element method. Especially, the effects of both the overlay weld repair and the stress concentration in notch-type thinned area on the piping integrity were investigated. The results are as follows; (1) The piping load carrying capacity may significantly decrease due to FAC. In special, the load carrying capacity of the main feedwater pipes was reduced by about 40 percent during about 140 months operation in Korean PWR plants. (2) By performing overlay weld repair at the outside of pipe, the piping load carrying capacity can increase and the stress concentration level in the thinned area can be reduced. (author)

  16. [Piping cinnamon] 791

    OpenAIRE

    W L H Skeen and Co

    2003-01-01

    279 x 211 mm. Showing female workers making cinnamon pipes. The cinnamon is placed on a low tripod formed from four sticks, and steadied with the operator's foot while the cuticle is scraped off with a small curved knife. Annotated '791' on the bottom right hand corner of the photograph. Date approximate.

  17. Aeronautical tubes and pipes

    Science.gov (United States)

    Beauclair, N.

    1984-12-01

    The main and subcomponent French suppliers of aircraft tubes and pipes are discussed, and the state of the industry is analyzed. Quality control is essential for tubes with regard to their i.d. and metallurgical compositions. French regulations do not allow welded seam tubes in hydraulic circuits unless no other form is available, and then rustproofed steel must be installed. The actual low level of orders for any run of tubes dictates that the product is only one of several among the manufacturers' line. Automation, both in NDT and quality control, assures that the tubes meet specifications. A total of 10 French companies participate in the industry, serving both civil and military needs, with some companies specializing only in titanium, steel, or aluminum materials. Concerns wishing to enter the market must upgrade their equipment to meet the higher aeronautical specifications and be prepared to furnish tubes and pipes that serve both functional and structural purposes simultaneously. Additionally, pipe-bending machines must also perform to tight specifications. Pipes can range from 0.2 mm exterior diameter to 40 mm, with wall thicknesses from 0.02 mm to 3 mm. A chart containing a list of manufacturers and their respective specifications and characteristics is presented, and a downtrend in production with reduction of personnel is noted.

  18. Probabilistic Risk Assessment: Piping Fragility due to Earthquake Fault Mechanisms

    Directory of Open Access Journals (Sweden)

    Bu Seog Ju

    2015-01-01

    Full Text Available A lifeline system, serving as an energy-supply system, is an essential component of urban infrastructure. In a hospital, for example, the piping system supplies elements essential for hospital operations, such as water and fire-suppression foam. Such nonstructural components, especially piping systems and their subcomponents, must remain operational and functional during earthquake-induced fires. But the behavior of piping systems as subjected to seismic ground motions is very complex, owing particularly to the nonlinearity affected by the existence of many connections such as T-joints and elbows. The present study carried out a probabilistic risk assessment on a hospital fire-protection piping system’s acceleration-sensitive 2-inch T-joint sprinkler components under seismic ground motions. Specifically, the system’s seismic capacity, using an experimental-test-based nonlinear finite element (FE model, was evaluated for the probability of failure under different earthquake-fault mechanisms including normal fault, reverse fault, strike-slip fault, and near-source ground motions. It was observed that the probabilistic failure of the T-joint of the fire-protection piping system varied significantly according to the fault mechanisms. The normal-fault mechanism led to a higher probability of system failure at locations 1 and 2. The strike-slip fault mechanism, contrastingly, affected the lowest fragility of the piping system at a higher PGA.

  19. An Overview of Long Duration Sodium Heat Pipe Tests

    Science.gov (United States)

    Rosenfeld, John H.; Ernst, Donald M.; Lindemuth, James E.; Sanzi, James L.; Geng, Steven M.; Zuo, Jon

    2004-02-01

    High temperature heat pipes are being evaluated for use in energy conversion applications such as fuel cells, gas turbine re-combustors, and Stirling cycle heat sources; with the resurgence of space nuclear power, additional applications include reactor heat removal elements and radiator elements. Long operating life and reliable performance are critical requirements for these applications. Accordingly long-term materials compatibility is being evaluated through the use of high temperature life test heat pipes. Thermacore, Inc. has carried out several sodium heat pipe life tests to establish long term operating reliability. Four sodium heat pipes have recently demonstrated favorable materials compatibility and heat transport characteristics at high operating temperatures in air over long time periods. A 316L stainless steel heat pipe with a sintered porous nickel wick structure and an integral brazed cartridge heater has successfully operated at 650C to 700C for over 115,000 hours without signs of failure. A second 316L stainless steel heat pipe with a specially-designed Inconel 601 rupture disk and a sintered nickel powder wick has demonstrated over 83,000 hours at 600C to 650C with similar success. A representative one-tenth segment Stirling Space Power Converter heat pipe with an Inconel 718 envelope and a stainless steel screen wick has operated for over 41,000 hours at nearly 700C. A hybrid (i.e. gas-fired and solar) heat pipe with a Haynes 230 envelope and a sintered porous nickel wick structure was operated for about 20,000 hours at nearly 700C without signs of degradation. These life test results collectively have demonstrated the potential for high temperature heat pipes to serve as reliable energy conversion system components for power applications that require long operating lifetime with high reliability. Detailed design specifications, operating history, and test results are described for each of these sodium heat pipes. Lessons learned and future life

  20. An Overview of Long Duration Sodium Heat Pipe Tests

    Science.gov (United States)

    Rosenfeld, John H.; Ernst, Donald M.; Lindemuth, James E.; Sanzi, James L.; Geng, Steven M.; Zuo, Jon

    2004-01-01

    High temperature heat pipes are being evaluated for use in energy conversion applications such as fuel cells, gas turbine re-combustors, and Stirling cycle heat sources; with the resurgence of space nuclear power, additional applications include reactor heat removal elements and radiator elements. Long operating life and reliable performance are critical requirements for these applications. Accordingly long-term materials compatibility is being evaluated through the use of high temperature life test heat pipes. Thermacore International, Inc., has carried out several sodium heat pipe life tests to establish long term operating reliability. Four sodium heat pipes have recently demonstrated favorable materials compatibility and heat transport characteristics at high operating temperatures in air over long time periods. A 3l6L stainless steel heat pipe with a sintered porous nickel wick structure and an integral brazed cartridge heater has successfully operated at 650 to 700 C for over 115,000 hours without signs of failure. A second 3l6L stainless steel heat pipe with a specially-designed Inconel 60 I rupture disk and a sintered nickel powder wick has demonstrated over 83,000 hours at 600 to 650 C with similar success. A representative one-tenth segment Stirling Space Power Converter heat pipe with an Inconel 718 envelope and a stainless steel screen wick has operated for over 41 ,000 hours at nearly 700 0c. A hybrid (i.e. gas-fired and solar) heat pipe with a Haynes 230 envelope and a sintered porous nickel wick structure was operated for about 20,000 hours at nearly 700 C without signs of degradation. These life test results collectively have demonstrated the potential for high temperature heat pipes to serve as reliable energy conversion system components for power applications that require long operating lifetime with high reliability, Detailed design specifications, operating hi story, and test results are described for each of these sodium heat pipes. Lessons

  1. Analysis of ultrasonic wave propagation in transversely isotropic austenitic welds

    International Nuclear Information System (INIS)

    Ultrasonic testing of austenitic welds is widely known to be difficult mainly due to the anisotropy and inhomogeneity of their elastic properties. This study investigates the physical phenomena of ultrasonic wave propagation and scattering in austenitic welds, modeled as homogeneous and transversely isotropic. The velocity and slowness surfaces are obtained for the transversely isotropic plane of austenitic welds, using the elasticity analysis. Also, the phenomena of wave generation, propagation and scattering in the same medium are simulated using the mass-spring lattice model. The numerical results show good qualitative agreement with the analytical results, and various waves in the numerical results are identified by comparing with the analytical results. Further development of this work will provide useful and practical results for the field ultrasonic testing of austenitic welds.

  2. MODULATED STRUCTURES AND ORDERING STRUCTURES IN ALLOYING AUSTENITIC MANGANESE STEEL

    Institute of Scientific and Technical Information of China (English)

    L. He; Z.H. Jin; J.D. Lu

    2001-01-01

    The microstructure of Fe-10Mn-2Cr-1.5C alloy has been investigated with transmission electron microscopy and X-ray diffractometer. The superlattice diffraction spots and satellite reflection pattrens have been observed in the present alloy, which means the appearence of the ordering structure and modulated structure in the alloy. It is also proved by X-ray diffraction analysis that the austenite in the alloy is more stable than that in traditional austenitic manganese steel. On the basis of this investigation,it is suggested that the C-Mn ordering clusters exist in austenitic manganese steel and the chromium can strengthen this effect by linking the weaker C-Mn couples together,which may play an important role in work hardening of austenitic manganese steel.

  3. Research on Mediate Temperature Decomposition of High Nitrogen Austenite

    Institute of Scientific and Technical Information of China (English)

    LI Xiao-ling; BEI Duo-hui; HU Ming-juan; ZHU Zu-chang

    2004-01-01

    The decomposed products from high nitrogen austenite aging at 225℃ were investigated by TEM. It is found that the shape of decomposition products inside the austenite grains is not regular and not strictly oriented. Preferential nucleation of γ-Fe4N at dislocations and grain boundaries has been observed. It also has been found that during the first stage of the high nitrogen austenite decomposition a large quantity of ultra-fine γ-Fe4N precipitate inside the parent austenite, which has been thought to be the undecomposed region before. The ultimate products are composed of highly dispersed α-Fe and γ-Fe4N, with both of them maintaining nanometer scale. The micro-hardness of them can be as high as900HV.

  4. Change in austenite transformation kinetics under hot rolling action

    International Nuclear Information System (INIS)

    The effect of hot plastic deformation on kinetics of austenite transformation both during continuous cooling and under isothermal conditions, is studied. Experiments are performed using the 40 Kh, 60 KhC2, 40KhNM and 30KhGSN2 steels. It is shown that hot working speeds up isothermal transformation of austenite of low- and medium alloyed steels in pearlite range. In medium-alloyed 30KhGSN2 40KhNM steels hot working does not speed up atherma.l austenite transformation in the pearlite range and somewhat hinders it in the bainite range, due to which hardenability must not reduce at high temperatUre thermomechanical treatment. The difference in the effect of hot working on isothermal and athermal austenite transformations is conditioned by the effect of after-deformation pauses, which are practically inevitable in cases of continuous cooling of products

  5. Chemically Induced Phase Transformation in Austenite by Focused Ion Beam

    Science.gov (United States)

    Basa, Adina; Thaulow, Christian; Barnoush, Afrooz

    2013-11-01

    A highly stable austenite phase in a super duplex stainless steel was subjected to a combination of different gallium ion doses at different acceleration voltages. It was shown that contrary to what is expected, an austenite to ferrite phase transformation occurred within the focused ion beam (FIB) milled regions. Chemical analysis of the FIB milled region proved that the gallium implantation preceded the FIB milling. High resolution electron backscatter diffraction analysis also showed that the phase transformation was not followed by the typical shear and plastic deformation expected from the martensitic transformation. On the basis of these observations, it was concluded that the change in the chemical composition of the austenite and the local increase in gallium, which is a ferrite stabilizer, results in the local selective transformation of austenite to ferrite.

  6. Effects of austenitizing temperature in quenched niobium steels

    International Nuclear Information System (INIS)

    Three steel compositions with varying Nb content were austenitized at different temperatures and quenched in cold water. Metallographic examination and hardness measurements provided a basis for explaining the hardening mechanism and the role of Nb on the process. (Author)

  7. Austenite grain growth calculation of 0.028% Nb steel

    Directory of Open Access Journals (Sweden)

    Priadi D.

    2011-01-01

    Full Text Available Modeling of microstructural evolution has become a powerful tool for materials and process design by providing quantitative relationships for microstructure, composition and processing. Insufficient attention has been paid to predicting the austenite grain growth of microalloyed steel and the effect of undissolved microalloys. In this research, we attempted to calculate a mathematical model for austenite grain growth of 0.028% Nb steel, which can account for abnormal grain growth. The quantitative calculation of austenite grain growth generated from this model fit well with the experimental grain growth data obtained during reheating of niobium steels. The results of this study showed that increasing the temperature increases the austenite grain size, with a sharp gradient observed at higher temperatures.

  8. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    International Nuclear Information System (INIS)

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The

  9. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Chopra, O. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Gruber, Eugene E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shack, William J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2010-06-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (≤3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC

  10. Components of the primary circuit of LWRs

    International Nuclear Information System (INIS)

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.)

  11. Petrol Pipe Line Telemonitoring Design

    OpenAIRE

    Abdelrasoul Jabar Alzubaidi

    2015-01-01

    Petrol pipe lines are subjected to different types of malfunction. The malfunction can happen due to technical faults or it may be due to a gangsters attack on the petrol pipes in order to hinder the petrol pumping operations.. The damage of the petrol pipes causes a loss of a large amount of petrol from the pipe lines. Petrol pipe damage also causes fires and pollution to the environments . Such operations causes a lot of loss in economy to the country concerned where the sabotag...

  12. EFFECT OF CHEMICAL COMPOSITION ON RETAINED AUSTENITE IN TRIP STEEL

    Institute of Scientific and Technical Information of China (English)

    Y. Chen; X. Chen; Q.F. Wang; G.L. Yuan; C.Y. Li; X.Y. Li; Y.X. Wang

    2002-01-01

    The systematic chemical compositions including common C, Si, Mn, Al, and micro- alloying elements of Ti and Nb were designed for high volume fraction of retained austenite as much as possible. The thermo-cycle experiments were conducted by using Gleeble 2000 thermo-dynamic test machine for finding the appropriate composition. The experimental results showed that chemical composition had a significant effect on retained austenite, and the appropriate compositions were determined for commercial production of TRIP steels.

  13. Hot-working behaviour of high-manganese austenitic steels

    OpenAIRE

    L.A. Dobrzański; A. Grajcar; W. Borek

    2008-01-01

    Purpose: The work consisted in investigation of newly elaborated high-manganese austenitic steels with Nb and Ti microadditions in variable conditions of hot-working.Design/methodology/approach: Determination of processes controlling strain hardening was carried out in continuous compression test using Gleeble 3800 thermo-mechanical simulator.Findings: It was found that they have austenite microstructure with numerous annealing twins in the initial state. Continuous compression tests ...

  14. Nickel-free austenitic stainless steels for medical applications

    OpenAIRE

    Ke Yang and Yibin Ren

    2010-01-01

    The adverse effects of nickel ions being released into the human body have prompted the development of high-nitrogen nickel-free austenitic stainless steels for medical applications. Nitrogen not only replaces nickel for austenitic structure stability but also much improves steel properties. Here we review the harmful effects associated with nickel in medical stainless steels, the advantages of nitrogen in stainless steels, and emphatically, the development of high-nitrogen nickel-free stainl...

  15. Austenitic steels for boiler elements in USC power plants

    OpenAIRE

    A. Zieliński

    2013-01-01

    Purpose: Characteristics of functional properties of austenitic-based steels used for construction of boilers with supercritical and ultra-supercritical steam parameters.Design/methodology/approach: For selected austenitic steels in as-received state and after long-term annealing microstructural investigations were carried out with scanning and transmission electron microscope.Findings: Selected characteristics of structure and functional properties of materials to be used for critical elemen...

  16. 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)

  17. Numerical analysis of residual stress in welded pipes and life assessment of pipes considering initiation and growth of SCC based on probabilistic fracture mechanics

    International Nuclear Information System (INIS)

    Stress corrosion cracking (SCC) has been observed near the welded zones of pipes made of austenitic stainless steel type 316L. Residual stress is an important factor for SCC. In the joining processes of pipes, butt welding is conducted after surface machining. Residual stress is generated by both processes. In this study, numerical analyses of the residual stress distribution by butt welding after surface machining were performed by the finite element method. The SCC initiation time was estimated by the residual stress at the inner surface. SCC growth analyses based on probability fracture mechanics were performed by using the SCC initiation time and the residual stress distribution. As a result, the residual stress distribution in the axial direction has high tensile stress of approximately 1600 MPa at the inner surface. The effect of SCC initiation time on leakage probability is not as significant as the effect of plastic strain on the crack growth rate. (author)

  18. Heat pipe applications workshop report

    Energy Technology Data Exchange (ETDEWEB)

    Ranken, W.A.

    1978-04-01

    The proceedings of the Heat Pipe Applications Workshop, held at the Los Alamos Scientific Laboratory October 20-21, 1977, are reported. This workshop, which brought together representatives of the Department of Energy and of a dozen industrial organizations actively engaged in the development and marketing of heat pipe equipment, was convened for the purpose of defining ways of accelerating the development and application of heat pipe technology. Recommendations from the three study groups formed by the participants are presented. These deal with such subjects as: (1) the problem encountered in obtaining support for the development of broadly applicable technologies, (2) the need for applications studies, (3) the establishment of a heat pipe technology center of excellence, (4) the role the Department of Energy might take with regard to heat pipe development and application, and (5) coordination of heat pipe industry efforts to raise the general level of understanding and acceptance of heat pipe solutions to heat control and transfer problems.

  19. Heat pipe applications workshop report

    International Nuclear Information System (INIS)

    The proceedings of the Heat Pipe Applications Workshop, held at the Los Alamos Scientific Laboratory October 20-21, 1977, are reported. This workshop, which brought together representatives of the Department of Energy and of a dozen industrial organizations actively engaged in the development and marketing of heat pipe equipment, was convened for the purpose of defining ways of accelerating the development and application of heat pipe technology. Recommendations from the three study groups formed by the participants are presented. These deal with such subjects as: (1) the problem encountered in obtaining support for the development of broadly applicable technologies, (2) the need for applications studies, (3) the establishment of a heat pipe technology center of excellence, (4) the role the Department of Energy might take with regard to heat pipe development and application, and (5) coordination of heat pipe industry efforts to raise the general level of understanding and acceptance of heat pipe solutions to heat control and transfer problems

  20. Effect of cyclic loading on the relaxation of residual stress in the butt-weld joints of nuclear reactor piping

    International Nuclear Information System (INIS)

    Highlights: • The accuracy of welding simulation is confirmed by comparing with experiments. • Relaxation of residual stress for piping weld due to cyclic load is investigated. • High tensile stress that occurs in front of crack tip is reduced by cyclic loading. • Mechanism of relaxation of residual stress due to cyclic loading is discussed. • Cyclic loading on the piping welds affects the suppression of crack growth. - Abstract: Weld residual stress is among the most important factors in stress corrosion cracking (SCC) of the austenitic stainless steels used for pressure boundary piping in nuclear power plants. To assess the integrity of piping, particularly over long-term operation, it is necessary to understand the effects of cyclic loading, such as that caused by an earthquake, on residual stress. In this study, finite element analyses were performed using an axisymmetric model of a 250A pipe butt weld composed of low-carbon Type 316L stainless steel. The moving heat source was simulated by a double ellipsoid model. The accuracy of the method was verified by comparing the calculated results with experimental measurements. Subsequent to the welding simulation and residual stress analysis, the effects of cyclic loading were studied by applying several axial cyclic loading patterns to the model, varying the maximum load. Higher loading caused greater relaxation of the weld residual stress near the piping welds. It was concluded that cyclic loading on piping butt welds suppresses the SCC growth by reducing the tensile residual stress at the inner surface

  1. Nondestructive quality control of multi-layer spot welds of great thickness made of austenitic basic material

    International Nuclear Information System (INIS)

    The ultrasonic testing of three layer spot welds of great thickness made of Austenitic sheets permits the classification of quality from the aspect of strength. Division into nonbinding - glueing - welding is possible without problems. The formation of cracks in the centre of the joint is a sure indication for a weld of higher strength. In a test of 4 components with a total of 196 spot welds, 11 glued positions could be determined with certainty. (orig.)

  2. Effect of ferrite formation on abnormal austenite grain coarsening in low-alloy steels during the hot rolling process

    Science.gov (United States)

    Asahi, Hitoshi; Yagi, Akira; Ueno, Masakatsu

    1998-05-01

    Abnormal coarsening of austenite (γ) grains occurred in low-alloy steels during a seamless pipe hotrolling process. Often, the grains became several hundred micrometers in diameter. This made it difficult to apply direct quenching to produce high-performance pipes. The phenomenon of grain coarsening was successfully reproduced using a thermomechanical simulator, and the factors which affected grain coarsening were clarified. The mechanism was found to be basically strain-induced grain rowth which occurred during reheating at around 930 °C. Furthermore, once a pipe temperature decreased to the dual-phase region after the minimal hot working and prior to the reheating process, the grain coarsening was more pronounced. It was understood that the formation of ferrite along grain boundaries had the role of reducing the migration of grain boundaries into neighboring grains, leaving a strain-free, recrystallized region behind. This abnormal grain coarsening was found to be effectively prevented by an addition of Nb, the content of which varied depending on the C content. The effect of the Nb addition was confirmed by an in-line test.

  3. Nitrogen segregation and blister formation of 316LN austenitic steels during electron beam welding

    International Nuclear Information System (INIS)

    Full text of publication follows: High nitrogen austenitic stainless steel (316LN) has been selected as the structure material in shield blanket and the gravity support system in ITER due to its excellent erosion/corrosion resistance, high strength and toughness. However, most of nitrogen in this steel exists in the form of solid solution. The nitrogen can segregate from the matrix material and form small blisters or defects in the welding area, resulting in mechanical property reduction, which should be considered in the design and manufacture processing. In this study, we have investigated the blister and defect formation processing during electron beam welding. Focused electron beam with 100-150 kV high voltage, 300-500 mA beam current has been applied to weld the 316LN austenitic stainless steel components under vacuum condition. The blister formation in the welding area has been observed by both SEM and TEM directly, and was further confirmed by the micro-area composition analysis. The size and density distribution of blisters and defects with the welding depth, the vacuum condition and electron beam parameters has been investigated. At the same time, the tensile strength of the welded components was examined and compared with that of the matrix material. In this report, the mechanism of nitrogen blisters formation and its effects on the mechanical property of the welding components has also been discussed. (authors)

  4. Refurbishment of the IEAR1 primary coolant system piping supports

    Energy Technology Data Exchange (ETDEWEB)

    Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel, E-mail: gfainer@ipen.br, E-mail: afaloppa@ipen.br, E-mail: calberto@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    A partial replacement of the IEA-R1 piping system was concluded in 2014. This paper presents the study and the structural analysis of the IEA-R1 primary circuit piping supports, considering all the changes involved in the replacement. The IEA-R1 is a nuclear reactor for research purposes designed by Babcox-Willcox that is operated by IPEN since 1957. The reactor life management and modernization program is being conducted for the last two decades and already resulted in a series of changes, especially on the reactor coolant system. This set of components, divided in primary and secondary circuit, is responsible for the circulation of water into the core to remove heat. In the ageing management program that includes regular inspection, some degradation was observed in the primary piping system. As result, the renewing of the piping system was conducted in 2014. Moreover the poor condition of some original piping supports gave rise to the refurbishment of all piping supports. The aim of the present work is to review the design of the primary system piping supports taking into account the current conditions after the changes and refurbishment. (author)

  5. Further development of probabilistic analysis method for lifetime determination of piping and vessels. Final report

    International Nuclear Information System (INIS)

    Within the framework of research project RS1196 the computer code PROST (Probabilistic Structure Calculation) for the quantitative evaluation of the structural reliability of pipe components has been further developed. Thereby models were provided and tested for the consideration of the damage mechanism 'stable crack growth' to determine leak and break probabilities in cylindrical structures of ferritic and austenitic reactor steels. These models are now additionally available to the model for the consideration of the damage mechanisms 'fatigue' and 'corrosion'. Moreover, a crack initiation model has been established supplementary to the treatment of initial cracks. Furthermore, the application range of the code was extended to the calculation of the growth of wall penetrating cracks. This is important for surface cracks growing until the formation of a stable leak. The calculation of the growth of the wall penetrating crack until break occurs improves the estimation of the break probability. For this purpose program modules were developed to be able to calculate stress intensity factors and critical crack lengths for wall penetrating cracks. In the frame of this work a restructuring of PROST was performed including possibilities to combine damage mechanisms during a calculation. Furthermore several additional fatigue crack growth laws were implemented. The implementation of methods to estimate leak areas and leak rates of wall penetrating cracks was completed by the inclusion of leak detection boundaries. The improved analysis methods were tested by calculation of cases treated already before. Furthermore comparative analyses have been performed for several tasks within the international activity BENCH-KJ. Altogether, the analyses show that with the provided flexible probabilistic analysis method quantitative determination of leak and break probabilities of a crack in a complex structure geometry under thermal-mechanical loading as

  6. Change of relative Gibbs energy of martensite and austenite alloys of Fe-Ni system in the pre-martensite temperature range

    International Nuclear Information System (INIS)

    Chemical potentials of the components of quenched Fe-Ni alloys (28.7-32.7 at. % Ni) with martensite and austenite structures have been found with the Touch Instant Electromotive Force method. Differences between Gibbs energies of martensite and austenite phases have been calculated in the temperature range of 253-315 K which characterize the relative thermodynamic stability of these metastable phases. By means of interpolation the temperatures were determined when Gibbs energies of alloys with both types of structures are the same. Non-chemical contribution into Gibbs energy of martensite transformation has been evaluated

  7. Stability of cracked pipe under seismic/dynamic displacement-controlled stresses. Subtask 1.2 final report

    International Nuclear Information System (INIS)

    Results of displacement-controlled pipe fracture experiments, analyses, and material characterization efforts performed within the International Piping Integrity Research Group, IPIRG, Program Subtask 1.2 are discussed. Effects of dynamic versus quasi-static and monotonic versus cyclic loading were evaluated for ductile tearing of two materials, A106 Grade B ferritic steel and TP304 austenitic steel. Twelve through-wall-cracked pipe experiments were conducted on 6-inch diameter Schedule 120 pipe at 288 C (550 F). The results indicated dynamic loading at seismic strain rates marginally increased the load-carrying capacity of austenitic steel. The ferritic steel tested was sensitive to dynamic strain-aging, and consequently, its load-carrying capacity decreased at dynamic strain rates. Two parameters were found to affect the apparent ductile crack growth resistance during cyclic loading, load ratio (R) and incremental plastic displacement that occurs in a cycle. Cyclic (R = 0) loading had minimal effect on ductile tearing for both materials. However, fully reversed loading decreased the load-carrying capacity and toughness for both materials. The incremental plastic displacement can be as important as the load ratio; however, it is harder to quantify from design stress reports. Large plastic displacements will minimize the effect of negative load ratios

  8. Magnetic properties of single crystalline expanded austenite obtained by plasma nitriding of austenitic stainless steel single crystals.

    Science.gov (United States)

    Menéndez, Enric; Templier, Claude; Garcia-Ramirez, Pablo; Santiso, José; Vantomme, André; Temst, Kristiaan; Nogués, Josep

    2013-10-23

    Ferromagnetic single crystalline [100], [110], and [111]-oriented expanded austenite is obtained by plasma nitriding of paramagnetic 316L austenitic stainless steel single crystals at either 300 or 400 °C. After nitriding at 400 °C, the [100] direction appears to constitute the magnetic easy axis due to the interplay between a large lattice expansion and the expected decomposition of the expanded austenite, which results in Fe- and Ni-enriched areas. However, a complex combination of uniaxial (i.e., twofold) and biaxial (i.e., fourfold) in-plane magnetic anisotropies is encountered. It is suggested that the former is related to residual stress-induced effects while the latter is associated to the in-plane projections of the cubic lattice symmetry. Increasing the processing temperature strengthens the biaxial in-plane anisotropy in detriment of the uniaxial contribution, in agreement with a more homogeneous structure of expanded austenite with lower residual stresses. In contrast to polycrystalline expanded austenite, single crystalline expanded austenite exhibits its magnetic easy axes along basic directions. PMID:24028676

  9. Improved testing characteristics of austenitic and mixed welds by inverse phase adaptation of phased array signals; Verbesserung der Pruefbarkeit von Austenit- und Mischschweissverbindungen durch inverse Phasenanpassung von Gruppenstrahlerzeitsignalen

    Energy Technology Data Exchange (ETDEWEB)

    Bulavinov, A.; Kroening, M.; Walte, F. [Fraunhofer Institut Zerstoerungsfreie Pruefverfahren, Saarbruecken (Germany); Reddy, K. [QNET Engineering Ltd., Chennai (India)

    2006-07-01

    Thin-walled welds (less than 10 mm wall thickness) in austenitic pipes and mixed welds are difficult to test using non-destructive methods. This is true especially for detection of closed cracks, which is impossible by radiographic testing and problematic by ultrasonic testing. Several projects in reactor safety research provided better understanding of the testing characteristics of thes compounds. The resulting rules for writing of test specifications and for qualification of the technology employed also showed the limits of radiographic and ultrasonic testing. One important result was the simulation of the propagation of ultrasonic waves in model descriptions of welded structures as a function of the shape of the product. This allows, in principle, the applicaiton of ultrasonic migration techniques which enable consideration of phase disturbances gy structural anisotropy when the time signals of the ultrasonic sensor elements are summed up. The applicability of this 'inverse phase adaptation' was proved in heterogeneous and anisotropic model test bodies and on test bodies of the reactor safety programme. The contribution outlines the fundamentals of the technology and presents preliminary findings. (orig.)

  10. Review of key predictive methods for the ductile unstable fracture in LWR pressure boundary piping

    International Nuclear Information System (INIS)

    This report presents the following key methodologies to predict the ductile unstable fracture in LWR pressure boundary piping. (1) a predictive method by net-section collapse stress criterion (2) a predictive method by crack ligament fracture stress criterion (3) a predictive method by flow stress criterion (4) a predictive method using J-integral tearing instability theory Predictive methods (1),(2) and (3) focuss on the evaluation of the net-section stress for a cracked pipe. An acceptance criterion for circumferential flaws based on the above (3) has been recently specified in the ASME Code Section X1 to assess the margin-to-fracture in the austenitic stainless steel piping. On the other hand, the predictive methodology (4) by the tearing instability analysis is the elastic-plastic fracture mechanics approach to assure the structural integrity of the nuclear piping. In addition, extensive verification test programs regarding the structural integrity assessment of the circumferentially cracked stainless steel piping are summarized. (author)

  11. Austenite Formation from Martensite in a 13Cr6Ni2Mo Supermartensitic Stainless Steel

    OpenAIRE

    Bojack, A.; Zhao, L; Morris, P. F.; Sietsma, J.

    2016-01-01

    The influence of austenitization treatment of a 13Cr6Ni2Mo supermartensitic stainless steel (X2CrNiMoV13-5-2) on austenite formation during reheating and on the fraction of austenite retained after tempering treatment is measured and analyzed. The results show the formation of austenite in two stages. This is probably due to inhomogeneous distribution of the austenite-stabilizing elements Ni and Mn, resulting from their slow diffusion from martensite into austenite and carbide and nitride dis...

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

    International Nuclear Information System (INIS)

    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

  13. A generic approach for a linear elastic fracture mechanics analysis of components containing residual stress

    International Nuclear Information System (INIS)

    A review of through thickness transverse residual stress distribution measurements in a number of components, manufactured from a range of steels, has been carried out. Residual stresses introduced by welding and mechanical deformation have been considered. The geometries consisted of welded T-plate joints, pipe butt joints, tube-on-plate joints, tubular Y-joints and tubular T-joints as well as cold bent tubes and repair welds. In addition, the collected data cover a range of engineering steels including ferritic, austenitic, C-Mn and Cr-Mo steels. The methods used to measure the residual stresses also varied. These included neutron diffraction, X-ray diffraction and deep hole drilling techniques. Measured residual stress data, normalised by their respective yield stress have shown an inverse linear correlation versus the normalised depth of the region containing the residual stress (up to 0.5 of the component thickness). A simplified generic residual stress profile based on a linear fit to the data is proposed for the case of a transverse residual tensile stress field. Whereas the profiles in assessment procedures are case specific the proposed linear profile can be varied to produce a combination of membrane and bending stress distributions to give lower or higher levels of conservatism on stress intensity factors, depending on the amount of case specific data available or the degree of safety required

  14. Analysis of water slug impact (water hammer) in steam pipes of NPP

    International Nuclear Information System (INIS)

    A severe water hammer can happen by water entrapped in steam pipe. Consequently, pipe or its components and restraints can be damaged and its function can be lost. This kind of water hammer is a potential danger for safety operation of power plants, especially nuclear power plants. This paper briefly describes generating of water slug and the calculation method for water slug impact force on piping bend, and gives a practical example of the calculation

  15. A pilot study of thermal fatigue in feed water pipes using laboratory tests

    International Nuclear Information System (INIS)

    This study, which is the first part of larger investigation of thermal fatigue cracking in feed water pipes, has shown that the experimental equipment used is quite suitable for the problem. Thick walled tubular speciments of the austenitic stainless steel SS 2333 has been cycled by internal intermittently repeated water quenching while the outer temperature has been held constant at 428 degreeC. The results obtained for crack initiation and even crack propagation are in excellent agreement with previously obtained isothermal low cycle fatigue data. (author)

  16. Corrosion and deposit evaluation in large diameter pipes using radiography

    International Nuclear Information System (INIS)

    of misinterpreting the failure of such pipe component. On the whole the TRT method overestimated the degree of penetrated corrosion attack in the tangential position in the neighbourhood of 9% and the Double Wall Technique (DWT) had a maximum underestimation of 1.3% of the corroded surface area of the pipe. The TRT measurement of depth of local corrosion (pitting) was within an accuracy of ±0.37 mm and the measurement of corroded surface area (pit diameter) using DWT was within an accuracy of ± 0.29 mm. These tolerance limits are 5% less than the wall thickness of the pipe. The statistical interpretation of the obtained results from the study can reliably be used to develop a baseline data for the investigated pipe, through which effective monitoring of corrosion and deposit can be executed. (au)

  17. Diffusion in flexible pipes

    Energy Technology Data Exchange (ETDEWEB)

    Brogaard Kristensen, S.

    2000-06-01

    This report describes the work done on modelling and simulation of the complex diffusion of gas through the wall of a flexible pipe. The diffusion and thus the pressure in annulus depends strongly on the diffusion and solubility parameters of the gas-polymer system and on the degree of blocking of the outer surface of the inner liner due to pressure reinforcements. The report evaluates the basis modelling required to describe the complex geometries and flow patterns. Qualitatively results of temperature and concentration profiles are shown in the report. For the program to serve any modelling purpose in 'real life' the results need to be validated and possibly the model needs corrections. Hopefully, a full-scale test of a flexible pipe will provide the required temperatures and pressures in annulus to validate the models. (EHS)

  18. Heat Pipes For Alyeska

    Science.gov (United States)

    1977-01-01

    The heat pipes job is to keep the arctic ground frozen. The permafrost soil alternately freezes and thaws with seasonal temperature changes causing surface dislocations and problems for the builders. In winter, a phenomenon called frost-heaving uplifts the soil. It is something like the creation of highway potholes by the freezing of rainwater below the roadbed, but frost-heaving exerts a far greater force. Thawing of the frost in the summer causes the soil to settle unevenly. Therefore it is necessary to keep the soil in a continually frozen state so the pipeline won't rupture. To solve this problem, McDonnell Douglas Corp. applied heat pipe principles in the design of the vertical supports that hold up the pipeline.

  19. Pipe and elbow ratcheting strain effects on predicted fatigue failure

    International Nuclear Information System (INIS)

    Pipe and elbow ratcheting strains and their effects on fatigue failures have been investigated in the Piping and Fitting Dynamic Reliability Research Program (PRDRRP). This program was sponsored by EPRI and NRC and was performed by General Electric from 1985 to 1988. Forty-one piping components have been tested, which consisted of elbows, tees, reducers, nozzles and supports. In addition, two piping systems were tested for seismic loads and two other systems were tested for seismic loads and two other systems were tested for water hammer loads. In the test program seismic time history inputs, typical of those expected to occur at nuclear plants were scaled up to 50 times in amplitude to investigate the dynamic behavior and the failure modes of the piping components and piping systems. Most of the systems were pressurized at room temperature and were tested to failure. High strain rossettes were installed at high stress locations to measure the cyclic strains and ratcheting strains. In addition, scratch marks were made on the high stress locations. The elongation of the scratch marks were recorded to evaluate the ratcheting strains for each of the seismic test runs. This paper presents the ratcheting strain data. The measured ratcheting strains were compared to the ratcheting strains prediction using the approach by both the Miller Model and Edmunds-Bear Model. The fatigue failure data were compared with fatigue analyses

  20. Removal of a section of the CMS beam pipe

    CERN Multimedia

    CERN Bulletin

    2013-01-01

    Over recent weeks, members of the TE-VSC group have been removing seven components of the beam pipe located at the heart of the CMS detector. The delicate operations involved have been performed in several stages as the detector opening work has progressed.   Of the seven components concerned, only the central vacuum pipe will be replaced. The other six will be stored in a special radiation-shielded area on the surface and subsequently reinstalled ready for the resumption of machine operation. The video below, which was filmed on 15 May, shows one of the seven components of the vacuum pipe - the HFCT2, located to the right of the interaction point – being brought up from the CMS cavern to the surface by the transport team at Point 5.

  1. LHCb: Beam Pipe

    CERN Multimedia

    LHCb, Collaboration

    2005-01-01

    The proton beams circulate in the accelerator in Ultra High Vacuum to make them interact only with each other when colliding at the interaction point. A special beam pipe "holds" the vacuum where they pass through the LHCb detector:it has to be mechanically very strong to stand the difference in pressure between the vacuum inside it and the air in the cavern but also be as transparent as possible for the particles originating in the proton−proton collisions.

  2. LHCb: Beam Pipe portrait

    CERN Multimedia

    LHCb, Collaboration

    2005-01-01

    The proton beams circulate in the accelerator in Ultra High Vacuum to make them interact only with each other when colliding at the interaction point. A special beam pipe "holds" the vacuum where they pass through the LHCb detector: it has to be mechanically very strong to stand the difference in pressure between the vacuum inside it and the air in the cavern but also be as transparent as possible for the particles originating in the proton−proton collisions.

  3. Recent progress in structural integrity assessment techniques for components subject to service-induced degradation

    International Nuclear Information System (INIS)

    Nuclear power plant components are exposed to a wide range of environmental and loading conditions which can cause degradation over time. Aging embrittlement, erosion-corrosion, irradiation embrittlement, stress corrosion cracking, and corrosion fatigue are examples of aging mechanisms which could reduce structural margins in reactor components. The degradation effects from these mechanisms have been seen more frequently with the aging of the early nuclear plants. Since there is a strong incentive for keeping these older plants running for longer periods of time without compromising safety, proper plant management to minimize damage from degradation mechanisms is extremely important. Structural margin assessment, monitoring, and maintenance are important elements of such a management plan. Significant progress has been recently made in the understanding, evaluation and monitoring of these degradation mechanisms. This has led also to new requirements in the ASME Code design basis for nuclear plants. Current state of understanding and new developments in the ASME Code to address some of these degradation mechanisms are covered in this paper. Cast stainless steels used in pump casings and valve bodies have been known to experience thermal aging embrittlement at reactor operating temperatures. Recent predictive models of thermal aging effects on material toughness, developed at Argonne National Lab are reviewed and applied to assess ASME Code structural margins of a reactor pump casing. A recent ASME Code Case provides methods for the evaluation and acceptance criteria for reactor pressure vessels having ductile fracture toughness values reduced below the requirements of 10CFR50 due to irradiation embrittlement. Background and application of this code case to an older BWR vessel is described. The occurrence of stress corrosion cracking in austenitic stainless steel piping highlighted the need for evaluation methods for structural margin assessment in piping

  4. Inservice inspection of primary circuit components of VVER 440-Type nuclear power plants

    International Nuclear Information System (INIS)

    The Technical Research Centre of Finland (VTT) has since 1979 performed the inservice inspections of the primary circuits, steam generators and other ASME XI safety class 1 components of the Loviisa reactors (VVER440). A presentation of the performance of the inservice inspections is given in this paper. The main volumetric examination method used in the inservice inspection is the ultrasonic examination. The primary piping of Loviisa reactors is made of austenitic steel and the conventional ultrasonic technique cannot be applied due to the strong attenuation of ultrasonic waves. The special technique developed for the ultrasonic inspection of the welds in the primary piping as well as for the welds of main gate valves will be presented. As these inspections especially in older reactors have to be carried out in a radioactive environment, mechanized inspection equipment has been constructed to perform the work. An inspection manipulator is also used to inspect the base metal of the primary circuit in areas where mixing up of hot and cool water can cause thermal shocks and consequently lead to cracking. Because the ultrasonic inspection of main gate valves is strongly restricted by the valve geometry and material properties, an acoustic emission technique has been developed to improve the reliability of the inservice inspections. For steam generator tubing the eddy current inspection technique has been applied. Due to the different design of steam generators in VVER440 reactors, the performance of the inspection differs from the practice normally applied in U-tube steam generators. For the inspection of studs in primary circuit components both ultrasonic and eddy current techniques have been developed. A mechanized inspection equipment has been constructed to perform both ultrasonic and eddy current inspections simultaneously for large studs. For smaller studs and for threads in stud holes e.g. in steam generators mechanized eddy current inspection technique has

  5. Vibration analysis for condition monitoring on piping elbow thinning

    International Nuclear Information System (INIS)

    The development of a condition monitoring system is highly required to evaluate the severity of erosion/corrosion phenomenon in the piping components and to take actions before some events should occur in nuclear power plants. Vibration analysis is conducted for condition monitoring of the dynamic behavior by wall-thinning effect. This paper deals with analysis of the local vibration of piping elbows, which are thinned by flow-accelerated corrosion (FAC). To evaluate a change of the dynamic characteristics due to the flow induced vibration by thinning the piping elbow, we have performed the simulated FAC test and the chemical FAC test with the three-axis micro piezoelectric accelerometer. With simulated FAC test, we found the possibility to identify and to measure the change of the characteristics of the vibration signal originated from the FAC phenomenon on a piping system

  6. Pipe inspection using the pipe crawler. Innovative technology summary report

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) continually seeks safer and more cost-effective remediation technologies for use in the decontamination and decommissioning (D and D) of nuclear facilities. In several of the buildings at the Fernald Site, there is piping that was used to transport process materials. As the demolition of these buildings occur, disposal of this piping has become a costly issue. Currently, all process piping is cut into ten-foot or less sections, the ends of the piping are wrapped and taped to prevent the release of any potential contaminants into the air, and the piping is placed in roll off boxes for eventual repackaging and shipment to the Nevada Test Site (NTS) for disposal. Alternatives that allow for the onsite disposal of process piping are greatly desired due to the potential for dramatic savings in current offsite disposal costs. No means is currently employed to allow for the adequate inspection of the interior of piping, and consequently, process piping has been assumed to be internally contaminated and thus routinely disposed of at NTS. The BTX-II system incorporates a high-resolution micro color camera with lightheads, cabling, a monitor, and a video recorder. The complete probe is capable of inspecting pipes with an internal diameter (ID) as small as 1.4 inches. By using readily interchangeable lightheads, the same system is capable of inspecting piping up to 24 inches in ID. The original development of the BTX system was for inspection of boiler tubes and small diameter pipes for build-up, pitting, and corrosion. However, the system is well suited for inspecting the interior of most types of piping and other small, confined areas. The report describes the technology, its performance, uses, cost, regulatory and policy issues, and lessons learned

  7. Pipe inspection using the pipe crawler. Innovative technology summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-05-01

    The US Department of Energy (DOE) continually seeks safer and more cost-effective remediation technologies for use in the decontamination and decommissioning (D and D) of nuclear facilities. In several of the buildings at the Fernald Site, there is piping that was used to transport process materials. As the demolition of these buildings occur, disposal of this piping has become a costly issue. Currently, all process piping is cut into ten-foot or less sections, the ends of the piping are wrapped and taped to prevent the release of any potential contaminants into the air, and the piping is placed in roll off boxes for eventual repackaging and shipment to the Nevada Test Site (NTS) for disposal. Alternatives that allow for the onsite disposal of process piping are greatly desired due to the potential for dramatic savings in current offsite disposal costs. No means is currently employed to allow for the adequate inspection of the interior of piping, and consequently, process piping has been assumed to be internally contaminated and thus routinely disposed of at NTS. The BTX-II system incorporates a high-resolution micro color camera with lightheads, cabling, a monitor, and a video recorder. The complete probe is capable of inspecting pipes with an internal diameter (ID) as small as 1.4 inches. By using readily interchangeable lightheads, the same system is capable of inspecting piping up to 24 inches in ID. The original development of the BTX system was for inspection of boiler tubes and small diameter pipes for build-up, pitting, and corrosion. However, the system is well suited for inspecting the interior of most types of piping and other small, confined areas. The report describes the technology, its performance, uses, cost, regulatory and policy issues, and lessons learned.

  8. The effect of cooling rate and austenite grain size on the austenite to ferrite transformation temperature and different ferrite morphologies in microalloyed steels

    International Nuclear Information System (INIS)

    The effect of different austenite grain size and different cooling rates on the austenite to ferrite transformation temperature and different ferrite morphologies in one Nb-microalloyed high strength low alloy steel has been investigated. Three different austenite grain sizes were selected and cooled at two different cooling rates for obtaining austenite to ferrite transformation temperature. Moreover, samples with specific austenite grain size have been quenched, partially, for investigation on the microstructural evolution. In order to assess the influence of austenite grain size on the ferrite transformation temperature, a temperature differences method is established and found to be a good way for detection of austenite to ferrite, pearlite and sometimes other ferrite morphologies transformation temperatures. The results obtained in this way show that increasing of austenite grain size and cooling rate has a significant influence on decreasing of the ferrite transformation temperature. Micrographs of different ferrite morphologies show that at high temperatures, where diffusion rates are higher, grain boundary ferrite nucleates. As the temperature is lowered and the driving force for ferrite formation increases, intragranular sites inside the austenite grains become operative as nucleation sites and suppress the grain boundary ferrite growth. The results indicate that increasing the austenite grain size increases the rate and volume fraction of intragranular ferrite in two different cooling rates. Moreover, by increasing of cooling rate, the austenite to ferrite transformation temperature decreases and volume fraction of intragranular ferrite increases.

  9. Ion-nitriding of austenitic stainless steels

    International Nuclear Information System (INIS)

    Although ion-nitriding is an extensively industrialized process enabling steel surfaces to be hardened by nitrogen diffusion, with a resulting increase in wear, seizure and fatigue resistance, its direct application to stainless steels, while enhancing their mechanical properties, also causes a marked degradation in their oxidation resistance. However, by adaption of the nitriding process, it is possible to maintain the improved wear resistant properties while retaining the oxidation resistance of the stainless steel. The controlled diffusion permits the growth of a nitrogen supersaturated austenite layer on parts made of stainless steel (AISI 304L and 316L) without chromium nitride precipitation. The diffusion layer remains stable during post heat treatments up to 650 F for 5,000 hrs and maintains a hardness of 900 HV. A very low and stable friction coefficient is achieved which provides good wear resistance against stainless steels under diverse conditions. Electrochemical and chemical tests in various media confirm the preservation of the stainless steel characteristics. An example of the application of this process is the treatment of Reactor Control Rod Cluster Assemblies (RCCAs) for Pressurized Water Nuclear Reactors

  10. He blisters on welded austenitic stainless steel

    International Nuclear Information System (INIS)

    Surface blisters of single-crystal and polycrystalline metals induced by He-ion irradiation have been investigated by many researchers and several blister-formation mechanisms have been proposed. But there is no report on what blister densities and blister sizes are to be expected on a welded 316 austenitic stainless steel in use as a fusion reactor material. An experiment was carried out, and details are given. The exfoliation of blisters was almost not observed until the total dose of 2 x 1022 ions m-2 was reached. A figure shows the blister densities for every increment in blister diameter of 0.5 μm on the base and weld metals. A second figure shows the corresponding blister densities on the base and weld metals annealed at 653 K for 4.5 ksec after He-ion irradiation. The total blister densities of the base metals decrease to 4.3 to 5.5 x 1010 blisters m-2 and the average blister sizes increase to 2.8 to 3.2 μm. This phenomenon indicates that the implanted He ions diffuse in the weld and base metals. The blister sizes on the weld metals are smaller than those on the base metals and the densities on the weld metals are greater than those on the base metals. (author)

  11. The Impact Of the Welded Joints Made Of X8CrNiTi18–10 Stainless Steel on the Reliability Estimation of Pipes

    Directory of Open Access Journals (Sweden)

    Raimondas Skindaras

    2011-02-01

    Full Text Available The chrome-nickel stainless steels of austenitic class applied in chemistry and energy industry are often used in the production of exceptional structures employed in an environment aggressive and dangerous for human life. Therefore, it is particularly significant for durability and reliability requirements. The article explores cracks that appeared in a tube made of X8CrNiTi18–10 austenitic steel. The examined pipe has worked for 90 000 hours under high temperature and pressure in an aggressive media. To establish reasons for developed cracks, chemical composition, strength, hardness testing and metallographic structures have been examined. The performed investigations will prevent from potential emergencies and help with a more accurate assessment of the pipes made of particularly this class of steel in order to ensure operational reliability and durability in the future.Article in Lithuanian

  12. 49 CFR 192.279 - Copper pipe.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Copper pipe. 192.279 Section 192.279... Copper pipe. Copper pipe may not be threaded except that copper pipe used for joining screw fittings or... heavier wall pipe listed in Table C1 of ASME/ANSI B16.5....

  13. Steels for large steam pipes

    International Nuclear Information System (INIS)

    Presented are the results of pilot-scale operation and scientific investigation of metal of pipes 500 to 1100 mm in diameter manufactured by various techniques from 15Kh1M1F steel. The structure and the properties of the metal of ten meltings were investigated for homogeneity in the following zones: base metal, near-weld zone and weld metal. It was found that the macrostructure of pipes cast by a centrifugal method has a fairly dense structure; there is homogeneity across and lengthwise of the pipe. After a heat treatment, the metal of centrifugally cast pipes has both high short-time and long-time properties. Introduction of centrifugally cast pipe manufacture opens possibilities for using larger pipes in steam conduits of hot intermediate superheating devices

  14. Creep of welded branched pipes

    OpenAIRE

    Rayner, Glen

    2004-01-01

    Creep failure of welds in high-temperature power plant steam piping systems is known to be a potential cause of plant failure. Creep behaviour of plain pipes with circumferential welds and cross-weld specimens have received fairly extensive attention. However, research into the creep behaviour of welded thick-walled branched steam pipes has received less attention. Consequently, this thesis addresses improving the understanding of the creep behaviour for this type of geometry. Numerical and a...

  15. Pipe Lines – External Corrosion

    OpenAIRE

    Dan Babor

    2008-01-01

    Two areas of corrosion occur in pipe lines: corrosion from the medium carried inside the pipes; corrosion attack upon the outside of the pipes (underground corrosion. Electrolytic processes are also involved in underground corrosion. Here the moisture content of the soil acts as an electrolyte, and the ions required to conduct the current are supplied by water-soluble salts (chlorides, sulfates, etc.) present in the soil. The nature and amount of these soluble materials can vary within a wide...

  16. Dose dependence of the microstructural evolution in neutron-irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Microstructural data on the evolution of the dislocation loop, cavity, and precipitate populations in neutron-irradiated austenitic stainless steels are reviewed in order to estimate the displacement damage levels needed to achieve the 'steady state' condition. The microstructural data can be conveniently divided into two temperature regimes. In the low temperature regime (below about 200 degrees C) the microstructure of austenitic stainless steel is dominated by 'black spot' defect clusters and faulted interstitial dislocation loops. The dose needed to approach saturation of the loop and defect cluster densities is generally on the order of 1 displacement per atom (dpa) in this regime. In the high temperature regime (∼300 to 700 degrees C), cavities, precipitates, loops and network dislocations are all produced during irradiation; doses in excess of 10 dpa are generally required to approach a 'steady state' microstructural condition. Due to complex interactions between the various microstructural components that form during irradiation, a secondary transient regime is typically observed in commercial stainless steels during irradiation at elevated temperatures. This slowly evolving secondary transient may extend to damage levels in excess of 50 dpa in typical 300-series stainless steels, and to >100 dpa in radiation-resistant developmental steels. The detailed evolution of any given microstructural component in the high-temperature regime is sensitive to slight variations in numerous experimental variables, including heat-to-heat composition changes and neutron spectrum

  17. Dynamic experiments on cracked pipes

    International Nuclear Information System (INIS)

    In order to apply the leak before break concept to piping systems, the behavior of cracked pipes under dynamic, and especially seismic loading must be studied. In a first phase, an experimental program on cracked stainless steel pipes under quasi-static monotonic loading has been conducted. In this paper, the dynamic tests on the same pipe geometry are described. These tests have been performed on a shaking table with a mono frequency input signal. The main parameter of the tests is the frequency of excitation versus the frequency of the system

  18. Flooding phenomena in inclined pipes

    International Nuclear Information System (INIS)

    The flooding phenomena for air-water counter-current two-phase flow in inclined pipes were investigated experimentally. The inner diameter was 16 mm. The examined pipe inclination were 30deg, 45deg and 60deg from horizontal and the pipe length ranged from 0.5 to 5.5 m respectively. The present results indicate that the pipe length affects the flooding mechanism and the onset of flooding velocity. Finally, a simple correlation to predict the void fraction at the onset of flooding is proposed. (author)

  19. Temperature compensating stiff pipe clamp

    International Nuclear Information System (INIS)

    A new type of non-integral pipe attachment for nuclear piping seismic restraint that allows the pipe free thermal diametric expansion without constraint when using dissimilar pipe and clamp material is described. The clamp has a high spring rate that can be controlled by variable stiffness parameters in the design. Described in detail are thermal constraint stress, load stress distribution, spring rates, load angles and design philosophy. Analytical methods of code design, fabrication techniques, cost benefits and lead time reduction techniques are presented. 5 refs

  20. X-ray stress measurement of high manganese austenitic steels

    International Nuclear Information System (INIS)

    By using psi0 oscillation method with CrKβ γ(311) diffraction, the X-ray stress measurement was made on five specimens, which were obtained by water toughening and various plastic working treatments (tensile pre-strain, hammered and explosive hardening) from Hadfield's high manganese austenitic steel, and their mechanical elastic constants, X-ray elastic constants and the accuracy of measurement were examined. The results obtained are as follows: (1) The mechanical elastic constant for 13% Mn austenitic steel after water toughening was 18900 kg/mm2. This value gradually decreased with increasing tensile pre-strain. But it changed little when the specimen was treated by hammered or explosive hardening. (2) The K-value (K sub(X) = -37.26 kg/mm2/deg) of austenitic steel adopted in the standard method of X-ray stress measurement is considered suitable for the X-ray stress measurement of 13% Mn austenitic steel which has not been subjected to severe plastic deformation. (3) The error in stress measurement Δσ in the X-ray stress measurement of high manganese austenitic steel becomes larger, depending more on the statistical fluctuation parameter, than that of ferritic steel. (author)

  1. Austenitic stainless steel patterning by plasma assisted diffusion treatments

    International Nuclear Information System (INIS)

    The new concept of surface texturing or surface patterning on austenitic stainless steel by plasma assisted diffusion treatment is presented in this paper. It allows the creation of uniform micro or nano relief with regularly shaped asperities or depressions. Plasma assisted diffusion treatments are based on the diffusion of nitrogen and/or carbon in a metallic material at moderate to elevated temperatures. Below 420 deg. C, a plasma assisted nitriding treatment of austenitic stainless steel produces a phase usually called expanded austenite. Expanded austenite is a metastable nitrogen supersaturated solid solution with a disordered fcc structure and a distorted lattice. The nitrided layer with the expanded austenite is highly enriched in nitrogen (from 10 to 35 at%) and submitted to high compressive residual stresses. From mechanical consideration, it is shown that the only possible deformation occurs in the direction perpendicular to the surface. Such an expansion of the layer from the initial surface of the substrate to the gas phase is used here for surface patterning of stainless steel parts. The surface patterning is performed by using masks (TEM grid) and multi-dipolar plasmas.

  2. Ferrite stability in duplex austenitic stainless steel welds

    International Nuclear Information System (INIS)

    The presence of ferrite in austenitic stainless steel welds is known to be beneficial in avoiding hot cracking problems. In particular, the primary delta ferrite mode of solidification is important. For alloy compositions in which primary ferrite forms, it has been shown that up to approximately 40% ferrite may exist in the as-solidified structures. With further cooling, the ferrite becomes unstable, transforming to austenite. However, under typical welding conditions, the cooling rate is sufficiently high to suppress the complete transformation of ferrite and some residual ferrite is retained. For example, for Type 308 austenitic stainless steel filler metal, gas-tungsten arc welds contain 6 to 10% ferrite, although under equilibrium conditions at elevated temperatures, this same alloy can be homogenized into a fully austenitic structure. Thus, it is clear the retained ferrite in such duplex structure welds is unstable and transforms during elevated temperature applications. The stability of ferrite was investigated by measuring its composition after several different thermal treatments. The composition was measured by means of analytical electron microscopy of thinned foils, and only the major constituents, iron, chromium, and nickel, were analyzed. The composition of ferrite was measured as a function of aging time and temperature. It was found that, during aging, the ferrite composition changes and approaches a metastable equilibrium limit before eventually transforming to sigma phase or austenite. This limiting composition was determined as a function of temperature

  3. In situ observations of austenite grain growth in Fe-C-Mn-Si super bainitic steel

    Institute of Scientific and Technical Information of China (English)

    Feng Liu; Guang Xu; Yu-long Zhang; Hai-jiang Hu; Lin-xin Zhou; Zheng-liang Xue

    2013-01-01

    In situ observations of austenite grain growth in Fe-C-Mn-Si super bainitic steel were conducted on a high-temperature laser scanning confocal microscope during continuous heating and subsequent isothermal holding at 850, 1000, and 1100◦C for 30 min. A grain growth model was proposed based on experimental results. It is indicated that the austenite grain size increases with austenitizing temperature and holding time. When the austenitizing temperature is above 1100◦C, the austenite grains grow rapidly, and abnormal austenite grains occur. In addition, the eff ect of heating rate on austenite grain growth was investigated, and the relation between austenite grains and bainite morphology after bainitic transformations was also discussed.

  4. Evaluation of Microstructure and Mechanical Properties in Dissimilar Austenitic/Super Duplex Stainless Steel Joint

    Science.gov (United States)

    Rahmani, Mehdi; Eghlimi, Abbas; Shamanian, Morteza

    2014-10-01

    To study the effect of chemical composition on microstructural features and mechanical properties of dissimilar joints between super duplex and austenitic stainless steels, welding was attempted by gas tungsten arc welding process with a super duplex (ER2594) and an austenitic (ER309LMo) stainless steel filler metal. While the austenitic weld metal had vermicular delta ferrite within austenitic matrix, super duplex stainless steel was mainly comprised of allotriomorphic grain boundary and Widmanstätten side plate austenite morphologies in the ferrite matrix. Also the heat-affected zone of austenitic base metal comprised of large austenite grains with little amounts of ferrite, whereas a coarse-grained ferritic region was observed in the heat-affected zone of super duplex base metal. Although both welded joints showed acceptable mechanical properties, the hardness and impact strength of the weld metal produced using super duplex filler metal were found to be better than that obtained by austenitic filler metal.

  5. A study of ceramic-lined composite steel pipes prepared by SHS centrifugal-thermite process

    Directory of Open Access Journals (Sweden)

    Li Yuxin

    2016-01-01

    Full Text Available Al2O3 ceramic-lined steel pipe was produced by self-propagating high-temperature synthesis centrifugal thermite process (SHS C-T process from Fe2O3 and Al as the raw materials. The composition, phase separation and microstructures were investigated. The result showed the ceramic lined pipe is composed of the three main layers of various compositions, which were subsequently determined to be Fe layer, the transition layer and the ceramic layer. Fe layer is composed of austenite and ferrite, the transition layer consisted of Al2O3 ceramic and Fe, the ceramic layer consisted of the dendritic-shaped Al2O3 and the spinel-shaped structured FeAl2O4.

  6. Neutron diffraction residual stress measurements on girth-welded 304 stainless steel pipes with weld metal deposited up to half and full pipe wall thickness

    International Nuclear Information System (INIS)

    The residual stress distribution has been measured in two girth-welded austenitic stainless steel pipe weldments using time-of-flight neutron diffraction. One had weld filler metal deposited up to half the pipe wall thickness, and one had weld metal deposited up to full pipe wall thickness. The aim of the work is to evaluate the evolution in residual stress profile on filling the weld, on which there is little experimental data, and where the selection of the correct hardening model used in finite element modelling can benefit greatly from an understanding of the intermediate residual stresses partway through the welding operation. The measured residual stresses are compared with those calculated by finite element modelling and measured using X-ray diffraction. The results show a change in the measured hoop stress at the weld toe from tension to compression between the half- and fully-filled weld. The finite element results show an overprediction of the residual stress, which may be a consequence of the simple isotropic hardening model applied. The results have implications for the likely occurrence of stress corrosion cracking in this important type of pipe-to-pipe weldment. Highlights: ► 304 steel girth welded with weld metal to half and full pipe wall thickness. ► Residual stresses measured by neutron and X-ray diffraction, and modelled by FE. ► Weld toe residual σhoop changes from tensile to compressive from half to fully-filled. ► FE model for the fully-filled weld gives higher stress levels than those measured. ► Discrepancy is attributed to the isotropic hardening model used.

  7. Effects of Thermocapillary Forces during Welding of 316L-Type Wrought, Cast and Powder Metallurgy Austenitic Stainless Steels

    CERN Document Server

    Sgobba, Stefano

    2003-01-01

    The Large Hadron Collider (LHC) is now under construction at the European Organization for Nuclear Research (CERN). This 27 km long accelerator requires 1248 superconducting dipole magnets operating at 1.9 K. The cold mass of the dipole magnets is closed by a shrinking cylinder with two longitudinal welds and two end covers at both extremities of the cylinder. The end covers, for which fabrication by welding, casting or Powder Metallurgy (PM) was considered, are dished-heads equipped with a number of protruding nozzles for the passage of the different cryogenic lines. Structural materials and welds must retain high strength and toughness at cryogenic temperature. AISI 316L-type austenitic stainless steel grades have been selected because of their mechanical properties, ductility, weldability and stability of the austenitic phase against low-temperature spontaneous martensitic transformation. 316LN is chosen for the fabrication of the end covers, while the interconnection components to be welded on the protrud...

  8. Finite element thermal analysis of the fusion welding of a P92 steel pipe

    Directory of Open Access Journals (Sweden)

    A. H. Yaghi

    2012-05-01

    Full Text Available Fusion welding is common in steel pipeline construction in fossil-fuel power generation plants. Steel pipes in service carry steam at high temperature and pressure, undergoing creep during years of service; their integrity is critical for the safe operation of a plant. The high-grade martensitic P92 steel is suitable for plant pipes for its enhanced creep strength. P92 steel pipes are usually joined together with a similar weld metal. Martensitic pipes are sometimes joined to austenitic steel pipes using nickel based weld consumables. Welding involves severe thermal cycles, inducing residual stresses in the welded structure, which, without post weld heat treatment (PWHT, can be detrimental to the integrity of the pipes. Welding residual stresses can be numerically simulated by applying the finite element (FE method in Abaqus. The simulation consists of a thermal analysis, determining the temperature history of the FE model, followed by a sequentially-coupled structural analysis, predicting residual stresses from the temperature history.

    In this paper, the FE thermal analysis of the arc welding of a typical P92 pipe is presented. The two parts of the P92 steel pipe are joined together using a dissimilar material, made of Inconel weld consumables, producing a multi-pass butt weld from 36 circumferential weld beads. Following the generation of the FE model, the FE mesh is controlled using Model Change in Abaqus to activate the weld elements for each bead at a time corresponding to weld deposition. The thermal analysis is simulated by applying a distributed heat flux to the model, the accuracy of which is judged by considering the fusion zones in both the parent pipe as well as the deposited weld metal. For realistic fusion zones, the heat flux must be prescribed in the deposited weld pass and also the adjacent pipe elements. The FE thermal results are validated by comparing experimental temperatures measured by five thermocouples on the

  9. Study on Domestication of Ultra-Supercritical P92 Steel Pipe Fittings

    Institute of Scientific and Technical Information of China (English)

    Cong Xiangzhou; Xu Guangxin; Wei Xiao; An Jinping; Peng Xiankuan; Hui Na; Ye Qing

    2007-01-01

    @@ P92 steel pipe fittings are key components for domestic ultra-supercritical power units.Although under booming development in some countries,presently only a few foreign units under commercial operation are using P92 steel and the experience on fabrication and operation of P92 steel pipe fittings is insufficient.

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

    International Nuclear Information System (INIS)

    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

  11. Numerical simulation of hydrogen-assisted crack initiation in austenitic-ferritic duplex steels

    International Nuclear Information System (INIS)

    Duplex stainless steels have been used for a long time in the offshore industry, since they have higher strength than conventional austenitic stainless steels and they exhibit a better ductility as well as an improved corrosion resistance in harsh environments compared to ferritic stainless steels. However, despite these good properties the literature shows some failure cases of duplex stainless steels in which hydrogen plays a crucial role for the cause of the damage. Numerical simulations can give a significant contribution in clarifying the damage mechanisms. Because they help to interpret experimental results as well as help to transfer results from laboratory tests to component tests and vice versa. So far, most numerical simulations of hydrogen-assisted material damage in duplex stainless steels were performed at the macroscopic scale. However, duplex stainless steels consist of approximately equal portions of austenite and δ-ferrite. Both phases have different mechanical properties as well as hydrogen transport properties. Thus, the sensitivity for hydrogen-assisted damage is different in both phases, too. Therefore, the objective of this research was to develop a numerical model of a duplex stainless steel microstructure enabling simulation of hydrogen transport, mechanical stresses and strains as well as crack initiation and propagation in both phases. Additionally, modern X-ray diffraction experiments were used in order to evaluate the influence of hydrogen on the phase specific mechanical properties. For the numerical simulation of the hydrogen transport it was shown, that hydrogen diffusion strongly depends on the alignment of austenite and δ-ferrite in the duplex stainless steel microstructure. Also, it was proven that the hydrogen transport is mainly realized by the ferritic phase and hydrogen is trapped in the austenitic phase. The numerical analysis of phase specific mechanical stresses and strains revealed that if the duplex stainless steel is

  12. Flexible PVDE comb transducers for excitation of axisymmetric guided waves in pipe

    International Nuclear Information System (INIS)

    Flexible PVDF pipe comb transducers are easy to install by wrapping around any size pipe. It is possible to mechanically couple these transducers to the pipe thereby eliminating the need to bond electrodes to the film and couple the transducer to the pipe. The simple fabrication process, installation, and affordability of these transducers makes them realistic candidates for condition based monitoring of some critical pipeline applications. These transducers are capable of exciting lower order axisymmetric modes with minimal radial displacement and maximum axial displacement as well as modes with both surface displacement components. This versatility is extremely important since under certain loading conditions modes with significant radial displacement are almost completely attenuated.

  13. Large-bore pipe decontamination

    International Nuclear Information System (INIS)

    The decontamination and decommissioning (D and D) of 1200 buildings within the US Department of Energy-Office of Environmental Management (DOE-EM) Complex will require the disposition of miles of pipe. The disposition of large-bore pipe, in particular, presents difficulties in the area of decontamination and characterization. The pipe is potentially contaminated internally as well as externally. This situation requires a system capable of decontaminating and characterizing both the inside and outside of the pipe. Current decontamination and characterization systems are not designed for application to this geometry, making the direct disposal of piping systems necessary in many cases. The pipe often creates voids in the disposal cell, which requires the pipe to be cut in half or filled with a grout material. These methods are labor intensive and costly to perform on large volumes of pipe. Direct disposal does not take advantage of recycling, which could provide monetary dividends. To facilitate the decontamination and characterization of large-bore piping and thereby reduce the volume of piping required for disposal, a detailed analysis will be conducted to document the pipe remediation problem set; determine potential technologies to solve this remediation problem set; design and laboratory test potential decontamination and characterization technologies; fabricate a prototype system; provide a cost-benefit analysis of the proposed system; and transfer the technology to industry. This report summarizes the activities performed during fiscal year 1997 and describes the planned activities for fiscal year 1998. Accomplishments for FY97 include the development of the applicable and relevant and appropriate regulations, the screening of decontamination and characterization technologies, and the selection and initial design of the decontamination system

  14. Large-bore pipe decontamination

    Energy Technology Data Exchange (ETDEWEB)

    Ebadian, M.A.

    1998-01-01

    The decontamination and decommissioning (D and D) of 1200 buildings within the US Department of Energy-Office of Environmental Management (DOE-EM) Complex will require the disposition of miles of pipe. The disposition of large-bore pipe, in particular, presents difficulties in the area of decontamination and characterization. The pipe is potentially contaminated internally as well as externally. This situation requires a system capable of decontaminating and characterizing both the inside and outside of the pipe. Current decontamination and characterization systems are not designed for application to this geometry, making the direct disposal of piping systems necessary in many cases. The pipe often creates voids in the disposal cell, which requires the pipe to be cut in half or filled with a grout material. These methods are labor intensive and costly to perform on large volumes of pipe. Direct disposal does not take advantage of recycling, which could provide monetary dividends. To facilitate the decontamination and characterization of large-bore piping and thereby reduce the volume of piping required for disposal, a detailed analysis will be conducted to document the pipe remediation problem set; determine potential technologies to solve this remediation problem set; design and laboratory test potential decontamination and characterization technologies; fabricate a prototype system; provide a cost-benefit analysis of the proposed system; and transfer the technology to industry. This report summarizes the activities performed during fiscal year 1997 and describes the planned activities for fiscal year 1998. Accomplishments for FY97 include the development of the applicable and relevant and appropriate regulations, the screening of decontamination and characterization technologies, and the selection and initial design of the decontamination system.

  15. Laser etching of austenitic stainless steels for micro-structural evaluation

    Science.gov (United States)

    Baghra, Chetan; Kumar, Aniruddha; Sathe, D. B.; Bhatt, R. B.; Behere, P. G.; Afzal, Mohd

    2015-06-01

    Etching is a key step in metallography to reveal microstructure of polished specimen under an optical microscope. A conventional technique for producing micro-structural contrast is chemical etching. As an alternate, laser etching is investigated since it does not involve use of corrosive reagents and it can be carried out without any physical contact with sample. Laser induced etching technique will be beneficial especially in nuclear industry where materials, being radioactive in nature, are handled inside a glove box. In this paper, experimental results of pulsed Nd-YAG laser based etching of few austenitic stainless steels such as SS 304, SS 316 LN and SS alloy D9 which are chosen as structural material for fabrication of various components of upcoming Prototype Fast Breeder Reactor (PFBR) at Kalpakkam India were reported. Laser etching was done by irradiating samples using nanosecond pulsed Nd-YAG laser beam which was transported into glass paneled glove box using optics. Experiments were carried out to understand effect of laser beam parameters such as wavelength, fluence, pulse repetition rate and number of exposures required for etching of austenitic stainless steel samples. Laser etching of PFBR fuel tube and plug welded joint was also carried to evaluate base metal grain size, depth of fusion at welded joint and heat affected zone in the base metal. Experimental results demonstrated that pulsed Nd-YAG laser etching is a fast and effortless technique which can be effectively employed for non-contact remote etching of austenitic stainless steels for micro-structural evaluation.

  16. Electrodynamic heat pipe

    Energy Technology Data Exchange (ETDEWEB)

    Shkilev, V.D.

    1982-01-01

    An electrohydrodynamic heat pipe consists of a housing in the form of a closed loop with rising and descending branches, in the first of which are located evaporator, ionizer, which is connected to a high voltage source, a nozzle and a collector of electrical charges. The second branch contains a condenser. In order to improve operating stability, the condenser is equipped with a collector for part of the condensate. It is connected by means of a dielectric tube to a nozzle, and the ionizer of the unit in the outgoing section of the tube is inserted within the nozzle along its access and faces the collector.

  17. Nickel-free austenitic stainless steels for medical applications

    Directory of Open Access Journals (Sweden)

    Ke Yang and Yibin Ren

    2010-01-01

    Full Text Available The adverse effects of nickel ions being released into the human body have prompted the development of high-nitrogen nickel-free austenitic stainless steels for medical applications. Nitrogen not only replaces nickel for austenitic structure stability but also much improves steel properties. Here we review the harmful effects associated with nickel in medical stainless steels, the advantages of nitrogen in stainless steels, and emphatically, the development of high-nitrogen nickel-free stainless steels for medical applications. By combining the benefits of stable austenitic structure, high strength and good plasticity, better corrosion and wear resistances, and superior biocompatibility compared to the currently used 316L stainless steel, the newly developed high-nitrogen nickel-free stainless steel is a reliable substitute for the conventional medical stainless steels.

  18. Propagation of fissures by fatigue in metastable austenitic steels

    International Nuclear Information System (INIS)

    Many works discuss martensitic transformation in austenitic stainless steels, and especially transformations induced by temperature or monotonic charges. Some studies have focused on the propagation of fissures by fatigue in metastable austenitic test pieces, which display reduced propagation speed of fissures when martensite is induced at the end of the fissure. However, controversy still persists with regard to the role of different parameters in the fatigue behavior of these steels. This work presents preliminary analysis results of fissure propagation by fatigue using test pieces obtained from 1 mm thick sheets of austenitic steel EN 1.4318 (AISI 301LN) with 17% Cr, 7% Ni, low C and alloyed with N. The tests were performed at R charge relations (relation between minimum and maximum charge) of 0.1, 0.3, 0.5 and 0.7. The results were analyzed applying the concepts of the two driving forces concept (cw)

  19. Effect of multiple austenitizing treatments on HT-9 steels

    International Nuclear Information System (INIS)

    The effect of multiple austenitizing treatments on the toughness of an Fe-12Cr-1.0Mo-0.5W-0.3V (HT-9) steel was studied. The resulting microstructures were characterized by their mechanical properties, precipitated carbide distribution, and fracture surface appearance. It was proposed that multiple transformations would refine the martensite structure and improve toughness. Optical and scanning electron microscopic observations revealed that the martensite packet structure was somewhat refined by a second austenite transformation. Transmission electron microscopy studies of carbon extraction replicas showed that this multiple step treatment had eliminated grain boundary carbide films seen in single treated specimens on prior austenite grain boundaries. The 0.2% yield strength, tensile strength, and elongation were relatively unchanged, but the toughness measured by fatigue pre-cracked Charpy impact tests increased for the multiple step specimens

  20. Microstructural characterisation of carbon implanted austenitic stainless steel

    International Nuclear Information System (INIS)

    Low carbon (316L) austenitic stainless steel has been implanted with carbon ions with a fluence of 5 x 1017 C ions/cm2 using an ion energy of 75 keV. The effect of carbon ion implantation on the microstructure of the austenitic steel has been examined in cross-section using transmission electron microscopy (TEM) both before and after implantation, and the implantation data correlated with a computer based simulation, TRIM (Transport and Range of Ions in Matter). It has been found that the high-fluence carbon ion implantation modified the microstructure of the steel, as demonstrated by the presence of two amorphous layers separated by a layer of expanded austenite