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Sample records for austenitic piping components

  1. Mechanized ultrasonic inspection of austenitic pipe systems

    International Nuclear Information System (INIS)

    Dressler, K.; Luecking, J.; Medenbach, S.

    1999-01-01

    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) [de

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  3. Integrity of austenitic stainless steel piping welds for nuclear service

    International Nuclear Information System (INIS)

    Canalini, A.; Lopes, L.R.

    1983-01-01

    A criterion applying K 1d concept was developed to determine the fracture mechanics properties of austenitic stainless steel nuclear piping welds. The critical dimensions, lenght and depth, for crack initiation were established and plotted in a chart. This study enables the dimensions of a discontinuity detected in an in-service inspection to be compared to the critical dimensions for crack initiation, and the indication can be judged critical or non-critical for the component. (author) [pt

  4. Investigations on the ratchetting behaviour of austenitic pipes

    International Nuclear Information System (INIS)

    Kraemer, D.; Krolop, S.; Scheffold, A.; Stegmeyer, R.

    1994-01-01

    Reversed bending tests at room temperature with pipes with and without internal pressure were carried out. The pipes were manufactured from the austenitic steel X10 CrNiNb 18 9. Under internal pressure ratchetting was observed in circumferential direction. The component tests were accompanied by numerical computations using a nonlinear kinematic hardening rule and superposed isotropic hardening. In total the constitutive model needed 13 parameters to be fitted when isotropic hardening resulted in a cyclic saturation. Uniaxial monotonic and cyclic loading tests served for characterizing the material. A reasonable parameter fitting with respect to describe ratchetting required load controlled nonzero mean-stress tests. On condition, that the loading will lead to cyclic saturation, ratchetting could be well predicted in the pipe with the found set of parameters. An extension of the isotropic hardening rule in the constitutive model was proposed allowing to describe various types of isotropic hardening. In a first step it was shown that under uniaxial conditions the extension reproduces continuous isotropic hardening up to incipient cracking quite well. (orig.)

  5. Sensitization development in austenitic stainless steel piping

    International Nuclear Information System (INIS)

    Bruemmer, S.M.; Page, R.E.; Atteridge, D.G.

    1984-10-01

    Pacific Northwest Laboratory and the Division of Engineering Technology of the US Nuclear Regulatory Commission are conducting a program to determine a method for evaluating welded and rapair-welded stainless steel piping for light-water reactor service. Validated models, based on experimental data, are being developed to predict the degree of sensitization (DOS) and the intergranular stress corrosion cracking (IGSCC) susceptibility in the heat-affected zone (HAZ) of the SS weldments. The cumulative effects of material composition, past fabrication procedures, past service exposure, weldment thermomechanical (TM) history, and projected post-repair component life are being considered. This program will measure and model the development of HAZ TM history and resultant sensitized microstructure in welded and repair-welded piping. An empirical correlation between a material's DOS and its susceptibility to SCC will be determined using slow strain rate tensile tests. Mill heat chemistries and processing/fabrication records already required in the nuclear industry will be used as input for initial DOS predictions

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

    International Nuclear Information System (INIS)

    Bloom, J.M.

    1995-01-01

    In 1990, the ASME Boiler and Pressure Vessel Code for Nuclear Components approved Code Case N-494 as an alternative procedure for evaluating laws in Light Water Reactor alterative procedure for evaluating flaws in Light Water Reactor (LWR) ferritic piping. The approach is an alternative to Appendix H of the ASME Code and alloys 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 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. A lower bound stress-strain curve was used to generate a PWFAD curve for the geometry of a part-through wall circumferential flaw in a cylinder under tension. Earlier work demonstrated that a cylinder under axial tension with a 50% flaw depth, 90 degrees in circumference, and radius to thickness of 10, produced a lower bound FAD curve. Validation of the new proposed Code Case procedure for austenitic piping was performed using actual pipe test data. Using the lower bound PWFAD curve, pipe test results were conservatively predicted. The resultant development of ht PWFAD curve for austenitic piping led to a revision of Code Case N-494 to include a procedure for assessment of flaws in austenitic piping

  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. Thermal aging evaluation of cast austenitic stainless steel pipe

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  9. Sensitivity of the magnetization curves of different austenitic stainless tube and pipe steels to mechanical fatigue

    International Nuclear Information System (INIS)

    Niffenegger, M.; Leber, H.J.

    2008-01-01

    In meta-stable austenitic stainless steels, fatigue is accompanied by a partial strain-induced transformation of paramagnetic austenite to ferromagnetic martensite [G.B. Olsen, M. Cohen, Kinetics of strain induced martensite nucleation, Metall. Trans. 6 (1975) 791-795]. The associated changes of magnetic properties as the eddy current impedance, magnetic permeability or the remanence field may serve as an indication for the degree of fatigue and therefore the remaining lifetime of a component, even though the exact causal relationship between martensite formation and fatigue is not fully understood. However, measuring these properties by magnetic methods may be limited by the low affinity for strain-induced martensite formation. Thus other methods have to be found which are able to detect very small changes of ferromagnetic contents. With this aim the influence of cyclic strain loading on the magnetization curves of the austenitic stainless tube and pipe steels TP 321, 347, 304L and 316L is analysed in the present paper. The measured characteristic magnetic properties, which are the saturation magnetization, residual magnetization, coercive field and the field dependent permeability (AC-magnetization), are sensitive to fatigue and the corresponding material changes (martensitic transformation). In particular, the AC-magnetization was found to be very sensitive to small changes of the amount of strain induced martensite and therefore also to the degree of fatigue. Hence we conclude that applying magnetic minor loops are promising for the non-destructive evaluation of fatigue in austenitic stainless steel, even if a very small amount of strain induced martensite is formed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

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

  13. Experimental analysis of austenitic stainless steel straight pipes and elbows under pressure and moment loadings

    International Nuclear Information System (INIS)

    Barrou, A.; Prost, J.P.; Delidais, M.

    1983-08-01

    In order to avoid undesirable plastic response in PWR primary system components, tests were performed on 1/10 scale pipes and elbows made from AISI 316 austenitic stainless steel. L/D ratios were from 0.56 to 4.50 mm, arc angles of elbows were 30 0 , 45 0 , 60 0 and 90 0 . Pipes were subjected to bending moments at 3 internal pressure levels. They were tested to determine the mode of failure and served as a reference for elbows. Elbows were subjected to in-plane (closing and opening) and out-of-plane bending moments, at 3 pressure and 2 temperature levels. During these tests, loadings and displacements of components were monitored. Ovalisation of sections was measured regularly. The experimental plastic collapse moment corresponding to excessive deformation was compared to the maximum allowable moment under Design conditions. The experimental plastic instability moment considered as a limit for functional capability was compared to the maximum allowable moment for level C and D service limits

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  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. Inter granular stress corrosion cracking of Ignalina NPP austenitic piping of outside diameter 325 mm

    International Nuclear Information System (INIS)

    Nedzinskas, L.; Klimasauskas, A.

    2003-01-01

    The Inter Granular Stress Corrosion Cracking (IGSCC) of Ignalina NPP main circulation circuit piping, produced from austenitic stainless steel is presented covering current performances and further 'Ageing Management' related actions and plans as well as experience (lessons learned) on solving IGSCC phenomenon, which is currently under investigations and no yet comprehensive answer how to avoid it. (author)

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

    International Nuclear Information System (INIS)

    Kamaya, Masayuki

    2012-01-01

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

  19. Fatigue crack growth in austenitic stainless steel piping

    International Nuclear Information System (INIS)

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

    1981-04-01

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

  20. Equipment for inspection of austenitic stainless steel pipe welds

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  1. Applying Ultrasonic Phased Array Technology to Examine Austenitic Coarse-Grained Structures for Light Water Reactor Piping

    International Nuclear Information System (INIS)

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

    2003-01-01

    Pacific Northwest Laboratory is evaluating the capabilities and limitations of phased array (PA) technology to detect service-type flaws in coarse-grained austenitic piping structures. The work is being sponsored by the U.S. Nuclear Regulatory Commission, Office of Research. This paper presents initial work involving the use of PA technology to determine the effectiveness of detecting and accurately characterizing flaws on the far-side of austenitic piping welds

  2. Structural integrity assessment of piping components

    International Nuclear Information System (INIS)

    Kushwaha, H.S.; Chattopadhyay, J.

    2008-01-01

    Integrity assessment of piping components is very essential for safe and reliable operation of power plants. Over the last several decades, considerable work has been done throughout the world to develop a methodology for integrity assessment of pipes and elbows, appropriate for the material involved. However, there is scope of further development/improvement of issues, particularly for pipe bends, that are important for accurate integrity assessment of piping. Considering this aspect, a comprehensive Component Integrity Test Program was initiated in 1998 at Bhabha Atomic Research Centre (BARC), India. In this program, both theoretical and experimental investigations were undertaken to address various issues related to the integrity assessment of pipes and elbows. Under the experimental investigations, fracture mechanics tests have been conducted on pipes and elbows of 200-400 mm nominal bore (NB) diameter with various crack configurations and sizes under different loading conditions. Tests on small tensile and three point bend specimens, machined from the tested pipes, have also been done to evaluate the actual stress-strain and fracture resistance properties of pipe/elbow material. The load-deflection curve and crack initiation loads predicted by non-linear finite element analysis matched well with the experimental results. The theoretical collapse moments of throughwall circumferentially cracked elbows, predicted by the recently developed equations, are found to be closer to the test data compared to the other existing equations. The role of stress triaxialities ahead of crack tip is also shown in the transferability of J-Resistance curve from specimen to component. The cyclic loading and system compliance effect on the load carrying capacity of piping components are investigated and new recommendations are made. (author)

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

    International Nuclear Information System (INIS)

    Foerster, K.; Grueter, L.; Setz, W.; Bhandari, S.; Debaene, J.P.; Faidy, C.; Schwalbe, K.H.

    1993-01-01

    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

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

    Directory of Open Access Journals (Sweden)

    Mushtaq Albdiry

    2017-11-01

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

  5. Failures of austenitic stainless steel components during storage: Case studies

    International Nuclear Information System (INIS)

    Shah, B.K.; Rastogi, P.K.; Sinha, A.K.; Kulkarni, P.G.

    1993-01-01

    Three studies of failures of austenitic stainless steel components during storage are described. In all cases, stress corrosion cracking was the failure mode by the action of residual stress alone. However, the source of residual stress was different for each case. Case 1 was the failure of a sample tube header for a pressurized heavy water reactor (PHWR). In Case 2, a heat exchanger shell failed during a hydrotest in a fertilizer plant. Cases concerned the cracking of type 304L plates used for spent fuel pool lining of a nuclear power station

  6. Investigations into the ratchetting behaviour of austenitic pipes

    International Nuclear Information System (INIS)

    Kraemer, D.; Krolop, S.; Scheffold, A.; Stegmeyer, R.

    1997-01-01

    In technical components subjected to cyclic loading, inelastic deformations cannot be excluded. In such cases, under certain conditions, small amounts of non-reversed plastic strain per cycle can accumulate to large strains, an effect commonly called ratchetting. The proof of ratchetting in complex structures is often possible by numerical methods only, e.g. the finite-element method. Describing cyclic plasticity and predicting ratchetting necessitate a suitable constitutive law. This paper describes the investigation of the ratchetting behaviour of thin-walled tubes under cyclic loading. Tests were performed and accompanied by finite-element computations using a non-linear kinematic hardening rule with superposed isotropic cyclic hardening. The constitutive law applied used a set of 13 material parameters. This paper discusses the requirements for uniaxial tests which meet the determination of a suitable set of parameters for describing ratchetting. To describe different kinds of isotropic hardening, an extension of the isotropic hardening rule is proposed. Under uniaxial conditions, continuous cyclic hardening is well reproduced with this extension. (orig.)

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

    International Nuclear Information System (INIS)

    Ihara, Ryohei; Hashimoto, Tadafumi; Mochizuki, Masahito

    2012-01-01

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

  8. Proof of fatigue strength of ferritic and austenitic nuclear components

    Energy Technology Data Exchange (ETDEWEB)

    Roos, E.; Herter, K.H.; Schuler, X.; Weissenberg, T. [Materialpruefungsanstalt, Univ. Stuttgart (Germany)

    2009-07-01

    For the construction, design and operation of nuclear components and systems the appropriate technical codes and standards provide material data, detailed stress analysis procedures and a design philosophy which guarantees a reliable behaviour of the structural components throughout the specified lifetime. Especially for cyclic stress evaluation the different codes and standards provide different fatigue analyses procedures to be performed considering the various mechanical and thermal loading histories and geometric complexities of the components. For the fatigue design curves used as limiting criteria the influence of different factors like e.g., environment, surface finish and temperature must be taken into consideration in an appropriate way. Fatigue tests were performed with low alloy steels as well as with Nb- and Ti-stabilized German austenitic stainless steels in air and simulated high temperature boiling water reactor environment. The experimental results are compared and valuated with the mean data curves in air as well as with mean data curves under high temperature water environment published in the international literature. (orig.)

  9. Smart manufacturing of complex shaped pipe components

    Science.gov (United States)

    Salchak, Y. A.; Kotelnikov, A. A.; Sednev, D. A.; Borikov, V. N.

    2018-03-01

    Manufacturing industry is constantly improving. Nowadays the most relevant trend is widespread automation and optimization of the production process. This paper represents a novel approach for smart manufacturing of steel pipe valves. The system includes two main parts: mechanical treatment and quality assurance units. Mechanical treatment is performed by application of the milling machine with implementation of computerized numerical control, whilst the quality assurance unit contains three testing modules for different tasks, such as X-ray testing, optical scanning and ultrasound testing modules. The advances of each of them provide reliable results that contain information about any failures of the technological process, any deviations of geometrical parameters of the valves. The system also allows detecting defects on the surface or in the inner structure of the component.

  10. Progress report on a NDT round robin on austenitic circumferential pipe welds

    International Nuclear Information System (INIS)

    Brast, G.; Maier, H.J.; Knoch, P.; Mletzko, U.

    1998-01-01

    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) [de

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

    International Nuclear Information System (INIS)

    Faure, F.; Leggatt, R.H.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    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)

    1996-07-01

    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.

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

    International Nuclear Information System (INIS)

    Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

    1996-07-01

    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. A countermeasure for external stress corrosion cracking in piping components by means of residual stress improvement on the outer surface

    International Nuclear Information System (INIS)

    Tanaka, Yasuhiro; Umemoto, Tadahiro

    1988-01-01

    Many techniques have been proposed as countermeasures for the External Stress Corrosion Cracking (ESCC) on austenitic stainless steel piping caused by sea salt particles. However, not one seems perfect. The method proposed here is an expansion of IHSI (Induction Heating Stress Improvement) which has been successfully implemented in many nuclear power plants as a remedy for Intergranular Stress Corrossion Cracking. The proposed method named EIHSI (External IHSI) can make the residual stress compressive on the outer surface of the piping components. In order to confirm the effectiveness of EIHSI, one series of tests were conducted on a weld joint between the pipe flange and the straight pipe. The measured residual stresses and also the results of the cracking test revealed that EIHSI is a superior method to suppress the ESCC. The outline of EIHSI and the verification tests are presented in this paper. (author)

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

    International Nuclear Information System (INIS)

    Champeix, L.; Baque, P.; Chairat, C.

    1980-04-01

    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 100 0 C) 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

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

    International Nuclear Information System (INIS)

    Shim, Deog Nam; Jung, In Cheol

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Nyman, R; Erixon, S; Tomic, B; Lydell, B

    1995-12-01

    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.

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

    International Nuclear Information System (INIS)

    Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

    1995-12-01

    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

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

    International Nuclear Information System (INIS)

    Hassan, T.; Matzen, V.C.

    1999-01-01

    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

  2. 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.

  3. Consideration of the environmental effects on fatigue behavior of austenitic components. Calculation methods and practical application

    International Nuclear Information System (INIS)

    Seichter, Johannes; Reese, Sven H.; Klucke, Dietmar

    2012-01-01

    During the last years environmental effects on the fatigue behavior of nuclear power plant components has worldwide been discussed controversial with respect to the transferability of laboratory data on real components. A publication from Argonne National Laboratory on experimental results concerning environmental effects (air and LWR coolant) on fatigue of austenitic steels included a proposal on calculation methods concerning the lifetime reduction due to environmental effects. This calculation method, i.e. multiplication of the usage factor by a F(en), has been included into the ASME Code, Section III, Division I, as Code Case N-792 (fatigue evaluations including environmental effects). The presented contribution evaluates the practical application of this calculation procedure and demonstrates the determination of the usage factor of an austenitic component under environmental exposure.

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

    International Nuclear Information System (INIS)

    Gilies, P.; Franco, C.; Cipiere, M.-F.; Ould, P.

    2005-01-01

    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)

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

    International Nuclear Information System (INIS)

    Lantukh, V.M.; Grebennik, V.S.; Kordinov, E.V.; Kesler, N.A.; Shchedrin, I.F.

    1987-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. Seismic Design of ITER Component Cooling Water System-1 Piping

    Science.gov (United States)

    Singh, Aditya P.; Jadhav, Mahesh; Sharma, Lalit K.; Gupta, Dinesh K.; Patel, Nirav; Ranjan, Rakesh; Gohil, Guman; Patel, Hiren; Dangi, Jinendra; Kumar, Mohit; Kumar, A. G. A.

    2017-04-01

    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.

  8. Corrosion of austenitic steels and their components in vanadium-containing chloride melts

    Science.gov (United States)

    Abramov, A. V.; Polovov, I. B.; Rebrin, O. I.; Lisienko, D. G.

    2014-08-01

    The corrosion of austenitic 12Kh18N10T, 10Kh17N13M2T, and 03Kh17N14M3 steels and their components (Cr, Fe, Ni, Mo) in NaCl-KCl-VCl2 melts with 5 wt % V at 750°C is studied. The rates and mechanisms of corrosion of the materials under these conditions are determined. The processes that occur during contact of the metals and steels with vanadium-containing chloride electrolytes are investigated.

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

    International Nuclear Information System (INIS)

    Cunto, Gabriel Giannini de

    2017-01-01

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

  10. Removal of Shippingport Station primary system components and piping

    International Nuclear Information System (INIS)

    LaGuardia, T.S.; Lipsett, S.M.

    1987-01-01

    The dismantling workscope for the Shippingport Station Decommissioning Project was divided into subtasks to permit the work to be subcontracted to the maximum extent practicable. Major subtasks were identified and described by Activity specifications which could then be grouped into logical work packages to be put out for bid. Two of the largest dismantling work packages, removal of piping and components, were grouped together and designated as Activity Specifications 4 and 5. TLG Services, Inc. and Cleveland Wrecking Company formed a Joint Venture to perform this work during a two-year period at a cost of approximately $7 million. The major portions of this dismantling workscope are described. The primary system components within this workscope consist of the stainless steel reactor coolant piping, check valves, reactor coolant pumps, steam generators, and reactor purification demineralizers and coolers. The work performed, the heavy rigging preparations and procedures, the cutting tools used, component draining/capping techniques to prevent spills, contamination containment, airborne control techniques, and lessons learned during the removal of these primary system components are described. Summaries of crew size and composition, labor hours, duration hours and radiation exposure to workers are provided and discussed briefly. The successful completion of this work is evidence of the engineering, planning, equipment, materials and labor pool available to remove large, radioactively contaminated components safely. This experience will help decommissioning planners to prepare for the removal of reactor components in future decommissioning

  11. Ultrasonic inspection of liquid-metal-filled austenitic stainless steel piping welds

    International Nuclear Information System (INIS)

    Mech, S.J.; Martin, J.D.

    1982-01-01

    The goal of this effort is to reliably detect a crack extending 25 to 50% through the wall of Schedule 40 sodium filled pipe at refueling temperatures (204 0 C [400 0 F]) using remote examination techniques. The task of demonstrating a prototype ultrasonic ISI system under simulated refueling conditions was laid out in two phases. The first phase was initiation of long-lead efforts which were key elements of a practical prototype system, including ultrasonic signal analysis efforts and laboratory prototype support systems. The second phase, dependent on successful completion of the first, consisted of development and demonstration of a prototype system in a simulated ISI environment

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

    International Nuclear Information System (INIS)

    Childs, W.J.

    1986-01-01

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

  13. Investigation of LWR environmental effect on fatigue lifetime of austenitic stainless steel component

    International Nuclear Information System (INIS)

    Kim, J. S.; Youm, H. K.; Jin, T. E.

    1999-01-01

    The fatigue lifetime of principal components in nuclear power plant is evaluated by using the design fatigue curves in ASME B and PV code during design process. However, it is inadequate to evaluate fatigue lifetime considering the LWR environmental effect by these design fatigue curves because these are presented only under atmosphere environment. Therefore, many studies are recently performed for the design fatigue curves considering LWR environmental effect and are presented that the design fatigue curves in ASME B and PV code can be non-conservative. In present paper, the limits and differences of the design fatigue curves considering environmental effect are presented. To investigate the change of fatigue lifetime according to each design fatigue curve, the CUFs for the pressurizer spray nozzle partly composed of austenitic stainless steel are calculated according to each one. Finally, if the evaluation result can not be satisfied with fatigue design requirement, the alternatives to reduce design cumulative usage factor are discussed. (author)

  14. Analysis methods for structure reliability of piping components

    International Nuclear Information System (INIS)

    Schimpfke, T.; Grebner, H.; Sievers, J.

    2004-01-01

    In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour (BMWA) GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The long-term objective of this development is to provide failure probabilities of passive components for probabilistic safety analysis of nuclear power plants. Up to now the code can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents some of the results of a benchmark analysis in the frame of the European project NURBIM (Nuclear Risk Based Inspection Methodologies for Passive Components). (orig.)

  15. FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

    International Nuclear Information System (INIS)

    Stover, R.L.; Beaver, T.R.; Chang, S.C.

    1983-01-01

    The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

  16. Remote-controlled welding during replacement of components and piping

    International Nuclear Information System (INIS)

    Faeser, K.; Huemmeler, A.; Pellkofer, D.

    1986-01-01

    Only on the basis of a thorough fundamental knowledge of nuclear power stations in general and the relevant codes and regulations in particular can extended repair measures, such as the replacement of components or pipelines, be planned and prepared. The application of effective decontamination procedures and shielding measures and a high degree of mechanization of the machining and welding operations will lead to a drastic reduction of the radiation load to which the personnel is exposed. By using highly sophisticated pipe assembling and welding systems the exposure period can be minimized. At the same time a very high level of quality is being reached. The close adherence to the schedule of individual detail operations confirms and justifies the necessity of thorough planning and training of personnel. It may be assumed that in the field of nuclear engineering some pioneer work has been done that will have a stimulating effect on other areas with similar or transferable applications. (orig.) [de

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

    International Nuclear Information System (INIS)

    Wais, E. A.; Rodabaugh, E. C.

    2005-01-01

    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

  18. Volatile organic components migrating from plastic pipes (HDPE, PEX and PVC) into drinking water.

    Science.gov (United States)

    Skjevrak, Ingun; Due, Anne; Gjerstad, Karl Olav; Herikstad, Hallgeir

    2003-04-01

    High-density polyethylene pipes (HDPE), crossbonded polyethylene pipes (PEX) and polyvinyl chloride (PVC) pipes for drinking water were tested with respect to migration of volatile organic components (VOC) to water. The odour of water in contact with plastic pipes was assessed according to the quantitative threshold odour number (TON) concept. A major migrating component from HDPE pipes was 2,4-di-tert-butyl-phenol (2,4-DTBP) which is a known degradation product from antioxidants such as Irgafos 168(R). In addition, a range of esters, aldehydes, ketones, aromatic hydrocarbons and terpenoids were identified as migration products from HDPE pipes. Water in contact with HDPE pipes was assessed with respect to TON, and values > or =4 were determined for five out of seven brands of HDPE pipes. The total amount of VOC released to water during three successive test periods were fairly constant for the HDPE pipes. Corresponding migration tests carried out for PEX pipes showed that VOC migrated in significant amounts into the test water, and TON >/=5 of the test water were observed in all tests. Several of the migrated VOC were not identified. Oxygenates predominated the identified VOC in the test water from PEX pipes. Migration tests of PVC pipes revealed few volatile migrants in the test samples and no significant odour of the test water.

  19. Magnetic Actuator with Multiple Vibration Components Arranged at Eccentric Positions for Use in Complex Piping

    Directory of Open Access Journals (Sweden)

    Hiroyuki Yaguchi

    2016-06-01

    Full Text Available This paper proposes a magnetic actuator using multiple vibration components to perform locomotion in a complex pipe with a 25 mm inner diameter. Due to the desire to increase the turning moment in a T-junction pipe, two vibration components were attached off-center to an acrylic plate with an eccentricity of 2 mm. The experimental results show that the magnetic actuator was able to move at 40.6 mm/s while pulling a load mass of 20 g in a pipe with an inner diameter of 25 mm. In addition, this magnetic actuator was able to move stably in U-junction and T-junction pipes. If a micro-camera is implemented in the future, the inspection of small complex pipes can be enabled. The possibility of inspection in pipes with a 25 mm inner diameter was shown by equipping the pipe with a micro-camera.

  20. Austenitic stainless steels and high strength copper alloys for fusion components

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.; Zinkle, S.J.; Alexander, D.J.; Stubbins, J.F.

    1998-01-01

    An austenitic stainless steel (316LN), an oxide-dispersion-strengthened copper alloy (GlidCop A125), and a precipitation-hardened copper alloy (Cu-Cr-Zr) are the primary structural materials for the ITER first wall/blanket and divertor systems. While there is a long experience of operating 316LN stainless steel in nuclear environments, there is no prior experience with the copper alloys in neutron environments. The ITER first wall (FW) consists of a stainless steel shield with a copper alloy heat sink bonded by hot isostatic pressing (HIP). The introduction of bi-layer structural material represents a new materials engineering challenge; the behavior of the bi-layer is determined by the properties of the individual components and by the nature of the bond interface. The development of the radiation damage microstructure in both classes of materials is summarized and the effects of radiation on deformation and fracture behavior are considered. The initial data on the mechanical testing of bi-layers indicate that the effectiveness of GlidCop A125 as a FW heat sink material is compromised by its strongly anisotropic fracture toughness and poor resistance to crack growth in a direction parallel to the bi-layer interface. (orig.)

  1. Manufacture of piping components for nuclear power plants

    International Nuclear Information System (INIS)

    Bartecek, R.

    1983-01-01

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

  2. Parametric Study on Ultimate Failure Criteria of Elbow Piping Components in Seismically Isolated NPP

    Energy Technology Data Exchange (ETDEWEB)

    Hahm, Dae Gi; Ki, Min Kyu [KAERI, Daejeon (Korea, Republic of); Jeon, Bub Gyu; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of)

    2016-05-15

    It is well known that the interface pipes between isolated and non-isolated structures will become the most critical in the seismically isolated NPPs. Therefore, seismic performance of such interface pipes should be evaluated comprehensively especially in terms of the seismic fragility capacity. To evaluate the seismic capacity of interface pipes in the isolated NPP, firstly, we should define the failure mode and failure criteria of critical pipe components. Hence, in this study, we performed the dynamic tests of elbow components which were installed in a seismically isolated NPP, and evaluated the ultimate failure mode and failure criteria by using the test results. To do this, we manufactured 25 critical elbow component specimens and performed cyclic loading tests under the internal pressure condition. The failure mode and failure criteria of a pipe component will be varied by the design parameters such as the internal pressure, pipe diameter, loading type, and loading amplitude. From the tests, we assessed the effects of the variation parameters onto the failure criteria. For the tests, we generated the seismic input protocol of relative displacement between the ends of elbow component. In this paper, elbow in piping system was defined as a fragile element and numerical model was updated by component test. Failure mode of piping component under seismic load was defined by the dynamic tests of ultimate pipe capacity. For the interface piping system, the seismic capacity should be carefully estimated since that the required displacement absorption capacity will be increased significantly by the adoption of the seismic isolation system. In this study, the dynamic tests were performed for the elbow components which were installed in an actual NPPs, and the ultimate failure mode and failure criteria were also evaluated by using the test results.

  3. Stress-free reference for neutron diffraction measurement of residual stress in butt-welded joints of austenitic stainless steel pipes

    International Nuclear Information System (INIS)

    Maekawa, Akira; Takahashi, Tsuneo; Tsuji, Takashi; Suzuki, Hiroshi; Moriai, Atsushi

    2012-01-01

    Stress-free lattice spacing d_0 has the most influence on reliability of neutron stress measurements made using an angle dispersive method. However, it is hard to evaluate the lattice spacing of welded structures and ductile materials such as stainless steel accurately. In this study, suitable measurement conditions for d_0 of welded pipe joints of austenitic stainless steel were discussed. The d_0 values derived from {311} and {111} reflections, which are often used in austenitic stainless steel for residual stress measurement, were examined. Comparison of the residual strains and stresses evaluated using the obtained d_0 and the finite element analysis showed that the way the d_0 values were chosen affected the measurement accuracy significantly. The stress measurement accuracy was remarkably improved when the {311} reflection was used and the proper d_0 value was chosen in the respective neutron diffraction measurements. For instance, for the axial diffraction measurements using the {311} reflection, it was recommended that only the axial d_0 value of the {311} reflection be used; the measurements using the {111} reflection were less accurate due to the large Young's modulus. Additionally, a lower diffraction angle was judged to be one of the factors leading to a decrease of the strain measurement accuracy. (author)

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

    International Nuclear Information System (INIS)

    Hoefler, A.; Grebner, H.

    1992-01-01

    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

  5. Monitoring of coolant temperature stratification on piping components in WWER-440 NPPs

    International Nuclear Information System (INIS)

    Hudcovsky, S.; Slanina, M.; Badiar, S.

    2001-01-01

    The presentation deals with the aims of non-standard temperature measurements installed on primary and secondary circuit in WWER-440 NPPs, explains reasons of coolant temperature stratification on the piping components. It describes methods of the measurements on pipings, range of installation of the temperature measurements in EBO and EMO units and illustrates results of measurements of coolant temperature stratification. (Authors)

  6. Ductile austenitic steel for fuel cans and core components of sodium cooled reactors

    International Nuclear Information System (INIS)

    Schaefer, L.

    1995-01-01

    Two austenitic steel melts of a new composition have been studied after irradiation in the PFR fast neutron flux, in the BR2 reactor, and in the Harwell V.E. Cyclotron. The investigations were focussed on helium embrittlement and irradiation induced swelling. (orig.)

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

    International Nuclear Information System (INIS)

    Vasil'chenko, G.S.; Dragunov, Yu.G.; Kabelevskij, M.G.; Kazantsev, A.G.; Kunavin, S.A.; Merinov, G.N.; Sokov, L.M.

    2000-01-01

    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 [ru

  8. Development on methods for evaluating structure reliability of piping components

    International Nuclear Information System (INIS)

    Schimpfke, T.; Grebner, H.; Peschke, J.; Sievers, J.

    2003-01-01

    In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour, GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The development is based on the experience achieved with applications of the public available US code PRAISE 3.10 (Piping Reliability Analysis Including Seismic Events), which was supplemented by additional features regarding the statistical evaluation and the crack orientation. PROST is designed to be more flexible to changes and supplementations. Up to now it can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents a parametric study on the influence by changing the method of stress intensity factor and limit load calculation and the statistical evaluation options on the leak probability of an exemplary pipe with postulated axial crack distribution. Furthermore the resulting leak probability of an exemplary pipe with postulated circumferential crack distribution is compared with the results of the modified PRAISE computer program. The intention of this investigation is to show trends. Therefore the resulting absolute values for probabilities should not be considered as realistic evaluations. (author)

  9. Corrosion behavior of austenitic steels and their components in niobium-containing chloride melts

    Science.gov (United States)

    Abramov, A. V.; Polovov, I. B.; Rebrin, O. I.; Volkovich, V. A.; Lisienko, D. G.

    2014-02-01

    The mechanism of corrosion of austenitic steels 12Kh18N10T, 10Kh17N13M2T, and 03Kh17N14M3 and metals Cr, Fe, Ni, and Mo in a NaCl-KCl-NbCl n ( n = 3.5, Nb content is 5 ± 0.1 wt %) melt at 750°C is studied. The metal and steel corrosion rates under these conditions are determined. The character of material fracture and the mechanisms of material corrosion are found.

  10. Sensitiaztion of austenitic stainless steels and its significance as regards stress-corrosion cracking of BWR pipe systems

    International Nuclear Information System (INIS)

    Roberts, W.; Otterberg, R.

    1984-05-01

    A critical literature evaluation dealing with sensitization of austenitic stainless steels and its importance in the context of intergranular stress-corrosion cracking (IGSCC) in high-temperature, oxygenated water is presented. The factors influencing the degree of sensitization are discussed, principally for type-304 stainless steels, both as regards sensitization arising as a result of isothermal holding within the critical temperature range and weld sensitization. The phenomenon of low-temperature sensitization is described and its potential significance under BWR operating conditions speculated upon. The principal features of and mechanisms controlling IGSCC of sensitized 304 steels in BWR-type environments are reviewed and some thoughts are given to the relevance of laboratory SCC testing in predicting the occurrence of cracking in actual BWR systems. Finally various countermeasures against IGSCC in existing and projected reactors are presented and discussed. (Author)

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

    International Nuclear Information System (INIS)

    Sudardjo; Histori; Triyadi, Ari

    1998-01-01

    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

  12. Sensitivity Analysis on Elbow Piping Components in Seismically Isolated NPP under Seismic Loading

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Hee Kun; Hahm, Dae Gi; Kim, Min Kyu [KAERI, Daejeon (Korea, Republic of); Jeon, Bub Gyu; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of)

    2016-05-15

    In this study, the FE model is verified using specimen test results and simulation with parameter variations are conducted. Effective parameters will randomly sampled and used as input values for simulations to be applied to the fragility analysis. pipelines are representative of them because they could undergo larger displacements when they are supported on both isolated and non-isolated structures simultaneously. Especially elbows are critical components of pipes under severed loading conditions such as earthquake action because strain is accumulated on them during the repeated bending of the pipe. Therefore, seismic performance of pipe elbow components should be examined thoroughly based on the fragility analysis. Fragility assessment of interface pipe should take different sources of uncertainty into account. However, selection of important sources and repeated tests with many random input values are very time consuming and expensive, so numerical analysis is commonly used. In the present study, finite element (FE) model of elbow component will be validated using the dynamic test results of elbow components. Using the verified model, sensitivity analysis will be implemented as a preliminary process of seismic fragility of piping system. Several important input parameters are selected and how the uncertainty of them are apportioned to the uncertainty of the elbow response is to be studied. Piping elbows are critical components under cyclic loading conditions as they are subjected large displacement. In a seismically isolated NPP, seismic capacity of piping system should be evaluated with caution. Seismic fragility assessment preliminarily needs parameter sensitivity analysis about the output of interest with different input parameter values.

  13. Pre- and post-calculations for crack opening and leak rate experiments on piping components within the HDR-program

    International Nuclear Information System (INIS)

    Grebner, H.; Hoefler, A.; Hunger, H.

    1991-01-01

    In this paper calculations to experiments on leak opening and leak rates of piping components are presented. The experiments are performed at the HDR-facility at Karlstein/Germany and up to now straight pipes and pipe branches were considered. Numerical and experimental results are compared. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-15

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

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

    International Nuclear Information System (INIS)

    Anton, D.L.; Lemkey, F.D.

    1986-01-01

    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

  16. Gap and impact of LMR [Liquid Metal Reactor] piping systems and reactor components

    International Nuclear Information System (INIS)

    Ma, D.C.; Gvildys, J.; Chang, Y.W.

    1987-01-01

    Because of high operation temperature, the LMR (Liquid Metal Reactor) plant is characterized by the thin-walled piping and components. Gaps are often present to allow free thermal expansion during normal plant operation. Under dynamic loadings, such as seismic excitation, if the relative displacement between the components exceeds the gap distance, impacts will occur. Since the components and piping become brittle over their design lifetime, impact is of important concern for it may lead to fractures of components and other serious effects. This paper deals with gap and impact problems in the LMR reactor components and piping systems. Emphasis is on the impacts due to seismic motion. Eight sections are contained in this paper. The gap and impact problems in LMR piping systems are described and a parametric study is performed on the effects of gap-induced support nonlinearity on the dynamics characteristics of the LMR piping systems. Gap and impact problems in the LMR reactor components are identified and their mathematical models are illustrated, and the gap and impact problems in the seismic reactor scram are discussed. The mathematical treatments of various impact models are also described. The uncertainties in the current seismic impact analyses of LMR components and structures are presented. An impact test on a 1/10-scale LMR thermal liner is described. The test results indicated that several clusters of natural modes can be excited by the impact force. The frequency content of the excited modes depends on the duration of the impact force; the shorter the duration, the higher the frequency content

  17. Statistical properties of material strength for reliability evaluation of components of fast reactors. Austenitic stainless steels

    International Nuclear Information System (INIS)

    Takaya, Shigeru; Sasaki, Naoto; Tomobe, Masato

    2015-03-01

    Many efforts have been made to implement the System Based Code concept of which objective is to optimize margins dispersed in several codes and standards. Failure probability is expected to be a promising quantitative index for optimization of margins, and statistical information for random variables is needed to evaluate failure probability. Material strength like tensile strength is an important random variable, but the statistical information has not been provided enough yet. In this report, statistical properties of material strength such as creep rupture time, steady creep strain rate, yield stress, tensile stress, flow stress, fatigue life and cyclic stress-strain curve, were estimated for SUS304 and 316FR steel, which are typical structural materials for fast reactors. Other austenitic stainless steels like SUS316 were also used for statistical estimation of some material properties such as fatigue life. These materials are registered in the JSME code of design and construction of fast reactors, so test data used for developing the code were used as much as possible in this report. (author)

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

    International Nuclear Information System (INIS)

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

    2006-07-01

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

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

    International Nuclear Information System (INIS)

    1980-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-07-01

    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)

  1. Applications of the TVO piping and component analysis and monitoring system (PAMS)

    Energy Technology Data Exchange (ETDEWEB)

    Smeekes, P. (Teollisuuden Voima Oy, Olkiluoto (Finland)); Kuuluvainen, O. (Rostedt Oy, Luvia (Finland)); Torkkeli, E. (FEMdata Oy, Haukilahti (Finland))

    2010-05-15

    To make fitness, safety and lifetime related assessments for piping and components, the amount of data to be managed is getting larger and larger. At the same time it is essential that the data is reliable, up-to-date, well traceable and easy and fast to obtain. At present the main focus of PAMS is still on piping, but in the future the component related databases and applications will be more and more developed. This paper presents a piping and component database system, consisting of separate geometrical, material, loading, result and document databases as well as current and future applications of the system. By means of a user configurable interface program the user can generate indata files, run application programs and define what data to write back into the result database. The data in the result database can subsequently be used in new input files to perform postprocessing on previous results, for instance fatigue analysis. crack growth analysis or RI-ISI. The system is intended to facilitate the analyses of piping and components and generate well-documented appendices comprising significant parts of the input and output and the associated source references. (orig.)

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

    International Nuclear Information System (INIS)

    Miroshnik, R.; Jeager, A.; Ben Haim, H.

    1998-01-01

    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

  3. Reliability Data Handbook for Piping Components in Nordic Nuclear Power Plants - R Book, Phase 2

    International Nuclear Information System (INIS)

    Hedtjaern Swaling, Vidar; Olsson, Anders

    2011-02-01

    This report presents results of a long research and development project financed by the regulatory body Straalsaekerhetsmyndigheten (SSM) (former SKI), the Swedish nuclear power plant licensees. The report presents a harmonized method for estimating Reliability Data for Piping Components in ASME code class 1 and 2 piping components (R-Book). Data in the R-Book is measured based on 'data driven' strategy. This first version of the R-Book comprises rupture frequencies and failure rates for all systems where ASME Code Class 1 or 2 events could be found in the OECD OPDE database. Nordic and Non-Nordic data are presented separately. Worldwide experience data is used to set up the relevant calculation cases, i.e. intersections of attributes for which there are at least one event present

  4. 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.

  5. How simulation of failure risk can improve structural reliability - application to pressurized components and pipes

    OpenAIRE

    Cioclov, Dimitru Dragos

    2013-01-01

    Probabilistic methods for failure risk assessment are introduced, with reference to load carrying structures, such as pressure vessels (PV) and components of pipes systems. The definition of the failure risk associated with structural integrity is made in the context of the general approach to structural reliability. Sources of risk are summarily outlined with emphasis on variability and uncertainties (V&U) which might be encountered in the analysis. To highlight the problem, in its practical...

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

    International Nuclear Information System (INIS)

    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

  7. 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

  8. Fatigue analysis for analytically overloaded piping components and valves in nuclear power plants

    International Nuclear Information System (INIS)

    Charalambus, B.

    1992-01-01

    Lately, in connection with life extension aspects of power plants, an increasingly accurate determination of the lifetime of components in nuclear stations is being required. In order to assess reliably current fatigue levels in piping systems, variables such as pressure, temperature, and resultant force and moment transients as well as analytical methods which take into account the real operational history must be considered. This paper presents a method for analyzing the transient heat transfer between fluid and pipe wall in order to investigate effects which until now have been assumed conservatively to be caused by a sudden jump in temperature. Further, an example is given showing that the K e factor approach in current design codes for performing simplified elastic-plastic fatigue analyses is conservative. (orig.)

  9. Corrosion of austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Silva, M C.M. [Instituto Nacional de Tecnologia, Rio de Janeiro (Brazil)

    1977-01-01

    Types of corrosion observed in a heat exchanger pipe and on a support of still of molasses fermented wort, both in austenitic stainless steel, are focused. Not only are the causes which might have had any kind of influence on them examined, but also the measures adopted in order to avoid and lessen its occurence.

  10. Feasibility study of the cut and weld operations by RH on the cooling pipes of ITER NB components

    International Nuclear Information System (INIS)

    Pineiro, Oscar; Fernandez, Carlos; Medrano, Mercedes; Liniers, Macarena; Botija, Jose; Alonso, Javier; Sarasola, Xabier; Damiani, Carlo

    2009-01-01

    The maintenance operations of ITER NB components inside the vessel - Beam Line Components (BLC's) involve the removal of the faulty component, its transport to the hot cell as well as the reverse operations of transport of the repaired/new component and its reinstallation inside the vessel. Prior to the removal of the BLC's the cooling pipes must be detached from the component following a procedure that applies to the cutting of the pipes and subsequent welding when the component is re-installed. The purpose of this study, conducted in the framework of EFDA, is to demonstrate the feasibility of the cut and weld operations on the water pipes of the BLC's using fully remote handling techniques. Viable technologies for the cut and weld operations have been identified within the study; in particular the following aspects will be presented in the paper: - Different strategies can be pursued in the detachment of the components depending on the number of cut and weld operations to be performed on the pipes. The selected strategy will impact on the procedure to be followed likewise on important aspects as the requirements of the flexible joints assembled on the pipes. - The existing cutting techniques have been examined in the light of the remotely performed pipe cutting at the NB cell. Modifications of commercial tools have been proposed in order to adapt them to the BLC's pipes requirements. The debris produced during the cutting process must be controlled and collected, therefore a cleaning system has been integrated in the adapted cutting tool referred above. - The existing welding techniques have been also examined and compared based on different criteria such as complexity, reliability, alignment tolerances, etc. TIG welding is the preferred technique as it stands out for its superior performance. The commercial tools identified need to be adapted to the NB environment. - The alignment of the pipes is a critical issue concerning the remote welding. A proper alignment

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

    International Nuclear Information System (INIS)

    Lo Frano, R.; Aquaro, D.; Fontani, E.; Pilo, F.

    2014-01-01

    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

  12. Application of PHADEC method for the decontamination of radioactive steam piping components

    International Nuclear Information System (INIS)

    Lo Frano, R.; Pilo, F.; Aquaro, D.

    2013-01-01

    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 Co 60 (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 components. Moreover the radioactivity in the crud thickness was measured. These values allowed finally to correlate the residence time in the acid attack ponds to the level of the achieved decontamination. (authors)

  13. Investigation of structure in the modular light pipe component for LED automotive lamp

    Science.gov (United States)

    Chen, Hsi-Chao; Zhou, Yang; Huang, Chien-Sheng; Jhong, Wan-Ling; Cheng, Bo-Wei; Jhang, Jhe-Ming

    2014-09-01

    Light-Emitting Diodes (LEDs) have the advantages of small length, long lifetime, fast response time (μs), low voltage, good mechanical properties and environmental protection. Furthermore, LEDs could replace the halogen lamps to avoid the mercury pollution and economize the use of energy. Therefore, the LEDs could instead of the traditional lamp in the future and became an important light source. The proposal of this study was to investigate the effects of the structure and length of the reflector component for a LED automotive lamp. The novel LED automotive lamp was assembled by several different modularization columnar. The optimized design of the different structure and the length to the reflector was simulated by software TracePro. The design result must met the vehicle regulation of United Nations Economic Commission for Europe (UNECE) such as ECE-R19 etc. The structure of the light pipe could be designed by two steps structure. Then constitute the proper structure and choose different power LED to meet the luminous intensity of the vehicle regulation. The simulation result shows the proper structure and length has the best total luminous flux and a high luminous efficiency for the system. Also, the stray light could meet the vehicle regulation of ECE R19. Finally, the experimental result of the selected structure and length of the light pipe could match the simulation result above 80%.

  14. 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.

  15. Feasibility study of the cut and weld operations by RH on the cooling pipes of ITER NB components

    Energy Technology Data Exchange (ETDEWEB)

    Pineiro, Oscar; Fernandez, Carlos [TECNATOM Avda. Montes de Oca 28700 S Sebastian de los Reyes, Madrid (Spain); Medrano, Mercedes [EURATOM-CIEMAT Association for Fusion. Avda. Complutense, 22. 28040 Madrid (Spain)], E-mail: mercedes.medrano@ciemat.es; Liniers, Macarena; Botija, Jose; Alonso, Javier; Sarasola, Xabier [EURATOM-CIEMAT Association for Fusion. Avda. Complutense, 22. 28040 Madrid (Spain); Damiani, Carlo [EFDA-Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2009-06-15

    The maintenance operations of ITER NB components inside the vessel - Beam Line Components (BLC's) involve the removal of the faulty component, its transport to the hot cell as well as the reverse operations of transport of the repaired/new component and its reinstallation inside the vessel. Prior to the removal of the BLC's the cooling pipes must be detached from the component following a procedure that applies to the cutting of the pipes and subsequent welding when the component is re-installed. The purpose of this study, conducted in the framework of EFDA, is to demonstrate the feasibility of the cut and weld operations on the water pipes of the BLC's using fully remote handling techniques. Viable technologies for the cut and weld operations have been identified within the study; in particular the following aspects will be presented in the paper: - Different strategies can be pursued in the detachment of the components depending on the number of cut and weld operations to be performed on the pipes. The selected strategy will impact on the procedure to be followed likewise on important aspects as the requirements of the flexible joints assembled on the pipes. - The existing cutting techniques have been examined in the light of the remotely performed pipe cutting at the NB cell. Modifications of commercial tools have been proposed in order to adapt them to the BLC's pipes requirements. The debris produced during the cutting process must be controlled and collected, therefore a cleaning system has been integrated in the adapted cutting tool referred above. - The existing welding techniques have been also examined and compared based on different criteria such as complexity, reliability, alignment tolerances, etc. TIG welding is the preferred technique as it stands out for its superior performance. The commercial tools identified need to be adapted to the NB environment. - The alignment of the pipes is a critical issue concerning the remote welding

  16. Critical element development of standard components for pipe welding/cutting by CO{sub 2} laser

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Kakudate, Satoshi; Nakahira, Masataka [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1994-11-01

    In D-T burning reactors such as International Thermonuclear Experimental Reactor(ITER), an internal access is inevitable for welding/cutting of cooling pipes of in-vessel components, because of spatial constraint due to a narrow port opening space. An internal-access pipe welding/cutting equipment is being developed in JAERI. Internal access is to approach through inside a pipe to a welding/cutting position, to use 10kW CO{sub 2} laser beam, and to be applicable to both welding and cutting with using a same processing head. A welding/cutting processing head with 10kW CO{sub 2} laser beam has been fabricated and the basic feasibility has been successfully demonstrated for studies of the internal-access pipe welding/cutting concept using 100-A stainless steel pipe with a thickness of 6.3mm. In this study, the optimum focal point of laser beam, laser power and traveling speed of the head have been investigated together with an adjusting mechanism of a relative distance between the head and the pipe wall. In addition, the radiation resistance of critical elements such as optical lens has been investigated. (author).

  17. 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)

  18. Analytical study for frequency effects on the EPRI/USNRC piping component tests. Part 1: Theoretical basis and model development

    International Nuclear Information System (INIS)

    Adams, T.M.; Branch, E.B.; Tagart, S.W. Jr.

    1994-01-01

    As part of the engineering effort for the Advanced Light Water Reactor the Advanced Reactor Corporation formed a Piping Technical Core Group to develop a set of improved ASME Boiler and Pressure Vessel Code, Section III design rules and approaches for ALWR plant piping and support design. The technical basis for the proposed changes to the ASME Boiler and Pressure Vessel Code developed by Technical Core Group for the design of piping relies heavily on the failure margins determined from the EPRI/USNRC piping component test program. The majority of the component tests forming the basis for the reported margins against failure were run with input frequency to natural frequency ratios (Ω/ω) in the range of 0.74 to 0.87. One concern investigated by the Technical Core Group was the effect which could exist on measured margins if the tests had been run at higher or lower frequency ratios than those in the limited frequency ratio range tested. Specifically, the concern investigated was that the proposed Technical Core Group Piping Stress Criteria will allow piping to be designed in the low frequency range (Ω/ω ≥ 2.0) for which there is little test data from the EPRI/USNRC test program. The purpose of this analytical study was to: (1) evaluate the potential for margin variation as a function of the frequency ratio (R ω = Ω/ω, where Ω is the forcing frequency and ω is the natural component frequency), (2) recommend a margin reduction factor (MRF) that could be applied to margins determined from the EPRI/USNRC test program to adjust those margins for potential margin variation with frequency ratio. Presented in this paper is the analytical approach and methodology, which are inelastic analysis, which was the basis of the study. Also, discussed is the development of the analytical model, the procedure used to benchmark the model to actual test results, and the various parameter studies conducted

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

    International Nuclear Information System (INIS)

    Lydell, Bengt; Olsson, Anders

    2008-01-01

    This report constitutes a planning document for a new RandD project to develop a piping component reliability parameter handbook for use in probabilistic safety assessment (PSA) and related activities. The Swedish acronym for this handbook is 'R-Book.' The objective of the project is to utilize the OECD Nuclear Energy Agency 'OECD Pipe Failure Data Exchange Project' (OPDE) database to derive piping component failure rates and rupture probabilities for input to internal flooding probabilistic safety assessment, high-energy line break' (HELB) analysis, risk-informed in-service inspection (RI-ISI) program development, and other activities related to PSA. This new RandD project is funded by member organizations of the Nordic PSA Group (NPSAG) - Forsmark AB, OKG AB, Ringhals AB, and the Swedish Nuclear Power Inspectorate (SKI). The history behind the current effort to produce a handbook of piping reliability parameters goes back to 1994 when SKI funded a 5-year RandD project to explore the viability of establishing an international database on the service experience with piping system components in commercial nuclear power plants. An underlying objective behind this 5-year program was to investigate the different options and possibilities for deriving pipe failure rates and rupture probabilities directly from service experience data as an alternative to probabilistic fracture mechanics. The RandD project culminated in an international piping reliability seminar held in the fall of 1997 in Sigtuna (Sweden) and a pilot project to demonstrate an application of the pipe failure database to the estimation of loss-of-coolant-accident (LOCA) frequency (SKI Report 98:30). A particularly important outcome of the 5-year project was a decision by SKI to transfer the pipe failure database including the lessons learned to an international cooperative effort under the auspices of the OECD Nuclear Energy Agency. Following on information exchange and planning meetings that were

  20. 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

  1. Intergranular corrosion in unserviced austenitic stainless steel pipes made of alloy 904L; Kornzerfall in nicht betriebsbeanspruchten rostfreien austenitischen Rohren aus Alloy 904L

    Energy Technology Data Exchange (ETDEWEB)

    Neidel, Andreas; Cagliyan, Erhan; Fischer, Boromir; Giller, Madeleine; Riesenbeck, Susanne [Siemens AG, Energy Sector, Berlin (Germany). Gasturbinenwerk Berlin

    2017-09-01

    Seamless tubes of the highly corrosion resistant austenitic steel 1.4539, X1NiCrMoCu25-20-5 (Alloy 904L) were observed to exhibit signs of inter-crystalline damage to a depth of several layers of grains and in particular on their internal surface. The material had been stored and had not been put into service. A number of hypotheses had been discussed to explain the predominant cause of the damage. Using optical light and scanning electron microscopy investigation techniques, clear evidence was obtained indicating it to be inter-crystalline corrosion due to the sensitisation of the grain boundaries. The most probable cause of this was determined to be the presence of residual deposits from the rolling process, which due to poor cleaning, had not been completely removed prior to the final solution annealing treatment. This explaining why predominantly the internal surface of the tubes was affected.

  2. Stress corrosion cracking of austenitic stainless steels in PWR primary water: an update of metallurgical investigations performed on French withdrawn components

    International Nuclear Information System (INIS)

    Boursier, J.M.; Gallet, S.; Rouillon, Y.; Bordes, P.

    2002-01-01

    Austenitic stainless steels (AISI 304, 304L, 316 and 316L) are largely used in Nuclear Power Plants because of their good resistance to corrosion and their satisfactory mechanical properties. Nevertheless, on various French PWR Nuclear Power Plants, several cases of corrosion have been encountered in auxiliary circuit portions where deleterious species and oxygen can be present. This paper focuses on the metallurgical investigations performed on pulled out components such as Canopy welds or 'dead legs' (auxiliary circuit portions connected to the main primary loops) in terms of cracking locations and degradation parameters. In addition, some comparisons between Nuclear Power Plant feedback and fundamental research and development studies are discussed, particularly in the scope of temperature, microstructure, stresses (applied and residual) and medium responsible for the degradation. (authors)

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

    International Nuclear Information System (INIS)

    Nyman, R.; Hegedus, D.; Tomic, B.; Lydell, B.

    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

  4. 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.

  5. Validation of the dynamic structural integrity of a nuclear piping component using static inelastic modelling technique

    International Nuclear Information System (INIS)

    Leonard, J.W.

    1975-01-01

    This work is concerned with the evaluation of a quasi-static method as applied to a swing check valve designed to provide emergency shut-off capability subsequent to a postulated break in a steam line. The impact analysis of swinging disk upon the valve seat is an asymmetric problem in dynamic elastoplasticity with potentially large displacements and strains resulting from the impact. To perform a quasi-static analysis for this component the disk and seat region of the valve was isolated from the piping system by special boundary elements and an elastic-plastic finite element model was generated assuming axisymmetric solid ring elements. An equivalent static axisymmetric incremental load system was used to approximate the nonsymmetric initial velocity of impact. Subsequent to the nonlinear incremental finite element analysis by a standard computer software package (MARC-CDC program), a special post-processing program was employed to calculate the incremental sum of external work due to the defined load system. Equating this external work to the initial kinetic energy of impact, parametric curves for displacements, stresses, and strains were obtained as functions of various levels of kinetic energy imparted to the valve at closure. To verify the conservative nature of the assumptions made in the quasi-static model, a comparison was made with a time-dependent, nonlinear, axisymmetric, elastic-plastic finite difference simulation. Another standard computer software package (PISCES-2DL) was used for this dynamic simulation. For a check-point value of initial impact kinetic energy, correlation between the quasi-static finite element and dynamic finite difference analyses is presented. Validations of the assumptions made in the quasi-static analysis and of the results obtained are discussed in detail

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

    International Nuclear Information System (INIS)

    2003-07-01

    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

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

    International Nuclear Information System (INIS)

    Ichikawa, Toshio

    2001-01-01

    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

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

    International Nuclear Information System (INIS)

    Dobre, Tanase; Parvulescu, Oana Cristina; Stoica, Anicuta; Iavorschi, Gustav

    2010-01-01

    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 experimental set-up and a data processing algorithm are shown to describe the cooling of a heat generating electronic device using heat pipes. A Thermalright SI-97 PC cooling system is employed as a case-study to determine the heat transfer characteristics of a fins cooler.

  9. Shippingport Station Decommissioning Project: Removal of piping and equipment and removal of primary system components

    International Nuclear Information System (INIS)

    1989-01-01

    This report is a technical synopsis of the removal of contaminated and non-contaminated piping and equipment from the Shippingport Station Decommissioning Project (SSDP). The information is provided as a part of the Technology Transfer Program to document dismantling activities in support of reactor decommissioning. 5 refs., 29 figs., 4 tabs

  10. 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)

  11. 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

    International Nuclear Information System (INIS)

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

    2004-01-01

    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

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

    International Nuclear Information System (INIS)

    Roos, E.; Herter, K.-H.; Julisch, P.; Otremba, F.; Schuler, X.

    2003-01-01

    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)

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

    International Nuclear Information System (INIS)

    Deppe, L.O.

    1987-01-01

    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) [pt

  14. Pipe connector

    International Nuclear Information System (INIS)

    Sullivan, T.E.; Pardini, J.A.

    1978-01-01

    A safety test facility for testing sodium-cooled nuclear reactor components includes a reactor vessel and a heat exchanger submerged in sodium in the tank. The reactor vessel and heat exchanger are connected by an expansion/deflection pipe coupling comprising a pair of coaxially and slidably engaged tubular elements having radially enlarged opposed end portions of which at least a part is of spherical contour adapted to engage conical sockets in the ends of pipes leading out of the reactor vessel and in to the heat exchanger. A spring surrounding the pipe coupling urges the end portions apart and into engagement with the spherical sockets. Since the pipe coupling is submerged in liquid a limited amount of leakage of sodium from the pipe can be tolerated

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

    International Nuclear Information System (INIS)

    Wintle, J.B.; Hayes, B.; Goldthorpe, M.R.

    2004-01-01

    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.)

  16. Ductile austenitic steel for fuel cans and core components of sodium cooled reactors; Ein duktiler austenitischer Stahl fuer Huellrohre und Kernkomponenten natriumgekuehlter Brueter

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, L.

    1995-08-01

    Two austenitic steel melts of a new composition have been studied after irradiation in the PFR fast neutron flux, in the BR2 reactor, and in the Harwell V.E. Cyclotron. The investigations were focussed on helium embrittlement and irradiation induced swelling. (orig.)

  17. 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

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

    International Nuclear Information System (INIS)

    Wesley, D.A.; Nakaki, D.K.; Hadidi-Tamjed, H.; Kipp, T.R.

    1990-10-01

    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

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

    International Nuclear Information System (INIS)

    Jong, Rudolf Peter de

    2004-01-01

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

  20. Development of a Versatile Ultrasonic Internal Pipe/Vessel Component Monitor for In-Service Inspection of Nuclear Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Searfass, Clifford T. [Structural Integrity Associates, Inc., State College, PA (United States); Malinowski, Owen M. [Structural Integrity Associates, Inc., State College, PA (United States); Van Velsor, Jason K. [Structural Integrity Associates, Inc., State College, PA (United States)

    2015-03-22

    The stated goal of this work was to develop a versatile system which could accurately measure vessel and valve internal vibrations and cavitation formation under in-service conditions in nuclear power plants, ultrasonically. The developed technology will benefit the nuclear power generation industry by allowing plant operators to monitor valve and vessel internals during operation. This will help reduce planned outages and plant component failures. During the course of this work, Structural Integrity Associates, Inc. gathered information from industry experts that target vibration amplitudes to be detected should be in the range of 0.001-in to 0.005-in (0.025-mm to 0.127-mm) and target vibration frequency ranges which should be detected were found to be between 0-Hz and 300-Hz. During the performed work, an ultrasonic measuring system was developed which utilized ultrasonic pulse-echo time-of-flight measurements to measure vibration frequency and amplitude. The developed system has been shown to be able to measure vibration amplitudes as low as 0.0008-in (0.020-mm) with vibration frequencies in the range of 17-Hz to 1000-Hz. Therefore, the developed system was able to meet the industry needs for vibration measurement. The developed ultrasonic system was also to be able to measure cavitation formation by monitoring the received ultrasonic time- and frequency-domain signals. This work also demonstrated the survivability of commercially available probes at temperatures up to 300-F for several weeks.

  1. Ice plugging of pipes using liquid nitrogen

    International Nuclear Information System (INIS)

    Twigg, R.J.

    1987-03-01

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

  2. Implementation of the structural integrity analysis for PWR primary components and piping

    International Nuclear Information System (INIS)

    Pellissier-Tanon, A.

    1982-01-01

    The trends on the definition, the assessment and the application of fracture strength evaluation methodology, which have arisen through experience in the design, construction and operation of French 900-MW plants are reviewed. The main features of the methodology proposed in a draft of Appendix ZG of the RCC-M code of practice for the design verification of fracture strength of primary components are presented. The research programs are surveyed and discussed from four viewpoints, first implementation of the LEFM analysis, secondly implementation of the fatigue crack propagation analysis, thirdly analysis of vessel integrity during emergency core cooling, and fourthly methodology for tear fracture analysis. (author)

  3. Evaluation of welds on a ferritic-austenitic stainless steel

    International Nuclear Information System (INIS)

    Pleva, J.; Johansson, B.

    1984-01-01

    Five different welding methods for the ferritic-austenitic steel 22Cr6Ni3MoN have been evaluated on mill welded heavy wall pipes. The corrosion resistance of the weld joints has been tested both in standard tests and in special environments, related to certain oil and gas wells. The tests were conclusive in that a welding procedure with the addition of sufficient amounts of filler metal should be employed. TIG welds without or with marginal filler addition showed poor resistance to pitting, and to boiling nitric acid. Contents of main alloying elements in ferrite and austenite phases have been measured and causes of corrosion attack in welds are discussed

  4. Plastics piping systems for industrial applications – Acrylonitrile-butadiene-styrene (ABS), unplasticized poly(vinyl chloride) (PVC-U) and chlorinated poly(vinyl chloride) (PVC-C) – Specifications for components and the system – Metric series

    CERN Document Server

    Deutsches Institut für Normung. Berlin

    2003-01-01

    Plastics piping systems for industrial applications – Acrylonitrile-butadiene-styrene (ABS), unplasticized poly(vinyl chloride) (PVC-U) and chlorinated poly(vinyl chloride) (PVC-C) – Specifications for components and the system – Metric series

  5. Plastics piping systems for industrial applications : acrylonitrile-butadiene- styrene (ABS), unplasticized poly(vinyl chloride) (PVC-U) and chlorinated poly(vinyl chloride) (PVC-C) : specifications for components and the system : metric series

    CERN Document Server

    International Organization for Standardization. Geneva

    2003-01-01

    Plastics piping systems for industrial applications : acrylonitrile-butadiene- styrene (ABS), unplasticized poly(vinyl chloride) (PVC-U) and chlorinated poly(vinyl chloride) (PVC-C) : specifications for components and the system : metric series

  6. Inspection of austenitic welds with ultrasonic phased array technology

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, A.; Fernandez, F. [Tecnatom (Spain); Dutruc, R.; Ferriere, R. [Metalscan (France)

    2011-07-01

    This series of slides presents the use of ultrasonic phased array technology in the inspection of austenitic welds. The inspection from outside surface (the inspection is performed in contact using wedges to couple the probe to the outer surface of the component) shows that longitudinal wave is the most adequate for perpendicular scans and transversal ultrasonic wave is the most adequate for parallel scans. Detection and length sizing are performed optimally in perpendicular scans. The inspection from inside surface shows: -) Good results in the detection of defects (Sizing has met the requirements imposed by the Authority of the Russian Federation); -) The new design of the mechanical equipment and of the numerous ultrasonic beams refracted by the array probes has increased the volume inspected. The design of the mechanical equipment has also allowed new areas to be inspected (example a piping weld that was not accessible from the outer surface; -) The ultrasonic procedure and Inspection System developed have been validated by the Authority of the Russian Federation. Phase array technique supplies solutions to solve accessibility concerns and improve the ultrasonic inspections of nuclear components

  7. 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.

  8. 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)

    Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

    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

  9. Investigation on field removed pipe sections in the PISC hot laboratories

    International Nuclear Information System (INIS)

    Cambini, M.; Crutzen, S.; Jehenson, P.

    1990-01-01

    Action no. 1 of PISC III (Programme for the Inspection of Steel Components): Real Contaminated Structures (RCS), seeks to collect results from specific investigations and limited round robin tests on real service induced defects in materials and structures of the primary circuit of Light Water Reactors. The hot cell facilities at JRC-Ispra are fully equipped for non destructive and destructive work on a collaborative basis. Cracked austenitic steel pipes coming from the primary circuit of the Muehleberg reactor (Switzerland) have been inspected in order to demonstrate the validity of the facilities for the examination of these contaminated pieces

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

    International Nuclear Information System (INIS)

    Eghlimi, Abbas; Shamanian, Morteza; Eskandarian, Masoomeh; Zabolian, Azam; Szpunar, Jerzy A.

    2015-01-01

    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. 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

  12. 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

  13. IAEA-NULIFE VERLIFE - Procedure for integrity and lifetime assessment of components and piping in WWER NPPs during operation - Tool for LTO

    International Nuclear Information System (INIS)

    Brumovsky, M.

    2012-01-01

    VERLIFE - 'Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation' was developed within the 5th Framework Programme of the European Union in 2003 and later upgraded within the 6th Framework Programme 'COVERS - Safety of WWER NPPs' of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for WWER type NPPs, as these codes were developed only for design and manufacture and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. To assure that VERLIFE Procedure will remain a living document, new 3-years IAEA project (in close cooperation with another project of the 6th Framework Programme of the European Union 'NULIFE - Plant Life Management of NPPs') has started in 2009. Final document, was approved by expert groups of the IAEA and NULIFE in June 28-30, 2011, and will be issued as 'IAEA/NULIFE Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs during Operation'. This document represents a necessary part for any integrity and lifetime assessment during operation that is a bases for further decision about safe and potential long term operation. To prepare documents like TLAA, it is necessary to have a tool that is able to evaluate lifetime of the main NPP components taking into account existing past operation as well as proposal for the future. (author)

  14. Pipe support

    International Nuclear Information System (INIS)

    Pollono, L.P.

    1979-01-01

    A pipe support for high temperature, thin-walled piping runs such as those used in nuclear systems is described. A section of the pipe to be suppported is encircled by a tubular inner member comprised of two walls with an annular space therebetween. Compacted load-bearing thermal insulation is encapsulated within the annular space, and the inner member is clamped to the pipe by a constant clamping force split-ring clamp. The clamp may be connected to pipe hangers which provide desired support for the pipe

  15. 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)

    Kolkoori, Sanjeevareddy

    2014-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kolkoori, Sanjeevareddy

    2014-07-01

    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

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

    International Nuclear Information System (INIS)

    Casteletti, L.C.; Fernandes, F.A.P.; Heck, S.C.; Gallego, J.

    2010-01-01

    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)

  18. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

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

  19. Ultrasonic testing of austenitic stainless steel welds

    International Nuclear Information System (INIS)

    Nishino, Shunichi; Hida, Yoshio; Yamamoto, Michio; Ando, Tomozumi; Shirai, Tasuku.

    1982-05-01

    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: 55 0 from SUS side, 45 0 from CS side) and in cast stainless steel welds (refraction angle: 45 0 , 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 45 0 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)

  20. The state-of-the Art of the Established Conventional and Advanced NDE-Techniques and Procedures for Defect Detection and Sizing - Part D: application of ultrasonic techniques for the inspection of austenitic components

    International Nuclear Information System (INIS)

    Edelmann, X.

    1997-01-01

    Ultrasonic examination of austenitic welds is possible in many cases but special techniques have often to be applied. The 'Handbook on the Ultrasonic Examination of Austenitic Welds' of the International Institute of Welding gives advice how to devise procedures for the detection, location and evaluation of ultrasonic indications of weld defects. In most circumstances it is necessary to use angled longitudinal wave probes. Austenitic weld examination is more complex and more expensive than ferritic weld examination. Detection and characterization of intergranular stress corrosion cracking has been improved by special qualification programs of the EPRI NDE Center. Dissimilar Metal welds are difficult to examine. Techniques have been developed for detection of circumferential and axial defects. Cast stainless steel structures are in general still extremely difficult to examine as a recent round robin test of the PISC program has shown. Mechanized examination with adequate data acquisition processing and presentation techniques enables better interpretation of examination results. (author)

  1. Austenitic stainless steels with cryogenic resistance

    International Nuclear Information System (INIS)

    Tarata, Daniela Florentina

    1999-01-01

    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

  2. Anelastic mechanical loss spectrometry of hydrogen in austenitic stainless steels

    International Nuclear Information System (INIS)

    Yagodzinskyy, Y.; Andronova, E.; Ivanchenko, M.; Haenninen, H.

    2009-01-01

    Atomic distribution of hydrogen, its elemental diffusion jumps and its interaction with dislocations in a number of austenitic stainless steels are studied with anelastic mechanical loss (AML) spectrometry in combination with the hydrogen thermal desorption method. Austenitic stainless steels of different chemical composition, namely, AISI 310, AISI 201, and AISI 301LN, as well as LDX 2101 duplex stainless steel are studied to clarify the role of different alloying elements on the hydrogen behavior. Activation analyses of the hydrogen Snoek-like peaks are performed with their decomposition to sets of Gaussian components. Fine structure of the composite hydrogen peaks is analyzed under the assumption that each component corresponds to diffusion transfer of hydrogen between octahedral positions with certain atomic compositions of the nearest neighbouring lattice sites. An additional component originating from hydrogen-dislocation interaction is considered. Binding energies for hydrogen-dislocation interaction are also estimated for the studied austenitic stainless steels.

  3. Evaluation of flawed-pipe experiments: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Zahoor, A.; Gamble, R.M.

    1986-11-01

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

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

    International Nuclear Information System (INIS)

    Mann, J.

    2015-01-01

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

  5. Residual stress studies of austenitic and ferritic steels

    International Nuclear Information System (INIS)

    Chrenko, R.M.

    1978-01-01

    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)

  6. Analysis of the influence of the anisotropy induced by cold rolling on duplex and super-austenitic stainless steels

    Directory of Open Access Journals (Sweden)

    Martino Labanti

    2010-07-01

    Full Text Available This report contains the results obtained from the mechanical characterization tests carried out on two different stainless steel (duplex 6%Ni, 22%Cr and super-austenitic 31%Ni, 28%Cr used for the manufacturing of pipes which are employed in the oil production. The activity has been performed in order to evaluate the effects of anisotropy, induced by cold rolling, on the mechanical characteristics of the investigated steels, measured in the three main directions. Considering the small size of the component, the method and the specimens used for the tests were not the standard one. The procedure carried out provided the strain measurement of the specimen during testing by means of resistive strain gages, bonded on the specimens.

  7. Pipe damping

    International Nuclear Information System (INIS)

    Ware, A.G.

    1985-01-01

    Studies are being conducted at the Idaho National Engineering Laboratory to determine whether an increase in the damping values used in seismic structural analyses of nuclear piping systems is justified. Increasing the allowable damping would allow fewer piping supports which could lead to safer, more reliable, and less costly piping systems. Test data from availble literature were examined to determine the important parameters contributing to piping system damping, and each was investigated in separate-effects tests. From the combined results a world pipe damping data bank was established and multiple regression analyses performed to assess the relative contributions of the various parameters. The program is being extended to determine damping applicable to higher frequency (33 to 100 Hz) fluid-induced loadings. The goals of the program are to establish a methodology for predicting piping system damping and to recommend revised guidelines for the damping values to be included in analyses

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

    International Nuclear Information System (INIS)

    2015-01-01

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

  9. Pipe damping

    International Nuclear Information System (INIS)

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

    1984-01-01

    A program has been developed to assess the available piping damping data, to generate additional data and conduct seperate effects tests, and to establish a plan for reporting and storing future test results into a data bank. This effort is providing some of the basis for developing higher allowable damping values for piping seismic analyses, which will potentially permit removal of a considerable number of piping supports, particularly snubbers. This in turn will lead to more flexible piping systems which will be less susceptible to thermal cracking, will be easier to maintain and inspect, as well as less costly

  10. Heat pipe

    International Nuclear Information System (INIS)

    Triggs, G.W.; Lightowlers, R.J.; Robinson, D.; Rice, G.

    1986-01-01

    A heat pipe for use in stabilising a specimen container for irradiation of specimens at substantially constant temperature within a liquid metal cooled fast reactor, comprises an evaporator section, a condenser section, an adiabatic section therebetween, and a gas reservoir, and contains a vapourisable substance such as sodium. The heat pipe further includes a three layer wick structure comprising an outer relatively fine mesh layer, a coarse intermediate layer and a fine mesh inner layer for promoting unimpeded return of condensate to the evaporation section of the heat pipe while enhancing heat transfer with the heat pipe wall and reducing entrainment of the condensate by the upwardly rising vapour. (author)

  11. Fatigue crack growth of 316NG austenitic stainless steel welds at 325 °C

    Science.gov (United States)

    Li, Y. F.; Xiao, J.; Chen, Y.; Zhou, J.; Qiu, S. Y.; Xu, Q.

    2018-02-01

    316NG austenitic stainless steel is a commonly-used material for primary coolant pipes of pressurized water reactor systems. These pipes are usually joined together by automated narrow gap welding process. In this study, welds were prepared by narrow gap welding on 316NG austenitic stainless steel pipes, and its microstructure of the welds was characterized. Then, fatigue crack growth tests were conducted at 325 °C. Precipitates enriched with Mn and Si were found in the fusion zone. The fatigue crack path was out of plane and secondary cracks initiated from the precipitate/matrix interface. A moderate acceleration of crack growth was also observed at 325°Cair and water (DO = ∼10 ppb) with f = 2 Hz.

  12. Austenitic stainless steel weld inspection

    International Nuclear Information System (INIS)

    Mech, S.J.; Emmons, J.S.; Michaels, T.E.

    1978-01-01

    Analytical techniques applied to ultrasonic waveforms obtained from inspection of austenitic stainless steel welds are described. Experimental results obtained from a variety of geometric and defect reflectors are presented. Specifically, frequency analyses parameters, such as simple moments of the power spectrum, cross-correlation techniques, and adaptive learning network analysis, all represent improvements over conventional time domain analysis of ultrasonic waveforms. Results for each of these methods are presented, and the overall inspection difficulties of austenitic stainless steel welds are discussed

  13. Countermeasure technologies against materials deterioration of nuclear power plant components

    International Nuclear Information System (INIS)

    2004-09-01

    This report was tentative safety standard on countermeasure technologies against materials deterioration of nuclear power plant components issued in 2004 on the base of the testing data obtained until March 2004, which was to be applied for technical evaluation for lifetime management of aged plants and preventive maintenance or repair of neutron irradiated components such as core shrouds and jet pumps. In order to prevent stress corrosion cracks (SCCs) of austenitic stainless steel welds of reactor components, thermal surface modification using laser beams was used on neutron irradiated materials with laser cladding or surface melting process methods by limiting heat input according to amount of accumulated helium so as to prevent crack initiation caused by helium bubble growth and coalescence. Laser cladding method of laser welding using molten sleeve set inside pipe surface to prevent SCCs of nickel-chromium-iron alloy welds, alloy 690 cladding method using tungsten inert gas (TIG) welding to prevent SCCs of nickel-chromium-iron alloy welds for dissimilar joints of pipes, and laser surface solid solution heat treatment method of laser irradiation on surfaces to prevent SCCs of austenitic stainless steel welds were also included as repair technologies. (T. Tanaka)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bush, S H; Heasler, P G; Dodge, R E

    1986-02-01

    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)

  15. Replaceable liquid nitrogen piping

    International Nuclear Information System (INIS)

    Yasujima, Yasuo; Sato, Kiyoshi; Sato, Masataka; Hongo, Toshio

    1982-01-01

    This liquid nitrogen piping with total length of about 50 m was made and installed to supply the liquid nitrogen for heat insulating shield to three superconducting magnets for deflection and large super-conducting magnet for detection in the π-meson beam line used for high energy physics experiment in the National Laboratory for High Energy Physics. The points considered in the design and manufacture stages are reported. In order to minimize the consumption of liquid nitrogen during transport, vacuum heat insulation method was adopted. The construction period and cost were reduced by the standardization of the components, the improvement of welding works and the elimination of ineffective works. For simplifying the maintenance, spare parts are always prepared. The construction and the procedure of assembling of the liquid nitrogen piping are described. The piping is of double-walled construction, and its low temperature part was made of SUS 316L. The super-insulation by aluminum vacuum evaporation and active carbon were attached on the external surface of the internal pipe. The final leak test and the heating degassing were performed. The tests on evacuation, transport capacity and heat entry are reported. By making the internal pipe into smaller size, the piping may be more efficient. (Kako, I.)

  16. Reliability analysis of stiff versus flexible piping

    International Nuclear Information System (INIS)

    Lu, S.C.

    1985-01-01

    The overall objective of this research project is to develop a technical basis for flexible piping designs which will improve piping reliability and minimize the use of pipe supports, snubbers, and pipe whip restraints. The current study was conducted to establish the necessary groundwork based on the piping reliability analysis. A confirmatory piping reliability assessment indicated that removing rigid supports and snubbers tends to either improve or affect very little the piping reliability. The authors then investigated a couple of changes to be implemented in Regulatory Guide (RG) 1.61 and RG 1.122 aimed at more flexible piping design. They concluded that these changes substantially reduce calculated piping responses and allow piping redesigns with significant reduction in number of supports and snubbers without violating ASME code requirements. Furthermore, the more flexible piping redesigns are capable of exhibiting reliability levels equal to or higher than the original stiffer design. An investigation of the malfunction of pipe whip restraints confirmed that the malfunction introduced higher thermal stresses and tended to reduce the overall piping reliability. Finally, support and component reliabilities were evaluated based on available fragility data. Results indicated that the support reliability usually exhibits a moderate decrease as the piping flexibility increases. Most on-line pumps and valves showed an insignificant reduction in reliability for a more flexible piping design

  17. Pipe support program at Pickering

    International Nuclear Information System (INIS)

    Sahazizian, L.A.; Jazic, Z.

    1997-01-01

    This paper describes the pipe support program at Pickering. The program addresses the highest priority in operating nuclear generating stations, safety. We present the need: safety, the process: managed and strategic, and the result: assurance of critical piping integrity. In the past, surveillance programs periodically inspected some systems, equipment, and individual components. This comprehensive program is based on a managed process that assesses risk to identify critical piping systems and supports and to develop a strategy for surveillance and maintenance. The strategy addresses all critical piping supports. Successful implementation of the program has provided assurance of critical piping and support integrity and has contributed to decreasing probability of pipe failure, reducing risk to worker and public safety, improving configuration management, and reducing probability of production losses. (author)

  18. Nondestructive testing of austenitic casting and dissimilar metal welds

    International Nuclear Information System (INIS)

    Lahdenperae, K.

    1995-01-01

    The publication is a literature study of nondestructive testing of dissimilar metal welds and cast austenitic components in PWR and BWR plants. A major key to the successful testing is a realistic mockup made of the materials to be tested. The inspectors must also be trained and validated using suitable mockups. (42 refs., 27 figs., 10 tabs.)

  19. Laser cladding crack repair of austenitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2009-06-01

    Full Text Available Laser cladding crack repair of austenitic stainless steel vessels subjected to internal water pressure was evaluated. The purpose of this investigation was to develop process parameters for in-situ repair of through-wall cracks in components...

  20. Development of new damping devices for piping

    International Nuclear Information System (INIS)

    Kobayashi, Hiroe

    1991-01-01

    An increase of the damping ratio is known to be very effective for the seismic design of a piping system. Increasing the damping ratio and reducing the seismic response of the piping system, the following three types of damping devices for piping systems are introduced: (1) visco-elastic damper, (2) elasto-plastic damper and (3) compact dynamic damper. The dynamic characteristics of these damping devices were investigated by the component test and the applicability of them to the piping system was confirmed by the vibration test using a three dimensional piping model. These damping devices are more effective than mechanical snubbers to reduce the vibration of the piping system. (author)

  1. Parametric calculations of fatigue-crack growth in piping

    International Nuclear Information System (INIS)

    Simonen, F.A.; Goodrich, C.W.

    1983-06-01

    This study presents calculations of the growth of piping flaws produced by fatigue. Flaw growth was predicted as a function of the initial flaw size, the level and number of stress cycles, the piping material, and environmental factors. The results indicate that the present flaw acceptance standards of ASME Section XI provide a relatively consistent set of allowable flaw sizes because the predicted life of flawed piping is relatively insensitive to pipe wall thickness, flaw aspect ratio, and piping material (ferritic versus austenitic). On the other hand, the results show that flaws that are acceptable under ASME Section XI can grow at unacceptable rates if the cyclic stresses are at the maximum level permitted by the design rules of ASME Section III. However, a review of the conservatisms inherent to the ASME code rules is presented to explain the low occurrence of piping fatigue failures in service. It is concluded that decreases in the allowable flaw sizes are not justified

  2. 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

  3. Austenitic Reversion of Cryo-rolled Ti-Stabilized Austenitic Stainless Steel: High-Resolution EBSD Investigation

    Science.gov (United States)

    Tiamiyu, A. A.; Odeshi, A. G.; Szpunar, J. A.

    2018-02-01

    In this study, AISI 321 austenitic stainless steel (ASS) was cryo-rolled and subsequently annealed at 650 and 800 °C to reverse BCC α'-martensite to FCC γ-austenite. The texture evolution associated with the reversion at the selected temperatures was investigated using high-resolution EBSD. After the reversion, TiC precipitates were observed to be more stable in 650 °C-annealed specimens than those reversed at 800 °C. {110} texture was mainly developed in specimens subjected to both annealing temperatures. However, specimens reversed at 650 °C have stronger texture than those annealed at 800 °C, even at the higher annealing time. The strong intensity of {110} texture component is attributed to the ability of AISI 321 ASS to memorize the crystallographic orientation of the deformed austenite, a phenomenon termed texture memory. The development of weaker texture in 800 °C-annealed specimens is attributed to the residual strain relief in grains, dissolution of grain boundary precipitates, and an increase in atomic migration along the grain boundaries. Based on the observed features of the reversed austenite grains and estimation from an existing model, it is suspected that the austenite reversion at 650 and 800 °C undergone diffusional and martensitic shear reversion, respectively.

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  5. Review: heat pipe heat exchangers at IROST

    OpenAIRE

    E. Azad

    2012-01-01

    The use of the heat pipe as a component in a heat recovery device has gained worldwide acceptance. Heat pipes are passive, highly reliable and offer high heat transfer rates. This study summarizes the investigation of different types of heat pipe heat recovery systems (HPHRSs). The studies are classified on the basis of the type of the HPHRS. This research is based on 30 years of experience on heat pipe and heat recovery systems that are presented in this study. Copyright , Oxford University ...

  6. IPM Pipe

    Science.gov (United States)

    Submit A Report View Reports List [+] View Reports Map [+] CDM Alert System Sign Up For Alerts User Login Annual Epidemic Histories Annual Season Summaries Contact Us ipmPIPE User Login Web Administrator Login

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

    Energy Technology Data Exchange (ETDEWEB)

    Blondé, R., E-mail: r.j.p.blonde@tudelft.nl [Fundamental Aspects of Materials and Energy, Faculty of Applied Sciences, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Materials Innovation Institute, Mekelweg 2, 2628 CD Delft (Netherlands); Jimenez-Melero, E., E-mail: enrique.jimenez-melero@manchester.ac.uk [Dalton Cumbrian Facility, The University of Manchester, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HA (United Kingdom); Zhao, L., E-mail: lie.zhao@tudelft.nl [Materials Innovation Institute, Mekelweg 2, 2628 CD Delft (Netherlands); Department of Materials Science and Engineering, Delft University of Technology, Mekelweg 2, 2628 CD Delft (Netherlands); Schell, N., E-mail: norbert.schell@hzg.de [Institute of Materials Research, Helmholtz-Zentrum Geesthacht, Max Planck Strasse 1, 21502 Geesthacht (Germany); Brück, E., E-mail: e.h.bruck@tudelft.nl [Fundamental Aspects of Materials and Energy, Faculty of Applied Sciences, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Zwaag, S. van der, E-mail: s.vanderzwaag@tudelft.nl [Novel Aerospace Materials Group, Faculty of Aerospace Engineering, Delft University of Technology, Kluyverweg 1, 2629 HS Delft (Netherlands); Dijk, N.H. van, E-mail: n.h.vandijk@tudelft.nl [Fundamental Aspects of Materials and Energy, Faculty of Applied Sciences, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-01-31

    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.

  8. Pipe grabber

    Energy Technology Data Exchange (ETDEWEB)

    Sharafutdinov, I.G.; Mubashirov, S.G.; Prokopov, O.I.

    1981-05-15

    A pipe grabber is suggested which contains a housing, clamping elements and centering mechanism with drive installed on the lower end of the housing. In order to improve the reliable operation of the pipe grabber, the centering mechanism is made in the form of a reinforced ringed flexible shaft, while the drive is made in the form of elastic rotating discs. In this case the direction of rotation of the discs and the flexible shaft is the opposite.

  9. Low cycle fatigue strength of austenitic stainless steel under large strain regime

    International Nuclear Information System (INIS)

    Sakai, Michiya; Saito, Kiyoshi; Matsuura, Shinichi

    1998-01-01

    In order to establish realistic seismic safety of nuclear power plants, it is necessary to clarify the failure mode of each components and prepare a damage evaluation method. The authors have proposed the damage evaluation method based on the fully numerical approach to evaluate the low cycle fatigue (LCF) failure under seismic loadings. This method has been validated by comparison with the dynamic failure tests of thin elbows which should be the one of the important components of the FBR primary piping system. However, since there exists limited LCF data, fatigue lives under large strain regime have been extrapolated by available fatigue data. In this study, LCF tests have been conducted over a large strain range from 2% to 10% on austenitic stainless steel SUS304. From the results, the regressive LCF curve has been proposed to modify the Wada's best-fit LCF curve under large strain regime. The usage factors calculated by author's numerical approach using proposed LCF curve have been improved to correct the underestimation of the fatigue damage. (author)

  10. Quantification by image analysis of grain size of the high temperature phase (austenite) of martensitic steels 9Cr-1Mo

    International Nuclear Information System (INIS)

    Barcelo, F.; Brachet, J.C.

    1993-01-01

    In martensitic steels, the austenitic grain size before transformation may influence mechanical properties. 9Cr-1Mo steel (EM10) is used in hexagonal pipes fabrication in fast neutrons reactors. Image analysis allows to quantify the older grain size in function of the austenization heat treatment conditions. (A.B.). 2 figs

  11. Oxygen isotope and trace element compositions of platiniferous dunite pipes of the Bushveld Complex, South Africa - Signals from a recycled mantle component?

    Science.gov (United States)

    Günther, T.; Haase, K. M.; Junge, M.; Oberthür, T.; Woelki, D.; Krumm, S.

    2018-06-01

    Platiniferous dunite pipes occur in the lower mafic/ultramafic portion of the Rustenburg Layered Suite of the Bushveld large igneous province (LIP). Olivine compositions in these pipes range from forsterite (Fo) 80 to 35 mol% and suggest crystallization from variably evolved magmas at high temperatures ( 1200 °C). The most primitive olivines are from a stock unit and have the highest contents of Ni (>0.15 wt%) and lowest contents of Mn (formation of the fayalitic olivines with its high Mn contents (>0.3 wt%). High δ18O values of olivine (5.7-7.0‰) and pyroxene (6.7-7.4‰) are akin to those of the Lower and Critical Zone of the Bushveld intrusion suggesting a common origin. The constant high O isotope ratios with variable Fo contents in the olivines are unlike trends observed in olivine phenocrysts in magmas forming by assimilation-fractional crystallization. We suggest that the high δ18O in the most primitive dunites reflect that of the primary melt of the Bushveld pipes, indicating either a bulk assimilation of crust prior to pipe formation or a contribution from recycled oceanic crust in the sub-continental lithospheric mantle (SCLM). The latter scenario is supported by the high Ni/Mn ratios in primitive pipe olivine that might be inherited from melting of a pyroxene-rich mantle source.

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

    International Nuclear Information System (INIS)

    1976-01-01

    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)

  13. Heat Pipes

    Science.gov (United States)

    1990-01-01

    Bobs Candies, Inc. produces some 24 million pounds of candy a year, much of it 'Christmas candy.' To meet Christmas demand, it must produce year-round. Thousands of cases of candy must be stored a good part of the year in two huge warehouses. The candy is very sensitive to temperature. The warehouses must be maintained at temperatures of 78-80 degrees Fahrenheit with relative humidities of 38- 42 percent. Such precise climate control of enormous buildings can be very expensive. In 1985, energy costs for the single warehouse ran to more than 57,000 for the year. NASA and the Florida Solar Energy Center (FSEC) were adapting heat pipe technology to control humidity in building environments. The heat pipes handle the jobs of precooling and reheating without using energy. The company contacted a FSEC systems engineer and from that contact eventually emerged a cooperative test project to install a heat pipe system at Bobs' warehouses, operate it for a period of time to determine accurately the cost benefits, and gather data applicable to development of future heat pipe systems. Installation was completed in mid-1987 and data collection is still in progress. In 1989, total energy cost for two warehouses, with the heat pipes complementing the air conditioning system was 28,706, and that figures out to a cost reduction.

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

    International Nuclear Information System (INIS)

    Haftirman; Maruhum Tua Lubis

    2009-01-01

    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

  15. Functional capability of piping systems

    International Nuclear Information System (INIS)

    Terao, D.; Rodabaugh, E.C.

    1992-11-01

    General Design Criterion I of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations requires, in part, that structures, systems, and components important to safety be designed to withstand the effects of earthquakes without a loss of capability to perform their safety function. ne function of a piping system is to convey fluids from one location to another. The functional capability of a piping system might be lost if, for example, the cross-sectional flow area of the pipe were deformed to such an extent that the required flow through the pipe would be restricted. The objective of this report is to examine the present rules in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, and potential changes to these rules, to determine if they are adequate for ensuring the functional capability of safety-related piping systems in nuclear power plants

  16. Finite-element analysis of flawed and unflawed pipe tests

    International Nuclear Information System (INIS)

    James, R.J.; Nickell, R.E.; Sullaway, M.F.

    1989-12-01

    Contemporary versions of the general purpose, nonlinear finite element program ABAQUS have been used in structural response verification exercises on flawed and unflawed austenitic stainless steel and ferritic steel piping. Among the topics examined, through comparison between ABAQUS calculations and test results, were: (1) the effect of using variations in the stress-strain relationship from the test article material on the calculated response; (2) the convergence properties of various finite element representations of the pipe geometry, using shell, beam and continuum models; (3) the effect of test system compliance; and (4) the validity of ABAQUS J-integral routines for flawed pipe evaluations. The study was culminated by the development and demonstration of a ''macroelement'' representation for the flawed pipe section. The macroelement can be inserted into an existing piping system model, in order to accurately treat the crack-opening and crack-closing static and dynamic response. 11 refs., 20 figs., 1 tab

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

    Directory of Open Access Journals (Sweden)

    Chang-Gi Han

    2016-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

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

    International Nuclear Information System (INIS)

    Seeberger, E.K.

    2000-01-01

    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/mm 2 ) 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)

  20. Modification of the Strength Anisotropy in an Austenitic ODS Steel

    International Nuclear Information System (INIS)

    Kim, T. K.; Jang, J.; Kim, S. H.; Lee, C. B.; Bae, C. S.; Kim, D. H.

    2007-01-01

    Among many candidate alloys for Gen IV reactors, the oxide dispersion strengthened (ODS) alloy is widely considered as a good candidate material for the in-reactor component, like cladding tube. The ODS alloy is well known due to its good high temperature strength, and excellent irradiation resistance. For the previous two decades in the nuclear community, the ODS alloy developments have been mostly focused on the ferritic martensitic (F-M) steel-based ones. On the other hand, the austenitic stainless steels (e.g. 316L or 316LN) have been used as a structural material due to its good high temperature strength and a good compatibility with a media. However, the austenitic stainless steel showed unfavorable characteristics in the dimensional stability under neutron irradiation and cracking behavior with the media. It is thus expected that the austenitic ODS steels restrain the dimension stability under neutron irradiation. However, the ODS alloys usually reveal the anisotropic characteristic in mechanical strength in the hoop and longitudinal directions, which is attributed to the grain morphology strongly developed parallel to the rolling direction with a high aspect ratio. This study focuses on a modification of the strength anisotropy of an austenitic ODS alloy by a recrystallization heat treatment

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

    International Nuclear Information System (INIS)

    Kim, In Soo; Kim, Sung Jin

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Natesan, K.; Chopra, D.K.; Zeman, G.J.; Smith, D.L.; Kassner, T.F.

    1977-01-01

    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)

  3. Ultrasonic inspection of austenitic welds

    Energy Technology Data Exchange (ETDEWEB)

    Tomlinson, J R; Wagg, A R; Whittle, M J [N.D.T. Applications Centre, CEGB, Manchester (United Kingdom)

    1980-11-01

    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)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  5. Microstructural characterization of primary coolant pipe steel

    International Nuclear Information System (INIS)

    Miller, M.K.; Bentley, J.

    1986-01-01

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

  6. Development of bore tools for pipe welding and cutting

    International Nuclear Information System (INIS)

    Oka, Kiyoshi; Ito, Akira; Takiguchi, Yuji

    1998-01-01

    In the International Thermonuclear Experimental Reactor (ITER), in-vessel components replacement and maintenance requires that connected cooling pipes be cut and removed beforehand and that new components be installed to which cooling pipes must be rewelded. All welding must be inspected for soundness after completion. These tasks require a new task concept for ensuring shielded areas and access from narrow ports. Thus, it became necessary to develop autonomous locomotion welding and cutting tools for branch and main pipes to weld pipes by in-pipe access; a system was proposed that cut and welded branch and main pipes after passing inside pipe curves, and elemental technologies developed. This paper introduces current development in tools for welding and cutting branch pipes and other tools for welding and cutting the main pipe. (author)

  7. Development of bore tools for pipe welding and cutting

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Ito, Akira; Takiguchi, Yuji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    In the International Thermonuclear Experimental Reactor (ITER), in-vessel components replacement and maintenance requires that connected cooling pipes be cut and removed beforehand and that new components be installed to which cooling pipes must be rewelded. All welding must be inspected for soundness after completion. These tasks require a new task concept for ensuring shielded areas and access from narrow ports. Thus, it became necessary to develop autonomous locomotion welding and cutting tools for branch and main pipes to weld pipes by in-pipe access; a system was proposed that cut and welded branch and main pipes after passing inside pipe curves, and elemental technologies developed. This paper introduces current development in tools for welding and cutting branch pipes and other tools for welding and cutting the main pipe. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-01-01

    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

  10. 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.

  11. 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

  12. 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 ...

  13. 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...

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  15. 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...

  16. 75 FR 973 - Certain Welded Stainless Steel Pipes From the Republic of Korea: Preliminary Results of...

    Science.gov (United States)

    2010-01-07

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

  17. Flexibility of trunnion piping elbows

    International Nuclear Information System (INIS)

    Lewis, G.D.; Chao, Y.J.

    1987-01-01

    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)

  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. Resolution of concerns in auxiliary feedwater piping

    International Nuclear Information System (INIS)

    Bain, R.A.; Testa, M.F.

    1994-01-01

    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

  20. Development of bore tools for pipe inspection

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Nakahira, Masataka; Taguchi, Kou; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    In the International Thermonuclear Reactor (ITER), replacement and maintenance on in-vessel components requires that all cooling pipes connected be cut and removed, that a new component be installed, and that all cooling pipes be rewelded. After welding is completed, welded area must be inspected for soundness. These tasks require a new work concept for securing shielded area and access from narrow ports. Tools had to be developed for nondestructive inspection and leak testing to evaluate pipe welding soundness by accessing areas from inside pipes using autonomous locomotion welding and cutting tools. A system was proposed for nondestructive inspection of branch pipes and the main pipe after passing through pipe curves, the same as for welding and cutting tool development. Nondestructive inspection and leak testing sensors were developed and the basic parameters were obtained. In addition, the inspection systems which can move inside pipes and conduct the nondestructive inspection and the leak testing were developed. In this paper, an introduction will be given to the current situation concerning the development of nondestructive inspection and leak testing machines for the branch pipes. (author)

  1. Asymptotic scalings of developing curved pipe flow

    Science.gov (United States)

    Ault, Jesse; Chen, Kevin; Stone, Howard

    2015-11-01

    Asymptotic velocity and pressure scalings are identified for the developing curved pipe flow problem in the limit of small pipe curvature and high Reynolds numbers. The continuity and Navier-Stokes equations in toroidal coordinates are linearized about Dean's analytical curved pipe flow solution (Dean 1927). Applying appropriate scaling arguments to the perturbation pressure and velocity components and taking the limits of small curvature and large Reynolds number yields a set of governing equations and boundary conditions for the perturbations, independent of any Reynolds number and pipe curvature dependence. Direct numerical simulations are used to confirm these scaling arguments. Fully developed straight pipe flow is simulated entering a curved pipe section for a range of Reynolds numbers and pipe-to-curvature radius ratios. The maximum values of the axial and secondary velocity perturbation components along with the maximum value of the pressure perturbation are plotted along the curved pipe section. The results collapse when the scaling arguments are applied. The numerically solved decay of the velocity perturbation is also used to determine the entrance/development lengths for the curved pipe flows, which are shown to scale linearly with the Reynolds number.

  2. 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).

  3. Plastics pipe couplings

    International Nuclear Information System (INIS)

    Glover, J.B.

    1980-07-01

    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.)

  4. Enhancing the capabilities of eddy current techniques for non-destructive evaluation of austenitic stainless steels

    International Nuclear Information System (INIS)

    Rao, B.P.C.; Thirunavukkarasu, S.; Sasi, B.; Jayakumar, T.; Baldev Raj

    2010-01-01

    Eddy current non-destructive evaluation (NDE) techniques find many applications during fabrication and in-service inspection of components made of stainless steel. In recent years, concurrent developments in electromagnetic field detection sensors such as giant magneto-resistive (GMR), giant-magneto impedance (GMI) and SQUIDs sensors, computers, microelectronics, and incorporating advanced signal and image processing techniques, have paved the way for enhancing the capabilities of existing eddy current (EC) techniques for examination of austenitic stainless steel (SS) plates, tubes and other geometries and several innovative methodologies have been developed. This paper highlights a few such applications in EC testing to austenitic stainless steel components used in fast reactors. (author)

  5. 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... Components § 192.193 Valve installation in plastic pipe. Each valve installed in plastic pipe must be designed so as to protect the plastic material against excessive torsional or shearing loads when the valve...

  6. Nondestructive characterization of austenitic stainless steels

    International Nuclear Information System (INIS)

    Jayakumar, T.; Kumar, Anish

    2010-01-01

    The paper presents an overview of the non-destructive methodologies developed at the authors' laboratory for characterization of various microstructural features, residual stresses and corrosion in austenitic stainless steels. Various non-destructive evaluation (NDE) parameters such as ultrasonic velocity, ultrasonic attenuation, spectral analysis of the ultrasonic signals, magnetic hysteresis parameters and eddy current amplitude have been used for characterization of grain size, precipitation behaviour, texture, recrystallization, thermomechanical processing, degree of sensitization, formation of martensite from metastable austenite, assessment of residual stresses, degree of sensitization and propensity for intergranular corrosion in different austenitic steels. (author)

  7. Ultrasonic inspection of austenitic welds

    International Nuclear Information System (INIS)

    Baikie, B.L.; Wagg, A.R.; Whittle, M.J.; Yapp, D.

    1976-01-01

    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

  8. 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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  10. Parametrical limits of SCC-susceptibility of austenitic and austenitic-ferritic Cr-Ni steels

    International Nuclear Information System (INIS)

    Starosvetskij, D.I.; Baru, R.L.; Bondarenko, A.I.; Bogoyavlenskij, V.L.; Timonin, V.A.

    1990-01-01

    Comparative investigations into corrosion cracking (CC) of austenitic (12Kh18N10T) and austenitic-ferritic (08Kh22N6T) chromium-nickel steels are performed for various chloride media in a wide range of chloride concentrations and temperatures. It is shown that the ratio between steels in terms of their CC-susceptibility is not definite and can undergo a reversal depending on parameters of medium, level and conditions of loading. Differences in mechanisms of corrosion cracking of austenitic and austenitic-ferritic steels are established

  11. Insulated pipe clamp design

    International Nuclear Information System (INIS)

    Anderson, M.J.; Hyde, L.L.; Wagner, S.E.; Severud, L.K.

    1980-01-01

    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. 5 refs

  12. Introduction to Loop Heat Pipes

    Science.gov (United States)

    Ku, Jentung

    2015-01-01

    This is the presentation file for the short course Introduction to Loop Heat Pipes, to be conducted at the 2015 Thermal Fluids and Analysis Workshop, August 3-7, 2015, Silver Spring, Maryland. This course will discuss operating principles and performance characteristics of a loop heat pipe. Topics include: 1) pressure profiles in the loop; 2) loop operating temperature; 3) operating temperature control; 4) loop startup; 4) loop shutdown; 5) loop transient behaviors; 6) sizing of loop components and determination of fluid inventory; 7) analytical modeling; 8) examples of flight applications; and 9) recent LHP developments.

  13. 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

  14. Constitutive modeling of metastable austenitic stainless steel

    NARCIS (Netherlands)

    Perdahcioglu, Emin Semih; Geijselaers, Hubertus J.M.; Huetink, Han; Khan, A.

    2010-01-01

    A physically based, macroscale constitutive model has been developed that can describe the complex mechanical behavior of metastable austenitic stainless steels. In the developed model a generalized model for the mechanically induced martensitic transformation is introduced. Mechanical tests have

  15. Consitutive modeling of metastable austenitic stainless steel

    NARCIS (Netherlands)

    Perdahcioglu, Emin Semih; Perdahcioglu, Emin Semih

    2008-01-01

    Metastable austenitic stainless steels combine high formability and high strength, which are generally opposing properties in materials. This property is a consequence of the martensitic phase transformation that takes place during deformation. This transformation is purely mechanically induced

  16. Heat pipes in modern heat exchangers

    International Nuclear Information System (INIS)

    Vasiliev, Leonard L.

    2005-01-01

    Heat pipes are very flexible systems with regard to effective thermal control. They can easily be implemented as heat exchangers inside sorption and vapour-compression heat pumps, refrigerators and other types of heat transfer devices. Their heat transfer coefficient in the evaporator and condenser zones is 10 3 -10 5 W/m 2 K, heat pipe thermal resistance is 0.01-0.03 K/W, therefore leading to smaller area and mass of heat exchangers. Miniature and micro heat pipes are welcomed for electronic components cooling and space two-phase thermal control systems. Loop heat pipes, pulsating heat pipes and sorption heat pipes are the novelty for modern heat exchangers. Heat pipe air preheaters are used in thermal power plants to preheat the secondary-primary air required for combustion of fuel in the boiler using the energy available in exhaust gases. Heat pipe solar collectors are promising for domestic use. This paper reviews mainly heat pipe developments in the Former Soviet Union Countries. Some new results obtained in USA and Europe are also included

  17. Insulated pipe clamp design

    International Nuclear Information System (INIS)

    Anderson, M.J.; Hyde, L.L.; Wagner, S.E.; Severud, L.K.

    1980-01-01

    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

  18. New assessment of feed water piping in GKN I including optimisation of piping supports

    International Nuclear Information System (INIS)

    Zaiss, W.; Heil, C.; Baier, B.; Manke, A.

    2003-01-01

    The quality of nuclear power plant components and piping is specified according to the then current state of knowledge. In operation, the quality can be reduced by ageing phenomena, so in-service quality assessment is constantly required. The contribution discusses the individual aspects of reassessment and its technical procedure, using the example of a feedwater pipe in the GKN I containment. (orig.) [de

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

    International Nuclear Information System (INIS)

    Suzuki, K.; Namita, Y.; Abe, H.; Ichihashi, I.; Suzuki, K.; Ishiwata, M.; Fujiwaka, T.; Yokota, H.

    2001-01-01

    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. Application of Leak Before Break concept in 316LN austenitic steel pipes welded using 316L; Aplicação do conceito 1vazamento antes da falha' (Leak Before Break) em tubulações de aço 316LN soldado com metal de adição 316L

    Energy Technology Data Exchange (ETDEWEB)

    Cunto, Gabriel Giannini de

    2017-07-01

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

  1. 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...

  2. Pipe drafting and design

    CERN Document Server

    Parisher, Roy A; Parisher

    2000-01-01

    Pipe designers and drafters provide thousands of piping drawings used in the layout of industrial and other facilities. The layouts must comply with safety codes, government standards, client specifications, budget, and start-up date. Pipe Drafting and Design, Second Edition provides step-by-step instructions to walk pipe designers and drafters and students in Engineering Design Graphics and Engineering Technology through the creation of piping arrangement and isometric drawings using symbols for fittings, flanges, valves, and mechanical equipment. The book is appropriate primarily for pipe

  3. High temperature heat pipe experiments in low earth orbit

    International Nuclear Information System (INIS)

    Woloshun, K.; Merrigan, M.A.; Sena, J.T.; Critchley, E.

    1993-01-01

    Although high temperature, liquid metal heat pipe radiators have become a standard component on most high power space power system designs, there is no experimental data on the operation of these heat pipes in a zero gravity or micro-gravity environment. Experiments to benchmark the transient and steady state performance of prototypical heat pipe space radiator elements are in preparation for testing in low earth orbit. It is anticipated that these heat pipes will be tested aborad the Space Shuttle in 1995. Three heat pipes will be tested in a cargo bay Get Away Special (GAS) canister. The heat pipes are SST/potassium, each with a different wick structure; homogeneous, arterial, and annular gap, the heat pipes have been designed, fabricated, and ground tested. In this paper, the heat pipe designs are specified, and transient and steady-state ground test data are presented

  4. Ultrasonic testing with the phased array method at the pipe connection inner edges in pipings

    International Nuclear Information System (INIS)

    Brekow, G.; Wuestenberg, H.; Hesselmann, H.; Rathgeb, W.

    1991-01-01

    Ultrasonic testing with the phased array method at the pipe connection inner edges in pipings. The pipe connection inner corner tests in feedwater lines to the main coolant pipe were carried out by Preussen-Elektra in cooperation with Siemens KWU and the BAM with the ultrasonic phased array method. The testing plan was developed by means of a computed model. For a trial of the testing plan, numerous ultrasonic measurements with the phased array method were carried out using a pipe test piece with TH-type inner edges, which was a 1:1 model of the reactor component to be tested. The data measured at several test notches in the pipe connection inner edge area covered by a plating of 6 mm were analyzed. (orig./MM) [de

  5. Computer simulation of LMFBR piping systems

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.W.; Fistedis, S.H.

    1977-01-01

    Integrity of piping systems is one of the main concerns of the safety issues of Liquid Metal Fast Breeder Reactors (LMFBR). Hypothetical core disruptive accidents (HCDA) and water-sodium interaction are two examples of sources of high pressure pulses that endanger the integrity of the heat transport piping systems of LMFBRs. Although plastic wall deformation attenuates pressure peaks so that only pressures slightly higher than the pipe yield pressure propagate along the system, the interaction of these pulses with the different components of the system, such as elbows, valves, heat exchangers, etc.; and with one another produce a complex system of pressure pulses that cause more plastic deformation and perhaps damage to components. A generalized piping component and a tee branching model are described. An optional tube bundle and interior rigid wall simulation model makes such a generalized component model suited for modelling of valves, reducers, expansions, and heat exchangers. The generalized component and the tee branching junction models are combined with the pipe-elbow loop model so that a more general piping system can be analyzed both hydrodynamically and structurally under the effect of simultaneous pressure pulses

  6. Some remarks on the analysis of stress-corrosion cracking of austenitic stainless-steel cladding

    International Nuclear Information System (INIS)

    Kupka, I.; Nrkous, P.

    1977-01-01

    Stress-corrosion cracking is greatly influenced by tensile stresses in the material. The occurrence of tensile stresses in the material under consideration results from residual stresses brought about during manufacturing processes and from stress caused by operation. In the case of an austenitic steel cladding the residual stresses arise in the course of welding and thermal treatment. The technique of residual stress measurement in austenitic cladding materials is described and the results are given. Both the longitudinal and transverse components of the stresses show in all cases similar behaviour not only prior to, but also after heat treatment. (J.B.)

  7. Nondestructive testing of austenitic casting and dissimilar metal welds; Kaksimetalliliitosten ja austeniittisten valujen testaustekniikoiden vertailu

    Energy Technology Data Exchange (ETDEWEB)

    Lahdenperae, K [VTT Manufacturing Technology, Espoo (Finland)

    1995-01-01

    The publication is a literature study of nondestructive testing of dissimilar metal welds and cast austenitic components in PWR and BWR plants. A major key to the successful testing is a realistic mockup made of the materials to be tested. The inspectors must also be trained and validated using suitable mockups. (42 refs., 27 figs., 10 tabs.).

  8. Simulation of radiation induced segregation and PWSCC susceptibility for austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto Koji; Yonezawa, Toshio; Iwamura, Toshihiko [Mitsubishi Heavy Industries Ltd., Takasago, Hyogo (Japan). Takasago R and D Center; Ajiki, Kazuhide [Mitsubishi Heavy Industries Ltd., Kobe (Japan). Kobe Shipyard and Machinery Works; Urata, Sigeru [General Office of Nuclear and Fossil Power Production, Kansai Electric Power Co., Inc., Osaka (Japan)

    2000-08-01

    Recently, irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels for core internal components materials become a subject of discussion in light water reactors (LWRs). IASCC has not been found in Pressurized Water Reactors (PWRs). However, the authors have investigated on the possibility of IASCC of austenitic stainless steels for core internal materials so as to be able to estimate the degradation of PWR plants up to the end of their lifetime. In this study, in order to verify the hypothetical that the IASCC in PWRs shall be caused by the primary water stress corrosion cracking (PWSCC) as a result of radiation induced segregation (RIS) at grain boundaries, the authors simulated RIS at grain boundaries of austenitic stainless steels based on previous study and estimated RIS tendency after long time operation. And the authors melted the test alloys whose bulk compositions simulated the grain boundary compositions of irradiated austenitic stainless steels and made clear chromium-nickel-silicon compositions for PWSCC susceptibility area in austenitic alloys by slow strain rate tensile (SSRT) test. (author)

  9. Simulation of radiation induced segregation and PWSCC susceptibility for austenitic stainless steels

    International Nuclear Information System (INIS)

    Fujimoto Koji; Yonezawa, Toshio; Iwamura, Toshihiko

    2000-01-01

    Recently, irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels for core internal components materials become a subject of discussion in light water reactors (LWRs). IASCC has not been found in Pressurized Water Reactors (PWRs). However, the authors have investigated on the possibility of IASCC of austenitic stainless steels for core internal materials so as to be able to estimate the degradation of PWR plants up to the end of their lifetime. In this study, in order to verify the hypothetical that the IASCC in PWRs shall be caused by the primary water stress corrosion cracking (PWSCC) as a result of radiation induced segregation (RIS) at grain boundaries, the authors simulated RIS at grain boundaries of austenitic stainless steels based on previous study and estimated RIS tendency after long time operation. And the authors melted the test alloys whose bulk compositions simulated the grain boundary compositions of irradiated austenitic stainless steels and made clear chromium-nickel-silicon compositions for PWSCC susceptibility area in austenitic alloys by slow strain rate tensile (SSRT) test. (author)

  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. Copyright © 2013 Elsevier B.V. All rights reserved.

  11. Ultrasonic inspection of austenitic welds

    International Nuclear Information System (INIS)

    Baikie, B.L.; Wagg, A.R.; Whittle, M.J.; Yapp, D.

    1976-01-01

    Optical and X-ray metallography combined with ultrasonic testing by compression waves was used for inspection of stainless steel weld metal produced by three different welding techniques. X-ray diffraction showed that each weld possessed a characteristic fibre textured structure which was shown by optical microscopy to be parallel to columnar grain boundaries. Metallographic evidence suggested that the development of fibre texture is due to the mechanism of competitive growth. From observations made as a result of optical metallographic examination the orientation of the fibre axis could be predicted if the weld geometry and welding procedure were known. Ultrasonic velocity and attenuation measurements as a continuous function of grain orientation, made on cylinders machined from weld samples, showed that attenuation was strongly orientation dependent. It was concluded that the sensitivity of ultrasonic inspection to small defects is unlikely to be as high for austenitic welds as for ferritic even when transmission is improved by modifying the welding procedure to improve the ultrasonic transmission. (U.K.)

  12. Application of Moessbauer effect to the study of austenite retained in low carbon steels

    International Nuclear Information System (INIS)

    Azevedo, A.L.T. de; Silva, E.G. da

    1979-01-01

    Moessbauer effect measurements were performed in two samples of low carbon, low alloy steels, one with a bainite granular microstructure and the other a martensitic one. The concentration of the retained austenite was determined in both samples by Moessbauer spectrometry and X radiation, a very good agreement for the sample with a greater austenite content having been observed. From the assumption that the carbon atoms in the f.c.c. matrix repel one another due to Coulomb interactions, giving origin to quadrupolar interactions, it was possible to determine carbon concentration in the MA (Martensite Austenite) components of bainite, the results being in good agreement with the one obtained from metallographic considerations. (I.C.R.) [pt

  13. Pipe failure probability - the Thomas paper revisited

    International Nuclear Information System (INIS)

    Lydell, B.O.Y.

    2000-01-01

    Almost twenty years ago, in Volume 2 of Reliability Engineering (the predecessor of Reliability Engineering and System Safety), a paper by H. M. Thomas of Rolls Royce and Associates Ltd. presented a generalized approach to the estimation of piping and vessel failure probability. The 'Thomas-approach' used insights from actual failure statistics to calculate the probability of leakage and conditional probability of rupture given leakage. It was intended for practitioners without access to data on the service experience with piping and piping system components. This article revisits the Thomas paper by drawing on insights from development of a new database on piping failures in commercial nuclear power plants worldwide (SKI-PIPE). Partially sponsored by the Swedish Nuclear Power Inspectorate (SKI), the R and D leading up to this note was performed during 1994-1999. Motivated by data requirements of reliability analysis and probabilistic safety assessment (PSA), the new database supports statistical analysis of piping failure data. Against the background of this database development program, the article reviews the applicability of the 'Thomas approach' in applied risk and reliability analysis. It addresses the question whether a new and expanded database on the service experience with piping systems would alter the original piping reliability correlation as suggested by H. M. Thomas

  14. Pipe-to-pipe impact program

    International Nuclear Information System (INIS)

    Alzheimer, J.M.; Bampton, M.C.C.; Friley, J.R.; Simonen, F.A.

    1984-06-01

    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

  15. 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

  16. Experimental benchmark for piping system dynamic response analyses

    International Nuclear Information System (INIS)

    Schott, G.A.; Mallett, R.H.

    1981-01-01

    The scope and status of a piping system dynamics test program are described. A 0.20-m nominal diameter test piping specimen is designed to be representative of main heat transport system piping of LMFBR plants. 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. 3 refs

  17. Experimental benchmark for piping system dynamic-response analyses

    International Nuclear Information System (INIS)

    1981-01-01

    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

  18. Safety design guide for pipe rupture protection for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for pipe rupture protection identifies high-energy systems in which pipe ruptures must be postulated to occur, as well as systems that must be protected from the dynamic effects of such ruptures. Dynamic effects considered in this SDG consist of pipe whip (including missiles generated by pipe ruptures, if any) and jet impingement, Requirements for protection against the dynamic effects of a postulated pipe rupture and method of protection of essential structures, systems and components are specified for these effects. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 5 refs. (Author) .new

  19. Effect of nitrogen and boron on weldability of austenitic stainless steels

    International Nuclear Information System (INIS)

    Bhaduri, A.K.; Albert, S.K.; Srinivasan, G.; Divya, M.; Das, C.R.

    2012-01-01

    Hot cracking is a major problem in the welding of austenitic stainless steels, particularly the fully austenitic grades. A group of alloys of enhanced-nitrogen 316LN austenitic stainless steel is being developed for structural components of the Indian Fast Reactor programme. Studying the hot cracking behaviour of this nitrogen-enhanced austenitic stainless steel is an important consideration during welding, as this material solidifies without any residual delta ferrite in the primary austenitic mode. Nitrogen has potent effects on the solidification microstructure, and hence has a strong influence on the hot cracking behaviour. Different heats of this material were investigated, which included fully austenitic stainless steels containing 0.070.22 wt% nitrogen. Also, borated austenitic stainless steels, such as type 304B4, have been widely used in the nuclear applications primarily due to its higher neutron absorption efficiency. Weldability is a major concern for this alloy due to the formation of low melting eutectic phase that is enriched with iron, chromium, molybdenum and boron. Fully austenitic stainless steels are prone to hot cracking during welding in the absence of a small amount of delta ferrite, especially for compositions rich in elements like boron that increases the tendency to form low melting eutectics. Detailed weldability investigations were carried out on a grade 304B4 stainless steel containing 1.3 wt% boron. Among the many approaches that have been used to determine the hot cracking susceptibility of different alloys, Variable-Restraint (Varestraint) weld test and Hot Ductility (Gleeble) tests are commonly used to evaluate the weldability of austenitic alloys. Hence, investigations on these materials consisted of detailed metallurgical characterization and weldability studies that included studying both the fusion zone and liquation cracking susceptibility, using Varestraint tests at 0.254.0%, strain levels and Gleeble (thermo

  20. Solar heating pipe

    Energy Technology Data Exchange (ETDEWEB)

    Hinson-Rider, G.

    1977-10-04

    A fluid carrying pipe is described having an integral transparent portion formed into a longitudinally extending cylindrical lens that focuses solar heat rays to a focal axis within the volume of the pipe. The pipe on the side opposite the lens has a heat ray absorbent coating for absorbing heat from light rays that pass through the focal axis.

  1. Constitutive modeling of metastable austenitic stainless steel (CD-rom)

    NARCIS (Netherlands)

    Perdahcioglu, Emin Semih; Geijselaers, Hubertus J.M.; Huetink, Han; Boisse, P.

    2008-01-01

    A stress-update algorithm is developed for austenitic metastable steels which undergo phase evolution during deformation. The material initially comprises only the soft and ductile austenite phase which due to the phenomenon of mechanically induced martensitic transformation, transforms completely

  2. Pipe whip analysis using the TEDEL code

    International Nuclear Information System (INIS)

    Millard, D.; Hoffmann, A.

    1985-02-01

    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

  3. Thinned pipe management program of Korean NPPs

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  4. Transformation in austenitic stainless steel sheet under different loading directions

    NARCIS (Netherlands)

    van den Boogaard, Antonius H.; Krauer, J.; Hora, P.

    2011-01-01

    The stress-strain relation for austenitic stainless steels is based on 2 main contributions: work hardening and a phase transformation from austenite to martensite. The transformation is highly temperature dependent. In most models for phase transformation from austenite to martensite, the stress

  5. Transformation in Austenitic Stainless Steel Sheet under Different Loading Directions

    NARCIS (Netherlands)

    van den Boogaard, Antonius H.; Krauer, J.; Hora, P.

    2011-01-01

    The stress-strain relation for austenitic stainless steels is based on 2 main contributions: work hardening and a phase transformation from austenite to martensite. The transformation is highly temperature dependent. In most models for phase transformation from austenite to martensite, the stress

  6. Piping dynamic analysis by the synthesis method

    International Nuclear Information System (INIS)

    Bezler, P.; Curreri, J.R.

    1976-01-01

    Since piping systems are a frequent source of noise and vibrations, their efficient dynamic analysis is imperative. As an alternate to more conventional analyses methods, an application of the synthesis method to piping vibrations analyses is demonstrated. Specifically, the technique is illustrated by determining the normal modes and natural frequencies of a composite bend from the normal mode and natural frequency data of two component parts. A comparison of the results to those derived for the composite bend by other techniques is made

  7. Longitudinal wave ultrasonic inspection of austenitic weldments

    International Nuclear Information System (INIS)

    Gray, B.S.; Hudgell, R.J.; Seed, H.

    1980-01-01

    Successful volumetric inspection of LMFBR primary circuits, and also much of the secondary circuit, is dependent on the availability of satisfactory examination procedures for austenitic welds. Application of conventional ultrasonic techniques is hampered by the anisotropic, textured structure of the weld metal and this paper describes development work on the use of longitudinal wave techniques. In addition to confirming the dominant effects of the weld structure on ultrasound propagation some results are given of studies utilising deliberately induced defects in Manual Metal Arc Welds in 50 mm plate together with preliminary work on the inspection of narrow austenitic welds fabricated by automatic processes. (author)

  8. Recrystallization induced plasticity in austenite and ferrite

    International Nuclear Information System (INIS)

    Huang Mingxin; Pineau, André; Bouaziz, Olivier; Vu, Trong-Dai

    2012-01-01

    Highlights: ► Plasticity can be induced by recrystallization in austenite and ferrite. ► Strain rate is proportional to recrystallization kinetics. ► Overall atomic flux selects a preferential direction may be the origin. - Abstract: New experimental evidences are provided to demonstrate that plastic strain can be induced by recrystallization in austenite and ferrite under an applied stress much smaller than their yield stresses. Such Recrystallization Induced Plasticity (RIP) phenomenon occurs because the overall atomic flux during recrystallization follows a preferential direction imposed by the applied stress.

  9. 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.

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

    International Nuclear Information System (INIS)

    Porollo, S.I.; Shulepin, S.V.; Konobeev, Y.V.; Garner, F.

    2007-01-01

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

    International Nuclear Information System (INIS)

    Ehrhart, W.S.; Elder, J.B.; Sprayberry, R.E.; Vande Kamp, R.W.

    1990-01-01

    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

  12. Some data of second sequence non standard austenitic ingot, A2

    International Nuclear Information System (INIS)

    Nurdin Effendi; Aziz K Jahja; Bandriana; Wisnu Ari Adi

    2012-01-01

    Synthesis of second sequence austenite stainless steel named A2 using extracted minerals from Indonesian mines has been carried out. The starting materials for austenite alloy consist of granular ferro scrap, nickel, ferro-chrome, ferro-manganese, and ferro-silicon. The second sequence composition differs from the former first sequence. This A2 sequence contained more nickel, meanwhile titanium element had not been added explicitly to it, and just been found from raw materials contents or impurities, as well as carbon content in the alloy. However before the actual alloying work started, the first important step was to carry out the determination of the fractional amount of each starting material necessary to form an austenite stainless steel alloy as specified. Once the component fraction of each base alloy-element was determined, the raw materials are weighed on the mini-balance. After the fractional quantities of each constituent have been computed, an appropriate amount of these base materials are weighed separately on the micro scale. The raw materials were then placed in the induction foundry furnace, which was operated by an electromagnetic inductive-thermal system. The foundry furnace system performs the stirring of the molten materials automatically. The homogenized molten metals were poured down into sand casting prepared in advance. Some of the austenite stainless steel were normalized at 600°C for 6 hours. The average density is 7.8 g cm -1 and the average hardness value of 'normalized' austenite stainless-steels is in the range of 460 on the Vickers scale. The microstructure observation concludes that an extensive portion of the sample's structure is dendritic and the surface turns out to be homogenous. X-ray diffraction analysis shows that the material belongs to the fcc crystallographic system, which fits in with the austenite class of the alloy. The experimental fractional elemental composition data acquired by OES method turn out to

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

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

    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

  14. Drill pipe bridge plug

    International Nuclear Information System (INIS)

    Winslow, D.W.; Brisco, D.P.

    1991-01-01

    This patent describes a method of stopping flow of fluid up through a pipe bore of a pipe string in a well. It comprises: lowering a bridge plug apparatus on a work string into the pipe string to a position where the pipe bore is to be closed; communicating the pipe bore below a packer of the bridge plug apparatus through the bridge plug apparatus with a low pressure zone above the packer to permit the fluid to flow up through the bridge plug apparatus; engaging the bridge plug apparatus with an internal upset of the pipe string; while the fluid is flowing up through the bridge plug apparatus, pulling upward on the work string and the bridge plug apparatus and thereby sealing the packer against the pipe bore; isolating the pipe bore below the packer from the low pressure zone above the packer and thereby stopping flow of the fluid up through the pipe bore; disconnecting the work string from the bridge plug apparatus; and maintaining the bridge plug apparatus in engagement with the internal upset and sealed against the pipe bore due to an upward pressure differential applied to the bridge plug apparatus by the fluid contained therebelow

  15. Welding metallurgy of austenitic stainless steels

    International Nuclear Information System (INIS)

    Ibrahim, A.N.

    1983-01-01

    Austenitic stainless steels welds are commonly found in nuclear reactor systems. The macrostructure and the transformation of delta -phase into γ - phase which occur during rapid solidification of such welds are discussed. Finally, several types of defects which are derived from the welding operation, particularly defects of crack type, are also discussed in brief. (author)

  16. Modeling of austenite to ferrite transformation

    Indian Academy of Sciences (India)

    395–398. c Indian Academy of Sciences. Modeling of austenite to ferrite transformation. MOHSEN KAZEMINEZHAD. ∗. Department of Materials Science and Engineering, Sharif University of Technology, Azadi Avenue, Tehran, Iran. MS received 17 January 2011; revised 9 July 2011. Abstract. In this research, an algorithm ...

  17. 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.

  18. Austenitic stainless steels for cryogenic service

    International Nuclear Information System (INIS)

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

    1985-01-01

    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

  19. Miniature Heat Pipes

    Science.gov (United States)

    1997-01-01

    Small Business Innovation Research contracts from Goddard Space Flight Center to Thermacore Inc. have fostered the company work on devices tagged "heat pipes" for space application. To control the extreme temperature ranges in space, heat pipes are important to spacecraft. The problem was to maintain an 8-watt central processing unit (CPU) at less than 90 C in a notebook computer using no power, with very little space available and without using forced convection. Thermacore's answer was in the design of a powder metal wick that transfers CPU heat from a tightly confined spot to an area near available air flow. The heat pipe technology permits a notebook computer to be operated in any position without loss of performance. Miniature heat pipe technology has successfully been applied, such as in Pentium Processor notebook computers. The company expects its heat pipes to accommodate desktop computers as well. Cellular phones, camcorders, and other hand-held electronics are forsible applications for heat pipes.

  20. 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 ...

  1. Development of support system for nuclear power plant piping

    International Nuclear Information System (INIS)

    Horino, Satoshi

    1987-01-01

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

  2. 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)

  3. Degradation mechanisms of small scale piping systems

    International Nuclear Information System (INIS)

    Bartonicek, J.; Koenig, G.; Blind, D.

    1996-01-01

    Operational experience shows that many degradation mechanisms can have an effect on small-scale piping systems. We can see from the analyses carried out that the degradation which has occurred is primarily linked with the fact that these piping systems were classified as being of low safety relevance. This is mainly due to such components being classified into low safety relevance category at the design stage, as well as to the low level of operational monitoring. Since in spite of the variety of designs and operational modes the degradation mechanisms detected may be attributed to the piping systems, we can make decisive statements on how to avoid such degradation mechanisms. Even small-scale piping systems may achieve guaranteed integrity in such cases by taking the appropriate action. (orig.) [de

  4. 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.

  5. Riser pipe elevator

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, W.; Jimenez, A.F.

    1987-09-08

    This patent describes a method for storing and retrieving a riser pipe, comprising the steps of: providing an upright annular magazine comprised of an inside annular wall and an outside annular wall, the magazine having an open top; storing the riser pipe in a substantially vertically oriented position within the annular magazine; and moving the riser pipe upwardly through the open top of the annular magazine at an angle to the vertical along at least a portion of the length of the riser pipe.

  6. Piping engineering and operation

    International Nuclear Information System (INIS)

    1993-01-01

    The conference 'Piping Engineering and Operation' was organized by the Institution of Mechanical Engineers in November/December 1993 to follow on from similar successful events of 1985 and 1989, which were attended by representatives from all sectors of the piping industry. Development of engineering and operation of piping systems in all aspects, including non-metallic materials, are highlighted. The range of issues covered represents a balance between current practices and implementation of future international standards. Twenty papers are printed. Two, which are concerned with pressurized pipes or steam lines in the nuclear industry, are indexed separately. (Author)

  7. Piping equipment; Materiel petrole

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

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

    International Nuclear Information System (INIS)

    Chellapandi, P.

    2013-01-01

    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

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

    International Nuclear Information System (INIS)

    Bertrand, C.; Ardellier, A.

    1996-01-01

    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

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

    International Nuclear Information System (INIS)

    Shack, W.J.

    1988-01-01

    Several types of environmental degradation of piping in light water reactor (LWR) power systems have already had significant economic impact on the industry. These include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping, erosion-corrosion of carbon steel piping in secondary systems, and a variety of types of fatigue failures. In addition, other problems have been identified that must be addressed in considering extended lifetimes for nuclear plants. These include the embrittlement of cast stainless steels after extended thermal aging at reactor operating temperatures and the effect of reactor environments on the design margin inherent in the ASME Section III fatigue design curves especially for carbon steel piping. These problems are being addressed by wide-ranging research programs in this country and abroad. The purpose of this review is to highlight some of the accomplishments of these programs and to note some of the remaining unanswered questions

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

    International Nuclear Information System (INIS)

    Shack, W.J.

    1987-08-01

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

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

    International Nuclear Information System (INIS)

    Green, D.R.

    1979-01-01

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

  13. Evaluation of J-integral estimation scheme for flawed throughwall pipes

    Energy Technology Data Exchange (ETDEWEB)

    Zahoor, A.

    1987-02-01

    The accuracy of the EPRI J-integral estimation scheme for pipes with throughwall cracks and subjected to pure bending was assessed using available experimental data on circumferentially flawed throughwall pipes. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. The results indicated that the EPRI J estimation scheme solutions are unnecessarily conservative compared to results from pipe experiments. As a result of these evaluations an improved J estimation scheme is developed, which is shown to have improved accuracy compared to the original EPRI J estimation scheme. These results imply that the flaw evaluation procedures in the ASME Code on austenitic piping welds are conservative. These results also have applications to the leak before break fracture mechanics analyses.

  14. Evaluation of J-integral estimation scheme for flawed throughwall pipes

    International Nuclear Information System (INIS)

    Zahoor, A.

    1987-01-01

    The accuracy of the EPRI J-integral estimation scheme for pipes with throughwall cracks and subjected to pure bending was assessed using available experimental data on circumferentially flawed throughwall pipes. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. The results indicated that the EPRI J estimation scheme solutions are unnecessarily conservative compared to results from pipe experiments. As a result of these evaluations an improved J estimation scheme is developed, which is shown to have improved accuracy compared to the original EPRI J estimation scheme. These results imply that the flaw evaluation procedures in the ASME Code on austenitic piping welds are conservative. These results also have applications to the leak before break fracture mechanics analyses. (orig.)

  15. Design rules for piping: Plastic stability of straight parts under level D loadings

    International Nuclear Information System (INIS)

    Touboul, F.; Ben Djidia, M.; Acker, D.

    1989-01-01

    Design rules for piping, elaborated for Fast Breeder Reactors, are based on analysis performed for Pressure Water Reactors. Interpretation of largely diversified straight parts tests, enable us to validate and improve existing rules and to propose a more suitable formula. Design rules for piping appear to be non conservative for austenitic thin tubes in bending or torsion. By introducing a B 2 coefficient, geometrically dependent, the gap between thin and thick tubes may be withheld. Conservatism of rules can be ensured by considering the allowable stress defined by ASME, Section III, Appendix F

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

    International Nuclear Information System (INIS)

    Steingrimsdottir, Kristin; Dahlberg, Magnus

    2011-10-01

    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

  17. Initiation model for intergranular stress corrosion cracking in BWR pipes

    International Nuclear Information System (INIS)

    Hishida, Mamoru; Kawakubo, Takashi; Nakagawa, Yuji; Arii, Mitsuru.

    1981-01-01

    Discussions were made on the keys of intergranular stress corrosion cracking of austenitic stainless steel in high-temperature water in laboratories and stress corrosion cracking incidents in operating plants. Based on these discussions, a model was set up of intergranular stress corrosion cracking initiation in BWR pipes. Regarding the model, it was presumed that the intergranular stress corrosion cracking initiates during start up periods whenever heat-affected zones in welded pipes are highly sensitized and suffer dynamic strain in transient water containing dissolved oxygen. A series of BWR start up simulation tests were made by using a flowing autoclave system with slow strain rate test equipment. Validity of the model was confirmed through the test results. (author)

  18. Formation of austenite in high Cr ferritic/martensitic steels by high fluence neutron irradiation

    Science.gov (United States)

    Lu, Z.; Faulkner, R. G.; Morgan, T. S.

    2008-12-01

    High Cr ferritic/martensitic steels are leading candidates for structural components of future fusion reactors and new generation fission reactors due to their excellent swelling resistance and thermal properties. A commercial grade 12%CrMoVNb ferritic/martensitic stainless steel in the form of parent plate and off-normal weld materials was fast neutron irradiated up to 33 dpa (1.1 × 10 -6 dpa/s) at 400 °C and 28 dpa (1.7 × 10 -6 dpa/s) at 465 °C, respectively. TEM investigation shows that the fully martensitic weld metal transformed to a duplex austenite/ferrite structure due to high fluence neutron irradiation, the austenite was heavily voided (˜15 vol.%) and the ferrite was relatively void-free; whilst no austenite phases were detected in plate steel. Thermodynamic and phase equilibria software MTDATA has been employed for the first time to investigate neutron irradiation-induced phase transformations. The neutron irradiation effect is introduced by adding additional Gibbs free energy into the system. This additional energy is produced by high energy neutron irradiation and can be estimated from the increased dislocation loop density caused by irradiation. Modelling results show that neutron irradiation reduces the ferrite/austenite transformation temperature, especially for high Ni weld metal. The calculated results exhibit good agreement with experimental observation.

  19. Piping data retrieval system (PDRS): An integrated package to aid piping layout

    International Nuclear Information System (INIS)

    Vyas, K.N.; Sharma, A.; Susandhi, R.; Basu, S.

    1986-01-01

    An integrated package to aid piping layout has been developed and implemented on PDP-11/34 system at Hall 7. The package allows various equipments to be modelled, consisting of primitive equipment components. The equipment layout for the plant can then be reproduced in the form of drawings such as plan, elevation, isometric or perspective. The package has the built in function to perform hidden line removal among equipments. Once the equipment layout is finalised, the package aids in superimposing the piping as per the specified pipe routine. The report discusses the general capabilities and the major input requirements for the package. (author)

  20. Laser cladding technology to small diameter pipes

    International Nuclear Information System (INIS)

    Fujimagari, H.; Hagiwara, M.; Kojima, T.

    2000-01-01

    A laser cladding method which produces a highly corrosion-resistant material coating layers (cladding) on the austenitic stainless steel (type 304 SS) pipe inner surface was developed to prevent SCC (stress corrosion cracking) occurrence. This technology is applicable to a narrow and long distance area from operators, because of the good accessibility of the YAG (yttrium-aluminum-garnet) laser beam that can be transmitted through an optical fiber. In this method a mixed paste metallic powder and heating-resistive organic solvent are firstly placed on the inner surface of a small pipe, and then a YAG laser beam transmitted through an optical fiber irradiates to the pasted area. A mixed paste will be melted and form a cladding layer subsequently. A cladding layer shows as excellent corrosion resistance property. This laser cladding (LC) method had already applied to several domestic nuclear power plants and had obtained a good reputation. This report introduces the outline of laser cladding technology, the developed equipment for practical application in the field, and the circumstance in actual plant application. (orig.)

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

    International Nuclear Information System (INIS)

    Moore, S.E.

    1975-11-01

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

  2. Transients in pipes

    International Nuclear Information System (INIS)

    Marchesin, D.; Paes-Leme, P.J.S.; Sampaio, R.

    1981-01-01

    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) [pt

  3. 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…

  4. Piping research program plan

    International Nuclear Information System (INIS)

    1988-09-01

    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)

  5. 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...

  6. Thermal fatigue cracking of austenitic stainless steels

    International Nuclear Information System (INIS)

    Fissolo, A.

    2001-01-01

    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 N i 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 >> N i , 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

  7. 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...

  8. 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.

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

    International Nuclear Information System (INIS)

    Stoller, R.E.

    1987-12-01

    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

  10. 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.

  11. The nucleation of austenite in ferritic ductile cast iron

    International Nuclear Information System (INIS)

    Chou, J.M.; Hon, M.H.; Lee, J.L.

    1992-01-01

    Austempered ductile cast iron has recently been receiving increasing attention because of its excellent combination of strength and ductility. Since the austenitization process has a significant influence on the mechanical properties of austempered ductile cast iron, several investigations on the nucleation sites of austenite and diffusion paths of carbon from spheroidal graphite have been reported in ferritic ductile cast iron. However, agreement on this subject has not ben reached. The purpose of this paper is to study the preferential nucleation sites of austenite during austenitization at two austenitizing temperatures in ferritic ductile cast iron. An attempt was made to understand the reasons which give rise to preferential austenite nucleation sites. The carbon diffusion paths from spheroidal graphite were also investigated

  12. J evaluation by simplified method for cracked pipes under mechanical loading

    International Nuclear Information System (INIS)

    Lacire, M.H.; Michel, B.; Gilles, P.

    2001-01-01

    The integrity of structures behaviour is an important subject for the nuclear reactor safety. Most of assessment methods of cracked components are based on the evaluation of the parameter J. However to avoid complex elastic-plastic finite element calculations of J, a simplified method has been jointly developed by CEA, EDF and Framatome. This method, called Js, is based on the reference stress approach and a new KI handbook. To validate this method, a complete set of 2D and 3D elastic-plastic finite element calculations of J have been performed on pipes (more than 300 calculations are available) for different types of part through wall crack (circumferential or longitudinal); mechanical loading (pressure, bending moment, axial load, torsion moment, and combination of these loading); different kind of materials (austenitic or ferritic steel). This paper presents a comparison between the simplified assessment of J and finite element results on these configurations for mechanical loading. Then, validity of the method is discussed and an applicability domain is proposed. (author)

  13. Diffractometry of expanded austenite using synchrotron radiation

    International Nuclear Information System (INIS)

    Fewell, M.P.; Priest, J.M.; Collins, G.A.; Short, K.T.

    2000-01-01

    Full text: The question of the structure of the nitrogen-rich surface layer produced in the nitriding of austenitic stainless steel has been controversial for some time. Diffractometry using conventional x-ray sources is routinely carried out on this material. The result universally seen shows an ostensibly f.c.c. lattice with a larger lattice parameter than that of the underlying austenite. The difficulty with this interpretation lies in the 200 reflection, which lies at slightly lower Bragg angle than expected on the basis of the 111, 220 and 311 reflections. This behaviour is seen in all work known to us, regardless of the grade of austenitic stainless steel or the details of the nitriding technique. It has been explained as due to a mixed f.c.c. phase with different grains having different lattice constants, or as due to a tetragonal distortion of the lattice or an f.c.c lattice with a high frequency of stacking faults, or as indicating a triclinic lattice with a unit cell having all sides equal and two angles equal

  14. Forecasting of mechanical - and structural behavior of 316 austenitic stainless steels by deformation charts

    International Nuclear Information System (INIS)

    Monteiro, S.N.

    1980-01-01

    The utilization of deformation charts applied to AISI 316 austenitic stainless steel with the purpose of foreseeing its behavior associated with structural and mechanical phenomena, is evaluated. The ocurrence of phenomena such as dynamic aging, martensite transformation, static aging, failure at creep curve, cells, subgrains and boundary slips is discussed in the different regions of the chart. A practical example of the charts' utilization for components of fast reactors is finally presented. (Author) [pt

  15. 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.

  16. Evaluation of aging of cast stainless steel components

    International Nuclear Information System (INIS)

    Chung, H.M.

    1991-02-01

    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

  17. Advanced ultrasonic techniques for nondestructive testing of austenitic and dissimilar welds in nuclear facilities

    Science.gov (United States)

    Juengert, Anne; Dugan, Sandra; Homann, Tobias; Mitzscherling, Steffen; Prager, Jens; Pudovikov, Sergey; Schwender, Thomas

    2018-04-01

    Austenitic stainless steel welds as well as dissimilar metal welds with nickel alloy filler material, used in safety relevant parts of nuclear power plants, still challenge the ultrasonic inspection. The weld material forms large oriented grains that lead, on the one hand, to high sound scattering and, on the other hand, to inhomogeneity and to the acoustic anisotropy of the weld structure. The ultrasonic wave fronts do not propagate linearly, as in ferritic weld joints, but along the curves, which depend on the specific grain structure of the weld. Due to the influence of these phenomena, it is difficult to analyze the inspection results and to classify the ultrasonic indications, which could be both from the weld geometry and from the material defects. A correct flaw sizing is not possible. In an ongoing research project, different techniques to improve the reliability of ultrasonic testing at these kinds of welds are investigated. In a first step (in the previous research project) two ultrasonic inspection techniques were developed and validated on plane test specimens with artificial and realistic flaws. In the ongoing project, these techniques are applied to circumferential pipe welds with longitudinal and transverse flaws. The technique developed at the Federal Institute for Materials Research and Testing (BAM) in Germany uses a combination of ray tracing and synthetic aperture focusing technique (SAFT). To investigate the unknown grain structure, the velocity distribution of weld-transmitting ultrasound waves is measured and used to model the weld by ray tracing. The second technique, developed at the Fraunhofer Institute for Nondestructive Testing (IZFP) in Germany, uses Sampling Phased Array (Full Matrix Capture) combined with the reverse phase matching (RPM) and the gradient elastic constant descent algorithm (GECDM). This inspection method is able to estimate the elastic constants of the columnar grains in the weld and offers an improvement of the

  18. 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.

  19. Crystallography of lath martensite and stabilization of retained austenite

    International Nuclear Information System (INIS)

    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 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. Comprehensive Deformation Analysis of a Newly Designed Ni-Free Duplex Stainless Steel with Enhanced Plasticity by Optimizing Austenite Stability

    DEFF Research Database (Denmark)

    Moallemi, Mohammad; Zarei-Hanzaki, Abbas; Eskandari, Mostafa

    2017-01-01

    A new metastable Ni-free duplex stainless steel has been designed with superior plasticity by optimizing austenite stability using thermodynamic calculations of stacking fault energy and with reference to literature findings. Several characterization methods comprising optical microscopy, magnetic......, including an ultimate tensile strength of ~900 MPa and elongation to fracture of ~94 pct due to the synergistic effects of transformation-induced plasticity and twinning-induced plasticity. The deformation mechanism of austenite is complex and includes deformation banding, strain-induced martensite...... formation, and deformation-induced twinning, while the ferrite phase mainly deforms by dislocation slip. Texture analysis indicates that the Copper and Rotated Brass textures in austenite (FCC phase) and {001}〈110〉 texture in ferrite and martensite (BCC phases) are the main active components during...

  1. Survey on application of probabilistic fracture mechanics approach to nuclear piping

    International Nuclear Information System (INIS)

    Kashima, Koichi

    1987-01-01

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

  2. 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

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

    International Nuclear Information System (INIS)

    Shcherbakov, E.N.; Kozlov, A.V.; Yagovitin, R.I.; Evseev, M.V.; Kinev, E.A.; Isobe, Y.; Sagisaka, M.; Okita, T.; Sekimura, N.; Garner, F.

    2007-01-01

    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 (∼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 -9 to 4 x 10 -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 ∼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)

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

    International Nuclear Information System (INIS)

    Garner, F.A.; Greenwood, L.R.; Harrod, D.L.

    1993-08-01

    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 in the pressurized water environment

  5. In Pipe Robot with Hybrid Locomotion System

    Directory of Open Access Journals (Sweden)

    Cristian Miclauş

    2015-06-01

    Full Text Available The first part of the paper covers aspects concerning in pipe robots and their components, such as hybrid locomotion systems and the adapting mechanisms used. The second part describes the inspection robot that was developed, which combines tracked and wheeled locomotion (hybrid locomotion. The end of the paper presents the advantages and disadvantages of the proposed robot.

  6. 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

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

    International Nuclear Information System (INIS)

    Kuzay, T.M.; Kasza, K.E.

    1985-01-01

    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 90 0 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

  8. Resolution of thermal striping issue downstream of a horizontal pipe elbow in stratified pipe flow. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Kuzay, T.M.; Kasza, K.E.

    1985-01-01

    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 90/sup 0/ 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.

  9. Evaluation of the influence of seismic restraint characteristics on breeder reactor piping systems

    International Nuclear Information System (INIS)

    Mello, R.M.; Pollono, L.P.

    1979-01-01

    For the Clinch River Breeder Reactor Plant (CRBRP) heat transport system piping within the reactor containment building, dynamic analyses of the piping loops have been performed to study the effect of restraint stiffness on the dynamic behavior of the piping. In addition, analysis and testing of typical CRBRP restraint system components have been performed for the purpose of quantifying and verifying the basic characteristics of the restraints used in the piping system dynamic analysis

  10. Analytical solution for stress, strain and plastic instability of pressurized pipes with volumetric flaws

    International Nuclear Information System (INIS)

    Cunha, Sérgio B.; Netto, Theodoro A.

    2012-01-01

    The mechanical behavior of internally pressurized pipes with volumetric flaws is analyzed. The two possible modes of circumferentially straining the pipe wall are identified and associated to hypothesized geometries. The radial deformation that takes place by bending the pipe wall is studied by means of axisymmetric flaws and the membrane strain developed by unequal hoop deformation is analyzed with the help of narrow axial flaws. Linear elastic shell solutions for stress and strain are developed, the plastic behavior is studied and the maximum hoop stress at the flaw is related to the undamaged pipe hoop stress by means of stress concentration factors. The stress concentration factors are employed to obtain equations predicting the pressure at which the pipe fails by plastic instability for both types of flaw. These analytical solutions are validated by comparison with burst tests on 3″ diameter pipes and finite element simulations. Forty-one burst tests were carried out and two materials with very dissimilar plastic behavior, carbon steel and austenitic stainless steel, were used in the experiments. Both the analytical and the numerical predictions showed good correlation with the experimentally observed burst pressures. - Highlights: ► An analytical model for the burst of a pipe with a volumetric flaw is developed. ► Deformation, strain and stress are modeled in the elastic and plastic domains. ► The model is comprehensively validated by experiments and numerical simulations. ► The burst pressure model’s accuracy is equivalent to finite element simulations.

  11. Heat pipe development

    Science.gov (United States)

    Bienart, W. B.

    1973-01-01

    The objective of this program was to investigate analytically and experimentally the performance of heat pipes with composite wicks--specifically, those having pedestal arteries and screwthread circumferential grooves. An analytical model was developed to describe the effects of screwthreads and screen secondary wicks on the transport capability of the artery. The model describes the hydrodynamics of the circumferential flow in triangular grooves with azimuthally varying capillary menisci and liquid cross-sections. Normalized results were obtained which give the influence of evaporator heat flux on the axial heat transport capability of the arterial wick. In order to evaluate the priming behavior of composite wicks under actual load conditions, an 'inverted' glass heat pipe was designed and constructed. The results obtained from the analysis and from the tests with the glass heat pipe were applied to the OAO-C Level 5 heat pipe, and an improved correlation between predicted and measured evaporator and transport performance were obtained.

  12. 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)

  13. Simplified pipe gun

    International Nuclear Information System (INIS)

    Sorensen, H.; Nordskov, A.; Sass, B.; Visler, T.

    1987-01-01

    A simplified version of a deuterium pellet gun based on the pipe gun principle is described. The pipe gun is made from a continuous tube of stainless steel and gas is fed in from the muzzle end only. It is indicated that the pellet length is determined by the temperature gradient along the barrel right outside the freezing cell. Velocities of around 1000 m/s with a scatter of +- 2% are obtained with a propellant gas pressure of 40 bar

  14. Stuck pipe prediction

    KAUST Repository

    Alzahrani, Majed

    2016-03-10

    Disclosed are various embodiments for a prediction application to predict a stuck pipe. A linear regression model is generated from hook load readings at corresponding bit depths. A current hook load reading at a current bit depth is compared with a normal hook load reading from the linear regression model. A current hook load greater than a normal hook load for a given bit depth indicates the likelihood of a stuck pipe.

  15. Stuck pipe prediction

    KAUST Repository

    Alzahrani, Majed; Alsolami, Fawaz; Chikalov, Igor; Algharbi, Salem; Aboudi, Faisal; Khudiri, Musab

    2016-01-01

    Disclosed are various embodiments for a prediction application to predict a stuck pipe. A linear regression model is generated from hook load readings at corresponding bit depths. A current hook load reading at a current bit depth is compared with a normal hook load reading from the linear regression model. A current hook load greater than a normal hook load for a given bit depth indicates the likelihood of a stuck pipe.

  16. Low temperature sensitization of austenitic stainless steel: an ageing effect during BWR service

    International Nuclear Information System (INIS)

    Shah, B.K.; Sinha, A.K.; Rastogi, P.K.; Kulkarni, P.G.

    1994-01-01

    Sensitization in austenitic stainless steel refers to chromium carbide precipitation at the grain boundaries with concomitant depletion of chromium below 12% near grain boundaries. This makes the material susceptible to either intergranular corrosion (IGC) or intergranular stress corrosion cracking (IGSCC). This effect is predominant whenever austenitic stainless steel is subjected to thermal exposure in the temperature range 723-1073K either during welding or during heat treatment. Low temperature sensitization (LTS) refers to sensitization at temperature below the typical range of sensitization i.e. 723-1073K. A prerequisite for LTS phenomenon is reported to be the presence of chromium carbide nuclei at the grain boundaries which can grow during boiling water reactor service even at a relatively lower temperature of around 560K. LTS can lead to failure of BWR pipe due to IGSCC. The paper reviews the phenomenological and mechanistic aspects of LTS. Studies carried out regarding effect of prior cold work on LTS are reported. Summary of the studies reported in literature to examine the occurrence of LTS during BWR service has also been included. (author). 10 refs., 3 figs

  17. Influence of alloy elements on physical and mechanical properties of single crystalline austenitic stainless steels

    International Nuclear Information System (INIS)

    Okamoto, Kazutaka; Kaneda, Junya; Yoshinari, Akira; Aono, Yasuhisa

    2000-01-01

    The single crystalline austenitic stainless steels based on 316 L were developed to improve their resistance to intergranular corrosion and stress corrosion cracking. However the mechanical properties of the single crystals were lower than those of polycrystalline. The precipitation hardening methods were applied to the single crystal for the increase of their mechanical strength by addition of niobium and heat treatments. In this paper, the influences of niobium concentration on the several physical and mechanical properties of these single crystalline austenitic stainless steels were studied. The thermal conductivity, coefficients of thermal expansion and elastic constants of the single crystals were almost the same as those of polycrystalline independently of the niobium concentration. The mechanical properties of the single crystals strongly depended on the niobium concentration and the orientation. In the specimen which contains 1.0 mass% niobium, 0.2% proof stress were remarkably improved; 370 MPa, 337 MPa and 403 MPa were obtained in , and orientations at the room temperature. The creep rupture strength and the high cycle fatigue strength were also improved by addition of niobium. In the -orientated specimen which contains 1.0 mass% niobium, the creep rupture strength at 873 K for 103 hours, 245 MPa and the high cycle fatigue strength at 773 K for 107 cycles, 220 MPa were obtained. Furthermore, the single crystalline pipe, bolts and nuts were successfully manufactured for the application of these single crystals. (author)

  18. Heat pipe dynamic behavior

    Science.gov (United States)

    Issacci, F.; Roche, G. L.; Klein, D. B.; Catton, I.

    1988-01-01

    The vapor flow in a heat pipe was mathematically modeled and the equations governing the transient behavior of the core were solved numerically. The modeled vapor flow is transient, axisymmetric (or two-dimensional) compressible viscous flow in a closed chamber. The two methods of solution are described. The more promising method failed (a mixed Galerkin finite difference method) whereas a more common finite difference method was successful. Preliminary results are presented showing that multi-dimensional flows need to be treated. A model of the liquid phase of a high temperature heat pipe was developed. The model is intended to be coupled to a vapor phase model for the complete solution of the heat pipe problem. The mathematical equations are formulated consistent with physical processes while allowing a computationally efficient solution. The model simulates time dependent characteristics of concern to the liquid phase including input phase change, output heat fluxes, liquid temperatures, container temperatures, liquid velocities, and liquid pressure. Preliminary results were obtained for two heat pipe startup cases. The heat pipe studied used lithium as the working fluid and an annular wick configuration. Recommendations for implementation based on the results obtained are presented. Experimental studies were initiated using a rectangular heat pipe. Both twin beam laser holography and laser Doppler anemometry were investigated. Preliminary experiments were completed and results are reported.

  19. Design analysis of liquid metal pipe supports

    International Nuclear Information System (INIS)

    Margolin, L.L.; LaSalle, F.R.

    1979-02-01

    Design guidelines pertinent to liquid metal pipe supports are presented. The numerous complex conditions affecting the support stiffness and strength are addressed in detail. Topics covered include modeling of supports for natural frequency and stiffness calculations, support hardware components, formulas for deflection due to torsion, plate bending, and out-of-plane flexibility. A sample analysis and a discussion on stress analysis of supports are included. Also presented are recommendations for design improvements for increasing the stiffness of pipe supports and which were utilized in the FFTF system

  20. Effect of smelting method on the austenite grain size and properties of heat-resisting pearlitic steel

    International Nuclear Information System (INIS)

    Balakhovskaya, M.B.; Khusainova, N.A.; Davlyatova, L.N.

    1975-01-01

    Influence of smelting method on austenite grain size and properties of refractory perlite steel were studied. An opportunity was found to increase the steel refractoriness without deteriorating its other properties. The steel 12Kh1MF of electric or common open-hearth smelting was used. The dependence of kinetics of austenite grain growth on the smelting method was studied in the temperature range 950 deg - 1200 deg C with 1 hour exposure. The grain size of austenite in steel is supposedly determined by aluminium nitrides and vanadium carbides. In tests of normalized (kept for 20 minutes at 950-980 deg C) and tempered (kept for 3 hours at 730 deg C) transverse (tangential) pipe cross-section samples the electric steel had higher impact viscosity than the open-hearth metal. At working temperatures (540 deg -580 deg C) the difference in viscosity has its minimum. Viscosity of both steels 12Kh1MF begins to sharply decrease from 20 deg C. However, electric steel has rather high viscosity even at - 40 deg C, while the open-hearth one becomes brittle as early as at - 20 deg C. Long-term strength tests at 580 deg C under stresses 10-14 kG/mm 2 show that the coarse-grain steel is more refractory, i.e. time till fracture of open-hearth steel samples is twice as long as that of electric steel samples

  1. Effect of smelting method on the austenite grain size and properties of heat-resisting pearlitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Balakhovskaya, M B; Khusainova, N A; Davlyatova, L N [Vsesoyuznyj Nauchno-Issledovatel' skij Teplotekhnicheskij Inst., Moscow (USSR)

    1975-12-01

    Influence of smelting method on austenite grain size and properties of refractory perlite steel were studied. An opportunity was found to increase the steel refractoriness without deteriorating its other properties. The steel 12Kh1MF of electric or common open-hearth smelting was used. The dependence of kinetics of austenite grain growth on the smelting method was studied in the temperature range 950 deg - 1200 deg C with 1 hour exposure. The grain size of austenite in steel is supposedly determined by aluminium nitrides and vanadium carbides. In tests of normalized (kept for 20 minutes at 950-980 deg C) and tempered (kept for 3 hours at 730 deg C) transverse (tangential) pipe cross-section samples the electric steel had higher impact viscosity than the open-hearth metal. At working temperatures (540 deg -580 deg C) the difference in viscosity has its minimum. Viscosity of both steels 12Kh1MF begins to sharply decrease from 20 deg C. However, electric steel has rather high viscosity even at /sup -/40 deg C, while the open-hearth one becomes brittle as early as at /sup -/20 deg C. Long-term strength tests at 580 deg C under stresses 10-14 kG/mm/sup 2/ show that the coarse-grain steel is more refractory, i.e. time till fracture of open-hearth steel samples is twice as long as that of electric steel samples.

  2. Fracture toughness of irradiated wrought and cast austenitic stainless steels in BWR environment

    International Nuclear Information System (INIS)

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

    2007-01-01

    Experimental data are presented on the fracture toughness of wrought and cast austenitic stainless steels (SSs) that were irradiated to a fluence of ∼ 1.5 x 10 21 n/cm 2 (E > 1 MeV) * (∼ 2.3 dpa) at 296-305 o C. To evaluate the possible effects of test environment and crack morphology on the fracture toughness of these steels, all tests were conducted in normal-water-chemistry boiling water reactor (BWR) environments at ∼ 289 o C. Companion tests were also conducted in air on the same material for comparison. The fracture toughness J-R curves for SS weld heat-affected-zone materials in BWR water were found to be comparable to those in air. However, the results of tests on sensitized Type 304 SS and thermally aged cast CF-8M steel suggested a possible effect of water environment. The available fracture toughness data on irradiated austenitic SSs were reviewed to assess the potential for radiation embrittlement of reactor-core internal components. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components are also discussed. (author)

  3. Seismic analysis of liquid metal reactor piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    To safely assess the adequacy of the LMR piping, a three-dimensional piping code, SHAPS, has been developed at Argonne National Laboratory. This code was initially intended for calculating hydrodynamic-wave propagation in a complex piping network. It has salient features for treating fluid transients of fluid-structure interactions for piping with in-line components. The code also provides excellent structural capabilities of computing stresses arising from internal pressurization and 3-D flexural motion of the piping system. As part of the development effort, the SHAPS code has been further augmented recently by introducing the capabilities of calculating piping response subjected to seismic excitations. This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis

  4. Comparative study of computational model for pipe whip analysis

    International Nuclear Information System (INIS)

    Koh, Sugoong; Lee, Young-Shin

    1993-01-01

    Many types of pipe whip restraints are installed to protect the structural components from the anticipated pipe whip phenomena of high energy lines in nuclear power plants. It is necessary to investigate these phenomena accurately in order to evaluate the acceptability of the pipe whip restraint design. Various research programs have been conducted in many countries to develop analytical methods and to verify the validity of the methods. In this study, various calculational models in ANSYS code and in ADLPIPE code, the general purpose finite element computer programs, were used to simulate the postulated pipe whips to obtain impact loads and the calculated results were compared with the specific experimental results from the sample pipe whip test for the U-shaped pipe whip restraints. Some calculational models, having the spring element between the pipe whip restraint and the pipe line, give reasonably good transient responses of the restraint forces compared with the experimental results, and could be useful in evaluating the acceptability of the pipe whip restraint design. (author)

  5. Microstructural evolution in deformed austenitic TWinning Induced Plasticity steels

    NARCIS (Netherlands)

    Van Tol, R.T.

    2014-01-01

    This thesis studies the effect of plastic deformation on the stability of the austenitic microstructure against martensitic transformation and diffusional decomposition and its role in the phenomenon of delayed fracture in austenitic manganese (Mn)-based TWinning Induced Plasticity (TWIP) steels.

  6. 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...

  7. Reverted austenite in PH 13-8 Mo maraging steels

    International Nuclear Information System (INIS)

    Schnitzer, Ronald; Radis, Rene; Noehrer, Matthias; Schober, Michael; Hochfellner, Rainer; Zinner, Silvia; Povoden-Karadeniz, E.; Kozeschnik, Ernst; Leitner, Harald

    2010-01-01

    The mechanical properties of maraging steels are strongly influenced by the presence of reverted austenite. In this study, the morphology and chemical composition of reverted austenite in a corrosion resistant maraging steel was characterized using transmission electron microscopy (TEM) and atom probe tomography (APT). Two types of austenite, i.e. granular and elongated, are present after aging at 575 o C, whereby the content of the latter increases during aging. The investigations revealed that the austenite phase is enriched in Ni, which prevents the transformation to martensite during cooling. Inside and next to the austenitc areas, Mo and Cr-rich carbides, which form during the aging treatment, were found. Various aging treatments were performed to obtain the activation energy for the formation of reverted austenite. Additionally, the experimental data are compared with thermodynamic and kinetic simulations. Based on these results and the chemical composition changes of the phases, a model for the formation of reverted austenite is presented. It is concluded that precipitation of B2-ordered NiAl and formation of reverted austenite take place simultaneously during aging and that dissolution of precipitates is not essential for the initial formation of reverted austenite.

  8. Probabilistic evaluation of main coolant pipe break indirectly induced by earthquakes Savannah River Project L and P Reactors

    International Nuclear Information System (INIS)

    Short, S.A.; Wesley, D.A.; Awadalla, N.G.; Kennedy, R.P.

    1989-01-01

    A probabilistic evaluation of seismically-induced indirect pipe break for the Savannah River Project (SRP) L- and P-Reactor main coolant (process water) piping has been conducted. Seismically-induced indirect pipe break can result primarily from: (1) failure of the anchorage of one or more of the components to which the pipe is anchored; or (2) failure of the pipe due to collapse of the structure. the potential for both types of seismically-induced indirect failures was identified during a seismic walkdown of the main coolant piping. This work involved: (1) identifying components or structures whose failure could result in pipe failure; (2) developing seismic capacities or fragilities of these components; (3) combining component fragilities to develop plant damage state fragilities; and (4) convolving the plant seismic fragilities with a probabilistic seismic hazard estimate for the site in order to obtain estimates of seismic risk in terms of annual probability of seismic-induced indirect pipe break

  9. Influence of laser shock peening on irradiation defects in austenitic stainless steels

    Science.gov (United States)

    Lu, Qiaofeng; Su, Qing; Wang, Fei; Zhang, Chenfei; Lu, Yongfeng; Nastasi, Michael; Cui, Bai

    2017-06-01

    The laser shock peening process can generate a dislocation network, stacking faults, and deformation twins in the near surface of austenitic stainless steels by the interaction of laser-driven shock waves with metals. In-situ transmission electron microscopy (TEM) irradiation studies suggest that these dislocations and incoherent twin boundaries can serve as effective sinks for the annihilation of irradiation defects. As a result, the irradiation resistance is improved as the density of irradiation defects in laser-peened stainless steels is much lower than that in untreated steels. After heating to 300 °C, a portion of the dislocations and stacking faults are annealed out while the deformation twins remain stable, which still provides improved irradiation resistance. These findings have important implications on the role of laser shock peening on the lifetime extension of austenitic stainless steel components in nuclear reactor environments.

  10. A review of compatibility of IFR fuel and austenitic stainless steel

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.

    1996-01-01

    Interdiffusion experiments have been conducted to investigate the compatibility of various austenitic stainless steels with U-Pu-Zr alloys, which are alloys to be employed as fuel for the Integral Fast Reactor being developed by Argonne National Laboratory. These tests have also studied the compatibility of austenitic stainless steels with fission products, like the minor actinides (Np and Am) and lanthanides (Ce and Nd), that are generated during the fission process in an IFR. This paper compares the results of these investigations in the context of fuel-cladding compatibility in IFR fuel elements, specifically focusing on the relative Interdiffusion behavior of the components and the types of phases that develop based on binary phase diagrams. Results of Interdiffusion tests are assessed in the light of observations derived from post-test examinations of actual irradiated fuel elements

  11. The effect of variations in carbon activity on the carburization of austenitic steels in sodium

    International Nuclear Information System (INIS)

    Gwyther, J.R.; Hobdell, M.R.; Hooper, A.J.

    1978-07-01

    Experience has shown that the liquid sodium coolant of fast breeder reactors is an effective carbon-transport medium; the resulting carburization of thin austenitic stainless steel components (eg IHX and fuel cladding) could adversely affect their mechanical integrity. The degree and nature of steel carburization depend, inter alia, on the carbon activity of the sodium environment. Exploratory tests are described in which specimens of austenitic stainless steel were carburized in sodium, the carbon activity of which was continuously monitored by a BNL electrochemical carbon meter. The sodium carbon activity was initially high, but decreased with time, simulating conditions equivalent to plant start-up or coolant clean-up following accidental oil ingress. The extent and nature of steel carburization was identified by metallography, electron microscopy, X-ray crystallography and chemical analysis. (author)

  12. 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.

  13. Phase transformation by fatigue in austenitic stainless steel

    International Nuclear Information System (INIS)

    Jo, Y.S.; Kwun, S.I.

    1988-01-01

    The effect of strain induced martensite on the fatigue behavior of AISI 304 austenitic stainless steel was investigated. During low cycle fatigue, the austenitic stainless steel showed a continuous cyclic hardening until fracture. The extent of cyclic hardening increased with decreasing austenite stability. The austenite stability was controlled by different aging time and temperature, which resulted in different carbide morphologies. The fatigue crack propagation rate near ΔK th varied also with the austenite stability inside the plastic zone at the crack up. Especially, the near-threshold fatigue crack propagation rate of the grain boundary carbide precipitated condition was the lowest. This was considered to be due to the roughness induced closure caused by intergranular facet. A new model for the intergranular facet formation and the fatigue crack propagation of grain boundary carbide precipitated condition was proposed. (Author)

  14. Development of solutions to benchmark piping problems

    Energy Technology Data Exchange (ETDEWEB)

    Reich, M; Chang, T Y; Prachuktam, S; Hartzman, M

    1977-12-01

    Benchmark problems and their solutions are presented. The problems consist in calculating the static and dynamic response of selected piping structures subjected to a variety of loading conditions. The structures range from simple pipe geometries to a representative full scale primary nuclear piping system, which includes the various components and their supports. These structures are assumed to behave in a linear elastic fashion only, i.e., they experience small deformations and small displacements with no existing gaps, and remain elastic through their entire response. The solutions were obtained by using the program EPIPE, which is a modification of the widely available program SAP IV. A brief outline of the theoretical background of this program and its verification is also included.

  15. Structural materials for ITER in-vessel component design

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

  16. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Doyen, J.J.; Lebey, J.

    1979-01-01

    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 280 0 C, 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 = 280 0 C). 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

  17. Damping Capacity of High Manganese Austenitic Stainless Steel with a Two Phase Mixed Structure of Martensite and Austenite

    International Nuclear Information System (INIS)

    Hwang, Tae Hyun; Kang, Chang-Yong

    2013-01-01

    The damping capacity of high manganese austenitic stainless steel with a two phase mixed structure of deformation-induced martensite and reversed austenite was studied. Reversed austenite with an ultra-fine grain size of less than 0.2 μm was obtained by reversion treatment. The two phase structure of deformation-induced martensite and reversed austenite was obtained by annealing treatment at a range of 500-700 °C and various times in cold rolled high manganese austenitic stainless steel. The damping capacity increased with an increasing annealing temperature and time. In high manganese stainless steel with the two phase mixed structure of martensite and austenite, the damping capacity decreased with an increasing volume fraction of deformation-induced martensite. Thus, the damping capacity was strongly affected by deformation-induced martensite. The results confirmed that austenitic stainless steel with a good combination of strength and damping capacity was obtained from the two phase mixed structure of austenite and martensite.

  18. Insulating jacket for heat sensitive components

    International Nuclear Information System (INIS)

    Class, G.

    1980-01-01

    The insulating jacket for long core components of sodium-cooled reactors consists of several layers of austenitic steel, between which a woven wire mesh of the same material is fitted. It is wound in the form of a spiral bandage on the core component. (DG) [de

  19. Key quality aspects for a new metallic composite pipe: corrosion testing, welding, weld inspection and manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Conder, Robert J.; Felton, Peter [Xodus Group Ltd., Aberdeen (United Kingdom); Smith, Richard [Shell Global Solutions Inc., Houston, TX (United States); Burke, Raymond [Pipestream Inc., Houston, TX (United States); Dikstra, Frits; Deleye, Xavier [Applus RTD Ltd., Rotterdam (Netherlands)

    2010-07-01

    XPipeTM is a new metallic composite pipe. This paper discusses three aspects of this new technology. The first subject is determination of the probability of hydrogen embrittlement by the XPipeTM manufacturing method. Two materials were analyzed in three tests: slow strain rate test, constant load test and notched tensile test. The results showed that the high strength steels used do not appear to be susceptible to hydrogen embrittlement. The second subject of this article is weld inspection. A non-destructive testing method of girth welds is developed to allow inspection of the thin-walled austenitic liner pipe. The results demonstrated that the welds can be inspected using the creeping wave technique. The third subject is quality control systems using the SCADA system, which maintains traceability of the materials and monitors and records all parameters during the production process. This system appears to be efficient in ensuring that the product pipe meets recognized quality standards.

  20. Surface improvement for inside surface of small diameter pipes by laser cladding technique

    International Nuclear Information System (INIS)

    Irisawa, Toshio; Morishige, Norio; Umemoto, Tadahiro; Ono, Kazumichi; Hamaoka, Tadashi; Tanaka, Atsushi

    1991-01-01

    A laser cladding technique has been used for surface improvement in controlling the composition of a metal surface. Recent high power YAG laser development gives an opportunity to use this laser cladding technique for various applications. A YAG laser beam can be transmitted through an optical fiber for a long distance and through narrow spaces. YAG laser cladding was studied for developing alloy steel to prevent stress corrosion cracking in austenitic stainless steel piping. In order to make a cladding layer, mixed metal powder was on the inside surface of the piping using an organic binder. Subsequently the powder beds were melted with a YAG laser beam transmitted through an optical fiber. This paper introduces the Laser cladding technique for surface improvement for the inside surface of a small diameter pipe. (author)

  1. Acoustical holographic Siamese image technique for imaging radial cracks in reactor piping

    International Nuclear Information System (INIS)

    Collins, H.D.; Gribble, R.P.

    1985-04-01

    This paper describes a unique technique (i.e., ''Siamese imaging'') for imaging quasi-vertical defects in reactor pipe weldments. The Siamese image is a bi-symmetrical view of the inner surface defect. Image construction geometry consists of two probes (i.e., source/receiver) operating either from opposite sides or the same side of the defect to be imaged. As the probes are scanned across a lower surface connected defect, they encounter two images - first the normal upright image and then the inverted image. The final integrated image consists of two images connected along their baselines, thus we call it a ''Siamese image.'' The experimental imaging results on simulated and natural cracks in reactor piping weldments graphically illustrate this unique technique. Excellent images of mechanical fatique and thermal cracks were obtained on ferritic and austenitic piping

  2. Austenite strengthening and softening during hot deformation

    International Nuclear Information System (INIS)

    Tushinskij, L.I.; Vlasov, V.S.; Kazimirova, I.E.; Tokarev, A.O.

    1981-01-01

    Processes of formation of austenite structure of 20 and 12Kh18N10T steels during hot deformation and postdeformation isothermal holdings have been investigated by the methods of analysis of curves of hot deformation, high-temperature metallography and light microscopy. Deformation has been exercised by extention in vacuum with average 4x10 -2 s -1 rate. Deformation temperatures of steel 20 are 930 and 1000 deg C, of steel 12Kh18N10T - 1100 deg C. It is stated that dynamic recrystallization takes place in both investigated steels during hot deformation. In the carbonic steel it is developed by shifting sections of high-angular boundaries, flow stress in this case remains constant. Recrystallization is developed by subgrain coalescence in austenite steel, that brings about preservation of increased defect density in recrystallized volumes. As a result strengthening of steel is continued up to fracture during the increase of the deformation degree. Postdeformation weakening of 12Kh18N10T steel is slowed down as compared with weakening of carbonic steel [ru

  3. Magnetic properties of cyclically deformed austenite

    Energy Technology Data Exchange (ETDEWEB)

    Das, Arpan, E-mail: dasarpan1@yahoo.co.in

    2014-06-01

    In meta-stable austenitic stainless steels, low cycle fatigue deformation is accompanied by a partial stress/strain-induced solid state phase transformation of paramagnetic γ(fcc) austenite phase to ferromagnetic α{sup /}(bcc) martensite. The measured characteristic of magnetic properties, which are the saturation magnetization, susceptibility, coercivity, retentivity, and the area under the magnetic hysteresis loop are sensitive to the total strain amplitude imposed and the corresponding material behaviour. The morphologies and nucleation characteristics of deformation induced martensites (i.e., ϵ(hcp), α{sup /}(bcc)) have been investigated through analytical transmission electron microscope. It has been observed that deformation induced martensites can nucleate at a number of sites (i.e., shear band intersections, isolated shear bands, shear band–grain boundary intersection, grain boundary triple points, etc.) through multiple transformation sequences: γ(fcc)→ϵ(hcp), γ(fcc)→ϵ(hcp)→α{sup /}(bcc), γ(fcc)→ deformation twin →α{sup /}(bcc) and γ(fcc)→α{sup /}(bcc). - Highlights: • LCF tests were done at various strain amplitudes of 304LNSS. • Quantification of martensite was done through ferritecope. • Magnetic properties were characterised through VSM. • Correlation of magnetic properties with the cyclic plastic response was done. • TEM was done to investigate the transformation micro-mechanisms.

  4. Degradation of austenitic stainless steel (SS) light water ractor (LWR) core internals due to neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Appajosula S., E-mail: Appajosula.Rao@nrc.gov

    2014-04-01

    Austenitic stainless steels (SSs) are extensively being used in the fabrication of light water reactor (LWR) core internal components. It is because these steels have relatively high ductility, fracture toughness and moderate strength. However, the LWR internal components exposure to neutron irradiation over an extended period of plant operation degrades the materials mechanical properties such as the fracture toughness. This paper summarizes some of the results of the existing open literature data on irradiation assisted stress corrosion cracking (IASCC) of 316 CW steels that have been published by the United States Nuclear Regulatory Commission (USNRC), industry, academia, and other research agencies.

  5. Electron backscatter and X-ray diffraction studies on the deformation and annealing textures of austenitic stainless steel 310S

    Energy Technology Data Exchange (ETDEWEB)

    Nezakat, Majid, E-mail: majid.nezakat@usask.ca [Canadian Light Source Inc., 44 Innovation Boulevard, Saskatoon, SK, S7N 2V3 (Canada); Akhiani, Hamed [Westpower Equipment Ltd., 4451 54 Avenue South East, Calgary, AB T2C 2A2 (Canada); Sabet, Seyed Morteza [Department of Ocean and Mechanical Engineering, Florida Atlantic University, Boca Raton, FL 33431 (United States); Szpunar, Jerzy [Department of Mechanical Engineering, University of Saskatchewan, 57 Campus Drive, Saskatoon, SK, S7N 5A9 (Canada)

    2017-01-15

    We studied the texture evolution of thermo-mechanically processed austenitic stainless steel 310S. This alloy was cold rolled up to 90% reduction in thickness and subsequently annealed at 1050 °C. At the early stages of deformation, strain-induced martensite was formed from deformed austenite. By increasing the deformation level, slip mechanism was found to be insufficient to accommodate higher deformation strains. Our results demonstrated that twinning is the dominant deformation mechanism at higher deformation levels. Results also showed that cold rolling in unidirectional and cross rolling modes results in Goss/Brass and Brass dominant textures in deformed samples, respectively. Similar texture components are observed after annealing. Thus, the annealing texture was greatly affected by texture of the deformed parent phase and martensite did not contribute as it showed an athermal reversion during annealing. Results also showed that when the fraction of martensite exceeds a critical point, its grain boundaries impeded the movement of austenite grain boundaries during annealing. As a result, recrystallization incubation time would increase. This caused an incomplete recrystallization of highly deformed samples, which led to a rational drop in the intensity of the texture components. - Highlights: •Thermo-mechanical processing through different cold rolling modes can induce different textures. •Martensite reversion is athermal during annealing. •Higher fraction of deformation-induced martensite can increase the annealing time required for complete recrystallization. •Annealing texture is mainly influenced by the deformation texture of austenite.

  6. Characterization of the austenitic stability of metastable austenitic stainless steel with regard to its formability

    Science.gov (United States)

    Schneider, Matthias; Liewald, Mathias

    2018-05-01

    During the last decade, the stainless steel market showed a growing volume of 3-5% p.a.. The austenitic grades are losing market shares to ferritic or 200-series grades due to the high nickel price, but still playing the most important role within the stainless steel market. Austenitic stainless steel is characterized by the strain-induced martensite formation, causing the TRIP-effect (Transformation Induced Plasticity) which is responsible for good formability and high strength. The TRIP-effect itself is highly dependent on the forming temperature, the strain as well as the chemical composition which has a direct influence on the stability of the austenite. Today the austenitic stability is usually characterized by the so called Md30-temperature, which was introduced by Angel and enhanced by several researches, particularly Nohara. It is an empirical formula based on the chemical composition and the grain size of a given material, calculating the temperature which is necessary to gain a 50 % martensite formation after 30 % of elongation in a tensile test. A higher Md30-temperature indicates a lower stability and therefore a higher tendency towards martensite formation. The main disadvantage of Md30 -temperature is the fact that it is not based on forming parameters and only describes a single point instead of the whole forming process. In this paper, an experimental set up for measuring martensite and temperature evolution in a non-isothermal tensile test is presented, which is based on works of Hänsel and Schmid. With this set up, the martensite formation rate for different steels of the steel grade EN 1.4301 and EN 1.4310 is measured. Based on these results a new austenitic stability criterion is defined. This criterion and the determined Md30-temperatures are related to the stretch formability of the materials. The results show that the new IFU criterion is with regard to the formability a much more useful characteristic number for metastable austenitic steels

  7. Characterization of radioactive contamination inside pipes with the Pipe Explorer{trademark} system. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Cremer, C.D.; Kendrick, D.T.; Lowry, W.; Cramer, E.

    1997-09-30

    The Department of Energy (DOE) is currently in the process of decommissioning and dismantling many of its nuclear materials processing facilities that have been in use for several decades. Site managers throughout the DOE complex must employ the safest and most cost effective means to characterize, remediate and recycle or dispose of hundreds of miles of potentially contaminated piping and duct work. The DOE discovered that standard characterization methods were inadequate for its pipes, drains, and ducts because many of the systems are buried or encased. In response to the DOE`s need for a more specialized characterization technique, Science and Engineering Associates, Inc. (SEA) developed the Pipe Explorer{trademark} system through a DOE Office of Science and Technology (OST) contract administered through the Federal Energy Technology Center (FETC). The purpose of this report is to serve as a comprehensive overview of all phases of the Pipe Explorer{trademark} development project. The report is divided into 6 sections. Section 2 of the report provides an overview of the Pipe Explorer{trademark} system, including the operating principles of using an inverting membrane to tow sensors into pipes. The basic components of the characterization system are also described. Descriptions of the various deployment systems are given in Section 3 along with descriptions of the capabilities of the deployment systems. During the course of the development project 7 types of survey instruments were demonstrated with the Pipe Explorer{trademark} and are a part of the basic toolbox of instruments available for use with the system. These survey tools are described in Section 4 along with their typical performance specifications. The 4 demonstrations of the system are described chronologically in Section 5. The report concludes with a summary of the history, status, and future of the Pipe Explorer{trademark} system in Section 6.

  8. Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system. Final report

    International Nuclear Information System (INIS)

    Cremer, C.D.; Kendrick, D.T.; Lowry, W.; Cramer, E.

    1997-01-01

    The Department of Energy (DOE) is currently in the process of decommissioning and dismantling many of its nuclear materials processing facilities that have been in use for several decades. Site managers throughout the DOE complex must employ the safest and most cost effective means to characterize, remediate and recycle or dispose of hundreds of miles of potentially contaminated piping and duct work. The DOE discovered that standard characterization methods were inadequate for its pipes, drains, and ducts because many of the systems are buried or encased. In response to the DOE's need for a more specialized characterization technique, Science and Engineering Associates, Inc. (SEA) developed the Pipe Explorer trademark system through a DOE Office of Science and Technology (OST) contract administered through the Federal Energy Technology Center (FETC). The purpose of this report is to serve as a comprehensive overview of all phases of the Pipe Explorer trademark development project. The report is divided into 6 sections. Section 2 of the report provides an overview of the Pipe Explorer trademark system, including the operating principles of using an inverting membrane to tow sensors into pipes. The basic components of the characterization system are also described. Descriptions of the various deployment systems are given in Section 3 along with descriptions of the capabilities of the deployment systems. During the course of the development project 7 types of survey instruments were demonstrated with the Pipe Explorer trademark and are a part of the basic toolbox of instruments available for use with the system. These survey tools are described in Section 4 along with their typical performance specifications. The 4 demonstrations of the system are described chronologically in Section 5. The report concludes with a summary of the history, status, and future of the Pipe Explorer trademark system in Section 6

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

    International Nuclear Information System (INIS)

    Xue Fei; Yu Weiwei; Wang Zhaoxi; Ma Qinzheng; Liu Wei

    2012-01-01

    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)

  10. Crack Resistance of Welded Joints of Pipe Steels of Strength Class K60 of Different Alloying Systems

    Science.gov (United States)

    Tabatchikova, T. I.; Tereshchenko, N. A.; Yakovleva, I. L.; Makovetskii, A. N.; Shander, S. V.

    2018-03-01

    The crack resistance of welded joints of pipe steels of strength class K60 and different alloying systems is studied. The parameter of the crack tip opening displacement (CTOD) is shown to be dependent on the size of the austenite grains and on the morphology of bainite in the superheated region of the heat-affected zone of the weld. The crack resistance is shown to be controllable due to optimization of the alloying system.

  11. High cycle fatigue of austenitic stainless steels under random loading

    International Nuclear Information System (INIS)

    Gauthier, J.P.; Petrequin, P.

    1987-08-01

    To investigate reactor components, load control random fatigue tests were performed at 300 0 C and 550 0 C, on specimens from austenitic stainless steels plates in the transverse orientation. Random solicitations are produced on closed loop servo-hydraulic machines by a mini computer which generates random load sequence by the use of reduced Markovian matrix. The method has the advantage of taking into account the mean load for each cycle. The solicitations generated are those of a stationary gaussian process. Fatigue tests have been mainly performed in the endurance region of fatigue curve, with scattering determination using stair case method. Experimental results have been analysed aiming at determining design curves for components calculations, depending on irregularity factor and temperature. Analysis in term of mean square root fatigue limit calculation, shows that random loading gives more damage than constant amplitude loading. Damage calculations following Miner rule have been made using the probability density function for the case where the irregularity factor is nearest to 100 %. The Miner rule is too conservative for our results. A method using design curves including random loading effects with irregularity factor as an indexing parameter is proposed

  12. 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

  13. Response of the primary piping loop to an HCDA

    International Nuclear Information System (INIS)

    Chang, Y.W.; Moneim, M.T.A.; Wang, C.Y.; Gvildys, J.

    1975-01-01

    The paper describes a method for analyzing the response of the primary piping loop that consists of straight pipes, elbows, and other components connected in series and subject to hypothetical core disruptive accident (HCDA) loads at both ends of the loop. The complete hydrodynamic equations in two-dimensions, that include both the nonlinear convective and viscous dissipation terms are used for the fluid dynamics together with the implicit ICE technique. The external walls of the pipes and components are treated as thin shells in which the analysis accounts for the membrane and bending strength of the wall, elastic-plastic material behavior, as well as large deformation under the effect of transient loading conditions. In the straight pipes, the flow is assumed to be axisymmetric; in the elbow regions, the two dimensions considered are the r and theta directions. The flow in the other components is also assumed to be axisymmetric; the components are modeled as a circular cylinder, in which the radius of the cylinder can be varied to conform with the outside shape of the component and the flow area inside can be changed independently from the outside shape. However, they must remain axially symmetric. The method is applied to a piping loop which consists of six elastic-plastic pipes and five rigid elbows connected in series and subjected to pressure pulses at both ends of the loop

  14. Calculation of dynamic hydraulic forces in nuclear plant piping systems

    International Nuclear Information System (INIS)

    Choi, D.K.

    1982-01-01

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

  15. Qualitative and Quantitative Control of Wastewater Dual Wall Polyethylene Pipes

    Directory of Open Access Journals (Sweden)

    Mohammad Reza Salimi

    2008-09-01

    Full Text Available Pipes are the most important components of wastewater collection systems accounting for considerable costs in constructing such systems. In view of this and regarding the growing trend in design and execution of wastewater collection and transmission lines in recent years, various types of pipes have been introduced into the market. Selection of appropriate pipes and their qualitative and quantitative control, therefore, call for due consideration given their high cost share in collection systems. In this paper, efforts are made to consider various types of pipes used in (urban and rural wastewater collection networks in an attempt to signal the significance of qualitative and quantitative control of different dual wall polyethylene pipes used as sewers. Finally, the relevant issues regarding the methods and conditions for technical control and inspection of polyethylene sewer lines during construction and operation stages are provided.

  16. Heat-pipe Earth.

    Science.gov (United States)

    Moore, William B; Webb, A Alexander G

    2013-09-26

    The heat transport and lithospheric dynamics of early Earth are currently explained by plate tectonic and vertical tectonic models, but these do not offer a global synthesis consistent with the geologic record. Here we use numerical simulations and comparison with the geologic record to explore a heat-pipe model in which volcanism dominates surface heat transport. These simulations indicate that a cold and thick lithosphere developed as a result of frequent volcanic eruptions that advected surface materials downwards. Declining heat sources over time led to an abrupt transition to plate tectonics. Consistent with model predictions, the geologic record shows rapid volcanic resurfacing, contractional deformation, a low geothermal gradient across the bulk of the lithosphere and a rapid decrease in heat-pipe volcanism after initiation of plate tectonics. The heat-pipe Earth model therefore offers a coherent geodynamic framework in which to explore the evolution of our planet before the onset of plate tectonics.

  17. Pipe whip: a summary of the damage observed in BNL pipe-on-pipe impact tests

    International Nuclear Information System (INIS)

    Baum, M.R.

    1987-01-01

    This paper describes examples of the damage resulting from the impact of a whipping pipe on a nearby pressurised pipe. The work is a by-product of a study of the motion of a whipping pipe. The tests were conducted with small-diameter pipes mounted in rigid supports and hence the results are not directly applicable to large-scale plant applications where flexible support mountings are employed. The results illustrate the influence of whipping pipe energy, impact position and support type on the damage sustained by the target pipe. (author)

  18. Environmental Assisted Fatigue Evaluation of Direct Vessel Injection Piping Considering Thermal Stratification

    International Nuclear Information System (INIS)

    Kim, Taesoon; Lee, Dohwan

    2016-01-01

    As the environmentally assisted fatigue (EAF) due to the primary water conditions is to be a critical issue, the fatigue evaluation for the components and pipes exposed to light water reactor coolant conditions has become increasingly important. Therefore, many studies to evaluate the fatigue life of the components and pipes in LWR coolant environments on fatigue life of materials have been conducted. Among many components and pipes of nuclear power plants, the direct vessel injection piping is known to one of the most vulnerable pipe systems because of thermal stratification occurred in that systems. Thermal stratification occurs because the density of water changes significantly with temperature. In this study, fatigue analysis for DVI piping using finite element analysis has been conducted and those results showed that the results met design conditions related with the environmental fatigue evaluation of safety class 1 pipes in nuclear power plants. Structural and fatigue integrity for the DVI piping system that thermal stratification occurred during the plant operation has conducted. First of all, thermal distribution of the piping system is calculated by computational fluid dynamic analysis to analyze the structural integrity of that piping system. And the fatigue life evaluation considering environmental effects was carried out. Our results showed that the DVI piping system had enough structural integrity and fatigue life during the design lifetime of 60 years

  19. 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.

  20. Studies on Stress Corrosion Cracking of Super 304H Austenitic Stainless Steel

    Science.gov (United States)

    Prabha, B.; Sundaramoorthy, P.; Suresh, S.; Manimozhi, S.; Ravishankar, B.

    2009-12-01

    Stress corrosion cracking (SCC) is a common mode of failure encountered in boiler components especially in austenitic stainless steel tubes at high temperature and in chloride-rich water environment. Recently, a new type of austenitic stainless steels called Super304H stainless steel, containing 3% copper is being adopted for super critical boiler applications. The SCC behavior of this Super 304H stainless steel has not been widely reported in the literature. Many researchers have studied the SCC behavior of steels as per various standards. Among them, the ASTM standard G36 has been widely used for evaluation of SCC behavior of stainless steels. In this present work, the SCC behavior of austenitic Fe-Cr-Mn-Cu-N stainless steel, subjected to chloride environments at varying strain conditions as per ASTM standard G36 has been studied. The environments employed boiling solution of 45 wt.% of MgCl2 at 155 °C, for various strain conditions. The study reveals that the crack width increases with increase in strain level in Super 304H stainless steels.

  1. 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

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

    International Nuclear Information System (INIS)

    Maekawa, Akira; Noda, Michiyasu; Suzuki, Michiaki

    2009-01-01

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

  3. On the behavior of pressurized pipings under excessive-stresses caused by earthquake loadings

    International Nuclear Information System (INIS)

    Udoguchi, Y.; Akino, K.; Shibata, H.

    1975-01-01

    Five types of breaking experiments on pipe elements and piping structures had been carried out from 1971 to 1973 by the technical sub-committee of the Japan Electric Association under the leadership taken by Y. Udoguchi, one of the authors. One of the fruitful results was to realize the guillotine-type rupture of pipe element on a shaking table. However, it was also shown that the margin for the design is enough, and allowable stresses under earthquake loading are obtained by modifying those of the Emergency Condition of the ASME Code. The experiments effected were as follows: straight pipe elements, curved pipes and T-branch pipe connections, made of both ferritic and austenitic steels, were subjected to repeated bending moment, torsional moment and combined under pressurized condition. The pressure corresponded to their design value, but the stresses caused by such moments exceeded over their allowable stress of the Faulted Condition of the ASME Code. The wave patterns were both sinusoidal and natural earthquake records

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

    International Nuclear Information System (INIS)

    Bartholome, G.; Wellein, R.

    1992-01-01

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

  5. Pipe-to-pipe impact tests

    Energy Technology Data Exchange (ETDEWEB)

    Bampton, M C.C.; Alzheimer, J M; Friley, J R; Simonen, F A

    1985-11-01

    Existing licensing criteria express what damage shall be assumed for various pipe sizes as a consequence of a postulated break in a high energy system. The criteria are contained in Section 3.6.2 of the Standard Review Plan, and the purpose of the program described with this paper is to evaluate the impact criteria by means of a combined experimental and analytical approach. A series of tests has been completed. Evaluation of the test showed a deficiency in the range of test parameters. These deficiencies are being remedied by a second series of tests and a more powerful impact machine. A parallel analysis capability has been developed. This capability has been used to predict the damage for the first test series. The quality of predictions has been improved by tests that establish post-crush and bending relationships. Two outputs are expected from this project: data that may, or may not, necessitate changes to the criteria after appropriate value impact evaluations and an analytic capability for rapidly evaluating the potential for pipe whip damage after a postulated break. These outputs are to be contained in a value-impact document and a program final report. (orig.).

  6. Performance of buried pipe installation.

    Science.gov (United States)

    2010-05-01

    The purpose of this study is to determine the effects of geometric and mechanical parameters : characterizing the soil structure interaction developed in a buried pipe installation located under : roads/highways. The drainage pipes or culverts instal...

  7. Development of Ultra-Fine-Grained Structure in AISI 321 Austenitic Stainless Steel

    Science.gov (United States)

    Tiamiyu, A. A.; Szpunar, J. A.; Odeshi, A. G.; Oguocha, I.; Eskandari, M.

    2017-12-01

    Ultra-fine-grained (UFG) structure was developed in AISI 321 austenitic stainless steel (ASS) using cryogenic rolling followed by annealing treatments at 923 K, 973 K, 1023 K, and 1073 K (650 °C, 700 °C, 750 °C, and 800 °C) for different lengths of time. The α'-martensite to γ-austenite reversion behavior and the associated texture development were analyzed in the cryo-rolled specimens after annealing. The activation energy, Q, required for the reversion of α'-martensite to γ-austenite in the steel was estimated to be 80 kJ mol-1. TiC precipitates and unreversed triple junction α'-martensite played major roles in the development of UFG structure through the Zener pinning of grain boundaries. The optimum annealing temperature and time for the development of UFG structure in the cryo-rolled AISI 321 steel are (a) 923 K (650 °C) for approximately 28800 seconds and (b) 1023 K (750 °C) for 600 seconds, with average grain sizes of 0.22 and 0.31 µm, respectively. Annealing at 1023 K (750 °C) is considered a better alternative since the volume fraction of precipitated carbides in specimens annealed at 1023 K (750 °C) are less than those annealed at 923 K (650 °C). More so, the energy consumption during prolonged annealing time to achieve an UFG structure at 923 K (650 °C) is higher due to low phase reversion rate. The hardness of the UFG specimens is 195 pct greater than that of the as-received steel. The higher volume fraction of TiC precipitates in the UFG structure may be an additional source of hardening. Micro and macrotexture analysis indicated {110}〈uvw〉 as the major texture component of the austenite grains in the UFG structure. Its intensity is stronger in the specimen annealed at low temperatures.

  8. Resonant creep enhancement in austenitic stainless steels due to pulsed irradiation at low doses

    International Nuclear Information System (INIS)

    Kishimoto, N.; Amekura, H.; Saito, T.

    1994-01-01

    Steady-state irradiation creep of austenitic stainless steels has been extensively studied as one of the most important design parameters in fusion reactors. The steady-state irradiation creep has been evaluated using in-pile and light-ion experiments. Those creep compliances of various austenitic steels range in the vicinity of ε/Gσ = 10 -6 ∼10 -5 (dpa sm-bullet MPa) -1 , depending on chemical composition etc. The mechanism of steady-state irradiation creep has been elucidated, essentially in terms of stress-induced preferential absorption of point defects into dislocations, and their climb motion. From this standpoint, low doses such as 10 -3 ∼10 -1 dpa would not give rise to any serious creep, and the irradiation creep may not be a critical issue for the low-dose fusion devices including ITER. It is, however, possible that pulsed irradiation causes different creep behaviors from the steady-state one due to dynamic unbalance of interstitials and vacancies. The authors have actually observed anomalous creep enhancement due to pulsed irradiation in austenitic stainless steels. The resonant behavior of creep indicates that pulsed irradiation may cause significant deformation in austenitic steels even at such low doses and slow pulsing rates, especially for the SA-materials. The first-wall materials in plasma operation of ∼10 2 s may suffer from unexpected transient creep, even in the near-term fusion deices, such as ITER. Though this effect might be a transient effect for a relatively short period, it should be taken into account that the pulsed irradiation makes influences on stress relaxation of the fusion components and on the irradiation fatigue. The mechanism and the relevant behaviors of pulse-induced creep will be discussed in terms of a point-defect model based on the resonant interstitial enrichment

  9. Deformation behavior of austenitic stainless steel at deep cryogenic temperatures

    Science.gov (United States)

    Han, Wentuo; Liu, Yuchen; Wan, Farong; Liu, Pingping; Yi, Xiaoou; Zhan, Qian; Morrall, Daniel; Ohnuki, Somei

    2018-06-01

    The nonmagnetic austenite steels are the jacket materials for low-temperature superconductors of fusion reactors. The present work provides evidences that austenites transform to magnetic martensite when deformation with a high-strain is imposed at 77 K and 4.2 K. The 4.2 K test is characterized by serrated yielding that is related to the specific motion of dislocations and phase transformations. The in-situ transmission electron microscope (TEM) observations in nanoscale reveal that austenites achieve deformation by twinning under low-strain conditions at deep cryogenic temperatures. The generations of twins, martensitic transformations, and serrated yielding are in order of increasing difficulty.

  10. Phase 2 of the International Piping Integrity Research Group programme

    International Nuclear Information System (INIS)

    Darlaston, B.J.

    1994-01-01

    The results of phase 1 of the International Piping Integrity Research Group (IPIRG-1) programme have been widely reported. The significance of the results is reviewed briefly, in order to put the phase 2 programme into perspective. The success of phase 1 led the participants to consider further development and validation of pipe and pipe component fracture analysis technology as part of another international group programme (IPIRG-2). The benefits of combined funding and of the technical exchanges and interactions are considered to be of significant advantage and value. The phase 2 programme has been designed with the overall objective of developing and experimentally validating methods of predicting the fracture behaviour of nuclear reactor safety-related piping, to both normal operating and accident loads. The programme will add to the engineering estimation analysis methods that have been developed for straight pipes. The pipe system tests will expand the database to include seismic loadings and flaws in fittings, such as bends, elbows and tees, as well as ''short'' cracks. The results will be used to validate further the analytical methods, expand the capability to make fittings and extend the quasi-static results for the USNRC's new programme on short cracks in piping and piping welds. The IPIRG-2 programme is described to provide a clear understanding of the content, strategy, potential benefits and likely significance of the work. ((orig.))

  11. OPDE-The international pipe failure data exchange project

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, Bengt [OPDE Clearinghouse, 16917 S. Orchid Flower Trail, Vail, AZ 85641-2701 (United States)], E-mail: boylydell@msn.com; Riznic, Jovica [Canadian Nuclear Safety Commission, Operational Engineering Assessment Division, PO Box 1046, Station B, Ottawa, Ont. K1P 5S9 (Canada)], E-mail: jovica.riznic@cnsc-ccsn.gc.ca

    2008-08-15

    Certain member countries of the Organization for Economic Cooperation and development (OECD) in 2002 established the OECD pipe failure data exchange project (OPDE) to produce an international database on the piping service experience applicable to commercial nuclear power plants. OPDE is operated under the umbrella of the OECD Nuclear Energy Agency (NEA). The Project collects pipe failure data including service-induced wall thinning, part through-wall crack, pinhole leak, leak, and rupture/severance (i.e., events involving large through-wall flow rates up to and beyond the make-up capacity of engineered safeguards systems). The part through-wall events include degradation in excess of design code allowable for pipe wall thinning or crack depth. OPDE also addresses such degradation that could have generic implications regarding the reliability of in-service inspection. Currently the OPDE database includes approximately 3,700 records on pipe failure affecting ASME Code Class 1 through 3 and non-safety-related (non-Code) piping. This paper presents the motivations and objectives behind the establishment of the OPDE project. The paper also summarizes the unique data quality considerations that are associated with the reporting and recording of piping component degradation and failure. An overview of the database content is included to place it in perspective relative to past efforts to systematically collect and evaluate service experience data on piping performance. Finally, a brief summary is given of current database application studies.

  12. OPDE-The international pipe failure data exchange project

    International Nuclear Information System (INIS)

    Lydell, Bengt; Riznic, Jovica

    2008-01-01

    Certain member countries of the Organization for Economic Cooperation and development (OECD) in 2002 established the OECD pipe failure data exchange project (OPDE) to produce an international database on the piping service experience applicable to commercial nuclear power plants. OPDE is operated under the umbrella of the OECD Nuclear Energy Agency (NEA). The Project collects pipe failure data including service-induced wall thinning, part through-wall crack, pinhole leak, leak, and rupture/severance (i.e., events involving large through-wall flow rates up to and beyond the make-up capacity of engineered safeguards systems). The part through-wall events include degradation in excess of design code allowable for pipe wall thinning or crack depth. OPDE also addresses such degradation that could have generic implications regarding the reliability of in-service inspection. Currently the OPDE database includes approximately 3,700 records on pipe failure affecting ASME Code Class 1 through 3 and non-safety-related (non-Code) piping. This paper presents the motivations and objectives behind the establishment of the OPDE project. The paper also summarizes the unique data quality considerations that are associated with the reporting and recording of piping component degradation and failure. An overview of the database content is included to place it in perspective relative to past efforts to systematically collect and evaluate service experience data on piping performance. Finally, a brief summary is given of current database application studies

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

    International Nuclear Information System (INIS)

    Rastogi, Rohit; Bhasin, V.; Kushwaha, H.S.

    1997-01-01

    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)

  14. 76 FR 43981 - Circular Welded Austenitic Stainless Pressure Pipe From the People's Republic of China: Final...

    Science.gov (United States)

    2011-07-22

    ... merchandise; (2) it is at a level of economic development similar to that of the PRC pursuant to section 773(c... factors of production. No party submitted comments challenging the Department's selection of the primary... Preliminary Results, the Department received no comments challenging its finding that Jiuli TC/Jiuli SD Co...

  15. 76 FR 17819 - Circular Welded Austenitic Stainless Pressure Pipe From the People's Republic of China...

    Science.gov (United States)

    2011-03-31

    ... tubing, meeting ASTM A-554 or comparable domestic or foreign specifications; (2) boiler, heat exchanger... financial statements of Ratnamani Metals & Tubes, Ltd. (``Ratnamani'') and Jindal SAW, Ltd. (``Jindal'') for.... Failure to comply with this requirement could result in the Secretary's presumption that reimbursement of...

  16. Optimization of Pipe Networks

    DEFF Research Database (Denmark)

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

    1991-01-01

    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...

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

    International Nuclear Information System (INIS)

    Bellesia, B.

    2006-12-01

    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

  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. Effects of cooling rate, austenitizing temperature and austenite deformation on the transformation behavior of high-strength boron steel

    International Nuclear Information System (INIS)

    Mun, Dong Jun; Shin, Eun Joo; Choi, Young Won; Lee, Jae Sang; Koo, Yang Mo

    2012-01-01

    Highlights: ► Non-equilibrium segregation of B in steel depends strongly on the cooling rate. ► A higher austenitization temperature reduced the B hardenability effect. ► An increase in B concentration at γ grain boundaries accelerates the B precipitation. ► The loss of B hardenability effect is due to intragranular borocarbide precipitation. ► The controlled cooling after hot deformation increased the B hardenability effect. - Abstract: The phase transformation behavior of high-strength boron steel was studied considering the segregation and precipitation behavior of boron (B). The effects of cooling rate, austenitizing temperature and austenite deformation on the transformation behavior of B-bearing steel as compared with B-free steel were investigated by using dilatometry, microstructural observations and analysis of B distribution. The effects of these variables on hardenability were discussed in terms of non-equilibrium segregation mechanism and precipitation behavior of B. The retardation of austenite-to-ferrite transformation by B addition depends strongly on cooling rate (CR); this is mainly due to the phenomenon of non-equilibrium grain boundary segregation of B. The hardenability effect of B-bearing steel decreased at higher austenitizing temperature due to the precipitation of borocarbide along austenite grain boundaries. Analysis of B distribution by second ion mass spectroscopy confirmed that the grain boundary segregation of B occurred at low austenitizing temperature of 900 °C, whereas B precipitates were observed along austenite grain boundaries at high austenitizing temperature of 1200 °C. The significant increase in B concentration at austenite grain boundaries due to grain coarsening and a non-equilibrium segregation mechanism may lead to the B precipitation. In contrast, solute B segregated to austenite grain boundaries during cooling after heavy deformation became more stable because the increase in boundary area by grain

  20. Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

    International Nuclear Information System (INIS)

    Balkey, K.R.; Wallace, I.T.; Vaurio, J.K.

    1982-01-01

    One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results

  1. Cyclic deformation behaviour of austenitic steels at ambient and ...

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    Fatigue; cyclic deformation behaviour; metastable austenitic steel; .... Figure 4 shows a sequence of the basic diagrams which can be used to assess the fatigue .... well as the change of temperature and the development of the magnetic ...

  2. Low-temperature creep of austenitic stainless steels

    Science.gov (United States)

    Reed, R. P.; Walsh, R. P.

    2017-09-01

    Plastic deformation under constant load (creep) in austenitic stainless steels has been measured at temperatures ranging from 4 K to room temperature. Low-temperature creep data taken from past and unreported austenitic stainless steel studies are analyzed and reviewed. Creep at cryogenic temperatures of common austenitic steels, such as AISI 304, 310 316, and nitrogen-strengthened steels, such as 304HN and 3116LN, are included. Analyses suggests that logarithmic creep (creep strain dependent on the log of test time) best describe austenitic stainless steel behavior in the secondary creep stage and that the slope of creep strain versus log time is dependent on the applied stress/yield strength ratio. The role of cold work, strain-induced martensitic transformations, and stacking fault energy on low-temperature creep behavior is discussed. The engineering significance of creep on cryogenic structures is discussed in terms of the total creep strain under constant load over their operational lifetime at allowable stress levels.

  3. Formation and stabilization of reversed austenite in supermartensitic stainless steel

    DEFF Research Database (Denmark)

    Nießen, Frank; Grumsen, Flemming Bjerg; Hald, John

    2017-01-01

    of the reversed austenite phase fraction. Annealing at higher temperatures led to a gradual increase in hardness which was caused by formation of fresh martensite from reversed austenite. It was demonstrated that stabilization of reversed austenite is primarily based on chemical stabilization by partitioning......The formation and stabilization of reversed austenite upon inter-critical annealing was investigated in a X4CrNiMo16-5-1 (EN 1.4418) supermartensitic stainless steel by means of scanning electron microscopy, electron backscatter-diffraction, transmission electron microscopy, energy-dispersive X......-ray spectroscopy and dilatometry. The results were supported by thermodynamics and kinetics models, and hardness measurements. Isothermal annealing for 2 h in the temperature range of 475 to 650 °C led to gradual softening of the material which was related to tempering of martensite and the steady increase...

  4. Nanostructured Bainite-Austenite Steel for Armours Construction

    Directory of Open Access Journals (Sweden)

    Burian W.

    2014-10-01

    Full Text Available Nanostructured bainite-austenite steels are applied in the armours construction due to their excellent combination of strength and ductility which enables to lower the armour weight and to improve the protection efficiency. Mechanical properties of the bainite-austenite steels can be controlled in the wide range by chemical composition and heat treatment. In the paper the results of investigation comprising measuring of quasi - static mechanical properties, dynamic yield stress and firing tests of bainite-austenite steel NANOS-BA® are presented. Reported results show that the investigated bainite-austenite steel can be used for constructing add-on armour and that the armour fulfils requirements of protection level 2 of STANAG 4569. Obtained reduction in weight of the tested NANOS-BA® plates in comparison with the present solutions is about 30%.

  5. Welding of austenitic stainless steel with a high molybdenum content

    International Nuclear Information System (INIS)

    Liljas, A.; Holmberg, B.

    1984-01-01

    Welding of austenitic steel is discussed. Welding tests of AVESTA 250 SMO (six percent Mo) are reported. Welding without special additives can make the joints susceptible for corrosion in aggressive environments, e.g. sea water. (L.E.)

  6. Effects of austenitizing temperature in quenched niobium steels

    International Nuclear Information System (INIS)

    Mello, F.B.C. de; Assuncao, F.C.R.

    1980-01-01

    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) [pt

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

    Science.gov (United States)

    Basa, Adina; Thaulow, Christian; Barnoush, Afrooz

    2014-03-01

    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.

  8. Mechanical factors affecting reliability of pressure components (fatigue, cracking)

    International Nuclear Information System (INIS)

    Lebey, J.; Garnier, C.; Roche, R.; Barrachin, B.

    1978-01-01

    The reliability of a pressure component can be seriously affected by the formation and development of cracks. The experimental studies presented in this paper are devoted to three different aspects of crack propagation phenomena which have been relatively little described. In close connection with safety analyses of PWR, the authors study the influence of the environment by carrying out fatigue tests with samples bathed in hot pressurized water. Ferritic, austenitic and Incolloy 800 steels were used and the results are presented in the form of fatigue curves in the oligocyclic region. The second part of the paper relates to crack initiation cirteria in ductile steels weakened by notches. The CT samples used make it possible to study almost all types of fracture (ductile, intermediate and brittle). The use of two criteria based on the load limit and on the toughness of the material constitutes a practical way of evaluating crack propagation conditions. A series of tests carried out on notched spherical vessels of different size shows that large vessels are relatively brittle; fast unstable fracture is observed as size increases. Crack growth rate in PWR primary circuits (3/6 steel) is studied on piping elements (0.25 scale) subjected to cyclic stress variations (285 0 C and with pressure varying between 1 and 160 bar in each cycle). By calculating the stress intensity factor, correlation with results obtained in the laboratory on CT samples is possible. (author)

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

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.

    1983-06-01

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

  10. A study of the carbon distribution in retained austenite

    International Nuclear Information System (INIS)

    Scott, C.P.; Drillet, J.

    2007-01-01

    Cold-rolled and annealed transformation-induced plasticity (TRIP) steels were overaged to modify the carbon concentrations (C γ ) in retained austenite. Experimental C γ values were directly obtained by electron energy loss spectroscopy and compared with data derived from X-ray diffraction measurements of the austenite lattice parameter (a γ ). In this way, we evaluated the different expressions available in the literature relating C γ to a γ

  11. Heat pipe applications workshop report

    International Nuclear Information System (INIS)

    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

  12. 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)

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

    International Nuclear Information System (INIS)

    Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

    2010-01-01

    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. The

  14. Stresses in a curved pipe subject to an in-plane bending moment

    International Nuclear Information System (INIS)

    Hofmann, E.; Heeschen, U.

    1979-01-01

    The design of the KWU-primary component supports is mainly defined by the loads of the postulated pipe breaks. To estimate the maximum loading of a component support it is necessary to know the maximum in-plane bending moment (opening and closing) that can be transmitted by a pipe bend. Another reason for such information is that the displacements and distortions of the components cause higher stresses in elbows than in straight pipes. With a detailed knowledge of the deformation characteristic of a pipe bend an integrity analysis could be done without an expensive plastic system analysis. With this purpose in mind experiments were performed with straight pipes and pipe bends of different dimensions subject to in-plane bending moments. The experimental results give the ratio between the maximum transmittable moment of a pipe bend to that of a straight pipe or, the distortion of the end cross-sections and the flattening of the elbow cross-section. An attempt is made to derive simple expressions for estimating the behaviour at pipe elbows. Parallel to the experiments calculations were done for the straight pipe and elbow with a finite difference code with plastic capabilities. The results of the experiment and calculation are compared with the formulas of the ASME-Code section III subjection NB. (orig.)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  16. Analytical model of impedance in elliptical beam pipes

    CERN Document Server

    Pesah, Arthur Chalom

    2017-01-01

    Beam instabilities are among the main limitations in building higher intensity accelerators. Having a good impedance model for every accelerators is necessary in order to build components that minimize the probability of instabilities caused by the interaction beam-environment and to understand what piece to change in case of intensity increasing. Most of accelerator components have their impedance simulated with finite elements method (using softwares like CST Studio), but simple components such as circular or flat pipes are modeled analytically, with a decreasing computation time and an increasing precision compared to their simulated model. Elliptical beam pipes, while being a simple component present in some accelerators, still misses a good analytical model working for the hole range of velocities and frequencies. In this report, we present a general framework to study the impedance of elliptical pipes analytically. We developed a model for both longitudinal and transverse impedance, first in the case of...

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

    DEFF Research Database (Denmark)

    Chen, Ming; Alimadadi, Hossein; Molin, Sebastian

    2017-01-01

    Ferritic stainless steel interconnect plates are widely used in planar solid oxide fuel cell and electrolysis cell stacks. During stack production and operation, nickel from the Ni/yttria stabilized zirconia fuel electrode or from the Ni contact component layer diffuses into the interconnect 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 interconnect plates. In this work, kinetic modeling of the inter-diffusion between Ni and FeCr based ferritic...

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

    DEFF Research Database (Denmark)

    Chen, Ming; Molin, Sebastian; Zhang, L.

    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 modeling of the inter-diffusion between Ni and FeCr based ferritic stainless steel was conducted, using the CALPHAD...

  19. Elastic interaction between twins during tensile deformation of austenitic stainless steel

    DEFF Research Database (Denmark)

    Juul, Nicolai Ytterdal; Winther, Grethe; Dale, Darren

    2016-01-01

    . However, the components of the Type II stress normal to the twin boundary plane exhibit the same large variations as for the grain boundaries. Elastic grain interactions are therefore complex and must involve the entire set of neighbouring grains. The elastic-regime stress along the tensile direction......In austenite, the twin boundary normal is a common elastically stiff direction shared by the two twins, which may induce special interactions. By means of three-dimensional X-ray diffraction this elastic interaction has been analysed and compared to grains separated by conventional grain boundaries...

  20. Chemical decontamination of reactor components

    International Nuclear Information System (INIS)

    Riess, R.; Berthold, H.O.

    1977-08-01

    A solution for the decontamination of reactor components of the primary system was developed. This solution is a modification of the APAC- (Alkaline Permanganate Ammonium Citrate) system described in the literature. The most important advantage of the present solution over the APAC-method is that it does not induce any selective corrosion attack on materials like stainless steel (austenitic), Inconel 600 and Incoloy 800. (orig.) [de

  1. 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)

  2. 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)

  3. Waste pipe calculus

    International Nuclear Information System (INIS)

    Kaufman, A.M.

    1978-01-01

    A rapid method is presented for calculating transport in a network of one-dimensional flow paths or ''pipes''. The method defines a Green's function for each flow path and prescribes a method of combining these Green's functions to produce an overall Green's function for the flow path network. A unique feature of the method is the use of the Laplace transform of these Green's functions to carry out most of the calculations

  4. Pipe damping studies

    International Nuclear Information System (INIS)

    Ware, A.G.

    1986-01-01

    The Idaho National Engineering Laboratory (INEL) is conducting a research program to assist the United States Nuclear Regulatory Commission (USNRC) in determining best-estimate damping values for use in the dynamic analysis of nuclear power plant piping systems. This paper describes four tasks in the program that were undertaken in FY-86. In the first task, tests were conducted on a 5-in. INEL laboratory piping system and data were analyzed from a 6-in. laboratory system at the ANCO Engineers facility to investigate the parameters influencing damping in the seismic frequency range. Further tests were conducted on 3- and 5-in. INEL laboratory piping systems as the second task to determine damping values representative of vibrations in the 33 to 100 Hz range, typical of hydrodynamic transients. In the third task a statistical evaluation of the available damping data was conduted to determine probability distributions suitable for use in probabilistic risk assessments (PRAs), and the final task evaluated damping data at high strain levels

  5. Pipe clamp effects on thin-walled pipe design

    International Nuclear Information System (INIS)

    Lindquist, M.R.

    1980-01-01

    Clamp induced stresses in FFTF piping are sufficiently large to require structural assessment. The basic principles and procedures used in analyzing FFTF piping at clamp support locations for compliance with ASME Code rules are given. Typical results from a three-dimensional shell finite element pipe model with clamp loads applied over the clamp/pipe contact area are shown. Analyses performed to categorize clamp induced piping loads as primary or secondary in nature are described. The ELCLAMP Computer Code, which performs analyses at clamp locations combining clamp induced stresses with stresses from overall piping system loads, is discussed. Grouping and enveloping methods to reduce the number of individual clamp locations requiring analysis are described

  6. Stability of cracked pipe under seismic/dynamic displacement-controlled stresses. Subtask 1.2 final report

    International Nuclear Information System (INIS)

    Kramer, G.; Veith, P.; Marschall, C.

    1997-06-01

    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

  7. Refurbishment of the IEAR1 primary coolant system piping supports

    International Nuclear Information System (INIS)

    Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

    2015-01-01

    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)

  8. Piping structural design for the ITER thermal shield manifold

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Chang Hyun, E-mail: chnoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Chung, Wooho, E-mail: whchung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo; Kang, Kyoung-O. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bae, Jing Do; Cha, Jong Kook [Korea Marine Equipment Research Institute, Busan 606-806 (Korea, Republic of); Kim, Kyoung-Kyu [Mecha T& S, Jinju-si 660-843 (Korea, Republic of); Hamlyn-Harris, Craig; Hicks, Robby; Her, Namil; Jun, Chang-Hoon [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • We finalized piping design of ITER thermal shield manifold for procurement. • Support span is determined by stress and deflection limitation. • SQP, which is design optimization method, is used for the pipe design. • Benchmark analysis is performed to verify the analysis software. • Pipe design is verified by structural analyses. - Abstract: The thermal shield (TS) provides the thermal barrier in the ITER tokamak to minimize heat load transferred by thermal radiation from the hot components to the superconducting magnets operating at 4.2 K. The TS is actively cooled by 80 K pressurized helium gas which flows from the cold valve box to the cooling tubes on the TS panels via manifold piping. This paper describes the manifold piping design and analysis for the ITER thermal shield. First, maximum allowable span for the manifold support is calculated based on the simple beam theory. In order to accommodate the thermal contraction in the manifold feeder, a contraction loop is designed and applied. Sequential Quadratic Programming (SQP) method is used to determine the optimized dimensions of the contraction loop to ensure adequate flexibility of manifold pipe. Global structural behavior of the manifold is investigated when the thermal movement of the redundant (un-cooled) pipe is large.

  9. Equi-biaxial loading effect on austenitic stainless steel fatigue life

    Directory of Open Access Journals (Sweden)

    C. Gourdin

    2016-10-01

    Full Text Available Fatigue lifetime assessment is essential in the design of structures. Under-estimated predictions may result in unnecessary in service inspections. Conversely, over-estimated predictions may have serious consequences on the integrity of structures. In some nuclear power plant components, the fatigue loading may be equibiaxial because of thermal fatigue. So the potential impact of multiaxial loading on the fatigue life of components is a major concern. Meanwhile, few experimental data are available on austenitic stainless steels. It is essential to improve the fatigue assessment methodologies to take into account the potential equi-biaxial fatigue damage. Hence this requires obtaining experimental data on the considered material with a strain tensor in equibiaxial tension. Two calibration tests (with strain gauges and image correlation were used to obtain the relationship between the imposed deflection and the radial strain on the FABIME2 specimen. A numerical study has confirmed this relationship. Biaxial fatigue tests are carried out on two austenitic stainless steels for different values of the maximum deflection, and with a load ratio equal to -1. The interpretation of the experimental results requires the use of an appropriate definition of strain equivalent. In nuclear industry, two kinds of definition are used: von Mises and TRESCA strain equivalent. These results have permitted to estimate the impact of the equibiaxiality on the fatigue life of components

  10. Pipe inspection using the pipe crawler. Innovative technology summary report

    International Nuclear Information System (INIS)

    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

  11. Heat pipes and heat pipe exchangers for heat recovery systems

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliev, L L; Grakovich, L P; Kiselev, V G; Kurustalev, D K; Matveev, Yu

    1984-01-01

    Heat pipes and heat pipe exchangers are of great importance in power engineering as a means of recovering waste heat of industrial enterprises, solar energy, geothermal waters and deep soil. Heat pipes are highly effective heat transfer units for transferring thermal energy over large distance (tens of meters) with low temperature drops. Their heat transfer characteristics and reliable working for more than 10-15 yr permit the design of new systems with higher heat engineering parameters.

  12. 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.

  13. Investigation on field removed pipe sections in the PISC hot laboratories

    International Nuclear Information System (INIS)

    Cambini, M.; Crutzen, S.; Jehenson, P.; Bergh, R. Van den; Violin, F.

    1990-01-01

    Action No. 1 of PISC II: Real Contaminated Structures (RCS), seeks to collect results from specific investigations and limited round robin tests on real service induced defects in materials and structures of the primary circuit of Light Water Reactors. The hot cell facilities at JRC-Ispra are fully equipped for non destructive and destructive work on a collaborative basis. Cracked austenitic steel primary circuit pipes coming from the primary circuit of the Muhleberg reactor (Switzerland) have been inspected in order to demonstrate the validity of the facilities to examine these contaminated pieces. (author)

  14. Development and application of preventive maintenance technique for pipes using laser cladding method

    International Nuclear Information System (INIS)

    Hatakenaka, Hiroaki; Yamadera, Masao; Shiraiwa, Takanori.

    1995-01-01

    A laser cladding method which produces a highly corrosion-resisting coating (cladding) on the surface of the material was developed for the purpose of preventing stress corrosion cracking (SCC) in the austenitic stainless steel (Type 304). In this method, metallic powder paste is applied on the inner surface of pipes, and then a YAG laser beam is irradiated to the paste, which melts and forms a clad with excellent corrosion resistance. Recently, the laser cladding method was practically and successfully applied to the actual nuclear power plant in Japan. This report describes this laser cladding technique, the equipment, and actual works in the field. (author)

  15. Leak before break analysis for cracking at multiple weld locations in BWR recirculation piping

    International Nuclear Information System (INIS)

    Zahoor, A.; Gamble, R.

    1984-01-01

    Periodically over the past decade, intergranular stress corrosion cracking (IGSCC) has been found in austenitic stainless steel piping at Boiling Water Reactor facilities. The effect of IGSCC on piping integrity has been evaluated previously in various BWR Owners Group and NRC studies. In these studies, the analyses were performed assuming the presence of a crack at a single weld location in the pipe run. The purpose of this investigation was to compare the leak rate and potential for unstable crack extension associated with a throughwall crack for the following two conditions in a BWR recirculation system: (1) the recirculation piping contains part through cracks at multiple weld locations and a single throughwall crack, and (2) the piping contains only a throughwall crack at one weld location. Two type BWRs were evaluated; namely, the ring header and five individual loop designs. The results from the analyses indicate that the potential for unstable crack extension at large bending loads, and leak rate at normal operation are not affected by the presence of part through cracks at multiple weld locations. The differences in the respective calculated L/sub eff/ and leak rates for the single and multiply cracked conditions are less than 2%

  16. Austenitic alloys Fe-Ni-Cr dominating

    International Nuclear Information System (INIS)

    Gibson, R.C.; Korenko, M.K.

    1980-01-01

    Austenitic alloy essentially comprising 42 to 48% nickel, 11 to 13% chromium, 2.6 to 3.4% niobium, 0.2 to 1.2% silicon, 0.5 to 1.5% vanadium, 2.6 to 3.4% molybdenum, 0.1 to 0.3% aluminium, 0.1 to 0.3% titanium, 0.02 to 0.05% carbon, 0.002 to 0.015% boron, up to 0.06% zirconium, the balance being iron. The characteristic of this alloy is a conventional elasticity limit to within 2% of at least 450 MPa, with a maximum tensile strength of at least 500 MPa at a test temperature of 650 0 C after immersion annealing at 1038 0 C and 30% hardening. To this effect the invention concerns Ni-Cr-Fe high temperature alloys possessing excellent mechanical strength characteristics, that can be obtained with lower levels of nickel and chromium than those used in alloys of this kind in the present state of the technique, a higher amount of niobium than in the previous alloys and with the addition of 0.5 to 1.5% vanadium [fr

  17. Ion-nitriding of austenitic stainless steels

    International Nuclear Information System (INIS)

    Pacheco, O.; Hertz, D.; Lebrun, J.P.; Michel, H.

    1995-01-01

    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

  18. 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)

    Esmailian, M.

    2010-01-01

    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.

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

    International Nuclear Information System (INIS)

    Chakravarti, B.

    1996-01-01

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

  20. Corrosion and deposit evaluation in large diameter pipes using radiography

    International Nuclear Information System (INIS)

    Boateng, A.

    2012-01-01

    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)

  1. Effect of austenitization conditions on kinetics of isothermal transformation of austenite of structural steels

    International Nuclear Information System (INIS)

    Konopleva, E.V.; Bayazitov, V.M.; Abramov, O.V.; Kozlova, A.G.

    1987-01-01

    Effect of austenization of kinetics of pearlite and bainite transformations for steels with different carbon content differing by alloying character and degree has been investigated. Austenization temperature increase is shown to leads to retardation of ferrite-pearlite transformation in low- and medium-carbon alloyed steels. Step-like holding in the region of austenite stable state (850, 950 deg) after high-temperature heating (1100 deg C) increases the rate of transformation partially recovering its kinetics and decomposition velocity after low-temperature heating in steels alloyed advantageously with carbide-forming elements (08Kh2G2F, 30Kh3) and does not affect kinetics in the 35Kh, 30KhGSN2A, 45N5 steels. Increase of heating temperature and growth of an austenite grain cause considerable acceleration of bainite transformation, increase of the temperaure of bainite transformation beginning and increase of the transformation amplitude in the 08Kh2G2F, 30Kh3 steels and affect weakly kinetics in steels with mixed alloying (30KhGSN2A) or low-alloy one (35Kh). The bainite transformation rate in the 45N5 steelite does not depend on austenization. The effect of additional acceleration of bainite transformation as a result holding after high-temperature heating in those steels, where activation of transformation occurs with increase of heating temperature

  2. Impact test of components

    International Nuclear Information System (INIS)

    Borsoi, L.; Buland, P.; Labbe, P.

    1987-01-01

    Stops with gaps are currently used to support components and piping: it is simple, low cost, efficient and permits free thermal expansion. In order to keep the nonlinear nature of stops, such design is often modeled by beam elements (for the component) and nonlinear springs (for the stops). This paper deals with the validity and the limits of these models through the comparison of computational and experimental results. The experimental results come from impact laboratory tests on a simplified mockup. (orig.)

  3. Flat Miniature Heat Pipes for Electronics Cooling: State of the Art, Experimental and Theoretical Analysis

    OpenAIRE

    M.C. Zaghdoudi; S. Maalej; J. Mansouri; M.B.H. Sassi

    2011-01-01

    An experimental study is realized in order to verify the Mini Heat Pipe (MHP) concept for cooling high power dissipation electronic components and determines the potential advantages of constructing mini channels as an integrated part of a flat heat pipe. A Flat Mini Heat Pipe (FMHP) prototype including a capillary structure composed of parallel rectangular microchannels is manufactured and a filling apparatus is developed in order to charge the FMHP. The heat transfer im...

  4. Noncondensable gas accumulation phenomena in nuclear power plant piping

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Aoki, Kazuyoshi; Sato, Teruaki; Shida, Akira; Ichikawa, Nagayoshi; Nishikawa, Akira; Inagaki, Tetsuhiko

    2011-01-01

    In the case of the boiling water reactor, hydrogen and oxygen slightly exist in the main steam, because these noncondensable gases are generated by the radiolytic decomposition of the reactor water. BWR plants have taken measures to prevent noncondensable gas accumulation. However, in 2001, the detonation of noncondensable gases occurred at Hamaoka-1 and Brunsbuttel, resulting in ruptured piping. The accumulation phenomena of noncondensable gases in BWR closed piping must be investigated and understood in order to prevent similar events from occurring in the future. Therefore, an experimental study on noncondensable gas accumulation was carried out. The piping geometries for testing were classified and modeled after the piping of actual BWR plants. The test results showed that 1) noncondensable gases accumulate in vertical piping, 2) it is hard for noncondensable gases to accumulate in horizontal piping, and 3) noncondensable gases accumulate under low-pressure conditions. A simple accumulation analysis method was proposed. To evaluate noncondensable gas accumulation phenomena, the three component gases were treated as a mixture. It was assumed that the condensation amount of the vapor is small, because the piping is certainly wrapped with heat insulation material. Moreover, local thermal equilibrium was assumed. This analysis method was verified using the noncondensable gas accumulation test data on branch piping with a closed top. Moreover, an experimental study on drain trap piping was carried out. The test results showed that the noncondensable gases dissolved in the drain water were discharged from the drain trap, and Henry's law could be applied to evaluate the amount of dissolved noncondensable gases in the drain water. (author)

  5. Removal of a section of the CMS beam pipe

    CERN Document Server

    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.

  6. 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 approximate nonlinear solution for small-amplitude flutter oscillations is obtained using a fifth-order multiple scales perturbation method, and large-amplitude oscillations are examined by numerical integration of the autonomous model equations, using a path-following algorithm. The pulsating fluid component...

  7. Damping in LMFBR pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Barta, D.A.; Lindquist, M.R.; Renkey, E.J.; Ryan, J.A.

    1983-06-01

    LMFBR pipe systems typically utilize a thicker insulation package than that used on water plant pipe systems. They are supported with special insulated pipe clamps. Mechanical snubbers are employed to resist seismic loads. Recent laboratory testing has indicated that these features provide significantly more damping than presently allowed by Regulatory Guide 1.61 for water plant pipe systems. This paper presents results of additional in-situ vibration tests conducted on FFTF pipe systems. Pipe damping values obtained at various excitation levels are presented. Effects of filtering data to provide damping values at discrete frequencies and the alternate use of a single equivalent modal damping value are discussed. These tests further confirm that damping in typical LMFBR pipe systems is larger than presently used in pipe design. Although some increase in damping occurred with increased excitation amplitude, the effect was not significant. Recommendations are made to use an increased damping value for both the OBE and DBE seismic events in design of LMFBR pipe systems

  8. Dynamic experiments on cracked pipes

    International Nuclear Information System (INIS)

    Petit, M.; Brunet, G.; Buland, P.

    1991-01-01

    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

  9. Water hammer in elastic pipes

    International Nuclear Information System (INIS)

    Gale, J.; Tiselj, I.

    2002-01-01

    One dimensional two-fluid six-equation model of two-phase flow, that can be found in computer codes like RELAP5, TRAC, and CATHARE, was upgraded with additional terms, which enable modelling of the pressure waves in elastic pipes. It is known that pipe elasticity reduces the propagation velocity of the shock and other pressure waves in the piping systems. Equations that include the pipe elasticty terms are used in WAHA code, which is being developed within the WAHALoads project of 5't'h EU research program.(author)

  10. Mechanical assessment of local thinned pipings

    International Nuclear Information System (INIS)

    Meister, E.

    2007-01-01

    Local wall thinning is likely to be found in some piping systems of nuclear power plant under, for example, Flow Accelerated Corrosion in raw water systems or by loss of metal during the grinding of the weld seam. To assess the mechanical integrity in such situations, EDF/SEPTEN has developed calculation methods for the RSE-M (In Service Inspection Rules for the Mechanical components of PWR nuclear power islands) code. This paper focuses on the methodology used for internal pressure resistance evaluation based on limit load calculations. Beyond the Nuclear Safety classification and requirements given by the RSE-M code, this problem is general for Power Piping and the associated in service rules. (author) [fr

  11. Reversed austenite for enhancing ductility of martensitic stainless steel

    Science.gov (United States)

    Dieck, S.; Rosemann, P.; Kromm, A.; Halle, T.

    2017-03-01

    The novel heat treatment concept, “quenching and partitioning” (Q&P) has been developed for high strength steels with enhanced formability. This heat treatment involves quenching of austenite to a temperature between martensite start and finish, to receive a several amount of retained austenite. During the subsequent annealing treatment, the so called partitioning, the retained austenite is stabilized due to carbon diffusion, which results in enhanced formability and strength regarding strain induced austenite to martensite transformation. In this study a Q&P heat treatment was applied to a Fe-0.45C-0.65Mn-0.34Si-13.95Cr stainless martensite. Thereby the initial quench end temperature and the partitioning time were varied to characterize their influence on microstructural evolution. The microstructural changes were analysed by dilatometer measurements, X-ray diffraction and scanning electron microscopy, including electron back-scatter diffraction. Compression testing was made to examine the mechanical behaviour. It was found that an increasing partitioning time up to 30 min leads to an enhanced formability without loss in strength due to a higher amount of stabilized retained and reversed austenite as well as precipitation hardening.

  12. Modeling of Non-isothermal Austenite Formation in Spring Steel

    Science.gov (United States)

    Huang, He; Wang, Baoyu; Tang, Xuefeng; Li, Junling

    2017-12-01

    The austenitization kinetics description of spring steel 60Si2CrA plays an important role in providing guidelines for industrial production. The dilatometric curves of 60Si2CrA steel were measured using a dilatometer DIL805A at heating rates of 0.3 K to 50 K/s (0.3 °C/s to 50 °C/s). Based on the dilatometric curves, a unified kinetics model using the internal state variable (ISV) method was derived to describe the non-isothermal austenitization kinetics of 60Si2CrA, and the abovementioned model models the incubation and transition periods. The material constants in the model were determined using a genetic algorithm-based optimization technique. Additionally, good agreement between predicted and experimental volume fractions of transformed austenite was obtained, indicating that the model is effective for describing the austenitization kinetics of 60Si2CrA steel. Compared with other modeling methods of austenitization kinetics, this model, which uses the ISV method, has some advantages, such as a simple formula and explicit physics meaning, and can be probably used in engineering practice.

  13. Diffraction study of the retained austenite content in TRIP steels

    Energy Technology Data Exchange (ETDEWEB)

    Gnaeupel-Herold, T., E-mail: tg-h@nist.gov [NIST Center for Neuron Research, 100 Bureau Dr., Gaithersburg MD 20899-6102 (United States); University of Maryland, Department of Material Science and Engineering., College Park MD 20742-2142 (United States); Creuziger, A., E-mail: adam.creuziger@nist.gov [NIST Metallurgy Division, 100 Bureau Dr., Gaithersburg MD 20899-8553 (United States); Kent State University, Kent, OH 44242 (United States)

    2011-04-25

    Research highlights: {yields} Novel orientation averaging scheme for retained austenite content measurement. {yields} assumption of random grain orientation generally not justified. {yields} Averaging scheme allows to disregard texture. {yields} unlike Rietveld method, averaging method does not orientation density function. {yields} Two independent (hkl) are necessary for retained austenite content. - Abstract: The results of a study of using neutron diffraction for determining the retained austenite content of TRIP steels are presented. The study covers a wide area of materials, deformation modes (uniaxial, biaxial and plane strain), strains, and the retained austenite content as a result of these variables. It was determined using basic principles of statistics that a minimum of two reflections (hkl) for each phase is necessary to calculate a phase mass fraction and the associated standard deviation. Texture from processing the steel is the largest source of uncertainty. Through the method of complete orientation averaging described in this paper, the texture effect and with it the standard deviation of the austenite mass fraction can be substantially reduced, regardless of the type or severity of the texture.

  14. Solidification behavior of austenitic stainless steel filler metals

    International Nuclear Information System (INIS)

    David, S.A.; Goodwin, G.M.; Braski, D.N.

    1980-02-01

    Thermal analysis and interrupted solidification experiments on selected austenitic stainless steel filler metals provided an understanding of the solidification behavior of austenitic stainless steel welds. The sequences of phase separations found were for type 308 stainless steel filler metal, L + L + delta + L + delta + γ → γ + delta, and for type 310 stainless steel filler metal, L → L + γ → γ. In type 308 stainless steel filler metal, ferrite at room temperature was identified as either the untransformed primary delta-ferrite formed during the initial stages of solidification or the residual ferrite after Widmanstaetten austenite precipitation. Microprobe and scanning transmission electron microscope microanalyses revealed that solute extensively redistributes during the transformation of primary delta-ferrite to austenite, leading to enrichment and stabilization of ferrite by chromium. The type 310 stainless steel filler metal investigated solidifies by the primary crystallization of austenite, with the transformation going to completion at the solidus temperature. In our samples residual ferrite resulting from solute segregation was absent at the intercellular or interdendritic regions

  15. Methods allowing in-situ aging evaluation of austenitic-ferritic molded steels

    International Nuclear Information System (INIS)

    Paris, D.; Massoud, J.P.; Chicois, J.; Borelly, R.; Shoji, T.; Yi, Y.

    1997-01-01

    The molded elbows of the primary coolant circuit of the PWR type reactors are made with austenitic-ferritic stainless molded steel which is embrittled at the service temperature. To complement with the control programs lead from bench-scale experiments, it is wished to dispose to methods allowing to verify directly on the component that the estimated aging level is effectively reached. The development of non-destructive methods comes up against difficulties: the obtained signals are generally of weak amplitude; they have to be relevant of aging; the measures having not be carrying out on the component at the initial state, we do not dispose of a reference for each component. These difficulties are illustrated by the development of three methods: the electromagnetic sorting, the ferromagnetic noise and an electrochemical method. Two methods seem to be promising: the thermoelectric power and the small angle neutron scattering. (O.M.)

  16. Effect of Ti additions on the swelling of electron irradiated austenitic steels and Ni alloys

    International Nuclear Information System (INIS)

    Gilbon, D.; Didout, G.; Le Naour, L.; Levy, V.

    1979-01-01

    It has been shown that titanium is a beneficial additive for the swelling of austenitic steels. The amplitude of the effects observed depends much on the nature and concentration of the other additives in the austenitic matrix [fr

  17. 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.

  18. Dose dependence of the microstructural evolution in neutron-irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Zinkle, S.J.; Maziasz, P.J.; Stoller, R.E.

    1993-01-01

    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

  19. High cycle fatigue of austenitic stainless steels

    International Nuclear Information System (INIS)

    Gauthier, J.P.; Lehmann, D.; Picker

    1990-01-01

    This study concerns the evaluation of material data to be used in LMFBR design codes. High cycle fatigue properties of three austenitic stainless steels are evaluated: type AISI 316 (UKAEA tests), type AISI 316L (CEA tests) and type AISI 304 (Interatom tests). The data on these steels comprised some 550 data points from 14 casts. This data set covered a wide range of testing parameters: temperature from 20-625 0 C, frequency from 1-20 000 Hz, constant amplitude and random fatigue loading, with and without mean stress, etc. However, the testing conditions chosen by the three partners differed considerably because they had been fixed independently and not harmonized prior to the tests. This created considerable difficulties for the evaluations. Experimental procedures and statistical treatments used for the three subsets of data are described and discussed. Results are presented in tables and graphs. Although it is often difficult to single out the influence of each parameter due to the different testing conditions, several interesting conclusions can be drawn: The HCF properties of the three steels are consistent with the 0.2% proof stress, the fatigue limit being larger than the latter at temperatures above 550 0 C. The type 304 steel has lower tensile properties than the two other steels and hence also lower HCF properties. Parameters which clearly have a significant effect of HCF behaviour are mean stress or R-ratio (less in the non-endurance region than in the endurance region), temperature, cast or product. Other parameters have probably a weak or no effect but it is difficult to conclude due to insufficient data: environment, specimen orientation, frequency, specimen geometry

  20. Analytical solution for stress, strain and plastic instability of pressurized pipes with volumetric flaws

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, Sergio B., E-mail: sbcunha@petrobras.com.br [PETROBRAS/TRANSPETRO, Av. Pres. Vargas 328 - 7th floor, Rio de Janeiro, RJ 20091-060 (Brazil); Netto, Theodoro A., E-mail: tanetto@lts.coppe.ufrj.br [COPPE, Federal University ot Rio de Janeiro, Ocean Engineering Department, PO BOX 68508, Rio de Janeiro - RJ (Brazil)

    2012-01-15

    The mechanical behavior of internally pressurized pipes with volumetric flaws is analyzed. The two possible modes of circumferentially straining the pipe wall are identified and associated to hypothesized geometries. The radial deformation that takes place by bending the pipe wall is studied by means of axisymmetric flaws and the membrane strain developed by unequal hoop deformation is analyzed with the help of narrow axial flaws. Linear elastic shell solutions for stress and strain are developed, the plastic behavior is studied and the maximum hoop stress at the flaw is related to the undamaged pipe hoop stress by means of stress concentration factors. The stress concentration factors are employed to obtain equations predicting the pressure at which the pipe fails by plastic instability for both types of flaw. These analytical solutions are validated by comparison with burst tests on 3 Double-Prime diameter pipes and finite element simulations. Forty-one burst tests were carried out and two materials with very dissimilar plastic behavior, carbon steel and austenitic stainless steel, were used in the experiments. Both the analytical and the numerical predictions showed good correlation with the experimentally observed burst pressures. - Highlights: Black-Right-Pointing-Pointer An analytical model for the burst of a pipe with a volumetric flaw is developed. Black-Right-Pointing-Pointer Deformation, strain and stress are modeled in the elastic and plastic domains. Black-Right-Pointing-Pointer The model is comprehensively validated by experiments and numerical simulations. Black-Right-Pointing-Pointer The burst pressure model's accuracy is equivalent to finite element simulations.

  1. Large-bore pipe decontamination

    International Nuclear Information System (INIS)

    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

  2. Heat pipe and method of production of a heat pipe

    International Nuclear Information System (INIS)

    Kemp, R.S.

    1975-01-01

    The heat pipe consists of a copper pipe in which a capillary network or wick of heat-conducting material is arranged in direct contact with the pipe along its whole length. Furthermore, the interior space of the tube contains an evaporable liquid for pipe transfer. If water is used, the capillary network consists of, e.g., a phosphorus band network. To avoid contamination of the interior of the heat pipe during sealing, its ends are closed by mechanical deformation so that an arched or plane surface is obtained which is in direct contact with the network. After evacuation of the interior space, the remaining opening is closed with a tapered pin. The ratio wall thickness/tube diameter is between 0.01 and 0.6. (TK/AK) [de

  3. Numerical simulation of hydrogen-assisted crack initiation in austenitic-ferritic duplex steels

    International Nuclear Information System (INIS)

    Mente, Tobias

    2015-01-01

    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

  4. Comprehensive Deformation Analysis of a Newly Designed Ni-Free Duplex Stainless Steel with Enhanced Plasticity by Optimizing Austenite Stability

    Science.gov (United States)

    Moallemi, Mohammad; Zarei-Hanzaki, Abbas; Eskandari, Mostafa; Burrows, Andrew; Alimadadi, Hossein

    2017-08-01

    A new metastable Ni-free duplex stainless steel has been designed with superior plasticity by optimizing austenite stability using thermodynamic calculations of stacking fault energy and with reference to literature findings. Several characterization methods comprising optical microscopy, magnetic phase measurements, X-ray diffraction (XRD) and electron backscattered diffraction were employed to study the plastic deformation behavior and to identify the operating plasticity mechanisms. The results obtained show that the newly designed duplex alloy exhibits some extraordinary mechanical properties, including an ultimate tensile strength of 900 MPa and elongation to fracture of 94 pct due to the synergistic effects of transformation-induced plasticity and twinning-induced plasticity. The deformation mechanism of austenite is complex and includes deformation banding, strain-induced martensite formation, and deformation-induced twinning, while the ferrite phase mainly deforms by dislocation slip. Texture analysis indicates that the Copper and Rotated Brass textures in austenite (FCC phase) and {001} texture in ferrite and martensite (BCC phases) are the main active components during tensile deformation. The predominance of these components is logically related to the strain-induced martensite and/or twin formation.

  5. Evaluation of seismic margins for an in-plant piping system

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

    1991-01-01

    Earthquake experience as well as experiments indicate that, in general, piping systems are quite rugged in resisting seismic loadings. Therefore there is a basis to hold that the seismic margin against pipe failure is very high for systems designed according to current practice. However, there is very little data, either from tests or from earthquake experience, on the actual margin or excess capacity (against failure from seismic loading) of in-plant piping systems. Design of nuclear power plant piping systems in the US is governed by the criteria given in the ASME Boiler and Pressure Vessel (B ampersand PV) Code, which assure that pipe stresses are within specified allowable limits. Generally linear elastic analytical methods are used to determine the stresses in the pipe and forces in pipe supports. The objective of this study is to verify that piping designed according to current practice does indeed have a large margin against failure and to quantify the excess capacity for piping and dynamic pipe supports on the basis of data obtained in a series of high-level seismic experiments (designated SHAM) on an in-plant piping system at the HDR (Heissdampfreaktor) Test Facility in Germany. Note that in the present context, seismic margin refers to the deterministic excess capacities of piping or supports compared to their design capacities. The excess seismic capacities or margins of a prototypical in-plant piping system and its components are evaluated by comparing measured inputs and responses from high-level simulated seismic experiments with design loads and allowables. Large excess capacities are clearly demonstrated against pipe and overall system failure with the lower bound being about four. For snubbers the lower bound margin is estimated at two and for rigid strut supports at five. 4 refs., 2 figs., 2 tabs

  6. Components of the primary circuit of LWRs

    International Nuclear Information System (INIS)

    1980-01-01

    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.) [de

  7. Effect of multiple austenitizing treatments on HT-9 steels

    International Nuclear Information System (INIS)

    Emigh, R.A.

    1985-12-01

    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

  8. submitter Physical Properties of a High-Strength Austenitic Stainless Steel for the Precompression Structure of the ITER Central Solenoid

    CERN Document Server

    Sgobba, Stefano; Arauzo, Ana; Roussel, Pascal; Libeyre, Paul

    2016-01-01

    The ITER central solenoid (CS) consists of six independent coils kept together by a precompression support structure that must react vertical tensile loads and provide sufficient preload to maintain coil-to-coil contact when the solenoid is energized. The CS precompression system includes tie plates, lower and upper key blocks, load distribution and isolation plates and other attachment, support and insulating hardware. The tie plates operating at 4 K are manufactured starting from forgings in a high-strength austenitic stainless steel (FXM-19) with a stringent specification. Moreover, forged components for the lower and upper key blocks have to be provided in the same FXM-19 grade with comparably strict requirements. FXM-19 is a high-nitrogen austenitic stainless steel, featuring high strength and toughness, ready weldability, and forgeability. It features as well higher integral thermal contraction down to 4 K compared with the very high Mn steel grade selected for the CS coil jackets, hence providing an ad...

  9. 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...

  10. Seismic ratchet-fatigue failure of piping systems

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  11. Seismic ratchet-fatigue failure of piping systems

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  12. Neutron depolarisation study of the austenite grain size in TRIP steels

    International Nuclear Information System (INIS)

    Dijk, N.H. van; Zhao, L.; Rekveldt, M.Th.; Fredrikze, H.; Tegus, O.; Brueck, E.; Sietsma, J.; Zwaag, S. van der

    2004-01-01

    We have performed combined neutron depolarisation and magnetisation measurements in order to obtain an in situ determination of the average grain size and volume fraction of the retained austenite phase in TRIP steels. The average grain size of the retained austenite was found to decrease for an increase in austenite volume fraction at two different annealing temperatures

  13. Self-stabilization of untransformed austenite by hydrostatic pressure via martensitic transformation

    International Nuclear Information System (INIS)

    Nakada, Nobuo; Ishibashi, Yuji; Tsuchiyama, Toshihiro; Takaki, Setsuo

    2016-01-01

    For improving the understanding of austenite stability in steel, hydrostatic pressure in untransformed austenite that is generated via martensitic transformation was evaluated from macro- and micro-viewpoints, and its effect on austenite stability was investigated in a Fe-27%Ni austenitic alloy. X-ray diffractometry revealed that the lattice parameter of untransformed austenite is continuously decreased via martensitic transformation only when martensite becomes the dominant phase in the microstructure. This suggests that the untransformed austenite is isotropically compressed by the surrounding martensite grains, i.e., hydrostatic pressure is generated in untransformed austenite dynamically at a later stage of martensitic transformation. On the other hand, microscopic strain mapping using the electron backscatter diffraction technique indicated that a finer untransformed austenite grain has a higher hydrostatic pressure, while a high density of dislocations is also introduced in untransformed austenite near the austenite/martensite interface because of lattice-invariant shear characterized by non-thermoelastic martensitic transformation. Furthermore, it was experimentally demonstrated that the hydrostatic pressure stabilizes the untransformed austenite; however, the austenite stabilization effect alone is not large enough to fully explain a large gap between martensite start and finish temperatures in steel.

  14. Impact analyses after pipe rupture

    International Nuclear Information System (INIS)

    Chun, R.C.; Chuang, T.Y.

    1983-01-01

    Two of the French pipe whip experiments are reproduced with the computer code WIPS. The WIPS results are in good agreement with the experimental data and the French computer code TEDEL. This justifies the use of its pipe element in conjunction with its U-bar element in a simplified method of impact analyses

  15. Mechanical Behaviour of Lined Pipe

    NARCIS (Netherlands)

    Hilberink, A.

    2011-01-01

    Installing lined pipe by means of the reeling installation method seems to be an attractive combination, because it provides the opportunity of eliminating the demanding welds from the critical time offshore and instead preparing them onshore. However, reeling of lined pipe is not yet proven

  16. Pulsed TIG welding of pipes

    International Nuclear Information System (INIS)

    Killing, U.

    1989-01-01

    The present study investigates into the effects of impulse welding parameters on weld geometry in the joint welding of thin-walled sheets and pipes (d=2.5 mm), and it uses random samples of thick-walled sheets and pipes (d=10 mm), in fixed positions. (orig./MM) [de

  17. Oxidation resistant high creep strength austenitic stainless steel

    Science.gov (United States)

    Brady, Michael P.; Pint, Bruce A.; Liu, Chain-Tsuan; Maziasz, Philip J.; Yamamoto, Yukinori; Lu, Zhao P.

    2010-06-29

    An austenitic stainless steel displaying high temperature oxidation and creep resistance has a composition that includes in weight percent 15 to 21 Ni, 10 to 15 Cr, 2 to 3.5 Al, 0.1 to 1 Nb, and 0.05 to 0.15 C, and that is free of or has very low levels of N, Ti and V. The alloy forms an external continuous alumina protective scale to provide a high oxidation resistance at temperatures of 700 to 800.degree. C. and forms NbC nanocarbides and a stable essentially single phase fcc austenitic matrix microstructure to give high strength and high creep resistance at these temperatures.

  18. A generic approach for a linear elastic fracture mechanics analysis of components containing residual stress

    International Nuclear Information System (INIS)

    Lee, Hyeong Y.; Nikbin, Kamran M.; O'Dowd, Noel P.

    2005-01-01

    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

  19. Promethus Hot Leg Piping Concept

    International Nuclear Information System (INIS)

    AM Girbik; PA Dilorenzo

    2006-01-01

    The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactor (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept

  20. Irradiation creep and swelling of various austenitic alloys irradiated in PFR and FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Toloczko, M.B. [Pacific Northwest National Lab., Richland, WA (United States)] [and others

    1996-10-01

    In order to use data from surrogate neutron spectra for fusion applications, it is necessary to analyze the impact of environmental differences on property development. This is of particular importance in the study of irradiation creep and its interactions with void swelling, especially with respect to the difficulty of separation of creep strains from various non-creep strains. As part of an on-going creep data rescue and analysis effort, the current study focuses on comparative irradiations conducted on identical gas-pressurized tubes produced and constructed in the United States from austenitic steels (20% CW 316 and 20% CW D9), but irradiated in either the Prototype Fast Reactor (PFR) in the United Kingdom or the Fast Flux Test Facility in the United States. In PFR, Demountable Subassemblies (DMSA) serving as heat pipes were used without active temperature control. In FFTF the specimens were irradiated with active ({+-}{degrees}5C) temperature control. Whereas the FFTF irradiations involved a series of successive side-by-side irradiation, measurement and reinsertion of the same series of tubes, the PFR experiment utilized simultaneous irradiation at two axial positions in the heat pipe to achieve different fluences at different flux levels. The smaller size of the DMSA also necessitated a separation of the tubes at a given flux level into two groups (low-stress and high-stress) at slightly different axial positions, where the flux between the two groups varied {le}10%. Of particular interest in this study was the potential impact of the two types of separation on the derivation of creep coefficients.

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

    International Nuclear Information System (INIS)

    Greenstreet, W.L.

    1975-11-01

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

  2. Heat pipes. (Latest citations from the US Patent Bibliographic file with exemplary claims). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-01-01

    The bibliography contains citations of selected patents concerning the design, manufacture, and applications of heat pipes. The use of heat pipes in heat exchange systems for heat storage, heat transfer, and heat utilization is discussed. Applications include semiconductor cooling, use in engine components, and building cooling and heating. (Contains 250 citations and includes a subject term index and title list.)

  3. Development of pipe wall thinning prediction software 'FALSET'

    International Nuclear Information System (INIS)

    Yoneda, Kimitoshi; Morita, Ryo; Inada, Fumio; Fujiwara, Kazutoshi

    2012-01-01

    Pipe wall thinning in power plants has been managed for maintaining plant integrity and safety with great importance. The target thinning phenomena are Flow Accelerated Corrosion (FAC) and Liquid Droplet Impingement Erosion (LDI). At present, the management is based on thinning rate and residual lifetime evaluation using pipe wall thickness measurement results. For the future, more safety and improvement in the management is required, and in this sense, prediction method of wall thinning is willing to be introduced. Therefore, prediction model of FAC and LDI have been constructed in CRIEPI, and to utilize these models to actual plant piping management easily, prediction software 'FALSET' is developed. FALSET has equipped with essential function for pipe wall thinning management in power plants, as follows; (1) Information and condition input of plant piping system and its component, (2) Wall thinning rate evaluation with CRIEPI's FAC/LDI prediction model, (3) Loading of wall thickness measurement data files and graphics of data trend, (4) Residual lifetime evaluation considering both measured and predicted thinning rate, (5) Statistical process and graphics of thinning rate and residual lifetime for multi-piping systems. With further verification and improvement of each function, there will be a perspective for this FALSET to be utilized as a management tool in power plants. (author)

  4. Determination of leakage areas in nuclear piping

    International Nuclear Information System (INIS)

    Keim, E.

    1997-01-01

    For the design and operation of nuclear power plants the Leak-Before-Break (LBB) behavior of a piping component has to be shown. This means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safely detectable by a suitable monitoring system. The LBB-concept of Siemens/KWU is based on computer codes for the evaluation of critical crack lengths, crack openings, leakage areas and leakage rates, developed by Siemens/KWU. In the experience with the leak rate program is described while this paper deals with the computation of crack openings and leakage areas of longitudinal and circumferential cracks by means of fracture mechanics. The leakage areas are determined by the integration of the crack openings along the crack front, considering plasticity and geometrical effects. They are evaluated with respect to minimum values for the design of leak detection systems, and maximum values for controlling jet and reaction forces. By means of fracture mechanics LBB for subcritical cracks has to be shown and the calculation of leakage areas is the basis for quantitatively determining the discharge rate of leaking subcritical through-wall cracks. The analytical approach and its validation will be presented for two examples of complex structures. The first one is a pipe branch containing a circumferential crack and the second one is a pipe bend with a longitudinal crack

  5. Determination of leakage areas in nuclear piping

    Energy Technology Data Exchange (ETDEWEB)

    Keim, E. [Siemens/KWU, Erlangen (Germany)

    1997-04-01

    For the design and operation of nuclear power plants the Leak-Before-Break (LBB) behavior of a piping component has to be shown. This means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safely detectable by a suitable monitoring system. The LBB-concept of Siemens/KWU is based on computer codes for the evaluation of critical crack lengths, crack openings, leakage areas and leakage rates, developed by Siemens/KWU. In the experience with the leak rate program is described while this paper deals with the computation of crack openings and leakage areas of longitudinal and circumferential cracks by means of fracture mechanics. The leakage areas are determined by the integration of the crack openings along the crack front, considering plasticity and geometrical effects. They are evaluated with respect to minimum values for the design of leak detection systems, and maximum values for controlling jet and reaction forces. By means of fracture mechanics LBB for subcritical cracks has to be shown and the calculation of leakage areas is the basis for quantitatively determining the discharge rate of leaking subcritical through-wall cracks. The analytical approach and its validation will be presented for two examples of complex structures. The first one is a pipe branch containing a circumferential crack and the second one is a pipe bend with a longitudinal crack.

  6. Fluid structure interaction in piping systems

    Energy Technology Data Exchange (ETDEWEB)

    Svingen, Bjoernar

    1996-12-31

    The Dr. ing. thesis relates to an analysis of fluid structure interaction in piping systems in the frequency domain. The governing equations are the water hammer equations for the liquid, and the beam-equations for the structure. The fluid and structural equations are coupled through axial stresses and fluid continuity relations controlled by the contraction factor (Poisson coupling), and continuity and force relations at the boundaries (junction coupling). A computer program has been developed using the finite element method as a discretization technique both for the fluid and for the structure. This is made for permitting analyses of large systems including branches and loops, as well as including hydraulic piping components, and experiments are executed. Excitations are made in a frequency range from zero Hz and up to at least one thousand Hz. Frequency dependent friction is modelled as stiffness proportional Rayleigh damping both for the fluid and for the structure. With respect to the water hammer equations, stiffness proportional damping is seen as an artificial (bulk) viscosity term. A physical interpretation of this term in relation to transient/oscillating hydraulic pipe-friction is given. 77 refs., 72 figs., 4 tabs.

  7. Design criteria for piping and nozzles program

    International Nuclear Information System (INIS)

    Moore, S.E.; Bryson, J.W.

    1977-01-01

    This report reviews the activities and accomplishments of the Design Criteria for Piping and Nozzles program being conducted by the Oak Ridge National Laboratory for the period July 1, 1975, to September 30, 1976. The objectives of the program are to conduct integrated experimental and analytical stress analysis studies of piping system components and isolated and closely-spaced pressure vessel nozzles in order to confirm and/or improve the adequacy of structural design criteria and analytical methods used to assure the safe design of nuclear power plants. Activities this year included the development of a finite-element program for analyzing two closely spaced nozzles in a cylindrical pressure vessel; a limited-parameter study of vessels with isolated nozzles, finite-element studies of piping elbows, a fatigue test of an out-of-round elbow, summary and evaluation of experimental studies on the elastic-response and fatigue failure of tees, parameter studies on the behavior of flanged joints, publication of fifteen topical reports and papers on various experimental and analytical studies; and the development and acceptance of a number of design rules changes to the ASME Code. 2 figures, 2 tables

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-15

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

  10. Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities

    International Nuclear Information System (INIS)

    Lin, Chi-Wen; Antaki, G.; Bandyopadhyay, K.; Bush, S.H.; Costantino, C.; Kennedy, R.

    1995-01-01

    This paper presents the seismic design and evaluation guidelines for underground piping for the Department of Energy (DOE) High-Level-Waste (HLW) Facilities. The underground piping includes both single and double containment steel pipes and concrete pipes with steel lining, with particular emphasis on the double containment piping. The design and evaluation guidelines presented in this paper follow the generally accepted beam-on-elastic-foundation analysis principle and the inertial response calculation method, respectively, for piping directly in contact with the soil or contained in a jacket. A standard analysis procedure is described along with the discussion of factors deemed to be significant for the design of the underground piping. The following key considerations are addressed: the design feature and safety requirements for the inner (core) pipe and the outer pipe; the effect of soil strain and wave passage; assimilation of the necessary seismic and soil data; inertial response calculation for the inner pipe; determination of support anchor movement loads; combination of design loads; and code comparison. Specifications and justifications of the key parameters used, stress components to be calculated and the allowable stress and strain limits for code evaluation are presented

  11. Evolution behavior of nanohardness after thermal-aging and hydrogen-charging on austenite and strain-induced martensite in pre-strained austenitic stainless steel

    Science.gov (United States)

    Zheng, Yuanyuan; Zhou, Chengshuang; Hong, Yuanjian; Zheng, Jinyang; Zhang, Lin

    2018-05-01

    Nanoindentation has been used to study the effects of thermal-aging and hydrogen on the mechanical property of the metastable austenitic stainless steel. Thermal-aging at 473 K decreases the nanohardness of austenite, while it increases the nanohardness of strain-induced ɑ‧ martensite. Hydrogen-charging at 473 K increases the nanohardness of austenite, while it decreases the nanohardness of strain-induced ɑ‧ martensite. The opposite effect on austenite and ɑ‧ martensite is first found in the same pre-strained sample. This abnormal evolution behavior of hardness can be attributed to the interaction between dislocation and solute atoms (carbon and hydrogen). Carbon atoms are difficult to move and redistribute in austenite compared with ɑ‧ martensite. Therefore, the difference in the diffusivity of solute atoms between austenite and ɑ‧ martensite may result in the change of hardness.

  12. Evaluation of the plastic characteristics of piping products in relation to ASME code criteria

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.; Moore, S.E.

    1978-07-01

    Theories and test data relevant to the plastic characteristics of piping products are presented and compared with Code Equations in NB-3652 for Class 1 piping; in NC/ND-3652.2 for Class 2 and Class 3 piping. Comparisons are made for (a) straight pipe, (b) elbows, (c) branch connections, and (d) tees. The status of data (or lack of data) for other piping components is discussed. Comparisons are made between available data and the Code equations for two typical piping materials, SA106 Grade B and SA312 TP304, for Code Design Limits, and Service Limits A, B, C, and D. Conditions under which the Code Limits cannot be shown to be conservative from available data are pointed out. Based on the results of the study, recommendations for Code revisions are presented, along with recommendations for additional work

  13. Pulsatile turbulent flow through pipe bends at high Dean and Womersley numbers

    Science.gov (United States)

    Kalpakli, Athanasia; Örlü, Ramis; Tillmark, Nils; Alfredsson, P. Henrik

    2011-12-01

    Turbulent pulsatile flows through pipe bends are prevalent in internal combustion engine components which consist of bent pipe sections and branching conduits. Nonetheless, most of the studies related to pulsatile flows in pipe bends focus on incompressible, low Womersley and low Dean number flows, primarily because they aim in modeling blood flow, while internal combustion engine related flows have mainly been addressed in terms of integral quantities and consist of single point measurements. The present study aims at bridging the gap between these two fields by means of time-resolved stereoscopic particle image velocimetry measurements in a pipe bend with conditions that are close to those encountered in exhaust manifolds. The time/phase-resolved three-dimensional cross-sectional flow-field 3 pipe diameters downstream the pipe bend is captured and the interplay between different secondary motions throughout a pulse cycle is discussed.

  14. Pulsatile turbulent flow through pipe bends at high Dean and Womersley numbers

    International Nuclear Information System (INIS)

    Kalpakli, Athanasia; Örlü, Ramis; Tillmark, Nils; Alfredsson, P Henrik

    2011-01-01

    Turbulent pulsatile flows through pipe bends are prevalent in internal combustion engine components which consist of bent pipe sections and branching conduits. Nonetheless, most of the studies related to pulsatile flows in pipe bends focus on incompressible, low Womersley and low Dean number flows, primarily because they aim in modeling blood flow, while internal combustion engine related flows have mainly been addressed in terms of integral quantities and consist of single point measurements. The present study aims at bridging the gap between these two fields by means of time-resolved stereoscopic particle image velocimetry measurements in a pipe bend with conditions that are close to those encountered in exhaust manifolds. The time/phase-resolved three-dimensional cross-sectional flow-field 3 pipe diameters downstream the pipe bend is captured and the interplay between different secondary motions throughout a pulse cycle is discussed.

  15. Development of Austenitic ODS Strengthened Alloys for Very High Temperature Applications

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Robertson, Ian [Kyushu Univ. (Japan); Sehitoglu, Huseyin [Univ. of Illinois, Urbana-Champaign, IL (United States); Sofronis, Petros [Kyushu Univ. (Japan); Gewirth, Andrew [Kyushu Univ. (Japan)

    2015-04-22

    This “Blue Sky” project was directed at exploring the opportunities that would be gained by developing Oxide Dispersion Strengthened (ODS) alloys based on the Fe-Cr-Ni austenitic alloy system. A great deal of research effort has been directed toward ferritic and ferritic/martensitic ODS alloys which has resulted in reasonable advances in alloy properties. Similar gains should be possible with austenitic alloy which would also take advantage of other superior properties of that alloy system. The research effort was aimed at the developing an in-depth understanding of the microstructural-level strengthening effects of ODS particles in austentic alloys. This was accomplished on a variety of alloy compositions with the main focus on 304SS and 316SS compositions. A further goal was to develop an understanding other the role of ODS particles on crack propagation and creep performance. Since these later two properties require bulk alloy material which was not available, this work was carried out on promising austentic alloy systems which could later be enhanced with ODS strengthening. The research relied on a large variety of micro-analytical techniques, many of which were available through various scientific user facilities. Access to these facilities throughout the course of this work was instrumental in gathering complimentary data from various analysis techniques to form a well-rounded picture of the processes which control austenitic ODS alloy performance. Micromechanical testing of the austenitic ODS alloys confirmed their highly superior mechanical properties at elevated temperature from the enhanced strengthening effects. The study analyzed the microstructural mechanisms that provide this enhanced high temperature performance. The findings confirm that the smallest size ODS particles provide the most potent strengthening component. Larger particles and other thermally- driven precipitate structures were less effective contributors and, in some cases, limited

  16. Effects of LWR coolant environments on fatigue lives of austenitic stainless steels

    International Nuclear Information System (INIS)

    Chopra, O.K.; Gavenda, D.J.

    1997-01-01

    The ASME Boiler and Pressure Vessel Code fatigue design curves for structural materials do not explicitly address the effects of reactor coolant environments on fatigue life. Recent test data indicate a significant decrease in fatigue life of pressure vessel and piping materials in light water reactor (LWR) environments. Fatigue tests have been conducted on Types 304 and 316NG stainless steel in air and LWR environments to evaluate the effects of various material and loading variables, e.g., steel type, strain rate, dissolved oxygen (DO) in water, and strain range, on fatigue lives of these steels. The results confirm the significant decrease in fatigue life in water. The environmentally assisted decrease in fatigue life depends both on strain rate and DO content in water. A decrease in strain rate from 0.4 to 0.004%/s decreases fatigue life by a factor of ∼ 8. However, unlike carbon and low-alloy steels, environmental effects are more pronounced in low-DO than in high-DO water. At ∼ 0.004%/s strain rate, reduction in fatigue life in water containing <10 ppb D is greater by a factor of ∼ 2 than in water containing ≥ 200 ppb DO. Experimental results have been compared with estimates of fatigue life based on the statistical model. The formation and growth of fatigue cracks in austenitic stainless steels in air and LWR environments are discussed

  17. Waste pipe calculus extensions

    International Nuclear Information System (INIS)

    O'Connell, W.J.

    1979-01-01

    The waste pipe calculus provides a rapid method, using Laplace transforms, to calculate the transport of a pollutant such as nuclear waste, by a network of one-dimensional flow paths. The present note extends previous work as follows: (1) It provides an alternate approximation to the time-domain function (inverse Laplace transform) for the resulting transport. This algebraic approximation may be viewed as a simpler and more approximate model of the transport process. (2) It identifies two scalar quantities which may be used as summary consequence measures of the waste transport (or inversely, waste retention) system, and provides algebraic expressions for them. (3) It includes the effects of radioactive decay on the scalar quantity results, and further provides simplifying approximations for the cases of medium and long half-lives. This algebraic method can be used for quick approximate analyses of expected results, uncertainty and sensitivity, in evaluating selection and design choices for nuclear waste disposal systems

  18. Solar chemical heat pipe

    International Nuclear Information System (INIS)

    Levy, M.; Levitan, R.; Rosin, H.; Rubin, R.

    1991-08-01

    The performance of a solar chemical heat pipe was studied using CO 2 reforming of methane as a vehicle for storage and transport of solar energy. The endothermic reforming reaction was carried out in an Inconel reactor, packed with a Rh catalyst. The reactor was suspended in an insulated box receiver which was placed in the focal plane of the Schaeffer Solar Furnace of the Weizman Institute of Science. The exothermic methanation reaction was run in a 6-stage adiabatic reactor filled with the same Rh catalyst. Conversions of over 80% were achieved for both reactions. In the closed loop mode the products from the reformer and from the metanator were compressed into separate storage tanks. The two reactions were run either separately or 'on-line'. The complete process was repeated for over 60 cycles. The overall performance of the closed loop was quite satisfactory and scale-up work is in progress in the Solar Tower. (authors). 35 refs., 2 figs

  19. Subsea pipe dream

    Energy Technology Data Exchange (ETDEWEB)

    Balcombe, Mark

    1988-09-22

    The Gulf of Mexico is famous today mainly for the ferocity of its hurricanes. But for anyone in the oil industry, it is also known for the vast array of oil pipelines that criss-cross its stormy waters, and for the large number of pipeline-laying barges which install them. Soon many of these vessels could be steaming to British waters - not to escape the weather, but to cash in on a bonanza of pipe-laying activity which could soon take place offshore northern Europe. The construction of new pipelines off the UK, Norway and Netherlands will, however, present a new range of challenges for pipeline designers and builders. First and foremost is the Piper Alpha platform disaster, which could saddle the UK offshore industry with a Pound 500 million-plus bill for the installation of emergency shutdown valves (ESVs) on existing lines.

  20. Investigation on field method using strain measurement on pipe surface to measure pressure pulsation in piping systems

    International Nuclear Information System (INIS)

    Maekawa, Akira; Tsuji, Takashi; Takahashi, Tsuneo; Kato, Minoru

    2013-01-01

    Accurate evaluation of the occurrence location and amplitude of pressure pulsations in piping systems can lead to efficient plant maintenance by preventing fatigue failure of piping and components because the pulsations can be one of the main causes of vibration fatigue and acoustic noise in piping. A non-destructive field method to measure pressure pulsations easily and directly was proposed to replace conventional methods such as prediction using numerical simulations and estimation using locally installed pressure gauges. The proposed method was validated experimentally by measuring pulsating flow in a mock-up piping system. As a result, it was demonstrated that the method to combine strain measurement on the outer surface of pipe with the formula for thick-walled cylinders could measure amplitudes and behavior of the pressure pulsations with a practical accuracy. Factors affecting the measurement accuracy of the proposed method were also discussed. Furthermore, the applicability of the formula for thin-walled cylinders was examined for variously shaped pipes. (author)

  1. Grain size control method for the nozzles of AP1000 primary coolant pipes

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shenglong [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China); Sun, Yanhui [Collaborative Innovation Center of Steel Technology, University of Science & Technology Beijing, Beijing 100083 (China); Yang, Bin, E-mail: byang@ustb.edu.cn [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China); Collaborative Innovation Center of Steel Technology, University of Science & Technology Beijing, Beijing 100083 (China); Zhang, Mingxian [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China)

    2017-04-01

    Highlights: • Design a new forging technology for AP1000 primary coolant pipe. • Method combining FEM and scale-down experiments is adopted. • The grain size and distribution in simulation and experiment are consistent. • Get optimal forging parameters for production guiding. - Abstract: AP1000 primary coolant pipe is made of 316LN austenitic stainless steel. It is a large special-shaped pipe manufactured by integral forging technology. Owing to non-uniform temperature and deformation during forging, coarse grains often occur in the boss sections of the pipe especially in the nozzles’ parts. In the present study, a new forging technology was proposed to control the grain size. The finite element method was used to optimize the forging speed and friction coefficient, then the scale-down experiments were performed for comparison. The forging speed is suggested to be less than 20 mm/s, and effective lubricants should be used to decrease the friction coefficient. The errors of the grain size between the experiment and simulation are less than 20%.

  2. Grain size control method for the nozzles of AP1000 primary coolant pipes

    International Nuclear Information System (INIS)

    Wang, Shenglong; Sun, Yanhui; Yang, Bin; Zhang, Mingxian

    2017-01-01

    Highlights: • Design a new forging technology for AP1000 primary coolant pipe. • Method combining FEM and scale-down experiments is adopted. • The grain size and distribution in simulation and experiment are consistent. • Get optimal forging parameters for production guiding. - Abstract: AP1000 primary coolant pipe is made of 316LN austenitic stainless steel. It is a large special-shaped pipe manufactured by integral forging technology. Owing to non-uniform temperature and deformation during forging, coarse grains often occur in the boss sections of the pipe especially in the nozzles’ parts. In the present study, a new forging technology was proposed to control the grain size. The finite element method was used to optimize the forging speed and friction coefficient, then the scale-down experiments were performed for comparison. The forging speed is suggested to be less than 20 mm/s, and effective lubricants should be used to decrease the friction coefficient. The errors of the grain size between the experiment and simulation are less than 20%.

  3. Beam loss reduction by magnetic shielding using beam pipes and bellows of soft magnetic materials

    Science.gov (United States)

    Kamiya, J.; Ogiwara, N.; Hotchi, H.; Hayashi, N.; Kinsho, M.

    2014-11-01

    One of the main sources of beam loss in high power accelerators is unwanted stray magnetic fields from magnets near the beam line, which can distort the beam orbit. The most effective way to shield such magnetic fields is to perfectly surround the beam region without any gaps with a soft magnetic high permeability material. This leads to the manufacture of vacuum chambers (beam pipes and bellows) with soft magnetic materials. A Ni-Fe alloy (permalloy) was selected for the material of the pipe parts and outer bellows parts, while a ferritic stainless steel was selected for the flanges. An austenitic stainless steel, which is non-magnetic material, was used for the inner bellows for vacuum tightness. To achieve good magnetic shielding and vacuum performances, a heat treatment under high vacuum was applied during the manufacturing process of the vacuum chambers. Using this heat treatment, the ratio of the integrated magnetic flux density along the beam orbit between the inside and outside of the beam pipe and bellows became small enough to suppress beam orbit distortion. The outgassing rate of the materials with this heat treatment was reduced by one order magnitude compared to that without heat treatment. By installing the beam pipes and bellows of soft magnetic materials as part of the Japan Proton Accelerator Research Complex 3 GeV rapid cycling synchrotron beam line, the closed orbit distortion (COD) was reduced by more than 80%. In addition, a 95.5% beam survival ratio was achieved by this COD improvement.

  4. Characterization of microstructures in austenitic stainless steels by ultrasonics

    International Nuclear Information System (INIS)

    Raj, Baldev; Palanichamy, P.; Jayakumar, T.; Kumar, Anish; Vasudevan, M.; Shankar, P.

    2000-01-01

    Recently, many nondestructive techniques have been considered for microstructural characterization of materials to enable in-situ component assessment for pre-service quality and in-service performance. Ultrasonic parameters have been used for estimation of average grain size, evaluation of recrystallization after cold working, and characterization of Cr2N precipitation during thermal aging in different grades of austenitic stainless steels. Ultrasonic first back wall echo signals were obtained from several specimens of AISI type 316 stainless steel with different grain sizes. Shift in the spectral peak frequency and the change in the full width at half maximum of the autopower spectrum of the first back wall echo are correlated with the grain size in the range 30-150 microns. The advantages of this method are: (i) independence of variation in couplant conditions (ii), applicable even to highly attenuating materials, (iii) direct correlation of the ultrasonic parameters with yield strength and (iv) suitability for shop-floor applications. Recrystallization behavior (temperature range 973-1173 K and time durations 0.5-1000 h) of cold worked titanium modified 316 stainless steel (D9) has been characterized using ultrasonic velocity measurements. A velocity parameter derived using a combination of shear and longitudinal wave velocities is correlated with the degree of recrystallization. These velocity measurement could also identify onset, progress and completion of recrystallization more accurately as compared to hardness and strength measurements. Ultrasonic velocity measurements were performed in thermally aged (at 1123 K for 10 to 2000 h) nuclear grade 316 LN stainless steel. Change in velocity due to thermal aging treatment could be used to reveal the formation of (i) Cr-N clusters associated with high lattice strains, (ii) coherent Cr2N precipitation, (iii) loss of coherency and (iv) growth of incoherent Cr2N precipitates. Microstructural characterization by

  5. Superconducting pipes and levitating magnets.

    Science.gov (United States)

    Levin, Yan; Rizzato, Felipe B

    2006-12-01

    Motivated by a beautiful demonstration of the Faraday and the Lenz laws in which a small neodymium magnet falls slowly through a conducting nonferromagnetic tube, we consider the dynamics of a magnet falling coaxially through a superconducting pipe. Unlike the case of normal conducting pipes, in which the magnet quickly reaches the terminal velocity, inside a superconducting tube the magnet falls freely. On the other hand, to enter the pipe the magnet must overcome a large electromagnetic energy barrier. For sufficiently strong magnets, the barrier is so large that the magnet will not be able to penetrate it and will be levitated over the mouth of the pipe. We calculate the work that must done to force the magnet to enter a superconducting tube. The calculations show that superconducting pipes are very efficient at screening magnetic fields. For example, the magnetic field of a dipole at the center of a short pipe of radius a and length L approximately > a decays, in the axial direction, with a characteristic length xi approximately 0.26a. The efficient screening of the magnetic field might be useful for shielding highly sensitive superconducting quantum interference devices. Finally, the motion of the magnet through a superconducting pipe is compared and contrasted to the flow of ions through a trans-membrane channel.

  6. Dictionary of pressure vessel and piping technology

    International Nuclear Information System (INIS)

    Jentgen, L.; Schmitz, H.P.

    1986-01-01

    A specialised dictionary has been compiled containing the appropriate English and German terms in the following technical fields: materials science, welding, destructive and non-destructive testing, thermal and mass transfer, the design and construction in particular of pressure vessels, tanks, heat exchangers, piping, expansion joints, valves, and components associated with the above fields. This dictionary is the result of many years spent in evaluating technical terminology from the relevant American and British regulations, technical rules, standards, and specifications (see bibliography) and correlating these with the terminology of comparable German regulations, rules and standards, together with the essential technical literature. (orig.) [de

  7. Methodic recommendations on ultrasonic testing of pipeline austenitic butt joints

    International Nuclear Information System (INIS)

    Grebennik, V.S.; Lantukh, V.M.; Tajts, M.Z.; Ermolov, I.N.; Volkov, A.S.; Vyatskov, I.A.; Kesler, N.A.; Shchedrin, I.F.

    1989-01-01

    Recommendations for the application of ultrasonic testing of austenitic welded joints of the Du 500 pipelines with the walls 32-34 mm thick made of steel Kh18N10T are developed. The optimal values of the main parameters of ultrasonic testing are determined experimentally. Principles of calculation of the optimal parameters are considered. 1 ref.; 4 figs

  8. On the elusive crystal structure of expanded austenite

    DEFF Research Database (Denmark)

    Brink, Bastian; Ståhl, Kenny; Christiansen, Thomas Lundin

    2017-01-01

    No consistent structural description exists for expanded austenite that accurately accounts for the hkl-dependent peak shifts and broadening observed in diffraction experiments. The best available description for homogeneous samples is a face-centered cubic lattice with stacking faults. Here Deby...

  9. To the corrosion of austenitic steels in sodium loops

    International Nuclear Information System (INIS)

    Schad, M.

    1978-03-01

    This report describes the comparison between experimental corrosion and calculated corrosion effects on austenitic steels exposed to liquid sodium. As basis for the calculations served a diffusion model. The comparison showed that the model is able to predict the corrosion effects. In addition the model was used to calculate the corrosion effect along an actual fuel rod. (orig.) [de

  10. Phase Transformation of Metastable Austenite in Steel during Nano indentation

    International Nuclear Information System (INIS)

    Ahn, Taehong; Lee, Sung Bo; Han, Heung Nam; Park, Kyungtae

    2013-01-01

    These can produce geometrical softening accompanied by a sudden displacement excursion during load-controlled nanoindentation, which referred to in the literature as a pop-in. In this study, phase transformation of metastable austenite to stress-induced ε martensite which causes pop-ins during nanoindentation of steel will be reported and discussed. This study investigated the relationship between pop-in behavior of austenite in the early stage of nanoindentation and formation of ε martensite based on microstructural analyses. The load-displacement curve obtained from nanoindentation revealed stepwise pop-ins in the early stage of plastic deformation. From analyses of high resolution TEM images, a cluster of banded structure under the indent turned out a juxtaposition of (111) planes of γ austenite and (0001) planes of ε martensite. The calculation of displacement along indentation axis for (111) slip system by formation of ε martensite showed that geometrical softening can also occur by ε martensite formation when considering that the stress-induced ε martensite transformation is the predominant deformation mode in the early stage of plastic deformation and its monopartial nature as well. These microstructural investigations strongly suggest that the pop-in behavior in the early stage of plastic deformation of austenite is closely related to the formation of ε martensite

  11. Phase Transformation of Metastable Austenite in Steel during Nano indentation

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Taehong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Sung Bo; Han, Heung Nam [Seoul National Univ., Seoul (Korea, Republic of); Park, Kyungtae [Hanbat National Univ., Daejeon (Korea, Republic of)

    2013-05-15

    These can produce geometrical softening accompanied by a sudden displacement excursion during load-controlled nanoindentation, which referred to in the literature as a pop-in. In this study, phase transformation of metastable austenite to stress-induced ε martensite which causes pop-ins during nanoindentation of steel will be reported and discussed. This study investigated the relationship between pop-in behavior of austenite in the early stage of nanoindentation and formation of ε martensite based on microstructural analyses. The load-displacement curve obtained from nanoindentation revealed stepwise pop-ins in the early stage of plastic deformation. From analyses of high resolution TEM images, a cluster of banded structure under the indent turned out a juxtaposition of (111) planes of γ austenite and (0001) planes of ε martensite. The calculation of displacement along indentation axis for (111) slip system by formation of ε martensite showed that geometrical softening can also occur by ε martensite formation when considering that the stress-induced ε martensite transformation is the predominant deformation mode in the early stage of plastic deformation and its monopartial nature as well. These microstructural investigations strongly suggest that the pop-in behavior in the early stage of plastic deformation of austenite is closely related to the formation of ε martensite.

  12. Effect of shot peening on metastable austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Fargas, G., E-mail: gemma.fargas@upc.edu [CIEFMA - Departament de Ciència dels Materials i Enginyeria Metallúrgica, Universitat Politècnica de Catalunya, 08028 Barcelona (Spain); CRnE, Centre de Recerca en Nanoenginyeria, Universitat Politècnica de Catalunya, 08028 Barcelona (Spain); Roa, J.J.; Mateo, A. [CIEFMA - Departament de Ciència dels Materials i Enginyeria Metallúrgica, Universitat Politècnica de Catalunya, 08028 Barcelona (Spain); CRnE, Centre de Recerca en Nanoenginyeria, Universitat Politècnica de Catalunya, 08028 Barcelona (Spain)

    2015-08-12

    In this work, shot peening was performed in a metastable austenitic stainless steel EN 1.4318 (AISI 301LN) in order to evaluate its effect on austenite to martensite phase transformation and also the influence on the fatigue limit. Two different steel conditions were considered: annealed, i.e., with a fully austenitic microstructure, and cold rolled, consisting of a mixture of austenite and martensite. X-ray diffraction, electron back-scattered diffraction and focus ion beam, as well as nanoindentation techniques, were used to elucidate deformation mechanisms activated during shot peening and correlate with fatigue response. Results pointed out that extensive plastic deformation and phase transformation developed in annealed specimens as a consequence of shot peening. However, the increase of roughness and the generation of microcracks led to a limited fatigue limit improvement. In contrast, shot peened cold rolled specimens exhibited enhanced fatigue limit. In the latter case, the main factor that determined the influence on the fatigue response was the distance from the injector, followed successively by the exit speed of the shots and the coverage factor.

  13. Carbon diffusion in carbon-supersaturated ferrite and austenite

    Czech Academy of Sciences Publication Activity Database

    Čermák, Jiří; Král, Lubomír

    2014-01-01

    Roč. 586, FEB (2014), s. 129-135 ISSN 0925-8388 R&D Projects: GA ČR(CZ) GAP108/11/0148; GA MŠk(CZ) ED1.1.00/02.0068 Institutional support: RVO:68081723 Keywords : carbon diffusion * Carbon supersaturation * Carbon supersaturation * Ferrite * Austenite Subject RIV: BJ - Thermodynamics Impact factor: 2.999, year: 2014

  14. Factors which determine the swelling rate of austenitic stainless steels

    International Nuclear Information System (INIS)

    Garner, F.A.; Wolfer, W.G.

    1983-01-01

    Once void nucleation subsides, the swelling rate of many austenitic alloys becomes rather insensitive to variables that control the transient regime of swelling. Models are presented which describe the roles of nickel, chromium and silicon in void nucleation. The relative insensitivity of steady-state swelling to temperature, displacement rate and composition is also discussed

  15. 46 CFR 56.01-3 - Power boilers, external piping and appurtenances (Replaces 100.1.1, 100.1.2, 122.1, 132 and 133).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Power boilers, external piping and appurtenances... boilers, external piping and appurtenances (Replaces 100.1.1, 100.1.2, 122.1, 132 and 133). (a) Power boiler external piping and components must meet the requirements of this part and §§ 52.01-105, 52.01-110...

  16. Characterization of the dissimilar welding - austenitic stainless steel with filler metal of the nickel alloy

    International Nuclear Information System (INIS)

    Soares, Bruno Amorim; Schvartzman, Monica Maria de Abreu Mendonca; Campos, Wagner Reis da Costa

    2007-01-01

    In elevated temperature environments, austenitic stainless steel and nickel alloy has a superior corrosion resistance due to its high Cr content. Consequently, this alloys is widely used in nuclear reactors components and others plants of energy generation that burn fossil fuel or gas, chemical and petrochemical industries. The object of the present work was to research the welding of AISI 304 austenitic stainless steel using the nickel alloy filler metals, Inconel 625. Gas tungsten arc welding, mechanical and metallographic tests, and compositional analysis of the joint were used. A fundamental investigation was undertaken to characterize fusion boundary microstructure and to better understand the nature and character of boundaries that are associated with cracking in dissimilar welds. The results indicate that the microstructure of the fusion zone has a dendritic structure, inclusions, and precipitated phases containing Ti and Nb are present in the inter-dendritic region. In some parts near to the fusion line it can be seen a band in the weld, probably a eutectic phase with lower melting point than the AISI 304, were the cracking may be beginning by stress corrosion. (author)

  17. Overview of microstructural evolution in neutron-irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1993-01-01

    Austenitic stainless steels are important structural materials common to several different reactor systems, including light water and fast breeder fission, and magnetic fusion reactors (LWR, FBR, and MFR, respectively). The microstructures that develop in 300 series austenitic stainless steels during neutron irradiation at 60-700 C include combinations of dislocation loops and networks, bubbles and voids, and various kinds of precipitate phases (radiation-induced, or -enhanced or -modified thermal phases). Many property changes in these steels during neutron irradiation are directly or indirectly related to radiation-induced microstructural evolution. Even more important is the fact that radiation-resistance of such steels during either FBR or MFR irradiation is directly related to control of the evolving microstructure during such irradiation. The purpose of this paper is to provide an overview of the large and complex body of data accumulated from various fission reactor irradiation experiments conducted over the many years of research on microstructural evolution in this family of steels. The data can be organized into several different temperature regimes which then define the nature of the dominant microstructural components and their sensitivities to irradiation parameters (dose, helium/dpa ratio, dose rate) or metallurgical variables (alloy composition, pretreatment). The emphasis in this paper will be on the underlying mechanisms driving the microstructure to evolve during irradiation or those enabling microstructural stability related to radiation resistance. (orig.)

  18. Assessment of precipitates of isothermal aged austenitic stainless steel using measurement techniques of ultrasonic attenuation

    International Nuclear Information System (INIS)

    Kim, Hun Hee; Kim, Hak Joon; Song, Sung Jin; Lim, Byeong Soo; Kim, Kyung Cho

    2014-01-01

    AISI 316L stainless steel is widely used as a structural material of high temperature thermoelectric power plants, since austenitic stainless steel has excellent mechanical properties. However, creep damage is generated in these components, which are operated under a high temperature and high pressure environment. Several researches have been done on how microstructural changes of precipitates affect to the macroscopic mechanical properties. And they investigate the relation between ultrasonic parameters and metallurgical results. But, these studies are limited by experiment results only. In this paper, attenuations of ultrasonic with isothermal damaged AISI 316L stainless steel were measured. Also, simulation of ultrasonic attenuation with variation of area fraction and size of precipitates were performed. And, from the measured attenuations, metallographic data and simulation results, we investigate the relations between the ultrasonic attenuations and the material properties which is area fraction of precipitates for the isothermal damaged austenitic stainless steel specimens. And, we studied parametric study for investigation of the relation between ultrasonic parameters and metallurgical results of the isothermal damaged AISI 316L stainless steel specimens using numerical methods.

  19. Comparison of material property specifications of austenitic steels in fast breeder reactor technology

    International Nuclear Information System (INIS)

    Vanderborck, Y.; Van Mulders, E.

    1985-01-01

    Austenitic stainless steels are very widely used in components for European Fast Breeder Reactors. The Activity Group Nr.3 ''Materials'', within Working Group ''Codes and Standards'' of the Fast Reactor Co-Ordination Committee of the European Communities, has decided to initiate a study to compare the material property specifications of the austenitic stainless steel used in the European Fast Breeder Technology. Hence, this study would allow one to view rapidly the designation of a particular steel grade in different European countries and to compare given property values for a same grade. There were dissimilarities, differences or voids appear, it could lead to an attempt to complete and/or to uniformize the nationally given values, so that on a practical level interchangeability, availability and use ease design and construction work. A selection of the materials and of their properties has been made by the Working Group. Materials examined are Stainless Steel AISI 304, 304 L, 304 LN, 316, 316 L, 316 LN, 316''Ti stab.'', 316''Nb stab''., 321, 347

  20. Influence of laser shock peening on irradiation defects in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Qiaofeng [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Su, Qing [Nebraska Center for Energy Sciences Research, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Wang, Fei [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Zhang, Chenfei; Lu, Yongfeng [Department of Electrical Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nastasi, Michael [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Energy Sciences Research, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Materials and Nanoscience, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States); Cui, Bai, E-mail: bcui3@unl.edu [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Materials and Nanoscience, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States)

    2017-06-15

    The laser shock peening process can generate a dislocation network, stacking faults, and deformation twins in the near surface of austenitic stainless steels by the interaction of laser-driven shock waves with metals. In-situ transmission electron microscopy (TEM) irradiation studies suggest that these dislocations and incoherent twin boundaries can serve as effective sinks for the annihilation of irradiation defects. As a result, the irradiation resistance is improved as the density of irradiation defects in laser-peened stainless steels is much lower than that in untreated steels. After heating to 300 °C, a portion of the dislocations and stacking faults are annealed out while the deformation twins remain stable, which still provides improved irradiation resistance. These findings have important implications on the role of laser shock peening on the lifetime extension of austenitic stainless steel components in nuclear reactor environments. - Highlights: •Laser shock peening generates a dislocation network, stacking faults and deformation twins in stainless steels. •Dislocations and incoherent twin boundaries serve as effective sinks for the annihilation of irradiation defects. •Incoherent twin boundaries remain as stable and effective defect sinks at 300 °C.