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Sample records for alloy pressure tubes

  1. Development of Zirconium alloys (for pressure tubes)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Jung, Chung Hwan; Yim, Kyong Soo; Kim, Sung Soo; Baek, Jong Hyuk; Jeong, Yong Hwan; Kim, Kyong Ho; Cho, Hae Dong [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Hwang, S. K.; Kim, M. H. [Inha Univ., Incheon (Korea, Republic of); Kwon, S. I [Korea Univ., Seoul (Korea, Republic of); Kim, I. S. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

    1997-09-01

    The objective of this research is to set up the basic technologies for the evaluation of pressure tube integrity and to develop improved zirconium alloys to prevent pressure tube failures due to DHC and hydride blister caused by excessive creep-down of pressure tubes. The experimental procedure and facilities for characterization of pressure tubes were developed. The basic research related to a better understanding of the in-reactor performances of pressure tubes leads to noticeable findings for the first time : the microstructural effect on corrosion and hydrogen pick-up behavior of Zr-2.5Nb pressure tubes, texture effect on strength and DHC resistance and enhanced recrystallization by Fe in zirconium alloys and etc. Analytical methodology for the assessment of pressure tubes with surface flaws was set up. A joint research is being under way with AECL to determine the fracture toughness of O-8 at the EOL (End of Life) that had been quadruple melted and was taken out of the Wolsung Unit-1 after 10 year operation. In addition, pressure tube with texture controlled is being made along with VNINM in Russia as a joint project between KAERI and Russia. Finally, we succeeded in developing 4 different kinds of zirconium alloys with better corrosion resistance, low hydrogen pickup fraction and higher creep strength. (author). 121 refs., 65 tabs., 260 figs

  2. Development of heat treated Zr-2.5% Nb alloy tubes for pressure tubes

    International Nuclear Information System (INIS)

    Saibaba, N.; Jha, S.K.; Tonpe, S.

    2011-01-01

    Zr-2.5% Nb alloy is the candidate material for pressure tubes of Pressurized Heavy Water Reactors (PHWR), and are manufactured in cold working condition while heat treated pressure tubes are used in RBMK and FUGEN type of reactors. The diametral creep of these tubes is the life limiting factor. This paper presents the extensive work carried out for the optimization of process parameters to manufacture heat treated Zr-2.5% Nb pressure tubes. Extensive dilactometry study was carried out to establish the transus temperature for the alloy and the effect of soaking temperature and cooling rate on the microstructure was characterized. On the basis of the study, water quenching (at 883 deg C) in the a b region with 20-25% primary a phase was selected, further cold worked, aged and finally autoclaved. Mechanical properties of the finished tubes were found to be comparable to the cold worked route. Large number of full sized tubes of about 700 - 800 mm long was produced to establish the repeatability. (author)

  3. Delayed hydrogen cracking of zirconium alloy pressure tubes

    International Nuclear Information System (INIS)

    Jackman, A.H.; Dunn, J.T.

    1976-10-01

    After several years of almost continuous service, Pickering Units 3 and 4 have both experienced long outages to replace cracked pressure tubes. This report summarizes the status of the investigation into the cause of the cracks as of May 1976. The basic cause of the cracking was the presence of very high residual tensile stresses in the pressure tubes due to improper rolling procedures. These residual stresses are being reduced to acceptable levels by local stress relieving techniques at Bruce G.S. and in future reactors improvements in rolling procedures and changes in pressure tube specifications will prevent a recurrence of this problem. (author)

  4. Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

    Energy Technology Data Exchange (ETDEWEB)

    Chen, G.; Zhang, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Xu, D.K. [Environmental Corrosion Center, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Li, D.H. [Hunan Taohuajiang Nuclear Power Co., Ltd, Yiyang, 413000 (China); Chen, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Zhang, Z., E-mail: zhe.zhang@tju.edu.cn [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China)

    2017-06-15

    In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ{sub x} did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ{sub xa}. For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ{sub xa} and the internal pressure p{sub i}. The hoop ratcheting strain ɛ{sub θ} increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ{sub x} was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

  5. Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

    International Nuclear Information System (INIS)

    Chen, G.; Zhang, X.; Xu, D.K.; Li, D.H.; Chen, X.; Zhang, Z.

    2017-01-01

    In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ x did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ xa . For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ xa and the internal pressure p i . The hoop ratcheting strain ɛ θ increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ x was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

  6. Time-dependent leak behavior of flawed Alloy 600 tube specimens at constant pressure

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, Chi Bum, E-mail: bahn@anl.gov [Argonne National Laboratory, Argonne, IL 60439 (United States); Majumdar, Saurin [Argonne National Laboratory, Argonne, IL 60439 (United States); Harris, Charles [United States Nuclear Regulatory Commission, Rockville, MD 20852 (United States)

    2011-10-15

    Leak rate testing has been performed using Alloy 600 tube specimens with throughwall flaws. Some specimens have shown time-dependent leak behavior at constant pressure conditions. Fractographic characterization was performed to identify the time-dependent crack growth mechanism. The fracture surface of the specimens showed the typical features of ductile fracture, as well as the distinct crystallographic facets, typical of fatigue crack growth at low {Delta}K level. Structural vibration appears to have been caused by the oscillation of pressure, induced by a high-pressure pump used in a test facility, and by the water jet/tube structure interaction. Analyses of the leak behaviors and crack growth indicated that both the high-pressure pump and the water jet could significantly contribute to fatigue crack growth. To determine whether the fatigue crack growth during the leak testing can occur solely by the water jet effect, leak rate tests at constant pressure without the high-pressure pump need to be performed. - Highlights: > Leak rate of flawed Alloy 600 tubing increased at constant pressure condition. > Fractography revealed two cases: ductile tearing and crystallographic facets. > Crystallographic facets are typical features of fatigue crack growth at low {Delta}K. > Fatigue source could be water jet-induced vibration and/or high-pressure pump pulsation.

  7. Delayed hydride cracking behavior of Zr-2.5Nb alloy pressure tubes for PHWR700

    Energy Technology Data Exchange (ETDEWEB)

    Sunil, S.; Bind, A.K.; Khandelwal, H.K.; Singh, R.N., E-mail: rnsingh@barc.gov.in; Chakravartty, J.K.

    2015-11-15

    In order to attain improved in-reactor performance few prototypes pressure tubes of Zr-2.5Nb alloy were manufactured by employing forging to break the cast structure and to obtain more homogeneous microstructure. Both double forging and single forging were employed. The forged material was further processed by employing hot extrusion, cold pilgering and autoclaving. A detailed characterization in terms of mechanical properties and microstructure of the prototype tubes were carried for qualifying it for intended use as pressure tubes in PHWR700 reactors. In this work, Delayed Hydride Cracking (DHC) behavior of the forged Zr-2.5Nb pressure tube material characterized in terms of DHC velocity and threshold stress intensity factor associated with DHC (K{sub IH}) was compared with that of conventionally manufactured material in the temperature range of 200–283 °C. Activation energy associated with the DHC in this alloy was found to be ∼60 kJ/mol for the forged materials.

  8. Microstructure control of Zr-Nb-Sn alloy with Mo addition for HWR pressure tube application

    International Nuclear Information System (INIS)

    Hwang, S. K.; Kim, M. H.; Kim, J. H.; Kwon, S. I.; Kim, Y. S.

    1997-01-01

    As a basic research to develop the material for heavy water reactor pressure tube application the effect of Mo addition to Zr-Nb-Sn alloy was studied for the purpose of minimizing the amount of cold working while maintaining a high strength. To select the target alloy system we first designed various alloy compositions and chose Zr-Nb-Sn and Zr-Nb-Mo through multi-regression analysis of the relationship between the basic properties and the compositions. Plasma arc melting was used to produce the alloys and the microstructure change introduced by the processing steps including hot forging, beta-heat treatment, hot rolling, cold rolling and recrystallization heat treatment was investigated. Recrystallization of Zr-Nb-Sn was retarded by adding Mo and this resulted in a fine grain structure in Zr-Nb-Sn-Mo alloy. Beside the retarding effect recrystallization, Mo increased the amount of residual beta phase and showed an indication of precipitation hardening, which added up to the possibility of applying the alloy for the desired usage. (author)

  9. Evaluation of candidate Stirling engine heater tube alloys after 3500 hours exposure to high pressure doped hydrogen or helium

    Science.gov (United States)

    Misencik, J. A.; Titran, R. H.

    1984-01-01

    The heater head tubes of current prototype automotive Stirling engines are fabricated from alloy N-155, an alloy which contains 20 percent cobalt. Because the United States imports over 90 percent of the cobalt used in this country and resource supplies could not meet the demand imposed by automotive applications of cobalt in the heater head (tubes plus cylinders and regenerator housings), it is imperative that substitute alloys free of cobalt be identified. The research described herein focused on the heater head tubes. Sixteen alloys (15 potential substitutes plus the 20 percent Co N-155 alloy) were evaluated in the form of thin wall tubing in the NASA Lewis Research Center Stirling simulator materials diesel fuel fired test rigs. Tubes filled with either hydrogen doped with 1 percent CO2 or with helium at a gas pressure of 15 MPa and a temperature of 820 C were cyclic endurance tested for times up to 3500 hr. Results showed that two iron-nickel base superalloys, CG-27 and Pyromet 901 survived the 3500 hr endurance test. The remaining alloys failed by creep-rupture at times less than 3000 hr, however, several other alloys had superior lives to N-155. Results further showed that doping the hydrogen working fluid with 1 vol % CO2 is an effective means of reducing hydrogen permeability through all the alloy tubes investigated.

  10. Influence of hydrogen content on impact toughness of Zr-2.5Nb pressure tube alloy

    Energy Technology Data Exchange (ETDEWEB)

    Singh, R.N., E-mail: rnsingh@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Viswanathan, U.K.; Kumar, Sunil; Satheesh, P.M.; Anantharaman, S. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Stahle, P. [Division of Solid Mechanics, Lund University/LTH, SE22100 Lund (Sweden)

    2011-07-15

    Highlights: > For the first time impact behaviour of Zr-2.5Nb pressure tube material used in Indian Pressurized Heavy Water Reactor (IPHWR) as a function of hydrogen content and temperature is being reported. > The critical hydrogen concentration to cause low energy fracture at 25 and 200 deg. C is suggested. > The impact behaviour is rationalized in terms of hydrogen content, test temperature, microstructural features and state of stress ahead of a crack. - Abstract: Influence of hydrogen content on the impact toughness of Zr-2.5% Nb alloy was examined by carrying out instrumented drop weight tests in the temperature range of 25-250 deg. C using curved Charpy specimens fabricated from unirradiated pressure tubes of Indian Pressurized Heavy Water Reactor (IPHWR). Hydrogen content of the samples was between 10 and 170 ppm by weight (wppm). Sharp ductile-to-brittle-transition behaviour was demonstrated by hydrided materials. The temperature for the onset of transition increased with the increase in the hydrogen content of the specimens. The fracture surfaces of unhydrided specimen exhibited ductile fracture caused by micro void coalescence and tear ridges at lower temperatures and by fibrous fracture at intermediate and at higher temperatures. Except for the samples tested at the upper shelf energy levels, the fracture surfaces of all hydrided samples were suggestive of hydride assisted failure. In most cases the transverse cracks observed in the fracture path matched well with the hydride precipitate distribution and orientation.

  11. The elastic properties of zirconium alloy fuel cladding and pressure tubing materials

    International Nuclear Information System (INIS)

    Rosinger, H.E.; Northwood, D.O.

    1979-01-01

    A knowledge of the elastic properties of zirconium alloys is required in the mathematical modelling of cladding and pressure tubing performance. Until recently, little of this type of data was available, particularly at elevated temperatures. The dynamic elastic moduli of zircaloy-2, zircaloy-4, the alloys Zr-1.0 wt%Nb, Zr-2.5 wt%Nb and Marz grade zirconium have therefore been determined over the temperature range 275 to 1000 K. Young's modulus and shear modulus for all the zirconium alloys decrease with temperature and are expressed by empirical relations fitted to the data. The elastic properties are texture dependent and a detailed study has been conducted on the effect of texture on the elastic properties of Zr-1.0 wt% Nb over the temperature range 275 to 775 K. The results are compared with polycrystalline elastic constants computed from single crystal elastic constants, and the effect of texture on the dynamic elastic moduli is discussed in detail. (Auth.)

  12. Microstructural characterization and mechanical properties of Excel alloy pressure tube material

    Science.gov (United States)

    Sattari, Mohammad

    Microstructural characterization and mechanical properties of Excel (Zr-3.5%Sn-0.8%Mo-0.8%Nb), a dual phase alphaZr -hcp and betaZr-bcc pressure tube material, is discussed in the current study which is presented in manuscript format. Chapter 3 discusses phase transformation temperatures using different techniques such as quantitative metallography, differential scanning calorimetry (DSC), and electrical resistivity. It was found that the alphaZr → alphaZr+beta Zr and alphaZr+betaZr → betaZr transformation temperatures are in the range of 600-690°C and 960-970°C respectively. Also it was observed that upon quenching from temperatures below ˜860°C the martensitic transformation of betaZr to alpha'--hcp is halted and instead the microstructure transforms into retained Zr with o hexagonal precipitates inside betaZr grains. Chapter 4 deals with aging response of Excel alloy. Precipitation hardening was observed in samples water-quenched from high in the alphaZr+beta Zr or betaZr regions followed by aging. The optimum aging conditions were found to be 450°C for 1 hour. Transmission electron microscopy (TEM) showed dispersion of fine precipitates (˜10nm) inside the martensitic phase. Energy dispersive X-ray spectroscopy (EDS) showed the chemical composition of precipitates to be Zr-30wt%Mo-25wt%Nb-2wt%Fe. Electron crystallography using whole pattern symmetry of the convergent beam electron diffraction (CBED) patterns together with selected area diffraction (SAD) polycrystalline ring patterns, suggests the -6m2 point group for the precipitates belonging to hexagonal crystal structure, with a= 2.936 A and c=4.481 A, i.e. c/a =1.526. Crystallographic texture and high temperature tensile properties as well as creep-rupture properties of different microstructures are discussed in Chapter 5. Texture analysis showed that solution treatment high in the alpha Zr+betaZr or betaZr regions followed by water quenching or air cooling results in a more random texture compared

  13. Development of modified route for fabrication of Zr-2.5Nb alloy pressure tubes

    International Nuclear Information System (INIS)

    Saibaba, N.; Hemantha Rao, G.V.S.; Phani Babu, C.; Jha, S.K.; Ganesha, G.N.; Ramana Rao, S.V.; Kumar Vaibhaw; Dey, G.K.; Srivastava, D.; Neogy, S.; Mani Krishna, K.V.

    2013-01-01

    Different fabrication trials involving the variation in three important manufacturing stages of Zr-2.5%Nb pressure tube were partially undetaken. The variations were with respect of mode of breaking the cast structure of the ingot (forging vs extrution), ratio of hot extrusion and number of stages of subsequent cold work to produce the finished tube. It was observed that forging process resulted in superior performance in breaking the cast structure. Higher extrusion ratios resulted in more favorable texture and microstrucutre. Continuity of the beta phase in the final microstructure was observed to be more in case of route involving single cold work subsequent to hot extrusion. (author)

  14. Pressure tube type reactors

    International Nuclear Information System (INIS)

    Komada, Masaoki.

    1981-01-01

    Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

  15. Tensile properties of Zr-2.5 Nb pressure tube alloy between 25 and 800 degC

    International Nuclear Information System (INIS)

    Singh, R.N.; Kishore, R.; Sinha, T.K.; Banerjee, S.

    2000-10-01

    Tensile properties of zirconium-2.5 wt. % niobium pressure tube material were evaluated by uniaxial tension tests at temperatures between 25 and 800 degC and under strain-rates varying from 3.3 x 10 -5 to 3.3 x 10 -3 /s. Tests were carried out on specimens fabricated from the sections of finished (autoclaved) tubes as well as on those machined from the sections of cold worked (2 nd pilgered) tubes. Moreover, specimens fabricated from finished tubes belonging to twenty different heats were tested at 300 degC to study the heat to heat variation in tensile properties of this alloy. In order to study the effect of the crystallographic texture on the tensile properties, specimens oriented in longitudinal as well as, in transverse directions of the tubes were also tested. Results showed that both yield and ultimate tensile strengths of this alloy decreased monotonically with increasing test temperatures, with a rapid fall in strengths above a temperature of 350 degC (623 K). The tensile ductility did not change appreciably up to 400 degC (673K) but increased rapidly above this temperature. The observed results on the temperature dependence of the strength and ductility indicated the possible occurrence of dynamic strain-ageing in this alloy in the temperature range of 200-300 degC (473 to 573 K). The transverse specimens showed higher strengths and lower ductility as compared to those of the longitudinal specimens up to a temperature of 350 degC (623 K). Above 350 degC, the difference in the strengths and the ductility of the two types of the specimens, became negligibly small indicating that the texture did not appreciably influence the tensile properties of this alloy at temperatures exceeding 350 degC. The alloy developed extensive superplasticity (ductility exceeding 100 %), when tested in the temperature range of 650-800 degC. Maximum ductility values of 650 % for longitudinal and 900 % for the transverse orientation with strain-rate sensitivity (m) exceeding 0

  16. Influence of hydrogen content on fracture toughness of CWSR Zr-2.5Nb pressure tube alloy

    Science.gov (United States)

    Singh, R. N.; Bind, A. K.; Srinivasan, N. S.; Ståhle, P.

    2013-01-01

    In this work, influence of hydrogen and temperature on the fracture toughness parameters of unirradiated, cold worked and stress relieved (CWSR) Zr-2.5Nb pressure tube alloys used in Indian Pressurized Heavy Water Reactor is reported. The fracture toughness tests were carried out using 17 mm width curved compact tension specimens machined from gaseously hydrogen charged tube-sections. Metallography of the samples revealed that hydrides were predominantly oriented along axial-circumferential plane of the tube. Fracture toughness tests were carried out in the temperature range of 30-300 °C as per ASTM standard E-1820-06, with the crack length measured using direct current potential drop (DCPD) technique. The fracture toughness parameters (JQ, JMax and dJ/da), were determined. The critical crack length (CCL) for catastrophic failure was determined using a numerical method. It was observed that for a given test temperature, the fracture toughness parameters representing crack initiation (JQ) and crack propagation (JMax, and dJ/da) is practically unaffected by hydrogen content. Also, for given hydrogen content, all the aforementioned fracture toughness parameters increased with temperature to a saturation value.

  17. Fast fracture of a zirconium alloy pressure tube: cause and implications

    International Nuclear Information System (INIS)

    Price, E.G.; Cheadle, B.A.

    1985-12-01

    The cause of the unstable fracture of a Zircaloy-2 pressure tube in the core of a CANDU reactor is reviewed. Failure was associated with the presence of brittle zones of zirconium hydride which developed as a result of thermal gradient induced hydrogen diffusion. Unstable fracture occurred when the partial thickness crack reached an unstable length and the crack ran 2 meters along the tube and terminated by circumferential tearing. The partial thickness defect initiated and propagated to an unstable length by delayed hydride cracking is high compared to fatigue progression and increases exponentially with temperature. Delayed hydride cracking can be prevented by reducing residual stresses to a minimum and by high standards of non-destructive testing that ensures freedom from unacceptable defects. Future prevention of fast fracture is based upon the inspection of a limited number of fuel channels for the presence of defects and for conditions which can cause hydride build-up together with the periodic removal of Zr-2.5wt% Nb tubes to monitor their condition

  18. Age hardening of cold-worked Zr-2.5 wt% Nb pressure tube alloy

    International Nuclear Information System (INIS)

    Kishore, R.; Singh, R.N.; Dey, G.K.; Sinha, T.K.

    1992-01-01

    Specimens for hardness and tensile tests, machined from a cold-worked zirconium-2.5% niobium pressure tube, with their axes parallel to longitudinal and transverse directions, were aged for 1 hr. at 300-500 C. The age hardening behaviour was monitored by mechanical tests, electron-microscopy and x-ray diffraction. In addition a few studies were carried on longitudinal tension specimens subjected to prolonged ageing (100-1000 hrs) at 300 C. It was observed that the short-term (1 hour) thermal ageing of this material at 300-400 C caused an increase in both strength and hardness without affecting ductility. It appears that the observed age-hardening is due to precipitation hardening by a niobium-rich phase and softening by recovery of cold-work and that the phenomenon is influenced by crystallographic texture. Further it was noted that a prolonged ageing at 300 C upto 1000 hrs, did not cause any appreciable changes in strength and ductility of the material compared to those obtained by 1 hour ageing at the same temperature. (author). 11 refs., 3 figs., 2 tabs

  19. Pressure tube reactor

    International Nuclear Information System (INIS)

    Susuki, Akira; Murata, Shigeto; Minato, Akihiko.

    1993-01-01

    In a pressure tube reactor, a reactor core is constituted by arranging more than two units of a minimum unit combination of a moderator sealing pipe containing a calandria tube having moderators there between and a calandria tube and moderators. The upper header and a lower header of the calandria tank containing moderators are communicated by way of the moderator sealing tube. Further, a gravitationally dropping mechanism is disposed for injecting neutron absorbing liquid to a calandria gas injection portion. A ratio between a moderator volume and a fuel volume is defined as a function of the inner diameter of the moderator sealing tube, the outer diameter of the calandria tube and the diameter of fuel pellets, and has no influence to intervals of a pressure tube lattice. The interval of the pressure tube lattice is enlarged without increasing the size of the pressure tube, to improve production efficiency of the reactor and set a coolant void coefficient more negative, thereby enabling to improve self controllability and safety. Further, the reactor scram can be conducted by injecting neutron absorbing liquid. (N.H.)

  20. Flow behaviour of autoclaved, 20% cold worked, Zr-2.5Nb alloy pressure tube material in the temperature range of room temperature to 800 deg. C

    International Nuclear Information System (INIS)

    Dureja, A.K.; Sinha, S.K.; Srivastava, Ankit; Sinha, R.K.; Chakravartty, J.K.; Seshu, P.; Pawaskar, D.N.

    2011-01-01

    Pressure tube material of Indian Heavy Water Reactors is 20% cold-worked and stress relieved Zr-2.5Nb alloy. Inherent variability in the process parameters during the fabrication stages of pressure tube and also along the length of component have their effect on micro-structural and texture properties of the material, which in turn affect its strength parameters (yield strength and ultimate tensile strength) and flow characteristics. Data of tensile tests carried out in the temperature range from room temperature to 800 deg. C using the samples taken out from a single pressure tube have been used to develop correlations for characterizing the strength parameters' variation as a function of axial location along length of the tube and the test temperature. Applicability of Ramberg-Osgood, Holloman and Voce's correlations for defining the post yield behaviour of the material has been investigated. Effect of strain rate change on the deformation behaviour has also been studied.

  1. Pressure tube reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Kumasaka, Katsuyuki.

    1988-01-01

    Purpose: To remove the heat of reactor core using a great amount of moderators at the periphery of the reactor core as coolants. Constitution: Heat of a reactor core is removed by disposing a spontaneous recycling cooling device for cooling moderators in a moderator tank, without using additional power driven equipments. That is, a spontaneous recycling cooling device for cooling the moderators in the moderator tank is disposed. Further, the gap between the inner wall of a pressure tube guide pipe disposed through the vertical direction of a moderator tank and the outer wall of a pressure tube inserted through the guide pipe is made smaller than the rupture distortion caused by the thermal expansion upon overheating of the pressure tube and greater than the minimum gap required for heat shiels between the pressure tube and the pressure tube guide pipe during usual operation. In this way, even if such an accident as can not using a coolant cooling device comprising power driven equipment should occur in the pressure tube type reactor, the rise in the temperature of the reactor core can be retarded to obtain a margin with time. (Kamimura, M.)

  2. N Reactor pressure tube 1350 postirradiation examination

    International Nuclear Information System (INIS)

    Cook, D.J.

    1977-01-01

    The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

  3. Quantitative texture determination in pressure tube (Zr-2.5 Wt% Nb alloy) material as a function of cold work

    International Nuclear Information System (INIS)

    Dey, G.K.; Tewari, R.; Srivastava, D.; De, P.K.; Banerjee, S.; Kiran Kumar, M.; Samajdar, I.

    2003-06-01

    The texture studies on the pressure tube Zr-2.5 Nb alloy have mainly been confined to the determination of the basal pole distribution along certain direction or the inverse pole presentation in the material. This information though useful does not provide an insight into micro-textural development upon cold working. In the present study, complete bulk as well as micro texture development as a function of cold work has been obtained by determining orientation distribution function. In this work, two distinct starting microstructures of Zr-2.5 wt% Nb have been used -(a) single-phase α(hcp) martensitic structure and (b) two-phase, β(bcc) + α, Widmanstaetten structure. In the second case, the α phase was present in lamellar morphology and β stringers were sandwiched between these a lamella. In some instances single-phase α were present. However, both microstructures had similar starting crystallographic texture. Samples were deformed by unidirectional and cross rolling at room temperature. In the two-phase structure the changes in the bulk texture on cold rolling was found to be insignificant, while in the single-phase material noticeable textural changes were observed. Taylor type deformation texture models predicted textural changes in single-phase structure but failed to predict the observed lack of textural development in the two-phase material. Microtexture observations showed that a plates remained approximately single crystalline after cold rolling, while the β matrix underwent significant orientational changes. Based on microstructural and microtextural observations, a simple model is proposed in which the plastic flow is mainly confined to the β matrix within which the α plates are subjected to in-plane rigid body rotation. The model explains the observed lack of textural developments in the two-phase structure. (author)

  4. Pressure tube reactors

    International Nuclear Information System (INIS)

    Natori, Hisahide.

    1981-01-01

    Purpose: To improve the electrical power generation efficiency in a pressure tube reactor in which coolants and moderators are separated by feedwater heating with heat generated in heavy water and by decreasing the amount of steams to be extracted from the turbine. Constitution: A heat exchanger and a heavy water cooler are additionally provided to a conventional pressure tube reactor. The heat exchanger is disposed at the pre-stage of a low pressure feedwater heater series. High temperature heavy water heated in the core is passed through the primary side of the exchanger, while feedwater is passed through the secondary side. The cooler is disposed on the downstream of the heat exchanger in the flowing direction of the heavy water, in which heavy water from the heat exchanger is passed through the primary side and the auxiliary equipment cooling water is sent to the secondary side thereof. Accordingly, since extraction of heating steams is no more necessary, the steam can be used for the rotation of the turbine, and the electrical power generation efficiency can be improved. (Seki, T.)

  5. Microstructure and textural characterization of hot extruded Zr-2.5Nb alloy PHWR pressure tube fabricated by various ingot processing route

    International Nuclear Information System (INIS)

    Vaibhaw, Kumar; Jha, S.K.; Saibaba, N.; Neogy, S.; Mani Krishna, K.V.; Srivastava, D.; Dey, G.K.

    2011-01-01

    Zr-2.5 Nb alloys finds its applications as a pressure tube component in pressure tube type thermal reactors such as PHWRs and RBMK due to properties attributed such as low neutron absorption cross section, high temperature strength and corrosion resistance etc. Manufacturing of this life time components involves series of thermo-mechanical processes of hot working and cold working with intermediate annealing. The life time of Pressure tube are limited due to their diametral creep properties which is governed by metallurgical characteristics such as texture, microstructure dislocation density etc. The primary breakdown of cast structure in Vacuum Arc Melted ingot can be effected by either hot extrusion or forging in single or multiple stages before final hot extrusion step into the blank for manufacturing of seamless pressure tube. Elevated temperature deformation carried out in hot working above the recrystallization temperature would enable impositions of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on process parameters such as extrusion ratio, temperature and strain rate. Basic microstructure developed at this deformation stage has significant bearing on the final properties of the material fabricated with subsequent cold working steps. The major texture in α+β Zr-2.5 Nb alloy is established during final extrusion to blank which does not change significantly during subsequent cold pilgering. However, microstructure is modified significantly in subsequent cold working which can be effected by cold pilgering or cold drawing in single or multiple steps. Present paper brings out the various ingot processing routes using forging and or extrusion followed for fabrication of pressure tubes. The development of texture and microstructures has been discussed at the blank stage from these processing routes and also with respect to varying extrusion variable such as extrusion ratio

  6. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1976-01-01

    Object: To prevent excessive heat generation due to radiation of a pressure tube vessel. Structure: A pressure tube encasing therein a core comprises a dual construction comprising inner and outer tubes coaxially disposed. High speed cooling water is passed through the inner tube for cooling. In addition, in the outer periphery of said outer tube there is provided a forced cooling tube disposed coaxially thereto, into which cooling fluid, for example, such as moderator or reflector is forcibly passed. This forced cooling tube has its outer periphery surrounded by the vessel into which moderator or reflector is fed. By the provision of the dual construction of the pressure tube and the forced cooling tube, the vessel may be prevented from heat generation. (Ikeda, J.)

  7. Pressure tube reactor

    International Nuclear Information System (INIS)

    Matsumoto, Tomoyuki; Fujino, Michihira.

    1980-01-01

    Purpose: To equalize heavy water flow distribution by providing a nozzle for externally injecting heavy water from a vibration preventive plate to the upper portion to feed the heavy water in a pressure tube reactor and swallowing up heavy water in a calandria tank to supply the heavy water to the reactor core above the vibration preventive plate. Constitution: A moderator injection nozzle is mounted on the inner wall of a calandria tank. Heavy water is externally injected above the vibration preventive plate, and heavy water in the calandria tank is swallowed up to supply the heavy water to the core reactor above the vibration preventive plate. Therefore, the heavy water flow distribution can be equalized over the entire reactor core, and the distribution of neutron absorber dissolved in the heavy water is equalized. (Yoshihara, H.)

  8. Comparison of evaluation method for planar flaw in pressure tube

    International Nuclear Information System (INIS)

    Choi, Sung Nam; Kim, Hyung Nam; Yoo, Hyun Joo; Hwang, Won Gul

    2009-01-01

    CSA N285.4-94 requires the periodic inservice inspection and surveillance of pressure tubes in operating CANDU nuclear power reactors. If the inspection results reveal a flaw exceeding the acceptance criteria of the Code, the flaw must be evaluated to determine if the pressure is acceptable for continued service. Currently, the flaw evaluation methodology and acceptance criteria specified in CSA N285.8-05, 'Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors'. The Code is applicable to zirconium alloy pressure tubes. The evaluation methodology for a crack-like flaw is similar to that of FFSG(Fitness For Service Guideline for Zirconium alloy pressure in operation CANDU) used now. The object of this paper is to address the fracture initiation and plastic collapse evaluation for the planar flaw as it applies to the pressure tube on Wolsong NPP.

  9. Structural integrity evaluations of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Radu, Vasile

    2003-01-01

    The core of a CANDU-6 pressurized heavy water reactor consists of some hundred horizontal pressure tubes that are manufactured from a Zr-2.5%Nb alloy and which contain the fuel bundles. These tubes are susceptible to a damaging phenomenon known as Delayed Hydride Cracking (DHC). The Zr-2.5%Nb alloy is susceptible to DHC phenomenon when there is diffusion of hydrogen atoms to a service-induced flaws, followed by the hydride platelets formation on the certain crystallographic planes in the matrix material. Finally, the development of hydride regions at the flaw-tip will happened. These hydride regions are able to fracture under stress-temperature conditions (DHC initiation) and the cracks can extend and grow by DHC mechanism. Some studies have been focused on the potential to initiate DHC at the blunt flaws in a CANDU reactor pressure tube and a methodology for structural integrity evaluation was developed. The methodology based on the Failure Assessment Diagrams (FAD's) consists in an integrated graphical plot, where the fracture failure and plastic collapse are simultaneously evaluated by means of two non-dimensional variables (K r and L r ). These two variables represent the ratio of the applied value of either stress or stress intensity factor and the resistance parameter of corresponding magnitude (yield stress or fracture toughness, respectively). Once the plotting plane is determined by the variables K r and L r , the procedure defines a critical failure line that establishes the safe area. The paper will demonstrate the possibility to perform structural integrity evaluations by means of Failure Assessment Diagrams for flaws occurring in CANDU pressure tubes. (author)

  10. Pressure tube reactor

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro; Kaneto, Kunikazu.

    1979-01-01

    Purpose: To attain uniform fluid poison distribution in a calandria tank by downwardly projecting, at an equal distance to the reactor core, a spacer wall from the periphery of an anti-vibration plate in the vicinity of a heavy water flow passage in the periphery of the anti-vibration plate, thereby decrease the amount of heavy water flowing into the heavy water flow passage. Constitution: A projecting wall concentrical with a calandria tank is suspended vertically from the boundary side at the peripheral portion of an anti-vibration plate to a water heavy flow passage in the periphery of the anti-vibration plate. The projecting wall has such a vertical length as about equal to the width of the heavy water flow passage, prevents heavy water flowing through apertures of a control rod guide tube from entering into the heavy water passage and increases the ratio of heavy water that flows through the heavy water flow passage in the anti-vibration plate. Consequently, if the liquid poison density in heavy water is varied, the ununiform poison density in the calandria tank can be prevented. (Seki, T.)

  11. Characterization of tube support alloys

    International Nuclear Information System (INIS)

    Vaia, A.R.

    1985-01-01

    The involvement and relationship of carbon steel corrosion products in the tube denting phenomenon promoted an intensive research effort to: 1) understand, reproduce, and arrest the denting process, and 2) evaluate alternative tube support materials to provide additional corrosion resistance. The paper summarizes a corrosion testing program for the verification of type 405 stainless steel under acid or all volatile treatment conditions

  12. Rejection index for pressure tubes

    International Nuclear Information System (INIS)

    Mitchell, A.B.; Meneley, D.

    1989-10-01

    The objective of the present study was to establish a set of criteria (or Rejection Index) which could be used to decide whether a zirconium-2 1/2 w/o niobium pressure tube in a CANDU reactor should be removed from service due to in-service degradation. A critique of key issues associated with establishing a realistic rejection index was prepared. Areas of uncertainty in available information were identified and recommendations for further analysis and laboratory testing made. A Rejection Index based on the following limits has been recommended: 1) Limits related to design intent and normal operation: any garter spring must remain within the tolerance band specified for its design location; the annulus gas system must normally be operated in a circulating mode with a procedure in place for purging to prevent accumulation of deuterium. It must remain sensitive to leaks into any part of the systems; and pressure tube dimensions and distortions must be limited to maintain the fuel channels within the original design intent; 2) Limits related to defect tolerance: adequate time margins between occurrence of a leaking crack and unstable failure must be demonstrated for all fuel channels; long lap-type flaws are unacceptable; crack-like defects of any size are unacceptable; and score marks, frat marks and other defects with contoured profiles must fall below certain depth, length and stress intensity limits; and 3) Limits related to property degradation: at operating temperature each pressure tube must be demonstrated to have a critical length in excess of a stipulated value; the maximum equivalent hydrogen level in any pressure tube should not exceed a limit which should be defined taking into account the known history of that tube; the maximum equivalent hydrogen level in any rolled joint should not exceed a limit which is presently recommended as 200 ppm equivalent hydrogen; and the maximum diametral creep strain should be limited to less than 5%

  13. Development of zirconium alloy tube manufacturing technology

    International Nuclear Information System (INIS)

    Kim, In Kyu; Park, Chan Hyun; Lee, Seung Hwan; Chung, Sun Kyo

    2009-01-01

    In late 2004, Korea Nuclear Fuel Company (KNF) launched a government funded joint development program with Westinghouse Electric Co. (WEC) to establish zirconium alloy tube manufacturing technology in Korea. Through this program, KNF and WEC have developed a state of the art facility to manufacture high quality nuclear tubes. KNF performed equipment qualification tests for each manufacturing machine with the support of WEC, and independently carried out product qualification tests for each tube product to be commercially produced. Apart from those tests, characterization test program consisting of specification test and characterization test was developed by KNF and WEC to demonstrate to customers of KNF the quality equivalency of products manufactured by KNF and WEC plants respectively. As part of establishment of performance evaluation technology for zirconium alloy tube in Korea, KNF carried out analyses of materials produced for the characterization test program using the most advanced techniques. Thanks to the accomplishment of the development of zirconium alloy tube manufacturing technology, KNF is expected to acquire positive spin off benefits in terms of technology and economy in the near future

  14. Advanced pressure tube sampling tools

    International Nuclear Information System (INIS)

    Wittich, K.C.; King, J.M.

    2002-01-01

    Deuterium concentration is an important parameter that must be assessed to evaluate the Fitness for service of CANDU pressure tubes. In-reactor pressure tube sampling allows accurate deuterium concentration assessment to be made without the expenses associated with fuel channel removal. This technology, which AECL has developed over the past fifteen years, has become the standard method for deuterium concentration assessment. AECL is developing a multi-head tool that would reduce in-reactor handling overhead by allowing one tool to sequentially sample at all four axial pressure tube locations before removal from the reactor. Four sets of independent cutting heads, like those on the existing sampling tools, facilitate this incorporating proven technology demonstrated in over 1400 in-reactor samples taken to date. The multi-head tool is delivered by AECL's Advanced Delivery Machine or other similar delivery machines. Further, AECL has developed an automated sample handling system that receives and processes the tool once out of the reactor. This system retrieves samples from the tool, dries, weighs and places them in labelled vials which are then directed into shielded shipping flasks. The multi-head wet sampling tool and the automated sample handling system are based on proven technology and offer continued savings and dose reduction to utilities in a competitive electricity market. (author)

  15. Ballooning of CANDU pressure tube in local thermal transients

    International Nuclear Information System (INIS)

    Mihalache, Maria; Ionescu, Viorel

    2008-01-01

    In certain LOCA scenarios for the CANDU fuel channel, the ballooning of the pressure tube and contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator takes place through the contact area. If the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. In INR-Pitesti the DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after occurrence of the contact between the two tubes. The code contains few models: thermal creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This paper gives a DELOCA code description and the fuel channel behaviour analysis, in transient temperature conditions of the pressure tube, using the materials properties, time and temperature dependencies of these properties as obtained in the different laboratories of the world and in the INR - Pitesti in the last years. DELOCA computer code simulated the fuel channel response to the constant heating rates of inside pressure tube surface. The paper presents contact temperature and time dependencies on the heating rate, and the appropriate fitting functions. (authors)

  16. Operating performance of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Price, E.G.

    1989-04-01

    The performance of Zircaloy-2 and Zr-2.5 Nb pressure tubes in CANDU reactors is reviewed. The accelerated hydriding of Zircaloy-2 in reducing water chemistries can lower the toughness of this material and it is essential that defect-initiating phenomena, such as hydride blister formation from pressure tube to calandria tube contact, be prevented. Zr-2.5 Nb pressure tubes are performing well with low rates of hydrogen pick-up and good retention of material properties

  17. The stress-rupture behavior of tubes made from austenitic stainless steels and Ni-based alloys subjected to internal pressure

    International Nuclear Information System (INIS)

    Schaefer, L.; Kempe, H.

    1983-12-01

    The report outlines the stress-rupture results obtained on tubes tested as possible fuel rod cladding tubes for fast breeder reactors cooled with sodium, steam or gas. For the rupture elongations of some specimens showing a pronounced burst, higher values than in earlier reports are now indicated because of better evaluation techniques. The choice and comparisons of materials are explained, the calculations of stresses and strains are described, and reference is made to the own studies carried out to date of the parameters influencing creep-rupture behaviour. Minor modifications of the composition of an alloy and of the mechanical-thermal treatment of materials, respectively, are seen to produce clearcut changes in the stress-rupture properties. (orig.) [de

  18. Study of microstructure, texture and mechanical properties of Zr–2.5Nb alloy pressure tubes fabricated with different processing routes

    International Nuclear Information System (INIS)

    Saibaba, N.; Vaibhaw, Kumar; Neogy, S.; Mani Krishna, K.V.; Jha, S.K.; Phani Babu, C.; Ramana Rao, S.V.; Srivastava, D.; Dey, G.K.

    2013-01-01

    Different fabrication trials involving the variation in three important stages of Zr–2.5Nb pressure tube were undertaken. The variations were with respect to the mode of breaking the cast structure of the ingot (forging vs extrusion), the hot extrusion ratio and the number of subsequent cold work stages to produce the finished tube. It was observed that the forging process resulted in superior performance in breaking the cast structure. Higher extrusion ratios resulted in more favorable texture and microstructure. More continuity of the beta phase was observed in the final microstructure for the route involving the single cold work step subsequent to hot extrusion

  19. Method of repairing pressure tube type reactors

    International Nuclear Information System (INIS)

    Asada, Takashi.

    1983-01-01

    Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

  20. The resistance to PWSCC of explosively expanded Alloy 600 tube-to-tubesheet joints

    International Nuclear Information System (INIS)

    Gold, R.E.; Pement, F.W.; Tarabek, S.A.; Economy, G.

    1992-01-01

    Experimental evaluations were performed to determine the approximate magnitude of the residual stresses associated with explosively expanded steam generator tubing, and to assess the resistance to primary water stress corrosion cracking (PWSCC) of these expansions. Indexing of residual stresses was performed by means of magnesium chloride exposures of surrogate stainless steel mockups. The PWSCC resistance was evaluated by the testing of pressurized mockups of explosively expanded mill annealed Alloy 600 tubing in a highly accelerated Alloy 600 tubing in a highly accelerated steam test environment. Shot peening of the inside tube surfaces was demonstrated to be effective in modifying the residual stresses, providing additional resistance to PWSCC

  1. Dynamics of explosively imploded pressurized tubes

    Science.gov (United States)

    Szirti, Daniel; Loiseau, Jason; Higgins, Andrew; Tanguay, Vincent

    2011-04-01

    The detonation of an explosive layer surrounding a pressurized thin-walled tube causes the formation of a virtual piston that drives a precursor shock wave ahead of the detonation, generating very high temperatures and pressures in the gas contained within the tube. Such a device can be used as the driver for a high energy density shock tube or hypervelocity gas gun. The dynamics of the precursor shock wave were investigated for different tube sizes and initial fill pressures. Shock velocity and standoff distance were found to decrease with increasing fill pressure, mainly due to radial expansion of the tube. Adding a tamper can reduce this effect, but may increase jetting. A simple analytical model based on acoustic wave interactions was developed to calculate pump tube expansion and the resulting effect on the shock velocity and standoff distance. Results from this model agree quite well with experimental data.

  2. Assessment of the Polyacrylic Acid for an Ammonia Water Treatment and for Alloy 800NG SG Tube Material in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Lamouroux, Christine; You, Dominique; Plancque, Gabriel; Roy, Marc; Laire, Charles; Schnongs, Philippe

    2012-09-01

    To prevent the Steam Generators (SG) fouling by corrosion products or the Tube Support Plate (TSP) blockage the on-line injection of a dispersant such the Polyacrylic Acid (PAA) could be a relevant water treatment. Long-term trials performed in PWRs have shown that the PAA, injected at the SG inlet, facilitate the evacuation of the iron oxides by the SG blowdown. Given the ammonia treatment of the secondary water of the Belgian PWRs, the R and D program carried out was devoted to: - Verify the innocuousness of the PAA and its degradation products versus Alloy 800NG SCC susceptibility in case of over concentrations and sludge presence, - Assess the potential impact of the PAA and its thermal degradation products on the specific NH 3 water treatment. The main results can be summarized as following: The corrosion tests performed with PAA in case of over concentrations and sludge couldn't point out any negative effect of the dispersant on the SCC susceptibility of tubing materials such as Alloy 800NG. No significant modification of the tube oxide layer has been observed. At the SG operating temperature, the PAA is decomposed and a large spectrum from high to lower molecular weights polymers than the initial PAA arises. The fragmentation of the polymer into low molecular weight polyacrylic acids is obtained within 20 minutes and the average molecular weight is reduced by 50% from the original one. The thermal degradation products, their quantity and their kinetic of appearance, have been determined. The generated acetate concentration during the on-line dispersant application should remain low compared to the current values observed in the SG water. From the numerical simulation based on acetate concentration and on the kinetic law deduced from the experimental work, it can be concluded that in a 2-phase medium, the margin on the water pH compared to the neutral pH remains high. At 180 deg. C, no impact on the water pH is identified, taking into account realistic

  3. Delayed hydrogen cracking test design for pressure tubes

    International Nuclear Information System (INIS)

    Haddad, Roberto; Loberse, Antonio N.; Yawny, Alejandro A.; Riquelme, Pablo

    1999-01-01

    CANDU nuclear power stations pressure tubes of alloy Zr-2,5 % Nb present a cracking phenomenon known as delayed hydrogen cracking (DHC). This is a brittle fracture of zirconium hydrides that are developed by hydrogen due to aqueous corrosion on the metal surface. This hydrogen diffuses to the crack tip where brittle zirconium hydrides develops and promotes the crack propagation. A direct current potential decay (DCPD) technique has been developed to measure crack propagation rates on compact test (CT) samples machined from a non irradiated pressure tube. Those test samples were hydrogen charged by cathodic polarization in an acid solution and then pre cracked in a fatigue machine. This technique proved to be useful to measure crack propagation rates with at least 1% accuracy for DHC in pressure tubes. (author)

  4. N Reactor pressure tube 2566 postirradiation examination

    International Nuclear Information System (INIS)

    Scott, K.V.

    1978-01-01

    Pressure tube 2566 was removed from N Reactor in July, 1977 to initiate the postirradiation examination program required by the Technical Specifications. Destructive examination of the pressure tube, after a maximum accumulated fluence of 4.6 x 10 21 n/cm 2 (E > 1 MeV), was conducted at the Hanford Engineering Development Laboratory to determine the effects of reactor service on the mechanical properties and hydrogen absorption and corrosion characteristics of the pressure tube. Tube 2566 is the sixth tube removed for destructive examination since the initial reactor startup. Evaluation of test results reveal that no significant detrimental changes have occurred in the parameters studied, since the last tube was removed in 1974

  5. Delayed hydride cracking in Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    Mieza, Juan I.; Domizzi, Gladys; Vigna, Gustavo L.

    2007-01-01

    Zr-2.5 Nb alloy from CANDU pressure tubes are prone to failure by hydrogen intake. One of the degradation mechanisms is delayed hydride cracking, which is characterized by the velocity of cracking. In this work, we study the effect of beta zirconium phase transformation over delayed hydride cracking velocity in Zr-2.5 Nb alloy from pressure tubes. Acoustic emission technique was used for cracking detection. (author) [es

  6. Pressure tube rupture in a closed tank

    International Nuclear Information System (INIS)

    Khater, H.A.; Hadaller, G.I.; Stern, F.

    1985-06-01

    A study has been prepared on the feasibility of conducting pressure tube/calandria tube rupture tests in a closed tank, simulating a scaled-down calandria vessel. The study includes: i) a review of previous work, ii) an analytical investigation of the scaling problem of the calandria vessel and relevant in-core structures, iii) selection of a method for initiating pressure tube/calandria tube rupture, iv) a set of specifications for the test assembly, v) general arrangement drawings, vi) a proposal for a test matrix, vii) a survey and evaluation of existing facilities which could provide the required high pressure, temperature and fluid inventory, and viii) a cost estimate for the detailed design and construction, instrumentation, data acquisition and reduction, testing and reporting. The study concludes that it is both technically and practically feasible to conduct pressure tube rupture tests in a closed tank

  7. Low in reactor creep Zr-base alloy tubes

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Holt, R.A.

    1984-01-01

    This invention relates to zirconium alloy tubes especially for use in nuclear power reactors. More particularly it relates to quaternary 3.5 percent Sn, 1 percent Mo, 1 percent Nb, balance Zr alloy tubes which have been extruded, cold worked and heat treated to lower their dislocation density. In one embodiment the alloys are cold worked less than 5 percent and stress relieved to produce a low dislocation density and in another embodiment the alloys are cold worked up to about 50 percent and annealed to produce a very low dislocation density and also small equiaxed β grains

  8. Development of delayed hydride cracking resistant-pressure tube

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Kim, S. S.; Yim, K. S

    2000-10-01

    For the first time, we demonstrate that the pattern of nucleation and growth of a DHC crack is governed by the precipitation of hydrides so that the DHC velocity and K{sub IH} are determined by an angle of the cracking plane and the hydride habit plane 10.7. Since texture controls the distribution of the 10.7 habit plane in Zr-2.5Nb pressure tube, we draw a conclusion that a textural change in Zr-2.5Nb tube from a strong tangential texture to the radial texture shall increase the threshold stress intensity factor, K{sub IH}, and decrease the delayed hydride cracking velocity. This conclusion is also verified by a complimentary experiment showing a linear dependence of DHCV and K{sub IH} with an increase in the basal component in the cracking plane. On the basis of the study on the DHC mechanism and the effect of manufacturing processes on the properties of Zr-2.5Nb tube, we have established a manufacturing procedure to make pressure tubes with improved DHC resistance. The main features of the established manufacturing process consist in the two step-cold pilgering process and the intermediate heat treatment in the {alpha} + {beta} phase for Zr-2.5Nb alloy and in the {alpha} phase for Zr-1Nb-1.2Sn-0.4Fe alloy. The manufacturing of DHC resistant-pressure tubes of Zr-2.5Nb and Zr-1N-1.2Sn-0.4Fe was made in the ChMP zirconium plant in Russia under a joint research with Drs. Nikulina and Markelov in VNIINM (Russia). Zr-2.5Nb pressure tube made with the established manufacturing process has met all the specification requirements put by KAERI. Chracterization tests have been jointly conducted by VNIINM and KAERI. As expected, the Zr-2.5Nb tube made with the established procedure has improved DHC resistance compared to that of CANDU Zr-2.5Nb pressure tube used currently. The measured DHC velocity of the Zr-2.5Nb tube meets the target value (DHCV <5x10{sup -8} m/s) and its other properties also were equivalent to those of the CANDU Zr-2.5Nb tube used currently. The Zr-1Nb-1

  9. Modelling of pressure tube Quench using PDETWO

    International Nuclear Information System (INIS)

    Parlatan, Y.; Lei, Q.M.; Kwee, M.

    2004-01-01

    Transient two-dimensional heat conduction calculations have been carried out to determine the time-dependent temperature distribution in an overheated pressure tube during quenching with water. The purpose of the calculations is to provide input for evaluation of thermal (secondary) stresses in the pressure tube due to quench. The quench phenomenon in pressure tubes could occur in several hypothetical accident scenarios, including incidents involving intermittent buoyancy-induced flow during outages. In these scenarios, there will be two (radial and axial) or three dimensional temperature gradients, resulting in thermal stresses in the pressure tube, as the water front reaches and starts to cool down the hot pressure tube. The transient, two-dimensional heat conduction equation in the pressure tube during quench is solved using a FORTRAN package called PDETWO, available in the open literature for solving time-dependent coupled systems of non-linear partial differential equations over a two-dimensional rectangular region. This routine is based on finite difference solution of coupled, non-linear partial differential equations. Temperature gradient in the circumferential gradient is neglected for conservatism and convenience. The advancing water front is not modelled explicitly, and assumed to be at a uniform temperature and moving at a constant velocity inferred from experimental data. For outer surface and both ends of the pressure tube in the axial direction, a zero-heat flux boundary condition is assumed, while for the inner surface a moving water-quench front is assumed by appropriately varying the fluid temperature and the heat transfer coefficient. The pressure tube is assumed to be at a uniform temperature of 400 o C initially, to represent conditions expected during an intermittent buoyancy-influenced flow scenario. The results confirm the expectations that axial temperature gradients and associated heat fluxes are small in comparison with those in the

  10. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1975-01-01

    Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

  11. Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

    International Nuclear Information System (INIS)

    Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N.

    2000-09-01

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described

  12. Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Y M; Kim, Y S; Gong, U S; Kwon, S C; Kim, S S; Choo, K N

    2000-09-01

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

  13. Relaxation and corrosion resistance of alloy 800 used for steam generator tubes of ship borne boilers

    International Nuclear Information System (INIS)

    Corrieu, J.M.; Cortial, F.; Maillard, J.L.; Vernot-Loier, C.; Lebeau, M.

    1994-01-01

    The INCO ''INCOLOY 800'' trademark groups the Fe-Cr-Ni alloys containing 30 to 35% nickel, 19 to 23% chromium, 0,15 to 0,60% aluminium, 0,15 to 0,60% titanium and less than 0,10% carbon contents, used as construction materials for condenser and heat exchanger tubes. In parallel with water chemistry control and studies aimed at reducing the residual stresses resulting from tube expansion, studies have been conducted to a better understanding of this alloy, its metallurgy and its corrosion behaviour under accurately defined fabrication and heat treatment conditions. The purpose of this paper is to present the results of a behaviour study of INDRET alloy 800 concerning isothermal relaxation and effects of the said relaxation heat treatments on alloy microstructure studied with a transmission electron-chemical method to determine the sensitiveness to intergranular corrosion, and by electrochemistry in pressurized hot water. (authors). 4 figs., 5 tabs., 7 refs

  14. The transformation behaviour of the beta phase in Zr-2.5 wt% Nb pressure tubes

    International Nuclear Information System (INIS)

    Griffiths, M.; Winegar, J.E.

    1994-12-01

    A temperature-time-transformation (TTT) diagram has been developed for the β-phase in Zr-2.5 wt% Nb pressure tubes. The results show that the morphology and/or physical state of the β-phase in pressure tubes has a significant effect on the transformation behaviour compared with a bulk Zr-19 wt%Nb alloy. (author). 14 refs., 1 tab., 15 figs

  15. Delayed hydride cracking and elastic properties of Excel, a candidate CANDU-SCWR pressure tube material

    International Nuclear Information System (INIS)

    Pan, Z.L.

    2010-01-01

    Excel, a Zr alloy which contains 3.5%Sn, 0.8%Nb and 0.8%Mo, shows high strength, good corrosion resistance, excellent creep-resistance and dimension stability and thus is selected as a candidate pressure tube material for CANDU-SCWR. In the present work, the delayed hydride cracking properties (K IH and the DHC growth rates), the hydrogen solubility and elastic modulus were measured in the irradiated and unirradiated Excel pressure tube material. (author)

  16. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  17. Innovation in pressure tube life assessment

    International Nuclear Information System (INIS)

    Guler, B.; Kalenchuk, D.; Celovsky, A.

    2003-01-01

    The hydrogen equivalent concentration and the rate of hydrogen ingress (in particular, deuterium) in pressure tubes are important parameters that must be assessed to determine the fitness-for-service of CANDU reactors. This paper presents the latest refinement in a process referred to as 'Pressure Tube Sampling', which is the only fully qualified and proven method that allows accurate determination of both the hydrogen equivalent concentration and the rate deuterium ingress without performing an expensive fuel channel removal. Pressure Tube Sampling has evolved over the past fifteen years during which over 2,300 samples have been obtained from CANDU reactors around the globe. In-reactor sampling is the standard method for determining the hydrogen equivalent concentrations and deuterium ingress rates in CANDU reactors. Over the past fifteen years, continual improvements in the Pressure Tube Sampling process have resulted in: the capability to obtain circumferential and axial samples, reduced 'on-face' time, reduced cost, reduced dose to workers, and improved analysis accuracy. Most recently, the new Multi-Head Sampling Tool (MHST) has been developed that continues this trend by using one tool to sample at all four axial pressure tube locations in a single visit to the fuel channel, thereby further improving efficiency. In 2001 October, the MHST was successfully deployed at Wolsong 1 by AECL for Korea Hydro and Nuclear Power. The tool was delivered using their Advanced Delivery Machine (ADM) and a total of sixteen samples were obtained from four channels. A significant saving in time was achieved with a rate of one channel (four samples) being sampled every 2 1/2 hours. For a typical 10-channel campaign, this could equate to a 2 to 3 days time/saving, which is significant in terms of outage schedule, cost, and worker dose. This paper provides a description of some of the latest innovations, with specific details on site application, performance, and end results

  18. Characteristics of Pilger Die Materials for Nuclear Zirconium Alloy Tubes

    International Nuclear Information System (INIS)

    Park, Ki Bum; Kim, In Kyu; Park, Min Young; Kahng, Jong Yeol; Kim, Sun Doo

    2011-01-01

    KEPCO Nuclear Fuel Company's (KEPCO NF) tube manufacturing facility, Techno Special Alloy (TSA) Plant, has started cold pilgering operation since 2008. It is obvious that the cold pilgering process is one of the key processes controlling the quality and the characteristics of the tubes manufactured, i.e. nuclear zirconium alloy tube in KEPCO NF. Cold pilgering is a rolling process for forming metal tubes in which diameter and wall thickness are reduced in a number of forming steps, using ring dies at outside of the tube and a curved mandrel at inside to reduce tube cross sections by up to 90 percent. The OD size of tube is reduced by a pair of dies, and ID size and wall thickness is controlled simultaneously by mandrel. During the cold pilgering process, both tools are the critical components for providing qualified tube. Development of pilger die and mandrel has been a significant importance in the zirconium tube manufacturing and a major goal of KEPCO NF. The objective of this study is to evaluate the life time of pilger die during pilgering. Therefore, a comparison of the heat treatment and mechanical properties of between AISI 52100 and AISI H13 materials was made in this study

  19. Creep-Rupture Behavior of Ni-Based Alloy Tube Bends for A-USC Boilers

    Science.gov (United States)

    Shingledecker, John

    Advanced ultrasupercritical (A-USC) boiler designs will require the use of nickel-based alloys for superheaters and reheaters and thus tube bending will be required. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section II PG-19 limits the amount of cold-strain for boiler tube bends for austenitic materials. In this summary and analysis of research conducted to date, a number of candidate nickel-based A-USC alloys were evaluated. These alloys include alloy 230, alloy 617, and Inconel 740/740H. Uniaxial creep and novel structural tests and corresponding post-test analysis, which included physical measurements, simplified analytical analysis, and detailed microscopy, showed that different damage mechanisms may operate based on test conditions, alloy, and cold-strain levels. Overall, creep strength and ductility were reduced in all the alloys, but the degree of degradation varied substantially. The results support the current cold-strain limits now incorporated in ASME for these alloys for long-term A-USC boiler service.

  20. Gamma sensitivity of pressurized drift tubes

    International Nuclear Information System (INIS)

    Baranov, S.A.; Bojko, I.R.; Shelkov, G.A.; Ignatenko, M.A.

    1995-01-01

    Using a set of commonly used radioactive sources, the efficiency of pressurized drift tubes for gammas with energy from 5.9 keV up to 1.3 MeV has been measured. The tube was made of aluminium and filled with Ar, 15%CO 2 and 2.5%iC 4 H 10 gas mixture at 3 atm. The measured efficiency is compared with the results of the calculations in the frame of our simple model as well as with that of the Monte Carlo simulation using GEANT code. The results of our calculations are in agreement with experimental data, while GEANT simulation tends to give lower efficiency in the energy range of 200 keV γ <1300 keV. The average efficiency of the tube in the field of ATLAS gamma background is about 0.45%. 8 refs., 7 figs., 1 tab

  1. Calandria cooling structure in pressure tube reactor

    International Nuclear Information System (INIS)

    Hyugaji, Takenori; Sasada, Yasuhiro.

    1976-01-01

    Purpose: To contrive the structure of a heavy water distributing device in a pressure tube reactor thereby to reduce the variation in the cooling function thereof due to the welding deformation and installation error. Constitution: A heating water distributing plate is provided at the lower part of the upper tubular plate of a calandria tank to form a heavy water distributing chamber between both plates and a plurality of calandria tubes. Heavy water which has flowed in the upper part of the heavy water distributing plate from the heavy water inlet nozzle flows down through gaps formed around the calandria tubes, whereby the cooling of the calandria tank and the calandria tubes is carried out. In the above described calandria cooling structure, a heavy water distributing plate support is provided to secure the heavy water distributing plate and torus-shaped heavy water distributing rings are fixed to holes formed in the heavy water distributing plate penetrating through the calandria tubes thereby to form torus-shaped heavy water outlet ports each having a space. (Seki, T.)

  2. Microstructural examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

    Srivastava, D.; Tewari, R.; Dey, G.K.; Sah, D.N.; Banerjee, S.

    2005-01-01

    Irradiation induced microstructural changes in Zr alloys strongly influence the creep, growth and mechanical properties of pressure tube material. Since dimensional changes and mechanical property degradation can limit the life of pressure tube, it is essential to study and develop an understanding of the microstructure produced by neutron irradiation, by examining samples taken from the irradiated components. In the present work, an effort has been made to examine, microstructure of the Zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water Reactor (PHWR). The present work is a first step towards a comprehensive program of characterization of microstructure of reactor materials after irradiation to different fluence levels in power reactors. In this study, samples from a Zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit 1, for a period for 6.77 effective full power years (EFPYs), have been prepared and examined. The samples selected from the tube are expected to have a cumulative radiation damage of about 3 dpa. Samples prepared from the off cuts of RAPS-1 pressure tubes were also studied for examining the unirradiated microstructure of the material. The samples were examined in a 200kV JEOL 2000 FX microscope. This paper presents the distinct features observed in irradiated sample and a comprehensive comparison of the microstructures of the unirradiated and irradiated material. The effect of annealing on the annihilation of the defects generated during irradiation has been also studied. The bright field micrographs revealed that microstructure of the irradiated samples was different in many respects from the microstructure of the unirradiated samples. The presence of defect structure in the form of loops etc could be seen in the irradiated sample. These loops were mostly c-type loops lying in the basal plane. The dissolution and redistribution of the precipitates were

  3. Deformation of in-service pressure tubes

    International Nuclear Information System (INIS)

    Sarce, A.L.

    1993-01-01

    Candu type nuclear reactor pressure tubes suffer deformations during operation. This are consequences of irradiation growth and creep. By means of a computer code which takes into account the material microstructure, the above mentioned deformations are calculated, and results are compared with corresponding values measured at Embalse nuclear power plant. The calculations make explicit inclusion of intergranular stresses caused by an isotropy in the material. (author). 1 ref

  4. Delayed hydride cracking in zirconium alloys in pressure tube nuclear reactors. Final report of a coordinated research project 1998-2002

    International Nuclear Information System (INIS)

    2004-10-01

    This report describes all of the research work undertaken as part of the IAEA coordinated research project on hydrogen and hydride induced degradation of the mechanical and physical properties of zirconium based alloys, and includes a review of the state of the art in understanding crack propagation by Delayed Hydride Cracking (DHC), and details of the experimental procedures that have produced the most consistent set of DHC rates reported in an international round-robin exercise to this date. It was concluded that 1) the techniques for performing measurements of the rate of delayed hydride cracking in zirconium alloys have been transferred from the host laboratory to other countries; 2) by following a strict procedure, a very consistent set of values of crack velocity were obtained by both individual laboratories and between the different laboratories; 3) the results over a wide range of test temperatures from materials with various microstructures fitted into the current theoretical framework for delayed hydride cracking; 4) an inter-laboratory comparison of hydrogen analysis revealed the importance of calibration and led to improvements in measurement in the participating laboratories and 5) the success of the CRP in achieving its goals has led to the initiation of some national programmes

  5. Contrastive Analysis and Research on Negative Pressure Beam Tube System and Positive Pressure Beam Tube System for Mine Use

    Science.gov (United States)

    Wang, Xinyi; Shen, Jialong; Liu, Xinbo

    2018-01-01

    Against the technical defects of universally applicable beam tube monitoring system at present, such as air suction in the beam tube, line clogging, long sampling time, etc., the paper analyzes the current situation of the spontaneous combustion fire disaster forecast of mine in our country and these defects one by one. On this basis, the paper proposes a research thought that improving the positive pressure beam tube so as to substitute the negative pressure beam tube. Then, the paper introduces the beam tube monitoring system based on positive pressure technology through theoretical analysis and experiment. In the comparison with negative pressure beam tube, the paper concludes the advantage of the new system and draws the conclusion that the positive pressure beam tube is superior to the negative pressure beam tube system both in test result and test time. At last, the paper proposes prospect of the beam tube monitoring system based on positive pressure technology.

  6. 21 CFR 868.5860 - Pressure tubing and accessories.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Pressure tubing and accessories. 868.5860 Section... (CONTINUED) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Therapeutic Devices § 868.5860 Pressure tubing and accessories. (a) Identification. Pressure tubing and accessories are flexible or rigid devices intended to...

  7. Forming of Zr-4 alloy guide tube with varied diameters

    International Nuclear Information System (INIS)

    Wei Songyan; Tian Zhenye

    1989-10-01

    A new built-up mould method to manufacture Zr-4 alloy guide tubes with varied diameters at the middle of tube is introduced. The guide tube is used in nuclear power plants for guiding the control rods. This method has many advantages such as simple in forming, low cost of manufacturing, no need of special devices and favour of batch processing. The test results show that the accuracy of size, mechanical properties, resistance to corrosion, grain size and hydrogenate orientation of the end-products can meet the technical needs for nuclear reactor operation

  8. A statistical approach to the prediction of pressure tube fracture toughness

    International Nuclear Information System (INIS)

    Pandey, M.D.; Radford, D.D.

    2008-01-01

    The fracture toughness of the zirconium alloy (Zr-2.5Nb) is an important parameter in determining the flaw tolerance for operation of pressure tubes in a nuclear reactor. Fracture toughness data have been generated by performing rising pressure burst tests on sections of pressure tubes removed from operating reactors. The test data were used to generate a lower-bound fracture toughness curve, which is used in defining the operational limits of pressure tubes. The paper presents a comprehensive statistical analysis of burst test data and develops a multivariate statistical model to relate toughness with material chemistry, mechanical properties, and operational history. The proposed model can be useful in predicting fracture toughness of specific in-service pressure tubes, thereby minimizing conservatism associated with a generic lower-bound approach

  9. Moderator mixing after a pressure tube failure

    International Nuclear Information System (INIS)

    MacKinnon, J.C.; Fortman, R.A.; Hadaller, G.I.

    1997-01-01

    During a guaranteed shutdown state (GSS) in a CANDU reactor, there must be sufficient negative reactivity to ensure subcriticality in the event of a process failure. In one of the acceptable states, the reactor is kept subcritical by a high concentration of a neutron-absorbing chemical (the poison gadolinium nitrate) dissolved in the moderator (i.e., the moderator is guaranteed overpoisoned). A postulated accident scenario which is considered as a part of reactor safety analysis is the rupture of a fuel channel (i.e., a pressure tube/calandria tube break) when the reactor is in a GSS. If one of the channels in the core breaks (requiring a simultaneous failure of both the pressure tube and the surrounding calandria tube), coolant from the primary heat transport system will be discharged into the moderator, causing an associated displacement of fluid through relief ducts at the top of the calandria vessel. The incoming (unpoisoned) coolant may mix quickly with the moderator, or may mix slowly while displacing poisoned moderator through the relief ducts. The effectiveness of mixing generally depends on the break location, the coolant discharge rate and the moderator circulation. If an in-core loss of coolant accident occurred while the reactor is in this overpoisoned state, it must be guaranteed that even with the dilution of the poison by the incoming coolant the reactor will remain subcritical on both a local and global basis. This paper presents an overview of an experimental program in progress at the Moderator Test Facility at Stern Laboratories to investigate coolant/poison mixing for a simulated in-core fishmouth pressure tube/calandria tube rupture. The nominal system conditions investigated are of a reactor in a GSS, with coolant in the primary heat transport system at the same temperature as the heavily poisoned moderator, i.e., a depressurised 'cold' state. The results presented are those obtained during the commissioning of the modified Test Facility. The

  10. Observations and insights into Pb-assisted stress corrosion cracking of alloy 600 steam generator tubes

    International Nuclear Information System (INIS)

    Thomas, L.; Bruemmer, Stephen M.

    2005-01-01

    Pb-assisted stress-corrosion cracking (PbSCC) of Alloy 600 steam-generator tubing in high-temperature-water service and laboratory tests were studied by analytical transmission electron microscopy of cross-sectioned samples. Examinations of pulled tubes from many pressurized water reactors revealed lead in cracks from 11 of 17 samples. Comparisons of the degraded intergranular structures with ones produced in simple laboratory tests with PbO in near-neutral AVT water showed that the PbSCC characteristics in service tubing could be reproduced without complex chemistries and heat-flow conditions that can occur during plant operation. Observations of intergranular and transgranular cracks promoted by Pb in the test samples also provided new insights into the mechanisms of PbSCC in mill-annealed and thermally treated Alloy 600

  11. Selected reading on introduction to pressure tube technology

    International Nuclear Information System (INIS)

    Causey, A.R.; Coleman, C.E.; Ells, C.E.

    1981-10-01

    Four lectures on pressure tube technology were presented at Sheridan Park, Ontario, on 1981 June 1. The titles were 'Pressure Tubes and Their Operational Environment', 'Fabrication, Inspection and Properties of Current Production Pressure Tubes', 'In-Reactor Deformation of Fuel Channels', and 'Potential Failure Modes in Pressure Tubes'. This report lists the references used in preparing the lectures. It is intended to provide a starting point in reading for people who need to become familiar with pressure tube technology but have little prior knowledge of the topic

  12. Tube in zirconium base alloy for nuclear fuel assembly and manufacturing process of such a tube

    International Nuclear Information System (INIS)

    Mardon, J.P.; Senevat, J.; Charquet, D.

    1996-01-01

    This patent concerns the description and manufacturing guidelines of a zirconium alloy tube for fuel cladding or fuel assembly guiding. The alloy contains (in weight) 0.4 to 0.6% of tin, 0.5 to 0.8% of iron, 0.35 to 0.50% of vanadium and 0.1 to 0.18% of oxygen. The carbon and silicon tenors range from 100 to 180 ppm and from 80 to 120 ppm, respectively. The alloy contains only zirconium, plus inevitable impurities, and is completely recrystallized. Corrosion resistance tests were performed on tubes made of this alloy and compared to corrosion tests performed on zircaloy 4 tubes. These tests show a better corrosion resistance and a lower corrosion kinetics for the new alloy, even in presence of lithium and iodine, and a lower hydridation rate. The mechanical resistance of this alloy is slightly lower than the one of zircaloy 4 but becomes equivalent or slightly better after two irradiation cycles. The ductility remains always equal or better than for zircaloy 4. (J.S.)

  13. Nuclear power - replacement of pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

  14. Eccentric pressurized tube for measuring creep rupture

    International Nuclear Information System (INIS)

    Schwab, P.R.

    1981-01-01

    Creep rupture is a long term failure mode in structural materials that occurs at high temperatures and moderate stress levels. The deterioration of the material preceding rupture, termed creep damage, manifests itself in the formation of small cavities on grain boundaries. To measure creep damage, sometimes uniaxial tests are performed, sometimes density measurements are made, and sometimes the grain boundary cavities are measured by microscopy techniques. The purpose of the present research is to explore a new method of measuring creep rupture, which involves measuring the curvature of eccentric pressurized tubes. Theoretical investigations as well as the design, construction, and operation of an experimental apparatus are included in this research

  15. Fracture toughness of irradiated Zr-2.5Nb pressure tube from Indian PHWR

    Science.gov (United States)

    Shah, Priti Kotak; Dubey, J. S.; Shriwastaw, R. S.; Dhotre, M. P.; Bhandekar, A.; Pandit, K. M.; Anantharaman, S.; Singh, R. N.; Chakravartty, J. K.

    2015-03-01

    Fracture toughness of irradiated Zr-2.5Nb alloy pressure tube, fabricated by the cold pilgering and stress relieving route, was evaluated using disk compact tension type specimens. These specimens were punched out from the irradiated pressure tube (S-07), which was in service for about 8 effective full power years of reactor operation in the Kakrapar Atomic Power Station-2 (KAPS-2). The tests were carried out remotely inside a lead shielded enclosure. Crack growth during the test was measured using the direct current potential drop technique. The irradiated pressure tube showed low fracture toughness at 25 °C. The fracture toughness increased with increase in temperature up to 250 °C but was practically unaffected with further increase in temperature up to 300 °C. This paper discusses the fracture behavior of irradiated Indian pressure tube material and compares it with other data available.

  16. Measurement and analysis of pressure tube elongation in the Douglas Point reactor

    International Nuclear Information System (INIS)

    Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

    1980-02-01

    Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

  17. Deposition of magnetite particles onto alloy-800 steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Basset, M.; Arbeau, N.; McInerney, J.; Lister, D.H. [Univ. of New Brunswick, Dept. of Chemical Engineering, Fredericton, NB (Canada)

    1998-07-01

    Fouling is a particularly serious problem in the power generating industry. Deposits modify the thermalhydraulic characteristics of heat transfer surfaces by changing the resistance to heat transfer and the resistance to fluid flow, and, if thick enough, can harbour aggressive chemicals. Deposits are also implicated in the increase of radiation fields around working areas in the primary heat transfer systems of nuclear power plants. In order to understand the preliminary steps of the formation of corrosion product deposits on the outsides of steam generator tubes, a laboratory program has investigated the deposition of magnetite particles from suspension in water onto Alloy-800 surfaces under various conditions of flow, chemistry and boiling heat transfer. A recirculating loop made of stainless steel operating at less than 400kPa pressure, with a nominal coolant temperature of 90 degrees C, was equipped with a vertical glass column which housed a 2.5E-01m-long Alloy-800 boiler tube capable of generating a heat flux of 240kW/m{sup 2} . A concentration of suspended magnetite of 5.0E-03kg/m{sup 3} was maintained in the recirculating coolant, which was maintained at a pH of 7.5. The magnetite was synthesized with a sol-gel process, which was developed to produce reproducibly monodispersed, colloidal (<1{mu}m) and nearly spherical particles. A radiotracing method was used to characterize the deposit evolution with time and to quantify the removal of magnetite particles. The results from a series of deposition experiments are presented here. The deposition process is described in terms of a two-step mechanism: the transport step, involving the transport from the bulk of the liquid to the vicinity of the surface, followed by the attachment step, involving the attachment of the particle onto the surface. Under non-boiling heat transfer conditions, diffusion seems to be the dominant factor ruling deposition with a small contribution from thermophoresis; removal was

  18. Deposition of magnetite particles onto alloy-800 steam generator tubes

    International Nuclear Information System (INIS)

    Basset, M.; Arbeau, N.; McInerney, J.; Lister, D.H.

    1998-01-01

    Fouling is a particularly serious problem in the power generating industry. Deposits modify the thermalhydraulic characteristics of heat transfer surfaces by changing the resistance to heat transfer and the resistance to fluid flow, and, if thick enough, can harbour aggressive chemicals. Deposits are also implicated in the increase of radiation fields around working areas in the primary heat transfer systems of nuclear power plants. In order to understand the preliminary steps of the formation of corrosion product deposits on the outsides of steam generator tubes, a laboratory program has investigated the deposition of magnetite particles from suspension in water onto Alloy-800 surfaces under various conditions of flow, chemistry and boiling heat transfer. A recirculating loop made of stainless steel operating at less than 400kPa pressure, with a nominal coolant temperature of 90 degrees C, was equipped with a vertical glass column which housed a 2.5E-01m-long Alloy-800 boiler tube capable of generating a heat flux of 240kW/m 2 . A concentration of suspended magnetite of 5.0E-03kg/m 3 was maintained in the recirculating coolant, which was maintained at a pH of 7.5. The magnetite was synthesized with a sol-gel process, which was developed to produce reproducibly monodispersed, colloidal (<1μm) and nearly spherical particles. A radiotracing method was used to characterize the deposit evolution with time and to quantify the removal of magnetite particles. The results from a series of deposition experiments are presented here. The deposition process is described in terms of a two-step mechanism: the transport step, involving the transport from the bulk of the liquid to the vicinity of the surface, followed by the attachment step, involving the attachment of the particle onto the surface. Under non-boiling heat transfer conditions, diffusion seems to be the dominant factor ruling deposition with a small contribution from thermophoresis; removal was considered

  19. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Margen, P H; Ahlstroem, P E; Pershagen, B

    1961-04-15

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

  20. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    International Nuclear Information System (INIS)

    Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

    1961-04-01

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

  1. Joining method for pressure tube and martensitic stainless steel tube

    International Nuclear Information System (INIS)

    Kimoto, Hiroshi; Koike, Hiromitsu.

    1993-01-01

    In a joining portion of zirconium alloy and a stainless steel, the surface of martensitic stainless steel being in contact with Zr and Zr alloy is applied with a laser quenching solidification treatment before expanding joining of them to improve the surface. This can provide the surface with refined coagulated cell tissues and make deposits and impurities homogeneous and solubilized. As a result, the surface of the martensitic stainless steel has highly corrosion resistance, to suppress contact corrosion with Zr and Zr alloy. Accordingly, even if it is exposed to high temperature water of 200 to 350degC, failures of Zr and Zr alloy can be suppressed. (T.M.)

  2. In situ sampling for pressure tube deuterium concentration

    International Nuclear Information System (INIS)

    Harrington, A.J.; Kittmer, C.A.

    1988-01-01

    The present method of assessing the useful life of pressure tubes in CANDU (CANada Deuterium Uranium) reactors requires the periodic removal and examination of a tube. Special tooling was developed at Atomic Energy of Canada Limited (AECL) to obtain a sample of material from a pressure tube without removing the tube from the reactor. The sampling tool concept has been successfully used by Ontario Hydro during scheduled outages at the Pickering Nuclear Generating Station (PNGS). (author)

  3. Annular gap measurement between pressure tube and calandria tube by eddy current technique

    International Nuclear Information System (INIS)

    Bhole, V.M.; Rastogi, P.K.; Kulkarni, P.G.

    1992-01-01

    In pressurised heavy water reactor (PHWR) major distinguishing feature is that there are number of identical fuel channels in the reactor core. Each channel consists of pressure tube of Zr-2.5 Nb or zircaloy-2 through which high temperature, high pressure primary coolant is passing. The pressure tube contains fuel. Surrounding the pressure tube there is low pressure, cool heavy water (moderator). The moderator is thermally separated from coolant by the tube which is nominally concentric with pressure tube called calandria tube. There are four garter springs in the annular gap between pressure tube and calandria tube. During the life of the reactor there are number of factors by which the pressure tube sags, most important factors are irradiation creep, thermal creep, fuel load etc. Because of the sag of pressure tube it can touch the calandria tube resulting in formation of cold spot. This leads to hydrogen concentration at that spot by which the material at that place becomes brittle and can lead to catastrophic failure of pressure tube. There is no useful access for measurement of annular gap either through the gas annular space or from exterior of calandria tube. So the annular gap was measured from inside surface of pressure tube which is accessible. Eddy current technique was used for finding the gap. The paper describe the details of split coil design of bobbin probe, selection of operating point on normalised impedance diagram by choosing frequency. Experimental results on full scale mock up, and actual gap measurement in reactor channel, are also given. (author). 7 figs

  4. Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Ploc, R.A.; McRae, G.A

    1999-12-01

    Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable {beta}-Zr that is present as a thin layer encasing the predominant {alpha}-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded {beta}-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the {beta}-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

  5. Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    Ploc, R.A.; McRae, G.A.

    1999-12-01

    Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable β-Zr that is present as a thin layer encasing the predominant α-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded β-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the β-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

  6. Delayed hydride cracking in irradiated Zr-2.5 % Nb pressure tubes

    International Nuclear Information System (INIS)

    Cirimello, Pablo; Coronel, Pascual; Haddad, Roberto; Lafont, Claudio; Mizrahi, Rafael

    2003-01-01

    Pressure tubes in CANDU nuclear power plants are made of Zr-2.5 % Nb alloy, which is susceptible to a cracking process called Delayed Hydride Cracking (DHC). Measurement of DHC velocity on irradiated pressure tubes is essential to assure the validity of the Leak Before Break criterion. This work was performed on samples from two pressure tubes taken out of the Embalse NPP in 1995, belonging to fuel channels A-14 and L-12. DHC velocity in the axial direction was measured at 211 C degrees for samples taken from different axial positions, which allowed to study its dependence on fast neutron fluency and irradiation temperature. Non-irradiated material was also tested. It was found that DHC velocity results for the tested material were similar to those obtained for a great number of tubes irradiated in other CANDU plants. (author)

  7. High-temperature transient creep properties of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; Chow, C.K.

    2002-06-01

    During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)

  8. Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young-Jin [Power Engineering Research Institute, KEPCO Engineering and Construction, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2014-08-15

    Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor.

  9. Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

    International Nuclear Information System (INIS)

    Oh, Young-Jin; Chang, Yoon-Suk

    2014-01-01

    Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor

  10. Stress relief treatment of Alloy 600 steam generator tubing

    International Nuclear Information System (INIS)

    Rooyen, D. van; Cragnolino, C.

    1994-01-01

    The intergranular stress corrosion cracking (IGSCC) of Alloy 600 tubing in the primary side of operating steam generators is the subject of this investigation. The objective of the program was to examine the feasibility of heat treatment to alleviate the IGSCC problem. In addition to this, tests were also performed to examine the IGSCC susceptibility of nuclear grade Alloy 600 tubing obtained from various sources. Examination of temperature-time combinations that may hold potential for improved IGSCC resistance of the transition regions of tubes expanded into tube sheet holes was done. The combinations fall in two categories. One is of short duration and relatively high temperature, where induction is the best method of heating because the treatment only lasts from some tens of seconds to a few minutes. The other is carried out in a lower temperature range and lasts for several hours. This latter combination of temperatures and times is considered for the so-called global heat treatment of entire tube sheet. To assess the effect of these treatments, reverse U-bend testing in high purity deaerated water containing an overpressure of hydrogen was employed and several heats of Alloy 600 were compared in tests at 365 degrees C, which is well above actual operating temperatures of steam generators, but provides an accelerated test procedure. Results of furnace heating in the range of 550-610 degrees C indicated improvement in IGSCC resistance, with best performance after a heat treatment at 610 degrees C for nine hours. In addition to stress relief, carbide precipitation can also occur, and their relative contributions to the improvement is discussed

  11. Degradation of Alloy 800 steam generator tubing and its long-term behaviour predictions for plant life management

    International Nuclear Information System (INIS)

    Lu, Y.C.; Tapping, R.L.; Pandey, M.D.

    2009-01-01

    Alloy 800 tubing has a good service record in steam generators (SGs) in both German pressurized water reactors and CANDU 6 reactors, however, a recent comprehensive examination of several ex-service SG tubes removed from Darlington Nuclear Generating Station (DNGS) found that these SG tubes (which had experienced shallow pitting in service) were more susceptible to pitting corrosion in laboratory tests than a reference nuclear grade Alloy 800 tubing under SG crevice chemistry conditions. This was an unexpected finding and has raised questions about possible effects of in-service 'aging' on SG tubing. In addition, there has also been recent evidence that a few Alloy 800 tubes have experienced stress corrosion cracking (SCC) in some German pressurized water reactors (PWRs), possibly after many years of degradation-free service, although the inspection history of these tubes is not available to confirm that the reported degradation initiated recently. These findings suggest that Alloy 800 tubing may have some aging degradation susceptibility after many years of service. To provide support for a proactive SG aging management, a survey on the corrosion susceptibility of the archived Alloy 800 tubing from CANDU SGs under plausible crevice chemistry conditions was conducted to assess the potential material degradation issues in CANDU SGs. Experimental work was also performed to investigate the root cause leading to Alloy 800 SG tubing degradation. The results from this study suggested that a combination of negative factors; aggressive chemistry resulting from impurity ingress into the secondary side of the SGs, elevated electrochemical corrosion potential (ECP) during SG transients and surface strain/plastic deformation, might have led to the degradation of the ex-service SG tubing. The studies have shown that each of these conditions in isolation does not cause degradation of Alloy 800 SG tubing; a synergistic combination of factors is required. The OPEX and experimental

  12. Influence of tube spinning on formability of friction stir welded aluminum alloy tubes for hydroforming application

    Energy Technology Data Exchange (ETDEWEB)

    Wang, X.S. [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Hu, Z.L., E-mail: zhilihuhit@163.com [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Hubei Key Laboratory of Advanced Technology of Automobile Parts, Wuhan University of Technology, Wuhan 430070 (China); State Key Laboratory of Materials Processing and Die and Mould Technology, Huazhong University of Science and Technology (China); Yuan, S.J. [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Hua, L. [Hubei Key Laboratory of Advanced Technology of Automobile Parts, Wuhan University of Technology, Wuhan 430070 (China)

    2014-06-01

    Due to economic and ecological reasons, the application of tailor-welded blanks of aluminum alloy has gained more and more attention in manufacturing lightweight structures for automotives and aircrafts. In the study, the research was aimed to highlight the influence of spinning on the formability of FSW tubes. The microstructural characteristics of the FSW tubes during spinning were studied by electron backscattered diffraction (EBSD) and transmission electron microscopy (TEM). The formability of the FSW tubes with different spinning reduction was assessed by hydraulic bulge test. It is found that the spinning process shows a grain refinement of the tube. The grains of the FSW tube decrease with increasing thickness reduction, and the effect of grain refinement is more obvious for the BM compared to that of the weld. The difference of grain size and precipitates between the weld and BM leads to an asymmetric W-type microhardness distribution after spinning. The higher thickness reduction of the tube, the more uniform distribution of grains and precipitates it shows, and consequently results in more significant increase of strength. As compared with the result of tensile test, the tube after spinning shows better formability when the stress state changes from uniaxial to biaxial stress state.

  13. Failure maps for internally pressurized Zr-2.5% Nb pressure tubes with circumferential temperature variations

    International Nuclear Information System (INIS)

    Shewfelt, R.S.W.

    1986-01-01

    During some postulated loss-of-coolant accidents, the pressure tube temperature may rise before the internal pressure drops, causing the pressure tube to balloon. The temperature around the pressure tube circumference would likely be nonuniform, producing localized deformation that could possibly cause failure. The computer program, GRAD, was used to determine the circumferential temperature distribution required to cause an internally pressurized Zr-2.5% Nb pressure tube to fail before coming into full contact with its calandria tube. These results were used to construct failure maps. 7 refs

  14. Development and performance of inspection equipment for pressure tubes in Fugen

    International Nuclear Information System (INIS)

    Naruo, Kazuteru; Tanimoto, Ken-ichi; Ohta, Takeo; Nakamura, Takahisa; Imaizumi, Kiyoshi.

    1984-01-01

    The pressure tubes of Fugen are the important equipment as the many tubes compose the core, and since they are made of Zr-2.5% Nb alloy which has been used for the first time in Japan, they have become the object of monitoring (the follow-up investigation of the change of inside diameter, the presence of defects and so on) in addition to the in-service inspection. In this paper, on the inspection equipment for pressure tubes, that has been developed independently by the Power Reactor and Nuclear Fuel Development Corp. in order to carry out the ISI and monitoring, the course of development and the construction and the performance are reported, and the results of having used it for the fourth regular inspection of Fugen are described. The 10-year plan of the ISI and monitoring of pressure tubes is shown. The core of Fugen is composed of 224 pressure tubes, therefore, the inspection is carried out by sampling inspection. The monitoring is carried out on four tubes for the follow-up investigation and one tube that shows the severest operation history at the time of inspection. The equipment performs ultrasonic flaw detection, the measurement of inside diameter and the visual inspection of internal surface. (Kako, I.)

  15. Thermal creep behavior of N36 zirconium alloy cladding tube

    International Nuclear Information System (INIS)

    Wang, P.; Zhao, W.; Dai, X.

    2015-01-01

    N36 is an alloy containing Zr, Sn, Nb and Fe that is developed by China as a superior cladding material to meet the performance of PWR fuel assembly at the maximum fuel rod burn-up. The creep characteristics of N36 zirconium alloy cladding tube were investigated at temperature from 593 K to 723 K with stress ranging from 20 MPa to 160 MPa. Transitions in creep mechanisms were noted, showing the distinct three rate-controlled creep mechanisms for the alloy at test conditions. In the region of low stresses with stress exponent n ∼ 1 and activation energy Q ∼ (104±4) kJ.mol -1 , Coble creep, based on diffusion of materials through grain boundaries, is the dominant rate-controlling mechanism, which contributes to the creep deformation. The formation of slip bands acts as an accommodation mechanism. In the region of middle stress with stress exponent n ∼ 3 and activation energy Q ∼ (195±7) kJ.mol -1 , micro-creep, caused by viscous gliding of dislocations due to the interaction of O atoms with dislocations, controls the deformation. In the high stress region with stress exponent n ∼ 5-6 and activation energy Q ∼ (210±10) kJ.mol -1 , two mechanisms of the climb of edge dislocations (EDC) and the motion of jogged screw dislocation (MJS) contribute to rate controlling process. In test conditions N36 alloy cladding tube behaves a type of creep similar to that noted in class-I (A) alloys

  16. Ion-induced swelling of ODS ferritic alloy MA957 tubing to 500 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B., E-mail: mychailo.toloczko@pnnl.gov [Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Garner, F.A. [Radiation Effects Consulting, Richland, WA 99354 (United States); Voyevodin, V.N.; Bryk, V.V.; Borodin, O.V.; Mel’nychenko, V.V.; Kalchenko, A.S. [Kharkov Institute of Physics and Technology, Kharkov (Ukraine)

    2014-10-15

    In order to study the potential swelling behavior of the ODS ferritic alloy MA957 at very high dpa levels, specimens were prepared from pressurized tubes that were unirradiated archives of tubes previously irradiated in FFTF to doses as high as 110 dpa. These unirradiated specimens were irradiated with 1.8 MeV Cr{sup +} ions to doses ranging from 100 to 500 dpa and examined by transmission electron microscopy. No co-injection of helium or hydrogen was employed. It was shown that compared to several tempered ferritic/martensitic steels irradiated in the same facility, these tubes were rather resistant to void swelling, reaching a maximum value of only 4.5% at 500 dpa and 450 °C. In this fine-grained material, the distribution of swelling was strongly influenced by the presence of void denuded zones along the grain boundaries.

  17. Crack growth of throughwall flaw in Alloy 600 tube during leak testing

    International Nuclear Information System (INIS)

    Bahn, Chi Bum; Majumdar, Saurin

    2015-01-01

    Graphical abstract: - Highlights: • A series of leak testing was conducted at a constant pressure and room temperature. • The time-dependent increase in the leak rate was observed. • The fractography revealed slip offsets and crystallographic facets. • Time-dependent plasticity at the crack tip caused the slip offsets. • Fatigue by jet/structure interaction caused the crystallographic facets. - Abstract: We examined the issue of whether crack growth in a full thickness material can occur in a leaking crack. A series of leak tests was conducted at a room temperature and constant pressure (17.3 MPa) with Alloy 600 tube specimens containing a tight rectangular throughwall axial fatigue crack. To exclude a potential pulsation effect by a high pressure pump, the test water was pressurized by using high pressure nitrogen gas. Fractography showed that crack growth in the full thickness material can occur in the leaking crack by two mechanisms: time-dependent plasticity at the crack tip and fatigue induced by jet/structure interaction. The threshold leak rate at which the jet/structure interaction was triggered was between 1.3 and 3.3 L/min for the specific heat of the Alloy 600 tube tested

  18. The cracking of pressure tubes in the Pickering reactor

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.

    1978-01-01

    Small cracks in 17 of the 390 pressure tubes in Unit 3 of the 2056 MW (electrical) Pickering Generating Station and of 52 tubes in Unit 4, resulted in each of these units being out of service for many months. The cracks originated at areas of extremely high residual tensile stress produced by improper positioning of the rolling tool used during construction to join the pressure tube to its end-fitting. The mechanism of failure was delayed hydrogen cracking. (author)

  19. Development of rolled joints for zirconium-2.5 wt % niobium pressure tubes

    International Nuclear Information System (INIS)

    Madhusoodanan, K.; Sinha, R.K.; Samuel, K.A.; Joeman, V.

    1992-01-01

    Due to its higher strength and lower deuterium pick-up rate, as compared to the existing cold worked zircaloy-2 material, cold worked zirconium-2.5 wt% niobium (Zr-2.5%Nb) alloy is to be used as the pressure tube material in all forthcoming Indian PHWRs starting with KAPP-2. These pressure tubes, which carry the fuel bundles are to be joined to the S.S 403 end-fittings through rolled joints. Since the new pressure tubes have a lower wall thickness and higher room temperature yield stress, than zircaloy-2 tubes the design parameters of the rolled joint had to be developed afresh. Further, since Zr-2.5%Nb is susceptible to delayed hydride cracking, it is necessary to limit the residual stress near the rolled joint to a minimum. Since the high residual stress is due to the initial assembly clearance between the pressure tube and end-fitting, a modified rolled joint had to be developed, referred to as zero clearance rolled joint. This paper provides details of the work carried out at Reactor Engineering Division of Bhabha Atomic Research Centre, Bombay towards the development of the design of the rolled joint as well as the tooling and procedures required for achieving zero-clearance fit-ups at site. The requirements to be met by the Zr-2.5% Nb pressure tubes for achieving acceptable rolled joints are highlighted. (author). 5 refs., 6 figs., 3 tabs

  20. Evaluation of Fatigue Crack Initiation for Volumetric Flaw in Pressure Tube

    International Nuclear Information System (INIS)

    Choi, Sung Nam; Yoo, Hyun Joo

    2005-01-01

    CAN/CSA.N285.4-94 requires the periodic inservice inspection and surveillance of pressure tubes in operating CANDU nuclear power reactors. If the inspection results reveal a flaw exceeding the acceptance criteria of the Code, the flaw must be evaluated to determine if the pressure is acceptable for continued service. Currently, the flaw evaluation methodology and acceptance criteria specified in CSA-N285.05-2005, 'Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors'. The Code is applicable to zirconium alloy pressure tubes. The evaluation methodology for a crack-like flaw is similar to that of ASME B and PV Sec. XI, 'Inservice Inspection of Nuclear Power Plant Components'. However, the evaluation methodology for a blunt volumetric flaw is described in CSA-N285.05-2005 code. The object of this paper is to address the fatigue crack initiation evaluation for the blunt volumetric flaw as it applies to the pressure tube at Wolsong NPP

  1. Patterning of alloy precipitation through external pressure

    Science.gov (United States)

    Franklin, Jack A.

    Due to the nature of their microstructure, alloyed components have the benefit of meeting specific design goals across a wide range of electrical, thermal, and mechanical properties. In general by selecting the correct alloy system and applying a proper heat treatment it is possible to create a metallic sample whose properties achieve a unique set of design requirements. This dissertation presents an innovative processing technique intended to control both the location of formation and the growth rates of precipitates within metallic alloys in order to create multiple patterned areas of unique microstructure within a single sample. Specific experimental results for the Al-Cu alloy system will be shown. The control over precipitation is achieved by altering the conventional heat treatment process with an external surface load applied to selected locations during the quench and anneal. It is shown that the applied pressures affect both the rate and directionality of the atomic diffusion in regions close to the loaded surfaces. The control over growth rates is achieved by altering the enthalpic energy required for successful diffusion between lattice sites. Changes in the local chemical free energy required to direct the diffusion of atoms are established by introducing a non-uniform elastic strain energy field within the samples created by the patterned surface pressures. Either diffusion rates or atomic mobility can be selected as the dominating control process by varying the quench rate; with slower quenches having greater control over the mobility of the alloying elements. Results have shown control of Al2Cu precipitation over 100 microns on mechanically polished surfaces. Further experimental considerations presented will address consistency across sample ensembles. This includes repeatable pressure loading conditions and the chemical interaction between any furnace environments and both the alloy sample and metallic pressure loading devices.

  2. Tensile properties of quadruple melted Zr-2.5Nb pressure tubes evaluated from pressure tube offcuts

    International Nuclear Information System (INIS)

    Shah, Priti Kotak; Dubey, J.S.; Anantharaman, S.

    2013-12-01

    Rajasthan Atomic Power Station-2 (RAPS-2) is the first Pressurised Heavy Water Reactor (PHWR) in India having quadruple melted Zr-2.5Nb pressure tubes. Front-end and back-end off-cuts of sixteen pressure tubes were selected for studying the mechanical properties in axial and transverse directions of the tube. Tension tests were carried out at room temperature and at 300℃ using miniature tensile test specimens. The report presents the experimental details and discusses the base line tensile property data for the quadruple melted pressure tubes of RAPS-2. This data will be useful for the reactor life management. (author)

  3. Hydrogen concentration determination in pressure tube samples using differential scanning calorimetry (dsc)

    International Nuclear Information System (INIS)

    Marinescu, R.; Mincu, M.

    2015-01-01

    Zirconium alloys are widely used as a structural material in nuclear reactors. It is known that zirconium based cladding alloys absorb hydrogen as a result of service in a pressurized water reactor. Hydrogen absorbed (during operation of the reactor) in the zirconium alloy, out of which the pressure tube is made, is one of the major factors determining the life time of the pressure tube. For monitoring the hydrides, samples of the pressure tube are periodically taken and analyzed. At normal reactor operating temperature, hydrogen has limited solubility in the zirconium lattice and precipitates out of solid solution as zirconium hydride when the solid solubility is exceeded. As a consequences material characterization of Zr-2.5Nb CANDU pressure tubes is required after manufacturing but also during the operation to assess its structural integrity and to predict its behavior until the next in-service inspection. Hydrogen and deuterium concentration determination is one of the most important parameters to be evaluated during the experimental tests. Hydrogen present in zirconium alloys has a strong effect of weakening. Following the zirconium-hydrogen reaction, the resulting zirconium hydride precipitates in the mass of material. Weakening of the material, due to the presence of 10 ppm of precipitated hydrogen significantly affects some of its properties. The concentration of hydrogen in a sample can be determined by several methods, one of them being the differential scanning calorimetry (DSC). The principle of the method consists in measuring the difference between the amount of heat required to raise the temperature of a sample and a reference to a certain value. The experiments were made using a TA Instruments DSC Q2000 calorimeter. This paper contains experimental work for hydrogen concentration determination by Differential Scanning Calorimetry (DSC) method. Also, the reproducibility and accuracy of the method used at INR Pitesti are presented. (authors)

  4. Anisotropic deformation of Zr–2.5Nb pressure tube material at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Fong, R.W.L., E-mail: fongr@aecl.ca [Fuel and Fuel Channel Safety Branch, Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories, Chalk River, Ontario (Canada)

    2013-09-15

    Zr–2.5Nb alloy is used for the pressure tubes in CANDU® reactor fuel channels. In reactor, the pressure tube normally operates at 300 °C and experiences a primary coolant fluid internal pressure of approximately 10 MPa. Manufacturing and processing procedures generate an anisotropic state in the pressure tube which makes the tube stronger in the hoop (transverse) direction than in the axial (longitudinal) direction. This anisotropy condition is present for temperatures less than 500 °C. During postulated accident conditions where the material temperature could reach 1000 °C, it might be assumed that the high temperature and subsequent phase change would reduce the inherent anisotropy, and thus affect the deformation behaviour (ballooning) of the pressure tube. From constant-load, rapid-temperature-ramp, uniaxial deformation tests, the deformation rate in the longitudinal direction of the tube behaves differently than the deformation rate in the transverse direction of the tube. This anisotropic mechanical behaviour appears to persist at temperatures up to 1000 °C. This paper presents the results of high-temperature deformation tests using longitudinal and transverse specimens taken from as-received Zr–2.5Nb pressure tubes. It is shown that the anisotropic deformation behaviour observed at high temperatures is largely due to the stable crystallographic texture of the α-Zr phase constituent in the material that was previously observed by neutron diffraction measurements during heating at temperatures up to 1050 °C. The deformation behaviour is also influenced by the phase transformation occurring at high temperatures during heating. The effects of texture and phase transformation on the anisotropic deformation of as-received Zr–2.5Nb pressure tube material are discussed in the context of the tube ballooning behaviour. Because of the high temperatures in postulated accident scenarios, any irradiation damage will be annealed from the pressure tube material

  5. Investigation of impingement attack mechanism of copper alloy condenser tubes

    Energy Technology Data Exchange (ETDEWEB)

    Fukumura, Takuya; Nakajima, Nobuo; Arioka, Koji; Totsuka, Nobuo; Nakagawa, Tomokazu [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    In order to investigate generation and growth mechanisms of impingement attacks of sea water against copper alloy condenser tubes used in condensers of nuclear power plants, we took out condenser tubes from actual condensers, cut them into several pieces and carried out several material tests mainly for impinged spots. In addition water flow inside of a pit was analyzed. From the results of the investigation, it was found that all of impingement attacks were found in the marks left by sessile organisms and none were found in downstream of the marks as frequently proposed so far. At the pits generated inside the marks, iron coating was striped and zinc content was deficient in some cases. Combining these data and the result of flow analysis, we considered the following mechanism of the impingement attacks: sessile organisms clinging to the surface of the condenser tube and growth, occlusion of the tube, extinction and decomposition of sessile organisms, pollution corrosion under the organisms and cavity formation, occlusion removal by the cleaning, generation of impingement attacks by flow collision inside the cavity, growth of the impingement attacks. (author)

  6. Bulging of pressure tubes at hot spots under LOCA conditions

    International Nuclear Information System (INIS)

    Manu, C.; Shewfelt, R.S.W.; Wright, A.C.D.; Aboud, R.; Lau, J.H.K.; Sanderson, D.B.

    1996-01-01

    During certain postulated loss-of-coolant accidents (LOCA) in a CANDU reactor, some fuel channels can become highly voided within a very short time. Although the pressure tubes are heated mainly by convection and thermal radiation during the LOCA transient, additional heat flow occurs through the bearing pads that are in contact with the pressure tribe. This contact can lead to local hot spots and associated thermal stresses in the pressure tube wall. The two factors that affects the behavior of the pressure tubes during LOCA conditions are the internal pressure and the local heating. Although the effect of internal pressure and of axially uniform temperature has been studied elsewhere, the effect of the local heating on the pressure tube behavior has not been modelled before. This paper shows that the bulging of a pressure tube at a hot spot is the result of the thermal stresses that are developed in a pressure tube during a LOCA transient. To isolate the local heating effect from the internal pressure, a series of single-effect experiments was performed. In these experiments, sections of a CANDU pressure tube were subjected to local heating only. The thermal profile and the local deformation were measured function of time. To quantify the effect of the thermal stresses on the bulging of pressure tubes at hot spots and to develop numerical tools that can predict such bulging, finite element analyses were performed rising the ABAQUS finite element computer code. Use of the measured thermal profiles in the ABAQUS finite element analysis, resulted in very good agreement between the predicted and measured displacements. (author)

  7. Pressure loss characteristics of LSTF steam generator heat-transfer tubes. Pressure loss increase due to tube internal instruments

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro

    1994-11-01

    The steam generator of the Large-Scale Test Facility (LSTF) includes 141 heat-transfer U-tubes with different lengths. Six U-tubes among them are furnished with 15 or 17 probe-type instruments (conduction probe with a thermocouple; CPT) protuberant into the primary side of the U-tubes. Other 135 U-tubes are not instrumented. This results in different hydraulic conditions between the instrumented and non-instrumented U-tubes with the same length. A series of pressure loss characteristics tests was conducted at a test apparatus simulating both types of U-tube. The following pressure loss coefficient (K CPT ) was reduced as a function of Reynolds number (Re) from these tests under single-phase water flow conditions. K CPT =0.16 5600≤Re≤52820, K CPT =60.66xRe -0.688 2420≤Re≤5600, K CPT =2.664x10 6 Re -2.06 1371≤Re≤2420. The maximum uncertainty is 22%. By using these results, the total pressure loss coefficients of full length U-tubes were estimated. It is clarified that the total pressure loss of the shortest instrumented U-tube is equivalent to that of the middle-length non-instrumented U-tube and also that a middle-length instrumented U-tube is equivalent to the longest non-instrumented U-tube. Concludingly. it is important to take account of the CPT pressure loss mentioned above in estimation of fluid behavior at the non-instrumented U-tubes either by using the LSTF experiment data from the CPT-installed U-tubes or by using any analytical codes. (author)

  8. Evaluation of techniques for inspection and diagnostics of HWR pressure tubes

    International Nuclear Information System (INIS)

    Choi, Jong-Ho

    2008-01-01

    Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes for Heavy Water Reactors (HWRs), are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. The objective of the CRP was to inter-compare inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of the CRP, participants investigated the capability of different techniques to detect and characterize flaws. During the second phase, participants collaborated to determine the hydrogen concentration and to detect and characterize hydride blisters in zirconium alloy pressure tubes. Eight organizations from six countries, which operate HWRs, have participated in this CRP, Most of the techniques examined are well established and many of them are regularly used during in-service inspection of pressure tubes. The inter-comparison of these techniques provides a platform for identifying a particular technique (or a set of techniques), which is more accurate and reliable as compared to others for a specified task. The CRP also witnessed some new methodologies, which can be implemented on in-service inspection tools. These new techniques could complement the existing ones to overcome their limitations, thereby improving the reliability and accuracy of in-service inspection. This CRP also identified future areas of research and development. (author)

  9. Pressure vessels and methods of sealing leaky tubes disposed in pressure vessels

    International Nuclear Information System (INIS)

    Larson, G.C.

    1980-01-01

    This invention relates to pressure vessels and to methods of sealing leaky tubes in them and is especially applicable to pressure vessels in the form of sheet-and-tube type heat exchangers constructed with a large number of relatively small diameter tubes grouped in a bundle. To seal off a leaky tube in such a heat exchanger an explosive activated plug in the form of a hollow metal body is used, inserted at each end of the tube to be sealed. Using the arrangement of pressure vessel and associated tube sheets and the explosive activated plug method of sealing a leaky tube as described in this invention it is claimed that distortion of the adjacent tubes and the tube sheets is reduced when the explosive activated plugs are detonated. (U.K.)

  10. Sample summary report for ARG 1 pressure tube sample

    International Nuclear Information System (INIS)

    Belinco, C.

    2006-01-01

    The ARG 1 sample is made from an un-irradiated Zr-2.5% Nb pressure tube. The sample has 103.4 mm ID, 112 mm OD and approximately 500 mm length. A punch mark was made very close to one end of the sample. The punch mark indicates the 12 O'clock position and also identifies the face of the tube for making all the measurements. ARG 1 sample contains flaws on ID and OD surface. There was no intentional flaw within the wall of the pressure tube sample. Once the flaws are machined the pressure tube sample was covered from outside to hide the OD flaws. Approximately 50 mm length of pressure tube was left open at both the ends to facilitate the holding of sample in the fixtures for inspection. No flaw was machined in this zone of 50 mm on either end of the pressure tube sample. A total of 20 flaws were machined in ARG 1 sample. Out of these, 16 flaws were on the OD surface and the remaining 4 on the ID surface of the pressure tube. The flaws were characterized in to various groups like axial flaws, circumferential flaws, etc

  11. Study on Hydroforming of Magnesium Alloy Tube under Temperature Condition

    Science.gov (United States)

    Wang, Xinsong; Wang, Shouren; Zhang, Yongliang; Wang, Gaoqi; Guo, Peiquan; Qiao, Yang

    2018-01-01

    First of all, under 100 °C, 150 °C, 200 °C, 250 °C, 300 °C and 350 °C, respectively do the test of magnesium alloy AZ31B temperature tensile and the fracture of SEM electron microscopic scanning, studying the plastic forming ability under six different temperature. Secondly, observe and study the real stress-strain curves and fracture topography. Through observation and research can concluded that with the increase of temperature, the yield strength and tensile strength of AZ31B was increased, and the elongation rate and the plastic deformation capacity are increased obviously. Taking into account the actual production, energy consumption, and mold temperature resistance, 250 °Cwas the best molding temperature. Finally, under the temperature condition of 250 °C, the finite element simulation and simulation of magnesium alloy profiled tube were carried out by Dynaform, and the special wall and forming limit diagram of magnesium alloy were obtained. According to the forming wall thickness and forming limit diagram, the molding experiment can be optimized continuously.

  12. Boussignac continuous positive airway pressure for weaning with tracheostomy tubes

    NARCIS (Netherlands)

    Dieperink, Willem; Aarts, Leon P. H. J.; Rodgers, Michael G. G.; Delwig, Hans; Nijsten, Maarten W. N.

    2008-01-01

    Background: In patients who are weaned with a tracheostomy tube ( TT), continuous positive airway pressure ( CPAP) is frequently used. Dedicated CPAP systems or ventilators with bulky tubing are usually applied. However, CPAP can also be effective without a ventilator by the disposable Bous-signac

  13. Properties and application study of Inconel alloy tube made in China

    International Nuclear Information System (INIS)

    Yang Xiang; Su Xingwan; Wen Yan

    1997-01-01

    The mech-physical properties and the corrosion resistance properties of the SG tube of Inconel alloy made in China under any conditions are briefly presented, and the test and research for bending and expending the tubes have been performed. In the process of corrosion experiments the Inconel alloy tubes were compared with that of the same kind of materials made in foreign countries. The Inconel alloy tubes have better stress corrosion resistance cracking prosperities than Inconel 600 and Incoloy 800 when they were in the solutions which contained high concentrated chlorine ion and alkali at high temperature

  14. Remotised sliver sample removal from irradiated pressure tube

    Energy Technology Data Exchange (ETDEWEB)

    Rupani, B B; Sharma, B S.V.G.; Shyam, T V; Sinha, R K [Bhabha Atomic Research Centre, Bombay (India). Reactor Engineering Div.

    1994-12-31

    The life of irradiated pressure tubes of Pressurised Heavy Water Reactor (PHWR) can be assessed by analysing hydrogen content in sliver samples of pressure tube material. The sample can be obtained in the form of small slivers by scraping at any specified locations within the bore of the pressure tube of operating reactor. A tool namely Hydride Scraping Tool (HST) has been developed to obtain very thin slivers of hydride bearing samples from irradiated pressure tube. In order to save man-rem consumption during scraping operation, a feeding mechanism has also been designed and developed for axial positioning of scraping tool in the channel remotely. This paper covers general details about constructional features of the scraping tool, feeding mechanism and their control system. It also highlights the operational salient features and capabilities of the system. Possible applications of the feeding mechanism in other fields are also indicated. (author). 4 figs.

  15. Remotised sliver sample removal from irradiated pressure tube

    International Nuclear Information System (INIS)

    Rupani, B.B.; Sharma, B.S.V.G.; Shyam, T.V.; Sinha, R.K.

    1994-01-01

    The life of irradiated pressure tubes of Pressurised Heavy Water Reactor (PHWR) can be assessed by analysing hydrogen content in sliver samples of pressure tube material. The sample can be obtained in the form of small slivers by scraping at any specified locations within the bore of the pressure tube of operating reactor. A tool namely Hydride Scraping Tool (HST) has been developed to obtain very thin slivers of hydride bearing samples from irradiated pressure tube. In order to save man-rem consumption during scraping operation, a feeding mechanism has also been designed and developed for axial positioning of scraping tool in the channel remotely. This paper covers general details about constructional features of the scraping tool, feeding mechanism and their control system. It also highlights the operational salient features and capabilities of the system. Possible applications of the feeding mechanism in other fields are also indicated. (author). 4 figs

  16. Axial and transverse tensile properties Zr-2.5Nb pressure tube off-cuts

    International Nuclear Information System (INIS)

    Shah, Priti K.; Dubey, J.S.; Shriwastaw, R.S.; Balakrishnan, K.S.; Anantharaman, S.; Chakravartty, J.K.

    2011-01-01

    Zr-2.5Nb alloys in cold worked and stress relieved (CWSR) condition serves as pressure boundary for hot coolant in Indian Pressurized Heavy Water Reactor (IPHWR). Due to both microstructural and crystallographic anisotropy, the mechanical properties in general and fracture behavior in particular are anisotropic for this material. To understand the anisotropic mechanical behavior of the Zr-2.5Nb pressure tubes of IPHWRs, tension tests were carried out on the RAPS-2 and KAPS-2 pressure tube off-cuts. Off-cuts are the small pieces cut from the two ends of a pressure tube before installing it into the PHWR. Miniature flat tensile specimens (without applying any flattening treatment to the pressure tube) of both longitudinal (or axial) and transverse orientation were fabricated from the off-cuts. Tension tests were carried out both at room temperature and 300 deg C. The transverse specimens showed higher strength and lower elongation than that of the axial ones. The back-end off-cuts showed higher strength compared to front-end off-cuts along axial direction whereas the same is not true for transverse direction specimens. One typical plot obtained in the testing of one of the off-cut is shown. The paper will discuss about the importance of the work carried out, test specimens, test method and results obtained in detail

  17. Highlights of the metallurgical behaviour of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Price, E.G.

    1984-10-01

    This paper is an overview of the service induced metallurgical changes that take place in Zircaloy-2 and Zr-2.5 wt. percent Nb pressure tubes in CANDU reactors. It incorporates the findings of an evaluation program, that followed a significant pressure tube failure at Ontario Hydro's Pickering Nuclear Generating Station, and also provides valid reasons for continued confidence in the current CANDU design

  18. Consequences of pressure tube rupture on in-core components

    International Nuclear Information System (INIS)

    Hill, P.G.; Hauptmann, E.G.; Lee, V.

    1982-12-01

    An investigation has been made of the consequences of pressure tube rupture in calandria vessels of heavy water cooled and moderated reactors. The study included a review of previous experimental and analytical work, as well as supplementary investigations carried out to examine the validity of previous assumptions and findings. The central questions considered were: the possibility of a propagating pressure tube failure; damage to the calandria vessel; and damage to the shut-off-rod guide tubes of the reactor shut-down system. The results of the investigation do not indicate mechanisms of sufficient strength to cause propagating failure in a well-designed, well-operated reactor following a tube burst under normal operating conditions. However, not all the details of the physical processes involved in a tube burst have been revealed by existing experimental and analytical work

  19. Material reliability of Ni alloy electrodeposition for steam generator tube repair

    International Nuclear Information System (INIS)

    Kim, Dong Jin; Kim, Myong Jin; Kim, Joung Soo; Kim, Hong Pyo

    2007-01-01

    Due to the occasional occurrences of Stress Corrosion Cracking (SCC) in steam generator tubing (Alloy 600), degraded tubes are removed from service by plugging or are repaired for re-use. Since electrodeposition inside a tube dose not entail parent tube deformation, residual stress in the tube can be minimized. In this work, tube restoration via electrodeposition inside a steam generator tubing was performed after developing the following: an anode probe to be installed inside a tube, a degreasing condition to remove dirt and grease, an activation condition for surface oxide elimination, a tightly adhered strike layer forming condition between the electroforming layer and the Alloy 600 tube, and the condition for an electroforming layer. The reliability of the electrodeposited material, with a variation of material properties, was evaluated as a function of the electrodeposit position in the vertical direction of a tube using the developed anode. It has been noted that the variation of the material properties along the electrodeposit length was acceptable in a process margin. To improve the reliability of a material property, the causes of the variation occurrence were presumed, and an attempt to minimize the variation has been made. A Ni alloy electrodeposition process is suggested as a Primary Water Stress Corrosion Cracking (PWSCC) mitigation method for various components, including steam generator tubes. The Ni alloy electrodeposit formed inside a tube by using the installed assembly shows proper material properties as well as an excellent SCC resistance

  20. SCC of Alloy 600 in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Pascali, R.; Buzzanca, G.; Quaglia, G.M.; Ronchetti, C.

    1986-01-01

    The studies reported in this paper concern the evaluation of Alloy 600 and 690 behaviour in chemical agressive conditions simulating the concentration film on heat exchanging tube. The corrosion tests have been performed to evidence the influence of metallurgical conditions and different heats. Various devices for reproducing dead areas and steam blanketing have been designed and tested, such as, umbrellas, rings, thin deposits, etc. A system to reproduce the S.G. areas with thick deposits has been designed successively and set up in a previous series of tests, in boiling water at 56 kg/cm/sup 2/, 270 0 C and heat flux 45 W/cm/sup 2/. Caustic SCC tests have been carried out in adiabatic conditions also using small autoclaves

  1. Development of heat treated Zr-2.5 Wt% Nb pressure tube and its microstructural characterization using electron microscopy techniques

    International Nuclear Information System (INIS)

    Saibaba, N.

    2010-01-01

    Two phase Zr-2.5 wt % Nb alloy is widely used for manufacture of pressure tubes for pressurized heavy water reactors (PHWRs). These tubes are used in cold worked and stress relieved (CWSRs) condition and are manufactured by cold drawing or pilgering routes. The microstructure of the CWSR tube is characterized with presence of discontinuous β phase stringers sandwiched between elongated α-phase. Pressure tube undergoes dimensional changes and micro structural deterioration under the reactor operating conditions of temperature, pressure and neutron flux. This limits the life of the component and the availability of the power reactors. There is renewed interest in increasing the life of the pressure tube by bringing about a change in the microstructure of Zr-2.5 Nb material using various thermo mechanical processes during its manufacturing. Heat treatment of this two-phase alloy has been understood to uniquely stabilize the microstructure, which prevents degradation, under in-reactor service condition. This paper illustrates various heat treatment cycles carried out at intermediate cold working stage. Heat treatment involves solutionization of the Zr-2.5 wt % Nb tube from different temperatures followed by two types of quenching process viz, gas quenching and water quenching. The OIM-TEM studies were carried out for characterization of final tube. The technique confirmed the presence of β-phase relatively enriched in Nb content. The resulting SEM microstructures after ageing treatment at different soaking temperatures and time have been presented. Mechanical properties of heat treated pressure tubes, both at room temperature and elevated temperature have been compared with conventional CWSR pressure tube used in PHWRs. (author)

  2. Fatigue crack initiation at complex flaws in hydrided Zr-2.5%Nb samples from CANDU pressure tubes

    International Nuclear Information System (INIS)

    Stoica, L.; Radu, V.

    2016-01-01

    The paper addresses the phenomena which occur at locations where the oxide layer of the inner surface of CANDU tube pressure is damaged by the contact with the fuel element or due to the action of hard particles at the interface between the tube pressure and bearing pad of fuel element. In such situations generate defects, which most often are defects known as ''bearing pad fretting flaws'' or ''debris fretting flaws''. In this paper the experiments are completed in a series of previous works on the mechanical fatigue phenomenon on samples prepared from the pressure tube Zr-2.5% Nb alloy. The phenomenon of variable mechanical stress (or fatigue) may lead to initiation of cracks at the tip of volumetric flaws, according to the accumulation of hydrides, which then fractures and can propagate through the tube wall pressure due to the mechanism of type DHC (Delayed Hydride Cracking). (authors)

  3. Simulation studies on Tube End Expansion of AA2014 Alloy Tubes

    Science.gov (United States)

    Venugopal, L.; Prasad, N. E. C.; Geeta Krishna, P.; Praveen, L.

    2018-03-01

    End forming is defined as forming the end of tubular forms either by inverting the tube or by expanding it. It finds application in many fields such as in automotive and aerospace sectors as power transmission elements, fuel lines, exhaust pipes etc. The main aim of the present work is to expand the AA2014 alloy tubes with different die sets without any fracture. Deform 2D software was used for performing simulations on expanding the tubes with different die set (punch) values having differed forming angles (α = 15°, 30° and 45°) and expansion ratios (rp/r0 = 1.39, 1.53 and 1.67). Experiments were also conducted and the results correlate with the simulation results. The results shows that for the punch having less cone angle (α) values the linear displacement is more rather than higher cone angles. But in the case of higher cone angles the radial displacement is more than the linear displacement.

  4. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C.

    1997-01-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

  5. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  6. Study on manufacturing technology of fuel guide tube using HANA alloys

    International Nuclear Information System (INIS)

    Kim, Hyungil; Jung, Yangil; Park, Dongjun; Park, Jeongyong; Kim, Ilhyun; Choi, Byungkwon; Jeong, Yonghwan; Park, Sangyoon

    2013-04-01

    This research was focused on the study for the manufacturing technology of HANA alloys to crease the corrosion resistance of 30% as well as the to improve the strength of 10% when compared to the commercial zirconium alloys. The new manufacturing concept having higher corrosion resistance and strength than commercial alloy performance can be obtained in this research. This result was transferred to the KNF and, that will be commercialized. This research result can be summarized like this; Ο Parameter study to increase formability of HANA alloy tube - Study on alloy element and heat-treatment effect - Study on texture development mechanism - Study on final annealing effect Ο Out-of-pile performance evaluation of HANA alloy tube - Corrosion performance evaluation of HANA alloy manufactured at KNF - Mechanical performance evaluation of HANA alloy manufactured at KNF - Recrystallization behavior evaluation of HANA alloy manufactured at KNF - Texture characterization of HANA alloy manufactured at KNF - Microstructure characterization of HANA alloy manufactured at KNF Ο Manufacturing guideline setup to increase formability of HANA alloy tube - Manufacturing guideline setup to decrease surface defect - Manufacturing guideline setup to increase strength and corrosion resistance - Manufacturing guideline setup to control texture

  7. Assessment and management of ageing of major nuclear power plant components important to safety: CANDU pressure tubes

    International Nuclear Information System (INIS)

    1998-08-01

    The report documents the current practices for assessment and management of the ageing of the pressure tubes in CANDU reactors and Indian PHWTRs. Chapter headings are: fuel channel and pressure tube description, design basis for the fuel channel and pressure tube, degradation mechanisms and ageing concerns for pressure tubes, inspection and monitoring methods for pressure tubes,assessment methods and fitness-for-service guidelines for pressure tubes, mitigation methods for pressure tubes, and pressure tube ageing management programme

  8. TEM examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

    Srivastava, D.; Tewari, R.; Dey, G.K.; Sharma, B.P.; Sah, D.N.; Banerjee, Suparna; Sahoo, K.C.

    2005-09-01

    In the present work, microstructure of the zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water RAPP-1. Reactor (PHWR) has been examined for the first time using transmission electron microscope (TEM). The samples were obtained from a zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit -1, for a period for 6.77 effective full power years (EFPYs) and expected to have a cumulative radiation damage of about 3 dpa. In this study irradiated microstructure has been characterized and compared it with the microstructure of the unirradiated pressure tube samples. The effect of irradiation on the hydriding behaviour is also studied. (author)

  9. Numerical studies on heat transfer and pressure drop characteristics of flat finned tube bundles with various fin materials

    Science.gov (United States)

    Peng, Y.; Zhang, S. J.; Shen, F.; Wang, X. B.; Yang, X. R.; Yang, L. J.

    2017-11-01

    The air-cooled heat exchanger plays an important role in the field of industry like for example in thermal power plants. On the other hand, it can be used to remove core decay heat out of containment passively in case of a severe accident circumstance. Thus, research on the performance of fins in air-cooled heat exchangers can benefit the optimal design and operation of cooling systems in nuclear power plants. In this study, a CFD (Computational Fluid Dynamic) method is implemented to investigate the effects of inlet velocity, fin spacing and tube pitch on the flow and the heat transfer characteristics of flat fins constructed of various materials (316L stainless steel, copper-nickel alloy and aluminium). A three dimensional geometric model of flat finned tube bundles with fixed longitudinal tube pitch and transverse tube pitch is established. Results for the variation of the average convective heat transfer coefficient with respect to cooling air inlet velocity, fin spacing, tube pitch and fin material are obtained, as well as for the pressure drop of the cooling air passing through finned tube. It is shown that the increase of cooling air inlet velocity results in enhanced average convective heat transfer coefficient and decreasing pressure drop. Both fin spacing and tube pitch engender positive effects on pressure drop and have negative effects on heat transfer characteristics. Concerning the fin material, the heat transfer performance of copper-nickel alloy is superior to 316L stainless steel and inferior to aluminium.

  10. Excessively High Vapor Pressure of Al-based Amorphous Alloys

    Directory of Open Access Journals (Sweden)

    Jae Im Jeong

    2015-10-01

    Full Text Available Aluminum-based amorphous alloys exhibited an abnormally high vapor pressure at their approximate glass transition temperatures. The vapor pressure was confirmed by the formation of Al nanocrystallites from condensation, which was attributed to weight loss of the amorphous alloys. The amount of weight loss varied with the amorphous alloy compositions and was inversely proportional to their glass-forming ability. The vapor pressure of the amorphous alloys around 573 K was close to the vapor pressure of crystalline Al near its melting temperature, 873 K. Our results strongly suggest the possibility of fabricating nanocrystallites or thin films by evaporation at low temperatures.

  11. A survey on the corrosion susceptibility of Alloy 800 CANDU steam generator tubing materials

    International Nuclear Information System (INIS)

    Lu, Y.C.; Dupuis, M.; Burns, D.

    2008-01-01

    To provide support for a proactive steam generator (SG) aging management strategy, a survey on the corrosion susceptibility of the archived Alloy 800 tubing from CANDU SGs under plausible crevice chemistry conditions was conducted to assess the potential material degradation issues in CANDU SGs. Archived Alloy 800 samples were collected from four CANDU utilities. High-temperature electrochemical analysis was carried out to assess the corrosion susceptibility of the archived SG tubing under simulated CANDU crevice chemistry conditions at both 150 o C and 300 o C. The potentiodynamic polarization results obtained from the archived CANDU SG tubes were compared to the data from ex-service tubes removed from Darlington Nuclear Generating Station (DNGS) SGs and a reference nuclear grade Alloy 800 tubing. It was found that the removed Darlington SG tubes, with signs of in-service degradation, were more susceptible to pitting corrosion than the reference nuclear grade Alloy 800 tubing. At 150 o C, under the same neutral crevice chemistry conditions, the potentiodynamic polarization curve of the ex-service Darlington SG tubing has an active peak, which is a sign of propensity to crevice/underdeposit corrosion. This active peak was not observed in any of the potentiodynamic polarization curves of all archived Alloy 800 CANDU SG tubing indicating that archived CANDU SG tubes are less susceptible to the underdeposit corrosion under SG startup conditions. The corrosion behaviour of the archived Alloy 800 tubes from CANDU SG was similar to that of the reference nuclear grade Alloy 800 tubing. The results of this survey suggest that the Alloy 800 tubing materials used in the existing CANDU utilities (other than ex-service DNGS tubing) will continue to have reliable performance under specified CANDU operating conditions. Ex-service SG tubing from DNGS, although showing lower than average corrosion resistance, still has a wide acceptable operating margin and the in

  12. Pressure relief experiments on a cyclindrical carbon brick tube

    International Nuclear Information System (INIS)

    Lang, H.; Weise, H.J.; Ennen, P.

    1978-08-01

    Pressure relief experiments have been carried out on a carbon brick tube. The outer diameter of the specimen was 580 mm, the inner diameter 280 mm, the length 800 mm. The experiments were made with helium at the temperature of the environment. The measurements were carried out in the pressure range from 15 upto 39 bar. The pressure loss was measured dependent on the initial pressure and on time at 5 positions uniformly distributed over the thickness of the tube wall and in the pressure vessel. The maximum pressure transients occurred amounted to approximately 60 bar/second. The maximum overpressure with respect to the environment which occurred in the carbon brick during the relief experiments was about 3.3 bar. The measurements distinctly showed the presence and the effects of inhomogeneities in the sample material. No damages or changes in the carbon brick, which could be regarded as a consequence of the experiments, were found. (orig./GSC) [de

  13. Deadly pressure pneumothorax after withdrawal of misplaced feeding tube

    DEFF Research Database (Denmark)

    Andresen, Erik Nygaard; Frydland, Martin; Usinger, Lotte

    2016-01-01

    BACKGROUND: Many patients have a nasogastric feeding tube inserted during admission; however, misplacement is not uncommon. In this case report we present, to the best of our knowledge, the first documented fatality from pressure pneumothorax following nasogastric tube withdrawal. CASE PRESENTATION......, but our patient died less than an hour after withdrawal. The autopsy report stated that cause of death was tension pneumothorax, which developed following withdrawal of the misplaced feeding tube. CONCLUSIONS: The indications for insertion of nasogastric feeding tubes are many and the procedure...... is considered harmless; however, if the tube is misplaced there is good reason to be cautious on removal as this can unmask puncture of the pleura eliciting pneumothorax and, as this case report shows, result in an ultimately deadly tension pneumothorax....

  14. On random pressure pulses in the turbine draft tube

    Science.gov (United States)

    Kuibin, P. A.; Shtork, S. I.; Skripkin, S. G.; Tsoy, M. A.

    2017-04-01

    The flow in the conical part of the hydroturbine draft tube undergoes various instabilities due to deceleration and flow swirling at off-design operation points. In particular, the precessing vortex rope develops at part-load regimes in the draft tube. This rope induces periodical low-frequency pressure oscillations in the draft tube. Interaction of rotational (asynchronous) mode of disturbances with the elbow can bring to strong oscillations in the whole hydrodynamical system. Recent researches on flow structure in the discharge cone in a regime of free runner had revealed that helical-like vortex rope can be unstable itself. Some coils of helix close to each other and reconnection appears with generation of a vortex ring. The vortex ring moves toward the draft tube wall and downstream. The present research is focused on interaction of vortex ring with wall and generation of pressure pulses.

  15. Influence of the S/N ratio on the corrosion release of Alloy 690 tubes in a primary coolant

    International Nuclear Information System (INIS)

    Shim, Hee-Sang; Choi, Myung Sik; Kim, Kyung Mo; Seo, Myung Ji; Hur, Do Haeng; Choi, Tack-Sang; Yoo, One

    2014-01-01

    Alloy 690TT is a promising steam generator (SG) tube material of a pressurized water reactor due to its excellent resistance to stress corrosion cracking (SCC) that has caused problems in Alloy 600 as an old SG tube material. The qualities of this material have been managed thoroughly from manufacturing step under various specification regulations as well as in in-service step. For examples, the surface roughness are prescribed as the values less than 1.6 μm for the tube outside and 0.5 μm for the inside, respectively. In addition, the surface state and defect must be qualified through the eddy current test (ECT) and the ultrasonic test (UT) according to the ASME Section III, NB2550. Then, the signal-to-noise (S/N) ratio, which is measured using ECT bobbin probe, is the important criteria to determine the material and it shall be 15 to 1 or higher at the standard frequency for any fixed 0.5 m length of any tube. The corrosion behaviours of the Alloy 690TT under high-temperature pressurized primary water have been studied widely in a point of the SCC but discussed narrowly in a point of the corrosion release. In particular, the effect of the S/N ratio on the corrosion release of this material surface has been rarely investigated. In this work, we evaluate the influence of the S/N ratio on the corrosion release of Alloy 690 SG tubes. The specimens with different S/N ratio were selected through ECT bobbin inspection and a corrosion release test was conducted using a simulated primary circulation loop. The material properties and oxidation behaviours were investigated by surface profiler, scanning electron microscopy, transmission electron microscopy, grazing incidence X-ray diffraction and etc. As a result, the corrosion rate was matched preferably with the MRPC characteristics showing macroscopic surface state rather than with the bobbin S/N ratio results. (author)

  16. Holographic NDE of pressure tubes for Cirene nuclear reactor

    International Nuclear Information System (INIS)

    Di Chirico, G.; Pirodda, L.; Villani, A.

    1985-01-01

    Pressure tubes for CIRENE nuclear reactor can be subjected to fretting corrosion of the inner walls. The resulting marks exhibit different geometries, whose influence on the structural behaviour of the tubes has been evaluated by means of a real time holographic technique. The paper shows the results of this investigation. Position and shape of internal defects have been directly visualized by observing holographic fringe distorsions on the outside surface of the tubes. Furthermore, through the fringe patterns, circumferential stress values have also been obtained. (Author) [pt

  17. The temperature dependence of the tensile properties of thermally treated Alloy 690 tubing

    International Nuclear Information System (INIS)

    Harrod, D.L.; Gold, R.E.; Larsson, B.; Bjoerkman, G.

    1992-01-01

    Tensile tests were run in air on full tube cross-sections of 22.23 mm OD by 1.27 mm wall thickness Alloy 690 steam generator production tubes from ten (10) heats of material at eight (8) temperatures between room temperature and 760 degrees C. The tubing was manufactured to specification requirements consistent with the EPRI guidelines for Alloy 690 tubing. The room temperature stress-strain curves are described quite well by the Voce equation. Ductile fracture by dimpled rupture was observed at all test temperatures. The elevated temperature tensile properties are compared with design data given in the ASME Code

  18. Salvaging of service exposed cast alloy 625 cracker tubes of ammonia based Heavy Water Plants

    International Nuclear Information System (INIS)

    Kumar, Niraj; Misra, B.; Mahajan, M.P.; Mittra, J.; Sundararaman, M.; Chakravartty, J.K.

    2006-01-01

    In ammonia based heavy water plants, cracking of ammonia vapour, enriched in deuterium is carried out inside a cracker tube, packed with catalyst. These cracker tubes are made of alloy 625 (either wrought or cast) having dimensions of about 12.5 metres long, 88 mm outer diameter and 7.9 mm wall thickness. Seventy such tubes are housed in a typical ammonia cracker unit. The anticipated design life of such tube is 1,00,000 hrs. when operated at 720 degC based on creep as main degradation mechanism. Presently, these tubes are being operated at 680 degC skin temperature. Alloy 625 tubes are costly and normally not manufactured in India and are being imported. The cast alloy 625 cracker tubes have outlived their design life of 100,000 hrs. Therefore it has been decided to salvage the cast cracker tubes and extend the life further as it had already been done for wrought tubes. Similar to the earlier attempt of resolutionising of wrought alloy 625 tubes, efforts are in progress to salvage these cast tubes. In this study, cast tubes samples were subjected to solution-annealing treatment at two different temperatures, 1100degC and 1160degC respectively for two hrs. Mechanical properties along with the microstructure of the samples, which were resolutionized at 1160degC were comparable with that of virgin material. The 12.5 metres long cast alloy 625 cracker tubes will also be shortly solution-annealed in a specially designed resistance heating furnace after completing some more tests. (author)

  19. Hot hardness studies on zircaloy 2 pressure tube along three orientations

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Ravi, K.; Jarvis, T.; Sengupta, A.K.; Majumdar, S.; Tewari, R.; Shrivastava, D.; Dey, G.K.

    2002-01-01

    Zirconium based alloys are the natural choice for both the fuel element cans and in-core structural components in water cooled nuclear reactors. In this paper, the hot hardness behaviour of zircaloy 2 pressure tubes has been examined from room temperature to 400 degC using a hot hardness tester. For the purpose of comparison, the hardness of the as cast and room temperature rolled specimens has also been carried out. For this, the samples were cut along three orientations and hardness was measured in each of these directions using Vickers diamond pyramid indenter. The variation in hardness of the pressure tube samples show that the hardness was highest along circumferential direction and least along the axial direction. The room temperature rolled samples showed highest hardness along the rolling planes. These variations in hardness could be explained in terms of development of texture during working on the material. (author)

  20. A technique of taking samples from inside the pressure tube along the axis of the fuel channel

    International Nuclear Information System (INIS)

    Gyongyosi, T.

    2005-01-01

    Full text: The ageing process through its complex mechanisms affects in time more or less the component parts, the systems and the structures of the nuclear power plant. For CANDU type nuclear power plant the main component part in operation is the pressure tube, made from Zr - 2,5% Nb alloy, used in extreme hard operation conditions (static and dynamic loading from high pressure and temperature and high neutron flux). The pressure tube endures, in time, changing of the material and of the geometry. Additionally, the excessive hydrogen uptake initiates the Delayed Hydride Cracking (DHC) mechanism on the pressure tubes. For checking the evolution of the hydride / deuteration phenomena in the material of the pressure tube and especially to establish the real lifetime as compared to design lifetime it is useful to initiate, develop and apply a technology and a complex equipment for taking samples directly from inside the pressure tube, this enabling the determination of the hydrogen content. In the paper are showed briefly: - the evolution in time of the techniques for axial taking of the samples from inside of the pressure tube used in the CANDU 6 NPP; - the reasons that determined us to develop one of these technologies; - the technological facilities needed to apply it. (author)

  1. A technique of taking samples from inside the pressure tube along the axis of the fuel channel

    International Nuclear Information System (INIS)

    Gyongyosi, T.

    2005-01-01

    The ageing process through its complex mechanisms affects in time more or less the component parts, the systems and the structures of the nuclear power plant. For CANDU type nuclear power plant the main component part in operation is the pressure tube, made from Zr - 2,5% Nb alloy, used in extreme hard operation conditions (static and dynamic loading from high pressure and temperature and high neutron flux). The pressure tube endures, in time, changing of the material and of the geometry. Additionally, the excessive hydrogen uptake initiates the Delayed Hydride Cracking (DHC) mechanism on the pressure tubes. For checking the evolution of the hydride/deuteration phenomena in the material of the pressure tube and especially to establish the real lifetime as compared to design lifetime it is useful to initiate, develop and apply a technology and a complex equipment for taking samples directly from inside the pressure tube, this enabling the determination of the hydrogen content. In the paper are showed briefly: - the evolution in time of the techniques for axial taking of the samples from inside of the pressure tube used in the CANDU 6 NPP; - the reasons that determined us to develop one of these technologies; - the technological facilities needed to apply it. (author)

  2. Process for forming seamless tubing of zirconium or titanium alloys from welded precursors

    International Nuclear Information System (INIS)

    Sabol, G.P.; Barry, R.F.

    1987-01-01

    A process is described for forming seamless tubing of a material selected from zirconium, zirconium alloys, titanium, and titanium alloys, from welded precursor tubing of the material, having a heterogeneous structure resulting from the welding thereof. The process consists of: heating successive axial segments of the welded tubing, completely through the wall thereof, including the weld, to uniformly transform the heterogeneous, as welded, material into the beta phase; quenching the beta phase tubing segments, the heating and quenching effected sufficiently rapid enough to produce a fine sized beta grain structure completely throughout the precursor tubing, including the weld, and to prevent growth of beta grains within the material larger than 200 micrometers in diameter; and subsequently uniformly deforming the quenched precursor tubing by cold reduction steps to produce a seamless tubing of final size and shape

  3. Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2012-02-15

    Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

  4. Use of CATHENA to model calandria-tube/moderator heat transfer after pressure-tube/calandria-tube ballooning contact

    International Nuclear Information System (INIS)

    Fan, H.Z.; Bilanovic, Z.; Nitheanandan, T.

    2004-01-01

    A study was performed to assess the effect of the calandria-tube/moderator heat transfer after pressure-tube/calandria tube ballooning contact using CATHENA. Results of this study indicated that the analytical tool, CATHENA, can be applied for pool boiling heat transfer on the external surface of a large diameter tube, such as the calandria tube used in CANDU reactors. The methodology in such CANDU-generic study can be used to simulate the tube surface with multiple boiling regimes and to assess the benefits of closely coupling thermalhydraulics modelling and fuel/fuel channel behaviour modelling. CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a one-dimensional, two-fluid thermalhydraulic simulation code designed by AECL to analyse two-phase flow and heat transfer in piping networks. The detailed heat transfer package in CATHENA allows a connection to be established from the multiple solid surfaces of tubes to the surrounding large amount of moderator water, which acts as a heat sink during a postulated loss of coolant event. The generalized heat transfer package within CATHENA allows the tube walls to be divided into several layers in the radial direction and several sectors in the circumferential direction, to account for heat transfer conditions in these two directions. The CATHENA code with the generalized heat transfer package is capable of capturing key pool-boiling phenomena such as nucleate, transition and film boiling heat transfer as well as an ability to model the rewet phenomenon to some extent. A CATHENA input model was generated and used in simulations of selected contact boiling experiment test cases. The transient wall temperatures have been calculated in different portions of the calandria tube. By using this model an adequate agreement was achieved between CATHENA calculation and experimental measurement The CATHENA code enables one to investigate the transient and local thermal-mechanical behaviour of the calandria tube

  5. Shock tubes: compressions in the low pressure chamber

    International Nuclear Information System (INIS)

    Schins, H.; Giuliani, S.

    1986-01-01

    The gas shock tube used in these experiments consists of a low pressure chamber and a high pressure chamber, divided by a metal-diaphragm-to-rupture. In contrast to the shock mode of operation, where incident and reflected shocks in the low pressure chamber are studied which occur within 3.5 ms, in this work the compression mode of operation was studied, whose maxima occur (in the low pressure chamber) about 9 ms after rupture. Theoretical analysis was done with the finite element computer code EURDYN-1M, where the computation was carried out to 30 ms

  6. Inert medium (helium) irradiation testing of pressure tube samples

    International Nuclear Information System (INIS)

    Ancuta, M.; Radu, V.; Stefan, V.; Preda, M.

    2001-01-01

    Irradiation tests currently performed in C-5 capsule aim at obtaining data and information concerning behavior to irradiation of pressure tubes of CANDU type fuel channel, to evidence the factors limiting operation life span. A calculation code for analysis and prediction of pressure tube behavior should be based upon periodical inspection results, post irradiation examination of the removed from reactor pressure tubes as well as on the experimental results obtained with materials subjected to irradiation conditions identical with the operational ones. Mechanical behavior analysis should focus both complex thermal-mechanical type stresses and mechanical properties alteration under irradiation. The experimental results should be applied: - to evaluate the irradiation effects upon mechanical properties of Zr-2.5% Nb exposed to fluences up to 10 21 n·cm -2 ; - to gather data concerning the real stress / real deformation characteristic from which characteristic quantities can be deduced as, for instance, elasticity modulus, plasticity modulus, exponent of stress term in the Tsu-Berteles relation, to be used within the CANTUP simulation code describing pressure tube behavior, currently developed at INR Pitesti; - to develop prediction methods of pressure tube behavior and merging with in-service inspection procedure in order to forecast the life span and the proper timing for replacement before major failures occur. The samples irradiated in C-5 capsule were extracted from the ends of Zr-2.5% Nb pressure tubes resulting from Cernavoda NPP Unit 1. The samples for tensile tests were extracted on longitudinal and transversal directions of the pressure tube. The tests were carried out under following conditions: - test environment temperature, 260 - 280 deg.C; - testing medium, helium at 1 - 6 b pressure; - neutron flux (E n > 1 MeV), 1 - 2 · 10 13 ncm -2 s -1 ; - neutron fluence (E n > 1 MeV), 4 · 10 20 ncm -2 . The following characteristics were obtained from tensile

  7. High temperature deformation behavior of gradually pressurized zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Suzuki, Motoye

    1982-03-01

    In order to obtain preliminary perspectives on fuel cladding deformation behavior under changing temperature and pressure conditions in a hypothetical loss-of-coolant accident of PWR, a Zircaloy-4 tube burst test was conducted in both air and 99.97% Ar atomospheres. The tubes were directly heated by AC-current and maintained at various temperatures, and pressurized gradually until rupture occurred. Rupture circumferential strains were generally larger in Ar gas than in air and attained a maximum around 1100 K in both atmospheres. Some tube tested in air produced axially-extended long balloons, which proved not to be explained by such properties or ideas as effect of cooling on strain rate, superplasticity, geometrical plastic instability and stresses generated by surface oxide layer. A cause of the long balloon may be obtained in the anisotropy of the material structure. But even a qualitative analysis based on this property can not be made due to insufficient data of the anisotropy. (author)

  8. Pressure heat pumping in the orifice pulse-tube refrigerator

    International Nuclear Information System (INIS)

    Boer, P.C.T. de

    1996-01-01

    The mechanism by which heat is pumped as a result of pressure changes in an orifice pulse-tube refrigerator (OPTR) is analyzed thermodynamically. The thermodynamic cycle considered consists of four steps: (1) the pressure is increased by a factor π 1 due to motion of a piston in the heat exchanger at the warm end of the regenerator; (2) the pressure is decreased by a factor π 2 due to leakage out of the orifice; (3) the pressure is further decreased due to motion of the piston back to its original position; (4) the pressure is increased to its value at the start of the cycle due to leakage through the orifice back into the pulse tube. The regenerator and the heat exchangers are taken to be perfect. The pressure is assumed to be uniform during the entire cycle. The temperature profiles of the gas in the pulse tube after each step are derived analytically. Knowledge of the temperature at which gas enters the cold heat exchanger during steps 3 and 4 provides the heat removed per cycle from this exchanger. Knowledge of the pressure as a function of piston position provides the work done per cycle by the piston. The pressure heat pumping mechanism considered is effective only in the presence of a regenerator. Detailed results are presented for the heat removed per cycle, for the coefficient of performance, and for the refrigeration efficiency as a function of the compression ratio π 1 and the expansion ratio π 2 . Results are also given for the influence on performance of the ratio of specific heats. The results obtained are compared with corresponding results for the basic pulse-tube refrigerator (BPTR) operating by surface heat pumping

  9. Gas pressure in bubble attached to tube circular outlet

    OpenAIRE

    Salonen, A; Gay, Cyprien; Maestro, A; Drenckhan, W; Rio, Emmanuelle

    2016-01-01

    In the present Supplementary notes to our work ``Arresting bubble coarsening: A two-bubble experiment to investigate grain growth in presence of surface elasticity'' (accepted in EPL), we derive the expression of the gas pressure inside a bubble located above and attached to the circular outlet of a vertical tube.

  10. Method of detecting leakage from sealing attached to pressure tube

    International Nuclear Information System (INIS)

    Tomomatsu, Ken-ichi; Hayashi, Ken-ichi.

    1990-01-01

    The present invention provides a detection method for measuring the amount of water leaked from sealings attached to the lower end of a pressure tube. That is, the lower end of the pressure tube is sealed only by a metal sealing. A capturing vessel is placed under the pressure tube for capturing the leaked water dropping from the lower end of the pressure tube and the weight of the leaked water is measured on every capturing vessels to determine the amount of the leaked water. The leakage detection method based on the weight measurement has higher accuracy compared with a conventional volume measuring method using a water level gauge as described below. For example, if the volume of the captured water is 10cc, an error of about 0.1cc is caused by the volume measuring method using the water level gauge, whereas if 10g (10cc) weight of water is measured by using an accurate balance, error is only about 10 -4 g (10 -4 cc). Accordingly, the method of the present invention can measure at an accuracy about 1000 times as high as the conventional method. (I.S.)

  11. Characterization of magnetically impelled arc butt welded T11 tubes for high pressure applications

    Directory of Open Access Journals (Sweden)

    R. Sivasankari

    2015-09-01

    Full Text Available Magnetically impelled arc butt (MIAB welding is a pressure welding process used for joining of pipes and tubes with an external magnetic field affecting arc rotation along the tube circumference. In this work, MIAB welding of low alloy steel (T11 tubes were carried out to study the microstructural changes occurring in thermo-mechanically affected zone (TMAZ. To qualify the process for the welding applications where pressure could be up to 300 bar, the MIAB welds are studied with variations of arc current and arc rotation time. It is found that TMAZ shows higher hardness than that in base metal and displays higher weld tensile strength and ductility due to bainitic transformation. The effect of arc current on the weld interface is also detailed and is found to be defect free at higher values of arc currents. The results reveal that MIAB welded samples exhibits good structural property correlation for high pressure applications with an added benefit of enhanced productivity at lower cost. The study will enable the use of MIAB welding for high pressure applications in power and defence sectors.

  12. Pressure distribution over tube surfaces of tube bundle subjected to two phase cross flow

    International Nuclear Information System (INIS)

    Sim, Woo Gun

    2013-01-01

    Two phase vapor liquid flows exist in many shell and tube heat exchangers such as condensers, evaporators and nuclear steam generators. To understand the fluid dynamic forces acting on a structure subjected to a two phase flow, it is essential to obtain detailed information about the characteristics of a two phase flow. The characteristics of a two phase flow and the flow parameters were introduced, and then, an experiment was performed to evaluate the pressure loss in the tube bundles and the fluid dynamic force acting on the cylinder owing to the pressure distribution. A two phase flow was pre mixed at the entrance of the test section, and the experiments were undertaken using a normal triangular array of cylinders subjected to a two phase cross flow. The pressure loss along the flow direction in the tube bundles was measured to calculate the two phase friction multiplier, and the multiplier was compared with the analytical value. Furthermore, the circular distributions of the pressure on the cylinders were measured. Based on the distribution and the fundamental theory of two phase flow, the effects of the void fraction and mass flux per unit area on the pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure on the tube by a numerical method. It was found that for low mass fluxes, the measured two phase friction multipliers agree well with the analytical results, and good agreement for the effect of the void fraction on the drag coefficients, as calculated by the measured pressure distributions, is shown qualitatively, as compared to the existing experimental results

  13. An evaluation of the statistical variability in thermal expansion properties of steam generator tubesheet (SA-508) and tubing (Alloy-600TT)

    International Nuclear Information System (INIS)

    Riccardella, P.C.; Staples, J.F.; Kandra, J.T.

    2009-01-01

    Inspections of steam generator tubing are performed in U.S. PWRs as part of the Steam Generator Management Program. Westinghouse has recently completed a technical justification demonstrating that in steam generators with thermally treated Ni-Cr Alloy (Alloy 600TT) tubes that are hydraulically expanded into low alloy steel (SA-508) tubesheets, flaws in the region of the tubes below a certain distance from the top of the tubesheet, denoted H * , will not result in reactor coolant pressure boundary breach nor unacceptable primary-to-secondary leakage. This is because, even if a flaw in this region were to result in complete tube sever, if the length of undegraded tube in the tubesheet exceeds H*, neither operating nor accident loadings create sufficient pull-out forces to overcome the frictional forces between the tube and tubesheet. One key component of this technical justification is the differential thermal expansion between the tube and tubesheet, since a significant portion of the pullout strength of the hydraulically expanded tube-to-tubesheet joint is due to mechanical interference resulting from the larger expansion of the tubing relative to the tubesheet at a given temperature. To address this phenomenon, a detailed statistical evaluation of coefficient of thermal expansion (CTE) data for the tubesheet material (SA-508) and the tube material (thermally treated Alloy-600) was performed. Data used in the evaluation included existing test results obtained from a number of sources as well as extensive new laboratory data developed specifically for this purpose. The evaluation resulted in recommended statistical distributions of this property for the two materials including their means and probabilistic variability. In addition, it was determined that the CTE values reported in the ASME Code (Section II) represent reasonably conservative mean values for both the tubesheet and tubing material. (author)

  14. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

  15. Fuel bundle to pressure tube fretting in Bruce and Darlington

    Energy Technology Data Exchange (ETDEWEB)

    Norsworthy, A G; Ditschun, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs.

  16. Fuel bundle to pressure tube fretting in Bruce and Darlington

    International Nuclear Information System (INIS)

    Norsworthy, A.G.; Ditschun, A.

    1995-01-01

    As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs

  17. Experimental determination of thermal contact conductance between pressure and calandria tubes of Indian pressurised heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dureja, A.K., E-mail: akdureja@barc.gov.in [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Pawaskar, D.N.; Seshu, P. [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Mumbai (India); Sinha, S.K. [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Sinha, R.K. [Department of Atomic Energy, OYC, Near Gateway of India, Mumbai (India)

    2015-04-01

    Highlights: • We established an experimental facility to measure thermal contact conductance between disc shaped specimens. • We measured thermal contact conductance between Zr-2.5Nb alloy pressure tube (PT) material and Zr-4 calandria tube (CT) material. • We concluded that thermal contact conductance is a linear function of contact pressure for interface of PT and CT up to 10 MPa contact pressure. • We concluded that thermal contact conductance is a weak function of interface temperature. - Abstract: Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a tube type heavy water nuclear reactor, the pressure tube is expected to sag/balloon and come in contact with outer cooler calandria tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure tube and calandria tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper.

  18. Determination of delayed hydride cracking velocity of CANDU Zr-2.5Nb pressure tube

    International Nuclear Information System (INIS)

    Kim, Young Suk; Kim, Chan Jung; Rheem, Y. W.; Im, K. S.; Kwon, Sang Chul

    2000-07-01

    As agreed upon the contract with an IAEA Co-ordinated Research Project 'Hydrogen and Hydride Induced Degradation of the Mechanical and Physical Properties of Zirconium Based Alloys', we conducted DHC tests at 3 different temperatures of 144, 182 and 250 deg C on the curved compact tension specimens made from a Zr-2.5Nb pressure tube. Additional tests were carried out at 200 and 230 deg C with an aim to determine the activation energy for delayed hydride cracking. This report summarizes the results of DHC tests obtained so far. All the DHC tests were conducted in accordance with the procedures suggested by the Host Lab. 7 DHCV values determined at the same temperature such as 250 deg C show very low standard deviation, whose average values are very comparable to those reported by the participants. Thus, one of the most important results we have got is that we establish qualified DHC testing procedure through the IAEA CRP. An activation energy for DHC of unirradiated Zr-2.5Nb pressure tube was 49 KJ/mol which is very similar to the activation energy of 43 KJ/mol for irradiated Zr-2.5Nb pressure tubes. DHCV increased linearly with the hydrogen content up to around 25 ppm and then became saturated at higher hydrogen concentration

  19. Burst pressure and leak rate from fretted SG tubes

    International Nuclear Information System (INIS)

    Hwang, Seong Sik; Jung, Man Kyo; Kim, Hong Pyo; Kim, Joung Soo

    2005-01-01

    Steam generator(SG) tubes of a pressurized water reactor(PWR) have suffered from various types of corrosion, such as pitting, wastage and stress corrosion cracking (SCC) on both the primary and secondary side. Recently, fretting/wear degradation at the tube support region has been reported in some Korean nuclear power plants. In order to prevent the primary coolant from leaking to the secondary side, the tubes are repaired by a sleeving or plugging. It is important to establish the repair criteria to assure a reactor integrity and yet maintain the plugging ratio within the limits needed for an efficient operation. The objective of the burst test is to obtain a relationship between the burst/leak rate and the shape of the fretted flaws machined with an electro discharge machining (EDM)

  20. Flooding of a large, passive, pressure-tube LWR

    Energy Technology Data Exchange (ETDEWEB)

    Hejzlar, P.; Todreas, N.E.; Driscoll, M.J. [Massachusetts Institute of Technology, Cambridge, MA (United States)

    1995-09-01

    A reactor concept has been developed which can survive LOCA without scram and without replenishing primary coolant inventory. The proposed concept is a pressure tube type reactor similar to CANDU reactors, but differing in three key aspects: (1) a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles, (2) the heavy water coolant in the pressure tubes is replaced by light water, and (3) the calandria tank contains a low pressure gas instead of heavy water moderator. The gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents, it allows passive calandria flooding. This paper describes the thermal hydraulic characteristics of the gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube LWR concept. The flooding of the top row of fuel channels must be accomplished fast enough so that none of the critical components of the fuel channel exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. Two other considerations are important. The thermal shock experienced by the calandria and pressure tubes has been evaluated and shown to be within acceptable bounds. Finally, although complete flooding renders the reactor deeply subcritical, various steam/water densities can be hypothesized to be present during the flooding process which could cause reactivity to increase from the initially voided calandria case. One such hypothesis which leads to the maximum possible density of the steam/water mixture in the still unflooded calandria space is entrainment from the free surface. It is shown that the steam/water mixture density yielding the maximum reactivity peak cannot be achieved by entrainment because it exceeds thermohydraulically attainable densities of steam/water by an order of magnitude.

  1. Possible first occurrence of external corrosion on alloy 600TT tubes in France

    International Nuclear Information System (INIS)

    Boccanfuso, M.; Thebault, Y.; Massini, B.; Bigne, L.

    2015-01-01

    During the last decade, in different countries, several occurrences of external corrosion have been identified on steam generator (SG) tube bundles equipped with thermally treated 600 alloy. In France, this feedback leads EDF to enhance the SG inspection program. Nevertheless, until now, no damage of this type was reported. Recently, during in-service inspection at the Cattenom plant on a SG equipped with alloy 600TT tubes, Eddy current tests have highlighted a signal that could be related to external corrosion. The tube was removed and sent to the EDF hot laboratory for destructive examinations. Various exams were performed at different scales to characterize the causes of this NDT signal, the material properties and the residual stresses. The assessments carried out on the tube conclude that the source of the damage is external intergranular stress corrosion cracking, also called ODSCC (Outside Diameter Stress Corrosion Cracking) making it the first occurrence on the tube bundles made of alloy 600TT in the French fleet. This first case of 600 TT ODSCC in France is an unexpected and particular one, because of its altitude in the full mechanical rolling area. This is reinforced by the low number of occurrences noted to date (only one after nearly 30 years of operation of alloy 600TT tube bundles). International (Biblis) OPEX had identified recent IGSCC with cracks initiated and propagated in the tubesheet. For this case, the scenario considered requires highly restrictive conditions (tube in the sludge zone and on the periphery of the tube bundle, including the tube lane) and may explain the singular nature of the Cattenom tube

  2. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Prodea, Iosif; Catana, Alexandru

    2010-01-01

    The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

  3. Integrity evaluation of Alloy 600 RV head penetration tubes in Korean PWR plants

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Park, Sung Ho; Hong, Sung Yull; Choi, Kwang Hee

    1995-01-01

    The structural integrity assessment of Alloy 600 RV head penetration tubes has been an important issue for the economical and reliable operation of power plants. In this paper, an overview of the integrity evaluation program for the RV head penetration tubes in Korean nuclear power plants is presented. Since the crack growth mechanism of the penetration tube is due to the primary water stress corrosion cracking (PWSCC) which is mainly related to the stress at the tube, the present paper consists of three primary activities: the stress evaluation, the flaw evaluation, and data generation through material and mechanical tests. (author). 5 refs, 2 figs, 1 tab

  4. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

  5. Evaluation of crack propagation of alloy 600 tube in high temperature water, (1)

    International Nuclear Information System (INIS)

    Hirano, Hideo; Kawamura, H.; Kawamura, Kohji; Matsubara, Masaaki

    1990-01-01

    This report describes the analysis of stress intensity factors at cracks in alloy 600 steam generator tubes. Based on the results of the analysis, IGA/SCC tests were carried out to examine the effect of stress intensity and water quality on the crack propagation rate. The main test result are as follows: (1) Hoop stress was caused by the pressure difference between the internal and external surface of the steam generator tube. The calculated hoop stress was about 7 kg/mm 2 . In addition, the temperature difference between the internal and external surface caused thermal stress. The thermal stress was about 10 kg/mm 2 at the external surface and the one at the internal surface was about -10 kg/mm 2 . Total stress at the external and internal surface was 17 kg/mm 2 and -3 kg/mm 2 , respectively. (2) The stress intensity factor at the crack tip increased with increasing crack length. For a long crack, the stress intensity factor decreased with increasing crack number. However, for a short crack, the stress intensity factor decreased little with increasing crack number. (3) Under high stress-intensity conditions, i.e. 40∼50 kg·mm -3/2 , the IGA/SCC test showed that IGA/SCC propagated in AVT and AVT/boric-acid solution at 320degC and 350degC. However, the propagation rate was low. (author)

  6. Quality Management and Control of Low Pressure Cast Aluminum Alloy

    Science.gov (United States)

    Zhang, Dianxi; Zhang, Yanbo; Yang, Xiufan; Chen, Zhaosong; Jiang, Zelan

    2018-01-01

    This paper briefly reviews the history of low pressure casting and summarizes the major production processes of low pressure casting. It briefly introduces the quality management and control of low pressure cast aluminum alloy. The main processes include are: preparation of raw materials, Melting, refining, physical and chemical analysis, K-mode inspection, sand core, mold, heat treatment and so on.

  7. Detection of pressure tube leaks relying on moisture beetles only

    International Nuclear Information System (INIS)

    Kenchington, J.M.; Choi, A.; Jin, Y.

    2004-01-01

    A major decision was made for Pickering NGS A Annulus Gas System (ACS) that detection of a pressure tube (PT) leak should be achieved by using only moisture beetles and that dew point monitors would provide 'early warning' without status to shut down the reactor. Experience with Unit 3 has shown that dew point monitoring of pressure tube leaks was particularly subject to gas leaks and surface adsorption effects. Unit 4 was the first one to be converted during the full scale pressure tube replacement programme. Because of the fundamental change in design philosophy, moisture injection tests were carried out during commissioning to demonstrate that performance matched design. In particular it was necessary to show that leak before break (LBB) would be achieved if a leak occurred in the limiting string. Units 1 and 3 have since been converted. No decision has been taken to convert Pickering B units as gas leaks are small and no significant adsorption effects are anticipated. Hence dew point monitoring will not be impaired. (author)

  8. Design of experiments and equipment to test the ballooning characteristics of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Forrest, C.F.; Stern, F.; Hart, R.G.

    1992-01-01

    Experiments have been planned and an apparatus has been designed to enable creep testing of end-of-life pressure tube specimens in a LOCA environment. Effects that could be studied include: annealing of irradiation damage during transient heating; effects of hydride blisters on pressure tube ballooning strains; and, effects of uniformly-distributed hydrogen content on pressure tube ballooning strains. The proposed experimental program will consist of separate effects creep tests on pressure tube sections under transient heating conditions

  9. Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

    2009-01-01

    An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

  10. Optimization of drift gases for accuracy in pressurized drift tubes

    CERN Document Server

    Kirchner, J J; Dinner, A R; Fidkowski, K J; Wyatt, J H

    2001-01-01

    Modern detectors such as ATLAS use pressurized drift tubes to minimize diffusion and achieve high coordinate accuracy. However, the coordinate accuracy depends on the exact knowledge of converting measured times into coordinates. Linear space-time relationships are best for reconstruction, but difficult to achieve in the $E \\propto \\frac{1}{r}$ field. Previous mixtures, which contained methane or other organic quenchers, are disfavored because of ageing problems. From our studies of nitrogen and carbon dioxide, two mixtures with only small deviations from linearity were determined and measured. Scaling laws for different pressures and magnetic fields are also given.

  11. Optimization of drift gases for accuracy in pressurized drift tubes

    International Nuclear Information System (INIS)

    Kirchner, J.J.; Becker, U.J.; Dinner, R.B.; Fidkowski, K.J.; Wyatt, J.H.

    2001-01-01

    Modern detectors such as ATLAS use pressurized drift tubes to minimize diffusion and achieve high coordinate accuracy. However, the coordinate accuracy depends on the exact knowledge of converting measured times into coordinates. Linear space-time relationships are best for reconstruction, but difficult to achieve in the E∝1/r field. Previous mixtures, which contained methane or other organic quenchers, are disfavored because of ageing problems. From our studies of nitrogen and carbon dioxide, two mixtures with only small deviations from linearity were determined and measured. Scaling laws for different pressures and magnetic fields are also given

  12. Full length channel Pressure Tube sagging under completely voided full length pressure tube of an Indian PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Negi, Sujay, E-mail: negi.sujay@gmail.com [Indian Institute of Technology, Roorkee 247667 (India); Kumar, Ravi, E-mail: ravikfme@gmail.com [Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Bhabha Atomic Research Centre, Mumbai 400085 (India); Mukopadhyay, D., E-mail: dmukho@barc.gov.in [Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2017-03-15

    Highlights: • At 16 kW/m input, thermal stability was attained at 595 °C, without PT-CT contact. • At 20 kW/m step input, PT-CT contact occurred at 637 °C near bottom-center of the tube. • PT integrity was maintained throughout the experiment. - Abstract: An experimental investigation was conducted to simulate the sagging behavior of a full length Pressure Tube of a channel of 220 MWe Indian PHWR. The investigation aimed to recreate a condition resembling Loss of Coolant Accident (LOCA) with Emergency Core Cooling System (ECCS) failure in a nuclear power plant. A full length channel assembly immersed in moderator was subjected to electrical resistance heating of Pressure Tube (PT) to simulate the residual heat after shutting down of reactor. The temperature of PT started rising and the contact between PT and CT was established at the center of the tube where average bottom temperature was 637 °C. The integrity of PT was maintained throughout the experiment and the PT heat up was arrested on contact with the CT due to transfer of heat to the moderator.

  13. Analytical TEM of service-induced SCC in alloy 600TT steam generator tubing

    International Nuclear Information System (INIS)

    Wolfe, R.; Legras, L.; Boccanfuso; Martin, A.

    2015-01-01

    In 2008, Vogtle Electric Generating Plant Unit 1 performed tube pulls to confirm outside diameter stress corrosion cracking (ODSCC) in a steam generator with thermally treated Alloy 600TT tubing. Subsequent metallographic and other laboratory work attributed the cracking to the non-optimal microstructure of the tubing and the elevated residual stresses at the expansion transition. In the current work, analytical transmission electron microscopy was performed to gain a better understanding of this in-service cracking through a detailed characterization of the oxides and crack tips. These examinations, which are the first of this kind for U.S. Alloy 600TT tubing service cracks, detected lead (Pb) in the region of the top-of-tube sheet crevice, in oxides at the crack tips, and at degraded grain boundaries. In addition, sulfur was observed in oxides on the outside surface of the tube in the free span area. The presence of Pb at the crack tip and the lack of plasticity on the observed failure surfaces suggest that the environment played a predominant role in the cracking of this tubing with a non-optimal microstructure. The significance of the degradation will be discussed in the context of overall corrosion indications in Alloy 600TT steam generators in the United States. (authors)

  14. Neutron radiography of irradiated zircaloy coupons of pressure tubes from PHWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Gangotra, S; Ouseph, P M; Tamhane, A B; Singh, H N; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

    1994-12-31

    The Indian Pressurised Heavy Water Reactors (PHWR`s) are of CANDU type, consisting of 304 zircaloy-2 pressure tubes. These pressure tubes contain the fuel bundles, where the heat is generated and is removed by the heavy water flowing through these pressure tubes at high temperature and pressure. These pressure tubes are surrounded by the calandria tubes, and are separated from them by a pair of garter springs. Over a period of time, as a result of the irradiation creep and assisted by the displacement of the garter springs, the hot pressure tube may come in contact with the cold calandria tube. This would result in the hydrogen migrating to the cold contact location and formation of hydride blisters. These blisters could eventually rupture the pressure tube by the DHC (delayed hydrogen cracking) mechanism. 2 refs., 2 figs.

  15. Middle Ear Pressure Regulation - Complementary Action of the Mastoid and Eustachian Tube

    DEFF Research Database (Denmark)

    Gaihede, Michael; Dirckx, Joris J J; Jacobsen, Henrik

    2010-01-01

    , MEP counter-regulation presented as Eustachian tube openings with steep and fast pressure changes toward 0 Pa, whereas in others, gradual and slow pressure changes presented related to the mastoid; these changes sometimes crossed 0 Pa into opposite pressures. In many cases, combinations...... to continuous regulation of smaller pressures, whereas the tube was related to intermittent regulation of higher pressures....

  16. Middle Ear Pressure Regulation - Complementary Action of the Mastoid and Eustachian Tube

    DEFF Research Database (Denmark)

    Gaihede, Michael; Jacobsen, Henrik; Tveterås, Kjell

    , MEP counter-regulation presented as Eustachian tube openings with steep and fast pressure changes toward 0 Pa, whereas in others, gradual and slow pressure changes presented related to the mastoid; these changes sometimes crossed 0 Pa into opposite pressures. In many cases, combinations...... to continuous regulation of smaller pressures, whereas the tube was related to intermittent regulation of higher pressures....

  17. Apparatus for inspecting and repairing a pressurized-water reactor's steam generator heat exchanger tubes

    International Nuclear Information System (INIS)

    Mueller, O.; Roettger, H.; Kasti, H.; Hagen, H.G.

    1976-01-01

    Described is an apparatus provided for use with a pressurized-water reactor' steam generator having a manifold chamber enclosing the bottom side of a horizontal tube sheet having holes therethrough in which are mounted the tubes of a heat exchanger tube bundle. The manifold chamber has a manhole giving access to the tube's bottom side to permit internal inspection or repair of the tubes by registration of an end of a flexible guide conduit with the tube sheet holes and through which a flexible carrier can be guided for insertion via these holes in the tube sheet and through the tubes extending from the tube sheet's other side

  18. Method of reactivity control in pressure tube reactor

    International Nuclear Information System (INIS)

    Fukumura, Nobuo.

    1988-01-01

    Purpose: To provide a method of controlling reactivity in a pressure tube reactor at high conversion ratio intended for high burn-up degree. Method: Control tubes are inserted in heavy water moderator. Light water is filled in the tubes at the initial burning stage. Along with the advance of the burning, the light water is gradually removed and replaced with gases of less reactive nuclear reactivity with neutrons such as air or gaseous carbon dioxide. The tubes are made of less neutron absorbing material such as aluminum. By filling light water, infinite multiplication factor is reduced to suppress the reactivity at the initial burning stage. As light water is gradually removed and replaced with air, etc., it provides an effect like that elimination of heavy water moderator to increase the conversion ratio. Accordingly, nuclear fission materials are produced additionally by so much to extend the burn-up degree. In this way, it can provide excellent effect in realizing high burn-up ratio and high conversion ratio. (Kamimura, M.)

  19. The development of octagon Zr-4 alloy tube for heating reactors

    International Nuclear Information System (INIS)

    Yang Fanglin; Yang Yingli; Wang Guangshen

    1989-10-01

    The asymmetrical octagon Zr-4 alloy tubes which are used for fuel assembly in the heating reactor have been developed. The thickness of tube wall is 1.5 mm and the length is 1725 mm. The long side of the octagon is 138.7 0.3 +0.2 mm, the short side is 93.1 ± 0.1 mm. To manufacture these tubes a stretch draw forming processing method is adopted. The process is divided into two phases. In the first phase, a short draw mould is used to stretch the Zr-4 alloy tube. In the second phase, a long draw mould, its length is equal to the end-produt length, is used to complete the final processing. The size accuracy and repeatability of this method are excellent and can fully meet the design requirements

  20. A statistical method for draft tube pressure pulsation analysis

    International Nuclear Information System (INIS)

    Doerfler, P K; Ruchonnet, N

    2012-01-01

    Draft tube pressure pulsation (DTPP) in Francis turbines is composed of various components originating from different physical phenomena. These components may be separated because they differ by their spatial relationships and by their propagation mechanism. The first step for such an analysis was to distinguish between so-called synchronous and asynchronous pulsations; only approximately periodic phenomena could be described in this manner. However, less regular pulsations are always present, and these become important when turbines have to operate in the far off-design range, in particular at very low load. The statistical method described here permits to separate the stochastic (random) component from the two traditional 'regular' components. It works in connection with the standard technique of model testing with several pressure signals measured in draft tube cone. The difference between the individual signals and the averaged pressure signal, together with the coherence between the individual pressure signals is used for analysis. An example reveals that a generalized, non-periodic version of the asynchronous pulsation is important at low load.

  1. Experiment on the effects of contact between the pressure tube and the fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Y; Fujii, Y [Electric Power Development Co. Ltd., Tokyo (Japan); Kato, K [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

    1996-12-31

    The Advanced Thermal Reactor (ATR) is a pressure tube type reactor in which the fuel assembly is located close to the pressure tube. The ATR has a structure which is such that the thermal stretch of the fuel pin is not limited by the spacer if the fuel pin dries out. Accordingly. it is not thought that the fuel pin contacts the pressure tube due to large transformations around the Design Based Event (DBE). Nevertheless, the safety margin must be kept in case the over-DBE. We have confirmed in this experiment that the temperature of the pressure tube does not increase to the critical level when the fuel pin contacts the pressure tube and the functions of the pressure tube are maintained as a pressure boundary. Further, we analyzed the safety margin of the pressure tube using the data from this experiment and from code analysis. (author). 10 tabs., 32 figs.

  2. Rejuvenation of service exposed ammonia cracker tubes of cast Alloy 625 and their re-use

    Energy Technology Data Exchange (ETDEWEB)

    Singh, J.B., E-mail: jbsingh@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Verma, A. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Jaiswal, D.M.; Kumar, N.; Patel, R.D. [Heavy Water Board, Department of Atomic Energy, Anushakti Nagar, Mumbai 400094 (India); Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2015-09-17

    This study is an extension of a previous study undertaken to rejuvenate ammonia cracker tubes of Alloy 625 alloy that have been service exposed in heavy water plants for their full service life of 100,000 h. The service exposure caused significant microstructural modifications and deterioration in mechanical properties, and a solution annealing treatment of 2 h at 1160 °C rejuvenated all properties similar to those of the virgin alloy. The present study reports the evolution of microstructure and mechanical properties of a full service exposed centrifugally cast Alloy 625 tube that was put into service again for 55,000 h after receiving a rejuvenation treatment. During the second service, microstructural modifications, increase in strength and loss of ductility were on the lines of the work reported earlier. However, it was encouraging to observe that degraded properties after the second service life remained within the bounds of those of virgin and full service exposed tubes. The good performance of the rejuvenated tube during the second service life has been attributed to good control of operation parameters that limited the precipitation of grain boundary carbides during the first service life, which otherwise would have had a direct bearing on premature failure of tubes during their second service life.

  3. KTA 625 alloy tube with excellent corrosion resistance and heat resistance

    International Nuclear Information System (INIS)

    Fujiwara, Kazuo; Kadonaga, Toshiki; Kikuma, Seiji.

    1982-01-01

    The problems when seamless tubes are produced by using nickel base 625 alloy (61Ni-22Cr-9Mo-Cb) which is known as a corrosion resistant and heat resistant alloyF were examined, and the confirmation experiment was carried out on its corrosion resistance and heat resistance. Various difficulties have been experienced in the tube making owing to the characteristics due to the chemical composition, but they were able to be solved by the repeated experiments. As for the characteristics of the product, the corrosion resistance was excellent particularly in the environment containing high temperature, high concentration chloride, and also the heat resistance was excellent in the wide temperature range from normal temperature to 1000 deg C. From these facts, the wide fields of application are expected for these alloy tubes, including the evaporation and concentration equipment for radioactive wastes in atomic energy field. Expecting the increase of demand hereafter, Kobe Steel Ltd. examined the problems when seamless tubes are produced from the 625 alloy by Ugine Sejournet process. The aptitude for tube production such as the chemical composition, production process and the product characteristics, the corrosion resistance against chloride, hydrogen sulfide, polythionic and other acids,F the high temperature strength and oxidation resistance are reported. (Kako, I.)

  4. Brittle-fracture potential of irradiated Zircaloy-2 pressure tubes

    Science.gov (United States)

    Huang, F. H.

    1993-12-01

    Neutron irradiation can degrade the fracture toughness of Zircaloy-2 and may cause highly irradiated reactor components of this material to fail in a brittle manner. The effects of radiation embrittlement on the structural integrity of N Reactor pressure tubes are studied by performing KIc and JIc fracture toughness testing on samples cut from the Zircaloy-2 tubes periodically removed from the reactor. A fluence of 6 × 10 25n/ m2 ( E > 1 MeV) reduced the fracture toughness of the material by 40 to 50%. The fracture toughness values appear to saturate at 260°C with fluences above 3 × 10 25n/ m2 ( E > 1 MeV), but continue to decline with increasing fluence at temperatures below 177°C. Present and previous results obtained from irradiated pressure tubes indicate that the brittle-fracture potential of Zircaloy-2 increases with decreasing temperature and increasing fluence. Fractographic examinations of the fracture surfaces of irradiated samples reveal that circumferential hydride formation significantly influenced fracture morphology by providing sites for easy crack nucleation and leaving deep cracks. However, the deep cracks created at the hydride platelets in specimens containing less than 220 ppm hydrogen are not believed to be the major cause of degradation in postirradiation fracture toughness.

  5. Fracture Toughness Round Robin Test International in pressure tube materials

    International Nuclear Information System (INIS)

    Villagarcia, M.P.; Liendo, M.F.

    1993-01-01

    Part of the pressure tubes surveillance program of CANDU type reactors is to determine the fracture toughness using a special fracture specimen and test procedure. Atomic Energy of Canada Limited decided to hold a Round Robin Test International and 9 laboratories participated worldwide in which several pressure tube materials were selected: Zircaloy-2, Zr-2.5%Nb cold worked and Zr-2.5%Nb heat treated. The small specimens used held back the thickness and curvature of the tube. J-R curves at room temperature were obtained and the crack extension values were determined by electrical potential drop techniques. These values were compared with results generated from other laboratories and a bid scatter was founded. It could be due to slight variations in the test method or inhomogeneity of the materials and a statistical study must be done to see if there is any pattern. The next step for the Round Robin Test would be to make some modifications in the test method in order to reduce the scatter. (Author)

  6. Frictional pressure drop of high pressure steam-water two-phase flow in internally helical ribbed tubes

    International Nuclear Information System (INIS)

    Tingkuan, C.; Xuanzheng, C.

    1987-01-01

    It is well known that the internally helical ribbed tubes are effective in suppressing the dry-out in boiling tubes at high pressures, so they are widely used as furnace water wall tubes in modern large steam power boilers. Design of the boilers requires the data on frictional pressure drop characteristics of the ribbed tubes, but they are not sufficient now. This paper describes the experimental results on the adiabatic frictional pressure drop in both horizontal ribbed tubes with measured mean inside diameter of 11.69 mm and 35.42 mm at high pressure from 10 to 21 MPa, mass flow rate from 350 to 3800 kg/m/sup 2/s and steam quality from 0 to 1 in our high pressure electrically heated water loop. Simultaneously, both smooth tubes under the same conditions for comparison. Based on the tests the correlation for determining the frictional pressure drop of internally ribbed tubes are proposed

  7. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    International Nuclear Information System (INIS)

    Majumdar, S.; Kasza, K.

    2009-01-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  8. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S.; Kasza, K. [Argonne National Laboratory, Nuclear Energy Division, Lemont, Illinois (United States)

    2009-07-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  9. Fabrication of Aluminum Tubes Filled with Aluminum Alloy Foam by Friction Welding

    Directory of Open Access Journals (Sweden)

    Yoshihiko Hangai

    2015-10-01

    Full Text Available Aluminum foam is usually used as the core of composite materials by combining it with dense materials, such as in Al foam core sandwich panels and Al-foam-filled tubes, owing to its low tensile and bending strengths. In this study, all-Al foam-filled tubes consisting of ADC12 Al-Si-Cu die-cast aluminum alloy foam and a dense A1050 commercially pure Al tube with metal bonding were fabricated by friction welding. First, it was found that the ADC12 precursor was firmly bonded throughout the inner wall of the A1050 tube without a gap between the precursor and the tube by friction welding. No deformation of the tube or foaming of the precursor was observed during the friction welding. Next, it was shown that by heat treatment of an ADC12-precursor-bonded A1050 tube, gases generated by the decomposition of the blowing agent expand the softened ADC12 to produce the ADC12 foam interior of the dense A1050 tube. A holding time during the foaming process of approximately tH = 8.5 min with a holding temperature of 948 K was found to be suitable for obtaining a sound ADC12-foam-filled A1050 tube with sufficient foaming, almost uniform pore structures over the entire specimen, and no deformation or reduction in the thickness of the tube.

  10. An improved model to predict nonuniform deformation of Zr-2.5 Nb pressure tubes

    International Nuclear Information System (INIS)

    Lei, Q.M.; Fan, H.Z.

    1997-01-01

    Present circular pressure-tube ballooning models in most fuel channel codes assume that the pressure tube remains circular during ballooning. This model provides adequate predictions of pressure-tube ballooning behaviour when the pressure tube (PT) and the calandria tube (CT) are concentric and when a small (<100 degrees C) top-to-bottom circumferential temperature gradient is present on the pressure tube. However, nonconcentric ballooning is expected to occur under certain postulated CANDU (CANada Deuterium Uranium) accident conditions. This circular geometry assumption prevents the model from accurately predicting nonuniform pressure-tube straining and local PT/CT contact when the pressure tube is subjected to a large circumferential temperature gradient and consequently deforms in a noncircular pattern. This paper describes an improved model that predicts noncircular pressure-tube deformation. Use of this model (once fully validated) will reduce uncertainties in the prediction of pressure-tube ballooning during a postulated loss-of-coolant accident (LOCA) in a CANDU reactor. The noncircular deformation model considers a ring or cross-section of a pressure tube with unit axial length to calculate deformation in the radial and circumferential directions. The model keeps track of the thinning of the pressure-tube wall as well as the shape deviation from a reference circle. Such deviation is expressed in a cosine Fourier series for the lateral symmetry case. The coefficients of the series for the first m terms are calculated by solving a set of algebraic equations at each time step. The model also takes into account the effects of pressure-tube sag or bow on ballooning, using an input value of the offset distance between the centre of the calandria tube and the initial centre of the pressure tube for determining the position radius of the pressure tube. One significant improvement realized in using the noncircular deformation model is a more accurate prediction in

  11. PLUSS-A weldless leaktight sleeve for alloy 600/690 steam generator tubes

    International Nuclear Information System (INIS)

    Potz, F.; Bohmann, W.

    1998-01-01

    The ABB PLUSS sleeving represents a new SG tube repair technique qualified and approved to replace in the future most of the plugging as well as welded sleeving. Basically the advantages of an innovative combination of both alloys 600/690 and 800 are taken into consideration. The upper sleeve/SG tube-joint is hydraulically expanded stressing the SG tube only within the elastic range. The lower joint is hard rolled. The installation processes are simple and reproducible, fast, computerized and individually recorded. The operating temperature range of the sleeved SG-tube is effectively reduced so that any further corrosion is impeded. Both, sleeve and SG tube are fully inspectable by ECT. (author)

  12. Development of technology on the material surveillance of CANDU pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Kye Hoh; Han, Jung Hoh; Lee, Duk Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-05-01

    Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) Manufacture of test equipments and test (DHCV, hydriding, grip and tensile specimen etc). 23 figs, 6 tabs, 59 refs. (Author).

  13. First Research Coordination Meeting on Prediction of Axial and Radial Creep in HWR Pressure Tubes. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    Pressure tube deformation is a critical aging issue in operating Heavy Water Reactors (HWRs). According to the service year, horizontal pressure tubes have three kinds of deformation: diametral creep leading to the flow bypass and the penalty to critical heat flux for fuel rods, longitudinal creep leading to the interference of feeder pipes and/or with fuelling machine, and sagging leading to the interference with in-core components and potential contact between the pressure tube and calandria tube. The CRP scope includes the establishment of a database for pressure tube deformation, microstructure characterization of pressure tube materials collected from HWRs currently operating in Member States and development of a prediction model for pressure tube deformation

  14. Development of technology on the material surveillance of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Noh, Kye Hoh; Han, Jung Hoh; Lee, Duk Hyun

    1995-05-01

    Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) Manufacture of test equipments and test (DHCV, hydriding, grip and tensile specimen etc). 23 figs, 6 tabs, 59 refs. (Author)

  15. Characterization of Tubing from Advanced ODS alloy (FCRD-NFA1)

    International Nuclear Information System (INIS)

    Maloy, Stuart Andrew; Aydogan, Eda; Anderoglu, Osman; Lavender, Curt; Anderson, Iver; Rieken, Joel; Lewandowski, John; Hoelzer, Dave; Odette, George R.

    2016-01-01

    Fabrication methods are being developed and tested for producing fuel clad tubing of the advanced ODS 14YWT and FCRD-NFA1 ferritic alloys. Three fabrication methods were based on plastically deforming a machined thick-wall tube sample of the ODS alloys by pilgering, hydrostatic extrusion or drawing to decrease the outer diameter and wall thickness and increase the length of the final tube. The fourth fabrication method consisted of the additive manufacturing approach involving solid-state spray deposition (SSSD) of ball milled and annealed powder of 14YWT for producing thin-wall tubes. Of the four fabrication methods, two methods were successful at producing tubing for further characterization: production of tubing by high-velocity oxy-fuel spray forming and production of tubing using high-temperature hydrostatic extrusion. The characterization described shows through neutron diffraction the texture produced during extrusion while maintaining the beneficial oxide dispersion. In this research, the parameters for innovative thermal spray deposition and hot extrusion processing methods have been developed to produce the final nanostructured ferritic alloy (NFA) tubes having approximately 0.5 mm wall thickness. Effect of different processing routes on texture and grain boundary characteristics has been investigated. It was found that hydrostatic extrusion results in combination of plane strain and shear deformations which generate rolling textures of ?- and ?-fibers on and together with a shear texture of ?-fiber on and . On the other hand, multi-step plane strain deformation in cross directions leads to a strong rolling textures of ?- and ?-fiber on together with weak ?-fiber on . Even though the amount of the equivalent strain is similar, shear deformation leads to much lower texture indexes compared to the plane strain deformations. Moreover, while 50% of hot rolling brings about a large number of high-angle grain boundaries (HAB), 44% of shear deformation results

  16. Simulating Porous Magnetite Layer Deposited on Alloy 690TT Steam Generator Tubes.

    Science.gov (United States)

    Jeon, Soon-Hyeok; Son, Yeong-Ho; Choi, Won-Ik; Song, Geun Dong; Hur, Do Haeng

    2018-01-02

    In nuclear power plants, the main corrosion product that is deposited on the outside of steam generator tubes is porous magnetite. The objective of this study was to simulate porous magnetite that is deposited on thermally treated (TT) Alloy 690 steam generator tubes. A magnetite layer was electrodeposited on an Alloy 690TT substrate in an Fe(III)-triethanolamine solution. After electrodeposition, the dense magnetite layer was immersed to simulate porous magnetite deposits in alkaline solution for 50 days at room temperature. The dense morphology of the magnetite layer was changed to a porous structure by reductive dissolution reaction. The simulated porous magnetite layer was compared with flakes of steam generator tubes, which were collected from the secondary water system of a real nuclear power plant during sludge lancing. Possible nuclear research applications using simulated porous magnetite specimens are also proposed.

  17. Eddy current proximity measurement of perpendicular tubes from within pressure tubes in CANDU nuclear reactors

    Science.gov (United States)

    Bennett, P. F. D.; Underhill, P. R.; Morelli, J.; Krause, T. W.

    2018-04-01

    Fuel channels in CANDU® (CANada Deuterium Uranium) nuclear reactors consist of two non-concentric tubes; an inner pressure tube (PT) and a larger diameter calandria tube (CT). Up to 400 horizontally mounted fuel channels are contained within a calandria vessel, which also holds the heavy water moderator. Certain fuel channels pass perpendicularly over horizontally oriented tubes (nozzles) that are part of the reactor's liquid injection shutdown system (LISS). Due to sag, these fuel channels are at risk of coming into contact with the LISS nozzles. In the event of contact between the LISS nozzle and CT, flow-induced vibrations from within the moderator could lead to fretting and deformation of the CT. LISS nozzle proximity to CTs is currently measured optically from within the calandria vessel, but from outside the fuel channels. Measurement by an independent means would provide confidence in optical results and supplement cases where optical observations are not possible. Separation of PT and CT, known as gap, is monitored from within the PT using a transmit-receive eddy current probe. Investigation of the eddy current based gap probe as a tool to also measure proximity of LISS nozzles was carried out experimentally in this work. Eddy current response as a function of LISS-PT proximity was recorded. When PT-CT gap, PT wall thickness, PT resistivity and probe lift-off variations were not present this dependence could be used to determine the LISS-PT proximity. This method has the potential to provide LISS-CT proximity using existing gap measurement data. Obtaining LISS nozzle proximity at multiple inspection intervals could be used to provide an estimate of the time to LISS-CT contact, and thereby provide a means of optimizing maintenance schedules.

  18. Laser interferometer system for the measurement of creep in pressurized tubes

    International Nuclear Information System (INIS)

    Kirchner, T.L.

    1976-07-01

    A laser interferometer measurement system was developed to measure the length, diameter, and radius of various pressurized tube specimens. The machine measures and records profilometric data of the pressurized tubes prior to insertion in the reactor and then again after a predetermined fluence has been reached to determine the amount of creep which has occurred. This data provides a statistical basis for the description of steady-state in-reactor creep and creep rupture behavior of the reference fuel cladding and structural materials for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR). In addition, this data will be used to determine the relative in-reactor creep and creep rupture behavior of candidate alloys for advanced cladding and structural materials. The laser interferometer system, referred to as the Biaxial Creep Measurement Machine (BCMM), was built to meet or exceed design criteria such as: automatic measurement of the five biaxial creep specimens varying in size; complete automation of the machine using a mini-computer; complete specimen loading, unloading, and data processing in less than five minutes; storage of data on magnetic cassette tapes; quick-look data readout and error checking during each run to determine proper machine operation; and remote operation in a radioactive environment

  19. Fracture toughness behaviour using small CCT specimen of Zr-2.5Nb pressure tube materials

    International Nuclear Information System (INIS)

    Oh, Dong Joon; Kim, Young Suk; Ahn, Sang Bok; Im, Kyung Soo; Kwon, Sang Chul; Cheong, Yong Mu

    2001-03-01

    Fracture toughness of Zr-2.5Nb pressure tube is the essential data to estimate the CCL(critical crack length) for the concept of LBB(Leak-Before-Break) in PHWR. Zr-2.5Nb pressure tubes could be degraded due to the absorption of hydrogen from coolant and the irradiation. To investigate the fracture toughness behaviour such as J-resistance curves, dJ/da, and CCL of some Zr-alloys (CANDU-double, -quad, CW-E125, TMT-E125, E-635), the transverse tensile test and the fracture toughness test of small CCT (Curved Compact Tension) specimen with 17 mm width were carried out with the variation of testing temperature at different testing condition. To define the fracture mechanism of degradation, the fractographic comparison of fracture surface was performed using the stereoscope and SEM. In addition, the effect of non-uniformed pre-fatigue crack was also studied. In conclusion, CANDU double-melted was less tougher than CANDU quad-melted and the hydrogen embrittlement was found at room temperature. Finally, while the effect of non-uniformed pre-fatigue crack was considerable at room temperature, this effect was disappeared at 250-300 .deg. C

  20. Suppressing hydrogen ingress during aqueous corrosion of CANDU Zr-2.5 Nb pressure tube material

    International Nuclear Information System (INIS)

    Elmoselhi, M.B.; Donner, A.; Brennenstuhl, A.; Warr, B.D.; Ellis, P.J.; Evans, D.W.

    2002-01-01

    As a result of their special properties, including low neutron cross-section and intrinsic corrosion resistance, Zr alloys are used in the fabrication of nuclear core components, particularly fuel cladding (in most reactor types) and also Zr-2.5 Nb pressure tubes in CANDU trademark (Canada Deuterium Uranium) reactors. Corrosion and H uptake during service can limit the life of these components. Therefore, remedial action may be appropriate to slow the H uptake rate and prolong the working life of these reactor components. This work has explored the possibility of reducing H uptake in pressure tube material by incorporating an inhibiting agent into the corrosion environment. Two approaches have been tested, depositing a thin metallic film on the initial oxide surface and adding an inhibiting agent to the solution. The latter approach appears more practical. Screening experiments were conducted in short-term (∝30 day) exposures in high temperature (340 C) aqueous out-reactor environments, simulating the CANDU trademark heat transport coolant with various chemistries. Compounds tested included aluminum acetate, aluminum nitrate, lithium nitrate, rhodium nitrate and yttrium nitrate. Comparison of results from the aluminum nitrate additives and aluminum acetate additives suggests that the nitrate anion is the effective ingredient for H ingress inhibition. The nitrate anion appears to reduce the rate of H ingress regardless of the associated cation. However, each cation appears to affect the rate of corrosion differently. These cations were found to be incorporated in the oxide film. (authors)

  1. Wear behavior of 2-1/4 Cr-1Mo tubing against alloy 718 tube-support material in sodium-cooled steam generators

    International Nuclear Information System (INIS)

    Wilson, W.L.

    1983-05-01

    A series of prototypic steam generator 2-1/4 Cr-1 Mo tube/alloy 718 tube support plate wear tests were conducted in direct support of the Westinghouse Nuclear Components Division -- Breeder Reactor Components Project Large Scale steam Generator design. The initial objective was to verify the acceptable wear behavior of softer, ''over-aged'' alloy 718 support plate material. For all interfaces under all test conditions, resultant wear damage was adhesive in nature with varying amounts of 2-1/4 Cr-1 Mo tube material being adhesively transferred to the alloy 718 tube supports. Maximum tube wear depths exceeded the initially established design allowable limit of 127 μm (.005 in.) at 17 of the 18 interfaces tested. A decrease in contact stresses produced acceptable tube wear depths below a readjusted maximum design allowable value of 381 μm (.015 in.). Additional conservatisms associated with the simulation of a 40-year lifetime of rubbing in a one-week laboratory test provided further confidence that the 381 μm maximum tube wear allowance would not be exceeded in service. Softer, ''over-aged'' alloy 718 material was found to produce slightly less wear damage on 2-1/4 Cr-1 Mo tubing than fully age hardened material. Also, air formed oxide films on the alloy 718 reduced initial tube wear and delayed the onset of adhesive surface damage. However, at high surface stress levels, these films were not sufficiently stable to provide adequate long term protection from adhesive wear. The results of the present work and those of previous test programs suggest that the successful in-sodium tribological performance of 2-1/4 Cr-1 Mo/alloy 718 rubbing couples is dependent upon the presence of lubricative surface films, such as oxides and/or surface reaction or deposition products. 11 refs., 13 figs., 4 tabs

  2. Examination of steam generator alloy 800 NG tube from the Almaraz unit 2 NPP

    International Nuclear Information System (INIS)

    Diego, G. de; Gomez Briceno, D.; Maffiotte, C.; Baladia, M.; Arias, C.J.

    2015-01-01

    The steam generators of Almaraz Unit 2 were replaced in 1997 by the model 61W/D3 (Siemens) with Alloy 800NG steam generator tubes. Denting indications were firstly detected in 2006 in the SG-3. Crack indications were identified in 2009. At the end of 2011, three tubes were recovered from this steam generator to carry out destructive examination in order to identify the root cause of the tubes degradation. Analysis of deposits point out the existence of multiples elements in the removed OD (Outer Diameter) deposits as well as in the deposits at the free tube under sludge and at the transition zone. Deposits are more abundant at the transition zone than at free tube. About 10% Na concentration has been detected, whereas S and Cl appear in small concentrations. Si appears regularly and Cr, Ni concentrations in the deposits are similar. Multiple intergranular cracks have been detected at 3 mm above the last contact point between the tube and the TS (tube support), in a band of around 5 mm, practically in the whole perimeter of the tube. Fracture surface of crack-B was partially covered by a Si rich layer, whereas fracture surface of crack-A seems to be cleaner. However, no significant differences in composition, except higher amount of S in crack-B, were found in the deposits of both cracks. EDX mapping and Auger profiles point out Ni enrichment with slight Cr enrichment or depletion and Fe depletion. The comparison of Auger profiles with available results for Alloy 800 tested in caustic and acid sulfate environments seems to indicate that the environment inside the cracks detected in the tube R67C48 is neutral or moderately caustic

  3. Creep and stress rupture behaviour of zircaloy-2 and Zr-2.5% Nb alloy tubes at 573 K

    International Nuclear Information System (INIS)

    Laha, K.; Bhanu Sankara Rao, K.; Chandravathi, K.S.; Mannan, S.L.

    1992-01-01

    Zirconium alloys are extensively used for coolant tubes of pressurised heavy water reactors. The choice of these materials is based on their good corrosion resistance in water, low capture cross section for thermal neutrons and good mechanical properties. In this paper the results of an investigation performed on the creep and rupture behaviour of indigenously produced zircaloy-2 and Zr-2.5% Nb alloy are presented. Samples for creep testing were cut longitudinally from finished pressure tubes. Creep rupture tests were carried out in air under constant load conditions at 300 C employing five stress levels in the range 300-360 MPa. Zr-2.5% Nb alloy displayed higher rupture lives at all stress levels compared to zircaloy-2. Steady state creep rate of Zr-2.5%Nb was lower than that zircaloy-2 at identical stress levels. In the stress range of the experiments, the dependence of the steady state creep rate (ε s ) on applied stress (σ) for both the alloys could be represented by a power law, ε s =A σ n The stress sensitivity (n) for Zr-2.5% Nb was lower than that of zircaloy-2. For both the alloys the time to creep rupture t r was found related to the steady state creep rate through the modified Monkman-Grant relation (ε s ) α . t r = constant. Similar value of α was obtained for both the materials. Zr-2.5%Nb exhibited higher ductility (% elongation to rupture) compared to zircaloy-2 at stress levels ≥ 320 MPa. At lower stresses significant difference in ductility was not noticed. Percentage reduction in area was lower in Zr-2.5%Nb at all stress levels indicating better resistance for necking. The time for onset of tertiary was longer for Zr-2.5% Nb alloy. The proportion of life spent by Zr-2.5% Nb in steady state creep regime was higher compared to that of zircaloy-2. Metallographic investigations on longitudinal sections in both the alloys showed large number of intragranular pores close to the fracture surface. A few number of cracks which are characteristic of

  4. Remote ultrasonic characterisation of an irradiated pressure tube from RAPS-II

    Energy Technology Data Exchange (ETDEWEB)

    Gangotra, S; Muralidhar, S; Raut, S D; Ouseph, P M; Ghosh, J K; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

    1994-12-31

    The Rajasthan Atomic Power Station Unit-2 (RAPS-2) has reached a stage of operation where the contacting pressure tubes are suspect to failure as a result of irradiation creep and displacement of the garter springs, the hot pressure tube coming in contact with the cold calandria tube. To study and assess the safety of these pressure tubes, two channels believed to be in contact with the calandria tubes, have been removed from the reactor for detailed full length post irradiation examination. Some of the test results are presented. 2 refs., 3 figs., 1 tab.

  5. The suppression of dissolution for alloy 690 in high temperature and high pressure water with chromium ion implantation

    International Nuclear Information System (INIS)

    Shibata, Toshio; Fujimoto, Shinji; Ohtani, Saburou; Watanabe, Masanori; Hirao, Kyozo; Okumoto, Masaru; Shibaike, Hiroyuki.

    1994-01-01

    As the material of heat exchanger tubes for PWRs, the nickel alloys such as alloy 690 and alloy 600 have been used, but 58 Ni and 60 Co contained as an impurity elute in primary cooling water, and are radioactivated, in this way, they become the cause of radiation exposure. By increasing chromium concentration, the corrosion resistance of nickel alloys is improved, and for modern heat exchangers, the alloy 690, of which the chromium content is increased up to 30%, has been adopted, and excellent results have been obtained. In this research, aiming at the further reduction of radiation exposure, by increasing the chromium concentration in surface layer using ion implantation technology, the change of the corrosion behavior of alloy 690 in high temperature, high pressure water was investigated. The chemical composition of the alloy 690 used, and the making of plate specimens are shown. The polarization behavior of alloy 690 in 0.1 mol/l sulfuric acid deaerated at normal temperature is reported, and the effect of suppressing dissolution was remarkable in the specimens with much implantation. The electrochemical behavior of alloy 690 in simulated cooling water was investigated. Immobile case has high chromium content and is thin. (K.I.)

  6. CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2010-01-01

    The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

  7. Impact Fretting Wear Behavior of Alloy 690 Tubes in Dry and Deionized Water Conditions

    Institute of Scientific and Technical Information of China (English)

    Zhen-Bing Cai; Jin-Fang Peng; Hao Qian; Li-Chen Tang; Min-Hao Zhu

    2017-01-01

    The impact fretting wear has largely occurred at nuclear power device induced by the flow-induced vibration,and it will take potential hazards to the service of the equipment.However,the present study focuses on the tangential fretting wear of alloy 690 tubes.Research on impact fretting wear of alloy 690 tubes is limited and the related research is imminent.Therefore,impact fretting wear behavior of alloy 690 tubes against 304 stainless steels is investigated.Deionized water is used to simulate the flow environment of the equipment,and the dry environment is used for comparison.Varied analytical techniques are employed to characterize the wear and tribochemical behavior during impact fretting wear.Characterization results indicate that cracks occur at high impact load in both water and dry equipment;however,the water as a medium can significantly delay the cracking time.The crack propagation behavior shows a jagged shape in the water,but crack extended disorderly in dry equipment because the water changed the stress distribution and retarded the friction heat during the wear process.The SEM and XPS analysis shows that the main failure mechanisms of the tube under impact fretting are fatigue wear and friction oxidation.The effect of medium(water) on fretting wear is revealed,which plays a potential and promising role in the service of nuclear power device and other flow equipments.

  8. Applicability of Alignment and Combination Rules to Burst Pressure Prediction of Multiple-flawed Steam Generator Tube

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myeong Woo; Kim, Ji Seok; Kim, Yun Jae [Korea University, Seoul (Korea, Republic of); Jeon, Jun Young [Doosan Heavy Industries and Consruction, Seoul (Korea, Republic of); Lee, Dong Min [Korea Plant Service and Engineering, Technical Research and Development Institute, Naju (Korea, Republic of)

    2016-05-15

    Alignment and combination rules are provided by various codes and standards. These rules are used to determine whether multiple flaws should be treated as non-aligned or as coplanar, and independent or combined flaws. Experimental results on steam generator (SG) tube specimens containing multiple axial part-through-wall (PTW) flaws at room temperature (RT) are compared with assessment results based on the alignment and combination rules of the codes and standards. In case of axial collinear flaws, ASME, JSME, and BS7910 treated multiple flaws as independent flaws and API 579, A16, and FKM treated multiple flaws as combined single flaw. Assessment results of combined flaws were conservative. In case of axial non-aligned flaws, almost flaws were aligned and assessment results well correlate with experimental data. In case of axial parallel flaws, both effective flaw lengths of aligned flaws and separated flaws was are same because of each flaw length were same. This study investigates the applicability of alignment and combination rules for multiple flaws on the failure behavior of Alloy 690TT steam generator (SG) tubes that widely used in the nuclear power plan. Experimental data of burst tests on Alloy 690TT tubes with single and multiple flaws that conducted at room temperature (RT) by Kim el al. compared with the alignment rules of these codes and standards. Burst pressure of SG tubes with flaws are predicted using limit load solutions that provide by EPRI Handbook.

  9. Assessment of a Pressure Tube Rupture with a Poisoned Moderator

    International Nuclear Information System (INIS)

    Kim, S. R.; Kim, B. G.; Kim, S. C.; Kim, E. K.

    2005-01-01

    The postulated in-core LOCA has been analyzed and evaluated while the reactor is operating normally with a low moderator poison concentration for CANDU. However, when the reactor is operating with a relatively large amount of boron and/or gadolinium poison in the moderator, an assessment of the fuel integrity was required for the pressure tube rupture (PTR) accident. Poisoned moderator exists mainly during a startup after a prolonged shutdown lasting for more than one day. For the case of a reactor regulating system (RRS) working, the methodology of the PTR assessment with a poisoned moderator has been developed to determine the effective trip parameters, evaluate the fuel integrity, and establish the standard reactor start-up model for the Wolsong Nuclear Power Plants recently. The developed methodology and results are presented

  10. Polycrystalline models for the calculation of residual stresses in zirconium alloys tubes

    International Nuclear Information System (INIS)

    Signorelli, J.W.; Turner, P.A.; Lebensohn, R.A.; Pochettino, A.A.

    1995-01-01

    Tubes made of different Zirconium alloys are used in various types of reactors. The final texture of tubes as well as the distribution of residual stresses depend on the mechanical treatments done during their manufacturing process. The knowledge and prediction of both the final texture and the distribution of residual stresses in a tube for nuclear applications are of outstanding importance in relation with in-reactor performance of the tube, especially in what concerns to its irradiation creep and growth behaviour. The viscoplastic and the elastoplastic self consistent polycrystal models are used to investigate the influence of different mechanical treatments, performed during rolling processes on the final distribution of intergranular residual stresses of zirconium alloys tubes. The residual strains predictions with both formulations show a non linear dependence with the orientation, but they are qualitatively different. This discrepancy could be explain in terms of the relative plastic activity between the -type and -type deformation modes predicted with the viscoplastic and elastoplastic models. (author). 10 refs., 4 figs., 1 tab

  11. Modification of structural phase state in superficial layers of fuel tubes made of Zirconium alloys

    International Nuclear Information System (INIS)

    Volkov, N.; Kalin, B.; Pimenov, Y.; Timoshin, S.

    2011-01-01

    The paper presents the results obtained in developing the method for introduction of the required changes into states and properties of outer surface on fuel rod cladding made of zirconium alloys E110 and E635 through irradiation by radial Ar + ion beam with a broad energy spectrum. In particular, the paper demonstrates that ion beam treatment of the claddings surface, at the final stage of their fabrication, can upgrade substantially quality of outer tubular surface after mechanical polishing (the cleaner surface, the lower roughness, removal of technological transversal scratches). In addition, the ion beam irradiation results in higher micro-hardness of the modified layer and in better tribological parameters. Kinetic effects in growth of oxide films were studied for the tubular samples of zirconium alloys after ion-beam treatment (cleaning and polishing by radial Ar + ion beam). Also, corrosion tests of the tubular samples were carried out in water (at 350 0 C) and steam (at 350, 375 and 400 0 C) with duration up to 3000 hours. It was revealed that oxide layer consisting mainly of zirconium dioxide in monoclinic modification was formed on tubular surface after oxidation at 3500 0 C in water or steam. The oxidizing process in the pressurized steam created thicker oxide layer on tubular surface than that in the pressurized water. Experimental data were used to determine optimal conditions for ion-beam treatment of outer fuel tube surface. The tubular samples with the following geometrical parameters were investigated: length - up to 500 mm, diameter - 9,15 mm. Optimal regimes for ion-beam cleaning and polishing of the tubular samples were studied up to the process rate of 1 meter per minute. Within the frames of linear approximation, analytical relationships were derived for time dependent growth of oxide films and used to evaluate thickness of oxide film under test conditions (duration . up to 10000 hours). Thickness of oxide films can cover the range from 6

  12. Structural stability of high entropy alloys under pressure and temperature

    DEFF Research Database (Denmark)

    Ahmad, Azkar S.; Su, Y.; Liu, S. Y.

    2017-01-01

    The stability of high-entropy alloys (HEAs) is a key issue before their selection for industrial applications. In this study, in-situ high-pressure and high-temperature synchrotron radiation X-ray diffraction experiments have been performed on three typical HEAs Ni20Co20Fe20Mn20Cr20, Hf25Nb25Zr25Ti...

  13. Probabilistic fracture mechanics applied for DHC assessment in the cool-down transients for CANDU pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Radu, Vasile, E-mail: vasile.radu@nuclear.ro [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania); Roth, Maria [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania)

    2012-12-15

    For CANDU pressure tubes made from Zr-2.5%Nb alloy, the mechanism called delayed hydride cracking (DHC) is widely recognized as main mechanism responsible for crack initiation and propagation in the pipe wall. Generation of some blunt flaws at the inner pressure tube surface during refueling by fuel bundle bearing pad or by debris fretting, combined with hydrogen/deuterium up-take (20-40 ppm) from normal corrosion process with coolant, may lead to crack initiation and growth. The process is governed by hydrogen hysteresis of terminal solid solubility limits in Zirconium and the diffusion of hydrogen atoms in the stress gradient near to a stress spot (flaw). Creep and irradiation growth under normal operating conditions promote the specific mechanisms for Zirconium alloys, which result in circumferential expansion, accompanied by wall thinning and length increasing. These complicate damage mechanisms in the case of CANDU pressure tubes that are also are affected by irradiation environment in the reactor core. The structural integrity assessment of CANDU fuel channels is based on the technical requirements and methodology stated in the Canadian Standard N285.8. Usually it works with fracture mechanics principles in a deterministic manner. However, there are inherent uncertainties from the in-service inspection, which are associated with those from material properties determination; therefore a necessary conservatism in deterministic evaluation should be used. Probabilistic approach, based on fracture mechanics principle and appropriate limit state functions defined as fracture criteria, appears as a promising complementary way to evaluate structural integrity of CANDU pressure tubes. To perform this, one has to account for the uncertainties that are associated with the main parameters for pressure tube assessment, such as: flaws distribution and sizing, initial hydrogen concentration, fracture toughness, DHC rate and dimensional changes induced by long term

  14. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Determination of hydrogen concentration and blister characterization

    International Nuclear Information System (INIS)

    2009-03-01

    Heavy water reactors (HWRs) comprise significant numbers of today's operating nuclear power plants, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes, are an important factor in ensuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Intercomparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the framework of the IAEA's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of IAEA's project on advanced technologies for HWRs. The objective of the CRP was to compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP participants investigated the capability of different techniques to detect and characterize flaws. During the second phase participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in zirconium alloys. The intention was to identify the most effective pressure tube inspection and diagnostic methods and to identify further development needs. The organizations which participated in phase 2 of this CRP are: - Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL), Chalk River Laboratories (CRL), Canada; - Bhabha Atomic Research Centre (BARC), India; - Korea Atomic Energy Research Institute (KAERI), Republic of Korea; - National Institute for Research and Development for Technical Physics (NIRDTP), Romania; - Nuclear Non-Destructive Testing Research and Services (NNDT), Romania. IAEA-TECDOC-1499

  15. Endotracheal tube cuff pressure monitoring during neurosurgery - Manual vs. automatic method

    Directory of Open Access Journals (Sweden)

    Mukul Kumar Jain

    2011-01-01

    Full Text Available Background: Inflation and assessment of the endotracheal tube cuff pressure is often not appreciated as a critical aspect of endotracheal intubation. Appropriate endotracheal tube cuff pressure, endotracheal intubation seals the airway to prevent aspiration and provides for positive-pressure ventilation without air leak. Materials and Methods: Correlations between manual methods of assessing the pressure by an experienced anesthesiologists and assessment with maintenance of the pressure within the normal range by the automated pressure controller device were studied in 100 patients divided into two groups. In Group M, endotracheal tube cuff was inflated manually by a trained anesthesiologist and checked for its pressure hourly by cuff pressure monitor till the end of surgery. In Group C, endotracheal tube cuff was inflated by automated cuff pressure controller and pressure was maintained at 25-cm H 2 O throughout the surgeries. Repeated measure ANOVA was applied. Results: Repeated measure ANOVA results showed that average of endotracheal tube cuff pressure of 50 patients taken at seven different points is significantly different (F-value: 171.102, P-value: 0.000. Bonferroni correction test shows that average of endotracheal tube cuff pressure in all six groups are significantly different from constant group (P = 0.000. No case of laryngomalacia, tracheomalacia, tracheal stenosis, tracheoesophageal fistula or aspiration pneumonitis was observed. Conclusions: Endotracheal tube cuff pressure was significantly high when endotracheal tube cuff was inflated manually. The known complications of high endotracheal tube cuff pressure can be avoided if the cuff pressure controller device is used and manual methods cannot be relied upon for keeping the pressure within the recommended levels.

  16. Evaluation of High Temperature Corrosion Resistance of Finned Tubes Made of Austenitic Steel And Nickel Alloys

    Directory of Open Access Journals (Sweden)

    Turowska A.

    2016-06-01

    Full Text Available The purpose of the paper was to evaluate the resistance to high temperature corrosion of laser welded joints of finned tubes made of austenitic steel (304,304H and nickel alloys (Inconel 600, Inconel 625. The scope of the paper covered the performance of corrosion resistance tests in the atmosphere of simulated exhaust gases of the following chemical composition: 0.2% HCl, 0.08% SO2, 9.0% O2 and N2 in the temperature of 800°C for 1000 hours. One found out that both tubes made of austenitic steel and those made of nickel alloy displayed good resistance to corrosion and could be applied in the energy industry.

  17. Delayed Hydride Cracking Mechanism in Zirconium Alloys and Technical Requirements for In-Service Evaluation of Zr-2.5Nb Tubes with Flaws

    International Nuclear Information System (INIS)

    Kim, Young Suk

    2007-01-01

    In association with periodic inspection of CANDU nuclear power plant components, Canadian Standards Association issued CSA N285.8 in 2005 as technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors. This first version, CSA N285.8 involves procedures for, firstly, the evaluation of pressure tube flaws, secondly, the evaluation of pressure tube to calandria tube contact and, thirdly, the assessment of a reactor core, and material properties and derived quantities. The evaluation of pressure tube flaws includes delayed hydride cracking evaluation the procedures of which are stipulated based on the existing delayed hydride cracking models. For example, the evaluation of flaw-tip hydride precipitation during reactor cooldown involves a procedure to calculate the equilibrium hydrogen equivalent concentration in solution at the flaw tip, Htipas follows: Htip=Hfexp[- (VH delta no.)/RT], where Hf is the total bulk hydrogen equivalent concentration, VH partial molar volume of hydrogen in zirconium, δ a difference in hydrostatic stress between the bulk and the crack tip. When Htip ≥TSSP at temperature, then flaw-tip hydride is predicted to precipitate. Eq. (1) suggests that hydrogen concentration at the crack tip would increase due to an work energy given by the difference in the hydrostatic stress

  18. Database and prediction model for CANDU pressure tube diameter

    Energy Technology Data Exchange (ETDEWEB)

    Jung, J.Y.; Park, J.H. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

    2014-07-01

    The pressure tube (PT) diameter is basic data in evaluating the CCP (critical channel power) of a CANDU reactor. Since the CCP affects the operational margin directly, an accurate prediction of the PT diameter is important to assess the operational margin. However, the PT diameter increases by creep owing to the effects of irradiation by neutron flux, stress, and reactor operating temperatures during the plant service period. Thus, it has been necessary to collect the measured data of the PT diameter and establish a database (DB) and develop a prediction model of PT diameter. Accordingly, in this study, a DB for the measured PT diameter data was established and a neural network (NN) based diameter prediction model was developed. The established DB included not only the measured diameter data but also operating conditions such as the temperature, pressure, flux, and effective full power date. The currently developed NN based diameter prediction model considers only extrinsic variables such as the operating conditions, and will be enhanced to consider the effect of intrinsic variables such as the micro-structure of the PT material. (author)

  19. Experience in quality assurance of alloy D9 clad tubes for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Kapoor, K.; Prahlad, B.

    2012-01-01

    Stainless Steel Alloy D9 is the material for cladding in various sub-assemblies of Prototype Fast Breeder Reactor (PFBR). The fabrication, inspection, testing and supply of the clad tubes for the first core of PFBR is nearly completed. The paper also compares the specification requirements and the achieved results for some of the critical aspects which is arrived after completing supply against the first core requirement

  20. Compressed-tube pressure cell for optical studies at ocean pressures: Application to glucose mutarotation kinetics.

    Science.gov (United States)

    Lamelas, F J

    2016-12-01

    A self-contained compressed-tube pressure cell is tested to 25 MPa. The cell is very simple to construct and offers stable pressure control with optical access to fluid samples. The physical path length of light through the cell is large enough to measure optical activity. The entire system is relatively small and portable, and it is vibration-free, since a compressor is not used. Operation of the cell is demonstrated by measuring the mutarotation rate of aqueous glucose solutions at 25 °C. A logarithmic plot of the rate constant vs. pressure yields an activation volume for mutarotation of -22 cm 3 /mol, approximately twice the value measured previously at higher pressures.

  1. Temperature and pressure dependent osmotic pressure in liquid sodium-cesium alloys

    International Nuclear Information System (INIS)

    Rashid, R.I.M.A.

    1987-01-01

    The evaluation of the osmotic pressure in terms of the concentration fluctuations of mixtures and the equations of state of the pure liquids is considered. The temperature and pressure dependent experimentally measured concentration-concentration correlations in the long wavelength limit of liquid sodium-cesium alloys are used to demonstrate the appreciable dependence of the temperature and pressure on the osmotic pressure as a function of concentration. Introducing interchange energies as functions of temperature and pressure, our analysis is consistent with the Flory model. Thus, a formalism for evaluating the state dependent osmotic pressure is developed and our numerical work is considered to be an extension of the calculations of Rashid and March in the sense that a temperature and pressure dependent interchange energy parameter that more closely parameterizes the state dependent concentration fluctuations in the liquid alloys, is used. (author)

  2. A Study on Corrosion and Fretting Wear Resistance of Alloy 690 Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Won, Ju Jin; Min, Su Jung; Kim, Myeong Su; Kim, Kyu Tae [Dongguk Univ., Gyeongju (Korea, Republic of)

    2013-10-15

    In this article, the effects of such failures have on the materials of alloy 690 are assessed. The corroded volume variation and mass decreased continuously with time. However, the oxide volume changes in an irregular pattern since the oxide formed on the alloy 690 metal may be detached due to the flake formation. The amount of the fretting wear increased with time. It can be seen that the wear rate increased with time and reduced at the later time. The test results show that the ductility decreased as corrosion increases. Alloy 690 is broadly used as a material of nuclear power plant's steam generator tubes because of its excellent mechanical strength, corrosion properties, wear properties and stability at a high temperature. However, the tubes for nuclear power plant's steam generators become a major threat for lifetime management and efficient operation of nuclear power plant due to various corrosion and fretting wear failures caused by flow-induced vibration (FIV) that occurs between tubes.

  3. Thermal creep properties of alloy D9 stainless steel and 316 stainless steel fuel clad tubes

    International Nuclear Information System (INIS)

    Latha, S.; Mathew, M.D.; Parameswaran, P.; Bhanu Sankara Rao, K.; Mannan, S.L.

    2008-01-01

    Uniaxial thermal creep rupture properties of 20% cold worked alloy D9 stainless steel (alloy D9 SS) fuel clad tubes for fast breeder reactors have been evaluated at 973 K in the stress range 125-250 MPa. The rupture lives were in the range 90-8100 h. The results are compared with the properties of 20% cold worked type 316 stainless steel (316 SS) clad tubes. Alloy D9 SS were found to have higher creep rupture strengths, lower creep rates and lower rupture ductility than 316 SS. The deformation and damage processes were related through Monkman Grant relationship and modified Monkman Grant relationship. The creep damage tolerance parameter indicates that creep fracture takes place by intergranular cavitation. Precipitation of titanium carbides in the matrix and chromium carbides on the grain boundaries, dislocation substructure and twins were observed in transmission electron microscopic investigations of alloy D9 SS. The improvement in strength is attributed to the precipitation of fine titanium carbides in the matrix which prevents the recovery and recrystallisation of the cold worked microstructure

  4. Characterization of Tubing from Advanced ODS alloy (FCRD-NFA1)

    Energy Technology Data Exchange (ETDEWEB)

    Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lavender, Curt [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anderson, Iver [Ames Lab., Ames, IA (United States); Rieken, Joel [Ames Lab., Ames, IA (United States); Lewandowski, John [Case Western Reserve Univ., Cleveland, OH (United States); Hoelzer, Dave [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Odette, George R. [Univ. of California, Santa Barbara, CA (United States)

    2016-09-20

    Fabrication methods are being developed and tested for producing fuel clad tubing of the advanced ODS 14YWT and FCRD-NFA1 ferritic alloys. Three fabrication methods were based on plastically deforming a machined thick-wall tube sample of the ODS alloys by pilgering, hydrostatic extrusion or drawing to decrease the outer diameter and wall thickness and increase the length of the final tube. The fourth fabrication method consisted of the additive manufacturing approach involving solid-state spray deposition (SSSD) of ball milled and annealed powder of 14YWT for producing thin-wall tubes. Of the four fabrication methods, two methods were successful at producing tubing for further characterization: production of tubing by high-velocity oxy-fuel spray forming and production of tubing using high-temperature hydrostatic extrusion. The characterization described shows through neutron diffraction the texture produced during extrusion while maintaining the beneficial oxide dispersion. In this research, the parameters for innovative thermal spray deposition and hot extrusion processing methods have been developed to produce the final nanostructured ferritic alloy (NFA) tubes having approximately 0.5 mm wall thickness. Effect of different processing routes on texture and grain boundary characteristics has been investigated. It was found that hydrostatic extrusion results in combination of plane strain and shear deformations which generate rolling textures of α- and γ-fibers on {001}<110> and {111}<110> together with a shear texture of ζ-fiber on {011}<211> and {011}<011>. On the other hand, multi-step plane strain deformation in cross directions leads to a strong rolling textures of θ- and ε-fiber on {001}<110> together with weak γ-fiber on {111}<112>. Even though the amount of the equivalent strain is similar, shear deformation leads to much lower texture indexes compared to the plane strain deformations. Moreover, while 50% of hot rolling brings about a large number of

  5. Neutron flux measurements in C-9 capsule pressure tube

    International Nuclear Information System (INIS)

    Barbos, D.; Roth, C. S.; Gugiu, D.; Preda, M.

    2001-01-01

    C-9 capsule is a fuel testing facility in which the testing consists of a daily cycle ranging between the limits 100% power to 50% power. C-9 in-pile section with sample holder an instrumentation are introduced in G-9 and G-10 experimental channels. The experimental fuel channel has a maximum value when the in-pile section (pressure tube) is in G-9 channel and minimum value in G-10 channel. In this paper the main goals are determination or measurements of: - axial thermal neutron flux distribution in C-9 pressure tube both in G-9 and G-10 channel; - ratio of maximum neutron flux value in G-9 and the same value in G-9 channel and the same value in G-10 channel; - neutron flux-spectrum. On the basis of axial neutron flux distribution measurements, the experimental fuel element in sample holder position in set. Both axial neutron flux distribution of thermal neutrons and neutron flux-spectrum were performed using multi- foil activation technique. Activation rates were obtained by absolute measurements of the induced activity using gamma spectroscopy methods. To determine the axial thermal neutron flux distribution in G-9 and G-10, Cu 100% wire was irradiated at the reactor power of 2 MW. Ratio between the two maximum values, in G-9 and G-10 channels, is 2.55. Multi-foil activation method was used for neutron flux spectrum measurements. The neutron spectra and flux were obtained from reaction rate measurements by means of SAND 2 code. To obtain gamma-ray spectra, a HPGe detector connected to a multichannel analyzer was used. The spectrometer is absolute efficiency calibrated. The foils were irradiated at 2 MW reactor power in previously determined maximum flux position resulted from wire measurements. This reaction rates were normalized for 10 MW reactor power. Neutron self shielding corrections for the activation foils were applied. The self-shielding corrections are computed using Monte Carlo simulation methods. The measured integral flux is 1.1·10 14 n/cm 2 s

  6. Methodologies for assessment of the service life of pressure tubes in Indian PHWRs

    International Nuclear Information System (INIS)

    Sinha, R.K.; Sharma, A.; Madhusoodanan, K.; Sinha, S.K.; Malshe, U.D.

    1997-01-01

    For estimating safe service life of pressure tubes in Indian PHWRs, analytical methodologies have been developed to evaluate creep deformation, deuterium pick-up rate, blister growth at cold spot, and operating domain required for achieving leak-before-break. The paper provides an overview of these methodologies, and results of some studies carried out towards evolution of proposed fitness-for-service criteria for a pressure tube in contact with its calandria tube. (author)

  7. Status and Plans for work on pressure tube creep at AECL

    International Nuclear Information System (INIS)

    Bickel, Grant A.

    2013-01-01

    AECL research goals: • Develop empirical models to: – regress out operating conditions/extrinsic factors – rank relative strain behavior of measured in-service pressure tubes; • Correlate the ranked strains to manufacturing variables and the microstructure to: – Develop mechanistic insights – Optimize manufacturing/microstructure for improved pressure tube performance

  8. Specific aspects in the manufacturing and operating of CANDU reactor pressure tubes (P/T)

    International Nuclear Information System (INIS)

    Muscaloiu, C.

    1997-01-01

    The CANDU reactor design is based on a number of individual P/T in which nuclear fuel bundles are located. P/T are required to be operated in an environment of elevated temperature (300 o C), internal pressure (10 Mpa), fast neutron flux (E>1 MeV) and heavy water. The most suitable material which can provide the desired neutron economy and still maintain its mechanical properties along with corrosion resistance is zirconium alloys Zr+ 2.5 % Nb with the following composition: niobium, 2.5 to 2.8 weight percent; oxygen, 1,000 to 1,300 ppm; zirconium + allowed impurities - balance. A total of 380 pressure tubes are installed into reactor. Each pressure tube is attached at each end to a stainless steel end fitting by means of a grooved, expanded joint. The installation works were performed by ANM Bucuresti, under the technical support of General Electric Canada. The integrity of P/T after installation was examined as follows: - the surface of the rolled area on unrolled internal surface extending 25 mm beyond rolled area was inspected for irregularities by means of a boroscope; - all pressure tubes were subjected to the helium leak test after F/C installation. During P/T operating life periodical inspections according to Canadian Standard CSA N285.4 are performed. The selection of the P/T for inspection is based either on particular properties or on the operating conditions of the fuel channel. The inspection consists in: a) Base Line Inspection within 2 years period commencing after 7,000 EFPH of operation which will include a volumetric inspection over P/T full length and measurements of P/T sag, ID, wall thickness and F/C bearing positions; b) Periodic Inspection in the same conditions plus material surveillance (on the four most significant indication P/T detected during the Base Line Inspection). The inspection will be performed on 14 selected P/T. (author)

  9. The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

    1985-07-01

    On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

  10. SCC analysis of Alloy 600 tubes from a retired steam generator

    Science.gov (United States)

    Hwang, Seong Sik; Kim, Hong Pyo

    2013-09-01

    Steam generators (SG) equipped with Alloy 600 tubes of a Korean nuclear power plants were replaced with a new one having Alloy 690 tubes in 1998 after 20 years of operation. To set up a guide line for an examination of the other SG tubes, a metallographic examination of the defected tubes was carried out. A destructive analysis on 71 tubes was addressed, and a relation among the stress corrosion crack (SCC) defect location, defect depth, and location of the sludge pile was obtained. Tubes extracted from the retired SG were transferred to a hot laboratory. Detailed nondestructive analysis examinations were taken again at the laboratory, and the tubes were then destructively examined. The types and sizes of the cracks were characterized. The location and depth of the SCC were evaluated in terms of the location and height of the sludge. Most axial cracks were in the sludge pile, whereas the circumferential ones were around the top of the tube sheet (TTS) or below the TTS. Average defect depth of the axial cracks was deeper than that of the circumferential ones. Axial cracks at tube support plate (TSP) seem to be related with corrosion/sludge in crevice like at the TTS region. Circumferential cracks at TSP seem to be caused by tube denting at the upper part of the TSP. Tubes not having clear ECT signals for quantifying an ECT data-base. Tubes having no ECT signal. Tubes with a large ECT signal. Tubes with various types and sizes of flaws (primary water stress corrosion cracking (PWSCC), outside diameter stress corrosion cracking (ODSCC), Pit). Tubes with distinct PWSCC or ODSCC. Tubes were extracted from the RSG based on the field ECT with the criteria, and transferred to a hot laboratory at the Korea Atomic Energy Research Institute (KAERI) for destructive examination. A comprehensive ECT inspection was performed again at the hot laboratory to confirm the location of the cracks obtained from a field inspection. These exact locations of the defects were marked on the

  11. Study of creep collapse of tubes subject to external pressure at elevated temperature

    International Nuclear Information System (INIS)

    Takikawa, N.

    1982-01-01

    Intermediate heat exchanger (IHX) tubes of VHTR form the boundary between the primary and secondary coolants of the reactor. The tubes are subject to external pressures at a postulated secondary coolant depressurization accident, which might lead to creep collapse. Therefore, it is necessary to ensure the integrity against creep collapse by analysis. The objective of this work is to study a simplified analytical method for predicting collapse time of a curved tube subjected to an external pressure. The study is made based on the comparison of experimental collapse time of curved and straight tubes. Creep collapse tests were conducted under an elevated temperature and an external pressure. Test results showed that curved tubes had longer collapse time than straight tubes with the same cross sectional ovality. The simplified analytical method for a curved tube is proposed in this report, which is to compute collapse time of a straight tube with the same ovality. And in this method the computed time is considered as collapse time of the curved tube. The above test results show that this simplified method gives the conservative collapse time. And it is confirmed by additional IHX tube tests that the method is applicable to creep collapse analysis of IHX tubes

  12. Modelling of oxidation and hydriding behaviour of Zircaloy-2 pressure tubes in PHWR

    International Nuclear Information System (INIS)

    Sah, D.N.; Sunil Kumar; Khan, K.B.

    2002-01-01

    A computer model named DOCTOR (Deuteriding of Coolant Tubes during Operation of Reactor) has been developed for predicting the axial profile of oxide thickness and hydrogen (Deuterium) concentration in PHWR pressure tubes. This model is applicable to single channel or full core analysis. The main source of hydrogen is considered to be oxidation of pressure tube on the i.d. surface by high temperature coolant water. Three stages of oxidation is considered namely, pre- transition, post transition and accelerated. Oxidation rate is considered to be dependent on channel power, axial power/flux distribution, coolant temperature and pre-existing oxide thickness at the location. The kinetics parameters for oxidation model are derived from the actual measurement of oxide thickness on a number of pressure tubes examined in PIE Division. The input data required for the model are: channel power, channel power factor, axial flux distribution, coolant inlet temperature, critical oxide thickness, hydrogen pick up fraction, initial hydrogen in the material and time of operation (efpy). The model calculates the oxide layer thickness on the inside surface of the pressure tube along the length. The amount of hydrogen picked up by the pressure tube is calculated from the oxide thickness using hydrogen pick up fraction determined from the PIE data. The pressure tube length is divided into a number of axial segments for calculation. The temperature and fast neutron flux assumed to be constant in a given segment. The axial temperature profile calculated from the axial power profile in the channel is used for calculating the oxidation rate at various locations in the pressure tube. The model has been validated with PIE data of hydrogen equivalent measurement on a number of irradiated Zircaloy-2 pressure tubes of various PHWRs. The performance of the model in predicting the axial profile of hydrogen in the pressure tubes has been found to be good. (author)

  13. Operating envelope to minimize probability of fractures in Zircaloy-2 pressure tubes

    International Nuclear Information System (INIS)

    Azer, N.; Wong, H.

    1994-01-01

    The failure mode of primary concern with Candu pressure tubes is fast fracture of a through-wall axial crack, resulting from delayed hydride crack growth. The application of operating envelopes is demonstrated to minimize the probability of fracture in Zircaloy-2 pressure tubes based on Zr-2.5%Nb pressure tube experience. The technical basis for the development of the operating envelopes is also summarized. The operating envelope represents an area on the pressure versus temperature diagram within which the reactor may be operated without undue concern for pressure tube fracture. The envelopes presented address both normal operating conditions and the condition where a pressure tube leak has been detected. The examples in this paper are prepared to illustrate the methodology, and are not intended to be directly applicable to the operation of any specific reactor. The application of operating envelopes to minimized the probability of fracture in 80 mm diameter Zircaloy-2 pressure tubes has been discussed. Both normal operating and leaking pressure tube conditions have been considered. 3 refs., 4 figs

  14. Conservatism in methodologies for moderator subcooling sufficiency for fuel channel integrity upon pressure tube and calandria tube contact

    Energy Technology Data Exchange (ETDEWEB)

    Sun, L., E-mail: LSun@nbpower.com [Point Lepreau Generating Station, Lepreau, NB, (Canada)

    2015-07-01

    During a postulated large LOCA event in CANDU reactors, the pressure tube may balloon to contact with its surrounding calandria tube to transfer heat to the moderator. To confirm the integrity of the fuel channel in this case, many experiments have been performed in the last three decades. Based on the extant database of the pressure tube/calandria tube (PT/CT) contact, an analytical methodology was developed by Canadian Nuclear Industry to determine the sufficiency of moderator subcooling for fuel channel integrity. At the same time a semi-empirical methodology with an idea of Equivalent Moderator Subcooling (EMS) was also developed to judge the sufficiency of the moderator. In this work, some discussions were made over the two methodologies on their conservatism and it is demonstrated that the analytical approach is over conservative comparing with the EMS methodology. By using the EMS methodology, it is demonstrated that applying glass-peened calandria tubes, the requirement to moderator subcooling can be reduced by 10{sup o}C from that for smooth calandria tubes. (author)

  15. Demonstration of a shape memory alloy torque tube-based morphing radiator

    Science.gov (United States)

    Chong, Jorge B.; Walgren, Patrick; Hartl, Darren J.

    2018-03-01

    Long-distance crewed space exploration will require advanced thermal control systems (TCS) with the ability to handle a wide range of thermal loads. The ability of a TCS to adapt to the thermal environment is described by the turndown ratio. Developing radiators with high turndown ratios is critical for improving TCS technology. This paper describes a novel morphing radiator designed to achieve a high turndown ratio by varying its own radiative view factor and effective emissivity through the use of shape memory alloys (SMAs). This radiator features two SMA torque tubes cantilevered to a rigid fixture. The working fluid is transported within the SMA tubes through an annular flow system. In a cold environment, radiator panels fixed to the free ends of the tubes are oriented vertically in a parallel-plate fashion, where the high-emissivity interior faces have restricted views to the environment and heat rejection is minimized. When the system heats up, the tubes actuate by twisting in opposing directions, bringing the panels to a horizontal position with the interior faces exposed to maximize heat rejection. When the system cools down, the tubes twist in reverse, restoring the panels to the vertical orientation where heat rejection is again minimized. This variable heat rejection system has the potential for achieving higher turndown ratios than those of current state-of-the-art systems. A benchtop prototype has been designed and tested to demonstrate actuation and to explore internal heat transfer effects. Prototype design, testing, and results are herein described.

  16. Influence of hydride orientation on fracture toughness of CWSR Zr-2.5%Nb pressure tube material between RT and 300 °C

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Rishi K. [Engineering Directorate, Nuclear Power Corporation of India Limited, Mumbai 400094 (India); Department of Mechanical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India); Sunil, Saurav; Kumawat, B.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, R.N., E-mail: rnsingh@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Tewari, Asim [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India); Kashyap, B.P. [Department of Metallurgical Engineering and Materials Science, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India)

    2017-05-15

    An experimental setup was designed, fabricated and used to form radial hydrides in Zr-2.5%Nb alloy pressure tube spool. The design of setup was based on ensuring a hoop stress in the spool greater than threshold stress for reorientation of hydrides in this alloy, which was achieved by manipulating the thermal expansion coefficient of the plunger and pressure tube material and diametral interference between them. The experimental setup was loaded on a universal testing machine (UTM) fitted with an environmental chamber and subjected to a temperature cycle for the stress reorientation treatment. The metallographic examination of the hydrogen charged spools subjected to stress re-orientation treatment using this set up revealed formation of predominantly radial hydrides. The variation of fracture toughness of material containing radial hydride with test temperature showed typical ‘S’ curve behavior with transition temperatures more than that of the material containing circumferential hydride.

  17. Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

    2010-01-01

    In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

  18. Investigation of pressure drop in capillary tube for mixed refrigerant Joule-Thomson cryocooler

    International Nuclear Information System (INIS)

    Ardhapurkar, P. M.; Sridharan, Arunkumar; Atrey, M. D.

    2014-01-01

    A capillary tube is commonly used in small capacity refrigeration and air-conditioning systems. It is also a preferred expansion device in mixed refrigerant Joule-Thomson (MR J-T) cryocoolers, since it is inexpensive and simple in configuration. However, the flow inside a capillary tube is complex, since flashing process that occurs in case of refrigeration and air-conditioning systems is metastable. A mixture of refrigerants such as nitrogen, methane, ethane, propane and iso-butane expands below its inversion temperature in the capillary tube of MR J-T cryocooler and reaches cryogenic temperature. The mass flow rate of refrigerant mixture circulating through capillary tube depends on the pressure difference across it. There are many empirical correlations which predict pressure drop across the capillary tube. However, they have not been tested for refrigerant mixtures and for operating conditions of the cryocooler. The present paper assesses the existing empirical correlations for predicting overall pressure drop across the capillary tube for the MR J-T cryocooler. The empirical correlations refer to homogeneous as well as separated flow models. Experiments are carried out to measure the overall pressure drop across the capillary tube for the cooler. Three different compositions of refrigerant mixture are used to study the pressure drop variations. The predicted overall pressure drop across the capillary tube is compared with the experimentally obtained value. The predictions obtained using homogeneous model show better match with the experimental results compared to separated flow models

  19. Tube micro-fouling, boiling and steam pressure after chemical cleaning

    International Nuclear Information System (INIS)

    Hu, M.H.

    1998-01-01

    This paper presents steam pressure trends after chemical cleaning of steam generator tubes at four plants. The paper also presents tube fouling factor that serves as an objective parameter to assess tubing boiling conditions for understanding the steam pressure trend. Available water chemistry data helps substantiate the concept of tube micro-fouling, its effect on tubing boiling, and its impact on steam pressure. All four plants experienced a first mode of decreasing steam pressure in the post-cleaning operation. After 3 to 4 months of operation, the decreasing trend stopped for three plants and then restored to a pre-cleaning value or better. The fourth plant is soil in decreasing trend after 12 months of operation. Dissolved chemicals, such as silica, titanium can precipitate on tube surface. The precipitate micro-fouling can deactivate or eliminate boiling nucleation sites. Therefore, the first phase of the post-cleaning operation suffered a decrease in steam pressure or an increase in fouling factor. It appears that micro fouling by magnetite deposit can activate or create more bubble nucleation sites. Therefore, the magnetite deposit micro-fouling results in a decrease in fouling factor, and a recovery in steam pressure. Fully understanding the boiling characteristics of the tubing at brand new, fouled and cleaned conditions requires further study of tubing surface conditions. Such study should include boiling heat transfer tests and scanning electronic microscope examination. (author)

  20. Comparative study of water chemistry and surface oxide composition on alloy 600 steam generator tubing

    International Nuclear Information System (INIS)

    Bjoernkvist, L.; Norring, K.; Nyborg, L.

    1993-01-01

    The Ringhals 3 steam generators experience secondary IGSCC on the tubes at support plate locations. Its sister unit Ringhals 4 is so far without IGSCC. Extensive work has been carried out in order to determine the local chemistry in crevices and the composition of deposits and oxide films on the tubes. Hot soaks of the SG:s at zero power has been performed and the water chemistry in occluded crevices of the SGs was predicted to be alkaline, pH 300degreesC = 10. In addition to eddy current testing, a large number of tubes have been pulled and destructively examined. These analysis include SEM/EDS characterization of TSP crevice deposits and Auger electron spectroscopy (AES) with depth profiling to reveal the composition of the tube OD oxide film. The AES analysis show an outer oxide rich in Fe 3 O 4 , mostly deposited. The actual Alloy 600 oxide is found below the magnetite and is 1-2 μm thick. The composition profile of the oxide exhibits a Cr-depletion relative to Ni in the outer part of the oxide, whereas an enrichment is found in depth. In order to correlate the water chemistry to the oxide composition profiles and deposits on pulled tubes, reference samples were prepared in an autoclave. The environments were chosen similar to the predicted Ringhals 3 and 4 crevice chemistry. Exposure both in an alkaline (pH 320degreesC∼ 9.9) and an acidic (pH 320degreesC ∼4.3) environment, containing sodium, chloride and sulphate, was studied. Some samples were also found on the Alloy 600 samples exposed to alkaline environment. Thus the prediction of alkaline chemistry was verified. The enrichment of chromium relative to nickel was shown to be potential and time dependent resulting in an increased Cr/Ni ratio at Cr-max with increasing potential and time

  1. Hydride-induced degradation of hoop ductility in textured zirconium-alloy tubes: A theoretical analysis

    International Nuclear Information System (INIS)

    Qin, W.; Szpunar, J.A.; Kozinski, J.

    2012-01-01

    Hydride-induced degradation of hoop ductility in Zr-alloy tubular components has been studied for many years because of its importance in the nuclear industry. In this paper the role of intergranular and intragranular δ-hydrides in the degradation of ductility of the textured Zr-alloy tubes is investigated. The correlation among hydride distribution, orientation and morphology in the tubes is formulated based on thermodynamic modeling, and then analyzed. The results show that the applied stress, the crystallographic texture of α-Zr matrix, the grain-boundary structure, and the morphology and size of Zr grains simultaneously govern the site preference and the orientation of hydrides. A criterion is proposed to determine the threshold stress of hydride reorientation. The hoop ductility of the hydrided Zr tubes is discussed using the concept of macroscopic fracture strain. It is shown that the intergranular hydrides may be more deleterious to ductility than the intragranular ones. This work defines a general framework for understanding the relation of the microstructure of hydride-forming materials to embrittlement.

  2. Correlation between the critical heat flux and the fractal surface roughness of zirconium alloy tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; McRae, G.A.; Coleman, C.E.; Nitheanandan, T.; Sanderson, D.B.

    1999-10-01

    In CANDU fuel channels, Zircaloy calandria tubes isolate the hot pressure tubes from the cool heavy water moderator. The heavy-water moderator provides a backup heat sink during some postulated loss-of-coolant accidents. The decay heat from the fuel is transferred to the moderator to ensure fuel channel integrity during emergencies. Moderator temperature requirements are specified to ensure that the transfer of decay heat does not exceed the critical heat flux (CHF) on the outside surface of the calandria tube. An enhanced CHF provides increases in safety margin. Pool boiling experiments indicate the CHF is enhanced with glass-peening of the outside surface of the calandria tubes. The objective of this study was to evaluate the surface characteristics of glass-peened tubes and relate these characteristics to CHF. The micro-topologies of the tube surfaces were analysed using stereo-pair micrographs obtained from scanning electron microscopy (SEM) and photogrammetry techniques. A linear relationship correlated the CHF as a function of the 'fractal' surface roughness of the tubes. (author)

  3. Effects on Vocal Fold Collision and Phonation Threshold Pressure of Resonance Tube Phonation with Tube End in Water

    Science.gov (United States)

    Enflo, Laura; Sundberg, Johan; Romedahl, Camilla; McAllister, Anita

    2013-01-01

    Purpose: Resonance tube phonation in water (RTPW) or in air is a voice therapy method successfully used for treatment of several voice pathologies. Its effect on the voice has not been thoroughly studied. This investigation analyzes the effects of RTPW on collision and phonation threshold pressures (CTP and PTP), the lowest subglottal pressure…

  4. Brazing open cell reticulated copper foam to stainless steel tubing with vacuum furnace brazed gold/indium alloy plating

    Science.gov (United States)

    Howard, Stanley R [Windsor, SC; Korinko, Paul S [Aiken, SC

    2008-05-27

    A method of fabricating a heat exchanger includes brush electroplating plated layers for a brazing alloy onto a stainless steel tube in thin layers, over a nickel strike having a 1.3 .mu.m thickness. The resultant Au-18 In composition may be applied as a first layer of indium, 1.47 .mu.m thick, and a second layer of gold, 2.54 .mu.m thick. The order of plating helps control brazing erosion. Excessive amounts of brazing material are avoided by controlling the electroplating process. The reticulated copper foam rings are interference fit to the stainless steel tube, and in contact with the plated layers. The copper foam rings, the plated layers for brazing alloy, and the stainless steel tube are heated and cooled in a vacuum furnace at controlled rates, forming a bond of the copper foam rings to the stainless steel tube that improves heat transfer between the tube and the copper foam.

  5. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1996-08-01

    In Phase 1 of this project, a variety of developmental and commercial tubing alloys and claddings was exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, NF 709, 690 clad, and 671 clad for over 10,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy are being exposed for 4,000, 12,000, and 16,000 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after approximately 4,400 hours of exposure.

  6. Failure evaluation on a high-strength alloy SA213-T91 super heater tube of a power generation

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, J.; Purbolaksono, J.; Beng, L.C.; Ahmad, A. [University of Tenaga Nas, Kajang (Malaysia). Dept. of Mechanical Engineering

    2010-07-01

    This article presents failure investigation on a high-strength alloy SA213-T91 superheater tube. This failure is the first occurrence involving the material in Kapar Power Station Malaysia. The investigation includes visual inspections, hardness measurements, and microscopic examinations. The failed super-heater tube shows a wide open rupture with thin and blunt edges. Hardness readings on all the as-received tubes are used for estimating the operating metal temperature of the super-heater tubes. Microstructures of the failed tube show numerous creep cavities consisting of individual pores and chain of pores which form micro-and macro-cracks. The findings confirmed that the super-heater tube is failed by short-term overheating. Higher temperatures of the flue gas due to the inconsistent feeding of pulverized fuels into the burner is identified to cause overheating of the failed tube.

  7. Tensile strength of Zr-2.5 Nb pressure tubes: A statistical study

    Energy Technology Data Exchange (ETDEWEB)

    Shah, Priti Kotak, E-mail: pritik@barc.gov.in [Senior Scientist, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Dubey, J.S.; Datta, D.; Shriwastaw, R.S.; Rath, B.N.; Singh, R.N. [Senior Scientist, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Anantharaman, S. [Head, Post Irradiation Examination Division, Bhabha Atomic Research Centre, Mumbai (India); Chakravartty, J.K. [Director, Materials Group, Bhabha Atomic Research Centre, Mumbai (India)

    2015-12-15

    Highlights: • Tensile properties in axial and transverse direction for a number of Indian Zr-2.5 Nb PHWR pressure tubes. • Distribution of tensile properties of double-melted and quadruple-melted pressure tubes. • Tensile properties at front-end and back-end of the quadruple melted pressure tubes at room temperature and at 300 °C. - Abstract: In order to get an idea of the statistical variation in the tensile properties of the double-melted as well as quadruple melted Zr-2.5 Nb pressure tubes (PTs) and also the variation in tensile properties between the two ends of the pressure tubes, tension tests were carried out on around 50 pressure tube off-cuts. Longitudinal and transverse tensile specimens were prepared from these off-cuts of pressure tubes of double-melted and quadruple melted types. For quadruple melted pressure tubes the specimens were tested from both front-end and back-end off-cuts. Miniature flat tensile specimens having 1.8 mm width and 1.5 mm thickness and 7.6 mm gauge length were prepared from the pressure tube off-cuts without any flattening treatment. Tension tests were carried out in a screw-driven machine at room temperature and 300 °C for both front-end and back-end off-cuts of each of 16 pressure tubes. In general the transverse specimens showed higher yield strength (YS) and ultimate tensile strength (UTS) compared to the longitudinal specimens. Transverse specimens showed less strain hardening compared to the longitudinal specimens. The axial specimens showed higher uniform (UE) and total elongation (TE) compared to the transverse specimens. Double-melted pressure tubes showed relatively higher strength and lower elongation and larger standard deviation compared to the quadruple melted pressure tubes. Mean values of tensile properties showed that back-end off-cuts were relatively stronger and less ductile compared to the front-end off-cuts.

  8. Garter spring location of pressure tube for PHWR using eddy current testing methods

    International Nuclear Information System (INIS)

    Lee, Y. S.; Yang, D. J.; Jeong, H. K.

    2001-01-01

    There are garter springs between pressure tube and calandria tube for PHWR. If the space of these garter springs become to be changed, the sagging of tube is caused and the contact between the pressure tube and calandria tube will cause the tube to be failed. AECL has applied the eddy current testing methods using send-receive type probe for this purpose, but this study apply eddy current testing methods using bobbin differential type probe to detection of garter spring location. And we did the computer simulation using VIC-3D code and compared it with experiments results for inspection 1 ∼ 11kHz. The results was that the garter spring signal was successfully detected for every frequency, and 5 kHz was best

  9. Experimental residual stress evaluation of hydraulic expansion transitions in Alloy 690 steam generator tubing

    International Nuclear Information System (INIS)

    McGregor, R.; Doherty, P.; Hornbach, D.; Abdelsalam, U.

    1995-01-01

    Nuclear Steam Generator (SG) service reliability and longevity have been seriously affected worldwide by corrosion at the tube-to-tubesheet joint expansion. Current SG designs for new facilities and replacement projects enhance corrosion resistance through the use of advanced tubing materials and improved joint design and fabrication techniques. Here, transition zones of hydraulic expansions have undergone detailed experimental evaluation to define residual stress and cold-work distribution on and below the secondary-side surface. Using X-ray diffraction techniques, with supporting finite element analysis, variations are compared in tubing metallurgical condition, tube/pitch geometry, expansion pressure, and tube-to-hole clearance. Initial measurements to characterize the unexpanded tube reveal compressive stresses associated with a thin work-hardened layer on the outer surface of the tube. The gradient of cold-work was measured as 3% to 0% within .001 inch of the surface. The levels and character of residual stresses following hydraulic expansion are primarily dependent on this work-hardened surface layer and initial stress state that is unique to each tube fabrication process. Tensile stresses following expansion are less than 25% of the local yield stress and are found on the transition in a narrow circumferential band at the immediate tube surface (< .0002 inch/0.005 mm depth). The measurements otherwise indicate a predominance of compressive stresses on and below the secondary-side surface of the transition zone. Excellent resistance to SWSCC initiation is offered by the low levels of tensile stress and cold-work. Propagation of any possible cracking would be deterred by the compressive stress field that surrounds this small volume of tensile material

  10. Database for Pressure Tube Diameter and Operation Data of Wolsong NPP

    International Nuclear Information System (INIS)

    Jung, Jong Yeob; Kim, W. Y.; Bae, J. H.; Park, J. H.

    2010-12-01

    Pressure tube of CANDU reactor which is a long cylindrical shape of its diameter about 10 cm and length of about 6m, can be expanded toward both radial and axial directions due to irradiation under the high pressure and temperature condition. As the irradiation period increases, the radial expansion due to creep of the pressure tube increases. The radial expansion of the pressure tube comes out the reduction of the coolability and it results in the power deration. The objectives of the current work is to establish the database for the measured diametral data of pressure tube and operational data from Wolsong NPP as a preliminary work of developing the prediction model for pressure tube diameter. In order to develop the database, measured data for total 86 channels were collected from Wolsong NPP 1, 2, 3 and 4 and analyzed. Based on the provided data, the operational conditions such as an axial temperature and a pressure of the channel and neutron fluxes were derived. All data were analysed to derive the correlation between the pressure tube diameter and the other operational parameters

  11. Environmentally assisted fatigue evaluation model of alloy 690 steam generator tube in high temperature water

    International Nuclear Information System (INIS)

    Tan Jibo; Wu Xinqiang; Han Enhou; Wang Xiang; Liu Xiaoqiang; Xu Xuelian

    2015-01-01

    Nickel-based alloy 690 has been widely used as steam generator tube in light water reactor (LWR) nuclear power plants, which may suffer from corrosion fatigue during long-term service. Many researches and operating experience indicated that the effect of LWR environment could significantly reduce the fatigue life of structural materials. However. such an environmental degradation effect was not fully addressed in the current ASME code design fatigue curves. Therefore, the Regulatory Guide 1.207 issued by US NRC required a new NPP have to incorporate the environment effects into fatigue analyses. In the last few decades, researchers in USA and Japan systematically investigated the corrosion fatigue behavior of nuclear-grade structural materials in LWR environment. Then, ANL model and JSME model were proposed, which incorporated environmental effects, including temperature, dissolved oxygen (DO) and strain rate for the nickel-based alloys. Due to lack of experiment data on domestic materials, there is no related environmental fatigue design model in China. In the present work, based on the corrosion fatigue tests of a kind of boat-shaped specimen in borated and lithiated high temperature water, the corrosion fatigue behavior and environmentally assisted cracking mechanism of domestic Alloy 690 steam generator tube have been investigate. An IMR model for the nickel-based alloy was proposed. The environmental fatigue life correction factor (F en ) was established, which addressed the environmental factors, including temperature, strain rate and dissolved oxygen. The method to evaluate environmental fatigue damage of structural materials in NPPs was proposed. (authors)

  12. IGA resistance of TT Alloy 690 and concentration behavior of Broached Egg Crate tube support configuration

    International Nuclear Information System (INIS)

    Suzuki, S.; Kusakabe, T.; Yamamoto, H.; Arioka, K.; Ochi, T.

    1992-01-01

    In order to improve the reliability of the Steam Generator (SG), TT Alloy 690 and BEC (Broached Egg Crate) type tube support plate has been developed. Some tests are carried out to heighten the reliability for these improvements all the more and the following results are obtained. (1) SERT test (Slow Extension Rate Test) made clear that TT690 has less IGA susceptibility in comparison with MA600. (2) The alkaline susceptibility on the occurrence of IGA/SCC on TT690 and MA600 obtained by SERT corresponds to that obtained by Model Boiler test. (3) By model boiler test, superior concentration behaviors for BEC type tube support plate configuration have been recognized in comparison with Drill type. This result is obtained by the joint research of the five utilities (Kansai Epco, Hokkaido Epco, Shikoku Epco, Kyushu Epco, JAPCO) and MHI

  13. X-ray study of texture in zirconium alloy tubes and in graphite

    International Nuclear Information System (INIS)

    Skvortsov, V.V.; Alekseev, S.I.

    1987-01-01

    X-ray study of texture in zirconium alloy tubes and in graphite has been developed. The method is based on constructing coordinate grid of stereographic projection determining quantity and coordinates of points where measurements should be performed depending on a specimen slope pitch. Complete stereographic projection obtained so is a base both for constructing pole figures showing distribution normales of plane system being studied and for calculating texture coefficients determining property anisotropy in materials under investigation. This method can be applied to study texture in items of any materials independent of the item shape

  14. Leakage Characteristics of Dual-Cannula Fenestrated Tracheostomy Tubes during Positive Pressure Ventilation: A Bench Study

    Directory of Open Access Journals (Sweden)

    Thomas Berlet

    2016-01-01

    Full Text Available This study compared the leakage characteristics of different types of dual-cannula fenestrated tracheostomy tubes during positive pressure ventilation. Fenestrated Portex® Blue Line Ultra®, TRACOE® twist, or Rüsch® Traceofix® tracheostomy tubes equipped with nonfenestrated inner cannulas were tested in a tracheostomy-lung simulator. Transfenestration pressures and transfenestration leakage rates were measured during positive pressure ventilation. The impact of different ventilation modes, airway pressures, temperatures, and simulated static lung compliance settings on leakage characteristics was assessed. We observed substantial differences in transfenestration pressures and transfenestration leakage rates. The leakage rates of the best performing tubes were <3.5% of the delivered minute volume. At body temperature, the leakage rates of these tracheostomy tubes were <1%. The tracheal tube design was the main factor that determined the leakage characteristics. Careful tracheostomy tube selection permits the use of fenestrated tracheostomy tubes in patients receiving positive pressure ventilation immediately after stoma formation and minimises the risk of complications caused by transfenestration gas leakage, for example, subcutaneous emphysema.

  15. Plastic instability in short tubes with internal pressure

    International Nuclear Information System (INIS)

    Blass, A.; Moro, N.

    1981-01-01

    Corrections are presented to a previous formulation, which becomes more suitable to the analysis of the plastic instability of both short and long tubes. Existing experimental evidence confirms moderately this statement. (Author) [pt

  16. Delayed hydride cracking velocity and crack growth measurement using DCPD technique in Zr-2.5Nb pressure tube material

    International Nuclear Information System (INIS)

    Singh, R.N.; Kishore, R.; Roychaudhury, S.; Unnikrishnan, M.; Sinha, T.K.; De, P.K.; Banerjee, S.; Kumar, Santosh

    2000-12-01

    Nuclear structural materials have to perform under most demanding and exotic environmental conditions. Due to its unique properties dilute zirconium alloys are the only choice for in-core structural materials in water cooled nuclear reactors. Hydrogen related problems have been recognized as the life-limiting factor for the core components of Pressurized Heavy Water Reactors (PHWR). Delayed Hydride Cracking (Dhc) is one of them. In this study, Dhc crack growth has been monitored using Direct Current Potential Drop (Dcp) technique. Calibration curve between normalized Dcp output and normalized crack length was established at different test temperatures. Dhc velocity was measured along the axial direction of the Zirconium-2.5Niobium pressure tube material at 203 and 250 degree C. (author)

  17. Influence of deuterium content on tensile behavior of Zr-2.5Nb pressure tube material in the temperature range of ambient to 300 degC

    International Nuclear Information System (INIS)

    Bind, A.K.; Singh, R.N.; Chakravartty, J.K.; Dhandharia, Priyesh; Ghosh, Agnish; More, Nitin S.; Chhatre, A.G.; Vijayakumar, S.

    2011-08-01

    Tensile properties of autoclaved zirconium-2.5 wt. % niobium pressure tube material were evaluated by uniaxial tension tests at temperatures between 25 and 300 degC and under strain-rates of 1.075 x 10 -4 /s. Six number of Zr-2.5Nb alloy pressure tube spools of length 130 mm were obtained from pressure tube number 19-2557-2. Five spools were polished with abrasive paper to remove the oxide layer. These spools were gaseously charged with controlled amount of deuterium. The target deuterium concentrations were 25, 50, 75, 100 and 200 wppm of hydrogen equivalent. Ten samples were machined by EDM wire cutting from every spool. The tensile specimen axis was oriented along longitudinal direction of the tube. Metallographic examination of the deuterium charged samples suggested that the deuterides were predominantly circumferential deuterides. Analysis of tensile results showed that both yield and ultimate tensile strengths of this alloy decreased monotonically with increasing test temperatures. The tensile ductility decreased marginally with increase in test temperature from ambient to 300 degC. It was also observed that both strength and ductility appear to be unaffected by deuterium content at all temperatures, thereby suggesting that at least up to 200 wppm (Heq.) of deuterium tensile properties are not influenced by deuterium. (author)

  18. [Characterizing the passive opening of the eustachian tube in a hypo-/hyperbaric pressure chamber].

    Science.gov (United States)

    Meyer, M F; Mikolajczak, S; Luers, J C; Lotfipour, S; Beutner, D; Jumah, M D

    2013-09-01

    Beside arbitrary and not arbitrary active pressure equalization systems there is a passive equalization system via the Eustachian tube (ET) at pressure difference between the epipharyngeal space and the middle ear. Aim of this study was to characterize this passive equalization system in a hypobaric/hyperbaric pressure chamber by continuously measuring the tympanic impedance. In contrast to other studies, which are measured only in a hypobaric pressure chamber it is possible to include participants with Eustachian tube dysfunction (ETD). Following a fixed pressure profile 39 participants were exposed to phases of pressure rising and decompression. By continuously measuring the tympanic impedance in the pressure chamber it was possible to measure data of the Eustachian Tube opening Pressure (ETOP), Eustachian Tube closing pressure (ETCP) and Eustachian Tube opening duration (ETOD). In addition it was possible to characterize the gradient of pressure during decompression, while the ET was open. Beside the measurement of the arithmetic average of the ETOP (30.2 ± 15.1 mbar), ETCP (9.1 ± 7.7 mbar) and ETOD (0.65 ± 0.38 s) it was obvious that there are recurrent samples of pressure progression during the phase of tube opening. Generally it is possible to differentiate between the type of complete opening and partial opening. The fundamental characterization of the action of the passive tube opening, including the measurement of the ETOP, ETCP and ETOD, is a first step in understanding the physiological and pathophysiological function of the ET. © Georg Thieme Verlag KG Stuttgart · New York.

  19. Evaluation of oxides formed at high temperatures in Zr-2.5Nb pressure tubing

    Energy Technology Data Exchange (ETDEWEB)

    Kiran Kumar, N.A.P.; Szpunar, J.A., E-mail: kiraniitkgp@yahoo.com [Univ. of Saskatchewan, Saskatoon, Saskatchewan (Canada)

    2012-07-01

    The oxidation behavior of Zr-2.5Nb pressure tube samples has been studied at four different temperatures, i.e., 400°, 600°, 800°, and 1000°C. The amount of tetragonal phase is found to decrease with increase of temperature. The oxide texture of (002){sub m} and (111){sub m} type increased with the temperature from 400°C to 600°C, however at temperatures above 600°C the texture strength seems to diminish and the oxide layer becomes structurally unstable. Further, the impedance response is found to be dependent on the microstructure of the oxide film. For the sample oxidized at 400°C, Electrochemical Impedance Spectroscopy (EIS) spectra exhibited a two-time constant behavior, showing the formation of two-layer oxide film on the Zr-2.5Nb alloy, which correspond to a porous outer oxide and a barrier inner oxide, respectively. In addition, the samples were oxidized at constant temperature of 600°C with varying oxidation time. The observation shows that the oxide is more protective in the early stage of oxide growth. However, further growth of oxide film has resulted in degeneration of its protective character. (author)

  20. The role of strain localization in the fracture of irradiated pressure tube material

    International Nuclear Information System (INIS)

    Dutton, R.

    1989-04-01

    This report reviews those phenomena that lead to strain localization in zirconium alloys, with particular reference to the role played by the formation of shear bands in fracture processes. The important influence of plastic deformation, in general, on fracture mechanisms is emphasized. This is to be expected when elastic-plastic fracture mechanics is the chosen analytical technique. Intensely inhomogeneous characteristics of strain localization cause an abrupt bifurcation in the evolution of deformation strain and lead to plastic instability linked with intrinsic material behaviour (e.g., work softening) or of geometric origin (e.g., localized necking). Both of these effects are discussed in relation to measurable deformation parameters, such as the work hardening rate and strain rate sensitivity, which determine the degree of resistance to plastic instability. The modifying effect of irradiation on these quantities is given specific attention, the appropriate literature pertaining to Zircaloy and Zr-2.5% Nb being reviewed. Recommendations are made for a combined experimental and theoretical program to characterize strain localization and reduced ductility in irradiated cold-worked Zr-2.5% Nb pressure tube material. The relationship between the deformation properties and the fracture behaviour is discussed

  1. Deformation behavior of irradiated Zr-2.5Nb pressure tube material

    International Nuclear Information System (INIS)

    Himbeault, D.D.; Chow, C.K.; Puls, M.P.

    1994-01-01

    A study of the deformation behavior of irradiated highly textured Zr-2.5Nb pressure tube material in the temperature range of 30 degree C to 300 degree C was undertaken to understand better the mechanism for the deterioration of the fracture toughness with neutron irradiation. Strain localization behavior, believed to be a main contributor to reduced toughness, was observed in irradiated transverse tensile specimens at temperature greater than 100 degree C. The strain localization behavior was found to occur by the cooperative twinning of the highly textured grains of the material, resulting in a local softening of the material, where the flow than localizes. It is believed that the effect of the irradiation is to favor twinning at the expense of slip in the early stages of deformation. This effect becomes more pronounced at higher temperature, thus leading to the high-temperature strain localization behavior of the material. A limited amount of dislocation channeling was also observed; however, it is not considered to have a major role in the strain localization behavior of the material. Contrary to previous reports on irradiated zirconium alloys, static strain aging is observed in the irradiated material in the temperature range of 150 degree C to 300 degree C

  2. Chemical aspects of hydrogen ingress in zirconium and zircaloy pressure tubes: ageing management of Indian PHWR coolant channels - determination of hydrogen and deuterium

    International Nuclear Information System (INIS)

    Sayi, Y.S.; Shankaran, P.S.; Yadav, C.S.; Ramanjaneyulu, P.S.; Venugopal, V.; Ramakumar, K.L.; Chhapru, G.C.; Prasad, R.; Jain, H.C.; Sood, D.D.

    2009-02-01

    Pressurized heavy water reactors (PHWRs) use zirconium and zirconium based alloys as clad and coolant tubes since its beginning. The first ever zircaloy-2 pressure tube failure occurred in 1983 at Ontario Hydro's Pickering Unit 2 in Canada which necessitated a thorough examination of causes of such failure. The failure was attributed to massive hydriding at the failed spot of pressure tube. Continuous usage of zirconium alloys could result in their hydrogen and deuterium pick-up leading to hydrogen/ deuterium embrittlement. The life of the zircaloy coolant channels is dictated by hydrogen/deuterium content and hence ageing management of the pressure tubes is essential for ensuring their trouble-free usage. It is desirable to have a sound knowledge on the chemical aspects of zirconium and zirconium based alloys metallurgy, the mechanistic principles of hydrogen ingress into the pressure tubes during in reactor service, and identifying suitable analytical methodologies for precise and accurate determination of hydrogen in wafer thin sliver samples carved out from insides of pressure tubes without causing any structural damage so that it can continue to remain in service. This is desirable so that the ageing management does not result in cost-escalation. This report is divided in to three main parts. The first part deals with the chemical aspects of zirconium and zirconium based alloy metallurgy, the mechanism of hydrogen pick-up and hydride formation in zirconium matrix. The second part describes various methodologies and their limitations, available for hydrogen/deuterium determination. The third part deals in detail, about the extensive investigations carried out at Radioanalytical Chemistry Division (RACD) in Radiochemistry and Isotope Group for establishing an indigenously developed hot vacuum extraction system in combination with quadrupole mass spectrometry for precise determination of hydrogen and deuterium in wafer thin sliver sample of zircaloy. The

  3. Tube Plugging Criteria for the High-pressure Heaters of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyungnam; Cho, Nam-Cheoul; Lee, Kuk-hee [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, a method to establish the tube plugging criteria of BOP heat exchangers is introduced and the tube plugging criteria for the high pressure heaters of a nuclear power plant. This method relies on the similar plugging criteria used in the steam generator tubes. Power generation field urges nuclear power plants to reduce operating and maintaining costs to remain competitive. To reduce the cost by means of preventing the lowering thermal efficiency, the inspection of balance-of-plant heat exchanger, which was treated as not important work, becomes important. The tubing materials and tube thickness of heat exchangers in nuclear power plants are selected to withstand system temperature, pressure, and corrosion. But tubes have experienced leaks and failures and plugged based upon eddy current testing (ET) results. There are some problems for plugging the heat exchanger tubes since the criterion and its basis are not clearly described. For this reason, the criteria for the tube wall thickness are addressed in order to operate the heat exchangers in nuclear power plant without trouble during the cycle. The feed water heater is a kind of heat exchanger which raises the temperature of water supplied from the condenser. The heat source of high-pressure heaters is the extraction steam from the high-pressure turbine and moisture separator re-heater. If the tube wall of the heater is broken, the feed water flowing inside the tube intrudes to shell side. This forces the turbine to be stop in order to protect it. There are many codes and standards to be referred for calculating the minimum thickness of the heat exchanger tube in the designing stage. However, the codes and standards related to show the tube plugging criteria may not exist currently. A method to establish the tube plugging criteria of BOP heat exchangers is introduced and the tube plugging criteria for the high pressure heaters of Ulchin NPP No. 3 and 4. This method relies on the similar plugging

  4. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1] [Sample summary reports of pressure tube samples from Argentina, India, Canada, Republic of Korea, and Romania

    International Nuclear Information System (INIS)

    2006-05-01

    Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

  5. Thermal behaviour of pressure tube under fully and partially voided heating conditions using 19 pin fuel element simulator

    International Nuclear Information System (INIS)

    Yadav, Ashwini K.; Kumar, Ravi; Gupta, Akhilesh; Chatterjee, B.; Mukhopadhya, D.; Lele, H.G.

    2011-01-01

    In a nuclear reactor temperature can rise drastically during LOCA due to failure of heat transportation system and subsequently leads to mechanical deformations like sagging, ballooning and breaching of pressure tube. To understand the phenomenon an experiment has been carried out using 19 pin fuel element simulator. Main purpose of the experiment was to trace temperature profiles over the pressure tube, calandria tube and clad tubes of 220 MWe Indian Pressurised Heavy Water Reactor (IPHWR). The symmetrical heating of pressure tube of 1 m length was done through resistance heating of 19 pins under 13.5 kW power using a rectifier and the variation of temperatures over the circumference of pressure tube (PT), calandria tube (CT) and clad tubes were measured. The sagging of pressure tube was initiated at 460 deg C temperature and highest temperature attained was 650 deg C. The highest temperature attained by clad tubes was 680 deg C (over outer ring) and heat is dissipated to calandria vessel mainly due to radiation and natural convection. Again to simulate partially voided conditions, asymmetrical heating of pressure was carried out by injecting 8 kW power to upper 8 pins of fuel simulator. A maximum temperature difference of 295 deg C was observed over the circumference of pressure tube which highlights the magnitude of thermal stresses and its role in breaching of pressure tube under partially voided conditions. Integrity of pressure tube was retained during both symmetrical and asymmetrical heatup conditions. (author)

  6. Proposal for a Coordinated Research Project: Prediction of Axial and Radial Creep in Pressure Tubes

    International Nuclear Information System (INIS)

    Bozzano, Patricia B.

    2013-01-01

    Participation of Argentina: • hydrogen charge of 90 samples for hydrogen determination: IGF, HVEMS, DSC, Resistivity, DTA; • 17 samples for non destructive techniques; • 6 blisters in CANDU pressure tube sections for NDT evaluation

  7. Thermodynamic analysis of transition pressure of δ-stabilized binary plutonium alloys

    International Nuclear Information System (INIS)

    Wang Qinghui

    1992-01-01

    The transformation of δ-stabilized binary plutonium alloys to α-Pu was studies by thermodynamic analysis. A transition pressure-composition equation which can characterize the high pressure transformation from δ to α was derived. Values calculated by the equation and values measured by experiments of published references have the same tendency. the following facts can be explained properly by this equation. (1)The transformation pressure increases linearly with the amount of an alloying element. (2) The slope of the plot of transformation pressure versus composition of δ-Pu alloys is inversely proportional to the minimum amount of solute required to retain δ-phase at room temperature and pressure. (3) Curves showing the relationship between transformation pressure and composition of various δ-stabilized binary alloys interact at the same point of zero solute (transformation pressure axis). In addition, some transformation pressures from δ to α of δ-stabilized alloys are predicted by using the modified theoretical equation

  8. 76 FR 60083 - Carbon and Alloy Seamless Standard, Line, and Pressure Pipe From Japan and Romania

    Science.gov (United States)

    2011-09-28

    ... Alloy Seamless Standard, Line, and Pressure Pipe From Japan and Romania Determinations On the basis of... pressure pipe from Japan and Romania would be likely to lead to continuation or recurrence of material... regarding small- diameter carbon and alloy seamless standard, line, and pressure pipe from Romania...

  9. Pressure effects on Al89La6Ni5 amorphous alloy crystallization

    DEFF Research Database (Denmark)

    Zhuang, Yanxin; Jiang, Jianzhong; Zhou, T. J.

    2000-01-01

    The pressure effect on the crystallization of the Al89La6Ni5 amorphous alloy has been investigated by in situ high-pressure and high-temperature x-ray powder diffraction using synchrotron radiation. The amorphous alloy crystallizes in two steps in the pressure range studied (0-4 GPa). The first p...

  10. Two and dimensional heat analysis inside a high pressure electrical discharge tube

    International Nuclear Information System (INIS)

    Aghanajafi, C.; Dehghani, A. R.; Fallah Abbasi, M.

    2005-01-01

    This article represents the heat transfer analysis for a horizontal high pressure mercury steam tube. To get a more realistic numerical simulation, heat radiation at different wavelength width bands, has been used besides convection and conduction heat transfer. The analysis for different gases with different pressure in two and three dimensional cases has been investigated and the results compared with empirical and semi empirical values. The effect of the environmental temperature on the arc tube temperature is also studied

  11. Simulation and analysis of the thermal and deformation behaviour of `as-received` and `hydrided` pressure tubes used in the circumferential temperature distribution experiments (end of life/pressure tube behaviour)

    Energy Technology Data Exchange (ETDEWEB)

    Muir, W C; Bayoumi, M H [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    It is postulated that in-reactor pressure tubes may be subjected to radiation damage and dissolved deuterium which could change the pressure tube characteristics and lead to different behaviour than that of as-received pressure tubes under large LOCA (loss of coolant) conditions. A hydrided pressure tube was used to study the effect of dissolved hydrogen on thermal-mechanical behaviour. In the experiment, simulating an in-reactor (hydrided) pressure tube with circumferential differential temperature under boil-off conditions, the pressure tube ballooned into contact with the calandria tube. The pressure tube used in this experiment was hydrided in a furnace to a nominal value of 200 {mu}g/g dissolved hydrogen. This test was a repeat of the first supplementary boil-off test (S-5-1) which used an as-received pressure tube. The objective of this paper is to analyze the results obtained from the simulation of this Boil-Off test using the SMARTT computer code and to examine the effect of hydriding on the thermal and ballooning behaviour of the pressure tube by comparison with the results obtained from test S-5-1. A discussion of the results obtained from this comparison is presented together with an analysis of their application to the analysis of pressure tube behaviour in CANDU reactors. (author). 13 refs., 1 tab., 16 figs.

  12. Simulation and analysis of the thermal and deformation behaviour of 'as-received' and 'hydrided' pressure tubes used in the circumferential temperature distribution experiments (end of life/pressure tube behaviour)

    International Nuclear Information System (INIS)

    Muir, W.C.; Bayoumi, M.H.

    1995-01-01

    It is postulated that in-reactor pressure tubes may be subjected to radiation damage and dissolved deuterium which could change the pressure tube characteristics and lead to different behaviour than that of as-received pressure tubes under large LOCA (loss of coolant) conditions. A hydrided pressure tube was used to study the effect of dissolved hydrogen on thermal-mechanical behaviour. In the experiment, simulating an in-reactor (hydrided) pressure tube with circumferential differential temperature under boil-off conditions, the pressure tube ballooned into contact with the calandria tube. The pressure tube used in this experiment was hydrided in a furnace to a nominal value of 200 μg/g dissolved hydrogen. This test was a repeat of the first supplementary boil-off test (S-5-1) which used an as-received pressure tube. The objective of this paper is to analyze the results obtained from the simulation of this Boil-Off test using the SMARTT computer code and to examine the effect of hydriding on the thermal and ballooning behaviour of the pressure tube by comparison with the results obtained from test S-5-1. A discussion of the results obtained from this comparison is presented together with an analysis of their application to the analysis of pressure tube behaviour in CANDU reactors. (author). 13 refs., 1 tab., 16 figs

  13. Refrigerant charge, pressure drop, and condensation heat transfer in flattened tubes

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, M J; Newell, T A; Chato, J C [University of Illinois, Urbana, IL (United States). Dept. of Mechanical and Industrial Engineering; Infante Ferreira, C A [Delft University of Technology (Netherlands). Laboratory for Refrigeration and Indoor Climate Control

    2003-06-01

    Horizontal smooth and microfinned copper tubes with an approximate diameter of 9 mm were successively flattened in order to determine changes in flow field characteristics as a round tube is altered into a flattened tube profile. Refrigerants R134a and R410A were investigated over a mass flux range from 75 to 400 kg m{sup -2} s{sup -}2{sup 1} and a quality range from approximately 10-80%. For a given refrigerant mass flow rate, the results show that a significant reduction in refrigerant charge is possible. Pressure drop results show increases of pressure drop at a given mass flux and quality as a tube profile is flattened. Heat transfer results indicate enhancement of the condensation heat transfer coefficient as a tube is flattened. Flattened tubes with an 18{sup o} helix angle displayed the highest heat transfer coefficients. Smooth tubes and axial microfin tubes displayed similar levels of heat transfer enhancement. Heat transfer enhancement is dependent on the mass flux, quality and tube profile. (author)

  14. Stress relaxation analysis and irradiation creep and swelling in pressure tubes

    International Nuclear Information System (INIS)

    Beeston, J.M.; Burr, T.K.

    1979-01-01

    An analysis is presented of slit width test information on two pressure tubes that had been irradiated in test reactors. The analysis showed that differential swelling stresses and thermal stresses undergo relaxation. The mechanism responsible for the stress relaxation at temperatures less than 700 K was irradiation creep. Irradiation creep in thermal test reactor pressure tubes is evidently greater than it would be at equivalent conditions in fast reactors. The residual stresses observed in the slit width tests varied between 30 and 257 MPa and would act to reduce the operating stresses, thus allowing for increased service life of the tubes as compared with no stress relaxation

  15. SeZnSb alloy and its nano tubes, graphene composites properties

    Directory of Open Access Journals (Sweden)

    Abhay Kumar Singh

    2013-04-01

    Full Text Available Composite can alter the individual element physical property, could be useful to define the specific use of the material. Therefore, work demonstrates the synthesis of a new composition Se96-Zn2-Sb2 and its composites with 0.05% multi-walled carbon nano tubes and 0.05% bilayer graphene, in the glassy form. The diffused amorphous structure of the multi walled carbon nano tubes and bilayer gaphene in the Se96-Zn2-Sb2 alloy have been analyzed by using the Raman, X-ray photoluminescence spectroscopy, Furrier transmission infrared spectra, photoluminescence, UV/visible absorption spectroscopic measurements. The diffused prime Raman bands (G and D have been appeared for the multi walled carbon nano tubes and graphene composites, while the X-ray photoluminescence core energy levels peak shifts have been observed for the composite materials. Subsequently the photoluminescence property at room temperature and a drastic enhancement (upto 80% in infrared transmission percentage has been obtained for the bilayer graphene composite, along with optical energy band gaps for these materials have been evaluated 1.37, 1.39 and 1.41 eV.

  16. Externally fired gas turbine cycles with high temperature heat exchangers utilising Fe-based ODS alloy tubing

    International Nuclear Information System (INIS)

    Olsson, F.; Svensson, S.-A.; Duncan, R.

    2001-01-01

    This work is part of the BRITE / EuRAM Project 'Development of Torsional Grain Structures to Improve Biaxial Creep Performance of Fe-based ODS Alloy Tubing for Biomass Power Plant'. The main goal of this project is to heat exchanger tubes working at 1100 o C and above. The paper deals with design implications of a biomass power plant, using an indirectly fired gas turbine with a high temperature heat exchanger containing Fe-based ODS alloy tubing. In the current heat exchanger design, ODS alloy tubing is used in a radiant section, using a bayonet type tube arrangement. This enables the use of straight sections of ODS tubing and reduces the amount of material required. In order to assess the potential of the power plant system, thermodynamic calculations have been conducted. Both co-generation and condensing applications are studied and results so far indicate that the electrical efficiency is high, compared to values reached by conventional steam cycle power plants of the same size (approx. 5 MW e ). (author)

  17. Effect of tubing condensate on non-invasive positive pressure ventilators tested under simulated clinical conditions.

    Science.gov (United States)

    Hart, Diana Elizabeth; Forman, Mark; Veale, Andrew G

    2011-09-01

    Water condensate in the humidifier tubing can affect bi-level ventilation by narrowing tube diameter and increasing airflow resistance. We investigated room temperature and tubing type as ways to reduce condensate and its effect on bi-level triggering and pressure delivery. In this bench study, the aim was to test the hypothesis that a relationship exists between room temperature and tubing condensate. Using a patient simulator, a Res-med bi-level device was set to 18/8 cm H(2)O and run for 6 h at room temperatures of 16°C, 18°C and 20°C. The built-in humidifier was set to a low, medium or high setting while using unheated or insulated tubing or replaced with a humidifier using heated tubing. Humidifier output, condensate, mask pressure and triggering delay of the bi-level were measured at 1 and 6 h using an infrared hygrometer, metric weights, Honeywell pressure transducer and TSI pneumotach. When humidity output exceeded 17.5 mg H(2)O/L, inspiratory pressure fell by 2-15 cm H(2)O and triggering was delayed by 0.2-0.9 s. Heating the tubing avoided any such ventilatory effect whereas warmer room temperatures or insulating the tubing were of marginal benefit. Users of bi-level ventilators need to be aware of this problem and its solution. Bi-level humidifier tubing may need to be heated to ensure correct humidification, pressure delivery and triggering.

  18. AMPTRACT: an algebraic model for computing pressure tube circumferential and steam temperature transients under stratified channel coolant conditions

    International Nuclear Information System (INIS)

    Gulshani, P.; So, C.B.

    1986-10-01

    In a number of postulated accident scenarios in a CANDU reactor, some of the horizontal fuel channels are predicted to experience periods of stratified channel coolant condition which can lead to a circumferential temperature gradient around the pressure tube. To study pressure tube strain and integrity under stratified flow channel conditions, it is, necessary to determine the pressure tube circumferential temperature distribution. This paper presents an algebraic model, called AMPTRACT (Algebraic Model for Pressure Tube TRAnsient Circumferential Temperature), developed to give the transient temperature distribution in a closed form. AMPTRACT models the following modes of heat transfer: radiation from the outermost elements to the pressure tube and from the pressure to calandria tube, convection between the fuel elements and the pressure tube and superheated steam, and circumferential conduction from the exposed to submerged part of the pressure tube. An iterative procedure is used to solve the mass and energy equations in closed form for axial steam and fuel-sheath transient temperature distributions. The one-dimensional conduction equation is then solved to obtain the pressure tube circumferential transient temperature distribution in a cosine series expansion. In the limit of large times and in the absence of convection and radiation to the calandria tube, the predicted pressure tube temperature distribution reduces identically to a parabolic profile. In this limit, however, radiation cannot be ignored because the temperatures are generally high. Convection and radiation tend to flatten the parabolic distribution

  19. Numerical study on turbulent heat transfer and pressure drop of nanofluid in coiled tube-in-tube heat exchangers

    International Nuclear Information System (INIS)

    Aly, Wael I.A.

    2014-01-01

    Highlights: • The performance of helically coiled tube heat exchanger using nanofluid is modeled. • The 3D turbulent flow and conjugate heat transfer of CTITHE are solved using FVM. • The effects of nanoparticle concentration and curvature ratio are investigated. • The Gnielinski correlation for Nu for turbulent flow in helical tubes can be used for water-based Al 2 O 3 nanofluid. - Abstract: A computational fluid dynamics (CFD) study has been carried out to study the heat transfer and pressure drop characteristics of water-based Al 2 O 3 nanofluid flowing inside coiled tube-in-tube heat exchangers. The 3D realizable k–ε turbulent model with enhanced wall treatment was used. Temperature dependent thermophysical properties of nanofluid and water were used and heat exchangers were analyzed considering conjugate heat transfer from hot fluid in the inner-coiled tube to cold fluid in the annulus region. The overall performance of the tested heat exchangers was assessed based on the thermo-hydrodynamic performance index. Design parameters were in the range of; nanoparticles volume concentrations 0.5%, 1.0% and 2.0%, coil diameters 0.18, 0.24 and 0.30 m, inner tube and annulus sides flow rates from 2 to 5 LPM and 10 to 25 LPM, respectively. Nanofluid flows inside inner tube side or annular side. The results obtained showed a different behavior depending on the parameter selected for the comparison with the base fluid. Moreover, when compared at the same Re or Dn, the heat transfer coefficient increases by increasing the coil diameter and nanoparticles volume concentration. Also, the friction factor increases with the increase in curvature ratio and pressure drop penalty is negligible with increasing the nanoparticles volume concentration. Conventional correlations for predicting average heat transfer and friction factor in turbulent flow regime such as Gnielinski correlation and Mishra and Gupta correlation, respectively, for helical tubes are also valid for

  20. In vitro evaluation of the method effectiveness to limit inflation pressure cuffs of endotracheal tubes

    Directory of Open Access Journals (Sweden)

    Rafael de Macedo Coelho

    2016-04-01

    Full Text Available ABSTRACT BACKGROUND AND OBJECTIVE: Cuffs of tracheal tubes protect the lower airway from aspiration of gastric contents and facilitate ventilation, but may cause many complications, especially when the cuff pressure exceeds 30 cm H2O. This occurs in over 30% of conventional insufflations, so it is recommended to limit this pressure. In this study we evaluated the in vitro effectiveness of a method of limiting the cuff pressure to a range between 20 and 30 cm H2O. METHOD: Using an adapter to connect the tested tube to the anesthesia machine, the relief valve was regulated to 30 cm H2O, inflating the cuff by operating the rapid flow of oxygen button. There were 33 trials for each tube of three manufacturers, of five sizes (6.5-8.5, using three times inflation (10, 15 and 20 s, totaling 1485 tests. After inflation, the pressure obtained was measured with a manometer. Pressure >30 cm H2O or <20 cm H2O were considered failures. RESULTS: There were eight failures (0.5%, 95% CI: 0.1-0.9%, with all by pressures <20 cm H2O and after 10 s inflation (1.6%, 95% CI: 0 5-2.7%. One failure occurred with a 6.5 tube (0.3%, 95% CI: -0.3 to 0.9%, six with 7.0 tubes (2%, 95% CI: 0.4-3.6%, and one with a 7.5 tube (0.3%, 95% CI: -0.3 to 0.9%. CONCLUSION: This method was effective for inflating tracheal tube cuffs of different sizes and manufacturers, limiting its pressure to a range between 20 and 30 cm H2O, with a success rate of 99.5% (95% CI: 99.1-99.9%.

  1. Multiaxial creep of tubes of Alloy 800 and Alloy 617 at high temperature

    International Nuclear Information System (INIS)

    Penkalla, H.J.; Schubert, F.; Nickel, H.

    1989-01-01

    The deformation behaviour under multiaxial loading at temperature higher than 800 deg. C is strongly controlled by creep. For dimensioning and inelastic analysis the use of v. Mises theory and Norton's creep law for stationary creep are demonstrated for different combination of internal pressure and axial or torsional stress or strains. The experimental results are in satisfactory agreement with the theoretical predicted deformation behaviour if values for the coefficient k and n in Norton's creep law are used, which are close to the real creep resistance in the component. (author). 11 refs, 12 figs, 2 tabs

  2. THE EFFECTS OF AREA CONTRACTION ON SHOCK WAVE STRENGTH AND PEAK PRESSURE IN SHOCK TUBE

    Directory of Open Access Journals (Sweden)

    A. M. Mohsen

    2012-06-01

    Full Text Available This paper presents an experimental investigation into the effects of area contraction on shock wave strength and peak pressure in a shock tube. The shock tube is an important component of the short duration, high speed fluid flow test facility, available at the Universiti Tenaga Nasional (UNITEN, Malaysia. The area contraction was facilitated by positioning a bush adjacent to the primary diaphragm section, which separates the driver and driven sections. Experimental measurements were performed with and without the presence of the bush, at various diaphragm pressure ratios, which is the ratio of air pressure between the driver (high pressure and driven (low pressure sections. The instantaneous static pressure variations were measured at two locations close to the driven tube end wall, using high sensitivity pressure sensors, which allow the shock wave strength, shock wave speed and peak pressure to be analysed. The results reveal that the area contraction significantly reduces the shock wave strength, shock wave speed and peak pressure. At a diaphragm pressure ratio of 10, the shock wave strength decreases by 18%, the peak pressure decreases by 30% and the shock wave speed decreases by 8%.

  3. Precision tubes for high-pressure diesel injection lines; Praezisrohre fuer Hochdruck-Dieseleinspritzleitungen

    Energy Technology Data Exchange (ETDEWEB)

    Hagedorn, M.; Lechtenfeld, U.; Zaremba, A. [Mannesmann Praezisrohr GmbH, Hamm (Germany)

    2008-03-15

    The requirements on diesel injection lines raise because of increasing customers demands and more rigid environmental laws. In this context higher injection pressures effect both aspects positively. One important condition for increasing pressure levels is the economical provision of suitable injection lines. To reach this aim, Mannesmann Praezisrohr GmbH developed precision tubes for injection lines, which are fulfilling these increasing requirements. (orig.)

  4. Estimation of fracture toughness of Zr 2.5% Nb pressure tube of Pressurised Heavy Water Reactor using cyclic ball indentation technique

    Energy Technology Data Exchange (ETDEWEB)

    Chatterjee, S., E-mail: subrata@barc.gov.in; Panwar, Sanjay; Madhusoodanan, K.; Rama Rao, A.

    2016-08-15

    Highlights: • Measurement of fracture toughness of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situ Property Measurement System (IProMS) has been designed in house. • Conventional and IProMS tests conducted on pressure tube spool pieces having different mechanical properties. • Correlation has been established between the conventional and IProMS estimated fracture properties. - Abstract: In Pressurised Heavy Water Reactors (PHWRs) fuel bundles are located inside horizontal pressure tubes made up of Zr 2.5 wt% Nb alloy. Pressure tubes undergo degradation during its service life due to high pressure, high temperature and radiation environment. Measurement of mechanical properties of degraded pressure tubes is important for assessing their fitness for further operation. Presently as per safety guidelines imposed by the regulatory body, a few pre-decided pressure tubes are removed from the reactor core at regular intervals during the planned reactor shut down to carry out post irradiation examination (PIE) in a laboratory which consumes lots of man-rem and imposes economic penalties. Hence a system is indeed felt necessary which can carry out experimental trials for measurement of mechanical properties of pressure tubes under in situ conditions. The only way to accomplish this important objective is to develop a system based on an in situ measurement technique. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing an indentation test either on the outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ conditions. Considering the importance of such measurements, an In situ Property

  5. Experimental study on condensation heat transfer enhancement and pressure drop penalty factors in four microfin tubes

    Energy Technology Data Exchange (ETDEWEB)

    Han, D [Korea University, Seoul (Korea). Institute of Advanced Machinery Design; Lee, Kyu-Jung [Korea University, Seoul (Korea). Dept. of Mechanical Engineering

    2005-08-01

    Heat transfer and pressure drop characteristics of four microfin tubes were experimentally investigated for condensation of refrigerants R134a, R22, and R410A in four different test sections. The microfin tubes examined during this study consisted of 8.92, 6.46, 5.1, and 4 mm maximum inside diameter. The effect of mass flux, vapor quality, and refrigerants on condensation was investigated in terms of the heat transfer enhancement factor and the pressure drop penalty factor. The pressure drop penalty factor and the heat transfer enhancement factor showed a similar tendency for each tube at given vapor quality and mass flux. Based on the experimental data and the heat-momentum analogy, correlations for the condensation heat transfer coefficients in an annular flow regime and the frictional pressure drops are proposed. (author)

  6. Experimental study and modeling of high-temperature oxidation and phase transformation of cladding-tubes made in zirconium alloy

    International Nuclear Information System (INIS)

    Mazeres, Benoit

    2013-01-01

    One of the hypothetical accident studied in the field of the safety studies of Pressurized light Water Reactor (PWR) is the Loss-Of-Coolant-Accident (LOCA). In this scenario, zirconium alloy fuel claddings could undergo an important oxidation at high temperature (T≅ 1200 C) in a steam environment. Cladding tubes constitute the first confinement barrier of radioelements and then it is essential that they keep a certain level of ductility after quenching to ensure their integrity. These properties are directly related to the growth kinetics of both the oxide and the αZr(O) phase and also to the oxygen diffusion profile in the cladding tube after the transient. In this context, this work was dedicated to the understanding and the modeling of the both oxidation phenomenon and oxygen diffusion in zirconium based alloys at high temperature. The numerical tool (EKINOX-Zr) used in this thesis is based on a numerical resolution of a diffusion/reaction problem with equilibrium-conditions on three moving boundaries: gas/oxide, oxide/αZr(O), αZr(O)/βZr. EKINOX-Zr kinetics model is coupled with ThermoCalc software and the Zircobase database to take into account the influence of the alloying elements (Sn, Fe, Cr, Nb) but also the influence of hydrogen on the solubility of oxygen. This study focused on two parts of the LOCA scenario: the influence of a pre-oxide layer (formed in-service) and the effects of hydrogen. Thanks to the link between EKINOX-Zr and the thermodynamic database Zircobase, the hydrogen effects on oxygen solubility limit could be considered in the numerical simulations. Thus, simulations could reproduce the oxygen diffusion profiles measured in pre-hydrided samples. The existence of a thick pre-oxide layer on cladding tubes can induce a reduction of this pre-oxide layer before the growth of a high-temperature one during the high temperature dwell under steam. The first simulations performed using the numerical tool EKINOX-Zr showed that this particular

  7. Endotracheal tube cufi pressures in adult patients undergoing ...

    African Journals Online (AJOL)

    obstructing tracheal mucosal blood flow but high enough to form an effective seal when delivering PPV. Tracheal ... the capillary blood pressure supplying the trachea and is followed by ischaemia with inflammation. ... The aim of this study was to determine the ETT cuff pressures of patients receiving general anaesthesia at ...

  8. Endotracheal tube cuff pressure management in adult critical care ...

    African Journals Online (AJOL)

    of respondents performed cuff pressure measurements every 6 - 12 hours; 32% reported ... effectively when performing ETT cuff pressure management, to reduce practice variance, ... questionnaires were distributed to professional nurses working in .... MOV, CPM and the palpation method.3 No advantage of CPM over.

  9. Plugging of feed inlet tube upstands with Ni/Ti shape memory alloy plugs - Heysham 1 power station

    International Nuclear Information System (INIS)

    Mathews, A.J.

    1988-01-01

    The paper contains a description of a new approach for Plugging feed inlet tubes of Gas-Cooled Reactors. Instead of utilizing the original explosive method plugging by fitting a shape memory alloy plug into the upstand is being described. (author)

  10. Welding qualification procedure for fuel rods tubes of Zr-Sn alloys by the TIG automatic process

    International Nuclear Information System (INIS)

    1984-11-01

    It is presented the requirements to be used in the Welding qualification procedure for tubes of Zr-Sn alloys, specified in the ASTM B353 regulatory guide, used in the fabrication of fuel rods PWR reactors by the automatic TIG process. (E.G.) [pt

  11. Stress analysis of pressurized water reactor steam generator tube denting phenomena. Interim report

    International Nuclear Information System (INIS)

    Thomas, J.M.; Cipolla, R.C.; Ranjan, G.V.; Derbalian, G.

    1978-07-01

    In some Pressurized Water Reactor (PWR) steam generators, a corrosion product has formed on the carbon steel support plate in the crevice between the tube and support plate. The corrosion product occupies more volume than the original metal; the tube-to-support plate crevice volume is thus consumed with corrosion product, and further corrosive action results in a radially inward force on the tube and a radially outward force on the corroding support plate. This has resulted in indentation (''denting'') of the tube, accompanied by occasional cracking. Large in-plane deformation and cracking of support plates has also been observed in the most severely affected plants along with some serious side effects, such as deformation and cracking of inner row tube U-bends caused by support plate movement. Mechanical aspects of the tube denting phenomena have been studied using analytical models. The models used ranged from closed form analytical solutions to state-of-the-art numerical elastic-plastic computer program for moderate strains. It was found that tube dents, such as those observed in operating steam generators, are associated with yielding of both the tubes and support plates. Also studied were the stresses in tube U-bends caused by support plate flow slot deformation

  12. A Study On Critical Thinning In Thin-walled Tube Bending Of Al-Alloy 5052O Via Coupled Ductile Fracture Criteria

    International Nuclear Information System (INIS)

    Li Heng; Yang He; Zhan Mei

    2010-01-01

    Thin-walled tube bending(TWTB) method of Al-alloy tube has attracted wide applications in aerospace, aviation and automobile,etc. While, under in-plane double tensile stress states at the extrados of bending tube, the over-thinning induced ductile fracture is one dominant defect in Al-alloy tube bending. The main objective of this study is to predict the critical wall-thinning of Al-alloy tube bending by coupling two ductile fracture criteria(DFCs) into FE simulation. The DFCs include Continuum Damage Mechanics(CDM)-based model and GTN porous model. Through the uniaxial tensile test of the curved specimen, the basic material properties of the Al-alloy 5052O tube is obtained; via the inverse problem solution, the damage parameters of both the two fracture criteria are interatively determined. Thus the application study of the above DFCs in the TWTB is performed, and the more reasonable one is selected to obtain the critical thinning of Al-alloy tube in bending. The virtual damage initiation and evolution (when and where the ductile fracture occurs) in TWTB are investigated, and the fracture mechanisms of the voided Al-alloy tube in tube bending are consequently discussed.

  13. Performance evaluation of reactor operated zircaloy-2 pressure tubes of RAPS-1

    International Nuclear Information System (INIS)

    Chatterjee, S.; Ramadasan, E.; Balakrishnan, K.S.; Bahl, J.K.

    1992-01-01

    Detailed post irradiation examination was carried out on pressure tube sections from E-10, F-9 and F-10 locations of RAPS-1 after an in-reactor residence equivalent to 3.6 effective full power years. The F-10 pressure tube was studied in detail on sections obtained from one end to the other, whereas in the case of E-9 and F-9 pressure tubes only the end sections were examined. The studies carried out were visual examination, metallography, hydrogen i.e. H(D) analysis and mechanical testing at 300 C. Microstructural observations revealed uniform and random hydride/deuteride platelet distribution and absence of blisters or hydride segregation. The H(D) content in the F-10 pressure tube was found to vary in the range 6-12 ppm. The typical H(D) content in the three tubes was around 1 ppm. The H(D) pick-up evaluated from the observed oxide layer thickness was 8 ppm. Longitudinal tensile specimens fabricated from the F-10 pressure tube section and tested at 300 C exhibited increase in yield strength and tensile strength of 39% and 30% respectively. The residual uniform elongation was typically 1.8%. The observed changes in the tensile properties were found to be lower than those reported on unstressed specimens irradiated to similar neutron fluences. The observed hydrogen content and tensile properties obtained in F-10 pressure tube would not be detrimental under normal reactor operating conditions. (author). 10 refs., 4 figs., 2 tabs., 1 annexure

  14. 75 FR 69125 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China

    Science.gov (United States)

    2010-11-10

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

  15. Core construction in a pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto; Aoki, Katsutada.

    1975-01-01

    Object: To replace a centrally positioned fuel assembly of a fuel assembly unit with a reactor controlling machinery to decrease a distance between the fuel assemblies thereby saving use of heavy water and enhancing economy. Structure: A centrally positioned fuel assembly of a fuel assembly unit, which is composed of a plurality of fuel assemblies orderly arranged in lattice fashion, is replaced with a reactor controlling members such as control rods, poison tubes and the like to provide an arrangement of lattice-free type fuel assembly, thus reducing the pitch as small as possible. (Kamimura, M.)

  16. Advanced NDE (ANDE) and its application for pressure tube inspections in OPG reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jarron, D.; Trelinski, M.; Kretz, S. [Ontario Power Generation, Ajax, Ontario (Canada)]. E-mail: don.jarron@opg.com; mike.trelinski@opg.com; steve.kretz@opg.com

    2006-07-01

    Periodic and in-service inspections of CANDU fuel channels are essential for the proper assessment of the structural integrity of these vital components. The arrival of new delivery devices for fuel channel inspections (Universal Delivery Machine) has driven new methods for gathering and analyzing NDE data. The Advanced Non-Destructive Examination (ANDE) system has been designed and field implemented as a high speed data acquisition system to meet the requirements of the CSA N285.4 code. It was built from the solid foundation of CIGAR experience and uses cutting edge hardware and software to attain high speed data collection enabling relatively quick inspection of a large number of fuel channels. The capabilities of the ANDE inspection system include: Surface and volumetric inspection of pressure tube by ultrasonics; Flaw characterization by ultrasonics; Pressure tube diameter measurements; Pressure tube thickness measurements; Garter Spring location by Eddy Current; Garter Spring location by ultrasonics; Pressure tube sag measurement. In addition to the above, selected flaws/areas of a pressure tube can be replicated using a two plate ANDE replica tool. At the heart of the inspection system is a set of twelve ultrasonic probes positioned in such a way that the inspected areas are examined from various angles and directions and by various ultrasonic wave modes (shear and longitudinal). High frequency ultrasound used for the examinations allows for reliable detection of small flaws. Separate sensors have been installed on the inspection head for Garter Spring location and sag measurements. (author)

  17. Pressure tests to assess the significance of defects in boiler and superheater tubing

    International Nuclear Information System (INIS)

    Guest, J.C.; Hutchings, J.A.

    1975-01-01

    Internal pressure tests on 9 per cent Cr-1 per cent Mo steel tubing containing artificial defects demonstrated that the resultant loss of strength was less than a simple calculation based on the reduced tube thickness would suggest. Bursting tests on tubes containing longitudinal defects of varying length, depth and acuity showed notch strengthening at ambient temperature and at 550 0 C. A flow stress concept developed for simple bursting tests was shown to apply to creep conditions at 550 0 C. Results of creep and short-term bursting tests show that the length as well as the depth of the defect is an important factor affecting the life of bursting strength of the tubes. Defects less than 10 per cent of the tube thickness were found to have an insignificant effect. (author)

  18. Propagation of atmospheric-pressure ionization waves along the tapered tube

    Science.gov (United States)

    Xia, Yang; Wang, Wenchun; Liu, Dongping; Yan, Wen; Bi, Zhenhua; Ji, Longfei; Niu, Jinhai; Zhao, Yao

    2018-02-01

    Gas discharge in a small radius dielectric tube may result in atmospheric pressure plasma jets with high energy and density of electrons. In this study, the atmospheric pressure ionization waves (IWs) were generated inside a tapered tube. The propagation behaviors of IWs inside the tube were studied by using a spatially and temporally resolved optical detection system. Our measurements show that both the intensity and velocity of the IWs decrease dramatically when they propagate to the tapered region. After the taper, the velocity, intensity, and electron density of the IWs are improved with the tube inner diameter decreasing from 4.0 to 0.5 mm. Our analysis indicates that the local gas conductivity and surface charges may play a role in the propagation of the IWs under such a geometrical constraint, and the difference in the dynamics of the IWs after the taper can be related to the restriction in the size of IWs.

  19. The EL-4 reactor. Changing of a pressure tube on a test loop

    International Nuclear Information System (INIS)

    Foulquier, H.; Clara, P.

    1964-01-01

    Right from the beginning of the EL-4 project, the research convected with the overall design of the reactor was guided by the various technical specifications resulting from a justifiable concern about the reliability. The external and internal tubes of each layer situated in the reactor block had in particular to be interchangeable. The research alone into the dismantling of the external tube, i.e in fact the pressure tube, justified a certain number of full-scale tests on a model. The tests carried out under relevant conditions on a non-irradiated structure made it possible to define a complete ranger of of positioning and un-positioning sequences at a distance for such a pressure tube. (authors) [fr

  20. Natural convection in vertical tubes with variable properties and prescribed pressure at the end of the tube

    International Nuclear Information System (INIS)

    Almeida Rego, O.A. de; Fernandes, E.C.

    1983-01-01

    The analysis of free convection flow development in a heated vertical open tube was established in the present work for the air, with Prandtl number equal to 0.7 and for water with Prandtl numbers equal to 1.0, 2.5 and 5.0 with variable properties and prescribed pressure conditions at the end of the tube. It is considered that the flow is incompressible, laminar and stable and can be described by the continuity, momentum and energy equations with the usual boundary-layer assumptions. The equations were solved by finite difference method and from the velocity and temperature distributions many quantities such as dimensioless flow and heat rates and Nusselt numbers can be determined. (Author) [pt

  1. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Seitz, W.W.; Girshik, A. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1998-06-01

    In Phase 1 of this project, laboratory experiments were performed on a variety of developmental and commercial tubing alloys and claddings by exposing them to fireside corrosion tests which simulated a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, RA253MA, Fe{sub 3}Al + 5Cr, Ta-modified 310, NF 709, 690 clad, 671 clad, and 800HT for up to approximately 16,000 hours to the actual operating conditions of a 250-MW, coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, and controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy were exposed for 4,483, 11,348, and 15,883 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after the full 15,883 hours of exposure. A previous topical report has been issued for the 4,483 hours of exposure.

  2. 2D modeling of moderator flow and temperature distribution around a single channel after pressure tube/calandria tube contact

    International Nuclear Information System (INIS)

    Behdadi, A.; Luxat, J.C.

    2009-01-01

    A 2D computational fluid dynamics (CFD) model has been developed to calculate the moderator velocity field and temperature distribution around a single channel inside the moderator of a CANDU reactor after a postulated ballooning deformation of the pressure tube (PT) into contact with the calandria tube (CT). Following contact between the hot PT and the relatively cold CT, there is a spike in heat flux to the moderator surrounding the CT which may lead to sustained CT dryout. This can detrimentally affect channel integrity if the CT post-dryout temperature becomes sufficiently high to result in thermal creep strain deformation. The present research is focused on establishing the limits for dryout occurrence on the CTs for the situation in which pressure tube-calandria tube contact occurs. In order to consider different location of the channels inside the calandria, both upward and downward flow directions have been analyzed. The standard κ - ε turbulence model associated with logarithmic wall function is applied to predict the effects of turbulence. The governing equations are solved by the finite element software package COMSOL. The buoyancy driven natural convection on the outer surface of a CT has been analyzed to predict the flow and temperature distribution around the single CT considering the local moderator subcooling, wall temperature and heat flux. The model also shows the effect of high CT temperature on the flow and subcooling around the CTs at higher/lower elevation depending on the flow direction in the domain. According to the flow pattern and temperature distribution, it is predicted that stable film boiling generates in the stagnation region on the cylinder. (author)

  3. Effect of the surface film electric resistance on eddy current detectability of surface cracks in Alloy 600 tubes

    International Nuclear Information System (INIS)

    Saario, T.; Paine, J.P.N.

    1995-01-01

    The most widely used technique for NDE of steam generator tubing is eddy current. This technique can reliably detect cracks grown in sodium hydroxide environment only at depths greater than 50% through wall. However, cracking caused by thiosulphate solutions have been detected and sized at shallower depths. The disparity has been proposed to be caused by the different electric resistance of the crack wall surface films and corrosion products in the cracks formed in different environments. This work was undertaken to clarify the role of surface film electric resistance on the disparity found in eddy current detectability of surface cracks in alloy 600 tubes. The proposed model explaining the above mentioned disparity is the following. The detectability of tightly closed cracks by the eddy current technique depends on the electric resistance of the surface films of the crack walls. The nature and resistance of the films which form on the crack walls during operation depends on the composition of the solution inside the crack and close to the crack location. During cooling down of the steam generator, because of contraction and loss of internal pressurization, the cracks are rather tightly closed so that exchange of electrolyte and thus changes in the film properties become difficult. As a result, the surface condition prevailing at high temperature is preserved. If the environment is such that the films formed on the crack walls under operating conditions have low electric resistance, eddy current technique will fail to indicate these cracks or will underestimate the size of these cracks. However, if the electric resistance of the films is high, a tightly closed crack will resemble an open crack and will be easily indicated and correctly sized by eddy current technique

  4. Improvement of corrosion resistance of vanadium alloys in high-temperature pressurized water

    International Nuclear Information System (INIS)

    Fujiwara, Mitsuhiro; Sakamoto, Toshiya; Satou, Manabu; Hasegawa, Akira; Abe, Katsunori; Kaiuchi, Kazuo; Furuya, Takemi

    2005-01-01

    Corrosion tests in pressurized and vaporized water were conducted for V-based high Cr and Ti alloys and V-4Cr-4Ti type alloys containing minor elements such as Si, Al and Y. Weight losses were observed for every alloy after corrosion tests in pressurized water. It was apparent that addition of Cr effectively reduced the weight change in pressurized water. The weight loss of V-4Cr-4Ti type alloys in corrosion tests in vaporized water was also reduced as Cr content increased. The V-20Cr-4Ti alloy had a slight weight gain, almost same as that of SUS316, which had the best corrosion properties in the tested alloys. The elongation of alloys with in excess of 10% Cr was reduced as Cr content increased. The elongations of the V-12Cr-4Ti and the V-15Cr-4Ti alloys were significantly reduced by corrosion and cleavage fracture was observed reflecting hydrogen embrittlement. The reduced elongations of the alloys of the alloys were recovered to the same level of as annealed conditions after hydrogen degassing. After corrosion, the V-15Cr-4Ti-0.5Y alloy still kept enough elongation, suggesting that the addition of Y is effective to reduce the hydrogen embrittlement. (author)

  5. Expanded heat treatment to form residual compressive hoop stress on inner surface of zirconium alloy tubing

    International Nuclear Information System (INIS)

    Megata, Masao

    1997-01-01

    A specific heat treatment process that introduces hoop stress has been developed. This technique can produce zirconium alloy tubing with a residual compressive hoop stress near the inner surface by taking advantage of the mechanical anisotropy in hexagonal close-packed zirconium crystal. Since a crystal having its basal pole parallel to the tangential direction of the tubing is easier to exhibit plastic elongation under the hoop stress than that having its basal pole parallel to the radial direction, the plastic and elastic elongation can coexist under a certain set of temperature and hoop stress conditions. The mechanical anisotropy plays a role to extend the coexistent stress range. Thus, residual compressive hoop stress is formed at the inner surface where more plastic elongation occurs during the heat treatment. This process is referred to as expanded heat treatment. Since this is a fundamental crystallographic principle, it has various applications. The application to improve PCI/SCC (pellet cladding interaction/stress corrosion cracking) properties of water reactor fuel cladding is promising. Excellent results were obtained with laboratory-scale heat treatment and an out-reactor iodine SCC test. These results included an extension of the time to SCC failure. (author)

  6. Changes in the design, fabrication and setting of guide tube support pins in alloy X750

    International Nuclear Information System (INIS)

    Benhamou, C.; Chambrin, J.L.; Todeschini, P.; Champredonde, J.; Lemaire, E.

    2004-01-01

    As a consequence of a problem of stress corrosion cracking (SCC) encountered on guide tube support pins (GTSP) of first generation (1982) and of second generation (1987), EDF and Framatome decided in mars 1988 to launch an important program involving a complete overhaul of the design, the material used, the fabrication and the setting in reactor of GTSP. This program has led to the implementation in 900 MWe and 1300 MWe PWR of a new tube guide support pin called NG89. This implementation began in 1989, now 15 years later, 40% of the operating GTSP in 900 MWe and 1300 MWe PWR are of NG89 type, the oldest ones cumulate 105000 hours in service without negative feedback experience. The main features of the NG89 is: - to be made from an alloy X-750 containing boron (from 25 to 45 ppm) - to have a SCC threshold set at 720 MPa - to be machined from metal bars completely treated, - to have a rolling of the fillets, and - to undergo a shot blasting on the zones of the surface the most acted upon. (A.C.)

  7. Final report on development evaluation of Task Group 3 pressure tubes

    International Nuclear Information System (INIS)

    Fleck, R.G.; Price, E.G.; Cheadle, B.A.

    1983-11-01

    This report describes the production and evaluation of pressure tubes manufactured to the recommendations of Task Group 3 (TG3) of the Creep Engineering Design Plan. The Zr-2.5 wt percent Nb tubes were manufactured by modified production route to change their metallurgical structure and so reduce the in-service elongation rates. Three modified routes were investigated and a total of twenty-eight tubes produced. There were no difficulties in manufacture and the tubes satisfied the quality assurance and design specifications of reactor grade tubes. Metallurgical evaluation showed that the expected changes in microstructure had occurred but not to the extent anticipated. The TG3 tubes were found to have comparable properties to current tubes when tested for: tensile strength (irradiated and unirradiated); hydride cracking; stress to reorient hydrides; hydrogen diffusion; flaw tolerance; corrosion (irradiated and unirradiated); wear; rolled joint characteristics; irradiation creep and growth. Lower in-service elongation rates are expected for tubes produced by two of the modified routes

  8. Transfer of a cold atmospheric pressure plasma jet through a long flexible plastic tube

    International Nuclear Information System (INIS)

    Kostov, Konstantin G; Prysiazhnyi, Vadym; Honda, Roberto Y; Machida, Munemasa

    2015-01-01

    This work proposes an experimental configuration for the generation of a cold atmospheric pressure plasma jet at the downstream end of a long flexible plastic tube. The device consists of a cylindrical dielectric chamber where an insulated metal rod that serves as high-voltage electrode is inserted. The chamber is connected to a long (up to 4 m) commercial flexible plastic tube, equipped with a thin floating Cu wire. The wire penetrates a few mm inside the discharge chamber, passes freely (with no special support) along the plastic tube and terminates a few millimeters before the tube end. The system is flushed with Ar and the dielectric barrier discharge (DBD) is ignited inside the dielectric chamber by a low frequency ac power supply. The gas flow is guided by the plastic tube while the metal wire, when in contact with the plasma inside the DBD reactor, acquires plasma potential. There is no discharge inside the plastic tube, however an Ar plasma jet can be extracted from the downstream tube end. The jet obtained by this method is cold enough to be put in direct contact with human skin without an electric shock. Therefore, by using this approach an Ar plasma jet can be generated at the tip of a long plastic tube far from the high-voltage discharge region, which provides the safe operation conditions and device flexibility required for medical treatment. (paper)

  9. Tensile and fracture toughness characteristics of Zr-2.5Nb pressure tube

    International Nuclear Information System (INIS)

    Jung, H. C.; Kim, Y. S.; Ahn, S. B.; Kim, S. S.; Im, K. S.

    2004-01-01

    The object of this study is to evaluate the characteristics of tensile and fracture toughness of Zr-2.5Nb pressure tube. The transverse tensile tests were performed at various temperatures and the fracture toughness tests were carried out at room temperature using the CCT (curved compact tension) specimen. These specimens were directly machined from the pressure tube retaining original curvatures. Also, the fracture toughness of two sets of Zr-2.5Nb manufactured at different time was compared. The chemical analysis and the Vicker's hardness tests were performed at two sets of Zr-2.5Nb pressure tube. The Vicker's hardness value of SET-2 containing more oxygen and carbon relatively was higher about 11 than that of SET-1

  10. Application of the VAW tube digester for metallurgical pressure-leaching processes

    International Nuclear Information System (INIS)

    Kaempf, F.; Pietsch, H.B.

    1978-01-01

    Problems associated with the treatment of complex and refractory ores or concentrates, as well as those related to environmental factors, have led to increased interest in hydrometallurgy under elevated temperatures and pressures. Pressure leaching can be carried out in vertical, horizontal or spherical autoclaves equipped with mechanical agitators. If high throughput capacities are catered for, the division of a conventional plant into several units is inevitable. By contrast, the VAW (Vereinigte Aluminium-Werke Aktiengesellschaft) tube digester enables hydrometallurgical processes to be carried out under pressure and at a high temperature with the use of a basically simple technology, extremely high specific throughput and improved thermal economics being achieved. The advantages of the tube digester over vessel autoclaves are described, and details of laboratory investigations into the applicability of tube digesters to various metallurgical applications are given. Test results are given for the leaching of refractory uranium ores. (author)

  11. Fabrication of Zr-2.5Nb pressure tubes to minimise the harmful effects of trace elements

    International Nuclear Information System (INIS)

    Theaker, J.R.; Coleman, C.E.; Davis, L.; Graham, R.A.

    1994-03-01

    Trace elements can reduce the fracture resistance of Zr-2.5Nb pressure tubes. The effects of hydrogen as hydrides and oxygen as an alloy-strengthening agent are well known, but the contribution of carbon, phosphorus, chlorine and segregated oxygen has only recently been recognized. Carbides and phosphides are brittle particles, while chlorine segregates to form planes of weakness that produce fissures is associated with low toughness. With long hold times in the (α + β) region, oxygen partitions in the α-grains; such grains are hard, and if they survive fabrication may reduce the toughness of the finished tube. Through a co-operative program involving AECL and manufacturers, a series of manufacturing innovations and controls has been introduced that minimize these harmful effects. Hydrogen is present in Zr sponge as water, can be absorbed at each stage of tube fabrication, and needs to be carefully controlled, particularly during ingot breakdown and subsequent forging. Hydrogen concentrations in finished tubes have been reduced by a factor of three through the optimization of manufacturing processes and the implementation of new technology. Multiple vacuum arc melting, use of selected raw materials and intermediate ingot surface conditioning have resulted in much improved fracture toughness through the reduction of chlorine and phosphorus concentrations. Optimum distribution of oxygen may be achieved through changes to the extrusion process cycle. An understanding of the Zr-2.5Nb-C phase diagram, particularly the solubility of carbon at low concentrations, has resulted in the specification of a lower carbon concentration. (author). 12 refs., 6 tabs., 10 figs

  12. Long-term creep rupture strength of weldment of Fe-Ni based alloy as candidate tube and pipe for advanced USC boilers

    Energy Technology Data Exchange (ETDEWEB)

    Bao, Gang; Sato, Takashi [Babcok-Hitachi K.K., Hiroshima (Japan). Kure Research Laboratory; Marumoto, Yoshihide [Babcok-Hitachi K.K., Hiroshima (Japan). Kure Div.

    2010-07-01

    A lot of works have been going to develop 700C USC power plant in Europe and Japan. High strength Ni based alloys such as Alloy 617, Alloy 740 and Alloy 263 were the candidates for boiler tube and pipe in Europe, and Fe-Ni based alloy HR6W (45Ni-24Fe-23Cr-7W-Ti) is also a candidate for tube and pipe in Japan. One of the Key issues to achieve 700 C boilers is the welding process of these alloys. Authors investigated the weldability and the long-term creep rupture strength of HR6W tube. The weldments were investigated metallurgically to find proper welding procedure and creep rupture tests are ongoing exceed 38,000 hours. The long-term creep rupture strengths of the HST weld joints are similar to those of parent metals and integrity of the weldments was confirmed based on with other mechanical testing results. (orig.)

  13. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1

    International Nuclear Information System (INIS)

    2006-05-01

    Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

  14. Characterizing the active opening of the eustachian tube in a hypobaric/hyperbaric pressure chamber.

    Science.gov (United States)

    Mikolajczak, Stefanie; Meyer, Moritz Friedo; Hahn, Moritz; Korthäuer, Christine; Jumah, Masen Dirk; Hüttenbrink, Karl-Bernd; Grosheva, Maria; Luers, Jan Christoffer; Beutner, Dirk

    2015-01-01

    Active and passive opening of the Eustachian tube (ET) enables direct aeration of the middle ear and a pressure balance between middle ear and the ambient pressure. The aim of this study was to characterize standard values for the opening pressure (ETOP), the opening frequency (ETOF), and the opening duration (ETOD) for active tubal openings (Valsalva maneuver, swallowing) in healthy participants. In a hypobaric/hyperbaric pressure chamber, 30 healthy participants (19 women, 11 men; mean age, 25.57 ± 3.33 years) were exposed to a standardized profile of compression and decompression. The pressure values were recorded via continuous impedance measurement during the Valsalva maneuver and swallowing. Based on the data, standard curves were identified and the ETOP, ETOD, and ETOF were determined. Recurring patterns of the pressure curve during active tube opening for the Valsalva maneuver and for active swallowing were characterized. The mean value for the Valsalva maneuver for ETOP was 41.21 ± 17.38 mbar; for the ETOD, it was 2.65 ± 1.87 seconds. In the active pressure compensation by swallowing, the mean value for the ETOP was 29.91 ± 13.07 mbar; and for the ETOD, it was 0.82 ± 0.53 seconds. Standard values for the opening pressure of the tube and the tube opening duration for active tubal openings (Valsalva maneuver, swallowing) were described, and typical curve gradients for healthy subjects could be shown. This is another step toward analyzing the function of the tube in compression and decompression.

  15. Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

    International Nuclear Information System (INIS)

    Agostini, F.

    2008-07-01

    The complexity of the two-phase flow in a tube bundle presents important problems in the design and understanding of the physical phenomena taking place. The working conditions of an evaporator depend largely on the dynamics of the two-phase flow that in turn influence the heat exchange and the pressure drop of the system. A characterization of the flow dynamics, and possibly the identification of the flow pattern in the tube bundle, is thus expected to lead to a better understanding of the phenomena and to reveal on the mechanisms governing the tube bundle. Therefore, the present study aims at providing further insights into two-phase bundle flow through a new visualization system able to provide for the first time a view of the flow in the core of a tube bundle. In addition, the measurement of the light attenuation of a laser beam through the two-phase flow and measurement of the high frequency pressure fluctuations with a piezo-electric pressure transducer are used to characterize the flow. The design and the validation of this new instrumentation also provided a method for the detection of dry-out in tube bundles. This was achieved by a laser attenuation technique, flow visualization, and estimation of the power spectrum of the pressure fluctuation. The current investigation includes results for two different refrigerants, R134a and R236fa, three saturations temperatures T sat = 5, 10 and 15 °C, mass velocities ranging from 4 to 40 kg/sm² in adiabatic and diabatic conditions (several heat fluxes). Measurement of the local heat transfer coefficient and two-phase frictional pressure drop were obtained and utilized to improve the current prediction methods. The heat transfer and pressure drop data were supported by extensive characterization of the two-phase flow, which was to improve the understanding of the two-phase flow occurring in tube bundles. (author)

  16. Pressure tube replacement in Pickering NGS A units 1 and 2

    International Nuclear Information System (INIS)

    Irvine, H.S.; Bennett, E.J.; Talbot, K.H.

    1986-10-01

    Being able to technically and economically replace the most radioactive components (excluding the nuclear fuel) in operating reactors will help to ensure the ongoing acceptance of nuclear power as a viable energy source for the future. Ontario Hydro is well along the path to meeting the above objective for its CANDU-PHW reactors. Following the failure of a Zircaloy-II pressure tube in unit 2 of Pickering NGS A in August, 1983, Ontario Hydro has embarked on a program to replace all Zircaloy-II pressure tubes in units 1 and 2 at Pickering. This program integrates the in-house research, design, construction, and operating skills of a large utility (Ontario Hydro) with the skills of a national nuclear organization (Atomic Energy of Canada Limited) and the private engineering sector of the Canadian nuclear industry. The paper describes the background to the pressure tube failure in Pickering unit 2 and to the efforts incurred in understanding the failure mechanism and how similar failures are not expected for the zirconium-niobium pressure tube material used in all other large CANDU-PHW units after units 1 and 2 of Pickering NGS A. The tooling developed for the pressure tube replacement program is described as well as the organization to undertake the program in an operating nuclear station. The retubing of units 1 and 2 at Pickering NGS A is nearing a successful completion and shows the benefits of being able to integrate the various skills required for this success. Pressure tube replacement in a CANDU-PHW reactor is equivalent to replacement of the reactor vessel in a LWR. The fact that this replacement can be done economically and with acceptable radiation dose to workers augurs well for the continued viability of the use of nuclear energy for the benefit of mankind. (author)

  17. Flooding of a large, passive, pressure-tube light water reactor

    International Nuclear Information System (INIS)

    Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

    1997-01-01

    A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

  18. A pulsed eddy current probe for inspection of support plates from within Alloy-800 steam generator tubes

    International Nuclear Information System (INIS)

    Krause, T. W.; Babbar, V. K.; Underhill, P. R.

    2014-01-01

    Support plate degradation and fouling in nuclear steam generators (SGs) can lead to SG tube corrosion and loss of efficiency. Inspection and monitoring of these conditions can be integrated with preventive maintenance programs, thereby advancing station-life management processes. A prototype pulsed eddy current (PEC) probe, targeting inspection issues associated with SG tubes in SS410 tube support plate structures, has been developed using commercial finite element (FE) software. FE modeling was used to identify appropriate driver and pickup coil configurations for optimum sensitivity to changes in gap and offset for Alloy-800 SG tubes passing through 25 mm thick SS410 support plates. Experimental measurements using a probe that was manufactured based on the modeled configuration, were used to confirm the sensitivity of differential PEC signals to changes in relative position of the tube within the tube support plate holes. Models investigated the effect of shift and tilt of tube with respect to hole centers. Near hole centers and for small shifts, modeled signal amplitudes from the differentially connected coil pairs were observed to change linearly with tube shift. This was in agreement with experimentally measured TEC coil response. The work paves the way for development of a system targeting the inspection and evaluation of support plate structures in steam generators

  19. A pulsed eddy current probe for inspection of support plates from within Alloy-800 steam generator tubes

    Science.gov (United States)

    Krause, T. W.; Babbar, V. K.; Underhill, P. R.

    2014-02-01

    Support plate degradation and fouling in nuclear steam generators (SGs) can lead to SG tube corrosion and loss of efficiency. Inspection and monitoring of these conditions can be integrated with preventive maintenance programs, thereby advancing station-life management processes. A prototype pulsed eddy current (PEC) probe, targeting inspection issues associated with SG tubes in SS410 tube support plate structures, has been developed using commercial finite element (FE) software. FE modeling was used to identify appropriate driver and pickup coil configurations for optimum sensitivity to changes in gap and offset for Alloy-800 SG tubes passing through 25 mm thick SS410 support plates. Experimental measurements using a probe that was manufactured based on the modeled configuration, were used to confirm the sensitivity of differential PEC signals to changes in relative position of the tube within the tube support plate holes. Models investigated the effect of shift and tilt of tube with respect to hole centers. Near hole centers and for small shifts, modeled signal amplitudes from the differentially connected coil pairs were observed to change linearly with tube shift. This was in agreement with experimentally measured TEC coil response. The work paves the way for development of a system targeting the inspection and evaluation of support plate structures in steam generators.

  20. Development of Evaluation Technology of the Integrity of HWR Pressure Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Y. M.; Kim, Y. S.; Im, K. S.; Kim, K. S.; Ahn, S. B

    2007-06-15

    Zr-2.5Nb pressure tubes are one of the most critical structural components governing the lifetime of the heavy water reactors to carry fuel bundles and heavy coolant water inside. Since they are being degraded during their operation in reactors due to dimensional changes caused by creep and irradiation growth, neutron irradiation and delayed hydride cracking, it is required to evaluate their degradation by conducting material testing and examinations on the highly irradiated pressure tubes in hot cells and to keep tracking of their degradation behavior with operation time, which are the aim of this project.

  1. Prediction of pressure tube fretting-wear damage due to fuel vibration

    International Nuclear Information System (INIS)

    Yetisir, M.; Fisher, N.J.

    1997-01-01

    Fretting marks between fuel bundle bearing pads and pressure tubes have been observed at the inlet end of some Darlington Nuclear Generating Station (NGS) and Bruce NGS fuel channels. The excitation mechanisms that lead to fretting are not fully understood. In this paper, the possibility of bearing pad-to-pressure tube fretting due to turbulence-induced motion of the fuel element is investigated. Numerical simulations indicate that this mechanism by itself is not likely to cause the level of fretting experienced in Darlington and Bruce NGSs. (orig.)

  2. Characteristics of CANDU fuel bundles that caused pressure tube fretting at the bundle midplane

    Energy Technology Data Exchange (ETDEWEB)

    Dennier, D; Manzer, A M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Koehn, E [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    Detailed measurements on new bundles, and those that caused fretting during in- and out-reactor tests, have given insight into the factors responsible for fretting at the midplane of the inlet bundle. Bottom fuel elements that were attached near radial endplate spokes and had inboard bearing pads in the rolled joint cavity produced a significant portion of the observed fret marks. These elements are influenced by several driving forces that deflect the centre bearing pads towards the pressure tube surface. The evidence suggests that slight changes in bundle design may be possible to reduce pressure tube fretting. (author). 4 refs., 3 tabs., 8 figs.

  3. Prediction of pressure tube fretting-wear damage due to fuel vibration

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M; Fisher, N J [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    Fretting marks between fuel bundle bearing pads and pressure tubes have been observed at the inlet end of some Darlington NGS (nuclear generating station) and Bruce NGS fuel channels. The excitation mechanisms that lead to fretting are not fully understood. In this paper, the possibility of bearing pad-to-pressure tube fretting due to turbulence-induced motion of the fuel element is investigated. Numerical simulations indicate that this mechanism by itself is not likely to cause the level of fretting experienced in Darlington and Bruce NGS`s (nuclear generating stations). (author). 12 refs., 2 tabs., 11 figs.

  4. 77 FR 67336 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania...

    Science.gov (United States)

    2012-11-09

    ... Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania: Final Results of Antidumping... alloy seamless standard, line and pressure pipe from Romania. The period of review is August 1, 2010..., line and pressure pipe from Romania. See Certain Small Diameter Carbon and Alloy Seamless Standard...

  5. Influence of deformation conditions on texture formation and ductility in titanium alloys under hydrostatic pressure

    International Nuclear Information System (INIS)

    Dekun, A.M.; Kushakevich, S.A.; Adamesku, R.A.; Khmelinin, Yu.F.; Beresnev, B.I.; Shishmintsev, V.F.

    1982-01-01

    The influence of hot pressing parameters on microstructure, texture and mechanical properties of bars from titanium alloys VT1-0, VT5-1, (α-alloys) and VT3-1 (α+ν-alloy) has been investigated. Mechanical testing of samples has been performed under hydrostatic pressure from 200 to 800 MPa. It is shown that the temperature, deformation degree and type of the structure obtained exert a slight effect on mechanical properties of bars. The texture heterogeneity is more pronounced in α-alloys. It has been found that hydrostatic pressure during sample tensile testing improves their ductility characteristics

  6. Influence of pressure on the solid state phase transformation of Cu–Al–Bi alloy

    International Nuclear Information System (INIS)

    Gong, Li; Jian-Hua, Liu; Wen-Kui, Wang; Ri-Ping, Liu

    2010-01-01

    The solid state phase transformation of Cu-Al-Bi alloy under high pressure was investigated by x-ray diffraction, energy dispersive spectroscopy and transmission electron microscopy. Experimental results show that the initial crystalline phase in the Cu-Al-Bi alloy annealed at 750 °C under the pressures in the range of 0–6 GPa is α-Cu solid solution (named as α-Cu phase below), and high pressure has a great influence on the crystallisation process of the Cu-Al-Bi alloy. The grain size of the α-Cu phase decreases with increasing pressure as the pressure is below about 3 GPa, and then increases (P > 3 GPa). The mechanism for the effects of high pressure on the crystallisation process of the alloy has been discussed. (condensed matter: structure, thermal and mechanical properties)

  7. Mechanistic modeling of heat transfer process governing pressure tube-to-calandria tube contact and fuel channel failure

    International Nuclear Information System (INIS)

    Luxat, J.C.

    2002-01-01

    Heat transfer behaviour and phenomena associated with ballooning deformation of a pressure tube into contact with a calandria tube have been analyzed and mechanistic models have been developed to describe the heat transfer and thermal-mechanical processes. These mechanistic models are applied to analyze experiments performed in various COG funded Contact Boiling Test series. Particular attention is given in the modeling to characterization of the conditions for which fuel channel failure may occur. Mechanistic models describing the governing heat transfer and thermal-mechanical processes are presented. The technical basis for characterizing parameters of the models from the general heat transfer literature is described. The validity of the models is demonstrated by comparison with experimental data. Fuel channel integrity criteria are proposed which are based upon three necessary and sequential mechanisms: Onset of CHF and local drypatch formation at contact; sustained film boiling in the post-contact period; and creep strain to failure of the calandria tube while in sustained film boiling. (author)

  8. Characterization of mechanical properties of hydroxyapatite-silicon-multi walled carbon nano tubes composite coatings synthesized by EPD on NiTi alloys for biomedical application.

    Science.gov (United States)

    Khalili, Vida; Khalil-Allafi, Jafar; Sengstock, Christina; Motemani, Yahya; Paulsen, Alexander; Frenzel, Jan; Eggeler, Gunther; Köller, Manfred

    2016-06-01

    Release of Ni(1+) ions from NiTi alloy into tissue environment, biological response on the surface of NiTi and the allergic reaction of atopic people towards Ni are challengeable issues for biomedical application. In this study, composite coatings of hydroxyapatite-silicon multi walled carbon nano-tubes with 20wt% Silicon and 1wt% multi walled carbon nano-tubes of HA were deposited on a NiTi substrate using electrophoretic methods. The SEM images of coated samples exhibit a continuous and compact morphology for hydroxyapatite-silicon and hydroxyapatite-silicon-multi walled carbon nano-tubes coatings. Nano-indentation analysis on different locations of coatings represents the highest elastic modulus (45.8GPa) for HA-Si-MWCNTs which is between the elastic modulus of NiTi substrate (66.5GPa) and bone tissue (≈30GPa). This results in decrease of stress gradient on coating-substrate-bone interfaces during performance. The results of nano-scratch analysis show the highest critical distance of delamination (2.5mm) and normal load before failure (837mN) as well as highest critical contact pressure for hydroxyapatite-silicon-multi walled carbon nano-tubes coating. The cell culture results show that human mesenchymal stem cells are able to adhere and proliferate on the pure hydroxyapatite and composite coatings. The presence of both silicon and multi walled carbon nano-tubes (CS3) in the hydroxyapatite coating induce more adherence of viable human mesenchymal stem cells in contrast to the HA coated samples with only silicon (CS2). These results make hydroxyapatite-silicon-multi walled carbon nano-tubes a promising composite coating for future bone implant application. Copyright © 2016 Elsevier Ltd. All rights reserved.

  9. Automated Control of Endotracheal Tube Cuff Pressure during Simulated Flight

    Science.gov (United States)

    2016-06-21

    accomplished in the intensive care unit (ICU) with stand-alone devices as well as those integral to a ventilator [13,14]. We hypothesized that closed loop ... Administration approved automatic cuff pressure adjustment devices (Intellicuff, Hamilton Medical , Reno, NV; Pyton, ARM Medical , Bristol, CT; Cuff Sentry, Outcome...711th Human Performance Wing U.S. Air Force School of Aerospace Medicine Int’l Expeditionary Educ & Training Dept Air Force Expeditionary Medical

  10. Superelastic NiTi memory alloy micro-tube under tension - nucleation and propagation of martensite band

    International Nuclear Information System (INIS)

    Li, Z.Q.; Sun, Q.P.

    2000-01-01

    The superelastic behavior of polycrystalline NiTi shape memory alloy micro-tube under tension is studied experimentally. The nominal stress-strain curve of the micro-tube is recorded. By using a special surface coating it is found that the deformation of the tube is via the nucleation and propagation of stress-induced martensite band. The experiments show that the martensite nucleates in the form of a spiral lens-shaped narrow band that is inclined at 61 to the axis of loading when the stress reaches the peak of stress-strain curve. The width and the length of the band grew gradually with increase of loading and finally joined and merged into a single band. The subsequent deformation of the tube is realized by the propagation of this cylindrical martensite band. (orig.)

  11. The Sensitivity Analysis of Axial Pressure Tube Creep Profile for Dryout Power in PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Euiseung; Kim, Youngae [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Stern Laboratory performed the CHF tests with only one axial pressure tube creep profile per 3.3%, 5.1% peak crept channel and made CHF correlation including creep factor from the CHF test results. Wolsong nuclear power plants also have utilized the same CHF correlation derived by CNL. Pressure tube diameter creep rate is function of fast neutron, coolant temperature, and coolant pressure in a channel. It means that various axial pressure tube creep profiles exist in PHWR due to the history of operating conditions. Usually, CHF correlation is used during ROP(Regional Overpower Protection) Trip Setpoint Analysis or Safety Analysis in PHWR. The sensitivity analysis for CHF effects using various creep profiles is needed. This paper summarizes the comparison results of dryout power between CHF test creep profile and estimated creep profiles of Wolsong units. The effect of axial pressure tube creep profile for dryout power in fuel channel is evaluated by using Stern Lab. CHF test creep profile and 380 channel creep profiles of Wolsong. The dryout powers at 3.3% and 5.1% test conditions are slightly smaller when using 380 Wolsong channels creep profiles. These also show that the simulated dryout powers maintain consistency regardless of flow conditions.

  12. Phase martensitic transformation study in mechanically alloyed Ti{sub 50}Ni{sub 25}Fe{sub 25} alloy via high pressure

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Joao Cardoso de; Ferreira, Ailton da Silva, E-mail: joao.cardoso.lima@ufsc.br [Universidade Federal de Santa Catarina (UFSC), Florianopolis (Brazil); Rovani, Pablo Roberto; Pereira, Altair Soria [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre (Brazil)

    2016-07-01

    Full text: Alloys based on titanium and nickel with shape memory effect (SME) have been widely investigated due to potential use in different areas of science and technology, such as electronics, medicine, and space.1 Among them, the superalloys Ti-Ni-Fe show high corrosion resistance and good mechanical properties even at high temperatures that make them suitable for use in applications such as power plant components that work under aggressive conditions. At room temperature, the TiNi alloy has a monoclinic (B19'), known as the martensitic phase. With increasing temperature, the B19' phase transforms into a trigonal/hexagonal (B19) phase, known as the R- or pre martensitic phase, which, in its turn, transforms into a cubic (B2) structure, known as the austenitic phase. On cooling to room temperature, the reverse B2→B19→B19' phase transformations are observed. Since the B19↔B19' transformation occurs at a temperature low enough to inhibit diffusion-controlled processes, it belongs to a class of diffusionless phase transformations known as martensitic transformations. For this study, a Ti{sub 50}Ni{sub 25}Fe{sub 25} (B2) alloy was prepared by mechanical alloying, and the effects of high pressures up to 18 GPa will be presented. The structural changes with increasing pressure were followed by recording in situ angle-dispersive X-ray diffraction (ADXRD) diffractograms, in transmission geometry, using a long fine focus Mo X-ray tube and an imaging plate detector. The obtained results were already reported in Ref [1]. (1) A. S. Ferreira, P. R. Rovani, J. C. de Lima, A. S. Pereira, J. Appl. Phys. 117 (2015). (author)

  13. Conceptual design of a pressure tube light water reactor with variable moderator control

    International Nuclear Information System (INIS)

    Rachamin, R.; Fridman, E.; Galperin, A.

    2012-01-01

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  14. Hierarchically structured nanoporous carbon tubes for high pressure carbon dioxide adsorption

    Directory of Open Access Journals (Sweden)

    Julia Patzsch

    2017-05-01

    Full Text Available Mesoscopic, nanoporous carbon tubes were synthesized by a combination of the Stoeber process and the use of electrospun macrosized polystyrene fibres as structure directing templates. The obtained carbon tubes have a macroporous nature characterized by a thick wall structure and a high specific surface area of approximately 500 m²/g resulting from their micro- and mesopores. The micropore regime of the carbon tubes is composed of turbostratic graphitic areas observed in the microstructure. The employed templating process was also used for the synthesis of silicon carbide tubes. The characterization of all porous materials was performed by nitrogen adsorption at 77 K, Raman spectroscopy, infrared spectroscopy, thermal gravimetric analysis (TGA, scanning electron microscopy (SEM as well as transmission electron microscopy (TEM. The adsorption of carbon dioxide on the carbon tubes at 25 °C at pressures of up to 30 bar was studied using a volumetric method. At 26 bar, an adsorption capacity of 4.9 mmol/g was observed. This is comparable to the adsorption capacity of molecular sieves and vertically aligned carbon nanotubes. The high pressure adsorption process of CO2 was found to irreversibly change the microporous structure of the carbon tubes.

  15. Evidence of new high-pressure magnetic phases in Fe-Pt Invar alloy

    International Nuclear Information System (INIS)

    Matsushita, M.; Endo, S.; Miura, K.; Ono, F.

    2003-01-01

    To investigate the magnetic properties of disordered Fe 70 Pt 30 Invar alloy under high pressure, measurements of the real part of the AC susceptibility (χ) were made under pressure up to 7.5 GPa in the temperature range 4.2-385 K using a cubic anvil high-pressure apparatus. The Curie temperature (T C ) decreased with increasing pressure, and then, two new high-pressure magnetic phases appeared. These results show that the ferromagnetism of Fe-Pt Invar alloy becomes weaker, and the antiferromagnetic interaction becomes dominant with increasing pressure

  16. Tracheal tube and laryngeal mask cuff pressure during anaesthesia - mandatory monitoring is in need

    DEFF Research Database (Denmark)

    Rokamp, K.Z.; Secher, N.H.; Møller, Ann

    2010-01-01

    ABSTRACT: BACKGROUND: To prevent endothelium and nerve lesions, tracheal tube and laryngeal mask cuff pressure is to be maintained at a low level and yet be high enough to secure air sealing. METHOD: In a prospective quality-control study, 201 patients undergoing surgery during anaesthesia (without...... the use of nitrous oxide) were included for determination of the cuff pressure of the tracheal tubes and laryngeal masks. RESULTS: In the 119 patients provided with a tracheal tube, the median cuff pressure was 30 (range 8 - 100) cm H2O and the pressure exceeded 30 cm H2O (upper recommended level) for 54...... patients. In the 82 patients provided with a laryngeal mask, the cuff pressure was 95 (10 - 121) cm H2O and above 60 cm H2O (upper recommended level) for 56 patients and in 34 of these patients, the pressure exceeded the upper cuff gauge limit (120 cm H2O). There was no association between cuff pressure...

  17. Manufacture of thin-walled clad tubes by pressure welding of roll bonded sheets

    Science.gov (United States)

    Schmidt, Hans Christian; Grydin, Olexandr; Stolbchenko, Mykhailo; Homberg, Werner; Schaper, Mirko

    2017-10-01

    Clad tubes are commonly manufactured by fusion welding of roll bonded metal sheets or, mechanically, by hydroforming. In this work, a new approach towards the manufacture of thin-walled tubes with an outer diameter to wall thickness ratio of about 12 is investigated, involving the pressure welding of hot roll bonded aluminium-steel strips. By preparing non-welded edges during the roll bonding process, the strips can be zip-folded and (cold) pressure welded together. This process routine could be used to manufacture clad tubes in a continuous process. In order to investigate the process, sample tube sections with a wall thickness of 2.1 mm were manufactured by U-and O-bending from hot roll bonded aluminium-stainless steel strips. The forming and welding were carried out in a temperature range between RT and 400°C. It was found that, with the given geometry, a pressure weld is established at temperatures starting above 100°C. The tensile tests yield a maximum bond strength at 340°C. Micrograph images show a consistent weld of the aluminium layer over the whole tube section.

  18. Heat transfer and pressure drop in a tube bank inclined with respect to the flow

    Energy Technology Data Exchange (ETDEWEB)

    Yanez Moreno, A.A.

    1985-01-01

    This research is intended to lend understanding and to quantify the heat-transfer and fluid-flow characteristics for yawed tube banks in both staggered and in-line arrays. The investigated range of yaw angle was from 90 (crossflow) to 45/sup 0/, while the freestream Reynolds number (based on the tube diameter) ranged between 7000 and 45,000. The transverse and longitudinal center-to-center distances between the tubes were S/sub T//D = S/sub L//D = 2, respectively. The heat-transfer experiments were carried out on a row-by-row basis. Pressure drop measurements were made not only upstream and downstream of the tube bank but also within it. The patterns of fluid flow adjacent to the tubes were visualized using the oil-lampblack technique. A detailed study was carried out to determine the heat-transfer characteristics of a yawed single cylinder. The yaw angle range was between 90 and 30/sup 0/, and flow visualization was also performed. The pressure measurements showed that the overall dimensionless pressure drop for the staggered array is higher than that for the in-line array for a given Reynolds number or yaw. The flow-visualization patterns showed that the boundary layer separation depends on the yaw angle. For the single cylinder, the Nusselt number varied with the yaw angle in an undulating manner and did not correlate with the Independence Principle.

  19. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1987-01-01

    This patent describes a pressurized water reactor steam generator having spaced rows of heat transfer tubes through which primary coolant from the reactor flows, the tubes being of a U-shaped design, with the U-bend portions of the U-shaped tubes stabilized by antivibration bars. The improvement described here comprises expandable antivibration bars for stabilizing the U-bend portions of the U-shaped tubes, the expandable bars having a pair of adjustable rods, formed from a pair of rod sections affixed to a connector, one rod section of each of the pair of rod sections having a plurality of protrusions. Each of the protrusions has slidable surfaces thereon. The other rod section of each of the pair of rod sections has indentations, each of the indentations having slidable surfaces thereon complementary to the sliding surfaces of the protrusions, such that the rods are expandable from a first cross-sectional width less than the spacing between two adjacent rows of the tubes, to a second cross-sectional width greater than the first cross-sectional width. The expanded rods are adapted to contact tubes of the two adjacent rows of the tubes

  20. Oxide Dispersion Strengthened Fe(sub 3)Al-Based Alloy Tubes: Application Specific Development for the Power Generation Industry

    Energy Technology Data Exchange (ETDEWEB)

    Kad, B.K.

    1999-07-01

    A detailed and comprehensive research and development methodology is being prescribed to produce Oxide Dispersion Strengthened (ODS)-Fe3Al thin walled tubes, using powder extrusion methodologies, for eventual use at operating temperatures of up to 1100C in the power generation industry. A particular 'in service application' anomaly of Fe3Al-based alloys is that the environmental resistance is maintained up to 1200C, well beyond where such alloys retain sufficient mechanical strength. Grain boundary creep processes at such high temperatures are anticipated to be the dominant failure mechanism.

  1. Experimental and visual study on flow patterns and pressure drops in U-tubes

    International Nuclear Information System (INIS)

    Da Silva Lima, J. R.

    2011-01-01

    In single- and two-phase flow heat exchangers (in particular 'coils'), besides the straight tubes there are also many singularities, in particular the 180° return bends (also called return bends or U-bends). However, contrary to the literature concerning pressure drops and heat transfer in straight tubes, where many experimental data and predicting methods are available, only a limited number of studies concerning U-bends can be found. Neither reliable experimental data nor proven prediction methods are available. Indeed, flow structure, pressure drop and heat transfer in U-bends are an old unresolved design problem in the heat transfer industry. Thus, the present study aims at providing further insight on two-phase pressure drops and flows patterns in U-bends. Based on a new type of U-bend test section, an extensive experimental study was conducted. The experimental campaign covered five test sections with three internal diameters (7.8, 10.8 and 13.4 mm), five bend diameters (24.8, 31.7, 38.1, 54.8 and 66.1 mm), tested for three orientations (horizontal, vertical upflow and vertical downflow), two fluids (R134a and R410A), two saturation temperatures (5 and 10 °C) and mass velocities ranging from 150 to 1000 kg s -1 m -2 . The flow pattern observations identified were stratified-wavy, slug-stratified-wavy, intermittent, annular, dryout and mist flows. The effects of the U-bend on the flow patterns were also observed. A total of 5655 pressure drop data were measured at seven different locations in the test section ( straight tubes and U-bend) providing a total of almost 40,000 data points. The straight tube data were first used to improve the actual two-phase straight tube model of Moreno-Quibén and Thome. This updated model was then used to developed a two-phase U-bend pressure drop model. Based on a comparison between experimental and predicted values, it is concluded that the new two-phase frictional pressure drop model for U-bends successfully

  2. Electromechanical phase transition of a dielectric elastomer tube under internal pressure of constant mass

    Directory of Open Access Journals (Sweden)

    Song Che

    2017-05-01

    Full Text Available The electromechanical phase transition for a dielectric elastomer (DE tube has been demonstrated in recent experiments, where it is found that the unbulged phase gradually changed into bulged phase. Previous theoretical works only studied the transition process under pressure control condition, which is not consistent with the real experimental condition. This paper focuses on more complex features of the electromechanical phase transition under internal pressure of constant mass. We derive the equilibrium equations and the condition for coexistent states for a DE tube under an internal pressure, a voltage through the thickness and an axial force. We find that under mass control condition the voltage needed to maintain the phase transition increases as the process proceeds. We analyze the entire process of electromechanical phase transition and find that the evolution of configurations is also different from that for pressure control condition.

  3. Failure Pressure Estimates of Steam Generator Tubes Containing Wear-type Defects

    International Nuclear Information System (INIS)

    Yoon-Suk Chang; Jong-Min Kim; Nam-Su Huh; Young-Jin Kim; Seong Sik Hwang; Joung-Soo Kim

    2006-01-01

    It is commonly requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. The critical defect dimensions have been determined based on the concept of plastic instability. This criterion, however, is known to be too conservative for some locations and types of defects. In this context, the accurate failure estimation for steam generator tubes with a defect draws increasing attention. Although several guidelines have been developed and are used for assessing the integrity of defected tubes, most of these guidelines are related to stress corrosion cracking or wall-thinning phenomena. As some of steam generator tubes are also failed due to fretting and so on, alternative failure estimation schemes for relevant defects are required. In this paper, three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of steam generator tubes with different defect configurations; elliptical wastage type, wear scar type and rectangular wastage type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of the steam generator tube. After investigating the effect of key parameters such as wastage depth, wastage length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wastage region. Comparison of failure pressures predicted according to the proposed estimation scheme with some corresponding burst test data shows good agreement, which provides a confidence in the use of the proposed equations to assess the integrity of steam generator tubes with wear-type defects. (authors)

  4. Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

    International Nuclear Information System (INIS)

    Diamond, W.T.

    2010-01-01

    One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

  5. The application of ductile-fracture analysis to predictions of pressure-tube failure

    International Nuclear Information System (INIS)

    Simpson, L.A.

    1981-08-01

    Progress during the past six years towards establishing a method for predicting critical crack length in a reactor pressure tube, based on data from tests on small fracture-mechanics specimens, is reviewed. The disadvantages of relying on data from burst tests alone are described along with the benefits of a small-specimen method. It is clear from the work reviewed that only an approach that can account for the ability of the presssure tube material to increase its crack-growth resistance during stable crack extension is suitable for the prediction of critical crack length. A method that utilizes crack-growth resistance curves based on crack-opening displacement, or the J integral, is described, along with a large body of experimental data. It is concluded that the resistance curve approach provides a viable method for the analysis of fracture in pressure tubes that can greatly improve our understanding of the material's behaviour

  6. Plane strain analytical solutions for a functionally graded elastic-plastic pressurized tube

    International Nuclear Information System (INIS)

    Eraslan, Ahmet N.; Akis, Tolga

    2006-01-01

    Plane strain analytical solutions to functionally graded elastic and elastic-plastic pressurized tube problems are obtained in the framework of small deformation theory. The modulus of elasticity and the uniaxial yield limit of the tube material are assumed to vary radially according to two parametric parabolic forms. The analytical plastic model is based on Tresca's yield criterion, its associated flow rule and ideally plastic material behaviour. Elastic, partially plastic and fully plastic stress states are investigated. It is shown that the elastoplastic response of the functionally graded pressurized tube is affected significantly by the material nonhomogeneity. Different modes of plasticization may take place unlike the homogeneous case. It is also shown mathematically that the nonhomogeneous elastoplastic solution presented here reduces to that of a homogeneous one by appropriate choice of the material parameters

  7. Extrusion and drawing of zircaloy 2. Production of pressure tubes for EL-4

    International Nuclear Information System (INIS)

    Thevenet, J.

    1964-01-01

    The authors give briefly the physical mechanical and chemical properties of zircaloy 2, as far as the transformation of this alloy is concerned. Extrusion: After a few general remarks concerning the extrusion and co-extrusion, including a comparison of the deformation resistance of canning metals and of zircaloy 2, the following points are considered: - the difficulties occurring because of the use of this alloy: - atmosphere protection - adjustment on to the machine tools - low thermal conductivity - economy of the metal (price) - the factors affecting the quality of the extruded products extrusion under a copper can and under lubricant glass - fine grain structure - temperature homogeneity - working temperature The transformation cycle - '550 kg ingot - preliminary shape 'for drawing of EL-4 tubes (112 x 120 L 12 m)' - is described in detail (extrusion or forging of the φ = 340 ingot into φ = 220 billets, cutting into lengths and hot drilling at φ = 125, fixing into a copper can and rough extrusion). Drawing: The main difficulties are due to seizing of the tools and to the necessity of protecting the alloy from the atmosphere during annealings. A brief description is given of drawing out on a short mandrel, on a long mandrel, of laminating on a reducing machine and of the carrying out of an annealing, as well as of the production of EL-4 tubes (φ =107 x 113 L 430 m) by drawing out shapes having a size of 112 x 120 on long mandrels. Conclusion: It is possible by extrusion and drawing to produce zircaloy 2 tubes similar to those which may be obtained normally using stainless steel. (authors) [fr

  8. Miniaturised Prandtl tube with integrated pressure sensors for micro-thruster plume characterisation

    NARCIS (Netherlands)

    Dijkstra, Marcel; Ma, Kechun; de Boer, Meint J.; Groenesteijn, Jarno; Lötters, Joost Conrad; Wiegerink, Remco J.

    2014-01-01

    A miniaturised Prandtl-tube sensor incorporating a 6 mm long 40 μm diameter microchannel with integrated pressure sensors has been realised. The sensor has been designed for the characterisation of rarefied plume flow from a MEMS-based monopropellant propulsion system for high-accuracy attitude

  9. A software tool for evaluation of hydrogen ingress in CANDU pressure tubes

    International Nuclear Information System (INIS)

    Mihalache, Maria; Vasile, Radu; Deaconu, Mariea

    2009-01-01

    The prediction of hydrogen isotopes concentration into the body and in the rolled joints of operating pressure tubes as a function of reactor hot hours is very important in many fitness-for-service assessments and end of life estimates. The rolled joints are high stress zones with potential for delayed hydride cracking. Predictive models for assessing the long-term deuterium ingress in both body and rolled joint of the pressure tubes have been implemented in a software tool, ROHID, developed in INR-Pitesti. ROHID is a PC-based Windows application with a user-friendly interface that predicts the equivalent hydrogen ingress for Zr-2.5Nb pressure tubes. It uses colour-coded reactor core maps to display the predicted deuterium concentration as a function of time for selected axial locations. Plots of deuterium versus axial location and time for individual pressure tubes are also available. Also, the software tool can predict the exceeding of hydrogen terminal solid solubility (HTSS) from hydrides during precipitation and dissolving processes as a function of time and axial location. (authors)

  10. Pre and post garter spring repositioning ultrasonic inspection of pressure tubes

    International Nuclear Information System (INIS)

    Desimone, C.; Katchadjian, P.; Tacchia, Mauricio

    1997-01-01

    This paper present a description of the ultrasonic cracked hydride blister detections system used for pre and post inspection of pressure tubes during garter spring repositioning in CNE (Embalse Nuclear Power Station). Ultrasonic system setup configuration, transducers characteristics, blister detection head, calibration of parameters, operating procedure, records of ultrasonic inspections and evaluation. (author) [es

  11. Evaluation of stress intensity factor for craks in surface of tubes with internal pressure

    International Nuclear Information System (INIS)

    Cesari, F.; Hellen, T.K.

    1977-01-01

    In this report the authors have examined the different methods for calculation of the stress intensity factor in tubes subject at internal pressure with surface cracks. The analysis includes cracks in 2-D axialsymmetric and 3-D. Moreover the authors have clarified the difference between the ASME Sec.11 and the procedure more rigorous

  12. Evaluated Plan Stress Of Weld In Pressure Tube Using X Ray Diffraction Technique

    International Nuclear Information System (INIS)

    Phan Trong Phuc; Nguyen Duc Thanh; Luu Anh Tuyen

    2011-01-01

    X ray diffraction is a fundamental technique measuring stress, this technique has determined crystal strain in materials, from that determined stress in materials. This paper presents study of evaluating plane stress of weld in pressure tube, using modern XRD apparatus: X Pert Pro. (author)

  13. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Additional information

    International Nuclear Information System (INIS)

    2009-03-01

    The reports from Argentina, Canada, India, Korea and Romania are presented concerning the projects carried out under the Coordinated Research Program (CRP) I3.30.10 of the International Agency for Atomic Energy - Vienna related to 'Intercomparison of Techniques for Pressure Tube Inspection and Diagnostics'

  14. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P. C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  15. Heat transfer and pressure drop of condensation of hydrocarbons in tubes

    Science.gov (United States)

    Fries, Simon; Skusa, Severin; Luke, Andrea

    2018-03-01

    The heat transfer coefficient and pressure drop are investigated for propane. Two different mild steel plain tubes and saturation pressures are considered for varying mass flux and vapour quality. The pressure drop is compared to the Friedel-Correlation with two different approaches to determine the friction factor. The first is calculation as proposed by Friedel and the second is through single phase pressure drop investigations. For lower vapour qualities the experimental results are in better agreement with the approach of the calculated friction factor. For higher vapour qualities the experimental friction factor is more precise. The pressure drop increases for a decreasing tube diameter and saturation pressure. The circumferential temperature profile and heat transfer coefficients are shown for a constant vapour quality at varying mass fluxes. The subcooling is highest for the bottom of the tube and lowest for the top. The average subcooling as well as the circumferential deviation decreases for rising mass fluxes. The averaged heat transfer coefficients are compared to the model proposed by Thome and Cavallini. The experimental results are in good agreement with both correlations, however the trend is better described with the correlation from Thome. The experimental heat transfer coefficients are under predicted by Thome and over predicted by Cavallini.

  16. The study on pressure oscillation and heat transfer characteristics of oscillating capillary tube heat pipe

    International Nuclear Information System (INIS)

    Kim, Jong Soo; Bui, Ngoc Hung; Jung, Hyun Seok; Lee, Wook Hyun

    2003-01-01

    In the present study, the characteristics of pressure oscillation and heat transfer performance in an oscillating capillary tube heat pipe were experimentally investigated with respect to the heat flux, the charging ratio of working fluid, and the inclination angle to the horizontal orientation. The experimental results showed that the frequency of pressure oscillation was between 0.1 Hz and 1.5 Hz at the charging ratio of 40 vol.%. The saturation pressure of working fluid in the oscillating capillary tube heat pipe increased as the heat flux was increased. Also, as the charging ratio of working fluid was increased, the amplitude of pressure oscillation increased. When the pressure waves were symmetric sinusoidal waves at the charging ratios of 40 vol.% and 60 vol.%, the heat transfer performance was improved. At the charging ratios of 20 vol.% and 80 vol.%, the waveforms of pressure oscillation were more complicated, and the heat transfer performance reduced. At the charging ratio of 40 vol.%, the heat transfer performance of the OCHP was at the best when the inclination angle was 90 .deg., the pressure wave was a sinusoidal waveform, the pressure difference was at the least, the oscillation amplitude was at the least, and the frequency of pressure oscillation was the highest

  17. The development of an auto-sealing system using an electrically shrinkable tube under a low-pressure condition

    Energy Technology Data Exchange (ETDEWEB)

    Okano, Yoshihiro; Kitagawa, Takao [NKK Corp, Tsu, Mie (Japan); Shoji, Norio [NKK Corp., Yokohama (Japan); Namioka, Toshiyuki [Nippon Kokan Koji Corp., Yokohama (Japan). Research and Development Dept.; Komura, Minoru [Nitto Denko Corp., Fukaya, Saitama (Japan)

    1997-04-01

    This article describes the development of a system to create high quality, automatic sealing of field joints of polyethylene coated pipelines. The system uses a combination of an electrically heated shrinkable tube and a low-pressure chamber. The self-heating shrinkable tube includes electric heater wires that heat when connected to electricity. A method was developed to eliminate air trapped between the tube and the steel pipe by shrinking the tube under a low-pressure condition. The low-pressure condition was automatic and easily attained by using a vacuum chamber. It was verified that the system produced high quality sealing of the field joints.

  18. Influence of microstructure on stress corrosion cracking susceptibility of alloys 600 and 690 in primary water of pressurized water reactors

    International Nuclear Information System (INIS)

    Kergaravat, J.F.

    1996-01-01

    The mechanism(s) responsible for the stress corrosion cracking (SCC) of Alloy 600 steam generator tubes of pressurized water reactors remain misunderstood in spite of numerous studies on the subject. This failure mode presents several experimental similarities with intergranular creep fracture of austenitic stainless steels. As far as intergranular creep fracture is concerned, grain boundary sliding (GBS) was proved to favor failure. The aim of this work is to check the role played by GBS during SCC. It takes into account chemical (chromium content) and microstructural parameters (grain size, precipitation distribution and density). Therefore, to get a complete set of micro-structurally different samples, we have prepared solution annealed specimens (1100 deg C, 20 min., water quenched) from industrial tubes of Alloys 600 and 690. Each specimen was crept at 500 deg C (400 MPa), 430 deg C (425 MPa) and 360 deg C (475 MPa). Before testing, every sample were engraved with a 7 μm wide fiducial grid. This grid has allowed us to measure GBS after creep testing. GBS was observed for industrial and solution annealed samples for the three testing temperatures. GBS amplitude depends'on chromium content: for micro-structurally identical specimens, Alloy 600 exhibits more GB strain than Alloy 690. It also strongly depends on grain boundary precipitation characteristics: carbide free boundaries slide more easily. During in situ straining experiments performed in a transmission electronic microscope, GBS was evidenced at 320 deg C for Alloy 600 industrial samples. It consists in grain boundary dislocation motion in the interface plane. These dislocations originate from perfect dislocations gliding in the grain interior, encountering grain boundary and spreading in it. Metallic intergranular carbides provide strong obstacles to GBS so stress enhancements arise against them. These stress enhancements are released by micro-twin emission. Constant extension rate tensile tests were

  19. Evaporation of R134a in a horizontal herringbone microfin tube: heat transfer and pressure drop

    Energy Technology Data Exchange (ETDEWEB)

    Wellsandt, S; Vamling, L [Chalmers University of Technology, Gothenburg (Sweden). Department of Chemical Engineering and Environmental Science, Heat and Power Technology

    2005-09-01

    An experimental investigation of in-tube evaporation of R134a has been carried out for a 4 m long herringbone microfin tube with an outer diameter of 9.53 mm. Measured local heat transfer coefficients and pressure losses are reported for evaporation temperatures between -0.7 and 10.1 {sup o}C and mass flow rates between 162 and 366 kg m{sup -2} s{sup -1}. Results from this work are compared to experimental results from literature as well as predicted values from some available helical microfin correlations. Differences in heat transfer mechanisms between helical and herringbone microfin tubes are discussed, as heat transfer coefficients in the investigated herringbone tube tend to peak at lower vapour qualities compared to helical microfins. Correlations developed for helical microfin tubes generally predict experimental values within {+-}30% for vapour qualities below 50%. However, at higher qualities none of the correlations are able to reflect the early peak of heat transfer coefficients. Predicted pressure gradients reproduce measured values in general within {+-}20%. (author)

  20. Some engineering aspects of the investigation into the cracking of pressure tubes in the Pickering reactors

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Towgood, G.R.; Hunter, T.A.

    1976-01-01

    In August 1974, Pickering Unit 3 (514 MWe) was shutdown for a period of 8 months because of cracks in 17 of the 390 pressure tubes. The cracks were a result of incorrect installation procedures during construction. Improper positioning of the rolling tool used to join the Zr-2.5 wt% Nb pressure tube to the end fitting produced very high residual tensile stresses. High stresses in combination with periods with the tubes cold caused the cracking. Crack propagation was by fracture of hydrides which are brittle when cold. Subsequent investigation confirmed that properly rolled joints are not susceptible to such cracking. The resources of Canadian industry, Ontario Hydro and Atomic Energy of Canada were coordinated to find engineering solutions to the crack program. The defective tubes were removed from reactor, thoroughly examined to identify the cause of the cracks, and thoroughly tested to prove safety. Non-destructive techniques were quickly adopted for inspection of tubes in Pickering. Tools and procedures for retubing the 17 channels were prepared and Pickering Unit 3 was returned to service at the end of March 1975. (author)

  1. Development of NDT techniques for the inspection of WSGHWR pressure tubes

    International Nuclear Information System (INIS)

    Gray, B.S.; Highmore, P.J.; Rudlin, J.R.; Cooper, A.G.

    1979-01-01

    The fuel for the Steam Generating Heavy Water Reactor at Winfrith Heath is contained in vertical Zircaloy pressure tubes and is cooled by boiling light water. This paper describes the development of NDT techniques for the inservice examination of the pressure tubes to provide continuing assurance of the absence of axial crack-like defects. The resultant equipment has to operate in water-filled tubes in the presence of the radiation field due to the irradiated fuel elements in adjacent tubes. Also, a layer of surface oxide on the inside of the tubes has been found to significantly affect the behaviour of a prototype inspection device. To provide adequate sensitivity in these conditions, without the occurrence of unnecessary spurious indications, a combination of techniques has been developed. This involves the use of ultrasonics in both pulse-echo and 'pitch and catch' mode together with a single frequency eddy current technique. Laboratory work using artificial defects is described and also how the development programme was modified to accommodate the results of in-reactor tests using a prototype device. Reference is also made to the development of CCTV equipment to provide a supplementary visual examination. (author)

  2. Determination of dislocation density in Zr-2.5Nb pressure tubes by x-ray

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Isaenkova, Perlovich; Cheong, Y. M.; Kim, S. S.; Yim, K. S.; Kwon, Sang Chul

    2000-11-01

    For X-ray determination of the dislocation density in CANDU Zr-2.5%Nb pressure tubes, a program was developed, using the Fourier analysis of X-ray line profiles and calculation of dislocation density by values of the coherent block size and the lattice distortion. The coincidence of obtained values of c- and a-dislocations with those, determined by the X-ray method for the same tube in AECL, was assumed to be the main criterion of validity of the developed program. The final variant of the program allowed to attain a rather close coincidence of calculated dislocation densities with results of AECL. The dislocation density was determined in all the zirconium grains with different orientations based on the texture of the stree-relieved CANDU tube. The complete distribution of c-dislocation density in -Zr grains depecding on their crystallographic orientations was constructed. The distribution of a-dislocation density within the texture maximum at L-direction, containing prismatic axes of all grains, was constructed as well. The analysis of obtained distributions testifies that -Zr grains of the stree-relieved CANDU tube significantly differ in their dislocation densities. Plotted diagrams of correlation between the dislocation density and the pole density allow to estimate the actual connection between texture and dislocation distribution in the studied tube. The distributions of volume fractions of all the zirconium grains depending on their dislocation density were calculated both for c- and a-dislocations. The distributions characterizes quantitatively the inhomogeneity of substructure conditions in the stress-relieved CANDU tube. the optimal procedure for determination of Nb content in {beta}-phases of CANDU Zr-2.5%Nb pressure tubes was also established.

  3. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  4. Hydrophilic film polymerized on the inner surface of PMMA tube by an atmospheric pressure plasma jet

    Science.gov (United States)

    Yin, Mengmeng; Huang, Jun; Yu, Jinsong; Chen, Guangliang; Qu, Shanqing

    2017-07-01

    Polymethyl methacrylate (PMMA) tube is widely used in biomedical and mechanical engineering fields. However, it is hampered for some special applications as the inner surface of PMMA tube exhibts a hydrophobic characteristic. The aim of this work is to explore the hydrophilic modification of the inner surface of the PMMA tubes using an atmospheric pressure plasma jet (APPJ) system that incorporates the acylic acid monomer (AA). Polar groups were grafted onto the inner surface of PMMA tube via the reactive radicals (•OH, •H, •O) generated in the Ar/O2/AA plasma, which were observed by the optical emission spectroscopy (OES). The deposition of the PAA thin layer on the PMMA surface was verified through the ATR-FTIR spectra, which clearly showed the strengthened stretching vibration of the carbonyl group (C=O) at 1700 cm-1. The XPS data show that the carbon ratios of C-OH/R and COOH/R groups increased from 9.50% and 0.07% to 13.49% and 17.07% respectively when a discharge power of 50 W was used in the APPJ system. As a result, the static water contat angle (WCA) of the modified inner surface of PMMA tube decreased from 100° to 48°. Furthermore, the biocompatibility of the APP modified PMMA tubes was illustrated by the study of the adhesion of the cultured MC3T3-E1 osteocyte cells, which exhibted a significantly enhanced adhesion density.

  5. Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

    International Nuclear Information System (INIS)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement

  6. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Seitz, W.W. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1997-12-01

    In Phase 1 a variety of developmental and commercial tubing alloys and claddings were exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347 RA-85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 Ta modified, NF 709, 690 clad, and 671 clad for approximately 4,000, 12,000, and 16,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were assembled on an air-cooled, retractable corrosion probe, the probe was installed in the reheater activity of the boiler and controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. The results will be presented for the preliminary metallurgical examination of the corrosion probe samples after 16,000 hours of exposure. Continued metallurgical and interpretive analysis is still on going.

  7. Design of a high-pressure single pulse shock tube for chemical kinetic investigations

    International Nuclear Information System (INIS)

    Tranter, R. S.; Brezinsky, K.; Fulle, D.

    2001-01-01

    A single pulse shock tube has been designed and constructed in order to achieve extremely high pressures and temperatures to facilitate gas-phase chemical kinetic experiments. Postshock pressures of greater than 1000 atmospheres have been obtained. Temperatures greater than 1400 K have been achieved and, in principle, temperatures greater than 2000 K are easily attainable. These high temperatures and pressures permit the investigation of hydrocarbon species pyrolysis and oxidation reactions. Since these reactions occur on the time scale of 0.5--2 ms the shock tube has been constructed with an adjustable length driven section that permits variation of reaction viewing times. For any given reaction viewing time, samples can be withdrawn through a specially constructed automated sampling apparatus for subsequent species analysis with gas chromatography and mass spectrometry. The details of the design and construction that have permitted the successful generation of very high-pressure shocks in this unique apparatus are described. Additional information is provided concerning the diaphragms used in the high-pressure shock tube

  8. Corrosion in PWR steam generator tubes made of alloy 600TT: overview of operating experience, NDE and safety issues

    International Nuclear Information System (INIS)

    Curieres, I. de; Sollier, T.; Delaval, C.

    2015-01-01

    About 60 PWR plants worldwide are operating with steam generator tubes made of alloy 600TT, among which 27 are located in France. This alloy is susceptible to corrosion, both on the primary and secondary side in every fleet, though with different kinetics or extent. It is noteworthy that many of the primary side corrosion issues can be clearly explained by design or operating conditions. However, studies show that all the secondary side issues are much hardly explained by simple considerations. This paper will give an overview of the international operating experience of this alloy and indicate the associated controllability and safety-related issues. An emphasis will be put on the manufacturing, chemistry and specificities of the different fleets. The French situation will be reviewed in this frame. (authors)

  9. Correlations of CO2 at supercritical pressures in a vertical circular tube

    International Nuclear Information System (INIS)

    Li Zhihui; Jiang Peixue

    2010-01-01

    The experiment results of convection heat transfer of CO 2 at supercritical pressures in a 2 mm diameter vertical circular tube for upward flow and downward flow were analyzed for pressures ranging from 78 to 95 bar, inlet temperatures from to 25 to 40 degree C, and inlet Re numbers from 3000 to 20000. The results were compared with some well known empirical correlations for the heat transfer without buoyancy effects and the heat transfer with strong buoyancy effects. It is found that there is a big deviation between the experiment results and empirical correlations. Based on the experiment data, correlations are developed for the local Nusselt correlations of CO 2 at supercritical pressures in vertical circular tubes.(authors)

  10. A review of current knowledge on the effects of hydrogen on the pressure tubes of Ontario Hydro operating reactors

    International Nuclear Information System (INIS)

    Leger, M.

    1982-01-01

    Since the occurrence of cracking in Zr-2.5 wt% Nb pressure tubes in Pickering 'A' units 3 and 4 in 1974/75 a great deal of information on the behaviour of hydrogen in pressure tube materials has been generated through research effort by both AECL and Ontario Hydro. In order to use this information effectively and to provide direction and co-ordination for ongoing research, a review of available information and current concerns on hydrogen in pressure tubes was undertaken. The review was divided into two main areas of interest: hydrogen ingress and hydride effects. The uncertainties in the rates of hydrogen ingress into the pressure tubes have been found to be very large. On the basis of current knowledge, predictions of the future behaviour of pressure tubes due to hydride effects are extremely difficult

  11. Reducing non value adding aluminium alloy in production of parts through high pressure die casting

    CSIR Research Space (South Africa)

    Pereira, MFVT

    2010-10-01

    Full Text Available in the cast part feed system, including overflows. CSIR intends using the results of this research for further development and application of high temperature die construction materials in high pressure die casting processes of light metal alloys...

  12. Turbulent convective heat transfer of methane at supercritical pressure in a helical coiled tube

    Science.gov (United States)

    Wang, Chenggang; Sun, Baokun; Lin, Wei; He, Fan; You, Yingqiang; Yu, Jiuyang

    2018-02-01

    The heat transfer of methane at supercritical pressure in a helically coiled tube was numerically investigated using the Reynolds Stress Model under constant wall temperature. The effects of mass flux ( G), inlet pressure ( P in) and buoyancy force on the heat transfer behaviors were discussed in detail. Results show that the light fluid with higher temperature appears near the inner wall of the helically coiled tube. When the bulk temperature is less than or approach to the pseudocritical temperature ( T pc ), the combined effects of buoyancy force and centrifugal force make heavy fluid with lower temperature appear near the outer-right of the helically coiled tube. Beyond the T pc , the heavy fluid with lower temperature moves from the outer-right region to the outer region owing to the centrifugal force. The buoyancy force caused by density variation, which can be characterized by Gr/ Re 2 and Gr/ Re 2.7, enhances the heat transfer coefficient ( h) when the bulk temperature is less than or near the T pc , and the h experiences oscillation due to the buoyancy force. The oscillation is reduced progressively with the increase of G. Moreover, h reaches its peak value near the T pc . Higher G could improve the heat transfer performance in the whole temperature range. The peak value of h depends on P in. A new correlation was proposed for methane at supercritical pressure convective heat transfer in the helical tube, which shows a good agreement with the present simulated results.

  13. Computer modelling of eddy current probes for ISI of pressure tube/calandria tube assemblies in PHWRs

    International Nuclear Information System (INIS)

    Rao, B.P.C.; Shyamsunder, M.T.; Bhattacharya, D.K.; Raj, Baldev

    1992-01-01

    Non-destructive Evaluation (NDE) plays a major role in ensuring the safe and reliable operation of PHWRs which are the mainstay of India's nuclear power programme. An important in-service inspection (ISI) requirement in these reactors is carried out through Eddy Current Testing (ECT) of the pressure tube (PT)/calandria tube (CT) assemblies. The material of construction of these assemblies is zircaloy-2. The two main objectives of this ISI are the detection of garter spring between CT and PT and the profiling of gap between CT and PT. The paper discusses the work carried out at the authors' laboratory on the development of ECT probes for ISI of PT/CT assemblies. Emphasis has been given on the work done on the design and optimisation of the probes using computer modeling. A 2-D finite element code has been developed for this purpose. The code is developed around a diffusion equation which can be derived from Maxwell's equations governing the electromagnetic phenomenon. An axisymmetry has been considered, since the probes are bobbin type. Results of impedance plane outputs obtained by modelling and those by experiments using actual probes have shown good matching. Salient features of an indigenously developed interactive PC based data acquisition, analysis and retrieval system to cater to ISI of PC/CT assemblies are described. (author). 10 refs., 7 figs

  14. Study on drop pressure and flow distribution of double-tube heat exchanger

    International Nuclear Information System (INIS)

    Liu Junqiang; Chen Minghui; Hu Yumin; Li Rizhu; Kong Dechun; Zhang Weijie

    2007-01-01

    The parallel connection channel pressure drop characters of the double-tube bundle heat exchange were experimentally investigated in this paper in order to find out how the flow of the heat exchanger is distributed and then to optimize the structure of heat exchanger according to the flow distribution. A double-tube bundle heat exchanger was built according to the similarity criteria. The experiment system was also built to test the optimization of the heat exchanger. The experiment results reveal that the calculating model is reliable and decreasing pipe space to optimize the heat exchanger is reasonable. (authors)

  15. Heat exchanging tube behaviour in steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Pastor, D.; Oertel, K.

    1979-01-01

    Based on a comprehensive failure statistics, materials corrosion chemistry and thermohydraulics problems of the tubings of steam generators are considered. A historical review of failures in the tubings of steam generators in pressurized water reactors reflects the often successless measures by designers, manufacturers and operating organizations for preventing failures, especially with regard to materials selection and water regime. It is stated that laboratory tests could not give sufficient information about safe and stable operation of nuclear steam generators unless real constructive, hydrodynamic, thermodynamical and chemical conditions of operation had been taken into account. (author)

  16. Laser-Doppler vibrating tube densimeter for measurements at high temperatures and pressures

    International Nuclear Information System (INIS)

    Aida, Tsutomu; Yamazaki, Ai; Akutsu, Makoto; Ono, Takumi; Kanno, Akihiro; Hoshina, Taka-aki; Ota, Masaki; Watanabe, Masaru; Sato, Yoshiyuki; Smith, Richard L. Jr.; Inomata, Hiroshi

    2007-01-01

    A laser-Doppler vibrometer was used to measure the vibration of a vibrating tube densimeter for measuring P-V-T data at high temperatures and pressures. The apparatus developed allowed the control of the residence time of the sample so that decomposition at high temperatures could be minimized. A function generator and piezoelectric crystal was used to excite the U-shaped tube in one of its normal modes of vibration. Densities of methanol-water mixtures are reported for at 673 K and 40 MPa with an uncertainty of 0.009 g/cm 3

  17. Study on the manufacturing process, causes of the pressure tube failure and methods for improving its performance

    Energy Technology Data Exchange (ETDEWEB)

    You, Ho Sik; Jeong, Jin Kon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Manufacturing processes of Zr-2.5Nb pressure tube used in CANDU reactor, effects of impurities on the properties of the pressure tube, experiences and causes of the pressure tube cracking accident and the development programs on the fuel channel at AECL have been described. Fabrication processes on the pressure tube have been explained in detail from the sponge production step to the final product. Test methods that are performed to verify the integrity of the final product have also been described. Most of the pressure tube rupture accidents were caused by DHC (Delayed Hydride Cracking). In cases of Pickering units 3 and 4 and Bruce unit 2, excessive residual stresses induced by improper rolled joint process had played a role to cause DHC. In Pickering unit 2, cracks formed by contact between pressure and calandria tubes due to the movement of garter spring were direct cause of failure. After the accidents, a lot of R and D programs on each component of the fuel channel have been carried out. The study on the improvement of manufacturing processes such as increasing cold working rate, performing the intermediate and final annealing and adding the third element like Fe, V, Cr for enhancing the pressure tube performance are on progress. To suppress hydrogen uptake into the pressure tube, the methods such as zirconia coating on the pressure tube, Cr-plating on the end fitting and placing the yttrium getter on the pressure tube are considered. Experiments on each test specimen are currently under way. Owing to such an effort, more advanced fuel channel can be installed in the next CANDU reactor. 6 tabs., 20 figs., 20 refs. (Author).

  18. Overview of research trends and problems on Cr-Mo low alloy steels for pressure vessel

    International Nuclear Information System (INIS)

    Chi, Byung Ha; Kim, Jeong Tae

    2000-01-01

    Cr-Mo low alloy steels have been used for a long time for pressure vessel due to its excellent corrosion resistance, high temperature strength and toughness. The paper reviewed the latest trends on material development and some problems on Cr-Mo low alloy steel for pressure vessel, such as elevated temperature strength, hardenability, synergetic effect between temper and hydrogen embrittlement, hydrogen attack and hydrogen induced disbonding of overlay weld-cladding

  19. Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

    International Nuclear Information System (INIS)

    Royen Van, E.

    2011-11-01

    The complexity of two-phase flow boiling on a tube bundle presents many challenges to the understanding of the physical phenomena taking place. It is important to quantify these numerous heat flow mechanisms in order to better describe the performance of tube bundles as a function of the operational conditions. In the present study, the bundle boiling facility at the Laboratory of Heat and Mass Transfer (LTCM) was modified to obtain high-speed videos to characterise the two-phase regimes and some bubble dynamics of the boiling process. It was then used to measure heat transfer on single tubes and in bundle boiling conditions. Pressure drop measurements were also made during adiabatic and diabatic bundle conditions. New enhanced boiling tubes from Wolverine Tube Inc. (Turbo-B5) and the Wieland-Werke AG (Gewa-B5) were investigated using R134a and R236fa as test fluids. The tests were carried out at saturation temperatures T sat of 5 °C and 15 °C, mass flow rates from 4 to 35 kg/m 2 s and heat fluxes from 15 to 70 kW/m 2 , typical of actual operating conditions. The flow pattern investigation was conducted using visual observations from a borescope inserted in the middle of the bundle. Measurements of the light attenuation of a laser beam through the intertube two-phase flow and local pressure fluctuations with piezo-electric pressure transducers were also taken to further help in characterising the complex flow. Pressure drop measurements and data reduction procedures were revised and used to develop new, improved frictional pressure drop prediction methods for adiabatic and diabatic two-phase conditions. The physical phenomena governing the enhanced tube evaporation process and their effects on the performance of tube bundles were investigated and insight gained. A new method based on a theoretical analysis of thin film evaporation was used to propose a new correlating parameter. A large new database of local heat transfer coefficients were obtained and then

  20. Effect of Ovality on Maximum External Pressure of Helically Coiled Steam Generator Tubes with a Rectangular Wear

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Dong In; Lim, Eun Mo; Huh, Nam Su [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of); Choi, Shin Beom; Yu, Je Yong; Kim, Ji Ho; Choi, Suhn [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    A structural integrity of steam generator tubes of nuclear power plants is one of crucial parameters for safe operation of nuclear power plants. Thus, many studies have been made to provide engineering methods to assess integrity of defective tubes of commercial nuclear power plants considering its operating environments and defect characteristics. As described above, the geometric and operating conditions of steam generator tubes in integral reactor are significantly different from those of commercial reactor. Therefore, the structural integrity assessment of defective tubes of integral reactor taking into account its own operating conditions and geometric characteristics, i. e., external pressure and helically coiled shape, should be made to demonstrate compliance with the current design criteria. Also, ovality is very specific characteristics of the helically coiled tube because it is occurred during the coiling processes. The wear, occurring from FIV (Flow Induced Vibration) and so on, is main degradation of steam generator tube. In the present study, maximum external pressure of helically coiled steam generator tube with wear is predicted based on the detailed 3-dimensional finite element analysis. As for shape of wear defect, the rectangular shape is considered. In particular, the effect of ovality on the maximum external pressure of helically coiled tubes with rectangular shaped wear is investigated. In the present work, the maximum external pressure of helically coiled steam generator tube with rectangular shaped wear is investigated via detailed 3-D FE analyses. In order to cover a practical range of geometries for defective tube, the variables affecting the maximum external pressure were systematically varied. In particular, the effect of tube ovality on the maximum external pressure is evaluated. It is expected that the present results can be used as a technical backgrounds for establishing a practical structural integrity assessment guideline of

  1. Fuel-element vibration and bearing pad to pressure tube fretting

    International Nuclear Information System (INIS)

    Fisher, N.J.; Taylor, C.E.; Pettigrew, M.J.

    1990-08-01

    Fuel channel operation under boiling condition results in increased flow velocities, which may lead to unacceptable fuel-element vibration and bearing pad to pressure tube fretting. The existing endurance test database does not fully cover the range of future channel operating conditions. In particular, after refuelling, some channels for future designs may operate with two-phase flow conditions outside the range of endurance test conditions. Full-scale endurance testing at realistic steam-water conditions involves substantial energy costs. Therefore, fundamental laboratory investigations were conducted to define and endurance test matrix which adequately envelops the future range of operating conditions while minimizing both the number of tests and the energy requirement of individual tests. The main focus of the laboratory investigations was to establish the relationships between: fuel channel flow conditions and fuel-element vibration; and fuel-element vibration and bearing pad to pressure tube fretting. The vibration response of a single fuel element was measured over a wide range of operating conditions covering realistic fuel channel conditions and simulated endurance testing conditions. For higher void fractions, the vibration amplitudes measured in air/water were much higher than in steam/water, while for low void fractions, the amplitudes were similar. The measured amplitudes in steam/water varied very little over the range of temperature and pressure investigated. The effects of temperature, pressure tube oxide thickness, vibration amplitude and bearing pad manufacturer on pressure tube fretting were investigated. The fretting rate is extremely temperature dependent. For vibration amplitudes about three or four times greater than expected in-reactor conditions, peak fretting rates were observed in the 225 to 286 degrees C temperature range. Fretting rates were seven times less at the higher temperatures of 300 and 315 degrees C, and the lower temperatures

  2. Microstructure and mechanical properties of an Al–Mg alloy solidified under high pressures

    International Nuclear Information System (INIS)

    Jie, J.C.; Zou, C.M.; Brosh, E.; Wang, H.W.; Wei, Z.J.; Li, T.J.

    2013-01-01

    Highlights: •Al–42.2Mg alloy was solidified under pressures of 1, 2, and 3 GPa and the microstructure analyzed. •A thermodynamic calculation of the Al–Mg phase diagram at high pressures was performed. •The phase content changes from predominantly γ-Al 12 Mg 17 at 1 GPa to FCC solid solution at 3 GPa. •The β-Al 3 Mg 2 is predicted to remain stable at low temperatures but is not observed. •The alloy solidified at high pressure has remarkably enhanced ultimate tensile strength. -- Abstract: Phase formation, the microstructure and its evolution, and the mechanical properties of an Al–42.2 at.% Mg alloy solidified under high pressures were investigated. After solidification at pressures of 1 GPa and 2 GPa, the main phase is the γ phase, richer in Al than in equilibrium condition. When the pressure is further increased to 3 GPa, the main phase is the supersaturated Al(Mg) solid solution with Mg solubility up to 41.6 at.%. Unlike in similar alloys solidified at ambient pressure, the β phase does not appear. Calculated high-pressure phase diagrams of the Al–Mg system show that although the stability range of the β phase is diminished with pressure, it is still thermodynamically stable at room temperature. Hence, the disappearance of the β phase is interpreted as kinetic suppression, due to the slow diffusion rate at high pressures, which inhibits solid–solid reactions. The Al–42.2 at.% Mg alloy solidified under 3 GPa has remarkably enhanced ultimate tensile strength compared to the alloy solidified under normal atmospheric pressure

  3. Heat transfer test in a tube using CO2 at supercritical pressures

    International Nuclear Information System (INIS)

    Kim, Hwan Yeol; Kim, Hyungrae; Song, Jin Ho; Cho, Bong Hyun; Bae, Yoon Yeong

    2005-01-01

    Heat transfer test facility, which is named as SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), has been constructed in KAERI for the study of heat transfer and pressure drop characteristics in a single tube, single rod and rod bundle at supercritical CO 2 conditions. The tests with supercritical water are difficult it terms of cost and effort, since the critical pressure and temperature of water are as high as 22.12 MPa and 374.14degC. As a substitute for water, CO 2 is selected for the test since the critical pressure and temperature of CO 2 are 7.38 MPa and 31.05degC that are much lower than those of water. This paper describes the design characteristics of the SPHINX and the experimental investigations on the heat transfer and pressure drop of a vertical single tube with an inside diameter of 4.4 mm with upward flow of supercritical CO 2 . The geometry of the single tube is the same as that of Kyushu University test performed with Freon (R22) for the direct comparison of a medium effect. The tests were performed with various heat and mass fluxes at a given pressure. The range of mass flux is 400∼1200 kg/m 2 s and the heat flux is chosen up to 150 kW/m 2 . The selected pressure are 7.75, 8.12, and 8.85 MPa. The test results are investigated and compared with the previous tests. (author)

  4. Pressure Measurements on a Deforming Surface in Response to an Underwater Explosion in a Water-Filled Aluminum Tube

    Directory of Open Access Journals (Sweden)

    G. Chambers

    2001-01-01

    Full Text Available Experiments have been conducted to benchmark DYSMAS computer code calculations for the dynamic interaction of water with cylindrical structures. Small explosive charges were suspended using hypodermic needle tubing inside Al tubes filled with distilled water. Pressures were measured during shock loading by tourmaline crystal, carbon resistor and ytterbium foil gages bonded to the tube using a variety of adhesives. Comparable calculated and measured pressures were obtained for the explosive charges used, with some gages surviving long enough to record results after cavitation with the tube wall.

  5. A study on the critical heat flux for annuli and round tubes under low pressure conditions

    International Nuclear Information System (INIS)

    Park, Jae Wook

    1997-02-01

    This study aims to reveal the characteristics of the critical heat flux (CHF) of internally heated concentric annuli and vertical round tubes in low-pressure and low-flow (LPLF) conditions. Although many efforts have been devote to the subject of the CHF during the last forty years, the information on the CHF phenomenon for LPLF conditions is still very limited. The applicable ranges of the CHF correlations for annuli and round tubes are concentrate on the operating conditions of nuclear power plant (NPP), namely high-pressure and high-flow (HPHF) conditions. these facts promoted to collect the reliable CHF data for LPLF conditions for both annuli and round tubes. The critical heat flux data for vertical flow boiling of water in annuli and round tubes at low pressures and low mass fluxes show the following trends: The observed CHF mechanism for annuli was changed in the order of flooding, churn-to-annular flow transition, and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. The observed parametric trends for annuli are consistent with the previous understanding except that the CHF for downward flow is considerably lower (up to 40%) than that for upward flow. The critical quality is much lower than that for round tubes at the same inlet conditions. The observed parametric trends for round tubes are generally consistent with the previous understanding except for system pressure an tube diameter effect. For the system pressure effect, it is observed that the pressure effect is complicated but not so large, whereas the existing CHF correlations do not present the parametric trend exactly. For tube diameter effect, the decreasing trends of CHF with respect to tube diameter was the general understanding so far, but in this region the CHF show a increasing trend of tube diameter. The prediction and the parametric trend analyses are performed by two view points, I.e., for fixed inlet conditions and for local

  6. Reconstruction of an acoustic pressure field in a resonance tube by particle image velocimetry.

    Science.gov (United States)

    Kuzuu, K; Hasegawa, S

    2015-11-01

    A technique for estimating an acoustic field in a resonance tube is suggested. The estimation of an acoustic field in a resonance tube is important for the development of the thermoacoustic engine, and can be conducted employing two sensors to measure pressure. While this measurement technique is known as the two-sensor method, care needs to be taken with the location of pressure sensors when conducting pressure measurements. In the present study, particle image velocimetry (PIV) is employed instead of a pressure measurement by a sensor, and two-dimensional velocity vector images are extracted as sequential data from only a one- time recording made by a video camera of PIV. The spatial velocity amplitude is obtained from those images, and a pressure distribution is calculated from velocity amplitudes at two points by extending the equations derived for the two-sensor method. By means of this method, problems relating to the locations and calibrations of multiple pressure sensors are avoided. Furthermore, to verify the accuracy of the present method, the experiments are conducted employing the conventional two-sensor method and laser Doppler velocimetry (LDV). Then, results by the proposed method are compared with those obtained with the two-sensor method and LDV.

  7. Comparison of stethoscope bell and diaphragm, and of stethoscope tube length, for clinical blood pressure measurement.

    Science.gov (United States)

    Liu, Chengyu; Griffiths, Clive; Murray, Alan; Zheng, Dingchang

    2016-06-01

    This study investigated the effect of stethoscope side and tube length on auscultatory blood pressure (BP) measurement. Thirty-two healthy participants were studied. For each participant, four measurements with different combinations of stethoscope characteristics (bell or diaphragm side, standard or short tube length) were each recorded at two repeat sessions, and eight Korotkoff sound recordings were played twice on separate days to one experienced listener to determine the systolic and diastolic BPs (SBP and DBP). Analysis of variance was carried out to study the measurement repeatability between the two repeat sessions and between the two BP determinations on separate days, as well as the effects of stethoscope side and tube length. There was no significant paired difference between the repeat sessions and between the repeat determinations for both SBP and DBP (all P-values>0.10, except the repeat session for SBP using short tube and diaphragm). The key result was that there was a small but significantly higher DBP on using the bell in comparison with the diaphragm (0.66 mmHg, P=0.007), and a significantly higher SBP on using the short tube in comparison with the standard length (0.77 mmHg, P=0.008). This study shows that stethoscope characteristics have only a small, although statistically significant, influence on clinical BP measurement. Although this helps understand the measurement technique and resolves questions in the published literature, the influence is not clinically significant.

  8. Study the Effect of Nasal Obstruction Surgery (Septoplasty on Eustachian Tube Function and Middle Ear Pressure

    Directory of Open Access Journals (Sweden)

    Rahim Davari

    2014-12-01

    Full Text Available Background & objectives: Deviation of the nasal septum is a common cause of unilateral or bilateral nasal airway obstruction and may follow nasal and midface trauma. Patient complaints of airway obstruction that are consistent with intranasal physical findings often lead to septoplasty and turbinate surgery. Severe nasal septal deviation leads to complete nasal obstruction and disturbs air passage from nostrils, however the effect of septal deviation and nasal obstruction surgery (septoplasty on Eustachian tube function and middle ear pressure is controversial and isn’t clear. Whereas many of surgeons do not believe the considerable effect of septoplasty on Eustachian tube function and middle ear pressure, so they do not recommend this procedure before middle ear surgery. On the other hand, some have an idea that septoplasty has significant effect on middle ear pressure suggesting this procedure before ear surgery like tympanoplasty.   Methods: This prospective analytical-descriptive study was conducted on seventy patients from 18 to 65 years of age who underwent septoplasty due to severe septal deviation leading to nasal obstruction in Beesat Hospital during one year (2012-2013. Middle ear pressure and Eustachian tube function on the septal deviated side and contralateral side before and after septoplasty (3 to 6 months later were measured through tympanometry and Eustachian tube function test (Toynbee test. The comparison between pre- and postoperative ETF tests and middle ear pressures was assessed using Paired –T test and p-value of less than 0.05 was considered as statistically significant.   Results : This study revealed that comparison between mean values of pressure in the deviated and contralateral side has no significant statistical difference before and after septoplasty. Also comparison between Eustachian function in the deviated side and contralateral side showed no significant difference before and after septoplasty.

  9. Pressure Drop Correlations of Single-Phase and Two-Phase Flow in Rolling Tubes

    International Nuclear Information System (INIS)

    Xia-xin Cao; Chang-qi Yan; Pu-zhen Gao; Zhong-ning Sun

    2006-01-01

    A series of experimental studies of frictional pressure drop for single phase and two-phase bubble flow in smooth rolling tubes were carried out. The tube inside diameters were 15 mm, 25 mm and 34.5 mm respectively, the rolling angles of tubes could be set as 10 deg. and 20 deg., and the rolling periods could be set as 5 s, 10 s and 15 s. Combining with the analysis of single-phase water motion, it was found that the traditional correlations for calculating single-phase frictional coefficient were not suitable for the rolling condition. Based on the experimental data, a new correlation for calculating single-phase frictional coefficient under rolling condition was presented, and the calculations not only agreed well with the experimental data, but also could display the periodically dynamic characteristics of frictional coefficients. Applying the new correlation to homogeneous flow model, two-phase frictional pressure drop of bubble flow in rolling tubes could be calculated, the results showed that the relative error between calculation and experimental data was less than ± 25%. (authors)

  10. Experimental investigation of pressure fluctuations caused by a vortex rope in a draft tube

    International Nuclear Information System (INIS)

    Kirschner, O; Ruprecht, A; Göde, E; Riedelbauch, S

    2012-01-01

    In the last years hydro power plants have taken the task of power-frequency control for the electrical grid. Therefore turbines in storage hydro power plants often operate outside their optimum. If Francis-turbines and pump-turbines operate at off-design conditions, a vortex rope in the draft tube can develop. The vortex rope can cause pressure oscillations. In addition to low frequencies caused by the rotation of the vortex rope and the harmonics of these frequencies, pressure fluctuations with higher frequencies can be observed in some operating points too. In this experimental investigation the flow structure and behavior of the vortex rope movement in the draft tube of a model pump-turbine are analyzed. The investigation focuses on the correlation of the pressure fluctuation frequency measured at the draft tube wall with the movement of the vortex rope. The movement of the vortex rope is analyzed by the velocity field in the draft tube which was measured with particle image velocimetry. Additionally, the vortex rope movement has been analyzed with the captures of high-speed-movies from the cavitating vortex rope. Besides the rotation of the vortex rope due to pressure fluctuation with low frequencies the results of the measurement also show a correlation between the rotation of the elliptical or deformed rope cross-section and the higher frequency pressure pulsation. An approximation shows that the frequencies of the pressure fluctuation and the movement of the vortex rope are also connected with the velocity of the flow. Taking into account the size and position of the cavitating vortex core as well as the velocity at the position of the surface of the cavitating vortex core the time-period of the rotation of the vortex core can be approximated. The results show that both, the low frequency pressure fluctuation and the higher frequency pressure fluctuation are correlating with the vortex rope movement. With this estimation, the period of the higher frequency

  11. CANDU pressure tube leak detection by annulus gas dew point measurement. A critical review

    Energy Technology Data Exchange (ETDEWEB)

    Greening, F.R. [CTS-NA, Tiverton, ON (Canada)

    2017-03-15

    In the event of a pressure tube leak from a small through-wall crack during CANDU reactor operations, there is a regulatory requirement - referred to as Leak Before Break (LBB) - for the licensee to demonstrate that there will be sufficient time for the leak to be detected and the reactor shut down before the crack grows to the critical size for fast-uncontrolled rupture. In all currently operating CANDU reactors, worldwide, this LBB requirement is met via continuous dew point measurements of the CO{sub 2} gas circulating in the reactor's Annulus Gas System (AGS). In this paper the historical development and current status of this leak detection capability is reviewed and the use of moisture injection tests as a verification procedure is critiqued. It is concluded that these tests do not represent AGS conditions that are to be expected in the event of a real pressure tube leak.

  12. Replacement of a cracked pressure tube in Bruce GS unit 2

    International Nuclear Information System (INIS)

    Dunn, J.T.

    1982-06-01

    In 1982 February, a primary heat transport system leak was detected in the annulus gas system by on-line instrumentation. The source of the leak was found to be a small axial crack in the pressure tube of fuel channel X-14. This fuel channel was removed and replaced by station maintenance staff, and the unit was returned to service five weeks after it had been shut down. The cracked pressure tube was sent to Chalk River Nuclear Laboratories for examination, and the crack was found to be very similar to those found in Pickering GS units 3 and 4 in 1974-75. It was caused by delayed hydride cracking during the period of high residual stress between the time of rolling and the pre-service stress relief

  13. CANDU pressure tube leak detection by annulus gas dew point measurement. A critical review

    International Nuclear Information System (INIS)

    Greening, F.R.

    2017-01-01

    In the event of a pressure tube leak from a small through-wall crack during CANDU reactor operations, there is a regulatory requirement - referred to as Leak Before Break (LBB) - for the licensee to demonstrate that there will be sufficient time for the leak to be detected and the reactor shut down before the crack grows to the critical size for fast-uncontrolled rupture. In all currently operating CANDU reactors, worldwide, this LBB requirement is met via continuous dew point measurements of the CO_2 gas circulating in the reactor's Annulus Gas System (AGS). In this paper the historical development and current status of this leak detection capability is reviewed and the use of moisture injection tests as a verification procedure is critiqued. It is concluded that these tests do not represent AGS conditions that are to be expected in the event of a real pressure tube leak.

  14. Delayed hydride cracking in Zr-2.5% wt Nb pressure tubes

    International Nuclear Information System (INIS)

    Cirimello, Pablo; Haddad, Roberto; Domizzi, Gladys

    2003-01-01

    During service, pressure tubes of CANDU nuclear power reactor are prone to suffer crack growth by delayed hydride cracking (DHC). For a given H 2 plus D 2 concentration there is a critical temperature (T c ) below which DHC may occur. In this work, T c was measured for CCT specimens cut from Zr-2.5 Wt % Nb pressure tubes. Hydrogen was added to the specimens to get concentrations of 40, 59 and 72 ppm. It was found that T c is higher than the corresponding precipitation temperature. The axial crack velocity (V p ) was also measured. Decreasing temperature from T c makes V p increase until a maximum is attained at a temperature close to precipitation temperature. At lower temperatures, in the presence of precipitated hydrides, decreasing temperature implies lower velocities, following an Arrhenius law: Vp=Aexp(-Q/RT), with an activation energy Q= 66 KJ/mol K. (author)

  15. Evaluation of a leaking crack in an irradiated CANDU pressure tube

    International Nuclear Information System (INIS)

    Coleman, C.E.; Simpson, L.A.

    1988-06-01

    Leak-before-break is used in CANDU reactors as part of the defence against rupture of the pressure tubes. Two important features of this technique are the action time available for detection of a leaking crack and the size of the leak allowing crack location. Support for continued reliance on leak-before-break is being obtained from experiments, on irradiated Zr-2.5 Nb pressure tubes attached to their end fittings, that simulate the behaviour of a leaking crack in a reactor. At reactor operating temperatures leaking cracks grow more slowly than dry cracks in the laboratory because they are cooled when pressurised water flashes to steam on their surface. These cracks remain stable till they are at least 70 mm long. From the results of these experiments the action time is at least 100 h. The leak rate increases rapidly when a through-wall crack extends a small amount, thus greatly assisting with crack location

  16. The Study of Heat Treatment Effects on Chromium Carbide Precipitation of 35Cr-45Ni-Nb Alloy for Repairing Furnace Tubes

    Directory of Open Access Journals (Sweden)

    Nakarin Srisuwan

    2016-01-01

    Full Text Available This paper presents a specific kind of failure in ethylene pyrolysis furnace tubes. It considers the case in which the tubes made of 35Cr-45Ni-Nb high temperature alloy failed to carburization, causing creep damage. The investigation found that used tubes became difficult to weld repair due to internal carburized layers of the tube. The microstructure and geochemical component of crystallized carbide at grain boundary of tube specimens were characterized by X-ray diffractometer (XRD, scanning electron microscopy (SEM with back-scattered electrons mode (BSE, and energy dispersive X-ray spectroscopy (EDS. Micro-hardness tests was performed to determine the hardness of the matrix and the compounds of new and used tube material. The testing result indicated that used tubes exhibited a higher hardness and higher degree of carburization compared to those of new tubes. The microstructure of used tubes also revealed coarse chromium carbide precipitation and a continuous carbide lattice at austenite grain boundaries. However, thermal heat treatment applied for developing tube weld repair could result in dissolving or breaking up chromium carbide with a decrease in hardness value. This procedure is recommended to improve the weldability of the 35Cr-45Ni-Nb used tubes alloy.

  17. Effects of Oxidation and fractal surface roughness on the wettability and critical heat flux of glass-peened zirconium alloy tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; Nitheanandan, T.; Bullock, C.D.; Slater, L.F.; McRae, G.A.

    2003-05-01

    Glass-bead peening the outside surfaces of zirconium alloy tubes has been shown to increase the Critical Heat Flux (CHF) in pool boiling of water. The CHF is found to correlate with the fractal roughness of the metal tube surfaces. In this study on the effect of oxidation on glass-peened surfaces, test measurements for CHF, surface wettability and roughness have been evaluated using various glass-peened and oxidized zirconium alloy tubes. The results show that oxidation changes the solid-liquid contact angle (i.e., decreases wettability of the metal-oxide surface), but does not change the fractal surface roughness, appreciably. Thus, oxidation of the glass-peened surfaces of zirconium alloy tubes is not expected to degrade the CHF enhancement obtained by glass-bead peening. (author)

  18. The formation and characteristics of hydride blisters in c.w. Zircaloy-2 pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Price, E G [ed.

    1994-09-01

    Under the auspices of the IAEA, a consultants` meeting was arranged in Vienna, 1994 July 25-29, at which a Canadian delegation, consisting of AECL and Ontario Hydro Technologies personnel, presented information on their knowledge of the behaviour of hydride blisters in Zircaloy-2 pressure tubes. This document contains the 10 papers presented by the Canadian delegation to the meeting. It is believed that they represent a good reference document on hydride blister phenomena.

  19. Remote-controlled television for locating leaking tubes in pressurized-water reactor steam generators

    International Nuclear Information System (INIS)

    Cormault, P.; Denis, J.

    1978-01-01

    The Scarabee system is designed for observation of the tubes in water boxes of pressurized-water reactor nuclear-power-station steam generators. It consists essentially of a camera and a projector used as a marker, both of which swivel freely. The whole unit is housed in a water-tight container which can easily be decontaminated. Remote control of camera and marker movement is carried out from a console. (author)

  20. Benchmark of WIMS-IST against MCNP for CANDU pressure tube fast fluxes

    International Nuclear Information System (INIS)

    Donders, R.E.; Douglas, S.R.

    2002-01-01

    Pressure tube fast-flux data in CANDU are currently calculated using the multi-group neutron transport code WIMS-IST. In this study, the WIMS-IST fast flux calculations are benchmarked against MCNP calculations (a Monte Carlo particle transport code), over the range of fuel burnup and coolant density in CANDU. The comparison shows good agreement between WIMS and MCNP, with WIMS fast fluxes being 1.5% to 4% lower than the MCNP values. The difference is smallest for fresh fuel, and increases with burnup. The fast flux gradient across the pressure tube (factor of 1.23 from inner edge to outer edge) is accurately calculated by WIMS. When reporting fast fluxes in pressure tubes, these are generally given as >1.000 MeV fluxes. For WIMS, this requires an extra conversion step, since the WIMS ENDF/B libraries do not have a group boundary at 1 MeV. The conversion step is based on a fictitious isotope ONEMEV in the WIMS nuclear data library. The conversion factor in WIMS was found to be about one percent too high. When providing >1 MeV fluxes from WIMS, this partially compensates for the slight under prediction of the fast flux. Pressure tube >1 MeV fluxes from WIMS are therefore 0.5% to 3% lower than MCNP values. To obtain accurate fast flux data, neutron transport calculations must be performed on a critical cell. For this study, all calculations were performed with radial albedo boundary conditions giving a critical cell. This required the use of an albedo version of MCNP, developed at AECL. (author)

  1. Some characteristics of the digitization pulses from high pressure neon-helium flash tubes

    International Nuclear Information System (INIS)

    Chan, D.S.K.; Leung, S.K.; Ng, L.K.

    1979-01-01

    Characteristics of the digitization output pulses from high pressure neon-helium flash tubes were studied under various operation conditions using square ultra-high voltage pulses. Properties reported by previous workers were compared. Two discharge mechanisms, the Townsend avalanche discharge and the streamer discharge, were observed to occur in sequence in some events. The output waveforms for both discharge mechanisms were studied in detail. The charge induced on a detecting probe was also estimated from the measured data. (Auth.)

  2. Non-destructive evaluation of stream generator tubes and pressure tubes from the PHWR reactors, using the rotating magnetic field method

    International Nuclear Information System (INIS)

    Premel, D.; Placko, D.; Grimberg, R.; Savin, A.

    2001-01-01

    This work presents a new type of eddy current transducer with a rotating magnetic field devoted to the inspection of steam generator tubes and pressure tubes from the PHWR reactors. A theoretical model has been developed that permits the calculations of the emf induced in the reception coils in the presence of the copper or magnetite deposits, anti-vibration railing and garter springs. (authors)

  3. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  4. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L.

    1997-01-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis

  5. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2011-07-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  6. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

    2011-01-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  7. Microstructure, elastic deformation behavior and mechanical properties of biomedical β-type titanium alloy thin-tube used for stents.

    Science.gov (United States)

    Tian, Yuxing; Yu, Zhentao; Ong, Chun Yee Aaron; Kent, Damon; Wang, Gui

    2015-05-01

    Cold-deformability and mechanical compatibility of the biomedical β-type titanium alloy are the foremost considerations for their application in stents, because the lower ductility restricts the cold-forming of thin-tube and unsatisfactory mechanical performance causes a failed tissue repair. In this paper, β-type titanium alloy (Ti-25Nb-3Zr-3Mo-2Sn, wt%) thin-tube fabricated by routine cold rolling is reported for the first time, and its elastic behavior and mechanical properties are discussed for the various microstructures. The as cold-rolled tube exhibits nonlinear elastic behavior with large recoverable strain of 2.3%. After annealing and aging, a nonlinear elasticity, considered as the intermediate stage between "double yielding" and normal linear elasticity, is attributable to a moderate precipitation of α phase. Quantitive relationships are established between volume fraction of α phase (Vα) and elastic modulus, strength as well as maximal recoverable strain (εmax-R), where the εmax-R of above 2.0% corresponds to the Vα range of 3-10%. It is considered that the "mechanical" stabilization of the (α+β) microstructure is a possible elastic mechanism for explaining the nonlinear elastic behavior. Copyright © 2015 Elsevier Ltd. All rights reserved.

  8. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    Science.gov (United States)

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  9. Creep behavior under internal pressure of zirconium alloy cladding oxidized in steam at high temperature

    International Nuclear Information System (INIS)

    Chosson, Raphael

    2014-01-01

    During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes creep under internal pressure and are oxidized on their outer surface at high temperature (HT). Claddings become stratified materials: zirconia and oxygen-stabilized α phase, called α(O), are formed on the outer surface of the cladding whereas the inner part remains in the β domain. The strengthening effect of oxidation on the cladding creep behavior under internal pressure has been highlighted at HT. In order to model this effect, the creep behavior of each layer had to be determined. This study focused on the characterization of the creep behavior of the α(O) phase at HT, through axial creep tests performed under vacuum on model materials, containing from 2 to 7 wt.% of oxygen and representative of the α(O) phase. For the first time, two creep flow regimes have been observed in this phase. Underlying physical mechanisms and relevant microstructural parameters have been discussed for each regime. The strengthening effect due to oxygen on the α(O) phase creep behavior at HT has been quantified and creep flow equations have been identified. A ductile to brittle transition criterion has been also suggested as a function of temperature and oxygen content. Relevance of the creep flow equations for each layer, identified in this study or from the literature, has been discussed. Then, a finite element model, describing the oxidized cladding as a stratified material, has been built. Based on this model, a fraction of the experimental strengthening during creep is predicted. (author) [fr

  10. Nb effect on Zr-alloy oxidation under high pressure steam at high temperatures

    International Nuclear Information System (INIS)

    Park, Kwangheon; Yang, Sungwoo; Kim, Kyutae

    2005-01-01

    The high-pressure steam effects on the oxidation of Zircaloy-4 (Zry-4) and Zirlo (Zry-1%Nb) claddings at high temperature have been analyzed. Test temperature range was 700-900degC, and pressures were 1-150 bars. High pressure-steam enhances oxidation of Zry-4, and the dependency of enhancement looks exponential to steam pressure. The origin of the oxidation enhancement turned out to be the formation of cracks in oxide. The loss of tetragonal phase by high-pressure steam seems related to the crack formation. Addition of Nb as an alloying element to Zr alloy reduces significantly the steam pressure effects on oxidation. The higher compressive stresses and the smaller fraction of tetragonal oxides in Zry-1%Nb seem to be the diminished effect of high-pressure steam on oxidation. (author)

  11. Anisotropic thermal creep of internally pressurized Zr-2.5Nb tubes

    International Nuclear Information System (INIS)

    Li, W.; Holt, R.A.

    2010-01-01

    The anisotropy of creep of internally pressurized cold-worked Zr-2.5Nb tubes with different crystallographic textures is reported. The stress exponent n was determined to be about three at transverse stresses from 100 to 250 MPa with an activation energy of ∼99.54 kJ/mol in the temperature range 300-400 o C. The stress exponent increased to ∼6 for transverse stresses from 250 to 325 MPa. From this data an experimental regime of 350 o C and 300 MPa was established in which dislocation glide is the likely strain-producing mechanism. Creep tests were carried out under these conditions on internally pressurized Zr-2.5Nb tubes with 18 different textures. Creep strain and creep anisotropy (ratio of axial to transverse steady-state creep rate, ε . A /ε . T ) exhibited strong dependence on crystallographic textures of the Zr-2.5Nb tubes. It was found that the values of (ε . A /ε . T ) increased as the difference between the resolved faction of basal plane normals in the transverse and radial directions (f T - f R ) increases. The tubes with the strongest radial texture showed a negative axial creep strain and a negative creep rate ratio (ε . A /ε . T ) and tubes with a strong transverse texture exhibited the positive values of steady-state creep rate ratio (ε . A /ε . T ) and good creep resistance in the transverse direction. These behaviors are qualitatively similar to those observed during irradiation creep, and also to the predictions of polycrystalline models for creep in which glide is the strain-producing mechanism and prismatic slip is the dominant system. A detailed analysis of the results using polycrystalline models may assist in understanding the anisotropy of irradiation creep.

  12. Flow instability research on steam generator with straight double-walled heat transfer tube for FBR. Pressure drop under high pressure condition

    International Nuclear Information System (INIS)

    Liu, Wei; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

    2008-01-01

    For the Steam Generator (SG) with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments were performed under high pressure condition in JAEA with using a straight tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper evaluates the pressure drop data with TRAC-BF1 code. The Pffan's correlation for single phase flow and the Martinelli-Nelson's two-phase flow multiplier are found can be well predicted the present pressure drop data under high pressure condition. (author)

  13. Adsorbate induced surface alloy formation investigated by near ambient pressure X-ray photoelectron spectroscopy

    DEFF Research Database (Denmark)

    Nierhoff, Anders Ulrik Fregerslev; Conradsen, Christian Nagstrup; McCarthy, David Norman

    2014-01-01

    for engineering of more active or selective catalyst materials. Dynamical surface changes on alloy surfaces due to the adsorption of reactants in high gas pressures are challenging to investigate using standard characterization tools. Here we apply synchrotron illuminated near ambient pressure X-ray photoelectron...

  14. Pressure-induced structural change in liquid GaIn eutectic alloy

    DEFF Research Database (Denmark)

    Yu, Q.; Ahmad, A. S.; Ståhl, Kenny

    2017-01-01

    Synchrotron x-ray diffraction reveals a pressure induced crystallization at about 3.4 GPa and a polymorphic transition near 10.3 GPa when compressed a liquid GaIn eutectic alloy up to ~13 GPa at room temperature in a diamond anvil cell. Upon decompression, the high pressure crystalline phase...

  15. 78 FR 63164 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania...

    Science.gov (United States)

    2013-10-23

    ... Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania: Final Results of Antidumping... carbon and alloy seamless standard, line and pressure pipe from Romania. For the final results we... pressure pipe from Romania.\\1\\ We invited interested parties to comment on the Preliminary Results. We...

  16. 78 FR 41369 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania...

    Science.gov (United States)

    2013-07-10

    ... Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania: Preliminary Results of..., line and pressure pipe (small diameter seamless pipe) from Romania. The period of review (POR) is... and Alloy Seamless Standard, Line and Pressure Pipe from Romania,'' dated concurrently with this...

  17. Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nikel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) seamless pipe and tube

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nikel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) seamless pipe and tube

  18. Statistical analysis and modelling of in-reactor diametral creep of Zr-2.5Nb pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jyrkama, Mikko I., E-mail: mjyrkama@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada); Bickel, Grant A., E-mail: grant.bickel@cnl.ca [Canadian Nuclear Laboratories, Chalk River Laboratories, Chalk River, ON, Canada K0J 1J0 (Canada); Pandey, Mahesh D., E-mail: mdpandey@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada)

    2016-04-15

    Highlights: • New and simple statistical model of pressure tube diametral creep. • Based on surveillance data of 328 pressure tubes from eight different CANDU reactors. • Uses weighted least squares (WLS) to regress out operating conditions. • The shape of the diametral creep profiles are predicted very well. • Provides insight and relative ranking of strain behaviour of in-service tubes. - Abstract: This paper presents the development of a simplified regression approach for modelling the diametral creep over time in Zr-2.5 wt% Nb pressure tubes used in CANDU reactors. The model is based on a large dataset of in-service inspection data of 328 different pressure tubes from eight different CANDU reactor units. The proposed weighted least squares (WLS) regression model is linear in time as a function of flux and temperature, with a temperature-dependent variance function. The model predicts the shape of the observed diametral creep profiles very well, and is useful not merely for prediction, but also for assessing tube-to-tube variability and manufacturing properties among the inspected tubes.

  19. Synthesis and characterization of Ni-Mo filler brazing alloy for Mo-W joining for microwave tube technology

    Directory of Open Access Journals (Sweden)

    Frank Ferrer Sene

    2013-04-01

    Full Text Available A brazing process based on Ni-Mo alloy was developed to join porous tungsten cathode bottom and dense molybdenum cathode body for microwave tubes manufacture. The Ni-Mo alloy was obtained by mixing and milling powders in the eutectic composition, and applied on the surface of the components. The brazing was made at 1400 °C by using induction heating in hydrogen for 5 minutes. Alumina surfaces were coated with the binder and analyzed by Energy Dispersive X-rays Fluorescence. The brazed samples were analyzed by Scanning Electron Microscopy coupled to Energy Dispersive Spectroscopy. Stress-strain tests were performed to determine the mechanical behavior of the joining. The quality of the brazing was evaluated by assuring the presence of a "meniscus" formed by the Ni-Mo alloy on the border of the tungsten and molybdenum joint, the absence of microstructural defects in the interface between the tungsten and molybdenum alloys, and the adhesion of the brazed components.

  20. Heat transfer and carryover of low pressure water in a heated vertical tube

    International Nuclear Information System (INIS)

    Smith, T.A.

    1976-01-01

    Local heat transfer coefficients in the stable film boiling and dispersed flow regimes were studied for the upward flow of low pressure water in a heated vertical tube. Wall temperatures were maintained constant with time and along the tube so that both axial and time temperature gradients approached zero. Heat flux along the tube was not constant but was applied so as to maintain a steady state temperature profile. A preheater was used to bring the liquid to saturation before it entered the main portion of the test section and in some cases the equilibrium quality was greater than zero at the entrance to the main test section. The test section was made of stainless steel, and the lower portion, the preheater, was heated directly by dc current. Copper block heat spikes were clamped to the upper test section and were used to apply the heat flux to maintain the wall temperature constant with time. Several theories for the different possible types of flow (laminar or turbulent, tube or film) were compared with the experimental data. The carry-over point for low flooding rates (1 inch/sec or less) was inferred from these comparisons and gave good agreement with the Plummer critical mass criterion for liquid carry-over

  1. Boiling heat transfer and dryout in helically coiled tubes under different pressure conditions

    International Nuclear Information System (INIS)

    Chung, Young-Jong; Bae, Kyoo-Hwan; Kim, Keung Koo; Lee, Won-Jae

    2014-01-01

    Highlights: • Heat transfer characteristics and dryout for helically coiled tube are performed. • A boiling heat transfer tends to increase with a pressure increase. • Dryout occurs at high quality test conditions investigated. • Steiner–Taborek’s correlation is predicted well based on the experimental results. - Abstract: A helically coiled once-through steam generator has been used widely during the past several decades for small nuclear power reactors. The heat transfer characteristics and dryout conditions are important to optimal design a helically coiled steam generator. Various experiments with the helically coiled tubes are performed to investigate the heat transfer characteristics and occurrence condition of a dryout. For the investigated experimental conditions, Steiner–Taborek’s correlation is predicted reasonably based on the experimental results. The pressure effect is important for the boiling heat transfer correlation. A boiling heat transfer tends to increase with a pressure increase. However, it is not affected by the pressure change at a low power and low mass flow rate. Dryout occurs at high quality test conditions investigated because a liquid film on the wall exists owing to a centrifugal force of the helical coil

  2. Complete Status Report Documenting Development of Friction Stir Welding for Joining Thin Wall Tubing of ODS Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bunn, Jeffrey R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gussev, Maxim N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    The development of friction stir welding (FSW) for joining thin sections of the advanced oxide dispersion strengthened (ODS) 14YWT ferritic alloy was initiated in Fuel Cycle Research and Development (FCRD), now the Nuclear Technology Research and Development (NTRD), in 2015. The first FSW experiment was conducted in late FY15 and successfully produced a bead-on-plate stir zone (SZ) on a 1 mm thick plate of 14YWT (SM13 heat). The goal of this research task is to ultimately demonstrate that FSW is a feasible method for joining thin wall (0.5 mm thick) tubing of 14YWT.

  3. Penetration of hydrogen isotopes through EhI 698 alloy at high pressure and temperature

    International Nuclear Information System (INIS)

    Bystritskij, V.M.; Voznyak, Ya.; Granovskij, V.B.

    1986-01-01

    The paper deals with investigations of the process of hydrogen and deuterium penetration through the high-temperature alloy EhI-698 at a pressure up to 1 kbar and temperature up to 1050 K. Parameters of the process obey Sieverts's law and can be described by Arrenius's and Vant-Goff's equations. The obtained results lead to a conclusion that the alloy EhI-698 is good for vessels to be employed in hydrogen media

  4. Experiments of draining and filling processes in a collapsible tube at high external pressure

    Science.gov (United States)

    Flaud, P.; Guesdon, P.; Fullana, J.-M.

    2012-02-01

    The venous circulation in the lower limb is mainly controlled by the muscular action of the calf. To study the mechanisms governing the venous draining and filling process in such a situation, an experimental setup, composed by a collapsible tube under external pressure, has been built. A valve preventing back flows is inserted at the bottom of the tube and allows to model two different configurations: physiological when the fluid flow is uni-directional and pathological when the fluid flows in both directions. Pressure and flow rate measurements are carried out at the inlet and outlet of the tube and an original optical device with three cameras is proposed to measure the instantaneous cross-sectional area. The experimental results (draining and filling with physiological or pathological valves) are confronted to a simple one-dimensional numerical model which completes the physical interpretation. One major observation is that the muscular contraction induces a fast emptying phase followed by a slow one controlled by viscous effects, and that a defect of the valve decreases, as expected, the ejected volume.

  5. Current Status of Development of High Nickel Low Alloy Steels for Commercial Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Chul; Lee, B. S.; Park, S. G.; Lee, K. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    SA508 Gr.3 Mn-Mo-Ni low alloy steels have been used for nuclear reactor pressure vessel steels up to now. Currently, the design goal of nuclear power plant is focusing at larger capacity and longer lifetime. Requirements of much bigger pressure vessels may cause critical problems in the manufacturing stage as well as for the welding stage. Application of higher strength steel may be required to overcome the technical problems. It is known that a higher strength and fracture toughness of low alloy steels such as SA508 Gr.4N low alloy steel could be achieved by increasing the Ni and Cr contents. Therefore, SA508 Gr.4N low alloy steel is very attractive as eligible RPV steel for the next generation PWR systems. In this report, we propose the possibility of SA508 Gr.4N low alloy steel for an application of next generation commercial RPV, based on the literature research result about development history of the RPV steels and SA508 specification. In addition, we have surveyed the research result of HSLA(High Strength Low Alloy steel), which has similar chemical compositions with SA508 Gr.4N, to understand the problems and the way of improvement of SA508 Gr.4N low alloy steel. And also, we have investigated eastern RPV steel(WWER-1000), which has higher Ni contents compared to western RPV steel.

  6. Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

    1989-01-01

    Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm 2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

  7. Derivation of Elastic Stress Concentration Factor Equations for Debris Fretting Flaws in Pressure Tubes of Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Jong Sung; Oh, Young Jin

    2014-01-01

    If volumetric flaws such as bearing pad fretting flaws and debris fretting flaws are detected in the pressure tubes of pressurized heavy water reactors during in-service inspection, the initiation of fatigue cracks and delayed hydrogen cracking from the detected volumetric flaws shall be assessed by using elastic stress concentration factors in accordance with CSA N285.8-05. The CSA N285.8-05 presents only an approximate formula based on linear elastic fracture mechanics for the debris fretting flaw. In this study, an engineering formula considering the geometric characteristics of the debris fretting flaw in detail was derived using two-dimensional finite element analysis and Kinectrics, Inc.'s engineering procedure with slight modifications. Comparing the application results obtained using the derived formula with the three-dimensional finite element analysis results, it is found that the results obtained using the derived formula agree well with the results of the finite element analysis

  8. Stress relief to prevent stress corrosion in the transition region of expanded Alloy 600 steam-generator tubing. Final report

    International Nuclear Information System (INIS)

    Woodward, J.; van Rooyen, D.

    1983-05-01

    The feasibility of preventing primary side roll transition cracking has been investigated, using induction heating to attain stress relief of expanded Ni-Cr-Fe Alloy 600 steam generator tubing. Work on rolled tubing and U-bends has shown that temperatures with which stress relief can be obtained range from 700 to 850 0 C, with lower temperatures in this range requiring longer times at temperature to provide the requisite reduction in residual stresses. No work has yet been done outside this range. Preliminary tests, using induction heating, have been carried out on a mock tube sheet assembly, designed to the dimensions of a typical steam generator, and have identified the type of heating/cooling cycle that would occur in the tube sheet during a stress relief operation. Preliminary results show that the times to reach the higher temperatures in the range observed to give stress relief, of the order of 850 0 C, can be as short as 8 seconds, and less with optimum coil design and power control

  9. MEASUREMENT OF ENDOTRACHEAL TUBE CUFF PRESSURE IN MECHANICALLYVENTILATED PATIENTS ON ARRIVAL TO INTENSIVE CARE UNIT - A CROSS-SECTIONAL STUDY

    Directory of Open Access Journals (Sweden)

    Arun Kumar Ajjappa

    2017-04-01

    Full Text Available BACKGROUND The monitoring of Endotracheal Tube (ETT cuff pressure in intubated patients on arrival to intensive care unit is very essential. The cuff pressure must be within an optimal range of 20-30cm H2O ensuring ventilation with no complications related to cuff overinflation and underinflation. This can be measured with a cuff pressure manometer. The aim of the study is to measure the endotracheal tube cuff pressure in patients on arrival to intensive care unit and to identify prevalence of endotracheal cuff underinflation and overinflation. MATERIALS AND METHODS A cross-sectional study was done on mechanically-ventilated patients who were intubated in casualty (emergency department on arrival to intensive care unit in S.S. Institute of Medical Sciences and Research Centre, Davangere. About 50 critically-ill patients intubated with a high volume, low pressure endotracheal tube were included in the study. An analogue manometer was used to measure the endotracheal tube cuff pressure. It was compared with the recommended level. The settings of mechanical ventilation, endotracheal tube size and peak airway pressure were recorded. RESULTS It was found that the mean cuff pressure was 64.10 cm of H2O with a standard deviation of 32.049. Of the measured cuff pressures, only 2% had pressures within an optimal range (20-30cm of H2O. 88% had cuff pressures more than 30cm of H2O. The mean peak airway pressure found to be 20.50cm of H2O with a Standard Deviation (SD of 5.064. CONCLUSION This study is done to emphasise the importance of cuff pressure measurement in all mechanically-ventilated patients as cuff pressure is found to be high in most of the patients admitted to intensive care unit. Complications of overinflation and underinflation can only be prevented if the acceptable cuff pressures are achieved.

  10. Denting of Inconel steam generator tubes in pressurized water reactors. Third informal report

    International Nuclear Information System (INIS)

    van Rooyen, D.; Weeks, J.R.

    1977-08-01

    The recent plant operating experience and laboratory test results on the phenomenon of denting in recirculating PWR steam generators is reviewed. Although denting was first reported only in plants that were converted from phosphate to AVT, it has now also been observed in plants still on phosphate, as well as in some that started on AVT. In some units, slightly abnormal eddy current signals have been observed at the top of the tube sheets. The degree of denting in operating steam generators may be related to the levels and duration of chloride inleakage. Chloride, however, is not the only active ingredient, and does not seem to give denting until local acid conditions arise; consequently, it may be necessary for soluble copper and/or nickel ions to be present to promote the denting reaction. Chloride concentrations in actively corroding crevices can increase by several orders of magnitude over the bulk coolant. It is thus difficult to develop a basis for Cl - specifications for secondary water. Maintaining Cl - low enough to prevent denting may be unmanageable without full flow condensate demineralization in coastal plants with copper alloy condensors and feedwater lines. Cathodic depolarization by oxidizing species are thought to promote the formation of acid chlorides in crevices and trigger the denting reactions; some ions may also catalyze the rapid formation of magnetite. These, and other mechanistic aspects of denting are discussed. The implications of the Inconel 600 tube defects at Ginna in non-dented areas, originating from the primary side, are also discussed

  11. Eustachian tube function and middle ear barotrauma associated with extremes in atmospheric pressure.

    Science.gov (United States)

    Miyazawa, T; Ueda, H; Yanagita, N

    1996-11-01

    Eustachian tube (ET) function was studied by means of sonotubometry and tubotympano-aerodynamography (TTAG) prior to and following exposure to hypobaric or hyperbaric conditions. Forty normal adults were subjected to hypobaric pressure. Fifty adults who underwent hyperbaric oxygen (HBO) therapy also were studied. Following hypobaric exposure, 14 of 80 ears (17.5%) exhibited middle ear barotrauma. Following hyperbaric exposure, 34 of 100 ears (34%) exhibited middle ear barotrauma. Dysfunction of the ET, characterized by altered active and passive opening capacity, was more prevalent following exposure to extremes in atmospheric pressure compared to baseline. The ET function, which was impaired after the first HBO treatment, improved gradually over the next 2 hours. Overall, however, ET function was worse after the seventh treatment. The patients who developed barotrauma exhibited worse ET function prior to hypobaric or hyperbaric exposure. Thus, abnormal ET function can be used to predict middle ear barotrauma prior to exposure to hypobaric or hyperbaric atmospheric pressure.

  12. Nonlinear vacuum gas flow through a short tube due to pressure and temperature gradients

    Energy Technology Data Exchange (ETDEWEB)

    Pantazis, Sarantis; Naris, Steryios; Tantos, Christos [Department of Mechanical Engineering, University of Thessaly, Pedion Areos, 38334 Volos (Greece); Valougeorgis, Dimitris, E-mail: diva@mie.uth.gr [Department of Mechanical Engineering, University of Thessaly, Pedion Areos, 38334 Volos (Greece); André, Julien; Millet, Francois; Perin, Jean Paul [Service des Basses Températures, UMR-E CEA/UJF-Grenoble 1, INAC, Grenoble, F-38054 (France)

    2013-10-15

    The flow of a rarefied gas through a tube due to both pressure and temperature gradients has been studied numerically. The main objective is to investigate the performance of a mechanical vacuum pump operating at low temperatures in order to increase the pumped mass flow rate. This type of pump is under development at CEA-Grenoble. The flow is modelled by the Shakhov kinetic model equation, which is solved by the discrete velocity method. Results are presented for certain geometry and flow parameters. Since according to the pump design the temperature driven flow is in the opposite direction than the main pressure driven flow, it has been found that for the operating pressure range studied here the net mass flow rate through the pump may be significantly reduced.

  13. Nonlinear vacuum gas flow through a short tube due to pressure and temperature gradients

    International Nuclear Information System (INIS)

    Pantazis, Sarantis; Naris, Steryios; Tantos, Christos; Valougeorgis, Dimitris; André, Julien; Millet, Francois; Perin, Jean Paul

    2013-01-01

    The flow of a rarefied gas through a tube due to both pressure and temperature gradients has been studied numerically. The main objective is to investigate the performance of a mechanical vacuum pump operating at low temperatures in order to increase the pumped mass flow rate. This type of pump is under development at CEA-Grenoble. The flow is modelled by the Shakhov kinetic model equation, which is solved by the discrete velocity method. Results are presented for certain geometry and flow parameters. Since according to the pump design the temperature driven flow is in the opposite direction than the main pressure driven flow, it has been found that for the operating pressure range studied here the net mass flow rate through the pump may be significantly reduced

  14. Effect of superficial velocity on vaporization pressure drop with propane in horizontal circular tube

    Science.gov (United States)

    Novianto, S.; Pamitran, A. S.; Nasruddin, Alhamid, M. I.

    2016-06-01

    Due to its friendly effect on the environment, natural refrigerants could be the best alternative refrigerant to replace conventional refrigerants. The present study was devoted to the effect of superficial velocity on vaporization pressure drop with propane in a horizontal circular tube with an inner diameter of 7.6 mm. The experiments were conditioned with 4 to 10 °C for saturation temperature, 9 to 20 kW/m2 for heat flux, and 250 to 380 kg/m2s for mass flux. It is shown here that increased heat flux may result in increasing vapor superficial velocity, and then increasing pressure drop. The present experimental results were evaluated with some existing correlations of pressure drop. The best prediction was evaluated by Lockhart-Martinelli (1949) with MARD 25.7%. In order to observe the experimental flow pattern, the present results were also mapped on the Wang flow pattern map.

  15. Monte Carlo Study on Gas Pressure Response of He-3 Tube in Neutron Porosity Logging

    Directory of Open Access Journals (Sweden)

    TIAN Li-li;ZHANG Feng;WANG Xin-guang;LIU Jun-tao

    2016-10-01

    Full Text Available Thermal neutrons are detected by (n,p reaction of Helium-3 tube in the compensated neutron logging. The helium gas pressure in the counting area influences neutron detection efficiency greatly, and then it is an important parameter for neutron porosity measurement accuracy. The variation law of counting rates of a near detector and a far one with helium gas pressure under different formation condition was simulated by Monte Carlo method. The results showed that with the increasing of helium pressure the counting rate of these detectors increased firstly and then leveled off. In addition, the neutron counting rate ratio and porosity sensitivity increased slightly, the porosity measurement error decreased exponentially, which improved the measurement accuracy. These research results can provide technical support for selecting the type of Helium-3 detector in developing neutron porosity logging.

  16. Direct numerical simulation of heat transfer to CO2 at supercritical pressure in a vertical tube

    International Nuclear Information System (INIS)

    Bae, Joong-Hun; Yoo, Jung-Yul; Choi, Hae-Cheon

    2003-01-01

    In the present study, the turbulent heat transfer to CO 2 at supercritical pressure in a vertical tube is investigated using Direct Numerical Simulation (DNS), where no turbulence model is adopted. Heat transfer to the supercritical pressure fluids is characterized by rapid variation of thermodynamic/ thermo-physical properties in the fluids. This change in properties occurs within a very narrow range of temperature across the so-called pseudo-critical temperature, causing a peculiar behavior of heat transfer characteristics. The buoyancy effects associated with very large changes in density proved to play a major role in turbulent heat transfer to supercritical pressure fluids. Depending on the degree of buoyancy effects, turbulent heat transfer may increase or significantly decrease, resulting in a local hot spot along the wall. Based on the results of the present DNS study combined with theoretical considerations for turbulent mixed convection heat transfer, the basic mechanism of this local heat transfer deterioration is explained

  17. Application of Deformable Templates for Recognizing Tracks Detected with High Pressure Drift Tubes

    International Nuclear Information System (INIS)

    Baginyan, S.; Baranov, S.; Glazov, A.; Ososkov, G.

    1994-01-01

    The modification of the deformable template method (DTM) application to the problem of track finding and track parameter determination for data detected with high pressure drift tubes (HPDT) in the design of ATLAS for the muon spectrometer experiment is proposed. Our DTM applications consist of two parts, according to two stages of the study. The first part relates to the stage of HPDT study on the CERN muon beam (BEAM-TEST) with the simplest one-prong events without noise signals, where the main obstacle is the left-right ambiguities for each tube. In the second part more complicated HPDT data are to be handled with noise signals. It was shown that the suggested DTM development solves the problem of track recognition and track parameter determination for both noiseless and noise data. Results are obtained on the real beam test data and on data simulating the muon spectrometer on the basis of HPDT. 14 refs., 10 figs

  18. A combined experimental and FE analysis procedure to evaluate tensile behavior of zircaloy pressure tubes

    International Nuclear Information System (INIS)

    Samal, M.K.; Vaze, K.K.; Balakrishnan, K.S.; Anantharaman, S.

    2012-01-01

    Determination of transverse mechanical properties from the ring type of specimens directly machined from the nuclear reactor pressure tubes is not straightforward because of the presence of combined membrane as well as bending stresses arising in the loaded condition. In this work, we have performed ring-tensile tests on the un-irradiated ring tensile specimen using two split semi-cylindrical mandrels as the loading device. A 3-D finite element (FE) analysis was performed in order to determine the material true stress-strain curve by comparing experimental load-displacement data with those predicted by FE analysis. In order to validate the methodology, miniaturized tensile specimens were machined from these tubes and tested. It was observed that the stress-strain data as obtained from ring tensile specimen could describe the load displacement curve of the miniaturized flat tensile specimen very well. (author)

  19. Evaluation of maintenance strategies for steam generator tubes in pressurized waster reactors. 2. Cost and profitability analyses

    International Nuclear Information System (INIS)

    Isobe, Y.; Sagisaka, M.; Yoshimura, S.; Yagawa, G.

    2000-01-01

    As an application of probabilistic fracture mechanics (PFM), risk-benefit analysis was carried out to evaluate maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). The analysis was conducted for SG tubes made of Inconel 600, and Inconel 690 as well assuming its crack initiation and crack propagation law based on Inconel 600 data. The following results were drawn from the analysis. Improvement of inspection accuracy reduces the maintenance costs significantly and is preferable from the viewpoint of profitability due to reduction of SG tube leakage and rupture. There is a certain region of SCC properties of SG tubes where sampling inspection is effective. (author)

  20. Compressibility of Ir-Os alloys under high pressure

    International Nuclear Information System (INIS)

    Yusenko, Kirill V.; Bykova, Elena; Bykov, Maxim; Gromilov, Sergey A.; Kurnosov, Alexander V.; Prescher, Clemens; Prakapenka, Vitali B.; Hanfland, Michael; Smaalen, Sander van; Margadonna, Serena; Dubrovinsky, Leonid S.

    2015-01-01

    Highlights: • fcc- and hcp-Ir-Os alloys were prepared from single-source precursors. • Their atomic volumes measured at ambient conditions using powder X-ray diffraction follow nearly linear dependence. • Compressibility of alloys have been studied up to 30 GPa at room temperature in diamond anvil cells. • Their bulk moduli increase with increasing osmium content. - Abstract: Several fcc- and hcp-structured Ir-Os alloys were prepared from single-source precursors in hydrogen atmosphere at 873 K. Their atomic volumes measured at ambient conditions using powder X-ray diffraction follow nearly linear dependence as a function of composition. Alloys have been studied up to 30 GPa at room temperature by means of synchrotron-based X-ray powder diffraction in diamond anvil cells. Their bulk moduli increase with increasing osmium content and show a deviation from linearity. Bulk modulus of hcp-Ir 0.20 Os 0.80 is identical to that of pure Os (411 GPa) within experimental errors. Peculiarities on fcc-Ir 0.80 Os 0.20 compressibility curve indicate possible changes of its electronic properties at ∼20 GPa

  1. An assessment of the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes of PHWR

    International Nuclear Information System (INIS)

    Sah, D.N.

    1992-01-01

    In view of the deleterious effect of hydriding on the operating life of zircaloy-2 pressure tubes in PHWRs there is an urgent need for the assessment of the status of the pressure tubes with respect to corrosion and hydrogen pick-up in the operating PHWRs. A model has been developed for analysing the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes under reactor operating conditions. This model predicts the axial profiles of oxide layer thickness and hydrogen pick-up in the pressure tubes as a function of the operating time of the reactor. The prediction of hydrogen pick-up by the model in the F-10 pressure tube of RAPS-I have been found to be in good agreement with the measured value of hydrogen content. This report gives a brief description of the model and its predictions on the present status of hydrogen pick-up in the pressure tubes of lead reactor RAPS-II. (author). 6 refs., 5 figs., 2 tabs

  2. Some observations of the pressure distribution in a tube bank for conditions of self generated acoustic resonance

    International Nuclear Information System (INIS)

    Fitzpatrick, J.A.; Donaldson, I.S.; McKnight, W.

    1979-01-01

    The results for mean and fluctuating pressure distributions around tubes in an in-line tube bank are presented for both non-resonant and self-excited acoustic standing wave resonant flow regimes. It is readily deduced that the nature of the flow in the bank is dramatically altered with the onset of acoustic resonance. The velocity gradients which appear across the bank with the onset of resonance would suggest regions of flow recirculation in the bank although no evidence of this was found. The spectra of fluctuating pressure on the duct roof in the bank and on tubes deep in the bank exhibited coherent peaks only during resonance. (author)

  3. Development of In-Service Inspection system for heat transfer tubes in the primary pressurized water cooler in the HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Shinozaki, Masayuki; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Wada, Shigeyuki

    1999-08-01

    The ISI (In-Service Inspection) system has been developed so as to maintain the structural integrity of heat transfer tubes in the primary pressurized water cooler in the HTTR (High Temperature Engineering Test Reactor). This system consists of eddy current probes, ultra-sonic probes, insertion and extraction units, positioning unit and so on. Verification and performance tests of the developed ISI system were carried out using mock-up heat transfer tubes in the primary pressurized water cooler. The constitution of the system, R and D results of the inspection probes, and verification and performance test results of the ISI system for heat transfer tubes are described in this paper. (author)

  4. High pressure thimble/guide tube seal fitting with built-in low pressure seal especially suitable for facilitated and more efficient nuclear reactor refueling service

    International Nuclear Information System (INIS)

    Bhatt, P.N.; Blaushield, R.M.

    1991-01-01

    This patent describes a HP/LP seal arrangement for an elongated guide tube and an elongated thimble disposed therein. The guide tube and thimble extending outwardly from the core of a nuclear reactor to a seal table where the guide tube is welded to the seal table to provide a high pressure seal relative thereto. It comprises: a tubular seal fitting disposed in alignment with the guide tube with the thimble extending therethrough on the low pressure side of the seal table; first high pressure sealing means coupling one end of the fitting to an end of the guide tube to prevent leakage from within the guide tube; inwardly facing thread means disposed adjacent the other and outer end of the seal fitting; a nut having an opening through which the thimble extends and further having outwardly facing threading in mating engagement with the fitting thread means; the fitting having a seal seat spaced longitudinally inwardly from the thread means and facing the fitting outer end and further disposed annularly about the inner surface of the fitting; deformable ring seal means; second releasable high pressure sealing means coupling the thimble to the outer end portion of the guide tube

  5. Microstructural studies on steam oxidised Zr-2.5%Nb pressure tube under simulated LOCA condition

    International Nuclear Information System (INIS)

    Banerjee, Suparna; Sawarn, Tapan K.; Pandit, K.M.; Anantharaman, S.; Srivastava, D.; Sah, D.N.

    2013-03-01

    Study of the microstructural evolution of Zr-2.5%Nb pressure tube material of Indian Pressurized Heavy Water Reactors (PHWRs) due to steam oxidation at high temperature (in the range 500-1050°C) was carried out on pressure tube coupons. Hydrogen pick up was less than 55 ppm in the samples oxidized at temperatures up to 850°C but high (250-400 ppm) in the samples oxidized in the β phase region (900°C and above). The microstructure of the samples oxidized above the α-Zr/β-Zr transition temperature showed from the surface inwards sequentially the presence of an oxide layer, an underlying oxygen stabilized α-Zr layer and a prior β-Zr phase containing hydride precipitates. An increase in the hardness was observed near the oxide-metal interface in the coupons oxidized above 900°C, due to formation of oxygen stabilized α-Zr layer. Higher hardness was also observed in the base metal in the samples oxidized at 1000 and 1050°C (author)

  6. Subchannel flow analysis in Candu and ACR pressure tubes with radial and axial diameter variation

    Energy Technology Data Exchange (ETDEWEB)

    Catana, A.; Prodea, L. [RAAN, Institute for Nuclear Research, Arges (Romania); Danila, N.; Prisecaru, I.; Dupleac, D. [Bucharest Univ. Politehnica(Romania)

    2007-07-01

    The Candu (Canada Deuterium Uranium) and ACR (Advanced Candu Reactor) are pressure tubes (PT) heavy water moderated reactors. Candu are heavy water and ACR are light water cooled reactors. The pressure tube is filled with 12 bundles, each consisting of 37 respectively 43 fuel rods. One Candu reactor is in operation at Cernavoda, Romania since 1996. ACR is a proposed advanced Candu. PT diameter variation has a significant impact on the thermal-hydraulic parameters. Almost all thermal-hydraulic parameters change, but some of them have a greater significance. In this work we have considered a set of radial and axial PT diameter variations both for Candu-600 and ACR-700 reactors using various types of fuel bundles. We can conclude the following: 1) some thermal-hydraulic parameters are significantly influenced: critical heat flux (CHF), pressure drop, or void fraction; 2) the most significant parameter CHF is worsening which reduces the safety margin; 3) some fuel types present a better thermal-hydraulic behavior; and 4) fuel bundles with fresh fuel or low burnup have a worse thermal-hydraulic behaviour than those at average burn-up.

  7. Subchannel flow analysis in Candu and ACR pressure tubes with radial and axial diameter variation

    International Nuclear Information System (INIS)

    Catana, A.; Prodea, L.; Danila, N.; Prisecaru, I.; Dupleac, D.

    2007-01-01

    The Candu (Canada Deuterium Uranium) and ACR (Advanced Candu Reactor) are pressure tubes (PT) heavy water moderated reactors. Candu are heavy water and ACR are light water cooled reactors. The pressure tube is filled with 12 bundles, each consisting of 37 respectively 43 fuel rods. One Candu reactor is in operation at Cernavoda, Romania since 1996. ACR is a proposed advanced Candu. PT diameter variation has a significant impact on the thermal-hydraulic parameters. Almost all thermal-hydraulic parameters change, but some of them have a greater significance. In this work we have considered a set of radial and axial PT diameter variations both for Candu-600 and ACR-700 reactors using various types of fuel bundles. We can conclude the following: 1) some thermal-hydraulic parameters are significantly influenced: critical heat flux (CHF), pressure drop, or void fraction; 2) the most significant parameter CHF is worsening which reduces the safety margin; 3) some fuel types present a better thermal-hydraulic behavior; and 4) fuel bundles with fresh fuel or low burnup have a worse thermal-hydraulic behaviour than those at average burn-up

  8. Experimental study of heat transfer and pressure drops for ammonia flowing inside a long tube

    International Nuclear Information System (INIS)

    Malek, A.; Colin, R.

    1985-01-01

    This report presents the results of the experimental study of heat transfer coefficients and pressure drops for boiling ammonia in a long tube. The scope of the tests discussed here corresponds to temperatures ranging from 30 to 70 0 C. This touches on various forthcoming applications, including binary cycles of nuclear power plants, as well as miscellaneous energy recovery cycles (heat pumps, geothermal energy, etc.). The results reported here of ammonia evaporators in the temperature range mentionned for two heat exchanger configurations: vertical and horizontal tubes. The correlations expressing the heat transfer coefficients cover the experimental results with a scatter of about +- 0.15% for the three parameters investigated: mass flow rate, heat load, and saturation pressure. As for pressure drops in two-phase flow, an equation expressing the weight of a column of liquid/vapour mixture is satisfactorily compared with the experimental results obtained here. The calculation of this weight is highly important for heat exchanger design, because it helps to predict the recirculation rate in the case of natural circulation. For some cases of evaporators, the calculation of this weight serves to predict the boiling lag in the lower part of the evaporator, which could give rise to low heat transfer coefficient [fr

  9. Flow and pressure drop fluctuations in a vertical tube subject to low frequency oscillations

    International Nuclear Information System (INIS)

    Pendyala, Rajashekhar; Jayanti, Sreenivas; Balakrishnan, A.R.

    2008-01-01

    Heat transfer and other equipment mounted on off-shore platforms may be subjected to low frequency oscillations. The effect of these oscillations, typically in the frequency range of 0.1-1 Hz, on the flow rate and pressure drop in a vertical tube has been studied experimentally in the present work. A 1.75 m-long vertical tube of inner diameter 0.016 m was mounted on a plate and the whole plate was subjected to oscillations in the vertical plane using a mechanical simulator capable of providing low frequency oscillations in the range of 8-30 cycles/min at an amplitude of 0.125 m. The effect of the oscillations on the flow rate and the pressure drop has been measured systematically in the Reynolds number range 500-6500. The induced flow rate fluctuations were found to be dependent on the Reynolds number with stronger fluctuations at lower Reynolds numbers. The effective friction factor, based on the mean pressure drop and the mean flow rate, was also found to be higher than expected. Correlations have been developed to quantify this Reynolds number dependence

  10. Flow and pressure drop fluctuations in a vertical tube subject to low frequency oscillations

    Energy Technology Data Exchange (ETDEWEB)

    Pendyala, Rajashekhar; Jayanti, Sreenivas [Department of Chemical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Balakrishnan, A.R. [Department of Chemical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India)], E-mail: arbala@iitm.ac.in

    2008-01-15

    Heat transfer and other equipment mounted on off-shore platforms may be subjected to low frequency oscillations. The effect of these oscillations, typically in the frequency range of 0.1-1 Hz, on the flow rate and pressure drop in a vertical tube has been studied experimentally in the present work. A 1.75 m-long vertical tube of inner diameter 0.016 m was mounted on a plate and the whole plate was subjected to oscillations in the vertical plane using a mechanical simulator capable of providing low frequency oscillations in the range of 8-30 cycles/min at an amplitude of 0.125 m. The effect of the oscillations on the flow rate and the pressure drop has been measured systematically in the Reynolds number range 500-6500. The induced flow rate fluctuations were found to be dependent on the Reynolds number with stronger fluctuations at lower Reynolds numbers. The effective friction factor, based on the mean pressure drop and the mean flow rate, was also found to be higher than expected. Correlations have been developed to quantify this Reynolds number dependence.

  11. Comparison of endotracheal tube cuff pressure values before and after training seminar.

    Science.gov (United States)

    Özcan, Ayça Tuba Dumanlı; Döğer, Cihan; But, Abdülkadir; Kutlu, Işık; Aksoy, Şemsi Mustafa

    2018-06-01

    It is recommended that endotracheal cuff (ETTc) pressure be between 20 and 30 cm H 2 O. In this present study, we intend to observe average cuff pressure values in our clinic and the change in these values after the training seminar. The cuff pressure values of 200 patients intubated following general anesthesia induction in the operating theatre were measured following intubation. One hundred patients whose values were measured before the training seminar held for all physician assistants, and 100 patients whose values were measured after the training seminar were regarded as Group 1 and Group 2, respectively. Cuff pressures of both groups were recorded, and the difference between them was shown. Moreover, cuff pressure values were explored according to the working period of the physician assistants. There was no significant difference between the groups in terms of age, gender and tube diameters. Statistically significant difference was found between cuff pressure values before and after the training (p values decreased, however no statistically significant different was found (p values and potential complications.

  12. Depressurization experiments on a plugged fibrous insulation in a horizontal pressure tube

    International Nuclear Information System (INIS)

    Lang, H.; Weise, H.J.; Ennen, P.

    1977-08-01

    Hot gas ducts for high-temperature reactors with a helium turbine are subject to additional operational loads not caused by the gas temperature. They include vibrations, caused by high gas velocities or by the sound fields emitted from the turbine, and stresses, originating from fast, short-time pressure changes. Such pressure changes occur as a rule if the generator coupled with the turbine has to be disconnected from the grid. In order to avoid no-load operation of the turbine a bypass between HP and LP side of the turbine is opened. As a consequence of this measure a sudden pressure drop occurs in the free flow cross-section causing differential pressures within the insulation. As the size of these differential pressures depends on the insulating material, the density of plugging, the kind of internals, and on the position and size of the depressurization borings, the pressure distributions in the insulation were measured on a test tube for the HP channel. (orig./RW) [de

  13. Exergoeconomic optimization of coaxial tube evaporators for cooling of high pressure gaseous hydrogen during vehicle fuelling

    International Nuclear Information System (INIS)

    Jensen, Jonas K.; Rothuizen, Erasmus D.; Markussen, Wiebke B.

    2014-01-01

    Highlights: • Three concepts of cooling hydrogen were identified. • A numerical heat transfer model of a coaxial-tube evaporator was built. • The cost of exergy destruction and capital investment cost was evaluated for a range of feasible solution. • The exergoeconomic optimum design for all three concepts was identified. • Cooling with a two-stage evaporator reduces total cost 45% compared to a one-stage evaporator. - Abstract: Gaseous hydrogen as an automotive fuel is reaching the point of commercial introduction. Development of hydrogen fuelling stations considering an acceptable fuelling time by cooling the hydrogen to −40 °C has started. This paper presents a design study of coaxial tube ammonia evaporators for three different concepts of hydrogen cooling, one one-stage and two two-stage processes. An exergoeconomic optimization is imposed to all three concepts to minimize the total cost. A numerical heat transfer model is developed in Engineer Equation Solver, using heat transfer and pressure drop correlations from the open literature. With this model the optimal choice of tube sizes and circuit numbers are found for all three concepts. The results show that cooling with a two-stage evaporator after the pressure reduction valve yields the lowest total cost, 45% lower than the highest, which is with a one-stage evaporator. The main contribution to the total cost was the cost associated with exergy destruction, the capital investment cost contributed with 5–14%. The main contribution to the exergy destruction was found to be thermally driven. The pressure driven exergy destruction accounted for 3–9%

  14. Synthesis of bulk nanocrystalline Pb-Sn-Te alloy under high pressure

    International Nuclear Information System (INIS)

    Zhu, P W; Chen, L X; Jia, X; Ma, H A; Ren, G Z; Guo, W L; Liu, H J; Zou, G T

    2002-01-01

    Pb-Sn-Te bulk nanocrystalline (NC) materials are prepared successfully by quenching melts under high pressure. The mean particle size is about 100 nm and the crystal structure is NaCl type. The mechanism of formation of the bulk NC alloy is explained: there is an increasing of the nucleation rate and a decrease in the growth rate of nuclei with increase of pressure during the solidification processes. The thermoelectric properties of Pb-Sn-Te bulk NC alloy are enhanced. This method is promising for producing thermoelectric materials with improved high-energy conversion efficiency

  15. Heat transfer test in a vertical tube using CO2 at supercritical pressures

    International Nuclear Information System (INIS)

    Kim, Hwan Yeol; Kim, Hyungrae; Song, Jin Ho; Cho, Bong Hyun; Bae, Yoon Yeong

    2007-01-01

    Heat transfer test facility, SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), was constructed at KAERI (Korea Atomic Energy Research Institute) for an investigation of the thermal-hydraulic behaviors of supercritical CO 2 at the various geometries of the test section. The test data will be used for the reactor core design of the SCWR (SuperCritical Water-cooled Reactor). As a working fluid, CO 2 was selected to make use of the low critical pressure and temperature of CO 2 compared with water. An experimental study was carried out in the SPHINX to investigate the characteristics of heat transfer and pressure drop at a vertical single tube with an inside diameter of 4.4 mm in case of an upward flow of supercritical CO 2 . The heat and mass fluxes were varied at a given pressure. The mass flux was in the range of 400-1,200 kg/m 2 s and the heat flux was chosen up to 150 kW/m 2 . The selected pressures were 7.75, 8.12, and 8.85 MPa. A heat transfer deterioration occurred at the lower mass fluxes. The experimental heat transfer coefficients were compared with the ones predicted by several existing correlations. The standard deviation was about 20% for each correlation and an apparent discrepancy was not found among the correlations. The major components of the pressure drop were a gravitational pressure drop and a frictional pressure drop. The frictional pressure drop increases as the mass flux and heat flux increase. (author)

  16. Effect of heavy ion irradiation and α+β phase heat treatment on oxide of Zr-2.5Nb pressure tube material

    Energy Technology Data Exchange (ETDEWEB)

    Choudhuri, Gargi, E-mail: gargi@barc.gov.in [Quality Assurance Division, BARC, Mumbai, 400085 (India); Mukherjee, P.; Gayathri, N. [Variable Energy Cyclotron Centre, Kolkata, 700064 (India); Kain, V.; Kiran Kumar, M.; Srivastava, D. [Material Science Division, BARC, Mumbai, 400085 (India); Basu, S. [Solid State Physics Division, BARC, Mumbai, 400085 (India); Mukherjee, D. [Quality Assurance Division, BARC, Mumbai, 400085 (India); Dey, G.K. [Material Science Division, BARC, Mumbai, 400085 (India)

    2017-06-15

    Effect of heavy-ion irradiation on the crystalline phase transformation of oxide of Zr-2.5Nb alloys has been studied. The steam-autoclaved oxide of pressure tube is irradiated with 306 KeV Ar{sup +9} ions at a dose of 3 × 10{sup 19} Ar{sup +9}/m{sup 2}. The damage profile has been estimated using “Stopping and Range of Ions in Matter” computer program. The variation of the crystal structure along the depth of the irradiated oxide have been characterized non-destructively by Grazing Incidence X-ray Diffraction technique and compared with unirradiated-oxide. The effect of different base metal microstructures on the characteristic of oxide has also been studied. Base metal microstructure as well as the cross-sectional oxide have been characterized using transmission electron microscope. Heavy ion irradiation can significantly alter the distribution of phases in the oxide of the alloy. The difference in chemical state of alloying element has also been found between unirradiated-oxide with that of irradiated-oxide using X-ray photo electron spectroscopy. Chemical state of Nb in steam autoclaved oxide is also altered when the base metal is α + β heat treated.

  17. Exergoeconomic optimization of coaxial tube evaporators for cooling of high pressure gaseous hydrogen during vehicle fuelling

    DEFF Research Database (Denmark)

    Jensen, Jonas Kjær; Rothuizen, Erasmus Damgaard; Markussen, Wiebke Brix

    2014-01-01

    Gaseous hydrogen as an automotive fuel is reaching the point of commercial introduction. Development of hydrogen fuelling stations considering an acceptable fuelling time by cooling the hydrogen to -40 C has started. This paper presents a design study of coaxial tube ammonia evaporators for three......-stage evaporator. The main contribution to the total cost was the cost associated with exergy destruction, the capital investment cost contributed with 5-14 %. The main contribution to the exergy destruction was found to be thermally driven. The pressure driven exergy destruction accounted for 3-9 %....

  18. Exact solution of unsteady flow generated by sinusoidal pressure gradient in a capillary tube

    Directory of Open Access Journals (Sweden)

    M. Abdulhameed

    2015-12-01

    Full Text Available In this paper, the mathematical modeling of unsteady second grade fluid in a capillary tube with sinusoidal pressure gradient is developed with non-homogenous boundary conditions. Exact analytical solutions for the velocity profiles have been obtained in explicit forms. These solutions are written as the sum of the steady and transient solutions for small and large times. For growing times, the starting solution reduces to the well-known periodic solution that coincides with the corresponding solution of a Newtonian fluid. Graphs representing the solutions are discussed.

  19. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1985-01-01

    An expandable antivibration bar for use in stabilizing the U-bend portion of heat transfer tubes in a pressurized water reactor steam generator comprises two adjustable rods connected together by an arcuate connector. The two adjustable rods preferably comprise two mating rod sections having complementary angular sliding surfaces thereon, with means provided to move the rod sections relative to each other along the sliding surfaces so as to expand the rods from a first mated cross-sectional width to a second larger cross-sectional width. The ends of the rod sections have means for aligning the two rod sections and maintaining them in alignment during expansion. (author)

  20. ToF-SIMS characterization of passivation layers of tubes of steam generators Alloy 690 for the nuclear industry: contribution to the understanding of the mechanisms

    International Nuclear Information System (INIS)

    Mazenc, Arnaud

    2013-01-01

    The nickel release in solution is responsible for part of the radioactive contamination of the primary circuit of the Pressurized Water Reactor (PWR) nuclear power plants. For safety issues, it is very important to limit this contamination. The oxide film formed on the steam generators (SG) tubes surface (in Alloy 690) plays an important role as diffusion barrier in the limitation of this phenomenon. It appears paramount to well control the phenomena of oxide film formation, which themselves are strongly impacted by the materials parameters of the tube (grain size, work hardening, etc.), as well as by the chemical conditions of the primary water (temperature, pH, etc.). The objective of this work is both to finely characterize oxide layer formed on SG tubes and to establish a relationship between corrosion and release. This study was carried out in three parts. The first part consists in the development of an extensive use of the Time of Flight Secondary Ion Mass Spectrometry (ToFSIMS) to characterize the oxide films. This methodology is based on elaboration of reference oxide films by oxidation of metallic samples. The second part of this work is dedicated to the investigation of the influence of three material parameters on the composition and structure of the oxide layer: crystallographic orientation, grain size, and content of chromium intergranular carbide precipitates. Considerable effects are observed for the crystallographic orientation as well as content of precipitates, while the grain size has no influence. Lastly, the last part presents simultaneously the influence of the physicochemical conditions of the primary water on the kinetics of release and the nature of the oxide layer. The temperature and the pH tend to be key parameters influencing corrosion and the release. (author)

  1. Thin-walled large-diameter zirconium alloy tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Price, E.G.; Richinson, P.J.

    1978-08-01

    The requirements of the thin-walled large-diameter Zircaloy-2 tubing used in CANDU reactors are reviewed. Strength, residual stress patterns, texture and prior deformation contribute to the stability of these tubes. The extent to which the present manufacturing route meets these requirements is discussed. (author)

  2. Measuring systolic arterial blood pressure. Possible errors from extension tubes or disposable transducer domes.

    Science.gov (United States)

    Rothe, C F; Kim, K C

    1980-11-01

    The purpose of this study was to evaluate the magnitude of possible error in the measurement of systolic blood pressure if disposable, built-in diaphragm, transducer domes or long extension tubes between the patient and pressure transducer are used. Sinusoidal or arterial pressure patterns were generated with specially designed equipment. With a long extension tube or trapped air bubbles, the resonant frequency of the catheter system was reduced so that the arterial pulse was amplified as it acted on the transducer and, thus, gave an erroneously high systolic pressure measurement. The authors found this error to be as much as 20 mm Hg. Trapped air bubbles, not stopcocks or connections, per se, lead to poor fidelity. The utility of a continuous catheter flush system (Sorenson, Intraflow) to estimate the resonant frequency and degree of damping of a catheter-transducer system is described, as are possibly erroneous conclusions. Given a rough estimate of the resonant frequency of a catheter-transducer system and the magnitude of overshoot in response to a pulse, the authors present a table to predict the magnitude of probable error. These studies confirm the variability and unreliability of static calibration that may occur using some safety diaphragm domes and show that the system frequency response is decreased if air bubbles are trapped between the diaphragms. The authors conclude that regular procedures should be established to evaluate the accuracy of the pressure measuring systems in use, the transducer should be placed as close to the patient as possible, the air bubbles should be assiduously eliminated from the system.

  3. Endotracheal tube cuff pressures during general anaesthesia while using air versus a 50% mixture of nitrous oxide and oxygen as inflating agents

    Directory of Open Access Journals (Sweden)

    Jesni Joseph Manissery

    2007-01-01

    Full Text Available The present study was aimed at assessing the efficacy of filling a 50% mixture of nitrous oxide : oxygen (50%N 2 O:O 2 in the endotracheal tube cuff to provide stable cuff pressures during general anaesthesia with 67%N 2 O. The endotracheal tube cuff pressures with air (control as the inflating agent in the tubes were found to have a total mean pressure of 62.60±12.33 at the end of one hour of general anaesthesia. When comparing the endotracheal tube cuff pressures in the Mallinckrodt tubes with that of the Portex tubes, with air as the inflating agent, the Portex tubes showed a significantly lower cuff pressures at the end of one hour. The endotracheal tube cuff pressures with 50%N 2 O:O 2 as the inflating agent showed a total mean pressure of 27.63 ± 3.221 at the end of one hour of general anaesthesia. This indicates that inflation of the cuff of the endotracheal tubes with a 50%N 2 O:O 2 rather than air maintains a stable intra cuff pressure. Therefore, the method of using a 50%N 2 O:O 2 for filling endotracheal tube cuff can be adopted for endotracheal tubes with high-volume, low-pressure cuffs to prevent both excessive cuff pressure and disruption of cuff seal, during general anaesthesia lasting up to one hour.

  4. Failure assessment and evaluation of critical crack length for a fresh Zr-2 pressure tube of an Indian PHWR

    International Nuclear Information System (INIS)

    Krishnan, Suresh; Bhasin, Vivek; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1996-01-01

    Fracture analysis of Zr-2 pressure tubes having through wall axial crack was done using finite element method. The analysis was done for tubes in as received condition. During reactor operation the mechanical properties of Zr-2 undergo changes. The analysis is valid for pressure tubes of newly commissioned reactors. The main aim of the study was to determine critical crack length of pressure tubes in normal operating conditions. Elastic plastic fracture analysis was done for different crack lengths to determine applied J-integral values. Tearing modulus instability concept was used to evaluate critical crack length. One of the important parameter studied was, the effect of crack face pressure, which leaking fluid exert on the crack faces/lips of through wall axial crack. Its effect was found to be significant for pressure tubes. It increases the applied J-integral values. Approximate analytical solutions which takes into account the plasticity ahead of crack tip, are available and widely used. These formulae do not take into account the crack face pressure. Since, for the present situation the effect of crack face pressure is significant hence, detailed finite analysis was necessary. Detailed 3D finite element analysis gives an insight into the variation of J-integral values over the thickness of pressure tube. It was found that J values are maximum at the middle layer of the tube. A peak factor on J values was defined and evaluated as ratio of maximum J to average J across the thickness, crack opening area for each length was also evaluated. The knowledge of crack opening area is useful for leak before break studies. The failure assessment was also done using Central Electricity Generating Board (CEGB) R-6 method considering the ductile tearing. The reserve factors (or safety margins) for different crack lengths was evaluated using R-6 method. (author). 30 refs., 21 figs., 34 tabs

  5. Analytical tools and methodologies for evaluation of residual life of contacting pressure tubes in the early generation of Indian PHWRs

    International Nuclear Information System (INIS)

    Sinha, S.K.; Madhusoodanan, K.; Rupani, B.B.; Sinha, R.K.

    2002-01-01

    In-service life of a contacting Zircaloy-2 pressure tube (PT) in the earlier generation of Indian PHWRs, is limited mainly due to the accelerated hydrogen pick-up and nucleation and growth of hydride blister(s) at the cold spot(s) formed on outside surface of pressure tube as a result of its contact with the calandria tube (CT). The activities involving development of the analytical models for simulating the degradation mechanisms leading to PT-CT contact and the methodologies for the revaluation of their safe life under such condition form the important part of our extensive programme for the life management of contacting pressure tubes. Since after the PT-CT contact, rate of hydrogen pick-up and nucleation and growth of hydride blisters govern the safe residual life of the pressure tube, two analytical models (a) hydrogen pick-up model ('HYCON') and (b) model for the nucleation and growth of hydride blister at the contact spot ('BLIST -2D') have been developed in-house to estimate the extent of degradation caused by them. Along with them, a methodology for evaluation of safe residual life has also been formulated for evaluating the safe residual life of the contacting channels. This paper gives the brief description of the models and the methodologies relevant for the contacting Zircaloy-2 pressure tubes. (author)

  6. A novel Fe–Cr–Nb matrix composite containing the TiB_2 neutron absorber synthesized by mechanical alloying and final hot isostatic pressing (HIP) in the Ti-tubing

    International Nuclear Information System (INIS)

    Litwa, Przemysław; Perkowski, Krzysztof; Zasada, Dariusz; Kobus, Izabela; Konopka, Gustaw; Czujko, Tomasz; Varin, Robert A.

    2016-01-01

    The Fe–Cr–Ti-Nb elemental powders were mechanically alloyed/ball milled with TiB_2 and a small quantity of Y_2O_3 ceramic to synthesize a novel Fe-based alloy-ceramic powder composite that could be processed by hot isostatic pressing (HIP) for a perceived potential application as a neutron absorber in nuclear reactors. After ball milling for the 30–80 h duration relatively uniform powders with micrometric sizes were produced. With increasing milling time a fraction of TiB_2 particles became covered with the much softer Fe-based alloy which resulted in the formation of a characteristic “core-mantel” structure. For the final HIP-ing process the mechanically alloyed powders were initially uniaxially pressed into rod-shaped compacts and then cold isostatically pressed (CIP-ed). Subsequently, the rod-shaped compacts were placed in the Ti-tubing and subjected to hot isostatic pressing (HIP) at 1150 °C/200 MPa pressure. The HIP-ing process resulted in the formation of the near-Ti and intermediate diffusional layers in the microstructure of HIP-ed samples which formed in accord with the Fe-Ti binary phase diagram. Those layers contain the phases such as α-Ti (HCP), the FeTi intermetallic and their hypo-eutectoid mixtures. In addition, needle-like particles were formed in both layers in accord with the Ti-B binary phase diagram. Nanohardness testing, using a Berkovich type diamond tip, shows that the nanohardness in the intermediate layer areas, corresponding to the composition of the hypo-eutectoid mixture of Ti-FeTi, equals 980.0 (±27.1) HV and correspondingly 1176.9 (±47.6) HV for the FeTi phase. The nanohardness in the sample's center in the areas with the fine mixture of Fe-based alloy and small TiB_2 particles equals 1048.3 (±201.8) HV. The average microhardness of samples HIP-ed from powders milled for 30 and 80 h is 588 HV and 733 HV, respectively. - Highlights: • A Fe–Cr–Nb-based composite with TiB_2 neutron absorbing ceramic was mechanically

  7. Denting of Inconel 600 steam generator tubes in pressurized water reactors

    International Nuclear Information System (INIS)

    Van Rooyen, D.; Weeks, J.R.

    1976-10-01

    Rapid, localized corrosion of carbon steel tube support plates (TSP) has led to cases of denting of steam generator tubes, due to the pressure of corrosion products formed in crevices between the tubes and TSP holes. The corrosion product is mainly magnetite (Fe 3 O 4 ), formed in ''run-away'' fashion as a result of local chemistry changes when an extended operation with phosphate (PO 4 ) treatment of the secondary coolant is followed by an all volatile treatment (AVT). The rate of the ''run-away'' magnetite formation, and therefore, the extent of damage will probably vary with the amounts of the harmful chemicals present and with temperature. Leaky condensers are felt to be responsible for the presence of Cl - ions, and for the observation that denting is more extensive in plants with salt water cooled condensers. It is possible that thermal cycles assist the denting process, both by mechanical and chemical ratchetting mechanisms. Recommendations are presented concerning the continued operation of plants with observed denting

  8. Use of pressurized eccentric tubes to study the effect of hydrostatic stress on swelling

    International Nuclear Information System (INIS)

    Wolfer, W.G.; Reiley, T.C.

    1977-05-01

    A technique for measuring the effect of hydrostatic stress on radiation-induced swelling is presented. This technique is based on the nonuniform hydrostatic stress that arises when an eccentric tube (a tube with inner and outer surfaces having dissimilar centers of revolution) is internally pressurized. The elastic analyses of the thin- and thick-walled eccentric tube are given. The elastic stress state is allowed to relax plastically, based on a constitutive law for deformation during neutron irradiation. In this case, the constitutive law contains a linearly stress-dependent deviatoric strain rate and a dilatation rate that is linearly dependent on hydrostatic stress. Emphasis is placed on the specimen design and experimental procedure for in-reactor experiments in which the coefficient relating hydrostatic stress and swelling is sought. It is shown that, for the 316L stainless steel specimens placed in EBR-II, we may expect that any appreciable effect of hydrostatic stress on swelling will be observable through changes in specimen curvature

  9. Phase composition and microhardness of rapidly quenched Al-Fe alloys after high pressure torsion deformation

    Energy Technology Data Exchange (ETDEWEB)

    Tcherdyntsev, V.V.; Kaloshkin, S.D.; Gunderov, D.V.; Afonina, E.A.; Brodova, I.G.; Stolyarov, V.V.; Baldokhin, Yu.V.; Shelekhov, E.V.; Tomilin, I.A

    2004-07-15

    Aluminium-based Al-Fe alloys with Fe content of 2, 8, and 10 wt.% were prepared by rapid quenching (RQ) from the melt at a rate of 10{sup 6} K/s. Structure of the alloys was examined by X-ray diffraction (XRD) and Moessbauer spectroscopy. Phase transformations of RQ alloys by high pressure torsion (HPT) were studied. Dependences of phase composition on the intensity of HPT were investigated. Microhardness measurements of HPT alloys show a considerable structural heterogeneity of specimens, the dependence of microhardness on the radius of the pills was found out. Phase composition and microhardness during the heating were investigated. At the initial step of heating (120-150 deg. C), an increase in microhardness was observed, whereas further heating leads to a decrease in the microhardness.

  10. 77 FR 21734 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line, and Pressure Pipe From Romania...

    Science.gov (United States)

    2012-04-11

    ... DEPARTMENT OF COMMERCE International Trade Administration [A-485-805] Certain Small Diameter Carbon and Alloy Seamless Standard, Line, and Pressure Pipe From Romania: Extension of Time Limit for... diameter carbon and alloy seamless standard, line and pressure pipe from Romania for the period August 1...

  11. Hydraulic performance evaluation of pressure compensating (pc) emitters and micro-tubing for drip irrigation system

    International Nuclear Information System (INIS)

    Mangrio, A.G.; Asif, M.; Jahangir, I.

    2013-01-01

    Drip irrigation system is necessary for those areas, where the water scarcity issues are present. The present study was conducted at the field station of Climate Change, Alternate Energy and Water Resources Institute (CAEWRI), National Agricultural Research Center (NARC), Islamabad, during 2013, regarding drip irrigation system. Drip irrigation system depends on uniform emitter application flow. All the emitters were tested and replicated thrice at pressure head (34 to 207Kpa) with an increment of 34 Kpa. The minimum and maximum discharges were 1.32 - 3.52, 3.36 - 5.42, and 43.22 - 100.99 Lph, with an average of 2.42, 4.63 and 73.66 Lph, for Bow Smith, RIS and Micro-tubing, respectively. It indicates that more than 90% of emission uniformity (EU) and uniformity coefficient (CU) for all Emitters, which shows excellent water application with least standard deviation, ranging 0.12 to 2.37, throughout the operating pressure heads in all emitters. An average coefficient of variation (CV) of all emitters were behaving less than 0.07, indicating an excellent class at all operating pressure heads between 34 to 207 Kpa. Moreover, the relationship of discharge and pressure of emitters indicates that discharge increased with the increase of pressure head. The Q-H curve plays key role in the selection of emitters. (author)

  12. Lattice dynamics, elasticity and magnetic abnormality in ordered crystalline alloys Fe3Pt at high pressures

    Science.gov (United States)

    Cheng, Tai-min; Yu, Guo-Liang; Su, Yong; Ge, Chong-Yuan; Zhang, Xin-Xin; Zhu, Lin; Li, Lin

    2018-05-01

    The ordered crystalline Invar alloy Fe3Pt is in a special magnetic critical state, under which the lattice dynamic stability of the system is extremely sensitive to external pressures. We studied the pressure dependence of enthalpy and magnetism of Fe3Pt in different crystalline alloys by using the first-principles projector augmented-wave method based on the density functional theory. Results show that the P4/mbm structure is the ground state structure and is more stable relative to other structures at pressures below 18.54 GPa. The total magnetic moments of L12, I4/mmm and DO22 structures decrease rapidly with pressure and oscillate near the ferromagnetic collapse critical pressure. At the pressure of 43 GPa, the ferrimagnetic property in DO22 structure becomes apparently strengthened and its volume increases rapidly. The lattice dynamics calculation for L12 structures at high pressures shows that the spontaneous magnetization of the system in ferromagnetic states induces the softening of the transverse acoustic phonon TA1 (M), and there exists a strong spontaneous volume magnetostriction at pressures below 26.95 GPa. Especially, the lattice dynamics stability is sensitive to pressure, in the pressure range between the ferromagnetic collapse critical pressure (41.9 GPa) and the magnetism completely disappearing pressure (57.25 GPa), and near the pressure of phase transition from L12 to P4/mbm structure (27.27 GPa). Moreover, the instability of magnetic structure leads to a prominent elastic modulus oscillation, and the spin polarizability of electrons near the Fermi level is very sensitive to pressures in that the pressure range. The pressure induces the stability of the phonon spectra of the system at pressures above 57.25 GPa.

  13. Oxide Dispersion Strengthened Fe3Al-Based Alloy Tubes: Application Specific Development for the Power Generation Industry

    Energy Technology Data Exchange (ETDEWEB)

    Kad, B.K.

    2002-02-08

    A detailed and comprehensive research and development methodology is being prescribed to produce Oxide Dispersion Strengthened (ODS)-Fe{sub 3}Al thin walled tubes, using powder extrusion methodologies, for eventual use at operating temperatures of up to 1100% in the power generation industry. A particular ''in service application'' anomaly of Fe{sub 3}Al-based alloys is that the environmental resistance is maintained up to 1200 C, well beyond where such alloys retain sufficient mechanical strength. Grain boundary creep processes at such high temperatures are anticipated to be the dominant failure mechanism. Thus, the challenges of this program are manifold: (1) to produce thin walled ODS-Fe{sub 3}Al tubes, employing powder extrusion methodologies, with (2) adequate increased strength for service at operating temperatures, and (3) to mitigate creep failures by enhancing the as-processed grain size in ODS-Fe{sub 3}Al tubes. Our research progress till date has resulted in the successful batch production of typically 8 Ft. lengths of 1-3/8 inch diameter, 1/8 inch wall thickness, ODS-Fe{sub 3}Al tubes via a proprietary single step extrusion consolidation process. The process parameters for such consolidation methodologies have been prescribed and evaluated as being routinely reproducible. Such processing parameters (i.e., extrusion ratios, temperature, can design etc.) were particularly guided by the need to effect post-extrusion recrystallization and grain growth at a sufficiently low temperature, while still meeting the creep requirement at service temperatures. Static recrystallization studies show that elongated grains (with their long axis parallel to the extrusion axis), typically 200-2000 {micro}m in diameter, and several millimeters long can be obtained routinely, at 1200 C. The growth kinetics are affected by the interstitial impurity content in the powder batches. For example complete recrystallization, across the tube wall thickness, is

  14. Life management of Zr 2.5% Nb pressure tube through estimation of fracture properties by cyclic ball indentation technique

    International Nuclear Information System (INIS)

    Chatterjee, S.; Madhusoodanan, K.; Rama Rao, A.

    2015-01-01

    In Pressurised Heavy Water Reactors (PHWRs) fuel bundles are located inside horizontal pressure tubes. Pressure tubes made up of Zr 2.5 wt% Nb undergo degradation during in-service environmental conditions. Measurement of mechanical properties of degraded pressure tubes is important for assessing its fitness for further service in the reactor. The only way to accomplish this important objective is to develop a system based on insitu measurement technique. Considering the importance of such measurement, an In-situ Property Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed indigenously. The remotely operable system is capable of carrying out indentation trial on the inside surface of the pressure tube and to estimate important mechanical properties like yield strength, ultimate tensile strength, hardness etc. It is known that fracture toughness is one of the important life limiting parameters of the pressure tube. Hence, five spool pieces of Zr 2.5 wt% Nb pressure tube of different mechanical properties have been used for estimation of fracture toughness by ball indentation method. Curved Compact Tension (CCT) specimens were also prepared from the five spool pieces for measurement of fracture toughness from conventional tests. The conventional fracture toughness values were used as reference data. A methodology has been developed to estimate the fracture properties of Zr 2.5 wt% Nb pressure tube material from the analysis of the ball indentation test data. This paper highlights the comparison between tensile properties measured from conventional tests and IProMS trials and relates the fracture toughness parameters measured from conventional tests with the IProMS estimated fracture properties like Indentation Energy to Fracture. (author)

  15. Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Jun-Lin; Zhou, Ke-Yi, E-mail: boiler@seu.edu.cn; Xu, Jian-Qun [Key Laboratory of Energy Thermal Conversion and Control of Ministry of Education, School of Energy and Environment, Southeast University, Nanjing 210096, Jiangsu Province (China); Wang, Xin-Meng; Tu, Yi-You [School of Materials Science and Engineering, Southeast University, Nanjing 210096, Jiangsu Province (China)

    2014-07-28

    Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

  16. Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

    Science.gov (United States)

    Huang, Jun-Lin; Zhou, Ke-Yi; Wang, Xin-Meng; Tu, Yi-You; Xu, Jian-Qun

    2014-07-01

    Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

  17. Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

    International Nuclear Information System (INIS)

    Huang, Jun-Lin; Zhou, Ke-Yi; Xu, Jian-Qun; Wang, Xin-Meng; Tu, Yi-You

    2014-01-01

    Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

  18. A Comparison of the Predicted Tube Plugging Rate for Alloy 600HTMA Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Boo, Myung Hwan; Kang, Yong Seok [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2010-10-15

    To manage components that are used in long term operations such as steam generation, it is important to know the tube plugging rate, which can cause the performance degradation. The life of components can be predicted by the method using determinism and probability theory. With a method using probability theory, damage prediction of tube is possible. In this study, damage prediction for steam generation (SG) tube is performed using Weibull distribution and predicted plugging rate (life) is compared with the simple sum plugging number and case by case (failure cause) plugging number

  19. Bruce and Darlington power pulse and pressure tube integrity programs -status 1995

    Energy Technology Data Exchange (ETDEWEB)

    Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Wylie, J [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

    1996-12-31

    The optimum solution to pressure tube fretting at the inlet of the Bruce and Darlington channels, a concern which became very serious following inspections in early 1992, is to remove the inlet bundle and operate with a 12 fuel bundle channel. During analysis of this operating mode a `power pulse` was identified which could occur during an inlet header break where all the fuel in the channel moved rapidly to the inlet of the channel. The pulse was unacceptable and the units were derated until solutions could be implemented. A number of solutions were identified and each station has begun implementation of their specific solution. Implementation has not been without problems and this paper provides a status report on the progress to date of the long bundle implementation solution for Bruce B and Darlington and the fuelling with the flow solution being implemented at Bruce A. Both types of solution have a significant impact on the original concern, fretting of the pressure tube. (author). 1 ref., 6 figs.

  20. A risk-informed approach to the assessment of DHC initiation in pressure tubes

    International Nuclear Informa