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Sample records for generator tubing materials

  1. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

    Corrosion resistance is the major consideration in selecting tubing material for CANDU steam generators. Corrosion, and additional considerations, lead to the following steam generator tubing material recommendations: for CANDU-BPHWR's (boiling pressurized heavy water reactors) low-cobalt Incoloy-800; for CANDU-PHWR's (pressurized, non-boiling, heavy water reactors), low-cobalt Monel-400

  2. Minimize corrosion degradation of steam generator tube materials

    International Nuclear Information System (INIS)

    Lu, Y.

    2006-01-01

    As part of a coordinated program, AECL is developing a set of tools to aid with the prediction and management of steam generator performance. Although stress corrosion cracking (of Alloy 800) has not been detected in any operating steam generator, for life management it is necessary to develop mechanistic models to predict the conditions under which stress corrosion cracking is plausible. Experimental data suggest that all steam generator tube materials are susceptible to corrosion degradation under some specific off-specification conditions. The tolerance to the chemistry upset for each steam generator tube alloy is different. Electrochemical corrosion behaviors of major steam generator tube alloys were studied under the plausible aggressive crevice chemistry conditions. The potential hazardous conditions leading to steam generator tube degradation and the conditions, which can minimize steam generator tube degradation have been determined. Recommended electrochemical corrosion potential/pH zones were defined for all major steam generator tube materials, including Alloys 600, 800, 690 and 400, under CANDU steam generator operating and startup conditions. Stress corrosion cracking tests and accelerated corrosion tests were carried out to verify and revise the recommended electrochemical corrosion potential/pH zones. Based on this information, utilities can prevent steam generator material degradation surprises by appropriate steam generator water chemistry management and increase the reliability of nuclear power generating stations. (author)

  3. Heat exchanger tubing materials for CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Taylor, G.F.

    1977-07-01

    The performance of steam generator tubing (nickel-chromium-iron alloy in NPD and nickel-copper alloy in Douglas Point and Pickering generating stations) has been outstanding and no corrosion-induced failures have occurred. The primary coolant will be allowed to boil in the 600 MW (electrical) CANDU-PHW reactors. An iron-nickel-chromium alloy has been selected for the steam generator tubing because it will result in lower radiation fields than the alloys used before. It is also more resistant than nickel-chromium-iron alloy to stress corrosion cracking in the high purity water of the primary circuit, an unlikely but conceivable hazard associated with higher operating temperatures. Austenitic alloy and ferritic-austenitic stainless steel tubing have been selected for the moderator coolers in CANDU reactors being designed and under construction. These materials will reduce the radiation fields around the moderator circuit while retaining the good resistance to corrosion in service water that has characterized the copper-nickel alloys now in use. Brass and bronze tubes in feedwater heaters and condensers have given satisfactory service but do, however, complicate corrosion control in the steam cycle and, to reduce the transport of corrosion products from the feedtrain to the steam generator, stainless steel is preferred for feedwater heaters and stainlss steel or titanium for condensers. (author)

  4. The SCC testing of nuclear steam generator tubing materials

    Science.gov (United States)

    Doherty, P. E.; Sarver, J. M.; Miglin, B. P.

    1996-05-01

    The integrity of heat-exchanger tubes in a nuclear reaction system is crucial for the safe operation of a power plant. In order to study the corrosion behavior of certain alloys, constant extension rate (CERT) tests were performed on alloy 690 and alloy 800 nuclear steam generator tubing specimens. In this article, the CERT test results (such as maximum stress achieved and crack morphology) are correlated to tubing microstructure, chemistry, and manufacturing processes.

  5. Evaluation on mechanical and corrosion properties of steam generator tubing materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Sup; Lee, Byong Whi; Lee, Sang Kyu; Lee, Young Ho; Kim, Jun Whan; Lee, Ju Seok; Kwon, Hyuk Sang; Kim, Su Jung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-06-15

    Steam generator is one of the major components of nuclear reactor pressure boundary. It's main function os transferring heat which generated in the reactor to turbine generator through steam generator tube. In these days, steam generator tubing materials of operating plant are used Inconel 600 alloys. But according to the operation time, there are many degradation phenomena which included mechanical damage due to flow induced vibration and corrosion damage due to PWSCC, IGA/SCC and pitting etc. Recently Inconel 690 alloys are selected as new and replacement steam generator tubes for domestic nuclear power plant. But there are few study about mechanical and corrosion properties of Inconel 600 and 690. The objectives of this study is to evaluate and compare mechanical and corrosion propertied of steam generator tube materials.

  6. 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

  7. Study on thermal and mechanical properties of U-tube materials for steam generator

    International Nuclear Information System (INIS)

    Rheu, Woo Suk; Kang, Young Hwan; Park, Jong Man; Joo, Ki Nam; Kim, Sung Soo; Maeng, Wan Young; Park, Se Jin

    1993-01-01

    Most of domestic nuclear plants have used I600 TT material for steam generator U-tube, and piled up the field experience. I600 HTMA and I690 TT, however, are recommended for an alternative of U-tube by ABB-CE since YK-3 and 4. Field experience of I600 HTMA and I690 TT have not compiled in the country, so it is concerned to select the future materials for U-tube. Thus, database on the thermal and mechanical properties of U-tube materials is very necessary for design documentations. In this study, the thermal, mechanical and metallugical properties were tested and evaluated to establish the database for steam generator U-tube. In addition, thermal conductivity of I600 and I690 was measured and compared statistically, providing a basic document for applying I690 to U-tube. The results will be used to improve the manufacturing process in order to increase the integrity of U-tube. (Author)

  8. Stress corrosion cracking of the tubing materials for nuclear steam generators in an environment containing lead

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Kim, Uh Chul; Lee, Eun Hee; Hwang, Seong Sik

    2004-01-01

    Steam generator tube materials show a high susceptibility to stress corrosion cracking (SCC) in an environment containing lead species and some nuclear power plants currently have degradation problems associated with lead-induced stress corrosion cracking in a caustic solution. Effects of an applied potential on SCC is tested for middle-annealed Alloy 600 specimens since their corrosion potential can be changed when lead oxide coexists with other oxidizing species like copper oxide in the sludge. In addition, all the steam generator tubing materials used for nuclear power plants being operated and currently under construction in Korea are tested in a caustic solution with lead oxide. (author)

  9. Wear studies of materials for tubes and anti-vibration bars in nuclear steam generators

    International Nuclear Information System (INIS)

    Ko, P.L.; Taponat, M.C.

    1995-01-01

    Wear occurs as a result of relative motion at the interface of two contacting bodies. In nuclear power steam generators, high flow rates can induce vibration of the tubes resulting in wear damage due to impact and sliding contacts between the tubes and their supports. A research project aiming to gain better understanding of the mechanisms and mechanics involved in vibratory wear and to develop a more versatile predictive wear model was carried out. Combinations of Inconel tubes against flat anti-vibration bars of 403 s.s. and electrolytic chrome plated Inconel 600 were tested under conditions of reciprocating sliding and impacting in water at room temperature and at 250 C. The results show that depending on the material combinations and the loading conditions distinctively different wear mechanisms and often drastically different wear rates can occur

  10. Probabilistic analysis of degradation incubation time of steam generator tubing materials

    International Nuclear Information System (INIS)

    Pandey, M.D.; Jyrkama, M.I.; Lu, Y.; Chi, L.

    2012-01-01

    The prediction of degradation free lifetime of steam generator (SG) tubing material is an important step in the life cycle management and decision for replacement of steam generators during the refurbishment of a nuclear station. Therefore, an extensive experimental research program has been undertaken by the Canadian Nuclear Industry to investigate the degradation of widely-used SG tubing alloys, namely, Alloy 600 TT, Alloy 690 TT, and Alloy 800. The corrosion related degradations of passive metals, such as pitting, crevice corrosion and stress corrosion cracking (SCC) etc. are assumed to start with the break down of the passive film at the tube-environment interface, which is characterized by the incubation time for passivity breakdown and then the degradation growth rate, and both are influenced by the chemical environment and coolant temperature. Since the incubation time and growth rate exhibit significant variability in the laboratory tests used to simulate these degradation processes, the use of probabilistic modeling is warranted. A pit is initiated with the breakdown of the passive film on the SG tubing surface. Upon exposure to aggressive environments, pitting corrosion may not initiate immediately, or may initiate and then re-passivate. The time required to initiate pitting corrosion is called the pitting incubation time, and that can be used to characterize the corrosion resistance of a material under specific test conditions. Pitting may be the precursor to other corrosion degradation mechanisms, such as environmentally-assisted cracking. This paper will provide an overview of the results of the first stage of experimental program in which samples of Alloy 600 TT, Alloy 690 TT, and Alloy 800 were tested under various temperatures and potentials and simulated crevice environments. The testing environment was chosen to represent layup, startup, and full operating conditions of the steam generators. Degradation incubation times for over 80 samples were

  11. Effect of heat treatment and composition on stress corrosion cracking of steam generation tubing materials

    International Nuclear Information System (INIS)

    Kim, H. P.; Hwang, S. S.; Kuk, I. H.; Kim, J. S.; Oh, C. Y.

    1998-01-01

    Effects of heat treatment and alloy composition on stress corrosion cracking (SCC) of steam generator tubing materials have been studied in 40% NaOH at 315.deg.C at potential of +200mV above corrosion potential using C-ring specimen and reverse U bend specimen. The tubing materials used were commercial Alloy 600, Alloy 690 and laboratory alloys, Ni-χCr-10Fe. Commercial Alloy 600, Alloy 690 were mill annealed or thermally treated.Laboratory alloy Ni-χCr-10Fe, and some of Alloy 600 and Alloy 690 were solution annealed. Polarization curves were measured to find out any relationship between SCC susceptibility and electrochemical behaviour. The variation in thermal treatment of Alloy 600 and Alloy 690 had no effect on polarization behaviour probably due to small area fraction of carbide and Cr depletion zone near grain boundary. In anodic polarization curves, the first and second anodic peaks at about 170mV and about at 260mV, respectively, above corrosion potential were independent of Cr content, whereas the third peak at 750mV above corrosion potential and passive current density in-creased with Cr content. SCC susceptibility decreased with Cr content and thermal treatment producing semicontinuous grain boundary decoration. Examination of cross sectional area of C-ring specimen showed deep SCC cracks for the alloys with less than 17%Cr and many shallow attacks for alloy 690. The role of Cr content in steam generator tubing materials and grain boundary carbide on SCC were discussed

  12. Steam generator tube failures

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

    1996-04-01

    A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service.

  13. Steam generator tube failures

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

    1996-04-01

    A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service

  14. 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

  15. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1984-10-01

    A review of the performance of steam generator tubes in 116 water-cooled nuclear power reactors showed that tubes were plugged at 54 (46 percent) of the reactors. The number of tubes removed from service decreased from 4 692 (0.30 percent) in 1981 to 3 222 (0.20 percent) in 1982. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that have used only volatile treatment, with or without condensate demineralization

  16. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1983-08-01

    A review of the performance of steam generator tubes in 110 water-cooled nuclear power reactors showed that tubes were plugged at 46 (42 percent) of the reactors. The number of tubes removed from service increased from 1900 (0.14 percent) in 1980 to 4692 (0.30 percent) in 1981. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that used all-volatile treatment since start-up. At one reactor a large number of degraded tubes were repaired by sleeving which is expected to become an important method of tube repair in the future

  17. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1982-04-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1980. Tube defects occurred at 38% of the 97 reactors surveyed. This is a marginal improvement over 1979 when defects occurred at 41% of the reactors. The number of failed tubes was also lower, 0.14% of the tubes in service in 1980 compared with 0.20% of those in service in 1979. Analysis of the causes of these failures indicates that stress corrosion cracking was the leading failure mechanism. Reactors that used all-volatile treatment of secondary water, with or without full-flow condensate demineralization since start-up showed the lowest incidence of corrosion-related defects

  18. Steam generator tube vibration study

    International Nuclear Information System (INIS)

    Enderlin, W.I.

    1986-01-01

    Chemical cleaning has been proposed to remove magnetite buildup in some pressurized water reactor steam generators. The US Nuclear Regulatory Commission (NRC) has expressed concern that such cleaning would combine with the tube denting caused by magnetite formation to enlarge tube/tube support plate clearances, increasing the level of flow-induced vibrations that could lead to unacceptably high tube wear and failure rates. In support of NRC, the Pacific Northwest Laboratory investigated whether such increased clearances would exacerbate tube fretting wear. Using a full-length scale model of a steam generator tube bundle, flow tests were conducted at an instrumented location through clearances representing as-built and post-cleaned tube conditions. Test results indicated little potential for increased tube wear as a result of chemical cleaning, under normal operating conditions at tube support locations similar to that tested

  19. Corrosion aspects of Ni-Cr-Fe based and Ni-Cu based steam generator tube materials

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, R.S., E-mail: rsdutta@barc.gov.i [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2009-09-01

    This paper reviews corrosion related issues of Ni-Cr-Fe based (in a general sense) and Ni-Cu based steam generator tube materials for nuclear power plants those have been dealt with for last more than four decades along with some updated information on corrosion research. The materials include austenitic stainless steels (SSs), Alloy 600, Monel 400, Alloy 800 and Alloy 690. Compatibility related issues of these alloys are briefly discussed along with the alloy chemistry and microstructure. For austenitic SSs, stress corrosion cracking (SCC) behaviour in high temperature aqueous environments is discussed. For Alloy 600, intergranular cracking in high temperature water including hydrogen-induced intergranular cracking is highlighted along with the interactions of material in various environments. In case of Monel 400, intergranular corrosion and pitting corrosion at ambient temperature and SCC behaviour at elevated temperature are briefly described. For Alloy 800, the discussion covers SCC behaviour, surface characterization and microstructural aspects of pitting, whereas hydrogen-related issues are also highlighted for Alloy 690.

  20. Corrosion aspects of Ni-Cr-Fe based and Ni-Cu based steam generator tube materials

    International Nuclear Information System (INIS)

    Dutta, R.S.

    2009-01-01

    This paper reviews corrosion related issues of Ni-Cr-Fe based (in a general sense) and Ni-Cu based steam generator tube materials for nuclear power plants those have been dealt with for last more than four decades along with some updated information on corrosion research. The materials include austenitic stainless steels (SSs), Alloy 600, Monel 400, Alloy 800 and Alloy 690. Compatibility related issues of these alloys are briefly discussed along with the alloy chemistry and microstructure. For austenitic SSs, stress corrosion cracking (SCC) behaviour in high temperature aqueous environments is discussed. For Alloy 600, intergranular cracking in high temperature water including hydrogen-induced intergranular cracking is highlighted along with the interactions of material in various environments. In case of Monel 400, intergranular corrosion and pitting corrosion at ambient temperature and SCC behaviour at elevated temperature are briefly described. For Alloy 800, the discussion covers SCC behaviour, surface characterization and microstructural aspects of pitting, whereas hydrogen-related issues are also highlighted for Alloy 690.

  1. Application of nano-sized TiO2 as an inhibitor of stress corrosion cracking in the steam generator tube materials.

    Science.gov (United States)

    Kim, Kyung Mo; Lee, Eun Hee; Kim, Uh Chul; Choi, Byung Seon

    2010-01-01

    Several chemicals were studied to suppress the damage due to a stress corrosion cracking (SCC) of the steam generator (SG) tubes in nuclear power plants. SCC tests were carried out to investigate the performance of TiO2 on several types of SG tube materials. The SCC tests were conducted by using an m-RUB specimen in a 10% NaOH solution at a temperature of 315 degrees C. The test with the addition of TiO2 showed a decrease in the SCC rate for the SG tubing materials. In order to improve the inhibition property in a crevice of TiO2, a sonochemical technique was applied to reduce the size of the TiO2 particle. From the SCC tests with the RUB specimen, the SG tube materials showed an enhanced cracking resistance with the addition of nano-sized TiO2 and the surface property was also changed.

  2. The Primary Water Stress Corrosion Cracking Mechanism of Alloy 600 Steam Generator Tubes: Materials Perspective

    International Nuclear Information System (INIS)

    Kim, Youngsuk; Kim, Sungsoo; Kim, Daewhan

    2013-01-01

    The problem is that intergranular (IG) cracking of austenitic Fe-Cr-Ni alloys occurs even in Ar with no corrosion or oxidation of grain boundaries being accompanied. This fact suggests that IG cracking has nothing to do with grain boundary (GB) corrosion or oxidation. This fact cast a doubt about the current notion that applied stresses are required to initiate IG cracking or PWSCC. These facts indicate that PWSCC is closely related to internal factors of materials, not to external factors such as grain boundary oxidation or corrosion or applied stresses. Given that austenitic alloys including Alloy 600 are a kind of solid solution alloys with alloying elements dissolved in the matrix as solutes, ordering of alloying elements of Fe, Cr and Ni occur in Alloy 600 during exposure to reactor operating condition. We suggest that atomic ordering is the main internal factor to govern PWSCC or IG cracking of austenitic Fe-Cr-Ni alloys because lattice contraction due to atomic ordering induces internal stresses which are large enough to cause GB cracking. The aim of this work is to provide experimental evidence for our suggestion. To this end, water quenching (WQ) or air cooling (AC) or furnace cooling (FC) was applied respectively to Alloy 600 after solution treatment at 1095 .deg. C for 0.5h to make Alloy 600 with either disorder (DO) or different degrees of short range order, respectively. Alloy 600 showed lattice contraction upon aging at 400 .deg. C whose extent increased with increasing cooling rate: the water-quenched (WQ) Alloy 600 exhibited the largest amount of lattice contraction than the furnace-cooled (FC) or air-cooled (AC) one. Yonezawa's experiments have indeed shown that the WQ-Alloy 600 with the largest amount of lattice contraction upon aging at 400 .deg. C is the most susceptible to PWSCC when compared to the AC- or FC-Alloy 600 with the lesser amount of lattice contraction. These observations demonstrate, for the first time, that PWSCC of Alloy 600 is

  3. Study on antioxidant experiment on forged steel tube sheet and tube hole for steam generator

    International Nuclear Information System (INIS)

    Zong Hai; Wang Detai; Ding Yang

    2012-01-01

    Antioxidant experiment on forged steel tube sheet and tube hole for steam generator was studied and the influence of different simulated heat treatments on the antioxidant performance of tube sheet and tube hole was made. The influence of different antioxidant methods on the size of tube hole was drawn. Furthermore, the change of size and weight of 18MnD5 forged steel tube sheet on the condition of different simulated heat treatments was also studied. The analytical results have proved reference information for the use of 18MnD5 material and for key processes of processing tube hole and wearing and expanding U-style tube. (authors)

  4. Steam generator tube integrity program

    International Nuclear Information System (INIS)

    Diercks, D.R.; Shack, W.J.

    1996-01-01

    The degradation of steam generator tubes in pressurized water nuclear reactors continues to be a serious problem, and the US Nuclear Regulatory Commission (NRC) is developing a performance-based rule and regulatory guide for steam generator tube integrity. To support the evaluation of industry-proposed implementation of these performance-based criteria, the NRC is sponsoring a new research program at Argonne National Laboratory on steam generator tubing degradation. The objective of the new program is to provide the necessary experimental data and predictive correlations and models that will permit the NRC to independently evaluate the integrity of steam generator tubes. The technical work in the program is divided into four tasks, (1) assessment of inspection reliability, (2) research on in-service inspection technology, (3) research on degradation modes and integrity, and (4) development of methodology and technical assessments for current and emerging regulatory issues. The objectives of and planned research activities under each of these four tasks are described here. (orig.)

  5. Steam generator materials

    International Nuclear Information System (INIS)

    Kim, Joung Soo; Han, J. H.; Kim, H. P.; Lim, Y. S.; Lee, D. H.; Suh, J. H.; Hwang, S. S.; Hur, D. H.; Kim, D. J.; Kim, Y. H.

    2002-05-01

    In order to keep the nuclear power plant(NPP)s safe and increase their operating efficiency, axial stress corrosion cracking(SCC)(IGA/IGSCC, PWSCC, PbSCC) test techniques were developed and SCC property data of the archive steam generator tubing materials having been used in nuclear power plants operating in Korea were produced. The data obtained in this study were data-based, which will be used to clarify the damage mechanisms, to operate the plants safely, and to increase the lifetime of the tubing. In addition, the basic technologies for the improvement of the SCC property of the tubing materials, for new SCC inhibition, for damaged tube repair, and for manufacturing processes of the tubing were developed. In the 1 phase of this long term research, basic SCC test data obtained from the archive steam generator tubing materials used in NPPs operating in Korea were established. These basic technologies developed in the 1 phase will be used in developing process optimization during the 2 phase in order to develop application technologies to the field nuclear power plants

  6. Tube Formation in Nanoscale Materials

    Directory of Open Access Journals (Sweden)

    Yan Chenglin

    2008-01-01

    Full Text Available Abstract The formation of tubular nanostructures normally requires layered, anisotropic, or pseudo-layered crystal structures, while inorganic compounds typically do not possess such structures, inorganic nanotubes thus have been a hot topic in the past decade. In this article, we review recent research activities on nanotubes fabrication and focus on three novel synthetic strategies for generating nanotubes from inorganic materials that do not have a layered structure. Specifically, thermal oxidation method based on gas–solid reaction to porous CuO nanotubes has been successfully established, semiconductor ZnS and Nb2O5nanotubes have been prepared by employing sacrificial template strategy based on liquid–solid reaction, and an in situ template method has been developed for the preparation of ZnO taper tubes through a chemical etching reaction. We have described the nanotube formation processes and illustrated the detailed key factors during their growth. The proposed mechanisms are presented for nanotube fabrication and the important pioneering studies are discussed on the rational design and fabrication of functional materials with tubular structures. It is the intention of this contribution to provide a brief account of these research activities.

  7. Tube Formation in Nanoscale Materials.

    Science.gov (United States)

    Yan, Chenglin; Liu, Jun; Liu, Fei; Wu, Junshu; Gao, Kun; Xue, Dongfeng

    2008-12-01

    The formation of tubular nanostructures normally requires layered, anisotropic, or pseudo-layered crystal structures, while inorganic compounds typically do not possess such structures, inorganic nanotubes thus have been a hot topic in the past decade. In this article, we review recent research activities on nanotubes fabrication and focus on three novel synthetic strategies for generating nanotubes from inorganic materials that do not have a layered structure. Specifically, thermal oxidation method based on gas-solid reaction to porous CuO nanotubes has been successfully established, semiconductor ZnS and Nb(2)O(5) nanotubes have been prepared by employing sacrificial template strategy based on liquid-solid reaction, and an in situ template method has been developed for the preparation of ZnO taper tubes through a chemical etching reaction. We have described the nanotube formation processes and illustrated the detailed key factors during their growth. The proposed mechanisms are presented for nanotube fabrication and the important pioneering studies are discussed on the rational design and fabrication of functional materials with tubular structures. It is the intention of this contribution to provide a brief account of these research activities.

  8. Fracture toughness determination in steam generator tubes

    International Nuclear Information System (INIS)

    Bergant M; Yawny, A; Perez Ipina, J

    2012-01-01

    The assessment of the structural integrity of steam generator tubes in nuclear power plants deserved increasing attention in the last years due to the negative impact related to their failures. In this context, elastic plastic fracture mechanics (EPFM) methodology appears as a potential tool for the analysis. The application of EPFM requires, necessarily, knowledge of two aspects, i.e., the driving force estimation in terms of an elastic plastic toughness parameter (e.g., J) and the experimental measurement of the fracture toughness of the material (e.g., the material J-resistance curve). The present work describes the development of a non standardized experimental technique aimed to determine J-resistance curves for steam generator tubes with circumferential through wall cracks. The tubes were made of Incoloy 800 (Ni: 30.0-35.0; Cr: 19.0-23.0; Fe: 35.5 min, % in weight). Due to its austenitic microstructure, this alloy shows very high toughness and is widely used in applications where a good corrosion resistance in aqueous environment or an excellent oxidation resistance in high temperature environment is required. Finally, a procedure for the structural integrity analysis of steam generator tubes with crack-like defects, based on a FAD diagram (Failure Assessment Diagram), is briefly described (author)

  9. Origami interleaved tube cellular materials

    International Nuclear Information System (INIS)

    Cheung, Kenneth C; Tachi, Tomohiro; Calisch, Sam; Miura, Koryo

    2014-01-01

    A novel origami cellular material based on a deployable cellular origami structure is described. The structure is bi-directionally flat-foldable in two orthogonal (x and y) directions and is relatively stiff in the third orthogonal (z) direction. While such mechanical orthotropicity is well known in cellular materials with extruded two dimensional geometry, the interleaved tube geometry presented here consists of two orthogonal axes of interleaved tubes with high interfacial surface area and relative volume that changes with fold-state. In addition, the foldability still allows for fabrication by a flat lamination process, similar to methods used for conventional expanded two dimensional cellular materials. This article presents the geometric characteristics of the structure together with corresponding kinematic and mechanical modeling, explaining the orthotropic elastic behavior of the structure with classical dimensional scaling analysis. (paper)

  10. In service inspection for steam generator tubes

    International Nuclear Information System (INIS)

    Comby, R.; Eyrolles, Ph.

    1988-01-01

    In this paper the authors show the means putting in place for examination of steam generators tubes. These means (eddy current probes, ultrasonic testing) associated with a knowledge on degradation phenomena allow mapping controlled tubes and limiting undesirable obturations [fr

  11. Failure analysis of retired steam generator tubings

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hong Pyo; Kim, J. S.; Hwang, S. S. and others

    2005-04-15

    Degradation of steam generator leads to forced outage and extension of outage, which causes increase in repair cost, cost of purchasing replacement power and radiation exposure of workers. Steam generator tube rupture incident occurred in Uljin 4 in 2002, which made public sensitive to nuclear power plant. To keep nuclear energy as a main energy source, integrity of steam generator should be demonstrated. Quantitative relationship between ECT(eddy current test) signal and crack size is needed in assesment of integrity of steam generator in pressurized water reactor. However, it is not fully established for application in industry. Retired steam generator of Kori 1 has many kinds of crack such as circumferential and axial primary water stress corrosion crack and outer diameter stress corrosion crack(ODSCC). So, it can be used in qualifying and improving ECT technology and in condition monitoring assesment for crack detected in ISI(in service inspection). In addition, examination of pulled tube of Kori 1 retired steam generator will give information about effectiveness of non welded sleeving technology which was employed to repair defect tubes and remedial action which was applied to mitigate ODSCC. In this project, hardware such as semi hot lab. for pulled tube examination and modification transportation cask for pulled tube and software such as procedure of transportation of radioactive steam generator tube and non-destructive and destructive examination of pulled tube were established. Non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in semi hot lab. Remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. Electrochemical decontamination technology for pulled tube was developed to reduce radiation exposure and enhance effectiveness of pulled tube examination. Multiparameter algorithm developed at ANL, USA was

  12. Destructive examination and analyses of pulled steam generator tubes

    International Nuclear Information System (INIS)

    Kim, Dong Jin; Kim, H. P.; Kim, J. S.; Lim, Y. S.; Hwang, S. S.; Kim, S. W.; Jeong, M. K.; Hong, J. H.; Kim, W. W.

    2011-07-01

    Steam generator model F in Kori 3, Younggwang 2 and Younggwang 1 as 950MWe PWR was provided by Westinghouse. Steam generator tube made of Alloy 600TT material (outer diameter 0.688'(17.475 mm), thickness 0.04'(1.016mm) in Blairsville was provided by Huntington alloys. Steam generator in Ulchin 4 as 1000MWe KHNP PWR was manufactured by Doosan heavy industry and steam generator tubes were manufactured by B and W (Bobcock and Wilcox). Alloy 600 MA was used as steam generator tubing material of outer diameter 19.05mm and thickness 1.07mm. Five tubes of Alloy 600TT which showed crack signal from non-destructive examination were pulled from Kori 3, Younggwang 2 and Younggwang 1. Two tubes which showed crack signal from non-destructive examination were pulled from Ulchin 4. For the pulled tubes, KAERI performed destructive examination. Through the destructive examination, the existence of cracks were confirmed and the cause of crack was investigated. Remedy was also suggested to mitigate the present circumstances. Stress corrosion cracking (SCC) was observed for Alloy 600TT tubes. It was recommended that the sludge content should be lowered, deleterious elements and MRI should be maintained continuously. For Ulchin 4, SCC was confirmed for Alloy 600MA tubes. It is necessary to lower and remove the sludge in the near term. In the longer term, replacement of steam generator was suggested

  13. Repair technology for steam generator tubes

    International Nuclear Information System (INIS)

    Kim, Seung Ho; Jung, Hyun Kyu; Jung, Seung Ho; Kim, Chang Hoi; Jung, Young Moo; Seo, Yong Chil; Kim, Jung Su; Seo, Moo Hong

    2001-02-01

    The most commonly used sleeving materials are thermally treated Alloy 600 and thermally treated Alloy 690 Alloy. Currently, thermally treated Alloy 690 and Alloy 800 are being offered although Alloy 800 has not been licensed in the US. To install sleeve, joint strength, leak tightness, PWSCC resistance, evaluation on process parameter range and the effect of equipments and procedures on repair plan and radiation damage have to be investigated before sleeving. ABB CE provides three type of leak tight Alloy 690 TIG welded and PLUSS sleeve. Currently, Direct Tube Repair technique using Nd:YAG laser has been developed by ABB CE and Westinghouse. FTI has brazed and kinetic sleeve designs for recirculating steam generator and hydraulic and rolled sleeve designs for one-through steam generators. Westinghouse provides HEJ, brazed and laser welded sleeve design. When sleeve is installed in order to repair the damaged S/G tubes, it is certain that defects can be occurred due to the plastic induced stress and thermal stress. Therefore it is important to minimize the residual stress. FTI provides the electrosleeve technique as a future repair candidate using electroplating

  14. Repair technology for steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Ho; Jung, Hyun Kyu; Jung, Seung Ho; Kim, Chang Hoi; Jung, Young Moo; Seo, Yong Chil; Kim, Jung Su; Seo, Moo Hong

    2001-02-01

    The most commonly used sleeving materials are thermally treated Alloy 600 and thermally treated Alloy 690 Alloy. Currently, thermally treated Alloy 690 and Alloy 800 are being offered although Alloy 800 has not been licensed in the US. To install sleeve, joint strength, leak tightness, PWSCC resistance, evaluation on process parameter range and the effect of equipments and procedures on repair plan and radiation damage have to be investigated before sleeving. ABB CE provides three type of leak tight Alloy 690 TIG welded and PLUSS sleeve. Currently, Direct Tube Repair technique using Nd:YAG laser has been developed by ABB CE and Westinghouse. FTI has brazed and kinetic sleeve designs for recirculating steam generator and hydraulic and rolled sleeve designs for one-through steam generators. Westinghouse provides HEJ, brazed and laser welded sleeve design. When sleeve is installed in order to repair the damaged S/G tubes, it is certain that defects can be occurred due to the plastic induced stress and thermal stress. Therefore it is important to minimize the residual stress. FTI provides the electrosleeve technique as a future repair candidate using electroplating.

  15. Corrosion performance of tube support materials

    International Nuclear Information System (INIS)

    Malagola, P.

    1985-01-01

    The problem of denting in steam generators leads to change in the conception of the tube support plates. A new material is now used for this component, a 13% Cr steel, which composition has been adjusted for weldability and mechanical resistance criteria. The geometry of trefoil support plate (TSP) has also been improved, using a broached TSP (quadrifoiled holes) instead of a drilled TSP. Tests have been performed on 13% Cr and C-steel broached TSP, and drilled TSP, to confirm the better resistance to denting of this new configuration

  16. Steam generator tube denting simulation testing

    International Nuclear Information System (INIS)

    Battaglia, P.J.; Singley, W.J.

    1978-02-01

    Tube denting has been reported in steam generators in a number of commercial nuclear power plants in recent years. In order to aid in understanding of the mechanism leading to tube denting in the steam generators, a Bettis laboratory test program was initiated to attempt to reproduce tube denting and to investigate the effects of chemistry, design, and temperature. The results of the tests indicate that denting can be reproduced in the small model steam generator test apparatus devised for this testing. Denting was observed under carbon steel support plates in seawater-contaminated secondary water in a test with hydrazine-ammonia chemistry and a primary water inlet temperature of 590 0 F and in a test with hydrazine-morpholine chemistry and a primary water inlet temperature of 545 0 F. The parameters of these two tests simulate conditions in a typical commercial steam generator and the Shippingport Atomic Power Station steam generators, respectively

  17. Circumferential cracking of steam generator tubes

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    1997-04-01

    On April 28, 1995, the U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 95-03, open-quote Circumferential Cracking of Steam Generator Tubes.close-quote GL 95-03 was issued to obtain information needed to verify licensee compliance with existing regulatory requirements regarding the integrity of steam generator tubes in domestic pressurized-water reactors (PWRs). This report briefly describes the design and function of domestic steam generators and summarizes the staff's assessment of the responses to GL 95-03. The report concludes with several observations related to steam generator operating experience. This report is intended to be representative of significant operating experience pertaining to circumferential cracking of steam generator tubes from April 1995 through December 1996. Operating experience prior to April 1995 is discussed throughout the report, as necessary, for completeness

  18. A Flue Gas Tube for Thermoelectric Generator

    DEFF Research Database (Denmark)

    2013-01-01

    The invention relates to a flue gas tube (FGT) (1) for generation of thermoelectric power having thermoelectric elements (8) that are integrated in the tube. The FTG may be used in combined heat and power (CHP) system (13) to produce directly electricity from waste heat from, e.g. a biomass boiler....... The CHP system may also be operated in a heating or cooling mode, thus being able to heat or cool water by feeding electricity to the system....

  19. Internal ultrasonic testing of steam generator tubes

    International Nuclear Information System (INIS)

    Furlan, J.; Soleille, G.; Chalaye, H.

    1983-01-01

    The ''in situ'' inspection of steam generator tubes uses generally Foucault currents before starting and along its life. This inspection aims at searching cracks and corrosion defects. The Foucault current method is quite badly adapted to ''closed crack'' detection, for it doesn't introduce neither resistivity or magnetic permeability variation, or lack of matter. More, it is sensible to the magnetic properties of the tube itself and to its environment (tubular or support plates). It is why, this first systematic inspection has to be completed by an ultrasonic one allowing to bring new elements in the uncertain cases. A device with an internal probe has been developed. It ''lights'' the tube wall with the aid of a transducer of which beam reflects on a mirror. Operating conditions are the same as for Foucault current testing, that is to say the probe moves inside the tube without rotation of the device (bent parts are excluded) [fr

  20. Influences of Alloying Element and Annealing on the Microstructure and Corrosion Resistance of Steam Generator Tubing Materials of Nuclear Power Plant (I)

    International Nuclear Information System (INIS)

    Kim, Young Sik; Pari, Yong Soo; Kuk, Il Hiun

    1996-01-01

    Influences of alloying elements and annealing heat treatments on Alloy 690 and Alloy 600 for steam generator tubing materials of nuclear power plants were studied. OM, SEM, TEM, and XRD analyses were used to study the microstructural changes of the alloys. Mechanical properties were investigated by means of tension tests and Rockwell hardness tests, and corrosion resistance was evaluated using the anodic polarization tests and the 65% boiling nitric acid immersion tests. Increasing the carbon content of Alloy 690, the hardness and tensile strength were increased, but the elongation and grain size were decreased. However, increasing the annealing temperature, the tensile strength and hardness were decreased, but the elongation and grain size were increased. Increasing the carbon content of Alloy 690, the results of the anodic polarization tests and the nitric acid immersion tests showed that the annealing temperature to reveal a minimum corrosion rate was increased. This behavior seemed to be due to the combination of the solid solution of carbon in the matrix and grain growth with annealing. In this work, the corrosion properties of Alloy 690 were better than that of Alloy 600, and the range of the optimum annealing temperature of Alloy 690 was from 1100 .deg. C to 1150 .deg. C

  1. Flow induced pulsations generated in corrugated tubes

    NARCIS (Netherlands)

    Belfroid, S.P.C.; Swindell, R.; Tummers, R.

    2008-01-01

    Corrugated tubes can produce a tonal noise when used for gas transport, for instance in the case of flexible risers. The whistling sound is generated by shear layer instability due to the boundary layer separation at each corrugation. This whistling is examined by investigating the frequency,

  2. Eddy current testing of steam generator tubes

    International Nuclear Information System (INIS)

    Neumaier, P.

    1981-01-01

    A rotating probe is described for improving the inspection of tubes and end plate in steam generators. The method allows a representation of the whole defect, consequently the observer is able to determine directly the type of defect, signal processing in-line or off-line is possible [fr

  3. Pulse tube coolers for Meteosat third generation

    International Nuclear Information System (INIS)

    Butterworth, James; Aigouy, Gérald; Chassaing, Clement; Debray, Benoît; Huguet, Alexandre

    2014-01-01

    Air Liquide's Large Pulse Tube Coolers (LPTC) will be used to cool the focal planes of the Infrared Sounder (IRS) and Flexible Combined Imager (FCI) instruments aboard the ESA/Eumetsat satellites Meteosat Third Generation (MTG). This cooler consists of an opposed piston linear compressor driving a pulse tube cold head and the associated drive electronics including temperature regulation and vibration cancellation algorithms. Preparations for flight qualification of the cooler are now underway. In this paper we present results of the optimization and qualification activities as well as an update on endurance testing

  4. Method to detect steam generator tube leakage

    International Nuclear Information System (INIS)

    Watabe, Kiyomi

    1994-01-01

    It is important for plant operation to detect minor leakages from the steam generator tube at an early stage, thus, leakage detection has been performed using a condenser air ejector gas monitor and a steam generator blow down monitor, etc. In this study highly-sensitive main steam line monitors have been developed in order to identify leakages in the steam generator more quickly and accurately. The performance of the monitors was verified and the demonstration test at the actual plant was conducted for their intended application to the plants. (author)

  5. U-tube steam generator predictions

    International Nuclear Information System (INIS)

    Hassan, Y.A.; Kalyanasundaram, M.

    1991-01-01

    This paper reports on the development of a RELAP5/MOD2 computer code model for a Model Boiler-2 U-tube steam generator (UTSG) to predict the thermal-hydraulic response of a UTSG during steady-state operation and for a loss-of-feedwater (LOF) transient. Steady-state conditions calculated by RELAP5 are compared with the measured data. The calculated heat transfer from the primary to the secondary side of the steam generator is found to be underpredicted by 30%. The heat transfer correlations used in existing thermal-hydraulic codes are developed for flow inside individual tubes and not for flow around tube bundles. Consequently, the secondary convective heat transfer is not accurately predicted by the codes. A revised version of the RELAP5 code with modified heat transfer correlations reasonably predicts the primary to the secondary heat transfer in bundle environments. Improved heat fluxes and heat transfer coefficients are obtained during steady-state and LOF accident transients. Steady-state behavior of the Semiscale MOD-2C steam generator is also computed with both the original and the revised versions of the code. Good agreement is achieved between the predictions and the test data when the modified heat transfer correlations are utilized

  6. Automated Diagnosis and Classification of Steam Generator Tube Defects

    International Nuclear Information System (INIS)

    Garcia, Gabe V.

    2004-01-01

    A major cause of failure in nuclear steam generators is tube degradation. Tube defects are divided into seven categories, one of which is intergranular attack/stress corrosion cracking (IGA/SCC). Defects of this type usually begin on the outer surface of the tubes and propagate both inward and laterally. In many cases these defects occur at or near the tube support plates. Several different methods exist for the nondestructive evaluation of nuclear steam generator tubes for defect characterization

  7. Automated Diagnosis and Classification of Steam Generator Tube Defects

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Gabe V. Garcia

    2004-10-01

    A major cause of failure in nuclear steam generators is tube degradation. Tube defects are divided into seven categories, one of which is intergranular attack/stress corrosion cracking (IGA/SCC). Defects of this type usually begin on the outer surface of the tubes and propagate both inward and laterally. In many cases these defects occur at or near the tube support plates. Several different methods exist for the nondestructive evaluation of nuclear steam generator tubes for defect characterization.

  8. Research perspectives on the evaluation of steam generator tube integrity

    International Nuclear Information System (INIS)

    Muscara, J.; Diercks, D.R.; Majumdar, S.; Kupperman, D.S.; Bakhtiari, S.; Shack, W.J.

    2002-01-01

    Industry efforts have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, stress corrosions cracking (SCC) and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding, there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by 'plug or repair on detection' because current NDE techniques for characterization of flaws and the knowledge of SCC crack growth rates are not accurate enough to permit continued operation. Replacement generators with improved designs and materials have performed well to date, but previous experience with the appearance of some types of SCC in Alloy 600 after 10 years or more of operation and laboratory results suggest additional understanding of corrosion performance of these materials is needed. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators. (author)

  9. Hideout in steam generator tube deposits

    International Nuclear Information System (INIS)

    Balakrishnan, P.V.; Franklin, K.J.; Turner, C.W.

    1998-05-01

    Hideout in deposits on steam generator tubes was studied using tubes coated with magnetite. Hideout from sodium chloride solutions at 279 degrees C was followed using an on-line high-temperature conductivity probe, as well as by chemical analysis of solution samples from the autoclave in which the studies were done. Significant hideout was observed only at a heat flux greater than 200 kW/m 2 , corresponding to a temperature drop greater than 2 degrees C across the deposits. The concentration factor resulting from the hideout increased highly non-linearly with the heat flux (varying as high as the fourth power of the heat flux). The decrease in the apparent concentration factor with increasing deposit thickness suggested that the pores in the deposit were occupied by a mixture of steam and water, which is consistent with the conclusion from the thermal conductivity measurements on deposits in a separate study. Analyses of the deposits after the hideout tests showed no evidence of any hidden-out solute species, probably due to the concentrations being very near the detection limits and to their escape from the deposit as the tests were being ended. This study showed that hideout in deposits may concentrate solutes in the steam generator bulk water by a factor as high as 2 x 10 3 . Corrosion was evident under the deposit in some tests, with some chromium enrichment on the surface of the tube. Chromium enrichment usually indicates an acidic environment, but the mobility required of chromium to become incorporated into the thick magnetite deposit may indicate corrosion under an alkaline environment. An alkaline environment could result from preferential accumulation of sodium in the solution in the deposit during the hideout process. (author)

  10. EP 1000 steam generator tube rupture analyses

    International Nuclear Information System (INIS)

    Saiu, G.; Frogheri, M.; Schulz, T.L.

    2001-01-01

    European electrical utility organizations together with Westinghouse and Ansaldo are participating in a program to utilize the Westinghouse passive nuclear plant technology to develop a plant which meets the European Utility Requirements (EUR) and is expected to be licensable in Europe. The program was initiated in 1994 and the plant is designated EP1000. The EP1000 design is notable for simplicity that comes from a reliance on passive safety systems to enhance plant safety. The use of passive safety systems has provided significant and measurable improvements in plant simplification, safety, reliability, investment protection and plant costs. These systems use only natural forces such as gravity, natural circulation, and compressed gas to provide the driving forces for the systems to adequately cool the reactor core following an initiating event. The EP1000 builds up on the Westinghouse passive nuclear plant technology to enhance plant safety and meet European Utility Requirements and specific European National Safety Criteria. This paper summarizes the main results of the Steam Generator Tube Rupture (SGTR) analysis activity, performed in Phase 2B of the European Passive Plant Program. The purpose of the study is to provide evidence that the passive safety system performance provides a significant improvement in terms of safety, providing significant margins to steam generator overfilling and reducing the need for operator actions. The behavior of the EP1000 plant following SGTR accidents has been analyzed by means of the RELAP5/Mod3.2 code. Sensitivity cases were performed, to address the impact of varying the number of steam generator tubes that rupture, and the potential adverse interactions that could result from operation of control systems (i.e., Chemical and Volume Control System, Startup Feedwater). Analyses have also been performed to define and verify improved protection system logic to avoid possible steam generator safety valve challenges both in the

  11. Finned Tube With Vortex Generators For A Heat Exchanger.

    Science.gov (United States)

    Sohal, Monohar S.; O'Brien, James E.

    2004-09-14

    A system for and method of manufacturing a finned tube for a heat exchanger is disclosed herein. A continuous fin strip is provided with at least one pair of vortex generators. A tube is rotated and linearly displaced while the continuous fin strip with vortex generators is spirally wrapped around the tube.

  12. Sleeve type repair of degraded nuclear steam generator tubes

    International Nuclear Information System (INIS)

    Ayres, P.S.; Stark, L.E.; Feldstein, J.G.; Fu, T.

    1986-01-01

    A sealable sleeve is described for insertion into the repair of a degraded tube which consists of: a hollow core inner member of the same material as the degraded tube; a thinner outer member of substantially pure nickel and resistant to corrosive attack, the outer member being metallurgically bonded with the inner member; an expanded portion of the sleeve at one end for positioning in the tube within a tube sheet; a multiplicity of grooves formed in and adjacent to the other end of the sleeve which extends into the free-standing portion of the tube beyond the tube sheet, and a noble metal braze material contained in the grooves

  13. Neutron generator tube ion source control apparatus

    International Nuclear Information System (INIS)

    Bridges, J.R.

    1982-01-01

    A pulsed neutron well logging system includes a neutron generator tube of the deuterium-tritium accelerator type and an ion source control apparatus providing extremely sharply time-defined neutron pulses. A low voltage control pulse supplied to an input by timing circuits turns a power FET on via a buffer-driver whereby a 2000 volt pulse is produced in the secondary of a pulse transformer and applied to the ion source of the tube. A rapid fall in this ion source control pulse is ensured by a quenching circuit wherein a one-shot responds to the falling edge of the control pulse and produces a 3 microsecond delay to compensate for the propagation delay. A second one-shot is triggered by the falling edge of the output of the first one-shot and gives an 8 microsecond pulse to turn on the power FET which, via an isolation transformer turns on a series-connected transistor to ground the secondary of the pulse transformer and the ion source. (author)

  14. Repair technique for steam generator tubes using electroforming

    International Nuclear Information System (INIS)

    Kim, Seung Ho; Jeong, Hyun Kyu; Seo, Moon Hong

    2001-07-01

    Pickering B CANDU Unit 5 had experienced leakage at sleeve/tube joint due to severe and local pitting in 1992δ1993. One year later, OHT developed electrosleeving techniques for steam generator tube repair which was applied at Pickering B CANDU Unit 5, Oconee Unit 1 and Callaway in 1994, 1995 and 1999 respectively. In the results of electrosleeved tube test, electrosleeve materials were stronger than mother tubes in mechanical properties and corrosion resistance under design criteria. Two analytical models were originally developed for estimating the failure temperature under severe accident transients. Electrosleeve, a structural layer of fine grained nickel is electroformed onto the strike by circulating an aqueous solution of Ni sulfate or sulfamate with NiCO3. The patents published by FTI said that the electrolyte for electroforming the structural layer contains a pinning agent to inhibit growth of metal grains in the electroformed layer. The pinning agent contains phosphoric, phosphorous acid, molybdenum. In localization of electrosleeving, there are some problems like as 1)low plating rate, 2)high residual stress, 3)alloy composition, 4)low material properties at high temperature. Ni-Fe plating exhibit anomalous codeposition; that is less noble metal, Fe, deposits preferentially to the more noble metal, Ni. Ductility decrease and residual stress increase with increase of Fe content in plate layer. Addition of particle size of 10δ400μm makes residual stress compressive in plate layer. Composite plating show excellent high temperature properties

  15. Radioactivity transport following steam generator tube rupture

    International Nuclear Information System (INIS)

    Hopenfeld, J.

    1985-03-01

    A review of the capabilities of the CITADEL computer code as well as plant experience to project radioactivity releases following a steam generator tube rupture in PWR's shows that certain experimental data are needed for reliable off-site dose predictions. This article defines five parameters which are the key for such predictions and discusses the functional dependence of these parameters on various operational variables. The above parameters can be used in conjunction with CITADEL or they can be inserted in the appropriate equations which then conveniently can be programmed as a subroutine in thermal-hydraulic system codes. A joint Westinghouse, Electric Power Research Institute and Nuclear Regulatory Commission Program aimed at obtaining the five parameters empirically is described

  16. Anatomy education for the YouTube generation.

    Science.gov (United States)

    Barry, Denis S; Marzouk, Fadi; Chulak-Oglu, Kyrylo; Bennett, Deirdre; Tierney, Paul; O'Keeffe, Gerard W

    2016-01-01

    Anatomy remains a cornerstone of medical education despite challenges that have seen a significant reduction in contact hours over recent decades; however, the rise of the "YouTube Generation" or "Generation Connected" (Gen C), offers new possibilities for anatomy education. Gen C, which consists of 80% Millennials, actively interact with social media and integrate it into their education experience. Most are willing to merge their online presence with their degree programs by engaging with course materials and sharing their knowledge freely using these platforms. This integration of social media into undergraduate learning, and the attitudes and mindset of Gen C, who routinely creates and publishes blogs, podcasts, and videos online, has changed traditional learning approaches and the student/teacher relationship. To gauge this, second year undergraduate medical and radiation therapy students (n = 73) were surveyed regarding their use of online social media in relation to anatomy learning. The vast majority of students had employed web-based platforms to source information with 78% using YouTube as their primary source of anatomy-related video clips. These findings suggest that the academic anatomy community may find value in the integration of social media into blended learning approaches in anatomy programs. This will ensure continued connection with the YouTube generation of students while also allowing for academic and ethical oversight regarding the use of online video clips whose provenance may not otherwise be known. © 2015 American Association of Anatomists.

  17. Material modeling for multistage tube hydroforming process simulation

    Science.gov (United States)

    Saboori, Mehdi

    The Aerospace industries of the 21st century demand the use of cutting edge materials and manufacturing technology. New manufacturing methods such as hydroforming are relatively new and are being used to produce commercial vehicles. This process allows for part consolidation and reducing the number of parts in an assembly compared to conventional methods such as stamping, press forming and welding of multiple components. Hydroforming in particular, provides an endless opportunity to achieve multiple crosssectional shapes in a single tube. A single tube can be pre-bent and subsequently hydroformed to create an entire component assembly instead of welding many smaller sheet metal sections together. The knowledge of tube hydroforming for aerospace materials is not well developed yet, thus new methods are required to predict and study the formability, and the critical forming limits for aerospace materials. In order to have a better understanding of the formability and the mechanical properties of aerospace materials, a novel online measurement approach based on free expansion test is developed using a 3D automated deformation measurement system (AramisRTM) to extract the coordinates of the bulge profile during the test. These coordinates are used to calculate the circumferential and longitudinal curvatures, which are utilized to determine the effective stresses and effective strains at different stages of the tube hydroforming process. In the second step, two different methods, a weighted average method and a new hardening function are utilized to define accurately the true stress-strain curve for post-necking regime of different aerospace alloys, such as inconel 718 (IN 718), stainless steel 321 (SS 321) and titanium (Ti6Al4V). The flow curves are employed in the simulation of the dome height test, which is utilized for generating the forming limit diagrams (FLDs). Then, the effect of stress triaxiality, the stress concentration factor and the effective plastic

  18. Development of safety evaluation technique of steam generator tubes for the next generation

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyuk Sang; Kim, I. S.; Ann, Se Jin; Lee, S. J.; Seo, M. S.; Lee, Y. H.; Kim, J. H.; Hong, J. G. [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-02-15

    Subject 1 - a technique for predicting the SCC susceptibility of steam generator tube material based on the repassivation kinetics was developed and the effects of Pb in the repassivation rate and SCC susceptibility rate of tube material was investigated with this technique. An alloy with a higher slope value of log i(t) vs. q(t) plot based on the current transient curve obtained by scratch test and a lower slope value log i(t) vs. l/q(t) plot (cBV) is repassivated faster with a more protective passive film and it can be predicted that it will show higher resistance to SCC. With PbO addition in all solution studied (pH 4, pH 10, Cl- containing pH 4), alloy 690TT showed decreased repassivation rate. So it can be predict that PbO addition lower the resistance of SCC of steam generator tune material. Subject 2 - SG wear testing of tube and support materials has been conducted at various load and sliding amplitude in air environment. The results showed effect of normal load and sliding amplitude on SG tube wear damage. It was also shown that, for predominantly sliding motion, the SG wear coefficient of work-rate model is lower for Inconel 690TT compared with inconel 600MA. SG tube wear data show that, for work-rates ranging from 4 to 25mW, average tube wear coefficient of 43.76{approx}54.05 X 10{sup 15} Pa{sup -1} for Inconel 600MA and 26.88{approx}33.94 X 10{sup -15} Pa{sup 1} for Inconel 690TT against 405 and 409 stainless steels.

  19. Experimental study of tube/support impact forces in multi-span PWR steam generator tubes

    International Nuclear Information System (INIS)

    Axisa, F.; Desseaux, A.; Gibert, R.J.

    1984-12-01

    The vibro-impact response of a straight part of a steam generator tube is investigated experimentally and using numerical simulation with the aim to relate tube overall dynamics with excitation and tube-support clearance. Configuration studied here corresponds to the tube being excited in only one direction at its first resonance presenting an antinode of vibration at the impacted support. Tests show namely that midspan displacement of tube is almost proportional to excitation level and clearance. Impact forces averaged over a cycle of vibration are almost proportional to excitation and poorly dependent on clearance. Results of numerical simulation are in fairly good agreement with test results

  20. On the distribution of temperatures in steam generator tubes at tube support plate Intersections

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.; Petelin, S.

    1995-01-01

    This analysis was initiated to examine the temperature fields in the steam generator tube in the vicinity of the tube support plates. It is assumed that the flow of the secondary coolant is severely disturbed there, which causes local heating of the tube surface. Different designs of tube support plates (a drilled hole - NE Krsko, broached trefoil and broached quatrefoil designs) were assessed and compared. Inside the drilled hole tube support plate, the temperature of the reactor coolant. Inside broached trefoil and quatrefoil support plates, the tube surface temperature reaches about 10K less than reactor coolant temperature. The most important result concerning the Krsko specific conditions is that the frequency of the detected defects can be correlated with the temperature of the tube outer surface and void fraction of the secondary coolant. (author)

  1. Corrosion and Rupture of Steam Generator Tubings in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-15

    This report is intended to provide corrosion engineers in the filed of nuclear energy with information on the corrosion and rupture behavior of steam generator tubing in PWRs. Various types of corrosion in PWR steam generator tubing have been reported all around the world, and countermeasures such as the addition of corrosion inhibitors, a water chemistry control, a tube plugging and sleeving have been applied. Steam generators equipped with alloy 600 tubing, which are not so resistant to a stress corrosion cracking (SCC), have generally been replaced with new steam generators made of alloy 690 TT (Thermally treated). Pull tube examination results which were performed of KAERI are summarized. The tubes were affected by a pitting, SCC, and a denting. Nondestructive examination method for the tubes and repair techniques are also reviewed. In addition, the regulatory guidance of some countries are reviewed. As a part of a tube integrity project in Korea, some results on a tube rupture and leak behaviors for axial cracks are also mentioned.

  2. Wear behavior of steam generator tubes in nuclear power plant operating condition

    International Nuclear Information System (INIS)

    Kim, In-Sup; Hong, Jin-Ki; Kim, Hyung-Nam; Jang, Ki-Sang

    2003-01-01

    Reciprocating sliding wear tests were performed on steam generator tubes materials at steam generator operating temperature. The material surfaces react with oxygen to form oxides. The oxide properties such as formation rate and mechanical properties are varied with the test temperature and alloy composition. So, it is important to investigate the wear properties of each steam generator tube materials in steam generator operating condition. The tests results indicated that the wear coefficient in work rate model of alloy 690 was faster than that of alloy 800. From the scanning electron microscopy observation, the wear scars were similar each other and worn surfaces were covered with oxide layers. It seemed that the oxide layers were formed by wear debris sintering or cold welding and these layer properties affected the wear rate of steam generator tube materials. (author)

  3. Characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation

    Energy Technology Data Exchange (ETDEWEB)

    Pagan, Sandra [Ontario Power Generation, 889 Brock Road, Pickering, Ontario (Canada); Duan Xinjian [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario (Canada)], E-mail: duanx@aecl.ca; Kozluk, Michael J. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario (Canada); Mills, Brian; Goszczynski, Guylaine [Kinectrics Inc., 800 Kipling Avenue, Toronto, Ontario (Canada)

    2009-03-15

    The Canadian Nuclear Standard CSA N285.4 requires the periodic metallurgical examination of removed ex-service steam generator tubes. This paper describes the practices used for the characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation (OPG). It shows that there is no degradation of mechanical properties of Monel 400 tubes after 7-18 effective full power years (EFPY) of operation and Incoloy 800 tubes after more than 10 EFPY of operation.

  4. 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

  5. Multifrequency eddy current testing of helical tubes of steam generators

    International Nuclear Information System (INIS)

    Pigeon, M.; David, B.

    1983-06-01

    In the event of a water-sodium reaction in a steam-generator of a fast breeder reactor, it is necessary to test the tubes close to the leak to evaluate the damage. In SUPERPHENIX, the tubes are about 100m long and are coiled on a dead body. This report describes the equipment and the technic to test such tubes with multifrequency eddy current technics [fr

  6. Development of stabilizers for steam generator tube repair

    International Nuclear Information System (INIS)

    Au-Yang, M.K.

    1987-01-01

    Fluidelastic stability, turbulence-induced and vortex-induced vibration analysis of different types of stabilizers for repairing steam generator tubes are presented. The performances of the different designs are compared with that of a common basis - the virgin tube. It was found that in addition to permitting the stabilized tubes to remain in operation, the sleeve has the additional merit of being the best performer of all the designs. In addition, it is adaptable to remote installation. (orig.)

  7. Process technology development of Ni electroplating in steam generator tube

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joung Soo; Kim, H. P.; Lim, Y. S.; Kim, S. S.; Hwang, S. S.; Yi, Y. S.; Kim, D. J.; Jeong, M. K

    2006-07-15

    Alloy 600 having superior resistance to corrosion is used as a steam generator tubing in Nuclear Power Plants. In spite of its high corrosion resistance, there are many tubes which have experienced the corrosion problems such as SCC, pitting under high temperature and high pressure environments of NPP, leading to a menace to the safety of NPPs as well as economical loss. A commonly applied approach to rehabilitation has been to repair the damaged areas of the tubes via the insertion of tubular sleeves which are either welded or mechanically bonded at their extremities to the host tube. Such intrusive sleeves have weak points, such as the crevices, the tube deformation and an introduction of stress onto the host tube which then usually requires stress relief to improve the in-service life. However a lot of problems including these during and after repairing can be solved by Ni electroplating having excellent corrosion resistance to such as SCC. This work is related to optimum process development for Ni electrodeposition inside damaged steam generator tubing for repairing and the damage prevention. The optimum electroplating process for planar specimens was developed and the electrodeposition was performed successfully inside tube specimens by using the modified and improved anode probe. The Ni-electrodeposit plated on the inner surfaces of the tube specimens was confirmed to show excellent SCC resistance. A multiple electrodeposition facility for simultaneous electroplating inside three tubes at the same time was built and proved to work properly.

  8. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

    A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

  9. Steam generator tube integrity requirements and operating experience in the United States

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    2009-01-01

    Steam generator tube integrity is important to the safe operation of pressurized-water reactors. For ensuring tube integrity, the U.S. Nuclear Regulatory Commission uses a regulatory framework that is largely performance based. This performance-based framework is supplemented with some prescriptive requirements. The framework recognizes that there are three combinations of tube materials and heat treatments currently used in the United States and that the operating experience depends, in part, on the type of material used. This paper summarizes the regulatory framework for ensuring steam generator tube integrity, it highlights the current status of steam generators, and it highlights some of the steam generator issues and challenges that exist in the United States. (author)

  10. Modeling and Simulation of U-tube Steam Generator

    Science.gov (United States)

    Zhang, Mingming; Fu, Zhongguang; Li, Jinyao; Wang, Mingfei

    2018-03-01

    The U-tube natural circulation steam generator was mainly researched with modeling and simulation in this article. The research is based on simuworks system simulation software platform. By analyzing the structural characteristics and the operating principle of U-tube steam generator, there are 14 control volumes in the model, including primary side, secondary side, down channel and steam plenum, etc. The model depends completely on conservation laws, and it is applied to make some simulation tests. The results show that the model is capable of simulating properly the dynamic response of U-tube steam generator.

  11. 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)

  12. Tests and analysis on steam generator tube failure propagation

    International Nuclear Information System (INIS)

    Tanabe, Hiromi

    1990-01-01

    The understanding of leak enlargement and failure propagation behavior is essential to select a design basis leak (DBL) of LMFBR steam generators. Therefore, various series of experiments, such as self-enlargement tests, target wastage tests, failure propagation tests were conducted in a wide range of leak using test facilities of SWAT at PNC/OEC. Especially, in the large leak tests, potential of overheating failure was investigated under a prototypical steam cooling condition inside target tubes. In the small leak, the difference of wastage resistivity was clarified among several tube materials such as 9-chrome steels. In regard to an analytical approach, a computer code LEAP (Leak Enlargement and Propagation) was developed on the basis of all of these experimental results. The code was used to validate the previously selected DBL of the prototype reactor, Monju, steam generator. This approach proved to be successful in spite of somewhat over-conservatism in the analysis. Moreover, LEAP clarified the effectiveness of a rapid steam dump and an enhanced leak detection system. The code improvement toward a realistic analysis is desired, however, to lessen the DBL for a future large plant and then the re-evaluation of the experimental data such as the size of secondary failure is under way. (author). 4 refs, 8 figs, 1 tab

  13. Vibration and wear characteristics of steam generator tubes

    International Nuclear Information System (INIS)

    Choi, Young Hwan

    2003-06-01

    This study investigates the fluid elastic instability characteristics of Steam Generator (SG) U-tubes with defect and the safety assessment of the potential for fretting-wear damages on Steam Generator (SG) U-tubes caused by foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions for determining the fluid elastic instability or fretting-wear parameters such as damping ratio, added mass and flow velocity are obtained from three-dimensional SG flow calculation using the ATHOS3 code. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, the wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the internal pressure on the vibration and fretting-wear characteristics of the tube

  14. Devices for investigation and intervention on steam generators tubes bundles

    International Nuclear Information System (INIS)

    Launay, J.P.; Sort, M.

    1986-01-01

    After a brief recall on the French regulation concerning pressure vessels, the authors describe the experience and the devices used by Framatome for closing, repairing, sleeving and shot peening for steam generators tubes bundles [fr

  15. Overview of steam generator tube degradation and integrity issues

    International Nuclear Information System (INIS)

    Diercks, D.R.; Shack, W.J.; Muscara, J.

    1996-10-01

    The degradation of steam generator tubes in pressurized water nuclear reactors continues to be a serious problem. Primary water stress corrosion cracking is commonly observed at the roll transition zone at U-bends, at tube denting locations, and occasionally in plugs and sleeves. Outer-diameter stress corrosion cracking and intergranular attack commonly occur near the tube support plate crevice, near the tube sheet in crevices or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of circumferential cracking at the RTZ on both the primary and secondary sides. Segmented axial cracking at the tubes support plate crevices is also becoming more common. Despite recent advances in in-service inspection technology, a clear need still exists for quantifying and improving the reliability of in- service inspection methods with respect to the probability of detection of the various types of flaws and their accurate sizing. Improved inspection technology and the increasing occurrence of such degradation modes as circumferential cracking, intergranular attack, and discontinuous axial cracking have led to the formulation of a new performance-based steam generator rule. This new rule would require the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes perform the required safety function over the next operating cycle. The new steam generator rule will also be applied to severe accident conditions to determine the continued serviceability of a steam generator with degraded tubes in the event of a severe accident. Preliminary analyses are being performed for a hypothetical severe accident scenario to determine whether failure will occur first in the steam generator tubes, which would lead to containment bypass, or instead in the hot leg nozzle or surge line, which would not

  16. Nano surface generation of grinding process using carbon nano tubes

    Indian Academy of Sciences (India)

    holes need different processing techniques. Conventional finishing methods used so far become almost impossible or cumbersome. In this paper, a nano material especially multi wall carbon nano tube is used in the machining process like ...

  17. A State of the Art Report on Wear Damage of Steam Generator Tubes

    International Nuclear Information System (INIS)

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

    2004-10-01

    The recent status on wear damage of steam generator tubes caused by flow-induced vibration was investigated, and the criteria for structural integrity evaluation of the wear-damaged tubes were reviewed. It was surveyed how the wear damage of tubes could be affected by main parameters, such as, materials properties and their combination, impact load and vibration amplitude/frequency, contact areas and diametral clearance between the tube and tube support plate, wear test duration, and test temperature. Finally, corrosive wear, which means the combined action of corrosion and wear simultaneously, was also surveyed in this report. There has been only a few works concerned on the wear damage of steam generator tubes in Korea, compared with the leading foreign research institutes. Especially, the experience related to the wear characteristics of Alloy 690, which has become a replacement material for Alloy 600 as steam generator tubes, is far from satisfactory. Systematic studies, therefore, concerned with structural integrity of tubes as well as improvement of were resistance of Alloy 690 in the PWR environment are needed

  18. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gorman, J.A.; Turner, A.P.L.

    1997-01-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses

  19. Steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

    Gorman, J.A.; Harris, J.E.; Lowenstein, D.B.

    1995-07-01

    The objectives of this project were to characterize defect mechanisms which could affect the integrity of steam generator tubes, to review and critique state-of-the-art Canadian and international steam generator tube fitness-for-service criteria and guidelines, and to obtain recommendations for criteria that could be used to assess fitness-for service guidelines for steam generator tubes containing defects in Canadian power plant service. Degradation mechanisms, that could affect CANDU steam generator tubes in Canada, have been characterized. The design standards and safety criteria that apply to steam generator tubing in nuclear power plant service in Canada and in Belgium, France, Japan, Spain, Sweden, and the USA have been reviewed and described. The fitness-for-service guidelines used for a variety of specific defect types in Canada and internationally have been evaluated and described in detail in order to highlight the considerations involved in developing such defect specific guidelines. Existing procedures for defect assessment and disposition have been identified, including inspection and examination practices. The approaches used in Canada and in Belgium, France, Japan, Spain, Sweden, and the USA for fitness-for-service guidelines were compared and contrasted for a variety of defect mechanisms. The strengths and weaknesses of the various approaches have been assessed. The report presents recommendations on approaches that may be adopted in the development of fitness-for-service guidelines for use in the dispositioning of steam generator tubing defects in Canada. (author). 175 refs., 2 tabs., 28 figs

  20. Development of the double-wall-tube steam generator. Evaluation of inner tube leak detection system

    International Nuclear Information System (INIS)

    Teraoku, Takuji; Kisohara, Naoyuki

    1995-01-01

    A double-wall-tube steam generator (DWT-SG) is considered to have possibility of eliminating a secondary heat transport system to realize a reliable and simplified FBR plant. Thus, basic tests for inner/outer tube leak detection and prototypical leak tests by use of the 1MWt DWT-SG model have been performed to evaluate the feasibility of DWT-SG. Their results demonstrated that the inner leak detection system can definitely detect a steam leak from an inner tube flaw. Analyses of the inner tube leak and detection behavior obtained in the 1MWt DWT-SG test enabled to estimate the performance of the inner tube detection system of the commercial DWT-SG system. (author)

  1. Structural and leakage integrity assessment of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Otruba, J.; Keilova, E.; Krhounek, V.; Turek, J.

    1996-01-01

    The leakage and plugging limits were derived for WWER steam generators based on leak and burst tests using tubes with axial part-through and through-wall defects. The following conclusions were arrived at: (i) The permissible primary-to-secondary leak rate with respect to the permissible through-wall defect size of WWER-440 and WWER-1000 steam generator tubes is 8 l/h. (ii) The primary-to-secondary leak rate is reduced by the blocking of the tube cracks by corrosion product particles and other substances. (iii) The rate of crack penetration through the tube wall is higher than the crack widening. (iv) The validity of the criterion of instability for tubes with through-wall cracks was confirmed experimentally. For the WWER-440 and WWER-1000 steam generators, the critical size of axial through-wall cracks, for the threshold primary-to-secondary pressure difference, is 13.8 and 12.0 mm, respectively. (v) The calculated leakage for the rupture of one tube and for the assumed extreme defects is two orders and one order of magnitude, respectively, higher than the proposed primary water leakage limit of 8 l/h. (vi) The experiments gave evidence that the use of the permissible thinning limit of 80% for the heat exchange tube plugging does not bring about uncontrollable leakage or unstable crack growth. This is consistent with experience gained at WWER-440 type nuclear power plants. 4 tabs., 5 figs., 9 refs

  2. Current Status on the Development of a Double Wall Tube Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ho Yun; Choi, Byoung Hae; Kim, Jong Man; Kim, Byung Ho

    2007-12-15

    A fast reactor, which uses sodium as a coolant, has a lot of merits as a next generation nuclear reactor. However, the possibility of a sodium-water reaction occurrence hinders the commercialization of this reactor. As one way to improve the reliability of a steam generator, a double-wall tube steam generator is being developed in GEN-4 program. In this report, the current state of the technical developments for a double-wall tube steam generator are reviewed and a future plan for the development of a double-wall tube steam generator is established. The current focuses of this research are an improvement of the heat transfer capability for a double-wall tube and the development of a proper leak detection method for the failure of a double-wall tube during a reactor operation. The ideal goal is an on-line leak detection of a double wall tube to prevent the sodium-water reaction. However, such a method is not developed as yet. An alternative method is being used to improve the reliability of a steam generator by performing a non-destructive test of a double wall tube during the refueling period of a reactor. In this method a straight double wall tube is employed to perform this test easily, but has a difficulty regarding an absorption of a thermal expansion of the used materials. If an on-line leak detection method is developed, the demerits of a straight double-wall tube are avoided by using a helical type double-wall tube, and the probability of a sodium-water reaction can be reduced to a level less than the design-based accident.

  3. Conceptual design of a bayonet-tube steam generator for the ALFRED lead-cooled reactor

    International Nuclear Information System (INIS)

    Damiani, Lorenzo; Montecucco, Massimo; Pini Prato, Alessandro

    2013-01-01

    Highlights: • Conceptual design of a bayonet-tube steam generator for a lead-cooled reactor demonstrator. • Steady-state simulations effected through RELAP 5. • Performance evaluation of the steam generator for different configurations of the bayonet tubes. - Abstract: The present paper is centred on the design of a bayonet tube steam generator, fundamental part of an innovative lead-cooled fast nuclear reactor (LFR). The construction of the LFR is the main objective of the European project named LEADER, of which Ansaldo Nucleare is an important member. The steam generator described in this paper is expected to be installed in a 300 MW thermal power demonstrator plant, named ALFRED. The investigations carried out in this work through the RELAP 5 code have first faced the sizing of the single bayonet tube and then the design of the whole heat exchanger. The configurations of the four coaxial tubes composing the single bayonet, the length of the bayonets and the materials employed have been investigated; the final heat exchanger configuration provides 510 bayonet tubes of 6 m active length with a thermal insulation between the inner descending tube and the rising annulus, assured by a special extremely insulating paint. The whole steam generator has shown its capability to reach the required exchanged power of 37.5 MW th , providing as output dry superheated steam at the desired temperature of 450 °C

  4. Structural and leakage integrity assessment of WWER steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Splichal, K.; Otruba, J. [Nuclear Research Inst., Rez (Switzerland)

    1997-12-31

    The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction. 10 refs.

  5. Flow induced vibration mock-up test for heat exchanger tubes of PWR steam generator

    International Nuclear Information System (INIS)

    Iwase, T.; Takai, M.; Uwagawa, S.; Nakamura, T.; Hirota, K.; Suzuta, T.

    2000-01-01

    It is one of the most important subjects to estimate the flow-related stability of the heat exchanger tubes. A large scale model steam generator has been developed to verify the stability of the tubes in the Japanese PWR steam generators for the two-phase flow-induced vibration and to accumulate related technical data of thermal-hydraulic and flow-induced vibration of U-bend tube bundle. The model steam generator has 230 U-bend tubes of 46 different radius and 5 columns for each of practical diameter and material, and the anti vibration bars are inserted into each spacing between tube arrays. The freon R123 has been used as the secondary side fluid in stead of water-steam two-phase. In the test, void fraction and interfacial velocities in U-bend and straight tube-bundle are measured with bi-optical probes, and vibration responses of some selected tubes are measured with strain gauges and accelerators. It is verified that the U-bend tubes are stable when they are supported as the design requires under normal and some over power no operating condition. The thermal hydraulic code FIT-III has been well verified with measured thermal and hydraulic data. (author)

  6. Effect of tube plugging in the thermalhydraulic performance of 'U' tube steam generators

    International Nuclear Information System (INIS)

    Braga, C.V.M.; Carajilescov, P.

    1981-05-01

    The thermalhydraulic performance of Angra II steam generator has been simulated using the model developed by Braga, C.V.M., 'Thermohydraulic model for steam generator of PWR power plants', in steady state, with plugging up to 40% of total number of tubes. (E.G.) [pt

  7. WWER steam generator tube structural and leakage integrity

    International Nuclear Information System (INIS)

    Splichal, K.; Krhounek, Vl.; Otruba, J.; Ruscak, M.

    1998-01-01

    The integrity of heat exchange tubes may influence the lifetime of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirements are to assure very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evaluation and heat exchange tubes plugging. The stress corrosion cracking and pitting are the main corrosion damages of WWER heat exchange tubes and are initiated from the outer surface. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through wall cracks, oriented preferentially in the axial direction. The paper presents the leakage and plugging limits for WWER steam generators, which have been determined from leak tests and burst tests. The tubes with axial part-through and through-wall defects have been used. The permissible value of primary to secondary leak rate was evaluated with respect to permissible axial through-wall defect size of WWER 440 and 1000 steam generator tubes. Blocking of the tube cracks by corrosion product particles and other compounds reduces the primary to secondary leak rate. The plugging limits involve the following factors: permissible tube wall thickness which determine further operation of the tubes with defects and assures their integrity under operating conditions and permissible size of a through-wall crack which is sufficiently stable under normal and accident conditions in relation to the critical crack length. For the evaluation of burst test of heat exchange tubes with longitudinal through-wall defects the instability criterion has been used and the dependence of the normalised burst pressure on the normalised length of an axial through-wall defect has been determined. The validity of the criterion of instability for WWER tubes with through

  8. Evaluation of EDTA based chemical formulations for the cleaning of monel-400 tubed steam generators

    International Nuclear Information System (INIS)

    Velmurugan, S.; Rufus, A.L.; Sathyaseelan, V.S.; Kumar, P.S.; Veena, S.N.; Srinivasan, M.P.; Narasimhan, S.V.

    1998-01-01

    The Steam Generator (SG) is an important component in any nuclear power plant which contributes significantly for the over all performance of the reactor. The failure of SG tubes occurs mainly by corrosion under accelerated conditions caused by fouling. There is continuous ingress of the corrosion products and ionic impurities from the condenser and feed train of the secondary heat transfer system. The corrosion products accumulate in the stagnant areas near the tube sheet, over the tube support plates and in the tube to tube support plate crevices. These accumulated deposits help to concentrate the aggressive impurities and induce a variety of corrosion processes affecting the structural materials and finally leading to failure of the SG tube. Scale forming impurities can deposit over the tube surfaces and result in reduction of heat transfer efficiency and over heating of the surfaces. Every effort is being made to control the transport of impurities to the steam generator. Increased blow down, installation of condensate polishers and use of all volatile amines have helped to reduce the corrosion product and ionic impurities input into the steam generators of PHWRs. Despite these efforts, failures of SG tubes in PHWRs have been reported. Hence, attempts are being made to develop chemical formulations to clean the deposits accumulated in the steam generators. The EPRI-SGOG chemical cleaning process has been tried with good success in steam generators of different designs including the steam generators of PHWRs. This paper discusses the work on the evaluation of EDTA based chemical cleaning formulations for monel-400 tubed steam generators of PHWRs. (author)

  9. Steam generator tube integrity program. Phase I report

    International Nuclear Information System (INIS)

    Alzheimer, J.M.; Clark, R.A.; Morris, C.J.; Vagins, M.

    1979-09-01

    The results are presented of the pressure tests performed as part of Phase I of the Steam Generator Tube Integrity (SGTI) program at Battelle Pacific Northwest Laboratory. These tests were performed to establish margin-to-failure predictions for mechanically defected Pressurized Water Reactor (PWR) steam generator tubing under operating and accident conditions. Defect geometries tested were selected because they simulate known or expected defects in PWR steam generators. These defect geometries are Electric Discharge Machining (EDM) slots, elliptical wastage, elliptical wastage plus through-wall slot, uniform thinning, denting, denting plus uniform thinning, and denting plus elliptical wastage. All defects were placed in tubing representative of that currently used in PWR steam generators

  10. An advanced tube wear and fatigue workstation to predict flow induced vibrations of steam generator tubes

    International Nuclear Information System (INIS)

    Gay, N.; Baratte, C.; Flesch, B.

    1997-01-01

    Flow induced tube vibration damage is a major concern for designers and operators of nuclear power plant steam generators (SG). The operating flow-induced vibrational behaviour has to be estimated accurately to allow a precise evaluation of the new safety margins in order to optimize the maintenance policy. For this purpose, an industrial 'Tube Wear and Fatigue Workstation', called 'GEVIBUS Workstation' and based on an advanced methodology for predictive analysis of flow-induced vibration of tube bundles subject to cross-flow has been developed at Electricite de France. The GEVIBUS Workstation is an interactive processor linking modules as: thermalhydraulic computation, parametric finite element builder, interface between finite element model, thermalhydraulic code and vibratory response computations, refining modelling of fluid-elastic and random forces, linear and non-linear dynamic response and the coupled fluid-structure system, evaluation of tube damage due to fatigue and wear, graphical outputs. Two practical applications are also presented in the paper; the first simulation refers to an experimental set-up consisting of a straight tube bundle subject to water cross-flow, while the second one deals with an industrial configuration which has been observed in some operating steam generators i.e., top tube support plate degradation. In the first case the GEVIBUS predictions in terms of tube displacement time histories and phase planes have been found in very good agreement with experiment. In the second application the GEVIBUS computation showed that a tube with localized degradation is much more stable than a tube located in an extended degradation zone. Important conclusions are also drawn concerning maintenance. (author)

  11. Anatomy Education for the YouTube Generation

    Science.gov (United States)

    Barry, Denis S.; Marzouk, Fadi; Chulak-Oglu, Kyrylo; Bennett, Deirdre; Tierney, Paul; O'Keeffe, Gerard W.

    2016-01-01

    Anatomy remains a cornerstone of medical education despite challenges that have seen a significant reduction in contact hours over recent decades; however, the rise of the "YouTube Generation" or "Generation Connected" (Gen C), offers new possibilities for anatomy education. Gen C, which consists of 80% Millennials, actively…

  12. Evaluation of a steam generator tube repair process using an explosive expansion techniuqe at TMI-1

    International Nuclear Information System (INIS)

    Rajan, J.; Shook, T.A.; Leonard, L.

    1983-01-01

    After a planned shutdown of Unit No. 1 at Three Mile Island, cracks were discovered in the primary side of steam generator tubes in the vicinity of the upper surface of the upper tubesheet. The nature of these cracks was later characterized as intergranular stress corrosion. The licensee, General Public Utilities Nuclear (GPUN), proposed to form a new tube-to-tubesheet seal below the cracks using a repair process wherein a detonating cord and polyethylene cartridge assembly inserted into the tube explosively expand the tube against the tubesheet. The explosive expansion process has had numerous applications over the years in the initial fabrication of heat exchanger tube-to-tubesheet assemblies and in repair processes using sleeving. However, this is the first use of this process in a steam generator to expand a previously rolled tube and to form a new seal between it and the tubesheet below a defective region in the tube. The seal obtained between the tube and tubesheet depends on the magnitude of explosive energy released in the detonating process. In this application, it is desired to obtain a mechanical bond rather than a metallurgical welding of the tube and tubesheet. A number of critical variables must be taken into account in order to obtain a successful mechanical seal. These include the explosive power of the detonating cord, the number of expansion shots used, the length of tube which is expanded, cartridge and tube diameters, the diameter of the tubesheet hole, the materials of the tube and tubesheet, and the condition of the surfaces at the time of repair. (orig./GL)

  13. Theoretical-experimental assessment of the variables affecting fretting of Atucha I nuclear power plant utility steam generators tubes

    International Nuclear Information System (INIS)

    Kulichevsky, Raul M.

    1995-01-01

    Fretting wear of Steam Generator tubes caused by flow induced vibrations generates uncertainty on their integrity. The knowledge of the controlling variables of the wear process may give a criterion to evaluate the tubes residual life. Information on vibratory response and dynamic interaction between tubes and their supports are prerequisites for understanding the relationship between fretting wear and tube vibration. Experimental results of the vibratory response of an Atucha-I nuclear power plant type U-tube, the influence of tube/support clearance on this response and a study of tube/support dynamic interaction, which allow the verification of a finite element model of this type of tubes, are presented in this work. Also wear results for the Incoloy 800/DIN 1.4550 austenitic stainless steel pair of materials and a first evaluation of the wear constant of this pair are presented. (author)

  14. Characterization of crevice deposits in Ringhals 3 steam generators tubes

    International Nuclear Information System (INIS)

    Lancha, A.M.; Gomez-Briceno, D.; Garcia-Mazario, M.; Maffiotte, C.; Mcllree, A.

    1998-01-01

    With the aim of characterizing the TSP crevices and tube deposits two tubes, including tube/SP intersections, from a retired Ringhals 3 steam generator have been destructively examined at CIEMAT. The characterization has been performed physically (porosity and microhardness mainly) and chemically (elemental composition and distribution, compounds identification and oxide layers analyses). In addition, a thorough study of the morphology, thickness, coloration, and distribution of the deposits, in both axial and radial directions, in several locations around the tube has been carried out. The results show a heterogeneous distribution of the corrosion products inside the crevice area, relative to thickness, composition and porosity. On the other hand a gap, with values from 4 μm to 140 μm, has been detected between the tube and the deposits in many locations around the tube. This gap might correspond to the shrinkage of the tube during the cooldown. Therefore, during the shutdowns, in presence of air and stagnant water, the existing species in the deposits close to this gap may become oxidized. This fact could be particularly important in the case of copper because it is well known that copper oxides can increase the electrochemical potential up to values in which the SCC susceptibility of alloy 600 appears. (authors)

  15. How to operate safely steam generators with multiple tube through-wall defects

    International Nuclear Information System (INIS)

    Hernalsteen, P.

    1993-01-01

    For a Nuclear Power Plant (NPP) of the Pressurized Water Reactor (PWR) type, the Steam Generator (SG) tube bundle represents the major but also the thinnest part of the primary pressure boundary. To the extent that no tube material has yet been identified to be immune to corrosion, defects may initiate in service and easily propagate through wall. While not a desirable feature, a Through Wall Deep (TWD) defect does not necessarily pose a threat to either the structural integrity or leaktightness and this paper shows how SG can (and indeed, do) operate safely and reliably while having many tubes affected by deep and even TWD defects

  16. In-vessel ITER tubing failure rates for selected materials and coolants

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Cadwallader, L.C. [EG& G Idaho Inc., Idaho Falls, ID (United States)

    1994-03-01

    Several materials have been suggested for fabrication of ITER in-vessel coolant tubing: beryllium, copper, Inconel, niobium, stainless steel, titanium, and vanadium. This report generates failure rates for the materials to identify the best performer from an operational safety and availability perspective. Coolant types considered in this report are helium gas, liquid lithium, liquid sodium, and water. Failure rates for the materials are generated by including the influence of ITER`s operating environment and anticipated tubing failure mechanisms with industrial operating experience failure rates. The analyses define tubing failure mechanisms for ITER as: intergranular attack, flow erosion, helium induced swelling, hydrogen damage, neutron irradiation embrittlement, cyclic fatigue, and thermal cycling. K-factors, multipliers, are developed to model each failure mechanism and are applied to industrial operating experience failure rates to generate tubing failure rates for ITER. The generated failure rates identify the best performer by its expected reliability. With an average leakage failure rate of 3.1e-10(m-hr){sup {minus}1}and an average rupture failure rate of 3.1e-11(m-hr){sup {minus}1}, titanium proved to be the best performer of the tubing materials. The failure rates generated in this report are intended to serve as comparison references for design safety and optimization studies. Actual material testing and analyses are required to validate the failure rates.

  17. Nickel electroplating as a remedy to steam generator tubing PWSCC

    International Nuclear Information System (INIS)

    Michaut, B.; Steltzlen, F.; Sala, B.; Laire, Ch.; Stubbe, J.

    1993-01-01

    Nickel plating appears to be a versatile process, as the application field, even if always used against PWSCC, is different from plant-to-plant. Its usage has been from a purely preventive action on tubes without defects, to a corrective action on through-wall cracked and leaking tubes. As a background for the large scale on-site operations of Doel 2 in 1990 (345 tubes) and Tihange 2 in 1992 (600 tubes), studies on four points are outlined, i.e. corrosion tests, stress measurements, sulfamate bath quality control, and in-service inspection. In conclusion, it appears that the nickel plating technique, following a case-by-case study, can often be a convenient remedy against Alloy 600 stress corrosion problems. New applications, in locations other than the steam generator field are under consideration

  18. Cubic Splines for Trachea and Bronchial Tubes Grid Generation

    Directory of Open Access Journals (Sweden)

    Eliandro Rodrigues Cirilo

    2006-02-01

    Full Text Available Grid generation plays an important role in the development of efficient numerical techniques for solving complex flows. Therefore, the present work develops a method for bidimensional blocks structured grid generation for geometries such as the trachea and bronchial tubes. A set of 55 blocks completes the geometry, whose contours are defined by cubic splines. Besides, this technique build on early ones because of its simplicity and efficiency in terms of very complex geometry grid generation.

  19. Leak location in a double wall tube steam generator. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Greene, D.A.; McMurtrie, K.R.; Gaubatz, D.C.

    1980-01-01

    An experimental and analytical study was made of an acoustic location system for use with double wall tube steam generators. A conceptual design of an acoustic leak location system was developed. This system has the potential for detecting the third fluid gas escaping from the annulus between the double walled tubes into the sodium. The acoustic system uses an array of accelerometers mounted onto the outside of the steam generator vessel. Electronic circuitry selects a number of accelerometers, forms them into an array which focuses within the vessel. The volume of the vessel is sequentially scanned for anomalous acoustic energy.

  20. 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)

  1. Structural integrity assessment of steam generator tubes deteriorated through primary water stress corrosion cracking in transition region of tube expansion

    International Nuclear Information System (INIS)

    Silveira, Helvecio Carlos Klinke da

    2002-01-01

    In PWR plants, steam generator tube degradation has been one of the most important economical concerns, besides causing operational safety problems. In this work, a survey of steam generator tube degradation modes is done. Degradation mechanisms and influence factors are introduced and discussed. The importance of stress corrosion cracking, especially in transition region of tube expansion zone, is underlined. The actual steam generator tube plugging criteria are conservative. Proposed alternative criteria are introduced and discussed. Distinction is done to structural integrity assessment of defective tubes. Real data of tube defect indications of axial cracks in expansion transition zone due to primary water stress corrosion cracking are used in analysis. Results allow discussing application aspects of deterministic and probabilistic criteria on structural integrity assessment of tubes with defect indications. Applied models are specifics, but the application of concept may be extended to other steam generator tube degradation modes. (author)

  2. Development of simulation technique and examination of mechanism for swelling of steam generator tube

    International Nuclear Information System (INIS)

    Kim, Seon Jin; Kim, Ki Nam; Park, Myung Chul; Shin, Gyeong Su; Cho, Jae Hwan

    2010-05-01

    This study was aimed at identifying the mechanism of the swelling through the development of simulation techniques for the swelling of steam generator tubes and correct understanding of swelling so as to evaluate the effect of swelling on soundness of steam generator based on results of the study. Test apparatus designed to simulate the tube swelling was fabricated and through a number of preliminary experiments at different conditions, swelling simulation was successfully completed. A tube swelling phenomenon is caused by a sort of ratcheting which is analyzable and is considered as a sort of fatigue phenomenon in which the stress is accumulated by action/reaction moment resulting from repeated impact of low energy, despite of hoop stress by internal pressure is significantly lower than yield strength of the material which has effect on hoop stress, overcoming the yield strength and causing the tubes to suffer plastic deformation

  3. Recent developments in plugging of steam generator tubes

    International Nuclear Information System (INIS)

    Buhay, S.; Abucay, R.C.

    1995-01-01

    Mechanical Plugging capability has been developed for Bruce Nuclear Generating Station (BNGS) steam generator (SG) tubes and Darlington Nuclear Generating Station (DNGS) SG tubes and tubesheet holes. The plug concept was a modified ABB/Combustion Engineering Inconel 690 plug with a nickel band, rolled into the tube or tubesheet hole from the primary side of the tubesheet. The qualification program included analytical justification of the plug body and experimental testing to verify the leak tightness of the rolled joint under conditions which meet or exceed all service or design requirements. Tools and procedures were developed and tested for manual and remote/robotic installation and removal of the mechanical plugs. Additionally, tools and procedures were developed to plug tubes/tubesheet holes at DNGS in the event the steam generator is recalled to service to act as a heat sink. A crew of Ontario Hydro personnel were trained and qualified for the installation of mechanical plugs for permanent and recall applications. During the DNGS Unit 4 spring 1995 outage, 6 tubes were plugged and the 'Recall Plugging Capability' was deployed and ready for use during a primary side SG tube removal. The mechanical plugs were installed manually with a typical 3 minute/plug in-bowl duration time with an average radiation dose of 12.5 mrem per plug. This compares favourably with manual plug welding during the same outage in the same SG bowl at approximately 15-30 minutes/plug in-bowl duration with an average radiation dose of 117 mrem/plug. (author)

  4. Stress corrosion cracking of a Kori 1 retired steam generator tube

    International Nuclear Information System (INIS)

    Kim, H.P.; Hwang, S.S.; Kim, D. J.; Kim, J. S.; Lim, Y.S.; Joung, M.K.

    2004-01-01

    The present work addressed the evolution trends of the Kori 1 retired steam generators tube degradation such as pitting, primary water stress corrosion cracking (PWSCC), and outer diameter stress corrosion cracking (ODSCC) using the Weibull distribution based on the repair of the tubing and introduced a failure analysis of the pulled out tubes from the Kori 1. A material and condenser replacement in the secondary side and a chemical cleaning of the steam generator changed the Weibull distribution for the pitting. An ingress of sea water through the condenser into the steam generator and an accumulation of chloride in the steam generator induced the pitting. A mechanism of a copper band formation within the corrosion product in a pit is proposed. Pitting seemed to have occurred in an acidic and oxidizing environment between 1978 and early 1990. The Weibull characteristic time and slope for a PWSCC is 25 year and 4.5, respectively. Axial PWSCC was only observed in the R16C35 tube and circumferential PWSCC was only observed in the R11C45 tube at the roll expansion transition. Some tubes that experienced extensive ODSCC rather than PWSCC in the roll transition seemed to be due to the impurities concentrated in the crevice which induce ODSCC, even though the stress in the roll transition of the primary side was higher than that in the secondary side. ODSCC seemed to have occurred in a caustic and slightly oxidizing environment from early 1990 to 1998. (authors)

  5. Cladding material, tube including such cladding material and methods of forming the same

    Science.gov (United States)

    Garnier, John E.; Griffith, George W.

    2016-03-01

    A multi-layered cladding material including a ceramic matrix composite and a metallic material, and a tube formed from the cladding material. The metallic material forms an inner liner of the tube and enables hermetic sealing of thereof. The metallic material at ends of the tube may be exposed and have an increased thickness enabling end cap welding. The metallic material may, optionally, be formed to infiltrate voids in the ceramic matrix composite, the ceramic matrix composite encapsulated by the metallic material. The ceramic matrix composite includes a fiber reinforcement and provides increased mechanical strength, stiffness, thermal shock resistance and high temperature load capacity to the metallic material of the inner liner. The tube may be used as a containment vessel for nuclear fuel used in a nuclear power plant or other reactor. Methods for forming the tube comprising the ceramic matrix composite and the metallic material are also disclosed.

  6. Flow-induced decentering and tube support interaction for steam generator tubes: experiment and physical interpretation

    International Nuclear Information System (INIS)

    Gay, N.; Granger, S.

    1992-11-01

    Maintaining PWR components under reliable operating conditions requires a complex design to prevent various damaging processes including flow-induced vibration and wear mechanisms. To improve the prediction of tube/support interaction and wear in PWR components, EDF has undertaken a comprehensive program oriented to both experimental and computational studies. The present paper illustrates one aspect of this program, related to the determination of contact forces between steam generator tubes and anti-vibration bars (AVBs). The dynamic, nonlinear behavior of a U-tube excited by an air cross-flow is investigated on the CLAVECIN experiment. Interesting and rather unexpected results have been obtained, by varying clearances and flow velocities. The paper is focused on four main points: (i) the originality of the experiment with a force measurement device located in flow; (ii) the importance of a refined data processing for accurately measuring contact forces; (iii) the presentation of the unexpected phenomena revealed in the CLAVECIN experiment, i.e. a flow-induced decentering of the tube which changed the initial tube/AVB clearance, and the consequences on tube/support interaction; (iv) the influence of the actual tube/support clearance in flow on wear mechanisms. The work, presented in the second part of this paper, concentrates exclusively on the physical interpretation of the flow-induced decentering phenomenon and on the theoretical analysis of its consequences on dynamic tube/support interaction. We show that the flow-induced decentering phenomenon can be generated by an unstable quasi-static coupling between the flexible tube and the confined flow, in the vicinity of the support system. This phenomenon is not specific to the CLAVECIN tests and it can be expected every time that a movable obstacle is subjected to confined flow. Moreover, in single-sided impacting conditions, the theoretical analysis confirms the linear relation, found in the CLAVECIN tests

  7. Strategy for assessment of WWER steam generator tube integrity. Report prepared within the framework of the coordinated research project on verification of WWER steam generator tube integrity

    International Nuclear Information System (INIS)

    2007-12-01

    Steam generator heat exchanger tube degradations happen in WWER Nuclear Power Plant (NPP). The situation varies from country to country and from NPP to NPP. More severe degradation is observed in WWER-1000 NPPs than in case of WWER-440s. The reasons for these differences could be, among others, differences in heat exchanger tube material (chemical composition, microstructure, residual stresses), in thermal and mechanical loadings, as well as differences in water chemistry. However, WWER steam generators had not been designed for eddy current testing which is the usual testing method in steam generators of western PWRs. Moreover, their supplier provided neither adequate methodology and criteria nor equipment for planning and implementing In-Service Inspection (ISI). Consequently, WWER steam generator ISI infrastructure was established with delay. Even today, there are still big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment (plugging criteria for defective tubes vary from 40 to 90% wall thickness degradation). Recognizing this situation, the WWER operating countries expressed their need for a joint effort to develop methodology to establish reasonable commonly accepted integrity assessment criteria for the heat exchanger tubes. The IAEA's programme related to steam generator life management is embedded into the systematic activity of its Technical Working Group on Life Management of Nuclear Power Plants (TWG-LMNPP). Under the advice of the TWG-LMNPP, an IAEA coordinated research project (CRP) on Verification of WWER Steam Generator Tube Integrity was launched in 2001. It was completed in 2005. Thirteen organizations involved in in-service inspection of steam generators in WWER operating countries participated: Croatia, Czech Republic, Finland, France, Hungary, Russian Federation, Slovakia, Spain, Ukraine, and the USA. The overall objective was to

  8. Experimental fretting-wear studies of steam generator materials

    International Nuclear Information System (INIS)

    Fisher, N.J.; Chow, A.B.; Weckwerth, M.K.

    1994-01-01

    Flow-induced vibration of steam generator tubes results in fretting-wear damage due to impacting and rubbing of the tubes against their supports. This damage can be predicted by computing tube response to flow-induced excitation forces using analytical techniques, and then relating this response to resultant wear damage using experimentally-derived wear coefficients. Fretting-wear of steam generator materials has been studied experimentally at Chalk River Laboratories for two decades. Tests are conducted in machines that simulate steam generator environmental conditions and tube-to-support dynamic interactions. Different tube and support materials, tube-to-support clearances and tube support geometries have been studied. As well, the effect of environmental conditions, such as temperature, oxygen content, pH and chemistry control additive, have been investigated. Early studies showed that damage was related to contact force as long as other parameters, such as geometry and motion were held constant. Later studies have shown that damage is related to a parameter called work-rate, which combines both contact force and sliding distance. Results of short- and long-term fretting-wear tests for CANDU steam generator materials at realistic environmental conditions are presented. These results demonstrate that work-rate is appropriate correlating parameter for impact-sliding interaction

  9. Implications of Steam Generator Fouling on the Degradation of Material and Thermal Performance

    Science.gov (United States)

    Turner, Carl W.

    Fouling of steam generators has a significant negative impact on the material and thermal performance the steam generators of pressurized water reactors. Corrosion products that originate from various components in the steam cycle of a nuclear power plant get pumped forward with the feed water to steam generators where they deposit on the tube bundle, tube support structure and the tube sheet. Heavy accumulation of deposit within the steam generator has led to some serious operational problems, including loss of thermal performance, under deposit corrosion, steam generator level oscillations, flow accelerated corrosion of carbon steel tube support plates and the failure of steam generator tubes due to high cycle fatigue.

  10. Diagnostic of corrosion defects in steam generator tubes using advanced signal processing from Eddy current testing

    International Nuclear Information System (INIS)

    Formigoni, Andre L.; Lopez, Luiz A.N.M.; Ting, Daniel K.S.

    2009-01-01

    Recently, the Brazilian Angra I PWR nuclear power plant went into a programmed shutdown for substitution of its Steam Generator (SG) which life was shortened due to stress corrosion in its tubes. The total cost of investment were around R$724 million. The signals generated during an Eddy-current Testing (ECT) inspection in SG tubes of nuclear plant allows for the localization and dimensioning of defects in the tubes. The defects related with corrosion generate complex signals that are difficult to analyze and are the most common cause in SG replacement in nuclear power plants around the world. The objective of this paper is the development of a methodology that allows for the characterization of corrosion signals by ECT inspections applied in the heat exchangers tubes of SG of a nuclear power plant. In this present work, the aim is to investigate distributed type defects by inducing controlled corrosion in sample tubes of different materials The ECT signals obtained from these samples tubes with corrosion implanted, will be analyzed using Zetec ECT equipment, the MIZ-17ET and its probes. The data acquisition will use a NI PC A/D CARD 700 card and the LabVIEW program. Subsequently, we will apply mathematical tools for signal processing like time windowed Fast Fourier transforms and Wavelets transforms, in MATLAB platform, which will allow effectiveness to remove the noises and to extract representative characteristics for the defect being analyzed. Previously obtained results as well as the proposal for the future work will be presented. (author)

  11. Life prediction of steam generator tubing due to stress corrosion crack using Monte Carlo Simulation

    International Nuclear Information System (INIS)

    Hu Jun; Liu Fei; Cheng Guangxu; Zhang Zaoxiao

    2011-01-01

    Highlights: → A life prediction model for SG tubing was proposed. → The initial crack length for SCC was determined. → Two failure modes called rupture mode and leak mode were considered. → A probabilistic life prediction code based on Monte Carlo method was developed. - Abstract: The failure of steam generator tubing is one of the main accidents that seriously affects the availability and safety of a nuclear power plant. In order to estimate the probability of the failure, a probabilistic model was established to predict the whole life-span and residual life of steam generator (SG) tubing. The failure investigated was stress corrosion cracking (SCC) after the generation of one through-wall axial crack. Two failure modes called rupture mode and leak mode based on probabilistic fracture mechanics were considered in this proposed model. It took into account the variance in tube geometry and material properties, and the variance in residual stresses and operating conditions, all of which govern the propagations of cracks. The proposed model was numerically calculated by using Monte Carlo Simulation (MCS). The plugging criteria were first verified and then the whole life-span and residual life of the SG tubing were obtained. Finally, important sensitivity analysis was also carried out to identify the most important parameters affecting the life of SG tubing. The results will be useful in developing optimum strategies for life-cycle management of the feedwater system in nuclear power plants.

  12. Development of a computer program to predict structural integrity against fretting wear of steam generator tubes: PIAT (program for integrity assessment of steam generator tubes)

    International Nuclear Information System (INIS)

    Park, Chi-Yong; Ryu, Ki-Wahn; Rhee, Huinam

    2013-01-01

    Highlights: ► We develop a computer code to assess the structural integrity of steam generator tubes. ► Flow-induced vibration of whole steam generator tubes can be analyzed systematically. ► The wear map is obtained to predict the wear depth of whole steam generator tubes. ► The structural integrity of steam generator tubes can be improved significantly. -- Abstract: Flow induced vibration of steam generator tubes potentially causes excessive fretting wear at the supports such as anti-vibration bars and tube support plates. For a reliable design of tubes against the flow-induced vibration related failure, the prediction of vibration and wear of tubes should be performed through complicated steps including the thermal-hydraulic analysis, dynamic modal analysis, evaluation of fluid-elastic instability, prediction of turbulence-induced vibration and wear depth for thousands of tubes. However, entire tubes cannot be evaluated within a limited time of design engineering by the conventional analysis methodology. In this paper, we describe an efficient computer program to assess the structural integrity of steam generator tubes against the flow-induced vibration related failure in a very systematic way. The program contains all the necessary thermal-hydraulic database of typical steam generators. It has a very special function to perform modal analysis for all thousands of tubes of a steam generator much faster than the conventional method. The program also performs fluid-elastic instability analysis and calculates the vibrational response to the turbulent flow excitation, and then can predict the wear depth for all tubes of a steam generator. Finally, we can generate the wear prediction map for whole tubes so that an efficient and practical steam generator maintenance management program is feasible. The utilization of the developed computer program for the design and maintenance of steam generators can significantly increase the structural integrity of steam

  13. Lessons learned from tubes pulled from French steam generators

    International Nuclear Information System (INIS)

    Berge, Ph.; Boursier, J.M.; Dallery, D.; De Keroulas, F.; Rouillon, Y.

    1998-01-01

    Since 1981, the Chinon Hot Laboratory has completed more than 380 metallurgical examinations of pulled French steam generator tubes. Electricite de France decided to perform such investigations from the very outset of the French nuclear program, in order to contribute to nuclear power plant safety. The main reasons for withdrawing tubes are to evaluate the degradation, to validate non destructive examination (NDE) techniques, to gain a better understanding of cracking phenomena, and to ensure that the criteria on which plugging operations are based remain conservative. Considerable experience has been accumulated in the field of primary water stress corrosion cracking (PWSCC), OD (secondary) side corrosion, leak and burst tests, and various tube plugging techniques. This paper focuses on the PWSCC phenomenon and on the secondary side corrosion process, and in particular, attempts to correlate French data from pulled tubes with the results of fundamental R and D studies. Finally, within the framework of the Nuclear Power Plant Safety and Maintenance Policy, all these results are discussed in terms of optimization of the field inspection of tube bundles and plugging criteria. (author)

  14. Steam Generator Tube Integrity Program: Surry Steam Generator Project, Hanford site, Richland, Benton County, Washington: Environmental assessment

    International Nuclear Information System (INIS)

    1980-03-01

    The US Nuclear Regulatory Commission (NRC) has placed a Nuclear Regulatory Research Order with the Richland Operations Office of the US Department of Energy (DOE) for expanded investigations at the DOE Pacific Northwest Laboratory (PNL) related to defective pressurized water reactor (PWR) steam generator tubing. This program, the Steam Generator Tube Integrity (SGTI) program, is sponsored by the Metallurgy and Materials Research Branch of the NRC Division of Reactor Safety Research. This research and testing program includes an additional task requiring extensive investigation of a degraded, out-of-service steam generator from a commercial nuclear power plant. This comprehensive testing program on an out-of-service generator will provide NRC with timely and valuable information related to pressurized water reactor primary system integrity and degradation with time. This report presents the environmental assessment of the removal, transport, and testing of the steam generator along with decontamination/decommissioning plans

  15. Process and device for locating a defective tube, particularly in the tube bundle of a steam generator

    International Nuclear Information System (INIS)

    Denis, Jean.

    1977-01-01

    A process is described for locating a defective tube, particularly in the tube bundle of a steam generator of the reversed U tube kind with the ends connected to a tube plate, marking with the bottom of the generator casing a space separated into two adjacent collectors, respectively for the inlet and outlet of a primary fluid flowing inside the tubes of the bundle, these being externally washed by a secondary vaporizing fluid. In this process a television camera that can be inserted into the casing is used. This process consists in transmitting to a display system outside the generator an image of the tube plate in each collector by means of a directional television camera and then to place over this image a luminous marker to locate the end or the faulty tube [fr

  16. 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

  17. Probe for detection of denting in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gerardin, J.P.; Germain, J.L.; Nio, J.C.

    1994-07-01

    In certain types of PWR steam generator, oxide deposits can lead to embedding, and subsequently to deformation of a tube (the phenomenon of ''denting''). Such embedding changes the vibratory behavior of the tubes and can result in fatigue cracking. This type of cracking can also be worsened in the event of improper assembly of the anti-vibration spacer bars supporting the U-bends. To prevent such incidents and provide for effective preventive condition-directed maintenance of its PWR steam generators, EDF has undertaken the study and development of a probe to detect this type of phenomenon. The studies began in 1990 and led to the building of an initial prototype probe. The principle behind the probe consists in inducing vibration in the U-bend and determining the main resonance modes of the tube. Measurements of frequency and amplitude and calculation of damping enable characterization of the mechanical behavior of the U-bend. The most important parameter is damping, for which the value must be sufficiently high to ensure that the tube is not subjected to major vibratory amplitudes during operation. Numerous tests have been performed with the first prototype version of the probe, on a mock-up in the test area and on one of the demounted steam generators on the Dampierre site. These different tests have enabled validation of the operating principle, fine-tuning the process, pinpointing certain mechanical problems in the probe design, and obtaining the first indications as to the real vibratory behavior of U-bends on a steam generator. On the basis of these preliminary tests, the specifications were drawn up for an industrial version of the probe. Following a call for bids and the choice of a manufacturer, work began on fabrication of a new probe model in 1993. This version was delivered at the end of 1993 and testing began in 1994. (authors). 5 figs., 2 tabs

  18. Steam Generator tube integrity -- US Nuclear Regulatory Commission perspective

    International Nuclear Information System (INIS)

    Murphy, E.L.; Sullivan, E.J.

    1997-01-01

    In the US, the current regulatory framework was developed in the 1970s when general wall thinning was the dominant degradation mechanism; and, as a result of changes in the forms of degradation being observed and improvements in inspection and tube repair technology, the regulatory framework needs to be updated. Operating experience indicates that the current U.S. requirements should be more stringent in some areas, while in other areas they are overly conservative. To date, this situation has been dealt with on a plant-specific basis in the US. However, the NRC staff is now developing a proposed steam generator rule as a generic framework for ensuring that the steam generator tubes are capable of performing their intended safety functions. This paper discusses the current U.S. regulatory framework for assuring steam generator (SG) tube integrity, the need to update this regulatory framework, the objectives of the new proposed rule, the US Nuclear Regulatory Commission (NRC) regulatory guide (RG) that will accompany the rule, how risk considerations affect the development of the new rule, and some outstanding issues relating to the rule that the NRC is still dealing with

  19. Prairie Island Nuclear Generating Plant steam generator owners group II: examination of 3 tubes removed from steam generator No. 12

    International Nuclear Information System (INIS)

    Kuchirka, P.; Madeyski, A.; Pearson, R.

    1986-01-01

    The No. 12 steam generator at Prairie Island Unit 1 has experienced some degradation of the hot leg tubes within the tube sheet region. The history of the degradation is given. A comparison of the rate of degradation of No. 12 steam generator with similar rates at Point Beach and Ginna is presen. The No. 11 steam generator has not experienced similar degradation. To investigate the cause and degree of degradation and to obtain correlation of field eddy current (EC) indications with actual conditions, the three hot leg tube segments were removed from the No. 12 steam generator for laboratory EC testing and destructive examination. Two of the tubes had field EC indications in the tube sheet region and one was apparently free of indications. The results of the testing and examinations are given. These tests showed that the lab EC correlated well with the destructive exam, but the field EC did not. The lab EC detected defects that are >35% thru wall, whereas the field EC detected defects that are >80%. The principle degradation was intergranular stress corrosion cracking in the tube sheet region with some uniform intergranular attack. The clean tube had randomly distributed IGA with a maximum depth of 15%. tube/tube sheet deposit chemical analysis does not support the existence of a caustic environment. The conclusions of this work are given

  20. Damage mechanisms and estimation of the frequency of leaks of steam generator tubes in German PWRs

    International Nuclear Information System (INIS)

    Reck, H.

    1992-01-01

    Operating experience of steam generator tubes in German PWRs has shown that so far there have only relatively few cases of damage been registered. The only steam generators with a high failure rate were exchanged in 1983. The material of the affected tubes was Inconel 600. The types of failure that occurred in the late 70's and early 80's were mainly wastage corrosion, which was thought to be the result of phosphate operating. After optimising the water chemistry and changing to ''high AVT'' operating, the failure rate decreased considerably. In total, about 973 of the 193335 tubes that were in operation were plugged because of wall-thinning or leaks. There have been 6 leaks, with the highest leakage volume being 40 liters per hour. 7 refs., 6 figs., 6 tabs

  1. Tube tightness survey during Phenix steam generator operation

    International Nuclear Information System (INIS)

    Cambillard, E.

    1976-01-01

    Phenix steam generators are once-through vessels with single-wall heat-exchange tubes. This design means that any leakage of water into the sodium must be detected as quickly as possible so that the installation can be shut down before extensive damage occurs. The detection of water leaks in Phenix steam generators is based on measurement of the concentration in the sodium, of hydrogen produced by the sodium-water reaction. Since the various modules--evaporators, superheaters, and reheaters--have no free sodium surfaces, detection of hydrogen in argon is not used in Phenix steam generators. The measurement systems employ a probe made of nickel tubes 0.3 mm thick. Hydrogen in the sodium diffuses into a chamber kept under vacuum by an ion pump. The hydrogen pressure in the chamber is measured by a quadrupole mass spectrometer. The nine measurement systems (three per steam generator) are calibrated by injecting hydrogen into the sodium of the secondary circuits. The data-processing computer calculates the hydrogen concentration in the sodium from the spectrometer signals and the probe temperatures, which are not regulated in Phenix; it generates instructions that enable the operator to act if a leak appears. So far, no leaks have been detected. These systems also make it possible to determine rates of hydrogen diffusion caused by corrosion of the steel walls on the water side

  2. Anisotropic deformation of Zr-2.5Nb pressure tube material at high temperatures

    Science.gov (United States)

    Fong, R. W. L.

    2013-09-01

    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 and

  3. Duplex-tube sodium-indication steam generator

    International Nuclear Information System (INIS)

    Matal, O.; Martoch, J.

    1984-01-01

    The steam generator with duplex tubes and sodium indication is connected to the main sodium input and output via the inlet and outlet chambers and has indication spaces connected to the interspaces of the duplex tubes. The first indication space is linked with the auxiliary inlet pipe to the inlet chamber and the second indication space is connected with the auxiliary pipe to the outlet chamber. Mounted to the auxiliary inlet pipe is at least one closure, i.e., a valve or electromagnetic stop. Mounted on the auxiliary outlet pipe is an indication sensor, e.g., a sodium flow sensor. At least one indication space is provided with an alarm sensor, e.g., a thermocouple, a pressure gauge and one sensor to monitor the hydrogen content of sodium. (J.P.)

  4. Preliminary design: duplex tube low-pressure saturated steam generator for large LMFBR plant. Final report

    International Nuclear Information System (INIS)

    Dawson, B.E.

    1979-10-01

    A preliminary design was completed for a steam generator which is applicable to a four loop LMFBR plant of 1000 MWe gross output employing dry and saturated steam conditions. Two steam generators of 364 MWt thermal capacity would be in each loop. The steam generator is a straight, duplex tube design with a shell bellows for thermal expansion and a single tubesheet with an integral leak detection system. The report outlines a proposed series of development tasks to provide additional design data for the duplex tube and to demonstrate the steam generator design with test modules. The design can be adapted readily as either the evaporator or the superheater for a superheated steam cycle with temperatures up to at least 430 0 C (800 0 F), which is below the creep range for the 2 1/4 Cr-1 Mo material

  5. Analysis of Hydrogen Generation and Accumulation in U-233 Tube Vaults

    International Nuclear Information System (INIS)

    Ally, M.R.; Willis, K.J.

    1999-01-01

    The purpose of the 233 U Safe Storage Program is to enhance the safe storage of 233 U-bearing materials. This report describes the work done at the Oak Ridge National Laboratory's Radiochemical Development Facility (RDF) to address questions related to possible hydrogen generation and accumulation in 233 U tube vaults. The objective of this effort was to verify assumptions in the mathematical model used to estimate the hydrogen content of the gaseous atmosphere that possibly could occur inside the tube vaults in Building 3019 and to evaluate proposed measures for mitigating any hydrogen concerns. A mathematical model was developed using conservative assumptions to evaluate possible hydrogen generation and accumulation in the tube vaults. The model concluded that an equilibrium concentration would be established below the lower flammability limit (LFL) of 4.1% hydrogen. The major assumptions used in the model that were validated are as follows: (1) The shield plug does not form a seal with the tube vault wall, thus allowing the hydrogen gas to diffuse past the shield plug to the upper section of the tube vault. (2) The tube vault end-cap leaks sufficiently to allow air to be drawn into the tube vault by the off-gas system, thereby purging hydrogen from the upper section of the tube vault. (3) Any hydrogen gas generated completely mixes with the other gases present in the lower section of the tube vault and does not stratify beneath the shield plug. (4) The diffusion coefficient determined from the literature for constant diffusion of hydrogen in air is valid. The coefficient is corrected for temperatures from 0 to 25 C. Another assumption used in the model, that hydrogen generated by radiolytic decomposition of hydrogen-bearing materials (e.g., moisture and plastic) leaks from the cans under steady-state condition, as opposed to a sudden release resulting from rupture of the can(s), was beyond the scope of this investigation. Several parameters from the original

  6. Thermal hydraulic characteristics of a double-walled tube advanced nuclear steam generator

    International Nuclear Information System (INIS)

    Cho, S.M.; Seltzer, A.H.

    1989-01-01

    In this paper the thermal hydraulic characteristics of double-walled tube steam generator designed for sodium-cooled nuclear reactors are presented. The double-walled tube construction, along with double-barrier welds for tube-to-tubesheet joints, virtually eliminates the probability of heat transfer tube failure. Considerations are given to the use of the internal core tube, helical vane swirl generator, external protector tube, and variably perforated flow baffles to improve thermal and hydraulic performance of the steam generator. These thermal hydraulic design features with a particular reference to a 432 MW PRISM steam generator are discussed

  7. LMFBR steam generator development: duplex bayonet tube steam generator. Volume II

    International Nuclear Information System (INIS)

    DeFur, D.D.

    1975-04-01

    This report represents the culmination of work performed in fulfillment of ERDA Contract AT(11-1)-2426, Task Agreement 2, in which alternate steam generator designs were developed and studied. The basic bayonet tube generator design previously developed by C-E under AEC Contract AT(11-1)-3031 was expanded by incorporating duplex heat transfer tubes to enhance the unit's overall safety and reliability. The effort consisted of providing and evaluating conceptual designs of the evaporator, superheater and reheater components for a large plant LMFBR steam generator (950 MWt per heat transport loop)

  8. Ultrasonic wall thickness gauging for ferritic steam generator tubing as an in-service inspection tool

    International Nuclear Information System (INIS)

    Haesen, W.M.J.; Tromp, Th.J.

    1980-01-01

    In-service inspection of LWR steam generators is more or less a standard routine operation. The situation can be very different for LMFBRs. For the SNR 300 (Kalkar Power Station) the situation is different because the steam generators have ferritic tubing. The tube walls are comparatively thick, 2 to 4.5 mm. During inservice examinations the steam generators will be drained on both sides, however on the sodium side a sodium film will be present. Furthermore the SNR 300 will have two types of steam generator. A straight tube design and a helical coil design will be used. Both types consist of a evaporator and superheater. The steam generators are of course not radioactive. It is obvious that in this case the eddy current (EC) technique is not an enviable inservice inspection tool. Basically EC is a surface flaw detection technique. Only the saturation magnetisation method will improve the EC technique sufficiently for ferritic material. However the 'in bore examination' with the saturation technique was, in case of the SNR 300 steam generator tubing, considered impossible since the inner diameters are fairly small. Furthermore sodium traces may influence the EC method. Although multifrequency methods can solve this problem, EC is not considered as a useful tool for examining ferritic tubing. Another method is to employ the 'stray flux' method which is under development with the TNO organization in Holland. The EC and stray flux method do have one drawback, these methods do not detect gradual changes in wall thickness. Ultrasonic examinations will be used in the SNR 300 as the main inspection tool for the steam generators. In this paper the reasons why ultrasonic examination was selected are explained. The results of the development work on this subject are discussed

  9. Development of a sealed-accelerator-tube neutron generator

    Science.gov (United States)

    Verbeke; Leung; Vujic

    2000-10-01

    Sealed-accelerator-tube neutron generators are being developed in Lawrence Berkeley National Laboratory (LBNL) for applications ranging from neutron radiography to boron neutron capture therapy and neutron activation analysis. The new generation of high-output neutron generators is based on the D-T fusion reaction, producing 14.1-MeV neutrons. The main components of the neutron tube--the ion source, the accelerator and the target--are all housed in a sealed metal container without external pumping. Thick-target neutron yield computations are performed in this paper to estimate the neutron yield of titanium and scandium targets. With an average deuteron beam current of 1 A and an energy of 120 keV, a time-averaged neutron production of approximately 10(14) n/s can be estimated for a tritiated target, for both pulsed and cw operations. In mixed deuteron/triton beam operation, a beam current of 2 A at 150 keV is required for the same neutron output. Recent experimental results on ion sources and accelerator columns are presented and discussed.

  10. Overview of steam generator tube-inspection technology

    International Nuclear Information System (INIS)

    Obrutsky, L.; Renaud, J.; Lakhan, R.

    2014-01-01

    Degradation of steam generator (SG) tubing due to both mechanical and corrosion modes has resulted in extensive repairs and replacement of SGs around the world. The variety of degradation modes challenges the integrity of SG tubing and, therefore, the stations' reliability. Inspection and monitoring aimed at timely detection and characterization of the degradation is a key element for ensuring tube integrity. Up to the early-70's, the in-service inspection of SG tubing was carried out using single-frequency eddy current testing (ET) bobbin coils, which were adequate for the detection of volumetric degradation. By the mid-80's, additional modes of degradation such as pitting, intergranular attack, and axial and circumferential inside or outside diameter stress corrosion cracking had to be addressed. The need for timely, fast detection and characterization of these diverse modes of degradation motivated the development in the 90's of inspection systems based on advanced probe technology coupled with versatile instruments operated by fast computers and remote communication systems. SG inspection systems have progressed in the new millennium to a much higher level of automation, efficiency and reliability. Also, the role of Non Destructive Evaluation (NDE) has evolved from simple detection tools to diagnostic tools that provide input into integrity assessment decisions, fitness-far-service and operational assessments. This new role was motivated by tighter regulatory requirements to assure the safety of the public and the environment, better SG life management strategies and often self-imposed regulations. It led to the development of advanced probe technologies, more reliable and versatile instruments and robotics, better training and qualification of personnel and better data management and analysis systems. This paper provides a brief historical perspective regarding the evolution of SG inspections and analyzes the motivations behind that evolution. It presents an

  11. New Media: Engaging and Educating the YouTube Generation

    Directory of Open Access Journals (Sweden)

    Anu Vedantham

    2011-12-01

    Full Text Available Today's undergraduates are clearly comfortable as consumers of technology and new media—purchasing ring tones for their cell phones and tunes for their iPods, text-messaging from handheld devices, scanning and tinkering with photos, keeping up with their Facebook friends and watching viral YouTube videos, sometimes all simultaneously. We share examples of classroom assignments integrated with library support services that engage today's undergraduates with academic materials in a variety of course contexts. We discuss how specific arrangements of library learning spaces and the alignment of space and staffing can help undergraduate students succeed with new media projects for class assignments.

  12. Laser cleaning of steam generator tubing based on acoustic emission technology

    International Nuclear Information System (INIS)

    Hou, Su-xia; Luo, Ji-jun; Shen, Tao; Li, Ru-song

    2015-01-01

    As a physical method, laser cleaning technology in equipment maintenance will be a good prospect. The experimental apparatus for laser cleaning of heat tubes in the steam generator was designed according to the results of theoretical analysis. There are two conclusions; one is that laser cleaning technology is attached importance to traditional methods. Which has advantages in saving on much manpower and material resource and it is a good cleaning method for heat tubes. The other is that the acoustic emission signal includes lots of information on the laser cleaning process, which can be used as real-time monitoring in laser cleaning processes. When the laser acts for 350 s, 100 % contaminants of heat tubes is cleaned off, and the sensor only receives weak AE signal at that time.

  13. Laser cleaning of steam generator tubing based on acoustic emission technology

    Energy Technology Data Exchange (ETDEWEB)

    Hou, Su-xia; Luo, Ji-jun; Shen, Tao; Li, Ru-song [Xi' an Hi-Tech Institute, Xi' an (China)

    2015-12-15

    As a physical method, laser cleaning technology in equipment maintenance will be a good prospect. The experimental apparatus for laser cleaning of heat tubes in the steam generator was designed according to the results of theoretical analysis. There are two conclusions; one is that laser cleaning technology is attached importance to traditional methods. Which has advantages in saving on much manpower and material resource and it is a good cleaning method for heat tubes. The other is that the acoustic emission signal includes lots of information on the laser cleaning process, which can be used as real-time monitoring in laser cleaning processes. When the laser acts for 350 s, 100 % contaminants of heat tubes is cleaned off, and the sensor only receives weak AE signal at that time.

  14. The development and application of overheating failure model of FBR steam generator tubes. 2

    International Nuclear Information System (INIS)

    Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi

    2001-11-01

    The JNC technical report 'The Development and Application of Overheating Failure Model of FBR Steam Generator Tubes' summarized the assessment method and its application for the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). This report describes the following items studied after the publication of the above technical report. 1. On the basis of the SWAT-3 experimental data, realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed. New correlations are cosine-shaped temperature profiles with 1,170 C maximum for the 100% and 40% Monju operating conditions, and those with 1,110 C maximum for the 10% condition. 2. For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the assessment. The revised assessment adopts the Katto's correlation for CHF, and the Condie-Bengston IV correlation for post-CHF. 3. Other additional examination for the assessment includes treatments of the whole heating effect (other than the local reaction zone) due to the sodium-water reaction, and the temperature-dependent thermal properties of the heat transfer tube material (2.25Cr-1Mo steel). The revised overheating tube failure assessment method has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% operating conditions when an initial leak is detected by the cover gas pressure detection system. The assessment for the SG system improved for the detection and blowdown systems shows even better safety margins against the overheating tube failure. (author)

  15. Entropy Generation of Shell and Double Concentric Tubes Heat Exchanger

    Directory of Open Access Journals (Sweden)

    basma abbas abdulmajeed

    2016-06-01

    Full Text Available Entropy generation was studied for new type of heat exchanger (shell and double concentric tubes heat exchanger. Parameters of hot oil flow rate, temperature of inlet hot oil and pressure drop were investigated with the concept of entropy generation. The results showed that the value of entropy generation increased with increasing the flow rate of hot oil and when cold water flow rate was doubled from 20 to 40 l/min, these values were larger. On the other hand, entropy generation increased with increasing the hot oil inlet temperature at a certain flow rate of hot oil. Furthermore, at a certain hot oil inlet temperature, the entropy generation increased with the pressure drop at different hot oil inlet flow rates. Finally, in order to keep up with modern technology, infrared thermography camera was used in order to measure the temperatures. The entropy generation was determined with lower values when infrared thermography camera was used to measure the temperatures, compared with the values obtained by using thermocouples.

  16. A compact nanosecond pulse generator for DBD tube characterization

    Science.gov (United States)

    Rai, S. K.; Dhakar, A. K.; Pal, U. N.

    2018-03-01

    High voltage pulses of very short duration and fast rise time are required for generating uniform and diffuse plasma under various operating conditions. Dielectric Barrier Discharge (DBD) has been generated by high voltage pulses of short duration and fast rise time to produce diffuse plasma in the discharge gap. The high voltage pulse power generators have been chosen according to the requirement for the DBD applications. In this paper, a compact solid-state unipolar pulse generator has been constructed for characterization of DBD plasma. This pulsar is designed to provide repetitive pulses of 315 ns pulse width, pulse amplitude up to 5 kV, and frequency variation up to 10 kHz. The amplitude of the output pulse depends on the dc input voltage. The output frequency has been varied by changing the trigger pulse frequency. The pulsar is capable of generating pulses of positive or negative polarity by changing the polarity of pulse transformer's secondary. Uniform and stable homogeneous dielectric barrier discharge plasma has been produced successfully in a xenon DBD tube at 400-mbar pressure using the developed high voltage pulse generator.

  17. Process for installing tubes in a steam generator

    International Nuclear Information System (INIS)

    Boula, G.; George, A.

    1988-01-01

    This process consists essentially to introduce the tubes by planar layers, to place antivibration bars above the layer and tensioning the bars with forces perpendicular to the layer, to check the play between the bars and the tubes and to replace the tubes beyond tolerance by other tubes [fr

  18. Influence of test tube material on subcooled flow boiling critical heat flux in short vertical tube

    International Nuclear Information System (INIS)

    Hata, Koichi; Shiotsu, Masahiro; Noda, Nobuaki

    2007-01-01

    The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u=4.0 to 13.3 m/s), the inlet subcoolings (ΔT sub,in =48.6 to 154.7 K), the inlet pressure (P in =735.2 to 969.0 kPa) and the increasing heat input (Q 0 exp(t/τ), τ=10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tube of inner diameter (d=6 mm), heated length (L=66 mm) and L/d=11 with the inner surface of rough finished (Surface roughness, Ra=3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tube of d=6 mm, L=60 mm and L/d=10 with Ra=0.18 μm and the Platinum (Pt) test tubes of d=3 and 6 mm, L=66.5 and 69.6 mm, and L/d=22.2 and 11.6 respectively with Ra=0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcoolings. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (author)

  19. Influence of Test Tube Material on Subcooled Flow Boiling Critical Heat Flux in Short Vertical Tube

    International Nuclear Information System (INIS)

    Koichi Hata; Masahiro Shiotsu; Nobuaki Noda

    2006-01-01

    The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u = 4.0 to 13.3 m/s), the inlet subcooling (ΔT sub,in = 48.6 to 154.7 K), the inlet pressure (P in = 735.2 to 969.0 kPa) and the increasing heat input (Q 0 exp(t/t), t = 10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tubes of inner diameters (d = 6 mm), heated lengths (L = 66 mm) and L/d = 11 with the inner surface of rough finished (Surface roughness, R a = 3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tubes of d = 6 mm, L = 60 mm and L/d = 10 with R a = 0.18 μm and the Platinum (Pt) test tubes of d = 3 and 6 mm, L = 66.5 and 69.6 mm, and L/d 22.2 and 11.6 respectively with R a = 0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcooling. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (authors)

  20. Development of LABGENE's steam generators tube to tubesheet welding qualification procedure

    Energy Technology Data Exchange (ETDEWEB)

    Pozzo, Renato Del [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)]. E-mail: delpozzo@ctmsp.mar.mil.br; Vieira, Guilherme Godinho [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). Centro Experimental ARAMAR; Patineti Filho, Eloi [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail: epatineti@yahoo.com.br

    2007-07-01

    The welding qualification procedure of LABGENE's Nuclear Electric Generation Laboratory - Steam Generators has special characteristics due to nuclear class 1 requirements, reduced dimensions of the LABGENE's equipment and combination of the materials involved with the tube to tubesheet welding. The welding procedure was performed using an automatic orbital welding machine without material addition. The weld joint was simulated using a sample made of a tube (ext. 12,7 BWG 18 x 90 mm) in SB-163 N08800 material and a plate (48 x 330 x 55 mm) in 20MnMoNi55 material, covered with 8 mm AWS E NiCrFe-3 cladding. For the development of the welding procedure, a lot of welding simulations were performed using machines and special devices designed for the dimensions of the pieces. Procedures related with operating, handling and cleaning conditions, essential to avoid the contamination of the pieces were issued. It was also developed a mixture of gases which contributed for the homogenising of the welding and also to avoid the appearance of cracks and defects on the weld joint. The results obtained with the performed tests fulfilled the requirements of the applied specifications and standards. The welding procedure was developed testing a lot of specimens removed from samples that were representatives of the equipment's tube to tubesheet welding. (author)

  1. Laboratory results of stress corrosion cracking of steam generator tubes in a complex environment - An update

    International Nuclear Information System (INIS)

    Horner, Olivier; Pavageau, Ellen-Mary; Vaillant, Francois; Bouvier, Odile de

    2004-01-01

    Stress corrosion cracking occurs in the flow-restricted areas on the secondary side of steam generator tubes of Pressured Water Reactors (PWR), where water pollutants are likely to concentrate. Chemical analyses carried out during the shutdowns gave some insight into the chemical composition of these areas, which has evolved during these last years (i.e. less sodium as pollutants). It has been modeled in laboratory by tests in two different typical environments: the sodium hydroxide and the sulfate environments. These models satisfactorily describe the secondary side corrosion of steam generator tubes for old plant units. Furthermore, a third typical environment - the complex environment - which corresponds to an All Volatile Treatment (AVT) environment containing alumina, silica, phosphate and acetic acid has been recently studied. This particular environment satisfactorily reproduces the composition of the deposits observed on the surface of the steam generator tubes as well as the degradation of the tubes. A review of the recent laboratory results obtained by considering the complex environment are presented here. Several tests have been carried out in order to study initiation and propagation of secondary side corrosion cracking for some selected materials in such an environment. 600 Thermally Treated (TT) alloy reveals to be less sensitive to secondary side corrosion cracking than 600 Mill Annealed (MA) alloy. Finally, the influence of some related factors like stress, temperature and environmental factors are discussed. (authors)

  2. Steam Generator tube plugging analysis of natural circulation conditions for NPP Krsko

    International Nuclear Information System (INIS)

    Bajs, T.; Mirkovic, D.

    1989-01-01

    Pump trip for NPP Krsko was analysed by deterministic approach. Analyses for 0% and 10% tube plugging were performed using computer code RELAP4/MOD6. The influence of steam generator tube plugging on natural circulation conditions is discussed. (author)

  3. 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)

  4. Laboratory study of corrosion of steam generator tubes: Preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Sala, B.; Organista, M. [Centre Technique Framatome, Le Creusot (France). Dept. Chimie-Corrosion; Henry, K.; Erre, R. [CNRS-CRMD, Orleans (France); Gelpi, A. [Framatome, Paris la Defense (France). Dept. Materiaux et Technologies; Cattant, F.; Dupin, M. [EDF, Avoine (France)

    1995-12-31

    The secondary side intergranular attack (IGA) and intergranular stress corrosion cracking (IGSCC) of steam generator tubes often occurs in crevices where impurities are concentrated, due to local elevated temperatures and restricted water flow. From the analysis of tubes pulled from plants, it is believed that alumino-silicates deposits and/or organic species may play a role in the development of IGA in near neutral environments. New observations suggest that similar environments and similar processes are operative inside the corroded grain boundaries. A former study using autoclave tests was mainly devoted to the formation of alumino-silicate deposits similar to those observed in plants. The present work pursued the study of local environments responsible for IGA/SC. It confirms former results on the catalytic decomposition of organic species into acetates and presents more details on the mechanism of formation of alumino-silicate deposits on alloy 600, particularly on the role of iron and, to a lesser extent, nickel cations. It was showed that, under the alumino-silicate deposits and in the presence of some organic species, a non-protective chromium rich layer may grow instead of the usual protective spinel oxide. The mechanism responsible for the formation of this layer is believed to involve interaction between iron and, to a lesser extent, nickel with silica and/or possible interaction between chromium and acetates. Preliminary capsule tests indicate that these conditions may induce the initiation of IGA.

  5. Inspection of ferromagnetic support structures from within alloy 800 steam generator tubes using pulsed eddy current

    Science.gov (United States)

    Buck, Jeremy Andrew

    Nondestructive testing is a critical aspect of component lifetime management. Nuclear steam generator (SG) tubes are the thinnest barrier between irradiated primary heat transport system and the secondary heat transport system, whose components are not rated for large radiation fields. Conventional eddy current testing (ECT) and ultrasonic testing are currently employed for inspecting SG tubes, with the former doing most inspections due to speed and reliability based on an understanding of how flaws affect coil impedance parameters when conductors are subjected to harmonically induced currents. However, when multiple degradation modes are present simultaneously near ferromagnetic materials, such as tube fretting, support structure corrosion, and magnetite fouling, ECT reliability decreases. Pulsed eddy current (PEC), which induces transient eddy currents via square wave excitation, has been considered in this thesis to simultaneously examine SG tube and support structure conditions. An array probe consisting of a central driver, coaxial with the tube, and an array of 8 sensing coils, was used in this thesis to perform laboratory measurements. The probe was delivered from the inner diameter (ID) of the SG tube, where support hole diameter, tube frets, and 2D off-centering were varied. When considering two variables simultaneously, scores obtained from a modified principal components analysis (MPCA) were sufficient for parameter extraction. In the case of hole ID variation with two dimensional tube off-centering (three parameters), multiple linear regression (MLR) of the MPCA scores provided good estimates of parameters. However, once a fourth variable, outer diameter tube frets, was introduced, MLR proved insufficient. Artificial neural networks (ANNs) were investigated in order to perform pattern recognition on the MPCA scores to simultaneously extract the four measurement parameters from the data. All models throughout this thesis were created and validated using

  6. Producing of Impedance Tube for Measurement of Acoustic Absorption Coefficient of Some Sound Absorber Materials

    Directory of Open Access Journals (Sweden)

    R. Golmohammadi

    2008-04-01

    Full Text Available Introduction & Objective: Noise is one of the most important harmful agents in work environment. In spit of industrial improvements, exposure with over permissible limit of noise is counted as one of the health complication of workers. In Iran, do not exact information of the absorption coefficient of acoustic materials. Iranian manufacturer have not laboratory for measured of sound absorbance of their products, therefore using of sound absorber is limited for noise control in industrial and non industrial constructions. The goal of this study was to design an impedance tube based on pressure method for measurement of the sound absorption coefficient of acoustic materials.Materials & Methods: In this study designing of measuring system and method of calculation of sound absorption based on a available equipment and relatively easy for measurement of the sound absorption coefficient related to ISO10534-1 was performed. Measuring system consist of heavy asbestos tube, a pure tone sound generator, calibrated sound level meter for measuring of some commonly of sound absorber materials was used. Results: In this study sound absorption coefficient of 23 types of available acoustic material in Iran was tested. Reliability of results by three repeat of measurement was tested. Results showed that the standard deviation of sound absorption coefficient of study materials was smaller than .Conclusion: The present study performed a necessary technology of designing and producing of impedance tube for determining of acoustical materials absorption coefficient in Iran.

  7. Implementation status of performance demonstration program for steam generator tubing analysts in Korea

    International Nuclear Information System (INIS)

    Cho, Chan Hee; Lee, Hee Jong; Yoo, Hyun Ju; Nam, Min Woo; Hong, Sung Yull

    2013-01-01

    Some essential components in nuclear power plants are periodically inspected using non destructive examinations, for example ultrasonic, eddy current and radiographic examinations, in order to determine their integrity. These components include nuclear power plant items such as vessels, containments, piping systems, pumps, valves, tubes and core support structure. Steam generator tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non radioactive sides of the nuclear power plant. There is potential that if a tube bursts while a plant is operating, radioactivity from the primary coolant system could escape directly to the atmosphere. Therefore, in service inspections are critical in maintaining steam generator tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due to its high inspection speed and flaw detectability on non magnetic tubes. However, it is not easy to analyze correctly eddy current signals because they are influenced by many factors. Therefore, the performance of eddy current data analysts for steam generator tubing should be demonstrated comprehensively. In Korea, the performance of steam generator tubing analysts has been demonstrated using the Qualified Data Analyst program. This paper describes the performance demonstration program for steam generator tubing analysts and its implementation results in Korea. The pass rate of domestic analysts for this program was 71.4%

  8. Generation Mechanism for Interlinked Flux Tubes on the Magnetopause

    Science.gov (United States)

    Farinas Perez, G.; Cardoso, F. R.; Sibeck, D.; Gonzalez, W. D.; Facskó, G.; Coxon, J. C.; Pembroke, A. D.

    2018-02-01

    We use a global magnetohydrodynamics simulation to analyze transient magnetic reconnection processes at the magnetopause. The solar wind conditions have been kept constant, and an interplanetary magnetic field with large duskward BY and southward BZ components has been imposed. Five flux transfer events (FTEs) with clear bipolar magnetic field signatures have been observed. We observed a peculiar structure defined as interlinked flux tubes (IFTs) in the first and fourth FTE, which had very different generation mechanisms. The first FTE originates as an IFTs and remains with this configuration until its final moment. However, the fourth FTE develops as a classical flux rope but changes its 3-D magnetic configuration to that of IFTs. This work studies the mechanism for generating IFTs. The growth of the resistive tearing instability has been identified as the cause for the first IFTs formation. We believe that the instability has been triggered by the accumulation of interplanetary magnetic field at the subsolar point where the grid resolution is very high. The evidence shows that two new reconnection lines form northward and southward of the subsolar region. The IFTs have been generated with all the classical signatures of a single flux rope. The other IFTs detected in the fourth FTE developed as a result of magnetic reconnection inside its complex and twisted magnetic fields, which leads to a change in the magnetic configuration from a flux rope of twisted magnetic field lines to IFTs.

  9. Effect of shot peening on steam generator tube cracking risks

    International Nuclear Information System (INIS)

    Lannoy, A.

    1993-01-01

    One of the main SG tube degradation modes in stress corrosion cracking on the primary side in the tube/tube plate roll transition zone. With a view to mitigating these cracking risks, on the 900 MWe PWR's plant, the SG inner tube walls were mechanically stress-relieved by shot peening. Between 1985 and 1988, periodic eddy current testing were performed during refuelling outages before and after the shot peening that allows to monitor the number and condition of affected tubes versus time in service. Statistical analysis were performed in order to test and quantify the effects of this treatment. (author). 3 figs., 2 refs

  10. Steam generator tube performance. Experience with water-cooled nuclear power reactors during 1985

    International Nuclear Information System (INIS)

    Tatone, O.S.; Tapping, R.L.

    1988-12-01

    The performance of steam generator tubes at water-cooled reactors during 1985 has been reviewed. Seventy-three of 168 reactors in the survey experienced tube degradation sufficient for the tubes to be plugged. The number of tubes plugged was 6837 or 0.28% of those in service. The leading cause of tube failure was stress corrosion cracking from the primary side. Stress corrosion cracking or intergranular attack from the secondary side and pitting were also major causes of tube failure. Unlike most previous years, fretting was a substantial problem at some reactors. Overall, corrosion continued to account for more than 80% of the defects. 20 refs

  11. Measurement by eddy currents of tube-antivibratory bar gap steam generators of PWR power plants

    International Nuclear Information System (INIS)

    Savin, E.; Bieth, M.; Floze, J.C.

    1990-01-01

    In steam generators tubes are maintained by AVB to limit vibrations amplitude induced by secondary fluid flow. After some years wear sometimes occurs. For gap measurement between tubes and AVB Framatome developed a method based on eddy current and using a probe rotating inside the tube [fr

  12. Experimental and analytical investigation of natural vibration of steam generator heat transfer tubes

    International Nuclear Information System (INIS)

    Han Liangbi; Shi Guolin; Yao Weida; Wang Yufen; Zhang Fugao; Ye Weijuan

    1987-11-01

    Experimental and analytical investigation of model steam generator heat transfer tubes with clearance and elastic supported effect was carried out. The experimental natural frequencies and normal modes of model tubes are found to be in good agreement with the corresponding analytical results. Both analytical and experimental results indicate that the antivibration bars between bends of tubes are effective

  13. Improved eddy-current inspection for steam generator tubing

    International Nuclear Information System (INIS)

    Dodd, C.V.; Pate, J.R.; Allen, J.D. Jr.

    1989-01-01

    Computer programs have been written to allow the analysis of different types of eddy-current probes and their performance under different steam generator test conditions. The probe types include the differential bobbin probe, the absolute bobbin probe, the pancake probe and the reflection probe. The generator test conditions include tube supports, copper deposits, magnetite deposits, denting, wastage, pitting, cracking and IGA. These studies are based mostly on computed values, with the limited number of test specimens available used to verify the computed results. The instrument readings were computed for a complete matrix of the different test conditions, and then the test conditions determined as a function of the readings by a least-squares technique. A comparison was made of the errors in fit and instrument drift for the different probe types. The computations of the change in instrument reading due to the defects have led to an ''inversion'' technique in which the defect properties can be computed from the instrument readings. This has been done both experimentally and analytically for each of these probe types. 3 refs., 13 figs., 1 tab

  14. YouTube Fridays: Engaging the Net Generation in 5 Minutes a Week

    Science.gov (United States)

    Liberatore, Matthew W.

    2010-01-01

    YouTube Fridays is a teaching tool that devotes the first five minutes of class each Friday to a YouTube video related to the course. Students select the videos, which expand the class's educational content in courses such as thermodynamics and material and energy balances. From assessments of two pilot studies using YouTube Fridays in Chemical…

  15. Specification of steam generator, condenser and regenerative heat exchanger materials for nuclear applications

    International Nuclear Information System (INIS)

    Jovasevic, J.V.; Stefanovic, V.M.; Spasic, Z.LJ.

    1977-01-01

    The basic standards specifications of materials for nuclear applications are selected. Seamless Ni-Cr-Fe alloy Tubes (Inconel-600) for steam generators, condensers and other heat exchangers can be employed instead of austenitic stainless steal or copper alloys tubes; supplementary requirements for these materials are given. Specifications of Ni-Cr-Fe alloy plate, sheet and strip for steam generator lower sub-assembly, U-bend seamless copper-alloy tubes for heat exchanger and condensers are also presented. At the end, steam generator channel head material is proposed in the specification for carbon-steel castings suitable for welding

  16. Eddy Current Signature Classification of Steam Generator Tube Defects Using A Learning Vector Quantization Neural Network

    International Nuclear Information System (INIS)

    Garcia, Gabe V.

    2005-01-01

    A major cause of failure in nuclear steam generators is degradation of their tubes. Although seven primary defect categories exist, one of the principal causes of tube failure is intergranular attack/stress corrosion cracking (IGA/SCC). This type of defect usually begins on the secondary side surface of the tubes and propagates both inwards and laterally. In many cases this defect is found at or near the tube support plates

  17. Eddy Current Signature Classification of Steam Generator Tube Defects Using A Learning Vector Quantization Neural Network

    Energy Technology Data Exchange (ETDEWEB)

    Gabe V. Garcia

    2005-01-03

    A major cause of failure in nuclear steam generators is degradation of their tubes. Although seven primary defect categories exist, one of the principal causes of tube failure is intergranular attack/stress corrosion cracking (IGA/SCC). This type of defect usually begins on the secondary side surface of the tubes and propagates both inwards and laterally. In many cases this defect is found at or near the tube support plates.

  18. Fatigue life prediction of autofrettage tubes using actual material behaviour

    International Nuclear Information System (INIS)

    Jahed, Hamid; Farshi, Behrooz; Hosseini, Mohammad

    2006-01-01

    There is a profound Bauschinger effect in the behaviour of high-strength steels used in autofrettaged tubes. This has led to development of methods capable of considering experimentally obtained (actual) material behaviour in residual stress calculations. The extension of these methods to life calculations is presented here. To estimate the life of autofrettaged tubes with a longitudinal surface crack emanating from the bore more accurately, instead of using idealized models, the experimental loading-unloading stress-strain behaviour is employed. The resulting stresses are then used to calculate stress intensity factors by the weight function method as input to fatigue life determination. Fatigue lives obtained using the actual material behaviour are then compared with the results of frequently used ideal models including those considering Bauschinger effect factors and strain hardening in unloading. Using standard fatigue crack growth relationships, life of the vessel is then calculated based on recommended initial and final crack length. It is shown that the life gain due to autofrettage above 70% overstrain is considerable

  19. FFT Analysis of the X-ray Tube Voltage Waveforms of High-Frequency Generators for Radiographic Systems

    International Nuclear Information System (INIS)

    Chida, K.; Saito, H.; Ito, D.; Shimura, H.; Zuguchi, M.; Takai, Y.

    2005-01-01

    Purpose: To present a novel method for analyzing the voltage waveform from high-frequency X-ray generators for radiographic systems. Material and Methods: The output signal of the actual voltage across the tube of a high-frequency generator was measured using the built-in voltage sense taps that are used for voltage regulation feedback in X-ray generators. The output signal was stored in an analyzing recorder, and the waveforms were analyzed using FFT analysis. The FFT analysis of high-frequency generators consisted of obtaining the power spectrum, comparing the major frequency components in the tube voltage waveforms, and examining the intensity of each frequency component. Results: FFT analysis enables an objective comparison of the complex tube voltage waveforms in high-frequency X-ray generators. FFT analysis detected the change in the X-ray tube voltage waveform that occurred when there were problems with the high-frequency generator. Conclusion: High-frequency X-ray generators are becoming the universal choice for radiographic systems. The X-ray tube voltage and its waveform are important features of an X-ray generator, and quality assurance (QA) is important, too. As a tool for engineers involved in the design and development of X-ray generators, we can see that our methods (FFT analysis) might have some value as a means of describing generator performance under varying conditions. Furthermore, since the X-ray tube voltage waveform of a high-frequency generator is complex, FFT analysis may be useful for QA of the waveform

  20. Prognostics for Steam Generator Tube Rupture using Markov Chain model

    International Nuclear Information System (INIS)

    Kim, Gibeom; Heo, Gyunyoung; Kim, Hyeonmin

    2016-01-01

    This paper will describe the prognostics method for evaluating and forecasting the ageing effect and demonstrate the procedure of prognostics for the Steam Generator Tube Rupture (SGTR) accident. Authors will propose the data-driven method so called MCMC (Markov Chain Monte Carlo) which is preferred to the physical-model method in terms of flexibility and availability. Degradation data is represented as growth of burst probability over time. Markov chain model is performed based on transition probability of state. And the state must be discrete variable. Therefore, burst probability that is continuous variable have to be changed into discrete variable to apply Markov chain model to the degradation data. The Markov chain model which is one of prognostics methods was described and the pilot demonstration for a SGTR accident was performed as a case study. The Markov chain model is strong since it is possible to be performed without physical models as long as enough data are available. However, in the case of the discrete Markov chain used in this study, there must be loss of information while the given data is discretized and assigned to the finite number of states. In this process, original information might not be reflected on prediction sufficiently. This should be noted as the limitation of discrete models. Now we will be studying on other prognostics methods such as GPM (General Path Model) which is also data-driven method as well as the particle filer which belongs to physical-model method and conducting comparison analysis

  1. Effect of crevice environment PH on corrosion damage of horizontal steam generator tubes

    International Nuclear Information System (INIS)

    Brozova, A.; Burda, J.; Splichal, K.

    2002-01-01

    In support of a project on lifetime calculation experiments were carried out to evaluate the resistance to environmentally assisted cracking (EAC) of steam generator tubes during operation. Estimations of the incubation period for crack initiation and the threshold K value, K Iscc , and the crack growth rate were made to predict evolution of damage in tube walls. The paper summarizes results of experiments of C ring specimen for the initiation testing and results of SENT (single edge notch tensile) specimen for the crack growth rate (CGR) testing. The specimens were exposed to concentrated environments at elevated temperatures simulating crevice environments in secondary side crevices in horizontal steam generators. The results show that the material of SG tubes is sensitive to transgranular environmentally assisted cracking in the three basic concentrated environments used, alkaline, neutral and acid. The most corrosive medium was the acid environment. The crack initiated practically immediately after acid environment exposure. The initiation process takes a long time in neutral and alkaline environments. The K Iscc values for environmentally assisted crack growth rate in alkaline and neutral concentrated environment were essentially the same. The crack growth rate was slightly higher for the neutral environment than for the alkaline one. Fracture patterns for the both environments were similar. (author)

  2. Dryout in sodium-heated helically-coiled steam generator tubes

    International Nuclear Information System (INIS)

    Tomita, Y.; Kosugi, T.; Kubota, J.; Nakajima, K.; Tsuchiya, T.

    1984-01-01

    Experimental research on the dryout phenomenon in sodium heated, helically coiled steam generator tubes was carried out. The fluctuation of the tube wall temperature caused by dryout was measured with thermocouples installed in the center of the tube wall. Empirical correlations of dryout quality were developed as functions of critical heat flux, water mass velocity and saturation pressure. These correlations confirmed that the design criterion of the MONJU steam generator was reasonable. (author)

  3. PWR steam generators tube integrity: plugging criteria for PWSCC in roll transition zone

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Cruz, Julio R.B.

    1999-01-01

    One of the most important causes for tube plugging in PWR (Pressurized Water Reactor) steam generators is the degradation mechanism called Primary Water Stress Corrosion Cracking (PWSCC) in roll transition zone (RTZ) near the tubesheet, mainly for Alloy 600 tubes. To avoid an excessive tube plugging, alternative criteria have been developed based on an approach that consists in withdrawing from service any tube containing a defect for which there is a high probability of a critical size under accident conditions to be reached during next operation cycle. Predictions of the number of tubes to be plugged can be done aiming at preventive maintenance and tube repair, and even a steam generator replacement, without a large and non-planned plant outage. This work presents important aspects related to tube plugging criteria for PWSCC in RTZ based on the risk of break after a leak detection. Calculations of allowable crack length and allowable leak rate for a particular situation are also shown. (author)

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

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1993-01-01

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

  5. The development and application of overheating failure model of FBR steam generator tubes. 3

    International Nuclear Information System (INIS)

    Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Wada, Yusaku; Miyakawa, Akira; Okabe, Ayao; Nakai, Ryodai; Hiroi, Hiroshi

    2002-03-01

    The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies. Major results obtained in the studies are as follows: 1. To evaluate the structural integrity of tube material, the strength standard for 2. 25Cr-1Mo steel was established taking account of time dependent effect based on the high temperature (700-1200degC) creep data. This standard has been validated with the tube rupture simulation test data. 2. The conditions for overheating by the high temperature reaction were determined by use of the SWAT-3 experimental data. The realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed as the cosine-shaped temperature profile. 3. For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the model. 4. The model has been validated with experimental data obtained by SWAT-3 and LLTR. The results were satisfactory with conservatism. The PFR superheater leak event in 1987 was studied, and the cause of event and the effectiveness of the improvement after the leak event could be identified by the analysis. 5. The model has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% water flow operating conditions when an initial leak is detected by the cover gas pressure detection system. (author)

  6. Improved technique to remove hardened sludge on top of Steam generator tube sheet

    Energy Technology Data Exchange (ETDEWEB)

    Baumgartl, R.

    2015-07-01

    Since many years the top of Steam Generator tube sheet is cleaned by high pressure water jets. In the standard process a multi-nozzle head is manipulated remote controlled inside the No tube lane. The high pressure water jets are directed between the inter-tube aisles. Inner bundle lancing enhanced the efficiency to remove hardened sludge at low flow areas above the tube sheet to a certain extent. For that the nozzle head is fed between the inner tube aisles thus reducing the work distance to a minimum. AREVA GmbH realized a hydraulic driven toothed blade to considerably raise the removal rate of the hardened sludge. (Author)

  7. Scale Thickness Measurement of Steam Generator Tubing Using Eddy Current Signal of Bobbin Coil

    International Nuclear Information System (INIS)

    Kim, Chang Soo; Um, Ki Soo; Kim, Jae Dong

    2012-01-01

    Steam generator is one of the major components of nuclear power plant and steam generator tubes are the pressure boundary between primary and secondary side, which makes them critical for nuclear safety. As the operating time of nuclear power plant increases, not only damage mechanisms but also scaled deposits on steam generator tubes are known to be problematic causing tube support flow hole blockage and heat fouling. The ability to assess the extent and location of scaled deposits on tubes became essential for management and maintenance of steam generator and eddy current bobbin data can be utilized to measure thickness of scale on tubes. In this paper, tube reference standards with various thickness of scaled deposit has been set up to provide information about the overall deposit condition of steam generator tubes, providing essential tool for steam generator management and maintenance to predict and prevent future damages. Also, methodology to automatically measure scale thickness on tubes has been developed and applied to field data to estimate overall scale amount.

  8. Stochastic modeling of inspection uncertainties and applications to pitting flaws in steam generator tubes

    International Nuclear Information System (INIS)

    Mao, D.; Yuan, X.-X.; Pandey, M.D.

    2009-01-01

    Steam generators (SG) are a major pressure retaining component of great safety significance in nuclear power plants. Due to various manufacturing, operation and maintenance activities, as well as material interaction with the surrounding chemical environment, the SG tubes have been subject to a number of degradation modes. Among them, the under-deposit pitting corrosion at outside surfaces of the SG tubes just on top of the tubesheet support plates has had a serious impact on the integrity of the SG tubes. This paper presents an advanced probabilistic model of pitting corrosion characterizing the inherent randomness of the pitting process and measurement uncertainties of the in-service inspection (ISI) data obtained from eddy current (EC) inspections. A Bayesian method based on Markov Chain Monte Carlo (MCMC) simulation is developed for estimating the model parameters. The proposed model is able to predict the actual pit number, the actual pit depth as well as the maximum pit depth, which is the main interest of the pitting corrosion model. (author)

  9. Production and testing of tubes for nuclear boiler steam generators

    International Nuclear Information System (INIS)

    Jacson, M.

    1977-01-01

    Vallourec, second pipe manufacturer in Europe, has developed a workshop for the production of nuclear heat exchanger tubes in its Montbard plant. This workshop, by its special construction, production engineering and handling procedures, has attained nuclear standards and can produce U-bended tubes from diameter 12 to 25 mm with a maximum length of 36 meters. Its annual out-put is 1.500.000 meters. The final dimensions are obtained by a cold rolling procedure, followed by an outside and inside degreasing, a solution annealing in a controlled atmosphere continuous type furnace, a surface grinding and an inside surface conditionning. The non-destructive tests: eddy currents, ultrasonic tests and thickness mesures are recorded on a single tube basis. The curving and packing procedures have been specially developed for this production [fr

  10. A study on extraction of the center point of steam generator tubes

    International Nuclear Information System (INIS)

    Cho, Jai Wan; Kim, Chang Hoi; Choi, Young Soo; Seo, Yong Chil; Kim, Seung Ho

    2002-01-01

    This paper describes extraction procedures for the center coordinates of steam generator tubes of Youngkwang nuclear power plant No. 6 unit. The centering coordinates of tubes are needed for monitoring whether ECT probe is exactly inserted into tube or not. The centering coordinates extraction procedure consists of two steps. The first step is to process the region with high contrast in entire image of steam generator tubes because the tube image tends not to have uniform contrast in entire image, which resulted from poor illuminations because steam generator bowl is sealed. Using the center points extracted in the first step and the geometry of tubes lined up in regular triangle patterns the centering coordinates of the rest region with low contrast are estimated. The straight lines, that is, from center point of a tube to the other center points of neighboring tubes in the horizontal, 60 .deg. C and 120 .deg. C directions are derived using center coordinates extracted only in a high contrast image region. Thus, the intersections of straight lines in horizontal direction and slant lines in regular triangular direction are adopted as the center coordinates of tubes in the rest image region with low contrast. The Chi-square interpolation method is used to determine the line's coefficients. In order to estimate the position and pose of camera assembly camera calibration method is also used. Using tubes geometry that tubes are placed on the tube sheet of steam generator with uniform pitch, 1 ( 25.4mm), in the triangular directions, on behalf of calibration chart, the camera calibration is carried out and the extrinsic parameters of camera assembly is estimated

  11. Fracture mechanics analysis of the steam generator tube after shot peening

    International Nuclear Information System (INIS)

    Shin, Kyu In; Jhung, Myung Jo; Choi, Young Hwan; Park, Jai Hak

    2003-01-01

    One of the main degradation of steam generator tubes is stress corrosion cracking induced by residual stress. The resulting damages can cause tube bursting or leakage of the primary water which contained radioactivity. Primary water stress corrosion crack occurs at the location of tube/tubesheet hard rolled transition zone. In order to investigate the effect of shot peening on stress corrosion cracking, stress intensity factors are calculated for the crack which is located in the induced residual stress field

  12. Automation of inspection methods for eddy current testing of steam generator tubes

    International Nuclear Information System (INIS)

    Meurgey, P.; Baumaire, A.

    1990-01-01

    Inspection of all the tubes of a steam generator when the reactor is stopped is required for some of these exchangers affected by stress corrosion cracking. Characterization of each crack, in each tube is made possible by the development of software for processing the signals from an eddy current probe. The ESTELLE software allows a rapid increase of tested tubes, more than 80,000 in 1989 [fr

  13. Associated-particle sealed-tube neutron probe: Detection of explosives, contraband, and nuclear materials

    International Nuclear Information System (INIS)

    Rhodes, E.; Dickerman, C.E.

    1996-01-01

    Continued research and development of the APSTNG shows the potential for practical field use of this technology for detection of explosives, contraband, and nuclear materials. The APSTNG (associated-particle sealed-tube generator) inspects the item to be examined using penetrating 14-MeV neutrons generated by the deuterium-tritium reaction inside a compact accelerator tube. An alpha detector built into the sealed tube detects the alpha-particle associated with each neutron emitted in a cone encompassing the volume to be inspected. Penetrating high-energy gamma-rays from the resulting neutron reactions identify specific nuclides inside the volume. Flight-times determined from the detection times of gamma-rays and alpha-particles separate the prompt and delayed gamma-ray spectra and allow a coarse 3-D image to be obtained of nuclides identified in the prompt spectrum. The generator and detectors can be on the same side of the inspected object, on opposite sides, or with intermediate orientations. Thus, spaces behind walls and other confined regions can be inspected. Signals from container walls can be discriminated against using the flight-time technique. No collimators or shielding are required, the neutron generator is relatively small, and commercial-grade electronics are employed. The use of 14-MeV neutrons yields a much higher cross-section for detecting nitrogen than that for systems based on thermal-neutron reactions alone, and the broad range of elements with significant 14-MeV neutron cross-sections extends explosives detection to other elements including low-nitrogen compounds, and allows detection of many other substances. Proof-of-concept experiments have been successfully performed for conventional explosives, chemical warfare agents, cocaine, and fissionable materials

  14. Preventive and corrective actions for tube degradation and new steam generator design concept

    International Nuclear Information System (INIS)

    Tsuge, A.; Hirano, H.; Sato, M.; Takamatsu, H.

    2004-01-01

    This paper describes the updated comprehensive overview on, (1) Tube degradation experiences through twenty three years operation of PWR Steam Generators in Japan. (2) Corrective and preventive techniques for tube repair operations, non-destructive examinations, and up-graded water chemistry control. (3) Strategy on the option of Steam Generator replacement. (4) Up-graded design features of Mitsubishi Steam Generator based on the long term operating experiences. (author)

  15. Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1978-02-01

    A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author)

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

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1992-03-01

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

  17. Studies on the permeation of hydrogen through steam generator tubes at high temperatures using an electrochemical method

    International Nuclear Information System (INIS)

    Giraudeau, F.; Yang, L.; Steward, F.R.; DeBouvier, O.

    1998-01-01

    The permeation of hydrogen through steam generator tubes at high temperatures (∼ 300 degrees C) has been studied using an electrochemical technique. With this technique, hydrogen is generated on one side of the tube and monitored on the other side. The time for the hydrogen to reach the other side is used to determine the diffusion coefficient of hydrogen in the tube. Boundary conditions at the entry and exit sides have been investigated separately. Preliminary studies were performed on Stainless Steel 316 and Nickel Alloy 800 to better understand the influence of the solution chemistry on the electrochemical evolution of hydrogen. The surface phenomena effect and the trapping effect are discussed to account for differences observed in the permeation response. The hydrogen permeation through oxides at the exit side has been studied. Two nickel alloys (Alloy 800 and Alloy 600), materials widely used for steam generator tubes, have been investigated. The tubes were prefilmed using two different treatments. The oxides were formed in dry air at high temperatures (300 degrees C to 600 degrees C), or in humid gas at 300 degrees C. The diffusion coefficients at 300 degrees C in Stainless Steel 316 and Alloy 800 were determined to be of the order of 10 -6 - 10 -7 cm 2 /s for the bare metal. This is in agreement with results obtained by gas phase permeation techniques in the literature. (author)

  18. New Media: Engaging and Educating the YouTube Generation

    Science.gov (United States)

    Vedantham, Anu; Hassen, Marjorie

    2011-01-01

    Today's undergraduates are clearly comfortable as consumers of technology and new media--purchasing ring tones for their cell phones and tunes for their iPods, text-messaging from handheld devices, scanning and tinkering with photos, keeping up with their Facebook friends and watching viral YouTube videos, sometimes all simultaneously. We share…

  19. Miniature, low-power X-ray tube using a microchannel electron generator electron source

    Science.gov (United States)

    Elam, Wm. Timothy (Inventor); Kelliher, Warren C. (Inventor); Hershyn, William (Inventor); DeLong, David P. (Inventor)

    2011-01-01

    Embodiments of the invention provide a novel, low-power X-ray tube and X-ray generating system. Embodiments of the invention use a multichannel electron generator as the electron source, thereby increasing reliability and decreasing power consumption of the X-ray tube. Unlike tubes using a conventional filament that must be heated by a current power source, embodiments of the invention require only a voltage power source, use very little current, and have no cooling requirements. The microchannel electron generator comprises one or more microchannel plates (MCPs), Each MCP comprises a honeycomb assembly of a plurality of annular components, which may be stacked to increase electron intensity. The multichannel electron generator used enables directional control of electron flow. In addition, the multichannel electron generator used is more robust than conventional filaments, making the resulting X-ray tube very shock and vibration resistant.

  20. Dedicated new descaling method to characterize corrosion and cation release of SG tubing materials

    International Nuclear Information System (INIS)

    Clauzel, Maryline; Guillodo, Michael; Foucault, Marc; Engler, Nathalie; Chahma, Farah

    2012-09-01

    PWR steam generators (SGs), due to the huge wetted surface, are the main source of corrosion product release in the primary coolant circuit. Corrosion products may be transported throughout the whole circuit, activated in the core, and redeposited all over circuit surfaces, resulting in an increase of activity buildup. Understanding the phenomena leading to corrosion product release from SG tubing materials is of primary importance to minimize the global dose integrated by workers and to optimize the reactor shutdown duration and environment releases. Lab scale testing devices are a way to investigate cation release and propose mitigation measures. The descaling technique is based on the specific dissolution of the oxides making possible, by gravimetry, to directly evaluate the total quantity of corroded metal and the quantity of released elements. This technique allows for a statistical study as several SG coupons are exposed in one single test and is usually well-adapted to tubing materials having high or medium cation release behaviors, but has been proven too less accurate for the most recent manufactured SG tubes having low cation release rates. An optimized descaling technique has been developed to allow for the study of low-releasing SG tubing materials. Several steps of the process have been reconsidered. The electropolishing of the coupon is now performed after a careful determination of the thickness of the perturbed layer on the tube outer and/or inner surface to completely remove it so as to limit as much as possible the release of electro-polished faces which are not matter of the study. The number of coupons exposed in the autoclave has been reduced to avoid any saturation of the water primary chemistry, and two kinds of control coupons have been prepared instead of one in the former descaling method to take into account the uncertainties due to the descaling process as well as the CP possible redeposition on the coupons during exposure. Another

  1. Independent tube verification and dynamic tracking in et inspection of nuclear steam generator

    International Nuclear Information System (INIS)

    Xiongzi, Li; Zhongxue, Gan; Lance, Fitzgibbons

    2001-01-01

    The full text follows. In the examination of pressure boundary tubes in steam generators of commercial pressurized water nuclear power plants (PWR's), it is critical to know exactly which particular tube is being accessed. There are no definitive landmarks or markings on the individual tubes. Today this is done manually, it is tedious, and interrupts the normal inspection work, and is difficult due to the presence of water on the tube surface, plug ends instead of tube openings in the field of view, and varying lighting quality. In order to eliminate the human error and increase the efficiency of operation, there is a need to identify tube position during the inspection process, independent of robot encoder position and motion. A process based on a Cognex MVS-8200 system and its application function package has been developed to independently identify tube locations. ABB Combustion Engineering Nuclear Power's Outage Services group, USPPL in collaboration with ABB Power Plant Laboratories' Advanced Computers and Controls department has developed a new vision-based Independent Tube Verification system (GENESIS-ITVS-TM ). The system employ's a model-based tube-shape detection algorithm and dynamic tracking methodology to detect the true tool position and its offsets from identified tube location. GENESIS-ITVS-TM is an automatic Independent Tube Verification System (ITVS). Independent tube verification is a tube validation technique using computer vision, and not using any robot position parameters. This process independently counts the tubes in the horizontal and vertical axes of the plane of the steam generator tube sheet as the work tool is moved. Thus it knows the true position in the steam generator, given a known starting point. This is analogous to the operator's method of counting tubes for verification, but it is automated. GENESIS-ITVS-TM works independent of the robot position, velocity, or acceleration. The tube position information is solely obtained from

  2. Experimental generation and observation of a super-resolution optical tube

    Directory of Open Access Journals (Sweden)

    Jianghua Xu

    2016-05-01

    Full Text Available We generated a super-resolution optical tube by tightly focusing a binary phase modulated azimuthally polarized laser beam. The binary phase modulation is achieved by a glass substrate with multi-belt concentric ring grooves. We also characterized the 3D beam profile by using a cross-shaped knife-edge fabricated on a silicon photo-detector. The size of the super-resolution dark spot in the tube is 0.32λ, which remains unchanged for ∼4λ within the tube. This optical tube may find applications in super-resolution microscopy, optical trapping and particle acceleration.

  3. Investigation of material efficient fin patterns for cost-effective operation of fin and tube heat exchanger

    DEFF Research Database (Denmark)

    Singh, Shobhana; Sørensen, Kim; Condra, Thomas Joseph

    2017-01-01

    and tube heat exchanger. Computational fluid dynamic models of fin and tube heat exchanger with different fin patterns are developed to investigate the fin pattern behavior on heat transfer and pressure loss performance data. In addition, the numerical results are utilized to analyze the engineering design......Design management of a thermal energy system is a critical part of identifying basic designs that meet large scale user demand under certain operating characteristics. Fin and tube heat exchangers are among the most commonly used thermal energy systems which are generating considerable interest...... scale-up heat exchanger configurations with each fin pattern focusing on the application of chosen fin and tube heat exchanger in marine exhaust gas boiler. The analysis highlights the impact of material efficient fin patterns investigated and predicts that the polynomial and sinusoidal fin patterns...

  4. Thermal-hydraulic tests of steam-generator tube-support-plate crevices. Volume 2. Appendixes I through S. Final report

    International Nuclear Information System (INIS)

    Cassell, D.S.; Vroom, D.W.

    1983-01-01

    A test program was conducted to determine for selected steam generator tube supports the thermal/hydraulic conditions at the inception of dryout as indicated by a tube wall temperature excursion, to determine the pressure drop across the supports, and to obtain photographic documentation of the flow upstream and downstream of the supports. A multi-tube steam generator model was used and testing performed over the range of typcal PWR steam generator operating conditions. These appendices contain information on instrumentation calibration, test model and loop calibration, error analysis, test model thermal-hydraulic analyses, index of lab materials and log sheets, index of two-phase flow still photographs, index of high speed movies and video, test data printouts, test model and loop fabrication drawings, procedure for silver brazing tubewall thermocouples, and procedure for esablishing tube-tube support line contact

  5. Economic evaluation of maintenance strategies for steam generator tubes using probabilistic fracture mechanics and financial method

    International Nuclear Information System (INIS)

    Sagisaka, Mitsuyuki; Isobe, Yoshihiro; Yoshimura, Shinobu; Yagawa, Genki

    2004-01-01

    As an application of probabilistic fracture mechanics (PFM) and a financial method, risk-benefit analyses were performed for the purpose of optimizing maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). Parameters such as in-service inspection (ISI) detection accuracy, ISI interval, sampling inspection, replacement of SGs and stress corrosion cracking (SCC) allowance operation were selected for sensitivity analyses. In the analysis of the operation introducing maintenance criteria, the effect of quantitative accuracy of the inspection was also taken into account. Although the analyses were mainly conducted for SG tubes made of Inconel 600 mill anneal (MA) materials, the analyses were also performed for SCC-resistant materials with making assumptions on their crack initiation probabilities and crack propagation laws. To justify whether or not it is worth while implementing the selected maintenance strategies in terms of an economic point of view, net present value (NPV) was calculated as an index which is one of the most fundamental financial indices for decision-making based on the discounted cash flow (DCF) method. (author)

  6. Simulation of stress corrosion crack growth in steam generator tubes

    International Nuclear Information System (INIS)

    Shin, K. I.; Park, J. H.; Joo, J. W.; Shin, E. S.; Kim, H. D.; Chung, H. S.

    2000-01-01

    Stress corrosion crack growth is simulated after assuming a small axial surface crack inside a S/G tube. Internal pressure and residual stresses are considered as applied forces. Stress intensity factors along crack front, variation of crack shape and crack growth rate are obtained and discussed. It is noticed that the aspect ratio of the crack is not depend on the initial crack shape but depend on the residual stress distribution

  7. Packaging material and flexible medical tubing containing thermally exfoliated graphite oxide

    Science.gov (United States)

    Prud'homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor)

    2011-01-01

    A packaging material or flexible medical tubing containing a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 m.sup.2/g to 2600 m.sup.2/g.

  8. Efficiency of defect specific maintenance od steam generator tubes: the case of ODSCC

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.

    1996-01-01

    The outside diameter stress corrosion cracking at tube support plates became the dominating ageing mechanism in steam generators tubes made of Inconel 600. A variety of maintenance approaches were developed and implemented worldwide to deal with this mechanism. Despite different philosophical and physical backgrounds implemented, all of the applied approaches satisfy the relevant regulatory requirements. For our purpose, the maintenance approach consist of: (1) inspection of tubes, (2) accepting or rejecting the defective tube and (3) plugging of rejected tubes. The problem of selecting an optimal maintenance approach is raised in the paper. Consequently, a method comparing the efficiency of applicable maintenance approaches is proposed. The efficiency is defined by three parameters: (a) number of plugged tubes, (b) probability of steam generator tube rupture and (c) predicted accidental leak rates through the defects. An original probabilistic model is proposed to quantify the probability of tube rupture, while procedures available in literature were used to define the accidental leak rates. The numerical example considers the data from Krsko NPP (Westinghouse 632 MWe). The maintenance approaches analyzed include: (i) no repair at all, (ii) traditional defect depth (40%) based maintenance, (iii) alternate plugging criterion (bobbin coil voltage as defined by EPRI and U.S. NRC) and (iv) combined traditional and alternate approach. Advantages of the defect specific approaches (iii) and (iv) over the traditional one (defect depth) are clearly shown. A brief discussion on the optimization of safe life of steam generator is given. (author)

  9. Laminar fluid flow and heat transfer in a fin-tube heat exchanger with vortex generators

    Energy Technology Data Exchange (ETDEWEB)

    Yanagihara, J.I.; Rodriques, R. Jr. [Polytechnic School of Univ. of Sao Paolo, Sao Paolo (Brazil). Dept. of Mechanical Engineering

    1996-12-31

    Development of heat transfer enhancement techniques for fin-tube heat exchangers has great importance in industry. In recent years, heat transfer augmentation by vortex generators has been considered for use in plate fin-tube heat exchangers. The present work describes a numerical investigation about the influence of delta winglet pairs of vortex generators on the flow structure and heat transfer of a plate fin-tube channel. The Navier-Stokes and Energy equations are solved by the finite volume method using a boundary-fitted coordinate system. The influence of vortex generators parameters such as position, angle of attack and aspect ratio were investigated. Local and global influences of vortex generators in heat transfer and flow losses were analyzed by comparison with a model using smooth fin. The results indicate great advantages of this type of geometry for application in plate fin-tube heat exchangers, in terms of large heat transfer enhancement and small pressure loss penalty. (author)

  10. New Generation of LTCC Materials

    OpenAIRE

    Valant, Matjaz; Suvorov, Danilo

    2004-01-01

    To reduce the complexity of LTCC systems and so accelerate the development of LTCC tapes with new functionalities it is necessary to reduce the number of phases within a particular tape. This can best be done by using glass-free single-phase ceramic systems. Such a material system consisting of low- and high-permittivity LTCC materials was developed based on Bi eulytite (permittivity; k’=16) and sillenite (k’=40) compounds and the δ-Bi2O3 solid solution with Nb2O5 (k’=90). The Ge and Si analo...

  11. Studies of the steam generator degraded tubes behavior on BRUTUS test loop

    Energy Technology Data Exchange (ETDEWEB)

    Chedeau, C.; Rassineux, B. [EDF/DER/MTC, Moret Sur Loing (France); Flesch, B. [EDF/EPN/DMAINT, Paris (France)] [and others

    1997-04-01

    Studies for the evaluation of steam generator tube bundle cracks in PWR power plants are described. Global tests of crack leak rates and numerical calculations of crack opening area are discussed in some detail. A brief overview of thermohydraulic studies and the development of a mechanical probabilistic design code is also given. The COMPROMIS computer code was used in the studies to quantify the influence of in-service inspections and maintenance work on the risk of a steam generator tube rupture.

  12. Steam generator tube integrity program: Annual report, August 1995--September 1996. Volume 2

    International Nuclear Information System (INIS)

    Diercks, D.R.; Bakhtiari, S.; Kasza, K.E.; Kupperman, D.S.; Majumdar, S.; Park, J.Y.; Shack, W.J.

    1998-02-01

    This report summarizes work performed by Argonne National Laboratory on the Steam Generator Tube Integrity Program from the inception of the program in August 1995 through September 1996. The program is divided into five tasks: (1) assessment of inspection reliability, (2) research on ISI (inservice-inspection) technology, (3) research on degradation modes and integrity, (4) tube removals from steam generators, and (5) program management. Under Task 1, progress is reported on the preparation of facilities and evaluation of nondestructive evaluation techniques for inspecting a mock-up steam generator for round-robin testing, the development of better ways to correlate failure pressure and leak rate with eddy current (EC) signals, the inspection of sleeved tubes, workshop and training activities, and the evaluation of emerging NDE technology. Results are reported in Task 2 on closed-form solutions and finite-element electromagnetic modeling of EC probe responses for various probe designs and flaw characteristics. In Task 3, facilities are being designed and built for the production of cracked tubes under aggressive and near-prototypical conditions and for the testing of flawed and unflawed tubes under normal operating, accident, and severe-accident conditions. Crack behavior and stability are also being modeled to provide guidance for test facility design, develop an improved understanding of the expected rupture behavior of tubes with circumferential cracks, and predict the behavior of flawed and unflawed tubes under severe accident conditions. Task 4 is concerned with the acquisition of tubes and tube sections from retired steam generators for use in the other research tasks. Progress on the acquisition of tubes from the Salem and McGuire 1 nuclear plants is reported

  13. The relative impact of sizing errors on steam generator tube failure probability

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.

    1998-01-01

    The Outside Diameter Stress Corrosion Cracking (ODSCC) at tube support plates is currently the major degradation mechanism affecting the steam generator tubes made of Inconel 600. This caused development and licensing of degradation specific maintenance approaches, which addressed two main failure modes of the degraded piping: tube rupture; and excessive leakage through degraded tubes. A methodology aiming at assessing the efficiency of a given set of possible maintenance approaches has already been proposed by the authors. It pointed out better performance of the degradation specific over generic approaches in (1) lower probability of single and multiple steam generator tube rupture (SGTR), (2) lower estimated accidental leak rates and (3) less tubes plugged. A sensitivity analysis was also performed pointing out the relative contributions of uncertain input parameters to the tube rupture probabilities. The dominant contribution was assigned to the uncertainties inherent to the regression models used to correlate the defect size and tube burst pressure. The uncertainties, which can be estimated from the in-service inspections, are further analysed in this paper. The defect growth was found to have significant and to some extent unrealistic impact on the probability of single tube rupture. Since the defect growth estimates were based on the past inspection records they strongly depend on the sizing errors. Therefore, an attempt was made to filter out the sizing errors and to arrive at more realistic estimates of the defect growth. The impact of different assumptions regarding sizing errors on the tube rupture probability was studied using a realistic numerical example. The data used is obtained from a series of inspection results from Krsko NPP with 2 Westinghouse D-4 steam generators. The results obtained are considered useful in safety assessment and maintenance of affected steam generators. (author)

  14. Steam generator tube rupture in an experimental facility scaled from a pressurized water reactor

    International Nuclear Information System (INIS)

    Loomis, G.G.

    1984-09-01

    Results from an experimental investigation of steam generator tube rupture in the Semiscale Mod-2B system are presented. From the experimental results, the characteristic system response signature for a wide range of number of tubes ruptured has been described. The tube rupture was assumed to occur during normal full power operation (15.6 MPa system pressure, 37 0 K core differential temperature). In addition, recovery scenarios involving operator actions were examined. The recovery scenarios included use of pressurizer auxiliary spray and internal heaters, steam generator feed and steam, primary feed and bleed, and main cooling pump operation. Recovery scenarios suggested by typical US pressurized water reactor emergency operating procedures were followed

  15. Materials for Reformer Furnace Tubes. History of evolution

    OpenAIRE

    M. Garbiak; W. Jasiński; B. Piekarski

    2011-01-01

    The paper discusses progress that has been made over the past sixty years in increasing the service life of centrifugally cast, creepresistant tubes operating in reformer furnaces. Attention was mainly focused on the principles of selection of the chemical composition of castings to improve their creep behaviour. The reasons accounting for withdrawal of tubes from service were indicated. Examples of chemical composition and mechanical properties obtained in creep-resistant Ni-Cr cast steel us...

  16. Leak behavior of steam generator tube-to-tubesheet joints under creep condition: Experimental study

    International Nuclear Information System (INIS)

    Bahn, Chi Bum; Majumdar, Saurin; Kasza, Ken E.; Shack, William J.

    2013-01-01

    To address concerns regarding excessive leakage from throughwall cracks in steam generator tube-to-tubesheet joints under severe accident conditions, leak rate testing was conducted using tube-to-collar joint specimens. The tube interior and the interface between tube and collar (crevice) were pressurized independently using nitrogen gas. The leak rate through the crevice was almost zero when the specimens were pressurized at ∼500 °C; this low leak rate is attributed to thermal mismatch effects preventing much leakage. The near zero leak rate was maintained until the onset of large leakage at higher temperatures. The leak rate behavior after the onset of the large leakage was not much affected by the crevice length or heat-to-heat variation of Alloy 600 tubes. This suggests that once the crevice gap opens, the creep rate of the low alloy steel collar becomes dominant. Specimens with different tube diameters behaved essentially the same way. To simulate a flawed steam generator tube in the tubesheet, the crevice region was pressurized through a hole in the tube. This simulation resulted in essentially the same behavior as those specimens whose tubes and crevices were pressurized independently. Oxidation of low alloy steel collars in air tests can increase the flow resistance, and thus tests using nitrogen gas would provide more conservative leak rate data. Highlights: ► Leak rates were measured by using tube-to-collar joint specimens under creep condition. ► Leak rate through the joint interface was almost zero at ∼500 °C due to thermal mismatch. ► The near zero leak rate was maintained until the onset of large leakage at ∼680 °C. ► The leak behavior after the onset of the large leakage was not affected by hydraulic expansion length or tube heats.

  17. Evaluation and field validation of Eddy-Current array probes for steam generator tube inspection

    International Nuclear Information System (INIS)

    Dodd, C.V.; Pate, J.R.

    1996-07-01

    The objective of the Improved Eddy-Current ISI for Steam Generator Tubing program is to upgrade and validate eddy-current inspections, including probes, instrumentation, and data processing techniques for inservice inspection of new, used, and repaired steam generator tubes; to improve defect detection, classification, and characterization as affected by diameter and thickness variations, denting, probe wobble, tube sheet, tube supports, copper and sludge deposits, even when defect types and other variables occur in combination; to transfer this advanced technology to NRC's mobile NDE laboratory and staff. This report describes the design of specialized high-speed 16-coil eddy-current array probes. Both pancake and reflection coils are considered. Test results from inspections using the probes in working steam generators are given. Computer programs developed for probe calculations are also supplied

  18. Thermal-hydraulic phenomena during reflux condensation cooling in steam generator tubes

    International Nuclear Information System (INIS)

    Hae, Yong Jeong; Bum, Nyun Kim; Kwangho, Lee

    1998-01-01

    The transitions of cooling mechanism in steam generator tubes during reflux condensation are studied. It is found that the transitions are closely related to the occurrence of flooding or counter-current flow limitation phenomena in steam generator tubes. As shown in the previous studies of other researchers, the transition from filmwise reflux condensation into total reflux condensation occurs when the flooding criterion suggested by Wallis is met. In this study, it is suggested that the transition from total reflux condensation to complete carry-over occurs depending on the tube height and cooling conditions. It is also shown that the flooding at SG tubes occurs before the flooding at hot leg when a reflux condensation mode is existing in steam generator. Though the thermal-hydraulic conditions during reflux condensation after a small-break loss-of-coolant accident have enough margin to the transition into carry-over, considerations for the prevention of primary coolant relocation should be provided

  19. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    Langford, O.M.; Peelman, H.E.

    1980-01-01

    A gas filled neutron tube in a nuclear well logging tool has a target an ion source voltage and a replenisher connected to ground. A negative high voltage is applied to the target by a power supply also providing a target current corresponding to the neutron output of the neutron generator tube. A constant current source provides a constant current. A network receiving the target current and the constant current provides a portion of the constant current as a replenisher current which is applied to the replenisher in a neutron generating tube. The network controls the magnitude of the replenisher current in accordance with the target current so as to control the neutron output of the neutron generating tube. (auth)

  20. Nano surface generation of grinding process using carbon nano tubes

    Indian Academy of Sciences (India)

    Nano surface finish has become an important parameter in the semiconductor, optical, electrical and mechanical industries. The materials used in these industries are classified as difficult to machine materials such as ceramics, glasses and silicon wafers. Machining of these materials up to nano accuracy is a great ...

  1. Electron beam generation and structure of defects in carbon and boron nitride nano-tubes

    Energy Technology Data Exchange (ETDEWEB)

    Zobelli, A

    2007-10-15

    The nature and role of defects is of primary importance to understand the physical properties of C and BN (boron nitride) single walled nano-tubes (SWNTs). Transmission electron microscopy (TEM) is a well known powerful tool to study the structure of defects in materials. However, in the case of SWNTs, the electron irradiation of the TEM may knock out atoms. This effect may alter the native structure of the tube, and has also been proposed as a potential tool for nano-engineering of nano-tubular structures. Here we develop a theoretical description of the irradiation mechanism. First, the anisotropy of the emission energy threshold is obtained via density functional based calculations. Then, we numerically derive the total Mott cross section for different emission sites of carbon and boron nitride nano-tubes with different chiralities. Using a dedicated STEM (Scanning Transmission Electron Microscope) microscope with experimental conditions optimised on the basis of derived cross-sections, we are able to control the generation of defects in nano-tubular systems. Either point or line defects can be obtained with a spatial resolution of a few nanometers. The structure, energetics and electronics of point and line defects in BN systems have been investigated. Stability of mono- and di- vacancy defects in hexagonal boron nitride layers is investigated, and their activation energies and reaction paths for diffusion have been derived using the nudged elastic band method (NEB) combined with density functional based techniques. We demonstrate that the appearance of extended linear defects under electron irradiation is more favorable than a random distribution of point defects and this is due to the existence of preferential sites for atom emission in the presence of pre-existing defects, rather than thermal vacancy nucleation and migration. (author)

  2. Electron beam generation and structure of defects in carbon and boron nitride nano-tubes

    International Nuclear Information System (INIS)

    Zobelli, A.

    2007-10-01

    The nature and role of defects is of primary importance to understand the physical properties of C and BN (boron nitride) single walled nano-tubes (SWNTs). Transmission electron microscopy (TEM) is a well known powerful tool to study the structure of defects in materials. However, in the case of SWNTs, the electron irradiation of the TEM may knock out atoms. This effect may alter the native structure of the tube, and has also been proposed as a potential tool for nano-engineering of nano-tubular structures. Here we develop a theoretical description of the irradiation mechanism. First, the anisotropy of the emission energy threshold is obtained via density functional based calculations. Then, we numerically derive the total Mott cross section for different emission sites of carbon and boron nitride nano-tubes with different chiralities. Using a dedicated STEM (Scanning Transmission Electron Microscope) microscope with experimental conditions optimised on the basis of derived cross-sections, we are able to control the generation of defects in nano-tubular systems. Either point or line defects can be obtained with a spatial resolution of a few nanometers. The structure, energetics and electronics of point and line defects in BN systems have been investigated. Stability of mono- and di- vacancy defects in hexagonal boron nitride layers is investigated, and their activation energies and reaction paths for diffusion have been derived using the nudged elastic band method (NEB) combined with density functional based techniques. We demonstrate that the appearance of extended linear defects under electron irradiation is more favorable than a random distribution of point defects and this is due to the existence of preferential sites for atom emission in the presence of pre-existing defects, rather than thermal vacancy nucleation and migration. (author)

  3. Analysis of potential for jet-impingement erosion from leaking steam generator tubes during severe accidents

    International Nuclear Information System (INIS)

    Majumdar, S.; Diercks, D. R.; Shack, W. J.

    2002-01-01

    This report summarizes analytical evaluation of crack-opening areas and leak rates of superheated steam through flaws in steam generator tubes and erosion of neighboring tubes due to jet impingement of superheated steam with entrained particles from core debris created during severe accidents. An analytical model for calculating crack-opening area as a function of time and temperature was validated with tests on tubes with machined flaws. A three-dimensional computational fluid dynamics code was used to calculate the jet velocity impinging on neighboring tubes as a function of tube spacing and crack-opening area. Erosion tests were conducted in a high-temperature, high-velocity erosion rig at the University of Cincinnati, using micrometer-sized nickel particles mixed in with high-temperature gas from a burner. The erosion results, together with analytical models, were used to estimate the erosive effects of superheated steam with entrained aerosols from the core during severe accidents

  4. New Generation of LTCC Materials

    Directory of Open Access Journals (Sweden)

    Valant, Matjaz

    2004-06-01

    Full Text Available To reduce the complexity of LTCC systems and so accelerate the development of LTCC tapes with new functionalities it is necessary to reduce the number of phases within a particular tape. This can best be done by using glass-free single-phase ceramic systems. Such a material system consisting of low- and high-permittivity LTCC materials was developed based on Bi eulytite (permittivity; k’=16 and sillenite (k’=40 compounds and the δ-Bi2O3 solid solution with Nb2O5 (k’=90. The Ge and Si analogues of the sillenites and eulytites, and the 0.75Bi2O3⋅0.25Nb2O5 solid solution meet the main requirements for LTCC with respect to their sintering behavior (Ts=850-900oC, their mutual chemical compatibility, their compatibility with a silver electrode as well as their dielectric properties.

    Para reducer la complejidad de los sistemas LTCC y así acelerar el desarrollo de láminas de LTCC con nuevas funcionalidades es necesario reducir el número de fases dentro de una determinada lámina. La mejor manera de hacer esto es usar sistemas cerámicos monofásicos libres de fase vítrea. Dicho sistema que consiste en materiales LTCC de baja- y alta-permitividad se ha desarrollado en base a compuestos de Bieulitita (permitividad; k’=16 y silenita (k’=40 y la solución sólida de δ-Bi2O3 con Nb2O5 (k’=90. Los análogos de Ge y Si de las silenitas y eulititas, y la solución sólida 0.75Bi2O3⋅0.25Nb2O5 cumplen los principales requerimientos de los LTCC respecto a su comportamiento de sinterización (Ts=850-900oC, su compatibilidad química mutua y su compatibilidad con electrodos de plata, así como en lo concerniente a sus propiedades dieléctricas.

  5. Steam generator secondary pH during a steam generator tube rupture

    International Nuclear Information System (INIS)

    Adams, J.P.; Peterson, E.S.

    1991-12-01

    The Nuclear Regulatory Commission requires utilizes to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the partitioning of iodine between liquid and vapor in steam generator secondary. Experimental studies have determined that the partitioning of iodine in water is very sensitive to the pH. Based on this experimental evidence, the NRC requested the INEL to perform an analytical assessment of secondary coolant system (SCS) pH during an SGTR. Design basis thermal and hydraulic calculations were used together with industry standard chemistry guidelines to determine the SCS chemical concentrations during an SGTR. These were used as input to the Facility for Analysis of Chemical Thermodynamics computer system to calculate the equilibrium pH in the SCS at various discrete time during an SGTR. The results of this analysis indicate that the SCS pH decreases from the initial value of 8.8 to approximately 6.5 by the end of the transient, independent of PWR design

  6. Measurement of Work Generation and Improvement in Performance of a Pulse Tube Engine

    Science.gov (United States)

    Hamaguchi, Kazuhiro; Futagi, Hiroaki; Yazaki, Taichi; Hiratsuka, Yoshikatsu

    Apart from double acting type engines, Stirling engines have either 2 pistons in 2 cylinders or 2 pistons in a single cylinder. Typically, the heater, regenerator and cooler are installed between the 2 pistons. The pulse tube engine, on the other hand, consists of a single piston in a single cylinder, a pulse tube, a heater, a regenerator, a cooler and a second cooler. For this paper, a simple prototype engine that uses air at normal atmospheric pressure as the working gas was fabricated. The oscillating velocity of the working gas in the pulse tube was measured using LDV, and the work flow emitting out of the pulse tube was observed. In addition, the effect of inserting heat storage material in the pulse tube on shaft power and indicated power was examined experimentally. A dramatic increase in the shaft power was achieved.

  7. Efek Perbedaan Jumlah dan Material Tube pada Distribusi Temperatur Tube Heat Exchanger dalam Kompor (Studi Kasus Di Industri Tempe Kecamatan Tenggilis Mejoyo Surabaya

    Directory of Open Access Journals (Sweden)

    Putu Angga Kristyawan

    2013-09-01

    Full Text Available Pemakaian heat exchanger pada kompor industri tempe di kelurahan Tenggilis Mejoyo Surabaya mengaplikasikan tube heat exchanger dengan jumlah tube 4 dan material tembaga. Pada pemakaian awal mampu memperlama penggunaan bahan bakar hingga 3 hari. Proses produksi heat exchanger memerlukan biaya hingga 2,5 juta rupiah. Untuk dapat mengatasi masalah tersebut maka dilakukan penelitian tentang performansi heat exchanger dengan memvariasikan material dan jumlah tube. Masalah ini disimulasikan dengan computational fluid dynamics. Simulasi dilakukan pada jumlah grid 388149 dan model turbulensi dengan nilai deviasi terhadap temperatur ukur sebesar 1,0%. Hasil penelitian menunjukkan bahwa performansi dengan jumlah tube 6 dan material besi memiliki performansi yang hanya berbeda sebesar 2395,188 Watt dan nilai temperatur keluaran hanya berbeda 2,338 K dengan 4 tube tembaga. Nilai investasi 6 tube besi juga lebih rendah dibandingkan dengan 4 tube tembaga senilai Rp 4.335.866,00 dan perbedaan nilai payback period hingga 4,22 bulan.

  8. 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

  9. A heat transfer study for vertical straight-tube steam generators heated by liquid metal

    International Nuclear Information System (INIS)

    Valette, M.

    1984-04-01

    A single-tube mockup of a vertical straight-tube steam generator heated by sodium-potassium alloy NaK was submitted to thermal and hydraulic testing in conditions representative of fast breeder reactor operation. The mockup consisted of a 10mm I.D. ferritic steel heat exchange tube centered inside a cylindrical stainless steel shell. The complete assembly was 20.9 meters long. Water flowed upward inside the exchange tube, and NaK flowed downward in the annular gap between the tube and the shell. The steam outlet pressure ranged from 90 to 195 bars, while the liquid metal temperature at the mockup inlet was between 480 and 580 0 C. The water flowrate in the tube ranged from 153 to 2460 kg.m -2 .s -1 . During the tests the fluid inlet and outlet temperatures, flowrate and pressures were measured, as was the NaK temperature profile over the full length of the device. The test results were subsequently compared with heat exchange and pressure drop values calculated using the standard formulas for straight-tube heat exchangers. The heat exchange coefficients predicted by these correlations in the boiling zone were found to be largely overestimated, while the calculated pressure drop values proved satisfactory. A set of modified correlations is proposed to account for the observed phenomena, and for use in designing commercial units, provided the sodium flow in the tube bundle is adequately distributed

  10. A calibration system for X-ray generators and tube factors

    International Nuclear Information System (INIS)

    Healey, T.; Dickson, D.G.; Greenwood, M.W.B.

    1979-01-01

    An apparatus (Machlett's Dynalyzer II system) is described that makes real-time dynamic tests on the output and performance of the X-ray tube and generator so that a single exposure gives information in mA and mAs; exposure time (ms); kVp anode to earth; kVp cathode to earth; kvp cathode to anode; tube filament current; line voltage and radiation output of the tube (mR). The method of use is described together with the results of comparisons made with other test equipment. Some novel design features are of particular interest. The results show that by using this apparatus present design characteristics of X-ray generator-control-tube systems are such that the accuracy of calibration can be improved by at least an order of magnitude. (author)

  11. A reappraisal of steam generator tube rupture in the French licensing process

    International Nuclear Information System (INIS)

    Conte, M.; Gouffon, A.; Moriette, P.

    1984-10-01

    Upon the examination of the safety options submitted by EDF (Electricite de France) for a new pressurized water reactor design (N4, 1400 MWe), the French safety authorities decided that the conventionnal list of events to take under consideration should be amended as follows: failure of 1 and 2 steam generator tubes. To meet these objectives, design improvements were decided and new operating criteria were required by the technical specifications. Various preventive measures have been adopted by EDF to reduce tube degradation risks at the design stage, at the secondary feedwater quality level, and concerning also the quality control. The radiological consequences of generator tube integrity failure can be mitigated if the primary coolant activity is low, the tube flow detection is rapid, the release time is short, and the operating procedure is suitable and easily implemented [fr

  12. A calculating method of tube-to-tubesheet joints design for steam generator

    International Nuclear Information System (INIS)

    Zhang Fuyuan

    1993-01-01

    A theoretical calculating method of the hydraulically expanded tube-to-tubesheet joints design is described. As a mathematical model, the total expanded process of the joints is divided in four stages. with the elastic and plastic theories, the stress, strain and displacement of the tube or tube and tubesheet are analysed by stages, then expansion pressure, deformation, residual stress and push-out force are evaluated. The method may be used to design the steam generators and steel tubular heat exchangers. The paper points out that the hydraulic-expansion plus local roller expansion (hybrid expansion) is better than the only hydraulic-expansion for the tube-to-tubesheet joints of the nuclear steam generators

  13. Eddy current technology for heat exchanger and steam generator tube inspection

    Energy Technology Data Exchange (ETDEWEB)

    Obrutsky, L.; Lepine, B.; Lu, J.; Cassidy, R.; Carter, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2004-07-01

    A variety of degradation modes can affect the integrity of both heat exchanger (HX) and balance of plant tubing, resulting in expensive repairs, tube plugging or replacement of tube bundles. One key component for ensuring tube integrity is inspection and monitoring for detection and characterization of the degradation. In-service inspection of HX and balance of plant tubing is usually carried out using eddy current (EC) bobbin coils, which are adequate for the detection of volumetric degradations. However, detection and quantification of additional modes of degradation such as pitting, intergranular attack (IGA), axial cracking and circumferential cracking require specialized probes. The need for timely, reliable detection and characterization of these modes of degradation is especially critical in Nuclear Generating Stations. Transmit-receive single-pass array probes, developed by AECL, offer high defect detectability in conjunction with fast and reliable inspection capabilities. They have strong directional properties, permitting probe optimization for circumferential or axial crack detection. Compared to impedance probes, they offer improved performance in the presence of variable lift-off. This EC technology can help resolve critical detection issues at susceptible areas, such as the rolled-joint transitions at the tubesheet, U-bends and tube-support intersections. This paper provides an overview of the operating principles and the capabilities of advanced ET inspection technology available for HX tube inspection. Examples of recent application of this technology in Nuclear Generating Stations (NGSs) are discussed. (author)

  14. Eddy current technology for heat exchanger and steam generator tube inspection

    International Nuclear Information System (INIS)

    Obrutsky, L.; Lepine, B.; Lu, J.; Cassidy, R.; Carter, J.

    2004-01-01

    A variety of degradation modes can affect the integrity of both heat exchanger (HX) and balance of plant tubing, resulting in expensive repairs, tube plugging or replacement of tube bundles. One key component for ensuring tube integrity is inspection and monitoring for detection and characterization of the degradation. In-service inspection of HX and balance of plant tubing is usually carried out using eddy current (EC) bobbin coils, which are adequate for the detection of volumetric degradations. However, detection and quantification of additional modes of degradation such as pitting, intergranular attack (IGA), axial cracking and circumferential cracking require specialized probes. The need for timely, reliable detection and characterization of these modes of degradation is especially critical in Nuclear Generating Stations. Transmit-receive single-pass array probes, developed by AECL, offer high defect detectability in conjunction with fast and reliable inspection capabilities. They have strong directional properties, permitting probe optimization for circumferential or axial crack detection. Compared to impedance probes, they offer improved performance in the presence of variable lift-off. This EC technology can help resolve critical detection issues at susceptible areas, such as the rolled-joint transitions at the tubesheet, U-bends and tube-support intersections. This paper provides an overview of the operating principles and the capabilities of advanced ET inspection technology available for HX tube inspection. Examples of recent application of this technology in Nuclear Generating Stations (NGSs) are discussed. (author)

  15. Intergranular corrosion on the secondary coolant side of french PWR steam generators tubes

    International Nuclear Information System (INIS)

    Nordmann, F.; Cattant, F.; Comby, R.

    1990-01-01

    Intergranular corrosion on the OD of steam generator tubes in French units, led only to a very few plugged tubes, contrarily to most of the countries. Non destructive and destructive examinations have shown that corrosion at tube support plate level increases moderately and is likely initiated by sodium hydroxide; in addition, above tubesheet, significant and sometimes high contents of lead have been noted. Up to now, selected remedies include chemistry specifications with low sodium concentrations obtained by additional mixed bed on makeup water and power decreases for hideout return, when necessary [fr

  16. Evaluation of sealed-tube neutron generators for the assay of fresh LWR fuel assemblies

    International Nuclear Information System (INIS)

    Cutter, J.; Lee, D.; Lindquist, L.O.; Menlove, H.O.; Caldwell, J.T.; Atencio, J.D.; Kunz, W.E.

    1981-11-01

    The use of sealed-tube neutron generators for the active assay of fresh light-water reactor fuel assemblies has been investigated. The results of experimental tests of the Kaman 801 generator are presented. Neutron yields, source moderation, and transportability are discussed. A comparison is made with the AmLi neutron source for use in the Coincidence Collar

  17. Technical basis for the CANDU steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

    Kozluk, M.J.; Scarth, D.A.; Graham, D.B.

    2002-01-01

    Active degradation mechanisms in steam generators and preheaters in Canadian CANDU T M generating stations are managed through Steam Generator Programs that incorporate tube inspection, maintenance (cleaning), fitness-for-service assessment, and preventative plugging as part of the overall steam generator management strategy. Steam generator and preheater tubes are inspected in accordance with the CSA Standard CAN/CSA-N285.4-94[l]. When a detected flaw indication does not satisfy the criteria of acceptance by examination, CSA-N285.4-94 permits a fitness-for-service assessment to determine acceptability. In 1999 Ontario Power Generation issued, for trial use, fitness-for-service guidelines for steam generator and preheater tubes in CANDU nuclear power plants. The main objectives of the Fitness-for-Service Guidelines are to provide reasonable assurance that tube structural integrity is maintained, and to provide reasonable assurance that there are adequate margins between estimated accumulated dose and applicable site dose limits. The Fitness-for-Service Guidelines are intended to provide industry-standard acceptance criteria and evaluation procedures for assessing the condition of steam generator and preheater tubes in terms of tube structural integrity, operational leak rate, and consequential leakage during an upset or abnormal event. This paper describes the technical basis for the minimum required safety factors specified in Table IC-1 of the Fitness-for-Service Guidelines and for the flaw models used to develop the flaw stability requirements in the nonmandatory, Appendix C of the Fitness-for-Service Guidelines. (author)

  18. Diagnosis of 3-dimensional geometry and stress corrosion cracking in steam generator tubes

    International Nuclear Information System (INIS)

    Lee, D.H.; Choi, M.S.; Hur, D.H.; Kim, K.M.; Han, J.H.; Song, M.H.

    2015-01-01

    Most of the corrosive degradations in steam generator tubes of nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, an expansion transition, u-bend, dent, bulge, etc. Therefore, accurate information on a geometric anomaly (precursor of degradation) in a tube is a prerequisite to the activity of pre- and in-service non destructive inspection for a precise and earlier detection of a defect in order to prevent a failure during an operation, and also for a root cause analysis of a failure. In this paper, a new diagnostic eddy current probe technology which has simultaneous dual function of a 3-dimensional geometry measurement and defect detection in steam generator tube is introduced. The D-Probe is a rotary type eddy current coil probe equipped with 3 different eddy current coil units (surface riding type plus-point and pancake coils for defect detection, and non-surface riding type shielded high frequency pancake coil for tube profile measurement). A specific data analysis software has been developed. By comparing the eddy current data from the defect with those from the geometric changes, the relationship between the degradation and geometric changes can be revealed. Also, it supplies information on tube location at which defect is most probable and thus, a more efficient detection of earlier degradation. The use of D-probe and analysis software has been demonstrated for steam generator tubes with various geometric anomalies in manufacturing and operating nuclear power plants

  19. Quasi-Monochromatic Flash X-Ray Generator Utilizing Disk-Cathode Molybdenum Tube

    Science.gov (United States)

    Sato, Eiichi; Sagae, Michiaki; Tanaka, Etsuro; Hayasi, Yasuomi; Germer, Rudolf; Mori, Hidezo; Kawai, Toshiaki; Ichimaru, Toshio; Sato, Shigehiro; Takayama, Kazuyoshi; Ido, Hideaki

    2004-10-01

    High-voltage condensers in a polarity-inversion two-stage Marx surge generator are charged from -40 to -60 kV using a power supply, and the electric charges in the condensers are discharged to an X-ray tube after closing the gap switches in the surge generator using a trigger device. The X-ray tube is a demountable diode, and the turbomolecular pump evacuates air from the tube with a pressure of approximately 1 mPa. Sharp K-series characteristic X-rays of molybdenum are produced without using a monochromatic filter, since the tube utilizes a disk cathode and a rod target, and bremsstrahlung rays are not emitted in the opposite direction to that of electron acceleration. The peak tube voltage increased with increasing charging voltage and increasing space between the target and cathode electrodes. At a charging voltage of -60 kV and a target-cathode space of 1.0 mm, the peak tube voltage and current were 110 kV and 0.75 kA, respectively. The pulse width ranged from 40 to 100 ns, and the maximum dimension of the X-ray source was 3.0 mm in diameter. The number of generator-produced K photons was approximately 7× 1014 photons/cm2\\cdots at 0.5 m from the source.

  20. Liquid metal fast breeder reactor steam generator: behaviour of heat exchange tubes in face of a through crack resulting in a contact between sodium and water

    International Nuclear Information System (INIS)

    Quinet, J.L.; Lannou, L.

    1978-01-01

    The results of a survey made Electricite de France on the behaviour of cracked tubes under operating conditions of an industrial steam generator are submitted in this communication. A comparison is made of the tube material: INCOLOY 800, 2 1/4 Cr-1 Mo, 9 Cr-2 Mo land to the initial leak. Finally, a description is given of the self-development process of a water leak into sodium. (author)

  1. Materials for generation-IV nuclear reactors

    International Nuclear Information System (INIS)

    Alvarez, M. G.

    2009-01-01

    Materials science and materials development are key issues for the implementation of innovative reactor systems such as those defined in the framework of the Generation IV. Six systems have been selected for Generation IV consideration: gas-cooled fast reactor, lead-cooled fast reactor, molten salt-cooled reactor, sodium-cooled fast reactor, supercritical water-cooled reactor, and very high temperature reactor. The structural materials need to resist much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. For this reason, the first consideration in the development of Generation-IV concepts is selection and deployment of materials that operate successfully in the aggressive operating environments expected in the Gen-IV concepts. This paper summarizes the Gen-IV operating environments and describes the various candidate materials under consideration for use in different structural applications. (author)

  2. Development of a multiphase shock tube for energetic materials characterization.

    Energy Technology Data Exchange (ETDEWEB)

    Castaneda, Jaime N.; Cooper, Marcia A.; Beresh, Steven Jay; Trott, Wayne Merle; Wagner, Justin L.; Kearney, Sean Patrick; Baer, Melvin R.; Pruett, Brian O. M.

    2010-06-01

    A novel multiphase shock tube to study particle dynamics in gas-solid flows has been constructed and tested. Currently, there is a gap in data for flows having particle volume fractions between the dusty and granular regimes. The primary purpose of this new facility is to fill that gap by providing high quality data of shock-particle interactions in flows having dense gas particle volume fractions. Towards this end, the facility aims to drive a shock into a spatially isotropic field, or curtain, of particles. Through bench-top experimentation, a method emerged for achieving this challenging task that involves the use of a gravity-fed contoured particle seeder. The seeding method is capable of producing fields of spatially isotropic particles having volume fractions of about 1 to 35%. The use of the seeder in combination with the shock tube allows for the testing of the impingement of a planar shock on a dense field of particles. The first experiments in the multiphase shock tube have been conducted and the facility is now operational.

  3. Nano surface generation of grinding process using carbon nano tubes

    Indian Academy of Sciences (India)

    1 TPa vs 70 GPa for aluminum, steel 2 Gpa and 700 GPa for C-fibre. The strength to weight ratio is 500 times greater than Aluminum (Sakas & Simpson 2007). Maximum strain will be. 10% much higher than any material. Thermal conductivity 3000 W/mK in the axial direction with small values in the radial direction.

  4. The new generation of packing materials

    International Nuclear Information System (INIS)

    Malikov, T.S.; Dzhonmurodov, A.S.; Usmanova, S.R.; Teshaev, Kh.I.; Mukhidinov, Z.K.

    2016-01-01

    Present article is devoted to new generation of packing materials. The methods of extraction and investigation of component composition and properties of whey protein, zein, carboxymethylcellulose, hyaluronic acid and pectins were elaborated in order to further application them in pharmaceutical industry as composite materials and for capsulation of medicines.

  5. High temperature technological heat exchangers and steam generators with helical coil assembly tube bundle

    International Nuclear Information System (INIS)

    Korotaev, O.J.; Mizonov, N.V.; Nikolaevsky, V.B.; Nazarov, E.K.

    1990-01-01

    Analysis of thermal hydraulics characteristics of nuclear steam generators with different tube bundle arrangements and waste heat boilers for ammonia production units was performed on the basis of operating experience results and research and development data. The present report involves the obtained information. The estimations of steam generator performances and repair-ability are given. The significant temperature profile of the primary and secondary coolant flows are attributed to all steam generator designs. The intermediate mixing is found to be an effective means of temperature profile overcoming. At present the only means to provide an effective mixing in heat exchangers of the following types: straight tubes, field tubes, platen tubes and multibank helical coil tubes (with complicated bend distribution along their length) are section arrangements in series in conjunction with forced and natural mixing in connecting lines. Development of the unificated system from mini helical coil assemblies allows to design and manufacture heat exchangers and steam generators within the wide range of operating conditions without additional expenses on the research and development work

  6. 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.

  7. Status of the steam generator tube circumferential ODSCC degradation experienced at the Doel 4 plant

    International Nuclear Information System (INIS)

    Roussel, G.

    1997-01-01

    Since the 1991 outage, the Doel Unit 4 nuclear power plant is known to be affected by circumferential outside diameter intergranular stress corrosion cracking at the hot leg tube expansion transition. Extensive non destructive examination inspections have shown the number of tubes affected by this problem as well as the size of the cracks to have been increasing for the three cycles up to 1993. As a result of the high percentage of tubes found non acceptable for continued service after the 1993 in-service inspection, about 1,700 mechanical sleeves were installed in the steam generators. During the 1994 outage, all the tubes sleeved during the 1993 outage were considered as potentially cracked to some extent at the upper hydraulic transition and were therefore not acceptable for continued service. They were subsequently repaired by laser welding. Furthermore all the tubes not sleeved during the 1993 outage were considered as not acceptable for continued service and were repaired by installing laser welded sleeves. During the 1995 outage, some unexpected degradation phenomena were evidenced in the sleeved tubes. This paper summarizes the status of the circumferential ODSCC experienced in the SG tubes of the Doel 4 plant as well as the other connected degradation phenomena

  8. Preliminary design study of removable integral steam generator units of the multiple helically wound tube type for a 1250 MW(th) H.T.G.C. reactor

    International Nuclear Information System (INIS)

    Gilli, P.V.; Fritz, K.; Lippitsch, J.; Sandri, A.H.; Weiss, B.

    1965-11-01

    The possibilities of designing a multiple steam generator for a 1250 MW(th) High Temperature Gas-Cooled Reactor, consisting of 18 units which are able to pass through 5 ft diam. holes in the integral prestressed concrete pressure vessel are investigated. A lay-out and design with bundles of multi-start helical tubes is evolved, particular attention being paid to the questions of tube blanking and removal of the unit, and of selection of materials for superheater and reheater tubes. Thermal and stress calculations have been carried out, using the Waagner-Biro Computer Code ADURHELIX. (author)

  9. Transmit-receive eddy current probes for defect detection and sizing in steam generator tubes

    International Nuclear Information System (INIS)

    Obrutsky, L.S.; Cecco, V.S.; Sullivan, S.P.

    1997-01-01

    Inspection of steam generator tubes in aging Nuclear Generating Stations is increasingly important. Defect detection and sizing, especially in defect prone areas such as the tubesheet, support plates and U-bend regions, are required to assess the fitness-for-service of the steam generators. Information about defect morphology is required to address operational integrity issues, i.e., risk of tube rupture, number of tubes at risk, consequential leakage. A major challenge continues to be the detection and sizing of circumferential cracks. Utilities around the world have experienced this type of tube failure. Conventional in-service inspection, performed with eddy current bobbin probes, is ineffectual in detecting circumferential cracks in tubing. It has been demonstrated in CANDU steam generators, with deformation, magnetite and copper deposits that multi-channel probes with transmit-receive eddy current coils are superior to those using surface impedance coils. Transmit-receive probes have strong directional properties, permitting probe optimization according to crack orientation. They are less sensitive to lift-off noise and magnetite deposits and possess good discrimination to internal defects. A single pass C3 array transmit-receive probe developed by AECL can detect and size circumferential stress corrosion cracks as shallow as 40% through-wall. Since its first trial in 1992, it has been used routinely for steam generator in-service inspection of four CANDU plants, preventing unscheduled shutdowns due to leaking steam generator tubes. More recently, a need has surfaced for simultaneous detection of both circumferential and axial cracks. The C5 probe was designed to address this concern. It combines transmit-receive array probe technology for equal sensitivity to axial and circumferential cracks with a bobbin probe for historical reference. This paper will discuss the operating principles of transmit-receive probes, along with inspection results

  10. Steady-state heat transfer in an inverted U-tube steam generator

    International Nuclear Information System (INIS)

    Boucher, T.J.

    1986-01-01

    Experimental results are presented involving U-tube steam generator tube bundle local heat transfer and fluid conditions during steady-state, full-power operations performed at high temperatures and pressures with conditions typical of a pressurized water reactor (15.0 MPa primary pressure, 600 K hot-leg fluid temperatures, 6.2 MPa secondary pressure). The MOD-2C facility represents the state-of-the-art in measurement of tube local heat transfer data and average tube bundle secondary fluid density at several elevations, which allows an estimate of the axial heat transfer and void distributions during steady-state and transient operations. The method of heat transfer data reduction is presented and the heat flux, secondary convective heat transfer coefficient, and void fraction distributions are quantified for steady-state, full-power operations

  11. Reduction of background luminance generated in the output screen of x-ray image intensifier tubes

    International Nuclear Information System (INIS)

    Tsuda, Motohisa; Kimura, Yutaro

    1985-01-01

    Background luminance of several origins introduced into an X-ray image intensifier tube deteriorates the contrast of its output image. The output screen, which consists of a phosphor layer usually prepared on a transparent substrate, generates influential background luminance in the X-ray image intensifier tube. A theoretical and experimental study on how reduce the background luminance of the output screen by increasing the thickness of the substrate is presented. In evaluating the effect of the background luminance, we calculated and measured the contrast ratio usually used as a characteristic index of the image intensifier tube. The theory and results from the experimental image intensifier tubes show good agreement. This method of background luminance reduction is not only more economical but also superior to the other methods utilizing dark substrates or optical fiber plates. (author)

  12. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mendler, O J; Takeuchi, K; Young, M Y

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

  13. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    International Nuclear Information System (INIS)

    Mendler, O.J.; Takeuchi, K.; Young, M.Y.

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results

  14. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    1977-01-01

    A means and method for energizing and regulating a neutron generator tube is described. It has a target, an ion source and a replenisher. A negative high voltage is applied to the target and the target current monitored. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordance with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current. (C.F.)

  15. Eddy-current tests on operational evaluation of steam generator tubes in nuclear power plants

    International Nuclear Information System (INIS)

    Lopez, Luiz Antonio Negro Martin; Ting, Daniel Kao Sun

    2000-01-01

    This paper presents a worldwide research on the technical and economical impacts due to failure in tube bundles of nuclear power plant steam generators. An Eddy current non destructive test using Foucault currents for the inspection and failure detection on the tubes, and also the main type of defects. The paper also presents the signals generated by a Zetec MIZ-40 test equipment. This paper also presents a brief description of an automatic system for data analysis which is under development by using a fuzzy logic and artificial intelligence

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

    OpenAIRE

    M. Głowacka; J. Łabanowski; S. Topolska

    2012-01-01

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

  17. Steam generator tube rupture: studies to improve plant procedure

    International Nuclear Information System (INIS)

    Tellier, N.; Zilliox, C.

    1984-10-01

    These accidents have the particularities to lead to atmospheric radioactive release and to be able to be determinated with appropriate operator actions. These radioactive releases are function of several parameters of which sensitivity is analyzed. The major part of the calculations were performed by EDF with an home made code called ''AXEL''. The main conclusions are: - the optimization of the safety injection monitoring to minimize radioactive releases to atmosphere, while ensuring the cooling of the core; - the radioactive releases to atmosphere are very low in any case but much more important if the filling of the steam generator secondary side cannot be avoided

  18. Wavelets transforms and fuzzy logic in the eddy-current inspection of nuclear power plants steam generator tubes

    International Nuclear Information System (INIS)

    Lopez, Luiz Antonio Negro Martin

    2002-01-01

    Nuclear power plants steam generators around the world have presented early damage history in their tubes, caused either by design errors or by inappropriate operation, which besides reducing the availability and the safety of the nuclear power plants it also generates heavy economical burden. To monitor the steam generators operational condition, the Eddy Current testing of their tubes is the non destructive method used to detect, localize, classify and to size the defects. The inspection is performed by inserting probes with coils in the tubes generating a signal correlated to the defect. These signals produced by the probe electric circuit are composed by the resistance and the inductive components which can be combined to produce a Lissajous figure in the complex plane. However, Eddy-Current signals contain noise which induce subjectivity inducing to errors in the inspector diagnosis. It is not uncommon to have different diagnosis from two inspectors about the same signal. The present work has the objective of supplying a methodology to analyze the signals which could help the inspector in the difficult task of interpreting the Eddy Current signals. It is proposed a method to remove the noise based on Wavelets Transforms. It is also proposed a normalization in the signal phase angle measurements. Furthermore, two additional characteristics are also studied, namely: the signal amplitudes and the widths of the Lissajous petals. The use of a Fuzzy Logic based inference engine is also developed and its use is demonstrated to be viable. The defects studied in this work are those which produces volumetric changes in the material. In order to test the proposed methodology, several artificial defects were produced in tubes using different types of materials like: brass, 316L stainless steel and Inconel 600 to produce a experimental data base. An Eddy-Current inspection equipment, the MIZ-17ET was used. Around 1000 time series signals of defects were acquired through

  19. Predicted wear on the tube outside surface due to foreign object in the secondary side of steam generator

    International Nuclear Information System (INIS)

    Kim, Hyung Nam; Cho, Nam Cheoul

    2012-01-01

    It is necessary to evaluate the effects of foreign objects on steam generator tubes and to use this information to take appropriate safety precautions to prevent nuclear accidents. Foreign objects may include loose parts from the feed water system and items lost by workers during o/h, and may flow into the secondary side of steam generators during operation. A foreign object could damage steam generator tube walls if there is relative motion between the tube and the foreign object. This is especially true for foreign objects that land on the tube sheet because the velocity of cross flow, which creates a contact force between the tube and foreign object, is relatively high there. During steam generator overhauls, foreign objects are detected by non destructive methods such as the visual test and/or the eddy current test. Confirmed foreign objects should be removed for nuclear safety. The Foreign Object Search and Retrieval System (FOSAR) can be used to remove foreign objects from the steam generators with a square tube array. However, the FOSAR cannot be used (or can be used in only a very restricted area, such as the outside of the tube bundle) in the steam generators with a triangular tube array. In order to continue nuclear power plant operations without removing foreign objects, the integrity of the steam generator tube must be verified. This paper introduces a practical method developed to evaluate the effects of foreign objects detected on tube sheets in the secondary sides of steam generators

  20. Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R. [Argonne National Lab., IL (United States)

    1997-02-01

    The objective of the workshop was to provide a working forum for the exchange of information by contributing experts on current issues related to PWR steam generator tube integrity. One hundred persons from 15 countries attended the workshop, including 36 from regulatory and nuclear policy agencies, 28 from research and development laboratories, 18 from nuclear vendors and consulting firms, and 18 from electrical utilities. The workshop opened with a plenary session; the first part of the session covered international steam generator regulatory practices and issues, featuring speakers from regulatory bodies in Belgium, France, Japan, Spain, and the US. In Part 2 of the plenary session, comprehensive technical overviews on steam generator tubing degradation, inspection, and integrity were presented by authorities in these fields from the US, France, and Belgium. Parallel working sessions on the second and third days of the workshop then developed findings and recommendations in the areas of (1) tubing degradation, (2) tubing inspection, (3) tubing integrity, (4) preventative and corrective measures, and (5) operational aspects and risk analysis. On the final day of the workshop, the working-session facilitators presented summaries of their sessions to the workshop attendees. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  1. Monochromatic flash x-ray generator utilizing a disk-cathode silver tube

    Science.gov (United States)

    Sato, Eiichi; Hayasi, Yasuomi; Germer, Rudolf K. F., II; Tanaka, Etsuro; Mori, Hidezo; Kawai, Toshiaki; Inoue, Takashi; Ogawa, Akira; Sato, Shigehiro; Ichimaru, Toshio; Takayama, Kazuyoshi; Onagawa, Jun; Ido, Hideaki

    2005-09-01

    The high-voltage condensers in a polarity-inversion two-stage Marx surge generator are charged from -50 to -70 kV by a power supply, and the electric charges in the condensers are discharged to an x-ray tube after closing gap switches in the surge generator with a trigger device. The x-ray tube is a demountable diode, and the turbomolecular pump evacuates air from the tube with a pressure of approximately 1 mPa. Clean silver Kα lines are produced using a 30-μm-thick palladium filter, since the tube utilizes a disk cathode and a rod target, and bremsstrahlung rays are not emitted in the opposite direction to that of electron acceleration. At a charging voltage of -70 kV, the instantaneous tube voltage and current are 90 kV and 0.8 kA, respectively. The x-ray pulse widths are approximately 80 ns, and the instantaneous number of generator-produced Kα photons is approximately 4×107photons/cm2 per pulse at 0.3 m from the source 3.0 mm in diameter.

  2. Monochromatic flash x-ray generator utilizing disk-cathode silver tube

    Science.gov (United States)

    Sato, Eiichi; Hayasi, Yasuomi; Germer, Rudolf K.; Tanaka, Etsuro; Mori, Hidezo; Kawai, Toshiaki; Ichimaru, Toshio; Takayama, Kazuyoshi; Ido, Hideaki

    2004-11-01

    The high-voltage condensers in a polarity-inversion two-stage Marx surge generator are charged from -50 to -70 kV by a power supply, and the electric charges in the condensers are discharged to an x-ray tube after closing gap switches in the surge generator with a trigger device. The x-ray tube is a demountable diode, and the turbomolecular pump evacuates air from the tube with a pressure of approximately 1 mPa. Clean silver Kα lines are produced using a 30 μm-thick palladium filter, since the tube utilizes a disk cathode and a rod target, and bremsstrahlung rays are not emitted in the opposite direction to that of electron acceleration. At a charging voltage of -70 kV, the instantaneous tube voltage and current were 90 kV and 0.8 kA, respectively. The x-ray pulse widths were approximately 80 ns, and the instantaneous number of generator-produced Kα photons was approximately 40 M photons/cm2 per pulse at 0.3 m from the source of 3.0 mm in diameter.

  3. Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants

    International Nuclear Information System (INIS)

    Diercks, D.R.

    1997-02-01

    The objective of the workshop was to provide a working forum for the exchange of information by contributing experts on current issues related to PWR steam generator tube integrity. One hundred persons from 15 countries attended the workshop, including 36 from regulatory and nuclear policy agencies, 28 from research and development laboratories, 18 from nuclear vendors and consulting firms, and 18 from electrical utilities. The workshop opened with a plenary session; the first part of the session covered international steam generator regulatory practices and issues, featuring speakers from regulatory bodies in Belgium, France, Japan, Spain, and the US. In Part 2 of the plenary session, comprehensive technical overviews on steam generator tubing degradation, inspection, and integrity were presented by authorities in these fields from the US, France, and Belgium. Parallel working sessions on the second and third days of the workshop then developed findings and recommendations in the areas of (1) tubing degradation, (2) tubing inspection, (3) tubing integrity, (4) preventative and corrective measures, and (5) operational aspects and risk analysis. On the final day of the workshop, the working-session facilitators presented summaries of their sessions to the workshop attendees. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  4. Electrochemical generation of mercury cold vapor and its in-situ trapping in gold-covered graphite tube atomizers

    International Nuclear Information System (INIS)

    Cerveny, Vaclav; Rychlovsky, Petr; Netolicka, Jarmila; Sima, Jan

    2007-01-01

    The combination of more efficient flow-through electrochemical mercury cold vapor generation with its in-situ trapping in a graphite tube atomizer is described. This coupled technique has been optimized to attain the maximum sensitivity for Hg determination and to minimize the limits of detection and determination. A laboratory constructed thin-layer flow-through cell with a platinum cathode served as the cold vapor generator. Various cathode arrangements with different active surface areas were tested. Automated sampling equipment for the graphite atomizer with an untreated fused silica capillary was used for the introduction of the mercury vapor. The inner surface of the graphite tube was covered with a gold foil placed against the sampling hole. The results attained for the electrochemical mercury cold vapor generation (an absolute limit of detection of 80 pg; peak absorbance, 3σ criterion) were compared with the traditional vapor generation using NaBH 4 as the reducing agent (an absolute limit of detection of 124 pg; peak absorbance, 3σ criterion). The repeatability at the 5 ng ml -1 level was better than 4.1% (RSD) for electrochemical mercury vapor generation and better than 5.6% for the chemical cold vapor generation. The proposed method was applied to the determination the of Hg contents in a certified reference material and in spiked river water samples

  5. Statistical prediction of the numbers of degraded tubes in nuclear power plant steam generators

    International Nuclear Information System (INIS)

    Gallucci, R.H.V.; Klisiewicz, J.W.; Craig, K.R.

    1990-01-01

    Corrosion of nuclear power plant steam generator (SG) tubes often necessitates plugging/sleeving, causing decreased SG thermal performance and possible SG replacement. Statistical methods have been developed to predict probabilistically the numbers of tubes degraded due to secondary side pitting, wastage, and intergranular attack/stress-corrosion cracking. Inspection data from two Combustion Engineering (C-E) plants have been converted into statistics representing defect formation and growth. Computer simulation programs have been generated to predict the numbers of tubes to be plugged/sleeved during future outages. The probabilistic predictions for both plants successfully have bounded subsequent observations. While so far applied only to C-E SGs for the three degradation phenomena, the statistical methodology is adaptable to other SG types and phenomena

  6. Hideout of sea water impurities in steam generator tube deposits: laboratory and field studies

    International Nuclear Information System (INIS)

    Balakrishnan, P.V.; Turner, C.W.; Thompson, R.; Sawochka, S.

    1996-01-01

    Sea water impurities hide out within thin (∼10 μm) deposits on steam generator tubes, as demonstrated by both laboratory studies using segments of fouled steam generator tubes pulled in 1992 from Crystal River-3 nuclear power station and field hideout return studies performed during recent plant shutdowns. Laboratory tests performed at 279 o C (534 o F) and heat fluxes ranging from 35 to 114 kW/m 2 (11,100 - 36,150 Btu/h.ft 2 ), conditions typical of the lower tubesheet to the first support plate region of a once-through steam generator, showed that impurity hideout can occur in thin free-span tube deposits. The extent of hideout increased with increasing heat flux. Soluble species, such as sodium and chloride ions, returned promptly to the bulk water from the deposits when the heat flux was turned off, whereas less soluble species, such as calcium sulfate and magnesium hydroxide, returned more slowly. Recent field hideout return studies performed at Crystal River-3 where the water level in the steam generators was maintained below the first tube support plate during the shutdown, thus wetting only the thin deposits in the free span and the small sludge pile, corroborate the laboratory findings, showing that hideout does indeed occur in the free-span regions of the tubes. These findings suggest that hideout within tube deposits has to be accounted for in the calculation of crevice chemistry from hideout return studies and in controlling the bulk chemistry using the molar ratio criterion. (author). 3 refs., 4 tabs., 3 figs

  7. Third-Generation Display Technology: Nominally Transparent Material

    Directory of Open Access Journals (Sweden)

    Charles Willow

    2010-12-01

    Full Text Available Display technology is reshaping the consumer, business, government, and even not-for-profit markets in the midst of the digital convergence, coupled with recent smart phones led by Apple, Inc. First-Generation (1G display technology was dominated by the Cathode Ray Tubes, followed by Liquid Crystal Display and Plasma in 2G. A radically innovative shift as a disruptive technology is expected to follow in 3G to utilize virtually any transparent material, which wirelessly connects to portable access points. This paper studies the feasibility of the 3G Display Technology (DT with Technology S-Curves, and presents possible business models and technology strategies which may be generated from it. Additional subsets of business models may be derived for a wide range of industry applications.

  8. Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

    International Nuclear Information System (INIS)

    Durbec, V.; Pitner, P.; Pages, D.; Riffard, T.; Flesch, B.

    1997-10-01

    Steam Generators (SG) of Pressurized Water Reactors have experienced world wide various types of tube degradations, mainly from stress corrosion cracking; because of this damage, primary-secondary leakage or tube rupture can occur. Safety against the risk of tube rupture is achieved through a combination of periodic in-service inspections (eddy current testing), surveillance of leaks during operation (leak before break concept) and tube plugging. In order to optimize the tube bundle SG maintenance, Electricite de France has developed a specific software named COMPROMIS. The model, based on probabilistic fracture mechanics makes it possible to quantify the influence of in service inspections and maintenance work on the risk of a SG Tube Rupture (SGTR), taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive examinations, crack initiation and propagation, critical sizes, leak before risk of break, etc...). This paper focuses on the leak rate calculation module and presents a sensitivity study of the influence of the leak before break on the conditional failure probability. (author)

  9. Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

    Energy Technology Data Exchange (ETDEWEB)

    Durbec, V.; Pitner, P.; Pages, D. [Electricite de France, 78 - Chatou (France). Research and Development Div.; Riffard, T. [Electricite de France, 69 - Villeurbanne (France). Engineering and Construction Div.; Flesch, B. [Electricite de France, 92 - Paris la Defense (France). Generation and Transmission Div.

    1997-10-01

    Steam Generators (SG) of Pressurized Water Reactors have experienced world wide various types of tube degradations, mainly from stress corrosion cracking; because of this damage, primary-secondary leakage or tube rupture can occur. Safety against the risk of tube rupture is achieved through a combination of periodic in-service inspections (eddy current testing), surveillance of leaks during operation (leak before break concept) and tube plugging. In order to optimize the tube bundle SG maintenance, Electricite de France has developed a specific software named COMPROMIS. The model, based on probabilistic fracture mechanics makes it possible to quantify the influence of in service inspections and maintenance work on the risk of a SG Tube Rupture (SGTR), taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive examinations, crack initiation and propagation, critical sizes, leak before risk of break, etc...). This paper focuses on the leak rate calculation module and presents a sensitivity study of the influence of the leak before break on the conditional failure probability. (author) 8 refs.

  10. Investigation of reliability of EC method for inspection of VVER steam generator tubes

    International Nuclear Information System (INIS)

    Corak, Z.

    2004-01-01

    Complete and accurate non-destructive examinations (NDE) data provides the basis for performing mitigating actions and corrective repairs. It is important that detection and characterization of flaws are done properly at an early stage. EPRI Document PWR Steam Generator Examination Guidelines recommends an approach that is intended to provide the following: Ensure accurate assessment of steam generator tube integrity; Extend the reliable, cost effective, operating life of the steam generators, and Maximize the availability of the unit. Steam Generator Eddy Current Data Analysis Performance Demonstration represents the culmination of the intense two-year industry effort in the development of a performance demonstration program for eddy current testing (ECT) of steam generator tubing. It is referred to as the Industry Database (IDB) and provides a capability for individual organizations to implement SG ECT performance demonstration programs in accordance with the requirements specified in Appendices G and H of the ISI Guidelines. The Appendix G of EPRI Document PWR Steam Generator Examination Guidelines specifies personnel training and qualification requirements for NDE personnel who analyze NDE data for PWR steam generator tubing. Its purpose is to insure a continuing uniform knowledge base and skill level for data analysis. The European methodology document is intended to provide a general framework for development of qualifications for the inspection of specific components to ensure they are developed in a consistent way throughout Europe while still allowing qualification to be tailored in detail to meet different nation requirements. In the European methodology document one will not find a detailed description of how the inspection of a specific component should be qualified. A recommended practice is a document produced by ENIQ to support the production of detailed qualification procedures by individual countries. VVER SG tubes are inspected by EC method but a

  11. Prevention of SCC occurring in an expansion transition region of steam generator tubing by Ni-plating in PWRs

    International Nuclear Information System (INIS)

    Kim, J.S.; Kim, M.J.; Kim, D.J.; Kim, H.P.

    2012-01-01

    Applicability of a nickel-plating technique was investigated for a possible proactive method to prevent stress corrosion cracking in the expansion transition region of pressurized water reactor steam generator tubing around the top of the tubesheet. The surface of steam generator tubes is plated with nickel in the region from the bottom ends of the U-tubes up to above the location where the tube is to be expanded, before the expansion process. The nickel-plated regions of the tubes are then inserted into the holes of the tube sheet and expanded to build up a steam generator. In order to verify the applicability and the effectiveness of the technique, mockup tests were performed for nickel-plated Alloy 600 HTMA tubes with hydraulic expansions. Integrity of the expanded nickel plating layers was examined and susceptibility to SCC was evaluated by using the C-ring and the slow strain rate tests in simulated pressurized water reactor environments. (author)

  12. Advanced Thermoelectric Materials for Radioisotope Thermoelectric Generators

    Science.gov (United States)

    Caillat, Thierry; Hunag, C.-K.; Cheng, S.; Chi, S. C.; Gogna, P.; Paik, J.; Ravi, V.; Firdosy, S.; Ewell, R.

    2008-01-01

    This slide presentation reviews the progress and processes involved in creating new and advanced thermoelectric materials to be used in the design of new radioiootope thermoelectric generators (RTGs). In a program with Department of Energy, NASA is working to develop the next generation of RTGs, that will provide significant benefits for deep space missions that NASA will perform. These RTG's are planned to be capable of delivering up to 17% system efficiency and over 12 W/kg specific power. The thermoelectric materials being developed are an important step in this process.

  13. Principal-Generated YouTube Video as a Method of Improving Parental Involvement

    Science.gov (United States)

    Richards, Joey

    2013-01-01

    The purpose of this study was to evaluate the involvement level of parents and reveal whether principal-generated YouTube videos for regular communication would enhance levels of parental involvement at one North Texas Christian Middle School (pseudonym). The following questions guided this study: 1. What is the beginning level of parental…

  14. Evaluation of ECT reliability for axial ODSCC in steam generator tubes

    International Nuclear Information System (INIS)

    Lee, Jae Bong; Park, Jai Hak; Kim, Hong Deok; Chung, Han Sub

    2010-01-01

    The integrity of steam generator tubes is usually evaluated based on eddy current test (ECT) results. Because detection capacity of the ECT is not perfect, all of the physical flaws, which actually exist in steam generator tubes, cannot be detected by ECT inspection. Therefore it is very important to analyze ECT reliability in the integrity assessment of steam generators. The reliability of an ECT inspection system is divided into reliability of inspection technique and reliability of quality of analyst. And the reliability of ECT results is also divided into reliability of size and reliability of detection. The reliability of ECT sizing is often characterized as a linear regression model relating true flaw size data to measured flaw size data. The reliability of detection is characterized in terms of probability of detection (POD), which is expressed as a function of flaw size. In this paper the reliability of an ECT inspection system is analyzed quantitatively. POD of the ECT inspection system for axial outside diameter stress corrosion cracks (ODSCC) in steam generator tubes is evaluated. Using a log-logistic regression model, POD is evaluated from hit (detection) and miss (no detection) binary data obtained from destructive and non-destructive inspections of cracked tubes. Crack length and crack depth are considered as variables in multivariate log-logistic regression and their effects on detection capacity are assessed using two-dimensional POD (2-D POD) surface. The reliability of detection is also analyzed using POD for inspection technique (POD T ) and POD for analyst (POD A ).

  15. Experimental facility design for study of fretting in steam generator tubes

    International Nuclear Information System (INIS)

    Balbiani, J.P.; Bergant, M.; Yawny, A.

    2012-01-01

    The design of an experimental facility for fretting wear testing of steam generator tubes under pressurized water up to 340 o C, is presented. The main component of the device consists in an autoclave which permits to recreate steam generator operating conditions. CAD CATIA V5R18, CAE ABAQUS and ASME Sec. VII Div. 1 (Rules for Construction of Pressure Vessels) were used along the design process. The design of the autoclave included the pressure vessel itself and the necessary flanges and nozzles. In addition, an axial dynamic sealing system was designed to allow for actuation from outside the pressure boundary. Complementary, typical tube - support contact conditions were analyzed and the principal variables affecting their mutual interaction determined. In addition, a simple device which allows performing fretting wear testing on steam generator tubes in air at room temperature was fabricated and the feasibility of a quantitative assessment of different aspects related with the fretting induced damage was explored. Characterization techniques available at Centro Atomico Bariloche, like light microscopy, scanning electron microscopy (SEM), energy dispersive analysis of X-ray (EDAX) and surface damage analysis by optic profilometry were shown to be appropriate for this aim. The designed facility will allow evaluating fretting damage of tubes - support combinations that might be used on the steam generator of the prototype reactor CAREM-25. It is also expected it could be applied to characterize fretting severity in other applications (nuclear fuel elements) (author)

  16. Stresses of steam generator U-tubes affecting stress corrosion cracking

    International Nuclear Information System (INIS)

    Yashima, S.; Ikenaga, H.; Nakamura, K.; Takaba, O.; Uragami, K.; Utsumi, H.

    1982-01-01

    Stress factors affecting U-bend cracking in the steam generators of PWR type reactors are discussed based on the results of stress corrosion cracking tests of Inconel 600 U-bend tube in polythionic acid solution subjected to the actual operating loads

  17. Extraction: a system for automatic eddy current diagnosis of steam generator tubes in nuclear power plants

    International Nuclear Information System (INIS)

    Georgel, B.; Zorgati, R.

    1994-01-01

    Improving speed and quality of Eddy Current non-destructive testing of steam generator tubes leads to automatize all processes that contribute to diagnosis. This paper describes how we use signal processing, pattern recognition and artificial intelligence to build a software package that is able to automatically provide an efficient diagnosis. (authors). 2 figs., 5 refs

  18. Steam generator tube integrity program. Semiannual report, August 1995--March 1996

    International Nuclear Information System (INIS)

    Diercks, D.R.; Bakhtiari, S.; Chopra, O.K.

    1997-04-01

    This report summarizes work performed by Argonne National Laboratory on the Steam Generator Tube Integrity Program from the inception of that program in August 1995 through March 1996. The program is divided into five tasks, namely (1) Assessment of Inspection Reliability, (2) Research on ISI (in-service-inspection) Technology, (3) Research on Degradation Modes and Integrity, (4) Development of Methodology and Technical Requirements for Current and Emerging Regulatory Issues, and (5) Program Management. Under Task 1, progress is reported on the preparation of and evaluation of nondestructive evaluation (NDE) techniques for inspecting a mock-up steam generator for round-robin testing, the development of better ways to correlate burst pressure and leak rate with eddy current (EC) signals, the inspection of sleeved tubes, workshop and training activities, and the evaluation of emerging NDE technology. Under Task 2, results are reported on closed-form solutions and finite element electromagnetic modeling of EC probe response for various probe designs and flaw characteristics. Under Task 3, facilities are being designed and built for the production of cracked tubes under aggressive and near-prototypical conditions and for the testing of flawed and unflawed tubes under normal operating, accident, and severe accident conditions. In addition, crack behavior and stability are being modeled to provide guidance on test facility design, to develop an improved understanding of the expected rupture behavior of tubes with circumferential cracks, and to predict the behavior of flawed and unflawed tubes under severe accident conditions. Task 4 is concerned with the cracking and failure of tubes that have been repaired by sleeving, and with a review of literature on this subject

  19. Effects of Support Structure Changes on Flow-induced Vibration Characteristics of Steam Generator Tubes

    International Nuclear Information System (INIS)

    Ryu, Ki Wahn; Park, Chi Yong; Rhee, Hui Nam

    2010-01-01

    Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability

  20. Nuclear material accounting: The next generation

    International Nuclear Information System (INIS)

    Kern, E.A.; McRae, L.P.; O'Callaghan, P.B.; Yearsley, D.

    1992-07-01

    The Westinghouse Hanford company (Westinghouse Hanford) and the Los Alamos National Laboratory (LANL) have undertaken a joint effort to develop a new generation material accounting system. The system will incorporate the latest advances in microcomputer hardware, software, and network technology. This system, the Local Area Network Material Accounting System (LANMAS), offers greater performance and functionality at a reduced overall cost. It also offers the possibility of establishing a standard among DOE and NRC facilities for material accounting. This report provides a discussion of this system

  1. Generation and growth rates of nonlinear distortions in a traveling wave tube.

    Science.gov (United States)

    Wöhlbier, John G; Dobson, Ian; Booske, And John H

    2002-11-01

    The structure of a steady state multifrequency model of a traveling wave tube amplifier is exploited to describe the generation of intermodulation frequencies and calculate their growth rates. The model describes the evolution of Fourier coefficients of circuit and electron beam quantities and has the form of differential equations with quadratic nonlinearities. Intermodulation frequencies are sequentially generated by the quadratic nonlinearities in a series solution of the differential equations. A formula for maximum intermodulation growth rates is derived and compared to simulation results.

  2. Generation and growth rates of nonlinear distortions in a traveling wave tube

    International Nuclear Information System (INIS)

    Woehlbier, John G.; Dobson, Ian; Booske, John H.

    2002-01-01

    The structure of a steady state multifrequency model of a traveling wave tube amplifier is exploited to describe the generation of intermodulation frequencies and calculate their growth rates. The model describes the evolution of Fourier coefficients of circuit and electron beam quantities and has the form of differential equations with quadratic nonlinearities. Intermodulation frequencies are sequentially generated by the quadratic nonlinearities in a series solution of the differential equations. A formula for maximum intermodulation growth rates is derived and compared to simulation results

  3. Friction generated ultrasound from geotechnical materials.

    Science.gov (United States)

    Tyler, T J; Hill, R; Lai, E

    2004-04-01

    Drilling is a process involved with product manufacturing and for civil engineers, site preparation. The usual requirement is for efficient material removal. In this study, the friction pair interaction generated by a drilling process provides ultrasound information related to parameters for the geotechnical material being drilled, where the drill bit has non-degrading ultrasonic characteristics and no essential requirement for material removal. This study has considered monitoring the ultrasonic signal generated by a drilling process, with a view to characterising the parameters of the geotechnical material being drilled and provides a novel method to identify or characterise ground structures. Drilling of geotechnical material systems, typically involve the interaction of a rotating probe and a granular composite medium. The applied load and angular velocity are measured to determine their relevance to the ultrasonic signal. Samples of granular materials have been graded into controlled grain size ranges. Attention has been focused on determining the effects on the ultrasound signal of grain size, bulk density and the water content of the granular material. A comparison between the various granular samples of the different grain sizes, density, water content and the associated ultrasonic signal has been done. The effect of each variable, and existing theory for these effects is commented upon. The broad aim of this research is to evaluate ultrasonic monitoring of drilling and assess its potential for real-time geotechnical ground condition monitoring applications and offer it as an alternative to existing methods.

  4. Regression analysis of pulsed eddy current signals for inspection of steam generator tube support structures

    International Nuclear Information System (INIS)

    Buck, J.; Underhill, P.R.; Mokros, S.G.; Morelli, J.; Krause, T.W.; Babbar, V.K.; Lepine, B.

    2015-01-01

    Nuclear steam generator (SG) support structure degradation and fouling can result in damage to SG tubes and loss of SG efficiency. Conventional eddy current technology is extensively used to detect cracks, frets at supports and other flaws, but has limited capabilities in the presence of multiple degradation modes or fouling. Pulsed eddy current (PEC) combined with principal components analysis (PCA) and multiple linear regression models was examined for the inspection of support structure degradation and SG tube off-centering with the goal of extending results to include additional degradation modes. (author)

  5. Application of probabilistic fracture mechanics to optimize the maintenance of PWR steam generator tubes

    International Nuclear Information System (INIS)

    Pitner, P.; Riffard, T.

    1993-09-01

    This paper describes the COMPROMIS code developed by Electricite de France (EDF) to optimize the tube bundle maintenance of steam generators (SG). The model, based on probabilistic fracture mechanics, makes it possible to quantify the influence of in-service inspections and maintenance work on the risk of an SG tube rupture, taking all significant parameters into account as random variables (initial defect size distribution, reliability of nondestructive detection and sizing, crack initiation and propagation, critical sizes, leak before risk of break, etc). (authors). 14 figs., 4 tabs., 12 refs

  6. Application of probabilistic fracture mechanics to estimate the risk of rupture of PWR steam generator tubes

    International Nuclear Information System (INIS)

    Pitner, P.; Riffard, T.; Granger, B.

    1992-01-01

    This paper describes the COMPROMIS code developed by Electricite de France (EDF) to optimize the tube bundle maintenance of steam generators. The model, based on probabilistic fracture mechanics, makes it possible to quantify the influence of in-service inspections and maintenance work on the risk of an SG tube rupture, taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive detection and sizing, crack initiation and propagation, critical sizes, leak before risk of break, etc.). (authors). 5 refs., 8 figs., 3 tabs

  7. Transmit-receive eddy current probes for defect detection and sizing in steam generator tubes

    International Nuclear Information System (INIS)

    Obrutsky, L.S.; Harasym, T.; Cecco, V.S.; Sullivan, S.P.

    1995-01-01

    Inspection of steam generator (SG) tubes in aging Nuclear Generating Stations is increasingly important. Defect detection and sizing, especially in defect prone areas such as the tubesheet, support plates and U-bend regions, are required to assess the fitness for service of the SG. Information about defect morphology is required to address operational integrity issues, i.e., risk of tube rupture, number of tubes at risk, consequential leakage. In 1991 Bruce Nuclear Generating Station A (BNGSA) was shut down because of SG tube failures due to circumferential stress-corrosion cracking (SCC) at the 40 o hot leg U-bend support plate. Eddy current inspection was performed using the best commercially available technology at that time which used pancake impedance coil probes. It was demonstrated that impedance probes were unable to detect any cracks except some that had propagated 100% through the tube wall. To address this problem AECL developed a new probe specifically designed for detecting circumferential SCC. This probe, denoted as C3, is a transmit-receive (T/R) multi-coil array probe with four or eight T/R units. Since the first field trial on 1000 tubes, it has been used routinely to inspect SG tubes at four CANDU plants for detection and sizing of SCC . The probe was able to detect cracks as shallow as 40% deep. Recently, a new T/R probe was designed to address the need of detecting circumferential and axial cracks simultaneously. The C5 incorporates T/R array probe technology with equal sensitivity to both type of cracks and a bobbin probe for historical comparison. T/R array probes with up to 24 T/R units have been used to inspect 12.9 to 22.2 mm diameter tubing for external as well as internal SCC in CANDU and PWR steam generators. This paper discusses the operating principles of T/R eddy current probes. It describes field experiences with detection and sizing of SCC with the C3 probe. Additionally, it explains the features of a C5 probe and its applications

  8. The PISC programme on defective steam generator tubes inspection. A status report

    International Nuclear Information System (INIS)

    Birac, C.; Comby, R.; Maciga, G.; Von Estorff, U.; Zanella, G.L.

    1994-06-01

    The general objective of the PISC Program (Programme for the Inspection of Steel Components) is to assess experimentally procedures and techniques in use for the in-service inspection of pressure components. The program is mainly a round robin test, the results of which are compared with real characteristics of the flaws obtained by destructive analysis. Materials tested are INCONEL 600 tubes, diameter 22.22 mm, wall thickness 1.27 mm. The technique applied is eddy current testing. The program of capability tests on loose tubes was started in 1990, the round robin tests ended in 1993. The preliminary results are presented. (R.P.). 8 refs., 9 figs., 4 tabs

  9. Stability of Balloon-Retention Gastrostomy Tubes with Different Concentrations of Contrast Material: In Vitro Study

    International Nuclear Information System (INIS)

    Lopera, Jorge E.; Alvarez, Alex; Trimmer, Clayton; Josephs, Shellie; Anderson, Matthew; Dolmatch, Bart

    2009-01-01

    The purpose of this study was to determine the performance of two balloon-retention-type gastrostomy tubes when the balloons are inflated with two types of contrast materials at different concentrations. Two commonly used balloon-retention-type tubes (MIC and Tri-Funnel) were inflated to the manufacturer's recommended volumes (4 and 20 cm 3 , respectively) with normal saline or normal saline plus different concentrations of contrast material. Five tubes of each brand were inflated with normal saline and 0%, 25%, 50%, 75%, and 100% contrast material dilutions, using either nonionic hyperosmolar contrast, or nonionic iso-osmolar contrast. The tubes were submerged in a glass basin containing a solution with a pH of 4. Every week the tubes were visually inspected to determine the integrity of the balloons, and the diameter of the balloons was measured with a caliper. The tests were repeated every week for a total of 12 weeks. The MIC balloons deflated slightly faster over time than the Tri-Funnel balloons. The Tri-Funnel balloons remained relatively stable over the study period for the different concentrations of contrast materials. The deflation rates of the MIC balloons were proportionally related to the concentration of saline and inversely related to the concentration of the contrast material. At high contrast material concentrations, solidification of the balloons was observed. In conclusion, this in vitro study confirms that the use of diluted amounts of nonionic contrast materials is safe for inflating the balloons of two types of balloon-retention feeding tubes. High concentrations of contrast could result in solidification of the balloons and should be avoided.

  10. Metallurgical problems in the exchange tube of a fast reactor steam generator

    International Nuclear Information System (INIS)

    Coriou, M.; Champeix, L.; Weisz, M.

    1980-10-01

    The design of the 1200 MWe Super Phenix power station steam generators is based on the following principles: once through helical tube exchangers which can be completely drained on the sodium side; the single wall exchange tubes are accessible to Foucault current testing during shutdowns. The authors explain the reasons for selecting the 800 Alloy for the exchange tubes. This choice was borne out by the results of several years of studies in the following areas: 6000 test hours with a 45 MWe model; corrosion test under stress in a water-steam and sodium plus caustic soda environment; resistance to creep and fatigue (effects of ageing and annealing, of the chemical compound); industrial feasibility, fabrication, utilization, bending, coiling, welding, testing. Concurrently, the EMl2 qualification finalizing has been pursued for the same application [fr

  11. The PISC programme on defective steam generator tubes inspection summary report

    International Nuclear Information System (INIS)

    Birac, C.; Comby, R.; Maciga, G.; Zanella, G.; Perez Prat, J.; Estorff, U. von

    1995-01-01

    The PISC III Actions are intended to extend the results and methodologies of the previous PISC exercises, i.e. the validation of the capabilities of the various examination techniques when used on real defects in real components under realistic conditions of inspection. The objective of this action is relatively close to that of the heavy structures programmes: the experimental evaluation of the performance of test procedures used for steam generator tubes in nuclear power plants during in-service or pre-service inspections. The exercise is a capability exercise consisting of Round Robin Tests on individual tubes including calibration, training and blind test tubes. In this paper the main conclusions from the RRT conducted in the framework of Action 5 will be presented and discussed. (author). 7 refs, 4 figs, 2 tabs

  12. The impact of NPP Krsko steam generator tube plugging on minimum DNBR at nominal conditions

    International Nuclear Information System (INIS)

    Lajtman, S.

    1996-01-01

    Typically, steam generator tube plugging (SGTP) both decreases the reactor coolant system (RCS) flow rate and the heat transfer surface area of the steam generator. At a constant thermal power and vessel outlet temperature, as tube plugging increases, the vessel average temperature, vessel inlet temperature and steam generator secondary side steam pressure decrease. This paper presents the analysis of impact of SGTP on Minimum Departure from Nucleate Boiling Ratio (MDNBR) at NPP Krsko (NEK), using the Improved Thermal Design Procedure (ITDP), WRB-1 correlation, and COBRA-III-C computer code. No credit was given to high plugging percentage region power reduction resulting from turbine volumetric flow limitations. MDNBR is found to be decreasing with increasing plugging, but not under the limiting values. (author)

  13. 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

  14. Coalescence model of two collinear cracks existing in steam generator tubes

    International Nuclear Information System (INIS)

    Moon, S.-I.; Chang, Y.-S.; Kim, Y.-J.; Park, Y.-W.; Song, M.-H.; Choi, Y.-H.; Lee, J.-H.

    2005-01-01

    The 40% of wall thickness criterion has been used as a plugging rule of steam generator tubes but it can be applicable just to a single-cracked tubes. In the previous studies preformed by the authors, a total of 10 local failure prediction models were introduced to estimate the coalescence load of two adjacent collinear through-wall cracks existing in thin plates, and the reaction force model and plastic zone contact model were selected as optimum models among them. The objective of this study is to verify the applicability of the proposed optimum local failure prediction models to the tubes with two collinear through-wall cracks. For this, a series of plastic collapse tests and finite element analyses were carried out using the tubes containing two collinear through-wall cracks. It has been shown that the proposed optimum failure models can predict the local failure behavior of two collinear through-wall cracks existing in tubes well. And a coalescence evaluation diagram was developed which can be used to determine whether the adjacent cracks detected by NED coalsece or not. (authors)

  15. Cascading pulse tubes on a large diaphragm pressure wave generator to increase liquefaction potential

    Science.gov (United States)

    Caughley, A.; Meier, J.; Nation, M.; Reynolds, H.; Boyle, C.; Tanchon, J.

    2017-12-01

    Fabrum Solutions, in collaboration with Absolut System and Callaghan Innovation, produce a range of large pulse tube cryocoolers based on metal diaphragm pressure wave generator technology (DPWG). The largest cryocooler consists of three in-line pulse tubes working in parallel on a 1000 cm3 swept volume DPWG. It has demonstrated 1280 W of refrigeration at 77 K, from 24 kW of input power and was subsequently incorporated into a liquefaction plant to produce liquid nitrogen for an industrial customer. The pulse tubes on the large cryocooler each produced 426 W of refrigeration at 77 K. However, pulse tubes can produce more refrigeration with higher efficiency at higher temperatures. This paper presents the results from experiments to increase overall liquefaction throughput by operating one or more pulse tubes at a higher temperature to pre-cool the incoming gas. The experiments showed that the effective cooling increased to 1500 W resulting in an increase in liquefaction rate from 13 to 16 l/hour.

  16. Continuous-wave radar to detect defects within heat exchangers and steam generator tubes.

    Energy Technology Data Exchange (ETDEWEB)

    Nassersharif, Bahram (New Mexico State University, Las Cruces, NM); Caffey, Thurlow Washburn Howell; Jedlicka, Russell P. (New Mexico State University, Las Cruces, NM); Garcia, Gabe V. (New Mexico State University, Las Cruces, NM); Rochau, Gary Eugene

    2003-01-01

    A major cause of failures in heat exchangers and steam generators in nuclear power plants is degradation of the tubes within them. The tube failure is often caused by the development of cracks that begin on the outer surface of the tube and propagate both inwards and laterally. A new technique was researched for detection of defects using a continuous-wave radar method within metal tubing. The experimental program resulted in a completed product development schedule and the design of an experimental apparatus for studying handling of the probe and data acquisition. These tests were completed as far as the prototypical probe performance allowed. The prototype probe design did not have sufficient sensitivity to detect a defect signal using the defined radar technique and did not allow successful completion of all of the project milestones. The best results from the prototype probe could not detect a tube defect using the radar principle. Though a more precision probe may be possible, the cost of design and construction was beyond the scope of the project. This report describes the probe development and the status of the design at the termination of the project.

  17. Experimental modeling of flow-induced vibration of multi-span U-tubes in a CANDU steam generator

    International Nuclear Information System (INIS)

    Mohany, A.; Feenstra, P.; Janzen, V.P.; Richard, R.

    2009-01-01

    Flow-induced vibration of the tubes in a nuclear steam generator is a concern for designers who are trying to increase the life span of these units. The dominant excitation mechanisms are fluidelastic instability and random turbulence excitation. The outermost U-bend region of the tubes is of greatest concern because the flow is almost perpendicular to the tube axis and the unsupported span is relatively long. The support system in this region must be well designed in order to minimize fretting wear of the tubes at the support locations. Much of the previous testing was conducted on straight single-span or cantilevered tubes in cross-flow. However, the dynamic response of steam generator multi-span U-tubes with clearance supports is expected to be different. Accurate modeling of the tube dynamics is important to properly simulate the dynamic interaction of the tube and supports. This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program was initiated at AECL to demonstrate that the tube support design for future CANDU steam generators will meet the stringent requirements associated with a 60 year design life. The main objective of the tests is to address the issue of in-plane and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with Anti-Vibration Bar (AVB) supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper. (author)

  18. Review of damages of nuclear power plants steam generator's tubes and way of detecting by using eddy current method

    International Nuclear Information System (INIS)

    Stanic, D.

    1996-01-01

    Steam generator tubing integrity is very important factor for reliable and safe operation of NPP. Several different types of tube degradation mechanisms were experienced in SG operation. To avoid possible tube rupture and primary-to-secondary leak, the EC examination of tubing should be performed. Different eddy current techniques may be used for detecting defects and theirs characterization. A comparison of data analysis results with pulled tube destructive metallography results can provide valuable insights in determining the capability of existing technology and provide guidance for procedure or technology improvements. (author)

  19. Evaluating Steam Generator Tubing Corrosion through Shutdown Nickel and Cobalt Releases

    International Nuclear Information System (INIS)

    Marks, Chuck; Little, Mike; Krull, Peter; Dennis Hussey; Kenny Epperson

    2012-09-01

    During power operation in PWRs, steam generator tubing corrodes. In PWRs with nickel alloy steam generator tubing this leads to the release of nickel into the coolant. While not structurally significant, this process leads to corrosion product deposition on the fuel surfaces that can threaten fuel integrity, provide a site for boron precipitation, and, through activation and subsequent release, lead to increased out-of-core radiation fields. During shutdown, decreases in temperature and pH and an increase in the oxidation potential lead to dissolution of some corrosion products from the core. This work evaluated the masses of corrosion products released during shutdown as a proxy for steam generator tubing corrosion rates. The masses were evaluated for trends with time (e.g., the number of cycles) and for the influence of design and operating features such as tubing manufacturer, plant design (e.g., three loop versus four loop), and operating chemistry program. This project utilized the EPRI PWR Chemistry Monitoring and Assessment database. Data from over 20 units, many over several cycles, were assessed. The focus was on corrosion product release from Alloy 690TT tubing and all data were from units that had replaced steam generators. Data were analyzed using models developed from corrosion rate test data reported in the literature with a heavy reliance on data from the EDF BOREAL testing. The most striking result of this analysis was a clear division between plants that exhibited corrosion with a falling rate (i.e., following an exponential decay as has been observed, for example, in the BOREAL testing) and those that showed a constant corrosion rate, sustained for many outages. This difference appears to be most closely correlated with the manufacturer of the tubing. Within the two distinct plant groups (decaying corrosion rate and constant corrosion rate), details of the trends were evaluated for correlation with zinc addition history, plant type, and operating

  20. Heavy ion irradiation effects in Zr excel alloy pressure tube material

    International Nuclear Information System (INIS)

    Idrees, Y.; Yao, Z.; Sattari, M.; Daymond, M.R.

    2012-01-01

    Zirconium Excel alloy (Zr-3.5wt.%Sn-0.8%Nb-0.8%Mo) is the candidate material for pressure tubes in the Generation-IV CANDU® Super Critical Water-cooled Reactor (SCWR) design. Changes in microstructure induced by neutron irradiation are known to have important consequences on the in-reactor deformation behavior. The in-situ ion irradiation technique has been employed to elucidate the irradiation damage in dual phase Zr-excel alloy (~60% hcp alpha and ~40% bcc beta). 1 MeV Kr ion irradiation experiments were conducted at different temperatures ranging from 100 o C-400 o C. Damage microstructures have been characterized by Transmission Electron Microscopy in both the alpha and beta phases at different temperatures after a maximum dose of 10 dpa. Several new observations including irradiation induced omega (ω) phase precipitation have been reported. The ω/β orientation relationship was determined by the detailed analysis of selected area diffraction patterns. In-situ irradiation provided an opportunity to observe the nucleation and growth of basal plane c-component loops. It has been shown that under Kr ion irradiation the c-loops start to nucleate and grow above a threshold dose, as has been observed for neutron irradiation. Furthermore, the role of temperature, material composition and pre-irradiation microstructure has been discussed in detail. (author)

  1. Risk assessment of severe accident-induced steam generator tube rupture

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    This report describes the basis, results, and related risk implications of an analysis performed by an ad hoc working group of the U.S. Nuclear Regulatory Commission (NRC) to assess the containment bypass potential attributable to steam generator tube rupture (SGTR) induced by severe accident conditions. The SGTR Severe Accident Working Group, comprised of staff members from the NRC`s Offices of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES), undertook the analysis beginning in December 1995 to support a proposed steam generator integrity rule. The work drew upon previous risk and thermal-hydraulic analyses of core damage sequences, with a focus on the Surry plant as a representative example. This analysis yielded new results, however, derived by predicting thermal-hydraulic conditions of selected severe accident scenarios using the SCDAP/RELAP5 computer code, flawed tube failure modeling, and tube failure probability estimates. These results, in terms of containment bypass probability, form the basis for the findings presented in this report. The representative calculation using Surry plant data indicates that some existing plants could be vulnerable to containment bypass resulting from tube failure during severe accidents. To specifically identify the population of plants that may pose a significant bypass risk would require more definitive analysis considering uncertainties in some assumptions and plant- and design-specific variables. 46 refs., 62 figs., 37 tabs.

  2. Risk assessment of severe accident-induced steam generator tube rupture

    International Nuclear Information System (INIS)

    1998-03-01

    This report describes the basis, results, and related risk implications of an analysis performed by an ad hoc working group of the U.S. Nuclear Regulatory Commission (NRC) to assess the containment bypass potential attributable to steam generator tube rupture (SGTR) induced by severe accident conditions. The SGTR Severe Accident Working Group, comprised of staff members from the NRC's Offices of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES), undertook the analysis beginning in December 1995 to support a proposed steam generator integrity rule. The work drew upon previous risk and thermal-hydraulic analyses of core damage sequences, with a focus on the Surry plant as a representative example. This analysis yielded new results, however, derived by predicting thermal-hydraulic conditions of selected severe accident scenarios using the SCDAP/RELAP5 computer code, flawed tube failure modeling, and tube failure probability estimates. These results, in terms of containment bypass probability, form the basis for the findings presented in this report. The representative calculation using Surry plant data indicates that some existing plants could be vulnerable to containment bypass resulting from tube failure during severe accidents. To specifically identify the population of plants that may pose a significant bypass risk would require more definitive analysis considering uncertainties in some assumptions and plant- and design-specific variables. 46 refs., 62 figs., 37 tabs

  3. 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)

  4. Wastage-resistant characteristics of 12Cr steel tube material. Small leak sodium-water reaction test

    International Nuclear Information System (INIS)

    Shimoyama, Kazuhito

    2004-03-01

    In the water leak accident of a steam generator designed for a sodium cooled reactor in the Feasibility Study, the localization of tube failure propagation by using an advanced water leak detector will be required from the viewpoints of the safety and economical efficiency of the plant. So far, the conventional knowledge and analytical tools have been used in the investigation and evaluation of water leak phenomenon; nevertheless, there was neither test data nor the study of quantitative evaluation on the corrosion behavior, so-called wastage-resistant characteristics, of 12Cr steel tube material in sodium-water reactions. Wastage tests for the 12Cr steel tube material were conducted in small water leaks by use of the Sodium-Water Reaction Test Rig (SWAT-1R), and the data of wastage rate were obtained in the parameter of water leak rate under the constant sodium temperature and distance between leak and target tubes. The test results lead to the following conclusions: (1) The wastage-resistibility of 12Cr steel is 1.6 times greater than that of 9Cr steel and is 2.7 times greater than that of 2.25Cr-1Mo steel. (2)The wastage-resistibility of 12Cr steel increases in smaller water leaks; especially in water leak rates of 1 g/sec or less, it is more excellent than that of SUS321 stainless steel used as Monju superheater tube material. (3) Based on the correlation of wastage rate for the 9Cr steel, the correlation for the 12Cr steel has been obtained to be used for the evaluation of tube failure propagation. As the correlation of wastage rate for the 12Cr steel is based on the correlation for the 9Cr steel, it gives enough conservatism in smaller water leaks. To serve in accurately evaluating the tube failure propagation in smaller water leaks, it is necessary to obtain new correlation of wastage rate for the 12Cr steel based on the data in the wide range of water leak rates. (author)

  5. Evaluation of materials for EPR power generation

    International Nuclear Information System (INIS)

    Mattas, R.F.; Stevens, H.C.; Misra, B.

    1979-01-01

    The blanket materials employed for heat generation in the Argonne Expermental Power Reactor (EPR) are evaluated. The EPR blanket consists of annealed Type 316 stainless steel sections cooled by pressurized water and Inconel 718 sections cooled by steam. The predicted lifetimes of the two different blanket sections are approximately 2 years of normal operation. The lifetime of annealed Type 316 stainless steel is limited by swelling considerations, while the lifetime of Inconel 718 is limited by ductility considerations

  6. Steam generators of Phenix: Measurement of the hydrogen concentration in sodium for detecting water leaks in the steam generator tubes

    International Nuclear Information System (INIS)

    Cambillard, E.; Lacroix, A.; Langlois, J.; Viala, J.

    1975-01-01

    The Phenix secondary circuits are provided with measurement systems of hydrogen concentration in sodium, that allow for the detection of possible water leaks in steam generators and the location of a faulty module. A measurement device consists of : a detector with nickel membranes of 0, 3 mm wall thickness, an ion pump with a 200 l/s flow rate, a quadrupole mass spectrometer and a calibrated hydrogen leak. The temperature correction is made automatically. The main tests carried out on the leak detection systems are reported. Since the first system operation (October 24, 1973), the measurements allowed us to obtain the hydrogen diffusion rates through the steam generator tube walls. (author)

  7. Phase Change Material Thermal Power Generator

    Science.gov (United States)

    Jones, Jack A.

    2013-01-01

    An innovative modification has been made to a previously patented design for the Phase Change Material (PCM) Thermal Generator, which works in water where ocean temperature alternatively melts wax in canisters, or allows the wax to re-solidify, causing high-pressure oil to flow through a hydraulic generator, thus creating electricity to charge a battery that powers the vehicle. In this modification, a similar thermal PCM device has been created that is heated and cooled by the air and solar radiation instead of using ocean temperature differences to change the PCM from solid to liquid. This innovation allows the device to use thermal energy to generate electricity on land, instead of just in the ocean.

  8. Improvements of defects sizing reliability in steam generators tubes through advanced NDT methods

    International Nuclear Information System (INIS)

    Benoist, B.; Gondard, C.

    1994-01-01

    As the population of nuclear power plants ages, new defects are appearing in steam generator tubes (stress corrosion cracking, corrosion pitting and intergranular corrosion). Utilities are requiring additional data to characterise defects after their detection, i.e. their depth, length and orientation, in order to optimise any tube plugging decision. Eddy current (E.C.) inspection is the reference Non Destructive Testing (NDT) method for SG's tubes inspection, due the long experience gained in the field and its rapidity. But in some cases, such as circumferential crack or multi-directional crack areas, its capabilities are limited (depth evaluation). Therefore, the application of ultrasonic (U.T) inspection as a complementary method can be helpful. We present in this paper new developments in the field of rotating probes testing and data processing which improve defect detection and sizing. Present on-site E.C inspections use axial bobbin coils running along all the length of the tubes. It allows a first and fast inspection for volumetric defects (deposits, wears, bumps,..). For the areas such as tube sheets and tube support plates, rotating probes have been developed in order to improve the circumferential resolution (detection of transversal defects). To emphasize our experience, inspection on the retired SG in DAMPIERRE has been undertaken with the rotating probe. Real E.C signals of primary wall stress corrosion cracking (PWSCC) will be presented. The detection procedure is based on visual examination of E.C. images (2D surface mappings or Lissajours figures). Image processing is used for automatic detection of defect signals. A first approach consists to use image enhancement techniques such as median filter. Sobel gradient, thresholding, binary morphology, to obtain a binary image leading to the defect areas. Results will be shown on artificial defects and on the DAMPIERRE signals. (Author) 10 refs

  9. Evaluation of the eddy-current method of inspecting steam generator tubing

    International Nuclear Information System (INIS)

    Flora, J.H.; Brown, S.D.; Weeks, J.R.

    1976-01-01

    The objective of this project has been to evaluate the eddy-current method of inspecting steam generator tubing by conducting a series of laboratory experiments with conventional eddy-current equipment. The experiments have involved obtaining eddy-current measurements on samples of 7/8-inch OD Inconel-600 tubing provided by the Westinghouse Nuclear Energy Systems Division. A variety of machined defects and some chemically induced flaws, such as stress corrosion cracks were fabricated in the tubing. Statistical evaluation of the data was employed to estimate the error encountered in measuring corrosion defects of various depths. It appears that the eddy-current technique can provide a reasonable measure of defect depth under certain conditions. On the other hand, the evaluation indicates that it is difficult to determine the depth of certain types of flaws with reliability and precision. Furthermore, although some defects as shallow as 10 percent of the tube wall could be detected, it was not possible to detect other types of flaws that were less than 40 percent deep even when the tube supports were not near the defects. The difficulty in detecting small volume flaws is attributed to low signal-to-noise ratio. Noise is a result of unwanted signals from test variables, such as wobble and variations in tube properties. The error in measurement of certain types of larger defects is associatedin part with test variables and also with the effects that the geometry of the defect has on the eddy-current signal patterns. The distortions in signal patterns caused by gradual wastage type defects and the poor reproducibility of signal patterns obtained from notches that represent stress corrosion cracks are described. Some developments that will rectify these detection and depth measurement problems are discussed

  10. Measurement with corrugated tubes of early-age autogenous shrinkage of cement-based material

    DEFF Research Database (Denmark)

    Tian, Qian; Jensen, Ole Mejlhede

    2009-01-01

    The use of a special corrugated mould enables transformation of volume strain into horizontal, linear strain measurement in the fluid stage. This allows continuous measurement of the autogenous shrinkage of cement-based materials since casting, and also effectively eliminates unwanted influence...... on the measuring results from gravity, temperature variation and mould restraint. In this paper the principle of the corrugated tube measurement is described. A systematic study was carried out on the influence on the measuring results of the material properties, size effects and encapsulated air in the corrugated...... tube. The experimental results show that there is a minor influence on the measuring results of the stiffness and size of the plastic tube as well as of the encapsulated air. However, the influence decreases with the hardening process and becomes negligible a few hours after final set....

  11. Free vibration analysis of a steam generator tube bundle with and without lateral support

    International Nuclear Information System (INIS)

    King, D.M.

    1979-04-01

    The vibrational modes and frequency characteristics of a pressurized water reactor (PWR) steam generator tube bundle assembly with and without lateral support in a fluid environment are analyzed. The idealized half-model was constructed using the SAP-IV finite element code. Free vibration analyses were performed for an in-air case and a submerged in-water case, each with different constraint conditions at steam generator tube bundle assembly support plates 10 and 11. These constraint conditions included having both support plates free, having both support plates fixed, and having support plate 11 free while support plate 10 was fixed. It was found that as the support plate constraints were removed, the frequency range for each case increased significantly

  12. Optical Fiber Demodulation System with High Performance for Assessing Fretting Damage of Steam Generator Tubes.

    Science.gov (United States)

    Huang, Peijian; Wang, Ning; Li, Junying; Zhu, Yong; Zhang, Jie; Xi, Zhide

    2018-01-12

    In order to access the fretting damage of the steam generator tube (SGT), a fast fiber Fabry-Perot (F-P) non-scanning correlation demodulation system based on a super luminescent light emitting diode (SLED) was performed. By demodulating the light signal coming out from the F-P force sensor, the radial collision force between the SGT and the tube support plate (TSP) was interrogated. For higher demodulation accuracy, the effects of the center wavelength, bandwidth, and spectrum noise of SLED were discussed in detail. Specially, a piezoelectric ceramic transducer (PZT) modulation method was developed to get rid of the interference of mode coupling induced by different types of fiber optics in the demodulation system. The reflectivity of optical wedge and F-P sensor was optimized. Finally, the demodulation system worked well in a 1:1 steam generator test loop and successfully demodulated a force signal of 32 N with a collision time of 2 ms.

  13. Application of numerical analysis techniques to eddy current testing for steam generator tubes

    International Nuclear Information System (INIS)

    Morimoto, Kazuo; Satake, Koji; Araki, Yasui; Morimura, Koichi; Tanaka, Michio; Shimizu, Naoya; Iwahashi, Yoichi

    1994-01-01

    This paper describes the application of numerical analysis to eddy current testing (ECT) for steam generator tubes. A symmetrical and three-dimensional sinusoidal steady state eddy current analysis code was developed. This code is formulated by future element method-boundary element method coupling techniques, in order not to regenerate the mesh data in the tube domain at every movement of the probe. The calculations were carried out under various conditions including those for various probe types, defect orientations and so on. Compared with the experimental data, it was shown that it is feasible to apply this code to actual use. Furthermore, we have developed a total eddy current analysis system which consists of an ECT calculation code, an automatic mesh generator for analysis, a database and display software for calculated results. ((orig.))

  14. Chemical preventive remedies for steam generators fouling and tube support plate blockages

    International Nuclear Information System (INIS)

    Alves Vieira, M.; Mayos, M.; Coquio, N.; Fourcroy, H.; Battesti, P.

    2010-01-01

    In 2006, EDF identified on several PWR units broached hole blockage on the upper Steam Generator (SG) Tube Support Plates (TSP). TSP blockage often occurs in association with secondary fouling. The units with copper alloys materials are more affected due the applied low pH 25 o C (9.20) all volatile treatment (AVT). Carbon steels materials are less protected against flow accelerated corrosion (FAC) and therefore more corrosion products enter the SGs through the final feed water (FFW). In parallel of chemical cleanings to remove oxides deposits in SGs, EDF has defined a strategy to improve operating conditions. It mainly relies on the removal of copper alloys materials to implement a high pH AVT (9.60) as a preventive remedy. However for some plants, copper alloys removal is not straightforward due to environmental constraints. EDF must indeed manage the implementation of a biocide treatment needed in closed loop cooling systems (as copper has a bacteriostatic effect on micro-organisms) and more generally must comply with discharge authorisations for chemical conditioning reagents or biocide reagent. An alternative conditioning was tested on the Dampierre 4 unit in 2007/2008 during 6 months to assess if operating at 9.40 was acceptable regarding the impacts on copper alloys materials. The perspective would be to implement it in the units where no biocide treatment can be applied on a short term. In parallel, other chemical conditionings or additives will be implemented or tested. First of all, EDF will carry out a trial test with APA in order to assess its efficiency on the removal of oxides deposits through SG blowdown. On the other hand, AVT with high pH ethanolamine (ETA) will be implemented as an alternative of ammonia and morpholine conditioning on some chosen plants. Ethanolamine is selected as a way to mitigate FAC kinetics in two-phase flow areas (reheaters or moisture heater separator) or to limit liquid releases. This paper provides the lessons of the

  15. Development Study of Cartridge/Crucible Tube Materials

    Science.gov (United States)

    McKechnie, Timothy N.; ODell, Scott J.

    1998-01-01

    The limitations of traditional alloys and the desire for improved performance for components is driving the increased utilization of refractory metals in tile space industry. From advanced propulsion systems to high temperature furnace components for microgravity processing, refractory metals are being used for their high melting temperatures and inherent chemical stability. Techniques have been developed to produce near net shape refractory metal components utilizing vacuum plasma spraying. Material utilization is very high, and laborious machining can be avoided. As-spray formed components have been tested and found to perform adequately. However, increased mechanical and thermal properties are needed. To improve these properties, post processing thermal treatments such as hydrogen sintering and vacuum annealing have been performed. Components formed from alloys of tungsten, rhenium, tantalum, niobium, and molybdenum are discussed and a metallurgical analyses detailing the results are presented. A qualitative comparison of mechanical properties is also included.

  16. Acoustic detection for water/steam leak from a tube of LMFBR steam generator

    International Nuclear Information System (INIS)

    Sonoda, Masataka; Shindo, Yoshihisa

    1989-01-01

    Acoustic leak detector is useful for detecting more quickly intermediate leak than the existing hydrogen detector and is available for identification of leak location on the accident of water/steam leak from a tube of LMFBR steam generator. This paper presents the overview of HALD (High frequency Acoustics Leak Detection) system, which is more sensitive for leak detection and lower cost of equipment for identification of leak location than a low frequency type detector. (author)

  17. 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)

  18. Processing of cartography from steam generator tubes using eddy current testing with an absolute coil

    International Nuclear Information System (INIS)

    Attaoui, P.; Benoist, B.; Besnard, R.; Sollier, T.; Gaillard, P.; Lengelle, R.

    1992-01-01

    This paper deals with the processing of electromagnetic cartography from steam generator tubes testing. These images are disturbed by background noise due to probe lift-off changes and by the rolling transition zone signal. Procedures which allow to obtain a flat cartography will be presented in part one. Then, using mathematical morphology tools on the cartography, expose the first results dealing with sizing and orientation of the defects. (author)

  19. Probabilistic evaluation of multiple failures for steam generators tubes by common mode

    International Nuclear Information System (INIS)

    Bloch, M.; Pierrey, J.L.; Dussarte, D.

    1987-11-01

    The reactor safety can be affected when systems or components are subject to phenomena conducting at a wear nontake in account in the conception. This paper presents a methodology which takes in account the non simultaneous failures resulting of this situation. To illustrate this purpose, we give an evaluation of risk of multiple failures for steam generators tubes by common mode (stress corrosion) when the reactor is in normal operation [fr

  20. Apparatus for steam generator tube wrapper spacer and support block removal

    International Nuclear Information System (INIS)

    Calhoun, G.L.; Cassette, A.J.

    1982-01-01

    A cutting torch is described that may be used to cut through the spacers and the support blocks in the annulus between the outer shell of a steam generator and a tube bundle wrapper. The torch is supported by a multi-section column made up of a number of interlocked separable aluminum channel sections carried by a motorized carriage movable vertically along a rack and guide member clamped to the wrapper

  1. Structural integrity assessments of steam generator tubes using the FAD methodology

    Energy Technology Data Exchange (ETDEWEB)

    Bergant, Marcos A., E-mail: marcos.bergant@cab.cnea.gov.ar [Gerencia CAREM, Centro Atómico Bariloche (CNEA), Av. Bustillo 9500, San Carlos de Bariloche 8400 (Argentina); Yawny, Alejandro A., E-mail: yawny@cab.cnea.gov.ar [División Física de Metales, Centro Atómico Bariloche (CNEA)/CONICET, Av. Bustillo 9500, San Carlos de Bariloche 8400 (Argentina); Perez Ipiña, Juan E., E-mail: juan.perezipina@fain.uncoma.edu.ar [Grupo Mecánica de Fractura, Universidad Nacional del Comahue/CONICET, Buenos Aires 1400, Neuquén 8300 (Argentina)

    2015-12-15

    Highlights: • The Failure Assessment Diagram (FAD) is used to assess cracked steam generator tubes. • Typical loading conditions and reported tensile and fracture properties are used. • The FAD is capable to predict the failure mode for different cracks and loads. • The FAD can be used to reduce the conservatism of the current plugging criteria. • Appropriate tensile and fracture properties at operating conditions are required. - Abstract: Steam generator tubes (SGTs) represents up to 60% of the total primary pressure retaining boundary area of a nuclear power plant. They have been found susceptible to diverse degradation mechanisms during service. Due to the significance of a SGT failure on the plant safe operation, nuclear regulatory authorities have established tube plugging or repairing criteria which are based on the defect depth. The widespreadly used “40% criterion” proposed in the 70s is an example whose use is still recommended in the last editions of the ASME Boiler and Pressure Vessel Code. In the present work, an alternative, more realistic and less conservative methodology for SGT integrity evaluation is proposed. It is based on the Failure Assessment Diagram (FAD) and takes advantage of the recent developments in non-destructive techniques which allow a more comprehensive characterization of tube defects, i.e., depth, length, orientation and type. The proposed approach has been applied to: the study of the influence of primary and secondary stresses on tube integrity; the prediction of failure mode (i.e., ductile fracture or plastic collapse) of defective SGTs for varied crack geometries and loading conditions; the analysis of the sensibility of tensile and fracture properties with temperature. The potentiality of the FAD as a comprehensive methodology for predicting the failure loads and failure modes of flawed SGTs is highlighted.

  2. The effect of the removal of steam generator tube ID deposits of heat transfer

    International Nuclear Information System (INIS)

    Klimas, S.J.; Miller, D.G.; Semmler, J.; Turner, C.W.

    1998-12-01

    The thermal resistance of boiler primary-side tube deposits from the Gentilly-2 Nuclear Generating Station (Hydro-Quebec) was evaluated by an experimental comparison of the heat-transfer rates between fouled samples and identical, factory-new, 'clean' tubing. The deposits were subsequently removed using either a chemical decontamination process (CAN-DEREM Plus) or a mechanical cleaning process (Siemens SIVABLAST) in two stages. After each removal, the thermal resistance of the remaining deposit was remeasured. The 90- to 150-μm-thick deposits on the inside diameter of steam generator cold-leg tubes were found to pose significant resistance to heat transfer (0.05 to 0.06 m 2 ·K/kW at 210 degrees C). However, the 10- to 30-μm-thick dense layers remaining on the tubes after the decontamination were found to have no measurable effect on the heat transfer. The thin, 2-μm tube deposit on the steam generator hot leg slightly enhanced heat transfer. The measured thermal resistance results in a calculated thermal conductivity of 1.5 W/m·K for the 90-μm-thick deposit. The 150-μm-thick deposits were found to consist of two layers: an outer surface layer having an average porosity of 50% and a conductivity of 2.3 W/m·K, and an inner layer having an average porosity of 5% and a conductivity of >3.0 W/m·K. The previous best estimate of the thermal conductivity was 1.4 W/m.K for the porous magnetite deposits that had formed with a thickness <90 μm on the primary side of nuclear steam generators. This work confirms this number, but also demonstrates that it is applicable only for porous, unconsolidated deposits. The conductivity increases for thicker deposits because of increasing deposit consolidation, particularly at the innermost layer adjacent to the tube metal. (author)

  3. The effect of the removal of steam generator tube ID deposits of heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Klimas, S.J.; Miller, D.G.; Semmler, J.; Turner, C.W

    1998-12-01

    The thermal resistance of boiler primary-side tube deposits from the Gentilly-2 Nuclear Generating Station (Hydro-Quebec) was evaluated by an experimental comparison of the heat-transfer rates between fouled samples and identical, factory-new, 'clean' tubing. The deposits were subsequently removed using either a chemical decontamination process (CAN-DEREM Plus) or a mechanical cleaning process (Siemens SIVABLAST) in two stages. After each removal, the thermal resistance of the remaining deposit was remeasured. The 90- to 150-{mu}m-thick deposits on the inside diameter of steam generator cold-leg tubes were found to pose significant resistance to heat transfer (0.05 to 0.06 m{sup 2}{center_dot}K/kW at 210 degrees C). However, the 10- to 30-{mu}m-thick dense layers remaining on the tubes after the decontamination were found to have no measurable effect on the heat transfer. The thin, 2-{mu}m tube deposit on the steam generator hot leg slightly enhanced heat transfer. The measured thermal resistance results in a calculated thermal conductivity of 1.5 W/m{center_dot}K for the 90-{mu}m-thick deposit. The 150-{mu}m-thick deposits were found to consist of two layers: an outer surface layer having an average porosity of 50% and a conductivity of 2.3 W/m{center_dot}K, and an inner layer having an average porosity of 5% and a conductivity of >3.0 W/m{center_dot}K. The previous best estimate of the thermal conductivity was 1.4 W/m.K for the porous magnetite deposits that had formed with a thickness <90 {mu}m on the primary side of nuclear steam generators. This work confirms this number, but also demonstrates that it is applicable only for porous, unconsolidated deposits. The conductivity increases for thicker deposits because of increasing deposit consolidation, particularly at the innermost layer adjacent to the tube metal. (author)

  4. Resistance of Incoloy 800 steam generator tube to pitting corrosion in PWR secondary water at 250°C

    Energy Technology Data Exchange (ETDEWEB)

    Schvartzman, Mônica M.A.M. [Pontifícia Universidade Católica de Minas Gerais (PUC-Minas), Belo Horizonte, MG (Brazil); Albuquerque, Adriana Silva de; Esteves, Luiza; Rabello, Emerson G.; Mansur, Fábio Abud, E-mail: monicacdtn@gmail.com, E-mail: asa@cdtn.br, E-mail: luiza.esteves@cdtn.br, E-mail: egr@cdtn.br, E-mail: fametalurgica@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The steam generator (SG) is one of the main components of a PWR, so the performance of this type of nuclear power plant depends to a large extent on the trouble-free operation of SGs. Its degradation significantly affects the overall plant performance. Alloy 800NG (Incoloy® 800) is a nickel-iron-chromium alloy used for steam generator tubes in PWRs due to their high strength, good workability and resistance to corrosion. This behavior is attributed to the protective oxide film formed on the metal surface by contact with the high temperature pressurized water. However, chloride is one of major SG impurities that cause the breakdown of the passive film and initiate localized corrosion in passive metals as Alloy 800NG. The aim of this study is to provide information about the pitting corrosion behavior of the Incoloy® 800 steam generator tube under normal secondary circuit parameters (250 deg C and 5 MPa) and abnormal conditions of operation (presence of chloride ions in the secondary water). For this, optical microscopy, XRD and EDS analysis and electrochemical tests have been carried out under simulated PWR secondary water operating conditions. The susceptibility to pitting corrosion was evaluated using electrochemical tests and the oxide layer formed on material was examined by means of scanning electron microscopy (SEM) and energy dispersive X-ray spectrometer (EDS) analyses. (author)

  5. Boreside rotating ultrasonic tester for wastage determination of LMFBR-type steam generator tubes

    International Nuclear Information System (INIS)

    Neely, H.H.; Renger, H.L.

    1979-01-01

    Large sodium-water reaction (SWR) leak tests are being run in near-prototypic steam generators at prototypic plant conditions of the Liquid Metal Fast Breeder Reactor (LMFBR). These tests simulate various types of steam tube failure at predetermined locations. A SWR results in a highly energetic-exothermic-caustic reaction which erodes neighboring tubes. A boreside-rotating ultrasonic inspection device was developed to measure wall thickness and inside diameter of the 2/one quarter/Cr-1 Mo, 10.1 mm I.D. steam tubes. Rotation of the UT beam yields a complimentary scan of the full tube in a single pass. The UT system was designed with a 15 MHz transducer in pulse-echo compression-wave mode at a pulse rate of 10,000/second. The UT beam is rotated at 20 r/s on a 1.27 mm pitch. System outputs are diameter, wall thickness, attitude, and axial position. Measurements are processed, then fed to a CRT and computer for later retrieval and plotting

  6. Impurities incorporation into magnetite scale formed on simulated steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, K.; Yamaguchi, K.; Koike, M. [Kyushu Electric Power Co., Inc. (Japan); Kawamura, H.; Hirano, H. [Central Research Inst. of Electric Power Industry (Japan); Yamada, Y.; Nakamura, T. [The Kansai Electric Power Co., Inc. (Japan)

    2002-07-01

    From a viewpoint of ensuring the integrity of steam generators (SGs) tubing in PWR plants, the research was made into how impurities in the secondary coolant are incorporated into magnetite (Fe{sub 3}O{sub 4}) scale formed on the tube in a laboratory test. We experimented with a method to form Fe{sub 3}O{sub 4} scale on a tube under a boiling heat transfer condition in the laboratory test, simulating the conditions of SG in the actual PWR plants. Based on the scale formation method, we investigated the incorporation of sulfur (S) into the scale. S is known as the most common impurity solved in the secondary coolant and a dominant factor in making heat transfer crevice environment acidic. The effects of sodium (Na) and silicon (Si), solved in test solution with S, on the S incorporation into scale were also investigated. The test resulted in a double-layered scale being formed on the tube surface, with the outer scale being porous and the inner scale dense. It was revealed that the S incorporation into scales was affected by the S concentration in the solution and existence of other impurities, such as Na and Si. (authors)

  7. Impurities incorporation into magnetite scale formed on simulated steam generator tubing

    International Nuclear Information System (INIS)

    Takahashi, K.; Yamaguchi, K.; Koike, M.; Kawamura, H.; Hirano, H.; Yamada, Y.; Nakamura, T.

    2002-01-01

    From a viewpoint of ensuring the integrity of steam generators (SGs) tubing in PWR plants, the research was made into how impurities in the secondary coolant are incorporated into magnetite (Fe 3 O 4 ) scale formed on the tube in a laboratory test. We experimented with a method to form Fe 3 O 4 scale on a tube under a boiling heat transfer condition in the laboratory test, simulating the conditions of SG in the actual PWR plants. Based on the scale formation method, we investigated the incorporation of sulfur (S) into the scale. S is known as the most common impurity solved in the secondary coolant and a dominant factor in making heat transfer crevice environment acidic. The effects of sodium (Na) and silicon (Si), solved in test solution with S, on the S incorporation into scale were also investigated. The test resulted in a double-layered scale being formed on the tube surface, with the outer scale being porous and the inner scale dense. It was revealed that the S incorporation into scales was affected by the S concentration in the solution and existence of other impurities, such as Na and Si. (authors)

  8. Gas Generation from Actinide Oxide Materials

    International Nuclear Information System (INIS)

    Bailey, George; Bluhm, Elizabeth; Lyman, John; Mason, Richard; Paffett, Mark; Polansky, Gary; Roberson, G. D.; Sherman, Martin; Veirs, Kirk; Worl, Laura

    2000-01-01

    This document captures relevant work performed in support of stabilization, packaging, and long term storage of plutonium metals and oxides. It concentrates on the issue of gas generation with specific emphasis on gas pressure and composition. Even more specifically, it summarizes the basis for asserting that materials loaded into a 3013 container according to the requirements of the 3013 Standard (DOE-STD-3013-2000) cannot exceed the container design pressure within the time frames or environmental conditions of either storage or transportation. Presently, materials stabilized and packaged according to the 3013 Standard are to be transported in certified packages (the certification process for the 9975 and the SAFKEG has yet to be completed) that do not rely on the containment capabilities of the 3013 container. Even though no reliance is placed on that container, this document shows that it is highly likely that the containment function will be maintained not only in storage but also during transportation, including hypothetical accident conditions. Further, this document, by summarizing materials-related data on gas generation, can point those involved in preparing Safety Analysis Reports for Packages (SARPs) to additional information needed to assess the ability of the primary containment vessel to contain the contents and any reaction products that might reasonably be produced by the contents

  9. Gas Generation from Actinide Oxide Materials

    Energy Technology Data Exchange (ETDEWEB)

    George Bailey; Elizabeth Bluhm; John Lyman; Richard Mason; Mark Paffett; Gary Polansky; G. D. Roberson; Martin Sherman; Kirk Veirs; Laura Worl

    2000-12-01

    This document captures relevant work performed in support of stabilization, packaging, and long term storage of plutonium metals and oxides. It concentrates on the issue of gas generation with specific emphasis on gas pressure and composition. Even more specifically, it summarizes the basis for asserting that materials loaded into a 3013 container according to the requirements of the 3013 Standard (DOE-STD-3013-2000) cannot exceed the container design pressure within the time frames or environmental conditions of either storage or transportation. Presently, materials stabilized and packaged according to the 3013 Standard are to be transported in certified packages (the certification process for the 9975 and the SAFKEG has yet to be completed) that do not rely on the containment capabilities of the 3013 container. Even though no reliance is placed on that container, this document shows that it is highly likely that the containment function will be maintained not only in storage but also during transportation, including hypothetical accident conditions. Further, this document, by summarizing materials-related data on gas generation, can point those involved in preparing Safety Analysis Reports for Packages (SARPs) to additional information needed to assess the ability of the primary containment vessel to contain the contents and any reaction products that might reasonably be produced by the contents.

  10. The effect of the number of condensed phases modeled on aerosol behavior during an induced steam generator tube rupture sequence

    International Nuclear Information System (INIS)

    Bixler, N.E.; Schaperow, J.H.

    1998-06-01

    VICTORIA is a mechanistic computer code designed to analyze fission product behavior within a nuclear reactor coolant system (RCS) during a severe accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS. A recently completed independent peer review of VICTORIA, while confirming the overall adequacy of the code, recommended a number of modeling improvements. One of these recommendations, to model three rather than a single condensed phase, is the focus of the work reported here. The recommendation has been implemented as an option so that either a single or three condensed phases can be treated. Both options have been employed in the study of fission product behavior during an induced steam generator tube rupture sequence. Differences in deposition patterns and mechanisms predicted using these two options are discussed

  11. CFD analysis of fin tube heat exchanger with a pair of delta winglet vortex generators

    International Nuclear Information System (INIS)

    Hwang, Seong Won; Kim, Dong Hwan; Min, June Kee; Jeong, Ji Hwan

    2012-01-01

    Among tubular heat exchangers, fin tube types are the most widely used in refrigeration and air-conditioning equipment. Efforts to enhance the performance of these heat exchangers included variations in the fin shape from a plain fin to a slit and louver type. In the context of heat transfer augmentation, the performance of vortex generators has also been investigated. Delta winglet vortex generators have recently attracted research interest, partly due to experimental data showing that their addition to fin-tube heat exchangers considerably reduces pressure loss at heat transfer capacity of nearly the same level. The efficiency of the delta winglet vortex generators widely varies depending on their size and shape, as well as the locations where they are implemented. In this paper, the flow field around delta winglet vortex generators in a common flow up arrangement was analyzed in terms of flow characteristics and heat transfer using computational fluid dynamics methods. Flow mixing due to vortices and delayed separation due to acceleration influence the overall fin performance. The fin with delta winglet vortex generators exhibited a pressure loss lower than that of a plain fin, and the heat transfer performance was enhanced at high air velocity or Reynolds number

  12. Five Tubes Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

    2010-12-01

    In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-CL-02 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Compared to the case of a single U-tube rupture test, opening frequency of the MSSVs in the intact steam generator (SG-2) was highly reduced after 500 seconds in the present SGTR-CL-02 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code

  13. Local chemical and thermal-hydraulic analysis of U-tube steam generators

    International Nuclear Information System (INIS)

    Lee, J.Y.; No, H.C.

    1990-01-01

    In order to know how pH distribution affects corrosion in a U-tube steam generator, a study of the combination of water chemistry and thermal-hydraulic conditions is suggested. A two-fluid (unequal velocity and unequal temperature) formulation is proposed to describe the convective transport of volatile species in each phase, and a spherical bubble model is developed on the basis of the penetration theory to describe the interfacial mass transfer. The thermal-hydraulic local conditions are obtained by the U-tube steam generator design analysis code FAUST which is based on the three-dimensional two-fluid model. The results of the present study are compared with dynamic equilibrium model calculations. This study shows that, in contrast with dynamic equilibrium calculations, the pH is lower in the cold-leg side than in the hot-leg side because of liquid recirculation. Just above the tube sheet, however, the lower void fraction in this region than that in the hot-leg region results in higher pH, which agrees with the prediction of the dynamic equilibrium model. (orig.)

  14. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    Science.gov (United States)

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    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.

  15. 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.

  16. Performance demonstration tests for eddy current inspection of steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

    Kurtz, R.J.; Heasler, P.G.; Anderson, C.M.

    1996-05-01

    This report describes the methodology and results for development of performance demonstration tests for eddy current (ET) inspection of steam generator tubes. Statistical test design principles were used to develop the performance demonstration tests. Thresholds on ET system inspection performance were selected to ensure that field inspection systems would have a high probability of detecting and and correctly sizing tube degradation. The technical basis for the ET system performance thresholds is presented in detail. Statistical test design calculations for probability of detection and flaw sizing tests are described. A recommended performance demonstration test based on the design calculations is presented. A computer program for grading the probability of detection portion of the performance demonstration test is given.

  17. Performance demonstration tests for eddy current inspection of steam generator tubing

    International Nuclear Information System (INIS)

    Kurtz, R.J.; Heasler, P.G.; Anderson, C.M.

    1996-05-01

    This report describes the methodology and results for development of performance demonstration tests for eddy current (ET) inspection of steam generator tubes. Statistical test design principles were used to develop the performance demonstration tests. Thresholds on ET system inspection performance were selected to ensure that field inspection systems would have a high probability of detecting and and correctly sizing tube degradation. The technical basis for the ET system performance thresholds is presented in detail. Statistical test design calculations for probability of detection and flaw sizing tests are described. A recommended performance demonstration test based on the design calculations is presented. A computer program for grading the probability of detection portion of the performance demonstration test is given

  18. Fin-and-tube heat exchanger enhancement with a combined herringbone and vortex generator design

    DEFF Research Database (Denmark)

    Välikangas, Turo; Singh, Shobhana; Sørensen, Kim

    2018-01-01

    Vortex generators (VGs) are the most commonly investigated enhancement methods in the field of improved heat exchangers. The aim of present work is to study the effect of VGs in a fin-and-tube heat exchanger (FTHE) with herringbone fin shape. The delta winglet VG design with length (s) and height...... (H) is selected based on previous studies. The investigated VG design is simple and considered realistic from the manufacturing point of view. The combined enhancement with herringbone fin and the VG is evaluated by simulating the conjugate heat transfer and the air flow. The structured mesh...... transfer in the herringbone fin but also decrease the pressure drop. The highest and longest investigated VG design is found to perform the best because of its ability to delay the flow detachment from the tube, to feed high kinetic energy flow to the recirculation zone and to create longitudinal vortices...

  19. Analysis of the influence of steam generator tube plugging on the large break loss of coolant accident in NPP Krsko

    International Nuclear Information System (INIS)

    Bizjak, S.; Stritar, A.

    1987-01-01

    The preliminary analysis of the influence of steam generator tube plugging to the large break LOCA behaviour of the NPP Krsko was performed. If 10% of the tubes are plugged, the peak cladding temperature reached is 37 K higher than the temperature reached after LOCA if no tubes were plugged. The decrease of the maximum peaking factor from 2.34 to 2.25 would compensate the influence of 10% plugged tubes. The analysis was not fully in compliance with the requirements of the conservative methodology. (author)

  20. Steam generator tube support plate degradation in French plants: maintenance strategy

    International Nuclear Information System (INIS)

    Gauchet, J.-P.; Gillet, N.; Stindel, M.

    1998-01-01

    This paper reports on the degradations of Steam Generator (SG) Tube Support Plates (TSPs) observed in French plants and the maintenance strategy adopted to continue operating the plant without any decrease of the required safety level. Only drilled carbon steel TSPs of early SGs are affected. Except the particular damage of the TSP8 of FESSENHEIM 2 caused by chemical cleaning procedures implemented in 1992, two main problems were observed almost exclusively on the upper TSP: Ligaments ruptured near the aseismic block located at 215 degrees. This degradation is perfectly detectable by bobbin coil inspection. It occurs very early in the life of the SG as can be seen from the records of previous inspections and no evolution of the signals was observed. This damage can be detected for 51M model SGs on several sites; Wastage of the ligaments resulting in enlargement of flow holes with in some cases complete consumption of a ligament. This damage was only observed for SGs of at GRAVELINES. This damage evolved cycle after cycle. Detailed studies were performed to analyze tubing behavior when a tube is not supported by the upper TSP because of missing ligaments. These studies evaluated the risk of vibratory instability, the behavior of both the TSP and the tubing in case of a seismic event or a LOCA and finally the behavior of the TSP in case of a Steam Line Break. Concerning vibratory instability it was possible to define zones where stability could not be demonstrated. Dampine, cables and sentinel plugs were then used when necessary to eliminate the risk of Steam Generator Tube Rupture (SGTR). For accidental conditions, it could be shown that no unacceptable damage occurs and that the core cooling function of the SG is always maintained if some tubes are plugged. From this analysis, It was possible to define the inspection programs for the different plants taking into account the specific situation of each plant regarding the damages detected. These programs include

  1. A quality assessment of cardiac auscultation material on YouTube.

    Science.gov (United States)

    Camm, Christian F; Sunderland, Nicholas; Camm, A John

    2013-02-01

    YouTube is a highly utilized Web site that contains a large amount of medical educational material. Although some studies have assessed the education material contained on the Web site, little analysis of cardiology content has been made. This study aimed to assess the quality of videos relating to heart sounds and murmurs contained on YouTube. We hypothesized that the quality of video files purporting to provide education on heart auscultation would be highly variable. Videos were searched for using the terms "heart sounds," "heart murmur," and "heart auscultation." A built-in educational filter was employed, and manual rejection of non-English language and nonrelated videos was undertaken. Remaining videos were analyzed for content, and suitable videos were scored using a purpose-built tool. YouTube search located 3350 videos in total, and of these, 22 were considered suitable for scoring. The average score was 4.07 out of 7 (standard deviation, 1.35). Six videos scored 5.5 or greater and 5 videos scoring 2.5 or less. There was no correlation between video score and YouTube indices of preference (hits, likes, dislikes, or search page). The quality of videos found in this study was highly variable. YouTube indications of preference were of no value in determining the value of video content. Therefore, teaching institutions or professional societies should endeavor to identify and highlight good online teaching resources. YouTube contains many videos relating to cardiac auscultation, but very few are valuable education resources. © 2012 Wiley Periodicals, Inc.

  2. 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

  3. EPR: steam generator tube rupture analysis in Finland and in France

    International Nuclear Information System (INIS)

    Israel, S.

    2006-01-01

    Different requirements between Finland and France lead EPR designer to define different features (system or action) for management of accidents on Olkiluoto 3 EPR that is under-construction in Finland compared to Flamanville 3 EPR that is foreseen in France. One of these differences concerns the management of Steam Generator Tube Rupture since no primary coolant (liquid and steam) release to the environment is allowed in Finland dislike in France where primary steam releases are not forbidden. This leads to define on Finnish EPR a strategy that anticipates mitigation action compared to French EPR and that only uses the unaffected steam generators. This strategy is intended to reduce the release to the environment. IRSN has analysed an other aspect of the Steam Generator Tube Rupture: the back-flow (flow of un-borated water from steam generator to the primary circuit). Indeed, if the Reactor Coolant Pumps have been shut down, the creation an un-borated water plug because of the back-flow could lead to reactivity accident in case of Reactor Coolant Pump restart. IRSN analysis shows that, using the current Olkiluoto 3 SGTR mitigation strategy and very penalizing assumptions, the amount of un-borated water transferred to the primary circuit on the Finnish EPR could be higher than on the French EPR in the long term. Discussions are going on between STUK and TVO to finalize the SGTR strategy so that both releases into the environment and risk of back-flow can be minimized. (author)

  4. Development of a 100 KV 10 a pulse generator on the basis of electron tubes for plasma immersion ion implantation

    International Nuclear Information System (INIS)

    Kaur, Mandeep; Barve, D.N.; Chakravarthy, D.P.

    2006-01-01

    The design of a high-voltage pulsing system on the basis of hard tube of hard tube for a plasma immersion ion implantation (PIII) facility is presented. A list of requirements, which have to be fulfilled by a high-voltage pulse generator to get best results and an optimum operation of the PIII system, is given. The requirement for the pulse generator can be fulfilled well using a pulse generator design, which employs a hard tube switch. The pulse generator design presented is optimized for PIII systems. The hard tube control can produce nearly rectangular pulses of any duration and repetition frequencies and is especially optimized for obtaining voltage rise times as short as possible. (author)

  5. High insulation foam glass material from waste cathode ray tube panel glass

    DEFF Research Database (Denmark)

    König, Jakob; Petersen, Rasmus Rosenlund; Yue, Yuanzheng

    Recycling of materials from obsolete equipment has become an important part of global waste management. With responsible collecting, dismantling and materials separation, majority of materials can be recycled. Cathode ray tube (CRT) glass represents as much as two-thirds of the weight of a TV...... parameters on the characteristics of foamed glass. CRT panel glass was crushed, milled and sieved below 63 m. Activated carbon used as a foaming agent and MnO2 as an ‘oxidizing’ agent were mixed with glass powders by means of a planetary ball mill. Foaming effect was observed in the temperature range...

  6. Study of a nonlinear system with shocks under broadband excitation. Application to a steam generator tube

    International Nuclear Information System (INIS)

    Thenint, Th.

    2011-01-01

    The steam generator is a heat exchanger and participates to the nuclear safety. Energy is transferred from the primary to the secondary fluid through many U-tubes maintained vertically by support plates. A sludge deposit tends to modify the boundary conditions and the secondary fluid flow. A fluid-elastic instability can then occur and lead to quick tube ruin. This thesis seeks a better understanding of the effect of contact nonlinearity on the dynamics of a tube in-air intermittently impacting the support plates and its consequences in regards with instability. The use of discretized contact conditions with circular obstacles distributed over the thickness of the plates and the use of enriched reduction bases allow quick and relevant nonlinear numerical simulations. These simulations are well correlated with experimental measurements and valid even with strong nonlinearity or negative modal damping. The evolution of power spectral densities (PSD) with growing excitation amplitude is analyzed: padding of the anti-resonances, peak shift and spread. It is then shown that an apparent stiffness associated with a permanent bilateral contact is pertinent to describe these transitions. In the case of an unstable linear system, one demonstrates that the nonlinearity keeps the responses bounded or stabilised, thus paving the way for future work with real or simulated fluid flows. (author)

  7. Calculation of relative tube/tube support plate displacements in steam generators under accident condition loads using non-linear dynamic analysis methodologies

    International Nuclear Information System (INIS)

    Smith, R.E.; Waisman, R.; Hu, M.H.; Frick, T.M.

    1995-01-01

    A non-linear analysis has been performed to determine relative motions between tubes and tube support plates (TSP) during a steam line break (SLB) event for steam generators. The SLB event results in blowdown of steam and water out of the steam generator. The fluid blowdown generates pressure drops across the TSPS, resulting in out-of-plane motion. The SLB induced pressure loads are calculated with a computer program that uses a drift-flux modeling of the two-phase flow. In order to determine the relative tube/TSP motions, a nonlinear dynamic time-history analysis is performed using a structural model that considers all of the significant component members relative to the tube support system. The dynamic response of the structure to the pressure loads is calculated using a special purpose computer program. This program links the various substructures at common degrees of freedom into a combined mass and stiffness matrix. The program accounts for structural non-linearities, including potential tube and TSP interaction at any given tube position. The program also accounts for structural damping as part of the dynamic response. Incorporating all of the above effects, the equations of motion are solved to give TSP displacements at the reduced set of DOF. Using the displacement results from the dynamic analysis, plate stresses are then calculated using the detailed component models. Displacements form the dynamic analysis are imposed as boundary conditions at the DOF locations, and the finite element program then solves for the overall distorted geometry. Calculations are also performed to assure that assumptions regarding elastic response of the various structural members and support points are valid

  8. THE EFFECTS OF SWIRL GENERATOR HAVING WINGS WITH HOLES ON HEAT TRANSFER AND PRESSURE DROP IN TUBE HEAT EXCHANGER

    Directory of Open Access Journals (Sweden)

    Zeki ARGUNHAN

    2006-02-01

    Full Text Available This paper examines the effect of turbulance creators on heat transfer and pressure drop used in concentric heat exchanger experimentaly. Heat exchanger has an inlet tube with 60 mm in diameter. The angle of swirl generators wings is 55º with each wing which has single, double, three and four holes. Swirl generators is designed to easily set to heat exchanger entrance. Air is passing through inner tube of heat exhanger as hot fluid and water is passing outer of inner tube as cool fluid.

  9. Utilization of a sealed-tube neutron generator for training and research

    International Nuclear Information System (INIS)

    Jonah, S.A.

    2000-01-01

    The development of a program in nuclear science and technology in Nigeria began in 1976 with the establishment of two research centers, namely, the Centre for Energy Research and Training, (CERT), Zaria and the Centre for Energy Research and Development (CERD), Ile-Ife. The choice of Neutron Activation Analysis (NAA) technique as a very effective method of training scientists in basic and applied nuclear research led to the purchase of two KAMAN A-711 Neutron Generators for the two research centers. At CERT, the neutron generator (code named ZARABUNG-1) was successfully installed and the first 14 MeV neutrons were produced through the technical assistance of the International Atomic Energy Agency (IAEA) in 1988. In 1991, a new tube-head was purchased and installed due to the expiration of the old tube. Following the completion of its permanent site, the neutron generator was re-located from the old site and re-installed at the permanent site of CERT in 1995. (author)

  10. Program for the development of design data: LMFBR steam generator materials

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-01-01

    Progress in fiscal year 1976 is reported for two tasks in the LMFBR steam generator materials development program. The primary objective of Task 18 is characterization of materials, mainly 2-1/4 Cr-1 Mo and austenitic stainless steels, to assure that satisfactory materials compatibility is achieved with LMFBR steam generator environments. Studies described include: the kinetics and magnitude of decarburization of 2-1/4 Cr-1 Mo in high temperature liquid sodium; decarburization effects on mechanical properties; stress corrosion susceptibility of 2-1/4 Cr-1 Mo in water/steam environments of LMFBR steam generators; and stainless steel mechanical properties within a carburizing sodium environment at high temperatures. In addition, two activities are in progress that directly support LMFBR steam generator design and fabrication. Under Subtask E, General Electric heads the working committee on steam generator materials for the Nuclear Systems Materials Handbook. Subtask F, an experimental effort, is investigating thermal degradation effects produced by particulate deposition (in sodium) on heat transfer surfaces. The Steam Generator Materials Qualification Task (Task 10-G) was initiated in support of the development of materials and processes used in Clinch River Breeder Reactor steam generators. The topics discussed are: LMFBR steam generator tubing; LMFBR steam generator tubesheet forgings (characterization of VAR/ESR melted 2-1/4 Cr-1 Mo); friction, wear and self-welding; nondestructive examination development; water chemistry studies; and transition joint development.

  11. Phenomenological modeling of eddy current signals with a view to characterizing steam generator tube flaws

    International Nuclear Information System (INIS)

    La, R.

    1997-01-01

    This work deals with the eddy current non-destructive test ing. Its long-term goal is to design an 'inverse model' for evaluating the geometry an d the dimensions of steam generator tube flaws from eddy current signals. The approach we adopted requires the preliminary knowledge of a 'forward model' that estimates the eddy current signal knowing the geometry and the dimensions of the flaws. A quasi-exhaustive study of the existing forward models showed their inadequacy to solve the inverse problem. Hence, we proposed to build a general forward model, appropriate to the inversion. Using a parametric approach, this model is phenomenological, i.e. it is based on observations made from results of a finite element code. For each position of the coil, the proposed forward model fist discretized the eddy current distribution into 'tubes of current'. A parametric description of the shape of these tubes is given according the system constituted of the coil and the tubes of current as a 'multi-transformer', their current signal, can then be deduced. The model was validated in the case of an axisymmetric configuration. Comparisons with both analytical and numerical models showed very good agreements. Then, the proposed model was applied to a three-dimensional configuration. Comparisons with experimental results are sufficiently conclusive to validate the approach to the construction of the phenomenological model. However, before envisaging the inverse problem, the computation time, still too long, ought to be reduced and the parametric description needs to be generalized to other three-dimensional configurations. (author)

  12. Numerical investigation of conjugate heat transfer and flow performance of a fin and tube heat exchanger with vortex generators

    DEFF Research Database (Denmark)

    Singh, Shobhana; Sørensen, Kim

    2017-01-01

    Vortex generator is considered as an effective device for augmentation of the thermal-hydraulic performance of a heat exchanger. The aim of present study is to examine the influence of vortex generators on a double fin and tube heat exchanger performance. Vortex generator of rectangular winglet...

  13. Experimental and theoretical investigations on safety of the SNR - straight-tube design steam generator with sodium-water reactions

    International Nuclear Information System (INIS)

    Dumm, K.; Sauermann, F.; Schnitker, W.; Welter, A.

    A number of large sodium-water reaction tests has been performed in a steam generator model in order to verify the layout criteria of the SNR straight-tube design steam generators under accident conditions. The experimental setup is described. The test results and their applicability to the SNR steam generators are given and discussed. (U.S.)

  14. Numerical investigation of heat transfer and entropy generation of laminar flow in helical tubes with various cross sections

    International Nuclear Information System (INIS)

    Kurnia, Jundika C.; Sasmito, Agus P.; Shamim, Tariq; Mujumdar, Arun S.

    2016-01-01

    Highlights: • Heat transfers of helical coiled tube with several cross section profiles are evaluated. • Helical tubes offer higher heat transfer and lower entropy generation. • Square cross-section generates the highest entropy, followed by ellipse and circular. • Study could serve as a guideline in designing an efficient helical tube heat exchanger. - Abstract: This study evaluates heat transfer performance and entropy generation of laminar flow in coiled tubes with various cross-sections geometries i.e. circular, ellipse and square, relatives to the straight tubes of similar cross-sections. A computational fluid dynamics model is developed and validated against empirical correlations. Good agreement is obtained within range of Reynolds and Dean numbers considered. Effect of geometry, wall temperature, Reynolds number and heating/cooling mode were examined. To evaluate the heat transfer performance of the coiled tube configurations, a parameter referred as Figure of Merit (FoM) is defined as the ratio heat transfer rate to the required pumping power. In addition, exergy analysis is carried out to examine the inefficiency of the coiled tube configurations. The results indicate that coiled tubes provide higher heat transfer rate. In addition, it was found to be more efficient as reflected by lower entropy generation as compared to straight tubes. Among the studied cross-section, square cross-section generates the highest entropy, followed by ellipse and circular counterpart. Entropy production from heat transfer contribution is two order-of-magnitude higher than that of entropy contribution from viscous dissipation. Cooling case produces slightly higher entropy than heating counterpart. Finally, this study can provide practical guideline to design more efficient coiled heat exchanger.

  15. Assessment of the integrity of degraded steam generator tube by the use of heterogeneous finite element method

    International Nuclear Information System (INIS)

    Duan, X.; Kozluk, M.; Pagan, S.; Mills, B.

    2006-01-01

    Steam generator tubes at Ontario Power Generation (OPG) have been experiencing a variety of degradations such as pitting, fretting wear, erosion-corrosion, thinning and denting. To assist with steam generator life cycle management, OPG has developed Fitness-For-Service Guidelines (FFSG) for steam generator tubes. The FFSG are intended to provide standard acceptance criteria and evaluation procedures for assessing the condition of steam generator tubes for structural integrity, operational leak rate, and consequential leakage during an upset or abnormal event. Based on inspection results in conjunction with representative, postulated distributions of flaws in the un-inspected tubes, the FFSG provide an acceptable method of satisfying the intent of CSA-N285.4 and justifying the continued operation of degraded steam generator tubes. Some non-mandatory empirical axial and circumferential flaw models are also provided in the FFSG for structural integrity assessments. The test data from the OPG Steam Generator Tube Test Program (SGTTP) showed that the FFSG axial flaw model is conservative for a wide range of defect morphologies. A defect-specific axial flaw model was proposed for lattice-bar fret defects in I800 tubes by utilizing the SGTTP database of extensive test results. A defect-specific flaw model for outer diameter (OD) pitting and inner diameter (ID) intergranular attack in Monel 400 tubes was also developed using the SGTTP test data. More tests have been scheduled to support the development of defect specific models for axial flaws (OD cracks or ID laps) in Monel 400 and to supplement the database for Monel 400 pits. This paper explores the use of simulated testing for use in developing defect specific flaw models to reduce the amount of expensive tests. The Heterogeneous Finite Element Model (HFEM) has been developed and successfully applied to predict the failure behaviour of ductile metals under various deformation modes, i.e. plane stress, plane strain and

  16. 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

  17. Nasal position of nasotracheal tubes: a retrospective analysis of intraoperatively generated three-dimensional X-rays during maxillofacial surgery.

    Science.gov (United States)

    Plümer, Lili; Schön, Gerhard; Klatt, Jan; Hanken, Henning; Schmelzle, Rainer; Pohlenz, Philipp

    2014-10-17

    The aim of this retrospective investigation was to evaluate the position of the nasotracheal tube in the nose and to show its anatomical relationship with the maxillary sinus ostium. Fifty data sets from patients who had undergone endonasal intubation were analyzed for tube positioning. There was a drop-out of eight data sets due to missing information concerning tube size and mode. Tube positioning was determined at the maxillary sinus ostium in the intraoperatively generated three-dimensional X-ray data sets. The type of tube, the tube size, and the presence of maxillary sinusitis were analyzed 30 minutes after intubation. The tube was positioned in the middle nasal meatus in 35 (83.3%) patients and not in the middle nasal meatus in 7 (16.7%) patients. The difference in comparison with equal distribution was significant (P tubes are positioned in the middle nasal meatus. This result can be part of the answer to the question of the causal relationship between position of the breathing tube and the onset of maxillary sinusitis. The indications for prolonged nasotracheal intubation instead of orotracheal intubation or early tracheostomy should be considered carefully.

  18. Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes; Revised September 3, 2003

    International Nuclear Information System (INIS)

    Rochau, Gary E.; Caffey, Thurlow W.H.; Bahram Nassersharif; Garcia, Gabe V.; Jedlicka, Russell P.

    2003-01-01

    OAK B204 Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003. A major cause of failures in heat exchangers and steam generators in nuclear power plants is degradation of the tubes within them. The tube failure is often caused by the development of cracks that begin on the outer surface of the tube and propagate both inwards and laterally. A new technique was researched for detection of defects using a continuous-wave radar method within metal tubing. The technique is 100% volumetric, and may find smaller defects, more rapidly, and less expensively than present methods. The project described in this report was a joint development effort between Sandia National Laboratories (SNL) and New Mexico State University (NMSU) funded by the US Department of Energy. The goal of the project was to research, design, and develop a new concept utilizing a continuous wave radar to detect defects inside metallic tubes and in particular nuclear plant steam generator tubing. The project was divided into four parallel tracks: computational modeling, experimental prototyping, thermo-mechanical design, and signal detection and analysis

  19. Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003

    Energy Technology Data Exchange (ETDEWEB)

    Gary E. Rochau and Thurlow W.H. Caffey, Sandia National Laboratories, Albuquerque, NM 87185-0740; Bahram Nassersharif and Gabe V. Garcia, Department of Mechanical Engineering, New Mexico State University, Las Cruces, NM 88003-8001; Russell P. Jedlicka, Klipsch School of Electrical and Computer Engineering, New Mexico State University, Las Cruces, NM 88003-8001

    2003-05-01

    OAK B204 Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003. A major cause of failures in heat exchangers and steam generators in nuclear power plants is degradation of the tubes within them. The tube failure is often caused by the development of cracks that begin on the outer surface of the tube and propagate both inwards and laterally. A new technique was researched for detection of defects using a continuous-wave radar method within metal tubing. The technique is 100% volumetric, and may find smaller defects, more rapidly, and less expensively than present methods. The project described in this report was a joint development effort between Sandia National Laboratories (SNL) and New Mexico State University (NMSU) funded by the US Department of Energy. The goal of the project was to research, design, and develop a new concept utilizing a continuous wave radar to detect defects inside metallic tubes and in particular nuclear plant steam generator tubing. The project was divided into four parallel tracks: computational modeling, experimental prototyping, thermo-mechanical design, and signal detection and analysis.

  20. Optical Fiber Demodulation System with High Performance for Assessing Fretting Damage of Steam Generator Tubes

    Directory of Open Access Journals (Sweden)

    Peijian Huang

    2018-01-01

    Full Text Available In order to access the fretting damage of the steam generator tube (SGT, a fast fiber Fabry-Perot (F-P non-scanning correlation demodulation system based on a super luminescent light emitting diode (SLED was performed. By demodulating the light signal coming out from the F-P force sensor, the radial collision force between the SGT and the tube support plate (TSP was interrogated. For higher demodulation accuracy, the effects of the center wavelength, bandwidth, and spectrum noise of SLED were discussed in detail. Specially, a piezoelectric ceramic transducer (PZT modulation method was developed to get rid of the interference of mode coupling induced by different types of fiber optics in the demodulation system. The reflectivity of optical wedge and F-P sensor was optimized. Finally, the demodulation system worked well in a 1:1 steam generator test loop and successfully demodulated a force signal of 32 N with a collision time of 2 ms.

  1. Evaluation of heat transfer tube failure propagation due to sodium-water reaction in steam generator

    International Nuclear Information System (INIS)

    Nei, Hiromichi

    1978-01-01

    An evaluation was made of heat transfer tube failure propagation due to sodium-water reaction wastage in a sodium heated steam generator, by comparing an empirically derived wastage equation with leak detector responses. The experimental data agreed well with the wastage equation even for different values of distance-to-nozzle diameter ratio, though the formula had been based on wastage data obtained for only one given distance. The time taken for failure propagation was estimated for a prototype steam generator, and compared with the responses characteristics of acoustic detectors and level gages. It was found that there exists a range of leak rate between 0.5 and 100 g/sec, where the level gage can play a useful role in leak detection. The acoustic detector can be expected to respond more rapidly than the cover gas pressure gage, if noise is kept below ten times the value observed in an experimental facility, SWAT-2. (auth.)

  2. Model-Based Interpretation and Experimental Verification of ECT Signals of Steam Generator Tubes

    International Nuclear Information System (INIS)

    Song, Sung Jin; Kim, Young Hwan; Kim, Eui Lae; Yim, Chang Jae; Lee, Jin Ho

    2004-01-01

    Model-based inversion tools for eddy current signals have been developed by combining neural networks and finite element modeling, for quantitative flaw characterization in steam generator tubes. In the present work, interpretation of experimental eddy current signals was carried out in order to validate the developed inversion tools. A database was constructed using the synthetic flaw signals generated by the finite element model. The hybrid neural networks composed of a PNN classifier and BPNN size estimators were trained using the synthetic signals. Experimental eddy current signals were obtained from axisymmetric artificial flaws. Interpretation of flaw signals was conducted by feeding the experimental signals into the neural networks. The interpretation was excellent, which shows that the developed inversion tools would be applicable to the Interpretation of real eddy current signals

  3. Melting of Nanoprticle-Enhanced Phase Change Material inside Shell and Tube Heat Exchanger

    Directory of Open Access Journals (Sweden)

    Seiyed Mohammad Javad Hosseini

    2013-01-01

    Full Text Available This paper presents a numerical study of melting of Nanoprticle-Enhanced phase change material (NEPCM inside a shell and tube heat exchanger using RT50 and copper particles as base material and nanoparticle, respectively. In this study, the effects of nanoparticles dispersion (, 0.03, and 0.05 on melting time, liquid fraction, and penetration length are investigated. The results show that the melting time decreases to 14.6% and the penetration length increases to 146% with increasing volume fraction of nanoparticle up to .

  4. Long term testing of materials for tube shielding, stage 2; Laangtidsprovning av tubskyddsmaterial, etapp 2

    Energy Technology Data Exchange (ETDEWEB)

    Norling, Rikard; Hjoernhede, Anders; Mattsson, Mattias

    2012-02-15

    Circulating Fluidized Bed (CFB) boilers are commonly used for combustion of biomass and are used to some extent for Waste-to-Energy (WtE) plants as well. The superheaters of the latter are for obvious reasons more prone to suffer from high temperature corrosion caused by the corrosive species in the fuel, mainly chlorides. Frequently the final (hottest) superheater is positioned in the loop seal, where the circulating bed material is returned to the furnace after being separated from the flue gas by a cyclone. The environment in the loop seal is relatively free of chlorides, since these primarily follow the flue gas into the convection pass. Hence, higher steam temperature can be allowed without excessive damage to the final superheater. On the other hand the superheaters, which are located in the convection pass, are more exposed to the corrosive species of the flue gas. Further, they are eroded by particles entrained in the gas flow. Particles and condensing gaseous species are to a large extent deposited on the superheaters, which limits the heat transfer and promotes corrosion. The deposits are regularly removed e.g. by soot blowers. The pressurized steam from soot blowers causes additional erosion damage to that caused by entrained particles. It shall be noted that the actual damage is caused by a combined mechanism of erosion and corrosion denoted erosion-corrosion, which usually results in dramatically accelerated wear. To avoid excessive erosion damage on the superheater tubes the first tube row of each bundle is protected by tube shielding. In its simplest form the shields are made from a steel sheet that has been bent into a semi-circular half-cylinder shell. These shields are attached onto the wind-side of the tubes by hangers. A typical material for tube shielding is the austenitic high temperature resistant stainless steel 253MA. Life of tube shielding depends on numerous factors such as boiler design, superheater location, fuel and operating

  5. Bifunctional thermoelectric tube made of tilted multilayer material as an alternative to standard heat exchangers.

    Science.gov (United States)

    Takahashi, Kouhei; Kanno, Tsutomu; Sakai, Akihiro; Tamaki, Hiromasa; Kusada, Hideo; Yamada, Yuka

    2013-01-01

    Enormously large amount of heat produced by human activities is now mostly wasted into the environment without use. To realize a sustainable society, it is important to develop practical solutions for waste heat recovery. Here, we demonstrate that a tubular thermoelectric device made of tilted multilayer of Bi(0.5)Sb(1.5)Te3/Ni provides a promising solution. The Bi(0.5)Sb(1.5)Te3/Ni tube allows tightly sealed fluid flow inside itself, and operates in analogy with the standard shell and tube heat exchanger. We show that it achieves perfect balance between efficient heat exchange and high-power generation with a heat transfer coefficient of 4.0 kW/m(2)K and a volume power density of 10 kW/m(3) using low-grade heat sources below 100°C. The Bi(0.5)Sb(1.5)Te3/Ni tube thus serves as a power generator and a heat exchanger within a single unit, which is advantageous for developing new cogeneration systems in factories, vessels, and automobiles where cooling of excess heat is routinely carried out.

  6. Ultrashort-pulse laser generated nanoparticles of energetic materials

    Science.gov (United States)

    Welle, Eric J [Niceville, NM; Tappan, Alexander S [Albuquerque, NM; Palmer, Jeremy A [Albuquerque, NM

    2010-08-03

    A process for generating nanoscale particles of energetic materials, such as explosive materials, using ultrashort-pulse laser irradiation. The use of ultrashort laser pulses in embodiments of this invention enables one to generate particles by laser ablation that retain the chemical identity of the starting material while avoiding ignition, deflagration, and detonation of the explosive material.

  7. Development of a novel miniature detonation-driven shock tube assembly that uses in situ generated oxyhydrogen mixture

    International Nuclear Information System (INIS)

    Janardhanraj, S.; Jagadeesh, G.

    2016-01-01

    A novel concept to generate miniature shockwaves in a safe, repeatable, and controllable manner in laboratory confinements using an in situ oxyhydrogen generator has been proposed and demonstrated. This method proves to be more advantageous than existing methods because there is flexibility to vary strength of the shockwave, there is no need for storage of high pressure gases, and there is minimal waste disposal. The required amount of oxyhydrogen mixture is generated using alkaline electrolysis that produces hydrogen and oxygen gases in stoichiometric quantity. The rate of oxyhydrogen mixture production for the newly designed oxyhydrogen generator is found to be around 8 ml/s experimentally. The oxyhydrogen generator is connected to the driver section of a specially designed 10 mm square miniature shock tube assembly. A numerical code that uses CANTERA software package is used to predict the properties of the driver gas in the miniature shock tube. This prediction along with the 1-D shock tube theory is used to calculate the properties of the generated shockwave and matches reasonably well with the experimentally obtained values for oxyhydrogen mixture fill pressures less than 2.5 bars. The miniature shock tube employs a modified tri-clover clamp assembly to facilitate quick changing of diaphragm and replaces the more cumbersome nut and bolt system of fastening components. The versatile nature of oxyhydrogen detonation-driven miniature shock tube opens up new horizons for shockwave-assisted interdisciplinary applications.

  8. Development of a novel miniature detonation-driven shock tube assembly that uses in situ generated oxyhydrogen mixture

    Energy Technology Data Exchange (ETDEWEB)

    Janardhanraj, S.; Jagadeesh, G., E-mail: jaggie@aero.iisc.ernet.in [Department of Aerospace Engineering, Indian Institute of Science, Bangalore 560012 (India)

    2016-08-15

    A novel concept to generate miniature shockwaves in a safe, repeatable, and controllable manner in laboratory confinements using an in situ oxyhydrogen generator has been proposed and demonstrated. This method proves to be more advantageous than existing methods because there is flexibility to vary strength of the shockwave, there is no need for storage of high pressure gases, and there is minimal waste disposal. The required amount of oxyhydrogen mixture is generated using alkaline electrolysis that produces hydrogen and oxygen gases in stoichiometric quantity. The rate of oxyhydrogen mixture production for the newly designed oxyhydrogen generator is found to be around 8 ml/s experimentally. The oxyhydrogen generator is connected to the driver section of a specially designed 10 mm square miniature shock tube assembly. A numerical code that uses CANTERA software package is used to predict the properties of the driver gas in the miniature shock tube. This prediction along with the 1-D shock tube theory is used to calculate the properties of the generated shockwave and matches reasonably well with the experimentally obtained values for oxyhydrogen mixture fill pressures less than 2.5 bars. The miniature shock tube employs a modified tri-clover clamp assembly to facilitate quick changing of diaphragm and replaces the more cumbersome nut and bolt system of fastening components. The versatile nature of oxyhydrogen detonation-driven miniature shock tube opens up new horizons for shockwave-assisted interdisciplinary applications.

  9. Probe-holding apparatus for holding a probe for checking steam generator tubes particularly in a nuclear reactor installation

    International Nuclear Information System (INIS)

    Adamowski, A.; Gagny; Gallet, G.; Lhermitte, J.; Monne, M.; Vautherot, G.

    1984-01-01

    Probe-holding apparatus for holding a probe for checking steam generator tubes particularly in a nuclear reactor installation. The apparatus comprises a telescopic arm supported via a ball and socket joint from a support mounted in or near an access aperture in a chamber at one end of the steam generator. A probe guide is carried by a carriage pivotally mounted at the other end of the telescopic arm. The carriage includes an endless belt having a series of spaced projections which engage into the ends of the tubes, the projections being spaced by a distance equal to the tube pitch or a multiple thereof. The belt is driven by a stepping motor in order to move the carriage and place the probe guide opposite different ones of the tubes

  10. COLENTEC: A new approach to investigate tube support plate clogging of Steam Generators

    International Nuclear Information System (INIS)

    Schindler, Patricia; Tevissen, Etienne; Pointeau, Veronique; Ungar, Alain

    2012-09-01

    Steam generators are crucial components of pressurized water reactors. To maintain their performance over thirty years, plant operators have been faced with a wide range of problems related in the major cases to corrosion issue: denting, intergranular attack, stress corrosion cracking... Recently (2006), the Tube Support Plate (TSP) clogging phenomena increased in some power plants of French nuclear park. In order to cope with this issue, EDF and CEA have launched a collaborative R and D program. It appears that steam generator clogging is potentially driven by a complex phenomenology involving thermochemistry and local mass transfer processes such as particles deposition or erosion. The study of TSP clogging would then take benefit of dedicated experiments reproducing both thermohydraulic and chemical conditions of the secondary circuit. Started in 2007, the COLENTEC project aims to provide these experimental data. The first objective is to reproduce the first stages of deposition of metallic oxides (mainly composed of magnetite Fe 3 O 4 s) on the TSP and then to point out the effect of thermohydraulic and chemical conditions on TSP clogging. This paper presents the ability of a facility composed of three thermohydraulic loops which reproduces the physical and chemical conditions of a steam generator in a test section made up of four SG primary tubes. This test section is a scale mock-up of a portion of the 8 th TSP. The thermal power is provided by a 500 kW electrical heating. Main chemical parameters including iron concentration are continuously monitored and adjusted by an additional low temperature loop connected with the two-phase flow. A first experimental phase will be based on the characterization of iron deposit on the TSP. During the second phase, injection of a radioactive tracer ( 59 Fe) and on-line γ-counting will allow us to quantify the deposition rates as a function of chemical or thermohydraulic parameters. (authors)

  11. Temperature measurements in steam generator tube plates of the advanced PWR (FDR) on the NS OTTO HAHN

    International Nuclear Information System (INIS)

    Klaeke, R.D.

    1980-01-01

    During inspection for surface cracks at the steam generator tube sheets on the feed water inlet sides in several scaling welds small cracks were found. Thermotensions are in view for their origination. To clear up the temperature conditions on the feed water inlets in the areas of the tube sheets a temperature measuring program was performed. The results showed that over a large power range during the steam generation on the feed water inlet side of the steam generator permanently temperature fluctuations took place. (orig.) [de

  12. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  13. Multifrequency eddy current system for steam generator tubing inspection. Volume 2. Analytical studies

    International Nuclear Information System (INIS)

    Libby, H.L.

    1979-04-01

    A multifrequency eddy current testing system has been developed to test nuclear steam generator tubes and has been evaluated on a steam generator mockup. Results to date show that use of more than one inspection frequency facilitates electronic assessment of flaw depth, thereby reducing reliance on visual interpretation of signal information by operators. Details on the system design and an evaluation of the system's performance on a steam generator mockup are provided. The system consists of a four frequency signal generator, which excites the inspection coil, followed by a Walsh function instrument which extracts information from any two of the four frequencies present in the composite test signal. The extracted information is processed to discriminate against unwanted signals, such as those from probe wobble, and is then transmitted to the defect decision circuitry for additional processing. Results of the mockup tests show that the system has a higher probability of flaw detection in many cases than does a conventional single frequency test. Tutorial information is presented on algebraic solutions of simultaneous equations and on representation and analysis of signals using orthogonal functions. Examples illustrating the design of the multifrequency inspection system are included. Also presented is an analytical study of several candidate means for implementing electronic assessment of flaw depth

  14. Next Generation TRD for CREAM Using Gas Straw Tubes and Foam Radiators

    Science.gov (United States)

    Malinin, A.; Ahn, H.S.; Fedin, O.; Ganel, O.; Han, J.H.; Kim, C.H.; Kim, K.C.; Lee, M.H.; Lutz, L.; Seo, E.S.; Walpole, P.; Wu, J.; Yoo, J.H.; Yoon, Y.S.; Zinn, S.Y.

    The Cosmic Ray Energetics And Mass (CREAM) experiment is designed to investigate the source, propagation and acceleration mechanism of high energy cosmic-ray nuclei, by directly measuring their energy and charge. Incorporating a transition radiation detector (TRD) provides an energy measurement complementary to the calorimeter, as well as additional track reconstruction capability. The next generation CREAM TRD is designed with 4 mm straw tubes to greatly improve tracking over the previous 20 mm tube design, thereby enhancing charge identification in the silicon charge detector (SCD). Plastic foam provides a weight-efficient radiator that doubles as a mechanical support for the straw layers. This design provides a compact, robust, reliable, low density detector to measure incident nucleus energy for 3 < Z < 30 nuclei in the Lorentz gamma factor range of 102-105. This paper discusses the new TRD design and the low power front end electronics used to achieve the large dynamic range required. Beam test results of a prototype TRD are also reported.

  15. Scale model test results for an inverted U-tube steam generator with comparisons to heat transfer correlations

    International Nuclear Information System (INIS)

    Boucher, T.J.

    1987-01-01

    To provide data for assessment and development of thermal-hydraulic computer codes, bottom main feedwater-line-break transient simulations were performed in a scale model (Semiscale Mod-2C) of a pressurized water reactor (PWR) with conditions typical of a PWR (15.0 MPa primary pressure, 600 K steam generator inlet plenum fluid temperatures, 6.2 MPa secondary pressure). The state-of-the-art measurements in the scale model (Type III) steam generator allow for the determination of U-tube steam generator allow for the determination of U-tube steam generator secondary component interactions, tube bundle local radial heat transfer, and tube bundle and riser vapor void fractions for steady state and transient operations. To enhance the understanding of the observed phenomena, the component interactions, local heat fluxes, local secondary convective heat transfer coefficients and local vapor void fractions are discussed for steady state, full-power and transient operations. Comparisons between the measurement-derived secondary convective heat transfer coefficients and those predicted by a number of correlations, including the Chen correlation currently used in thermal-hydraulic computer codes, show that none of the correlations adequately predict the data and points out the need for the formulation of a new correlation based on this experimental data. The unique information presented herein should be of the interest to anyone involved in modeling inverted U-tube steam generator thermal-hydraulics for forced convection boiling/vaporization heat transfer. 5 refs., 13 figs., 1 tab

  16. Applicability of chemical cleaning process to steam generator secondary side, (3). Effect of chemical cleaning on long term integrity of steam generator tube after chemical cleaning process

    International Nuclear Information System (INIS)

    Kawamura, Hirotaka; Fujiwara, Kazutoshi; Kanbe, Hiromi; Hirano, Hideo; Takiguchi, Hideki; Yoshino, Kouji; Yamamoto, Shuuichi; Shibata, Toshio; Ishigure, Kenkichi

    2006-01-01

    The application of the chemical cleaning process to dissolve and remove scales and sludge by chemicals is being planned at the Japanese pressurized water reactor (PWR) plant in order to maintain a designed heat transfer condition and to prevent the steam generator (SG) tube degradation. In this paper, the affects of the EPRI process and the KWU process on the long term integrity of SG tubing were investigated under the simulated SG condition using a SG model boiler test facility. No adverse effect of the both chemical cleaning processes on the long term integrity of SG tubing were observed. (author)

  17. YouTube as a crowd-generated water level archive.

    Science.gov (United States)

    Michelsen, N; Dirks, H; Schulz, S; Kempe, S; Al-Saud, M; Schüth, C

    2016-10-15

    In view of the substantial costs associated with classic monitoring networks, participatory data collection methods can be deemed a promising option to obtain complementary data. An emerging trend in this field is social media mining, i.e., harvesting of pre-existing, crowd-generated data from social media. Although this approach is participatory in a broader sense, the users are mostly not aware of their participation in research. Inspired by this novel development, we demonstrate in this study that it is possible to derive a water level time series from the analysis of multiple YouTube videos. As an example, we studied the recent water level rise in Dahl Hith, a Saudi Arabian cave. To do so, we screened 16 YouTube videos of the cave for suitable reference points (e.g., cave graffiti). Then, we visually estimated the distances between these points and the water level and traced their changes over time. To bridge YouTube hiatuses, we considered own photos taken during two site visits. For the time period 2013-2014, we estimate a rise of 9.5m. The fact that this rise occurred at a somewhat constant rate of roughly 0.4m per month points towards a new and permanent water source, possibly two nearby lakes formed from treated sewage effluent. An anomaly in the rising rate is noted for autumn 2013 (1.3m per month). As this increased pace coincides with a cluster of rain events, we deem rapid groundwater recharge along preferential flow paths a likely cause. Despite the sacrifice in precision, we believe that YouTube harvesting may represent a viable option to gather historical water levels in data-scarce settings and that it could be adapted to other environments (e.g., flood extents). In certain areas, it might provide an additional tool for the monitoring toolbox, thereby possibly delivering hydrological data for water resources management. Copyright © 2016 Elsevier B.V. All rights reserved.

  18. Generation and use of process maps for hot extrusion of seamless tubes for nuclear applications

    International Nuclear Information System (INIS)

    Vaibhaw, Kumar; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

    2009-01-01

    Full text: Hot extrusion is known as significant bulk deformation step in manufacturing of seamless tube production. Elevated temperature deformation carried out above the recrystallization temperature would enable imposition of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on extrusion process parameters such as temperature and strain rate (Ram speed). Basic microstructure developed at this deformation stage has significant bearing on the final properties of the material fabricated with subsequent cold working steps. Zirconium alloys and special nuclear grade austenitic stainless steels are two important groups of materials used as structural and core components in thermal and fast reactors world wide respectively. The properties of former alloy are very sensitive to the thermo mechanical fabrication steps initiated with hot extrusion due to their anisotropic deformation behaviour. However, nuclear grade austenitic stainless steels have many variants from their commercial grades in terms of micro and macro alloy chemistry. Factors such as these significantly affect the workability of the materials and require proper selection of extrusion parameters especially working temperature and extrusion speed plays a key role in the quality of the product. Modern developments in processing technology envisage the application of processing maps based on dynamic material model for selection of hot extrusion parameters. The present paper is aimed at bringing out significance of the map in selection of working domain with respect to the industrial process conditions for both groups of nuclear materials mentioned earlier. Developed process maps of certain alloys suggest use of extremely slow strain rate and low temperature extrusion which can not be achieved during bulk processing due to design of equipment and heat transfer constraints in industrial scale production. Attempts are made to highlight

  19. Material and fin pitch effect on frosting CO2 in a fin-and-tube heat exchanger

    Science.gov (United States)

    Bassila, Joseph; Toubassy, Joseph; Danlos, Amélie; Descombes, Georges; Clodic, Denis

    2017-02-01

    Cryo Pur technology uses cryogenic separation to remove water vapor and carbon dioxide from biogas, in order to obtain bio-methane. To cool down the biogas at a very low temperature, a fin-and-tube heat exchanger is designed. In order to improve the fin-and-tube heat exchanger performance, a model is developed to investigate the material and fin pitch on frosting carbon dioxide. This paper will study the effect of the tubes and the fins material, and the fin pitch effect. The purpose is to extend the duration of a frosting cycle.

  20. Steam-generator-tube-rupture transients for pressurized-water reactors

    International Nuclear Information System (INIS)

    Dobranich, D.; Henninger, R.J.; DeMuth, N.S.

    1982-01-01

    Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures

  1. Probability of detection model for the non-destructive inspection of steam generator tubes of PWRs

    Science.gov (United States)

    Yusa, N.

    2017-06-01

    This study proposes a probability of detection (POD) model to discuss the capability of non-destructive testing methods for the detection of stress corrosion cracks appearing in the steam generator tubes of pressurized water reactors. Three-dimensional finite element simulations were conducted to evaluate eddy current signals due to stress corrosion cracks. The simulations consider an absolute type pancake probe and model a stress corrosion crack as a region with a certain electrical conductivity inside to account for eddy currents flowing across a flaw. The probabilistic nature of a non-destructive test is simulated by varying the electrical conductivity of the modelled stress corrosion cracking. A two-dimensional POD model, which provides the POD as a function of the depth and length of a flaw, is presented together with a conventional POD model characterizing a flaw using a single parameter. The effect of the number of the samples on the PODs is also discussed.

  2. Improvements in the simulation of a main steam line break with steam generator tube rupture

    Science.gov (United States)

    Gallardo, Sergio; Querol, Andrea; Verdú, Gumersindo

    2014-06-01

    The result of simultaneous Main Steam Line Break (MSLB) and a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR) is a depressurization in the secondary and primary system because both systems are connected through the SGTR. The OECD/NEA ROSA-2 Test 5 performed in the Large Scale Test Facility (LSTF) reproduces these simultaneous breaks in a Pressurized Water Reactor (PWR). A simulation of this Test 5 was made with the thermal-hydraulic code TRACE5. Some discrepancies found, such as an underestimation of SG-A secondary pressure during the depressurization and overestimation of the primary pressure drop after the first Power Operated Relief Valve (PORV) opening can be improved increasing the nodalization of the Upper Head in the pressure vessel and meeting the actual fluid conditions of Upper Head during the transient.

  3. Development of intelligent Eddy Current Testing (ECT) system for PWR steam generator tube inspection

    International Nuclear Information System (INIS)

    Kawata, K.; Kawase, N.; Kurokawa, M.; Asada, Y.

    2005-01-01

    The intelligent ECT system was developed for the inspection of heat transfer tubes of the steam generator of the PWR plant. It consists of intelligent probe, data acquisition unit and data analysis system. The probe combines 24 channels inclined lay out drive coils and thin film pick-up coils with built-in electric circuits to provide high inspection capability equivalent to rotating coil ECT and high-speed inspection equivalent to conventional bobbin coil ECT. The advanced data analysis system that has filtering and automatic analysis functions is also developed to enable fast and precise analysis of large volume inspection data. The system was qualified by confirmation tests in FY 2003 to show thinned thickness sizing accuracy within ± 5%. (T. Tanaka)

  4. Radioactivity release vs probability for a steam generator tube rupture accident

    International Nuclear Information System (INIS)

    Buslik, A.J.; Hall, R.E.

    1978-01-01

    A calculation of the probability of obtaining various radioactivity releases from a steam generator tube rupture (SGTR) is presented. The only radioactive isotopes considered are Iodine-131 and Xe-133. The particular accident path considered consists of a double-ended guillotine SGTR followed by loss of offsite power (LOSP). If there is no loss of offsite power, and no system fault other than the SGTR, it is judged that the consequences will be minimal, since the amount of iodine released through the condenser air ejector is expected to be quite small; this is a consequence of the fact that the concentration of iodine in the vapor released from the condenser air ejector is very small compared to that dissolved in the condensate water. In addition, in some plants the condenser air ejector flow is automatically diverted to containment or a high-activity alarm. The analysis presented here is for a typical Westinghouse PWR such as described in RESAR-3S

  5. Impulsively Generated Sausage Waves in Coronal Tubes with Transversally Continuous Structuring

    Science.gov (United States)

    Yu, Hui; Li, Bo; Chen, Shao-Xia; Xiong, Ming; Guo, Ming-Zhe

    2016-12-01

    The frequency dependence of the longitudinal group speeds of trapped sausage waves plays an important role in determining impulsively generated wave trains, which have often been invoked to account for quasi-periodic signals in coronal loops. We examine how the group speeds ({v}{gr}) depend on angular frequency (ω) for sausage modes in pressureless coronal tubes with continuous transverse density distributions by solving the dispersion relation pertinent to the case where the density inhomogeneity of arbitrary form occurs in a transition layer of arbitrary thickness. We find that in addition to the transverse lengthscale l and density contrast {ρ }{{I}}/{ρ }{{e}}, the group speed behavior also depends on the detailed form of the density inhomogeneity. For parabolic profiles, {v}{gr} always decreases with ω first before increasing again, as happens for the much studied top-hat profiles. For linear profiles, however, the behavior of the ω -{v}{gr} curves is more complex. When {ρ }{{I}}/{ρ }{{e}}≲ 6, the curves become monotonical for large values of l. On the other hand, for higher density contrasts, a local maximum {v}{gr}\\max exists in addition to a local minimum {v}{gr}\\min when coronal tubes are diffuse. With time-dependent computations, we show that the different behavior of group speed curves, the characteristic speeds {v}{gr}\\min and {v}{gr}\\max in particular, is reflected in the temporal evolution and Morlet spectra of impulsively generated wave trains. We conclude that the observed quasi-periodic wave trains not only can be employed to probe such key parameters as density contrasts and profile steepness, but also have the potential to discriminate between the unknown forms of the transverse density distribution.

  6. SCC susceptibility and flaw tolerance evaluation for steam generator channel head materials

    International Nuclear Information System (INIS)

    Cothron, H.; Wolfe, R.

    2015-01-01

    Primary water stress corrosion cracking (PWSCC) has been reported in the divider plate assemblies of steam generators in operation outside of the United States. Evaluations have been performed to assess the susceptibility of U.S. steam generator channel head materials to PWSCC and the flaw tolerance of the channel head assembly in instances where cracking is assumed to occur. Earlier work concluded that the cracks reported in the foreign steam generators could not cause failure of the divider plate in the limiting U.S. steam generators during the design basis accident or normal operating conditions. Three additional cracking scenarios represent a potential breach of the primary pressure boundary of the channel head assembly: cracks initiating in the divider plate then propagating through the channel head cladding and into the low alloy steel shell material, cracks initiating in the tubesheet cladding then propagating to the tube-to-tubesheet weldments, and cracks initiating in the tube-to-tubesheet weld. Operating experience and literature were reviewed to determine the likelihood that cracks will propagate into the carbon steel channel head material and cause a breach in the primary pressure boundary. A flaw tolerance evaluation demonstrated that the structural integrity of the steam generator channel head is not compromised by a crack originating in the divider plate. Assumed axial and circumferential flaws in the steam generator channel head material remain well below the allowable flaw depths after 40 years of operation. Utilities can use the results from this analysis to determine the need to inspect steam generator Alloy 600 material in the divider plate assembly or the low-alloy steel channel head material in the period of operation beyond 40 years. (authors)

  7. Attenuation of hydrogen radicals traveling under flowing gas conditions through tubes of different materials

    International Nuclear Information System (INIS)

    Grubbs, R.K.; George, S.M.

    2006-01-01

    Hydrogen radical concentrations traveling under flowing gas conditions through tubes of different materials were measured using a dual thermocouple probe. The source of the hydrogen radicals was a toroidal radio frequency plasma source operating at 2.0 and 3.3 kW for H 2 pressures of 250 and 500 mTorr, respectively. The dual thermocouple probe was comprised of exposed and covered Pt/Pt13%Rh thermocouples. Hydrogen radicals recombined efficiently on the exposed thermocouple and the energy of formation of H 2 heated the thermocouple. The second thermocouple was covered by glass and was heated primarily by the ambient gas. The dual thermocouple probe was translated and measured temperatures at different distances from the hydrogen radical source. These temperature measurements were conducted at H 2 flow rates of 35 and 75 SCCM (SCCM denotes cubic centimeter per minute at STP) inside cylindrical tubes made of stainless steel, aluminum, quartz, and Pyrex. The hydrogen radical concentrations were obtained from the temperatures of the exposed and covered thermocouples. The hydrogen concentration decreased versus distance from the plasma source. After correcting for the H 2 gas flow using a reference frame transformation, the hydrogen radical concentration profiles yielded the atomic hydrogen recombination coefficient, γ, for the four materials. The methodology of measuring the hydrogen radical concentrations, the analysis of the results under flowing gas conditions, and the determination of the atomic hydrogen recombination coefficients for various materials will help facilitate the use of hydrogen radicals for thin film growth processes

  8. In situ microscopy reveals reversible cell wall swelling in kelp sieve tubes: one mechanism for turgor generation and flow control?

    Science.gov (United States)

    Knoblauch, Jan; Tepler Drobnitch, Sarah; Peters, Winfried S; Knoblauch, Michael

    2016-08-01

    Kelps, brown algae (Phaeophyceae) of the order Laminariales, possess sieve tubes for the symplasmic long-distance transport of photoassimilates that are evolutionarily unrelated but structurally similar to the tubes in the phloem of vascular plants. We visualized sieve tube structure and wound responses in fully functional, intact Bull Kelp (Nereocystis luetkeana [K. Mertens] Postels & Ruprecht 1840). In injured tubes, apparent slime plugs formed but were unlikely to cause sieve tube occlusion as they assembled at the downstream side of sieve plates. Cell walls expanded massively in the radial direction, reducing the volume of the wounded sieve elements by up to 90%. Ultrastructural examination showed that a layer of the immediate cell wall characterized by circumferential cellulose fibrils was responsible for swelling and suggested that alginates, abundant gelatinous polymers of the cell wall matrix, were involved. Wall swelling was rapid, reversible and depended on intracellular pressure, as demonstrated by pressure-injection of silicon oil. Our results revive the concept of turgor generation and buffering by swelling cell walls, which had fallen into oblivion over the last century. Because sieve tube transport is pressure-driven and controlled physically by tube diameter, a regulatory role of wall swelling in photoassimilate distribution is implied in kelps. © 2016 John Wiley & Sons Ltd.

  9. Research and development on probe inserting method into steam generator helically coiled tubes for in-service inspection

    International Nuclear Information System (INIS)

    Ohashi, Kiyoshi; Takahashi, Masao; Sagae, Makoto; Watanabe, Naoto

    1979-01-01

    Helically coiled tubes of steam generators (SG) in FBR are boundaries between sodium and water/steam. Therefore, to assure the integrity of tubes, it is necessary to inspect the tubes nondestructively for in service or after a sodium-water reaction accident. In order to make it possible to conduct in-service inspection of SG tubes, we have studied on eddy current probes and probe inserting methods. As for the probe inserting method, IHI designed a fluid driving type which consists of a model probe and signal cable with float balls and driven by air pressure force. Presented in this paper is the authors' report, which describes the fluid driving type as an effective method to insert an eddy current probe into helically coiled tubes. The outline of the test results is as follows: 1. It was possible to insert the probe into 65 meter length helically coiled tubes. 2. We could detected, as anticipated, a defect (outer circumferential wall thinning defect, 20% depth) on a test piece jointed with the helically coiled tubes. (author)

  10. 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.

  11. Support and tool displacement device for the attachment of a tube bundle on a tubular plate of a steam generator

    International Nuclear Information System (INIS)

    Morisot, M.; Werle, R.; Michaud, J.P.

    1983-01-01

    The steam generator is being assembled, disposed with its axis horizontal and its tubular plate vertical; the device described in this patent, allows to automatize the preparation stages of the tubular plate and the attachment of the bundle, to shorten the construction of the steam generator and to remove drudgeries done by hand on the tubular plate or the tubes of the bundle. The invention can be applied to the construction of PWR steam generators [fr

  12. Simulation Study on Material Property of Cantilever Piezoelectric Vibration Generator

    Directory of Open Access Journals (Sweden)

    Yan Zhen

    2014-06-01

    Full Text Available For increasing generating capacity of cantilever piezoelectric vibration generator with limited volume, relation between output voltage, inherent frequency and material parameter of unimorph, bimorph in series type and bimorph in parallel type piezoelectric vibration generator is analyzed respectively by mechanical model and finite element modeling. The results indicate PZT-4, PZT- 5A and PZT-5H piezoelectric materials and stainless steel, nickel alloy substrate material should be firstly chosen.

  13. GENERATION OF MAGNETOHYDRODYNAMIC WAVES IN LOW SOLAR ATMOSPHERIC FLUX TUBES BY PHOTOSPHERIC MOTIONS

    International Nuclear Information System (INIS)

    Mumford, S. J.; Fedun, V.; Erdélyi, R.

    2015-01-01

    Recent ground- and space-based observations reveal the presence of small-scale motions between convection cells in the solar photosphere. In these regions, small-scale magnetic flux tubes are generated via the interaction of granulation motion and the background magnetic field. This paper studies the effects of these motions on magnetohydrodynamic (MHD) wave excitation from broadband photospheric drivers. Numerical experiments of linear MHD wave propagation in a magnetic flux tube embedded in a realistic gravitationally stratified solar atmosphere between the photosphere and the low choromosphere (above β = 1) are performed. Horizontal and vertical velocity field drivers mimic granular buffeting and solar global oscillations. A uniform torsional driver as well as Archimedean and logarithmic spiral drivers mimic observed torsional motions in the solar photosphere. The results are analyzed using a novel method for extracting the parallel, perpendicular, and azimuthal components of the perturbations, which caters to both the linear and non-linear cases. Employing this method yields the identification of the wave modes excited in the numerical simulations and enables a comparison of excited modes via velocity perturbations and wave energy flux. The wave energy flux distribution is calculated to enable the quantification of the relative strengths of excited modes. The torsional drivers primarily excite Alfvén modes (≈60% of the total flux) with small contributions from the slow kink mode, and, for the logarithmic spiral driver, small amounts of slow sausage mode. The horizontal and vertical drivers primarily excite slow kink or fast sausage modes, respectively, with small variations dependent upon flux surface radius

  14. Generation of Magnetohydrodynamic Waves in Low Solar Atmospheric Flux Tubes by Photospheric Motions

    Science.gov (United States)

    Mumford, S. J.; Fedun, V.; Erdélyi, R.

    2015-01-01

    Recent ground- and space-based observations reveal the presence of small-scale motions between convection cells in the solar photosphere. In these regions, small-scale magnetic flux tubes are generated via the interaction of granulation motion and the background magnetic field. This paper studies the effects of these motions on magnetohydrodynamic (MHD) wave excitation from broadband photospheric drivers. Numerical experiments of linear MHD wave propagation in a magnetic flux tube embedded in a realistic gravitationally stratified solar atmosphere between the photosphere and the low choromosphere (above β = 1) are performed. Horizontal and vertical velocity field drivers mimic granular buffeting and solar global oscillations. A uniform torsional driver as well as Archimedean and logarithmic spiral drivers mimic observed torsional motions in the solar photosphere. The results are analyzed using a novel method for extracting the parallel, perpendicular, and azimuthal components of the perturbations, which caters to both the linear and non-linear cases. Employing this method yields the identification of the wave modes excited in the numerical simulations and enables a comparison of excited modes via velocity perturbations and wave energy flux. The wave energy flux distribution is calculated to enable the quantification of the relative strengths of excited modes. The torsional drivers primarily excite Alfvén modes (≈60% of the total flux) with small contributions from the slow kink mode, and, for the logarithmic spiral driver, small amounts of slow sausage mode. The horizontal and vertical drivers primarily excite slow kink or fast sausage modes, respectively, with small variations dependent upon flux surface radius.

  15. GENERATION OF MAGNETOHYDRODYNAMIC WAVES IN LOW SOLAR ATMOSPHERIC FLUX TUBES BY PHOTOSPHERIC MOTIONS

    Energy Technology Data Exchange (ETDEWEB)

    Mumford, S. J.; Fedun, V.; Erdélyi, R., E-mail: s.mumford@sheffield.ac.uk [Solar Physics and Space Plasma Research Centre (SP2RC), School of Mathematics and Statistics, The University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH UK (United Kingdom)

    2015-01-20

    Recent ground- and space-based observations reveal the presence of small-scale motions between convection cells in the solar photosphere. In these regions, small-scale magnetic flux tubes are generated via the interaction of granulation motion and the background magnetic field. This paper studies the effects of these motions on magnetohydrodynamic (MHD) wave excitation from broadband photospheric drivers. Numerical experiments of linear MHD wave propagation in a magnetic flux tube embedded in a realistic gravitationally stratified solar atmosphere between the photosphere and the low choromosphere (above β = 1) are performed. Horizontal and vertical velocity field drivers mimic granular buffeting and solar global oscillations. A uniform torsional driver as well as Archimedean and logarithmic spiral drivers mimic observed torsional motions in the solar photosphere. The results are analyzed using a novel method for extracting the parallel, perpendicular, and azimuthal components of the perturbations, which caters to both the linear and non-linear cases. Employing this method yields the identification of the wave modes excited in the numerical simulations and enables a comparison of excited modes via velocity perturbations and wave energy flux. The wave energy flux distribution is calculated to enable the quantification of the relative strengths of excited modes. The torsional drivers primarily excite Alfvén modes (≈60% of the total flux) with small contributions from the slow kink mode, and, for the logarithmic spiral driver, small amounts of slow sausage mode. The horizontal and vertical drivers primarily excite slow kink or fast sausage modes, respectively, with small variations dependent upon flux surface radius.

  16. A study on the deformation mechanism of inconel alloys for steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Hong, S. H.; Kim, H. Y.; Ahn, Y. C.; Lee, H. S.; Sohn, W. H.; Cha, S. I.; Bae, Y. H. [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

    2001-04-01

    The microstructure and the mechanical properties of lnconel 600 and 690 steam generator tube were investigated to develop the lnconel steam generator tube for nuclear power plant. The grain size and shape of lnconel 690 alloy were dependent on Mill Annealing temperature. It is shown that the smaller grain and the more serrated grain boundary, the higher tensile strength and creep resistance. The carbides were analyzed by using SEM and TEM after Thermal Treatment. Also, the quantitative analysis of carbide precipitation with Thermal Treatment was conducted by SPEED method. Thermal Treatment temperature was 705 deg C for lnconel 600 alloy and 720 deg C for lnconel 690 alloy. It is observed by XRD that the M{sub 23}C{sub 6} type and M{sub 7}C{sub 3}-type carbides were simultaneously precipitated in lnconel 600 alloy and only M{sub 23}C{sub 6}-type carbides were precipitated in lnconel 690 alloy. While the carbide length and thickness increased with increasing Thermal Treatment time, the growth rate of carbide length was higher than that of carbide thickness. The distance between carbides decreased for the first 15 hour Thermal Treatment for lnconel 600 alloy and 10 hour for lnconel 690 alloy, respectively, and then increased again. The distance between carbides decreased until the carbide precipitation reached at 90% of maximum precipitation because the carbides were precipitated and grown simultaneously. After 90% precipitation time, the precipitation ceased and only growth of the carbides took place, so the distance between carbides decreased. The serrations in stress-strain curves of in lnconel 690 could be interpreted in the temperature range from 200 deg C to 600 deg C. The phenomena and controlling mechanism of serrations were analyzed by investigating the critical strains for the onset of the serrations. The controlling mechanism for A1 serration was generation of vacancy by deformation, and those for A2 serration and B serrations were diffusion of

  17. Matching the laser generated p bunch into a crossbar-H drift tube linac

    Directory of Open Access Journals (Sweden)

    A. Almomani

    2012-05-01

    Full Text Available Proton bunches with energies up to 30 MeV have been measured at the PHELIX laser. Because of the laser-plasma interactions at a power density of about 4×10^{19}  W/cm^{2}, a total yield of 1.5×10^{13}  protons was produced. For the reference energy of 10 MeV, the yield within ±0.5  MeV was exceeding 10^{10}  protons. The important topic for a further acceleration of the laser generated bunch is the matching into the acceptance of an rf accelerator stage. With respect to the high space charge forces and the transit energy range, only drift tube linacs seem adequate for this purpose. A crossbar H-type (CH cavity was chosen as the linac structure. Optimum emittance values for the linac injection are compared with the available laser generated beam parameters. Options for beam matching into a CH structure by a pulsed magnetic solenoid and by using the simulation codes LASIN and LORASR are presented.

  18. Matching the laser generated p bunch into a crossbar-H drift tube linac

    Science.gov (United States)

    Almomani, A.; Droba, M.; Ratzinger, U.; Hofmann, I.

    2012-05-01

    Proton bunches with energies up to 30 MeV have been measured at the PHELIX laser. Because of the laser-plasma interactions at a power density of about 4×1019W/cm2, a total yield of 1.5×1013protons was produced. For the reference energy of 10 MeV, the yield within ±0.5MeV was exceeding 1010protons. The important topic for a further acceleration of the laser generated bunch is the matching into the acceptance of an rf accelerator stage. With respect to the high space charge forces and the transit energy range, only drift tube linacs seem adequate for this purpose. A crossbar H-type (CH) cavity was chosen as the linac structure. Optimum emittance values for the linac injection are compared with the available laser generated beam parameters. Options for beam matching into a CH structure by a pulsed magnetic solenoid and by using the simulation codes LASIN and LORASR are presented.

  19. The Thermal Hydraulics of Tube Support Fouling in Nuclear Steam Generators

    International Nuclear Information System (INIS)

    Rummens, Helena E.C.; Rogers, J.T.; Turner, C.W.

    2004-01-01

    It is hypothesized that the thermal-hydraulic environment plays a role in the fouling of tube supports in nuclear steam generators. Experiments were performed to simulate the thermal-hydraulic environment near various designs of supports. Pressure loss, local velocity, turbulence intensity, and local void fraction were measured to characterize the effect of the support. Fouling mechanisms specific to supports were inferred from these experimental data and from actual steam generator inspection results. An analytical model was developed to predict the rate of particulate deposition on the supports, to better understand the complex processes involved.This paper presents the following set of tools for assessing the fouling propensity of a given support design: (1) proposed fouling mechanisms, (2) criteria for support fouling propensity, (3) correlation of fouling with parameters such as mass flux and quality, (4) descriptions of experimental tools such as flow visualization and measurement of pressure-loss profiles, and (5) analytical tools.An important conclusion from this and our previous work is that the fouling propensity is greater with broached support plates, both trefoil and quatrefoil, than with lattice bar supports and formed bar supports, in which significant cross flows occur

  20. Comprehensive Modeling of U-Tube Steam Generators Using Extreme Learning Machines

    Science.gov (United States)

    Beyhan, Selami; Kavaklioglu, Kadir

    2015-10-01

    This paper proposes artificial neural network and fuzzy system-based extreme learning machines (ELM) for offline and online modeling of U-tube steam generators (UTSG). Water level of UTSG systems is predicted in a one-step-ahead fashion using nonlinear autoregressive with exogenous input (NARX) topology. Modeling data are generated using a well-known and widely accepted dynamic model reported in the literature. Model performances are analyzed with different number of neurons for the neural network and with different number of rules for the fuzzy system. UTSG models are built at different reactor power levels as well as full range that corresponds to all reactor operating powers. A quantitative comparison of the models are made using the root-mean-squared error (RMSE) and the minimum-descriptive-length (MDL) criteria. Furthermore, conventional back propagation learning-based neural and fuzzy models are also designed for comparing ELMs to classical artificial models. The advantages and disadvantages of the designed models are discussed.

  1. Safety evaluation report related to steam generator tube repair and return to operation Three Mile Island Nuclear Station, Unit No. 1 (Docket No. 50-289)

    International Nuclear Information System (INIS)

    Silver, H.

    1983-11-01

    Based on our evaluation of the steam generator tube repair method and of subsequent operation using the repaired steam generators, we conclude that the steam generator tube kinetic expansion process is acceptable, that applicable GDC have been met, and that there is reasonable assurance that the health and safety of the public will not be endangered by subsequent operation of the plant

  2. Generator for ionizing radiation

    International Nuclear Information System (INIS)

    Romanovskij, V.F.; Panasjuk, V.S.; Stepanov, B.M.; Ovtscharov, A.M.; Akimov, J.A.

    1979-01-01

    The X-ray, electron, or neutron generator contains a radiation source with an accelerating tube, whose shell encloses a resonance transformer, a subdivided tube insulator and a high-tension electrode for the accelerating tube. The accelerating tube can be evacuated. The high-tension winding of the resonance transformer lies within the tube insulator of the accelerating tube and the evacuated space between resonance transformer and tube insulator. The generator may be applied in medicine, in geophysical research or for activation analysis of materials. (DG) 891 HP/DG 892 BRE [de

  3. Phase Change Material Thermal Power Generator

    Science.gov (United States)

    Jones, Jack A. (Inventor); Chao, Yi (Inventor); Valdez, Thomas I. (Inventor)

    2014-01-01

    An energy producing device, for example a submersible vehicle for descending or ascending to different depths within water or ocean, is disclosed. The vehicle comprises a temperature-responsive material to which a hydraulic fluid is associated. A pressurized storage compartment stores the fluid as soon as the temperature-responsive material changes density. The storage compartment is connected with a hydraulic motor, and a valve allows fluid passage from the storage compartment to the hydraulic motor. An energy storage component, e.g. a battery, is connected with the hydraulic motor and is charged by the hydraulic motor when the hydraulic fluid passes through the hydraulic motor. Upon passage in the hydraulic motor, the fluid is stored in a further storage compartment and is then sent back to the area of the temperature-responsive material.

  4. X-ray output and percentage ripple in x-ray tube voltage. X-ray generators for mammography

    International Nuclear Information System (INIS)

    Miyazaki, Shigeru; Katoh, Yoh; Negishi, Toru; Abe, Shinji; Ogura, Izumi

    1998-01-01

    Various characteristics of x-ray generators used for mammography (tube voltage, tube current, percentage average error of irradiation time, percentage ripple of the tube voltage waveform, linearity, and reproducibility of the photographic effect) have already been clarified by the authors. In our more recent investigations, x-ray output and radiation quality as percentage ripple of the tube voltage waveform were evaluated using the dynamic study method with the aluminum filter specified in the International Electrotechnical Commission (IEC) standard. In addition, we also assessed the effects of fluctuation in percentage ripple of the tube voltage waveform on the x-ray spectrum. Based on the results obtained, the characteristics of an ideal x-ray generator for mammography are discussed. The results of this study showed that x-ray output differences in terms of percentage ripple ranged from 45% to 82% compared with that of a constant-potential high-voltage generator. With regard to radiation quality, differences of 0.01 to 0.02 mm were found in the half value layer using an aluminum filter. The thicker the x-ray absorber, the more marked the effects of percentage ripple. In terms of the x-ray spectrum, moreover, characteristic x-rays (at 17.4 and 19.5 keV) cannot be effectively used, although a molybdenum target or molybdenum filter is used. Based on these results, a constant potential high-voltage generator with percentage ripple of 4% or less in the tube voltage waveform should be employed for mammography. (author)

  5. Evaluation of nondestructive evaluation size measurement for integrity assessment of axial outside diameter stress corrosion cracking in steam generator tubes

    International Nuclear Information System (INIS)

    Joo, Kyung Mun; Hong, Jun Hee

    2015-01-01

    Recently, the initiation of outside diameter stress corrosion cracking (ODSCC) at the tube support plate region of domestic steam generators (SG) with Alloy 600 HTMA tubes has been increasing. As a result, SGs with Alloy 600 HTMA tubes must be replaced early or are scheduled to be replaced prior to their designed lifetime. ODSCC is one of the biggest threats to the integrity of SG tubes. Therefore, the accurate evaluation of tube integrity to determine ODSCC is needed. Eddy current testing (ECT) is conducted periodically, and its results could be input as parameters for evaluating the integrity of SG tubes. The reliability of an ECT inspection system depends on the performance of the inspection technique and ability of the analyst. The detection probability and ECT sizing error of degradation are considered to be the performance indices of a nondestructive evaluation (NDE) system. This paper introduces an optimized evaluation method for ECT, as well as the sizing error, including the analyst performance. This study was based on the results of a round robin program in which 10 inspection analysts from 5 different companies participated. The analysis of ECT sizing results was performed using a linear regression model relating the true defect size data to the measured ECT size data.

  6. Stress relieving procedure and facility by shot-peening the inside surface of NPP steam generators tubes

    International Nuclear Information System (INIS)

    Banica, I.; Maioru, H.

    1994-01-01

    Residual stress relieving of the transition zones between the deformed part and the non deformed part of the heat exchanger tubes expanded in tube sheets of the NPP equipment, is a technological problem attacked on international level as well as on national level through the continuing programme initiated by ICEMENERG. The most recent statistical data point out that over 75% of tube failures are taking place in the tube-to-tubesheet connection zone, a great number of them being produced in this area by intergranular attack and stress corrosion cracking. The increased occurrence of these incidents is explained first by the existence of residual stresses inside tube surfaces, induced by expanding the tubes. Relieving these residual stresses is the purpose of the outlined procedure and it is achieved by overlapping effects (compression stresses added over tensile stresses). In this paper aspects of the procedure are presented and also a facility is described for stress relieving by introducing compressive stresses from uniform and generalized collisions of the inside surface with micro balls of great kinetic energy carried by a pressurized gas. The stress relieving facility can be acted by remote control, the whole process being completely automatic. The procedure aims to the operation maintenance of the NPP steam generators. (Author)

  7. Eddy current magnetic bias x-probe qualification and inspection of steam generator Monel 400 tubing in Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

    Lepine, B.A.; Van Langen, J.; Obrutsky, L.

    2006-01-01

    This paper presents an overview of the x-probe MB 350 eddy current inspection array probe, for detection of open OD axial crack-like flaws in Monel 400 tubes at Pickering Nuclear Generating Station. This report contains a selection of inspection results from the field inspections performed with this probe during the 2003 and 2004 period at Pickering Nuclear Generating Station A and B. During the 2003 in-service eddy current inspection results of Pickering Nuclear Generating Station A (PNGS-A) Unit 2, a 13 mm (0.5 inch) long axial indication was detected by the CTR1 bobbin and CTR2-C4 array probes in Tube R25-C52 of Steam Generator (SG) 11 in the hot leg sludge pile region. An experimental magnetic bias X-probe, specially designed by Zetec for inspection of Monel 400 tubing, was deployed and the indication was characterized as a potential out diameter (OD) axially oriented crack. Post-inspection tube pulling and destructive examination confirmed the presence of an Environmentally Assisted Crack (EAC), approximately 80% deep and 13mm long. Due to the significance of this discovery, Ontario Power Generation (OPG) requested AECL to initiate a program for qualification of the X-probe MB 350 for the detection of OD axial cracks in medium to high magnetic permeability μ r Monel 400 PNGS-A and B steam generator tubing at different locations. The X-probe MB 350 subsequently has been deployed as a primary inspection probe for crack detection for PNGS steam generators. (author)

  8. Fouling of steam generator tubes in nuclear power plants. Laboratory tests to reproduce oxides deposition and chemical cleanings

    International Nuclear Information System (INIS)

    Goujon, C.; Bescond, A.; Mansour, C.; Delaunay, S.; Pauporte, T.; Bretelle, J-L.

    2014-01-01

    In the secondary circuit of nuclear Pressurized Water Reactors, magnetite (Fe 3 O 4 ) deposits lead to Steam Generators (SG) fouling decreasing thermal performances. As a counteraction, chemical cleanings, started in 1989, have become since 2006 the priority strategy to remove oxides deposited in SG of the EDF fleet. The use of chelating agents in chemical cleaning processes could affect the passive layer of SG tubes, and then modify their surface reactivity. To investigate this impact, a three steps R and D program was established: (1) reproduce deposits on SG tube surfaces using several techniques, (2) apply industrial chemical cleaning procedures and (3) study the redeposition of magnetite. First, SG tubes were fouled in a specific experimental loop, FORTRAND. In this device, magnetite and soluble iron were formed by carbon steel pipes corrosion in feedwater circuit representative conditions and released in the fluid. Then, corrosion products were flow-carried to the autoclave where their precipitation and deposition on heated SG tubes led to tubes fouling. Additional nickel base alloys substrates were also fouled by magnetite electrodeposition. Second, chemical cleaning processes were applied on fouled substrates and tubes in a specific experimental device, ECCLIPS. SG industrial cleaning processes timing and thermochemical conditions were strictly respected during these operations. Finally, fouling of cleaned substrates and tubes was performed in FORTRAND in the same experimental conditions as in the first step. At each step of the study, oxide composition and properties were investigated by surface characterizations. Comparison of oxide deposits before and after cleaning highlights the impact of chemical cleanings on tubes surface reactivity. (author)

  9. Video Captions for Online Courses: Do YouTube's Auto-Generated Captions Meet Deaf Students' Needs?

    Science.gov (United States)

    Parton, Becky Sue

    2016-01-01

    Providing captions for videos used in online courses is an area of interest for institutions of higher education. There are legal and ethical ramifications as well as time constraints to consider. Captioning tools are available, but some universities rely on the auto-generated YouTube captions. This study looked at a particular type of video--the…

  10. Creating a YouTube-Like Collaborative Environment in Mathematics: Integrating Animated Geogebra Constructions and Student-Generated Screencast Videos

    Science.gov (United States)

    Lazarus, Jill; Roulet, Geoffrey

    2013-01-01

    This article discusses the integration of student-generated GeoGebra applets and Jing screencast videos to create a YouTube-like medium for sharing in mathematics. The value of combining dynamic mathematics software and screencast videos for facilitating communication and representations in a digital era is demonstrated herein. We share our…

  11. Chemical treatment of deposits of junctions 'collector-tube' of horizontal steam generators

    International Nuclear Information System (INIS)

    Alkassem, S.N.

    2009-01-01

    A method of chemical treatment of deposits of junctions 'collector - tube' of horizontal steam generators of NPP with reactor VVER has been developed at the department of NPP of the Moscow Power Engineering Institute (Russia). The underexpanding zones of heat-exchanger tubes (HET) in the collector casing, forming fricative gaps with mass-transfer decrease plays a special role in corrosion damage of steam generator collectors. Moreover, if they are filled with porous deposit, then it can become an ideal place of potential concentration of aggressive impurity in the least thermal loading zone: just an accumulation of slime takes place in this zone most intensively. At present, there are series of methods for treating the deposits and for fighting against their formulation. One of the most effective widely used treatment methods is the chemical dissolution. Morpholine or Trilon-B can be used as reagents. The artificially created protective film of ceramic structures made of lithium ferrite on the surfaces of HETs reduces the corrosion-fatigue cracking process rate in the water with the parameters of the second contour. The stability of this film towards dissolution in contact with morpholine was experimentally tested and also a sufficiently durational presence of the film on the pipes' surfaces HETs and ring cracks has been verified by a repeated test. For reinforcing the protective effect, it is needed to maintain the film uniformity in the working process (microdosage of lithium hydroxide-LiOH). On the surface oxidized or polluted with deposits, first of all the LiOH is utilized for interaction with magnetite, then a formation of a coating of a mixed structure of lithium ferrite plus magnetite takes place. Since the combination of magnetite and lithium ferrite is insoluble in water then there is also no transfer of corrosion products in water from the protected surface. This circumstance strongly slows down the deposit formation process. Thus, iron-oxide deposits in

  12. Engaging the YouTube Google-Eyed Generation: Strategies for Using Web 2.0 in Teaching and Learning

    Science.gov (United States)

    Duffy, Peter

    2008-01-01

    YouTube, Podcasting, Blogs, Wikis and RSS are buzz words currently associated with the term Web 2.0 and represent a shifting pedagogical paradigm for the use of a new set of tools within education. The implication here is a possible shift from the basic archetypical vehicles used for (e)learning today (lecture notes, printed material, PowerPoint,…

  13. Accuracy of automatic tube compensation in new-generation mechanical ventilators.

    Science.gov (United States)

    Elsasser, Serge; Guttmann, Josef; Stocker, Reto; Mols, Georg; Priebe, Hans-Joachim; Haberthür, Christoph

    2003-11-01

    To compare performance of flow-adapted compensation of endotracheal tube resistance (automatic tube compensation, ATC) between the original ATC system and ATC systems incorporated in commercially available ventilators. Bench study. University research laboratory. The original ATC system, Dräger Evita 2 prototype, Dräger Evita 4, Puritan-Bennett 840. The four ventilators under investigation were alternatively connected via different sized endotracheal tubes and an artificial trachea to an active lung model. Test conditions consisted of two ventilatory modes (ATC vs. continuous positive airway pressure), three different sized endotracheal tubes (inner diameter 7.0, 8.0, and 9.0 mm), two ventilatory rates (15/min and 30/min), and four levels of positive end-expiratory pressure (0, 5, 10, and 15 cm H2O). Performance of tube compensation was assessed by the amount of tube-related (additional) work of breathing (WOBadd), which was calculated on the basis of pressure gradient across the endotracheal tube. Compared with continuous positive airway pressure, ATC reduced inspiratory WOBadd by 58%, 68%, 50%, and 97% when using the Evita 4, the Evita 2 prototype, the Puritan-Bennett 840, and the original ATC system, respectively. Depending on endotracheal tube diameter and ventilatory pattern, inspiratory WOBadd was 0.12-5.2 J/L with the original ATC system, 1.5-28.9 J/L with the Puritan-Bennett 840, 10.4-21.0 J/L with the Evita 2 prototype, and 10.1-36.1 J/L with the Evita 4 (difference between each ventilator at identical test situations, p ventilator (p <.025). Flow-adapted tube compensation by the original ATC system significantly reduced tube-related inspiratory and expiratory work of breathing. The commercially available ATC modes investigated here may be adequate for inspiratory but probably not for expiratory tube compensation.

  14. Single cells for forensic DNA analysis--from evidence material to test tube.

    Science.gov (United States)

    Brück, Simon; Evers, Heidrun; Heidorn, Frank; Müller, Ute; Kilper, Roland; Verhoff, Marcel A

    2011-01-01

    The purpose of this project was to develop a method that, while providing morphological quality control, allows single cells to be obtained from the surfaces of various evidence materials and be made available for DNA analysis in cases where only small amounts of cell material are present or where only mixed traces are found. With the SteREO Lumar.V12 stereomicroscope and UV unit from Zeiss, it was possible to detect and assess single epithelial cells on the surfaces of various objects (e.g., glass, plastic, metal). A digitally operated micromanipulator developed by aura optik was used to lift a single cell from the surface of evidence material and to transfer it to a conventional PCR tube or to an AmpliGrid(®) from Advalytix. The actual lifting of the cells was performed with microglobes that acted as carriers. The microglobes were held with microtweezers and were transferred to the DNA analysis receptacles along with the adhering cells. In a next step, the PCR can be carried out in this receptacle without removing the microglobe. Our method allows a single cell to be isolated directly from evidence material and be made available for forensic DNA analysis. © 2010 American Academy of Forensic Sciences.

  15. Processing and analysis techniques involving in-vessel material generation

    Science.gov (United States)

    Schabron, John F [Laramie, WY; Rovani, Jr., Joseph F.

    2011-01-25

    In at least one embodiment, the inventive technology relates to in-vessel generation of a material from a solution of interest as part of a processing and/or analysis operation. Preferred embodiments of the in-vessel material generation (e.g., in-vessel solid material generation) include precipitation; in certain embodiments, analysis and/or processing of the solution of interest may include dissolution of the material, perhaps as part of a successive dissolution protocol using solvents of increasing ability to dissolve. Applications include, but are by no means limited to estimation of a coking onset and solution (e.g., oil) fractionating.

  16. Expert system for eddy current signal analysis: non destructive testing of steam generator tubings

    International Nuclear Information System (INIS)

    Benoist, B.

    1991-01-01

    Automatic analysis, by computer, of defect signals in steam generator tubes, based on Eddy current multifrequency technique, is must often inefficient due to pilgrim noise. The first step is to use a method that allows us to eleminate the noise: the adaptative interpolation. Thanks to this method, which ensures reliable data on each channel, the analysis can be realised by taking into account the data corresponding to each basic or mixed channel. By correlating these diverse data, we can class the signals according to two types of defects: single defects (symmetrical), multiple defects (several in the same place). The second step is to use an expert system which allows a reliable diagnosis for whatever family the defect belongs to. According to this classification, analysis is continued and results in the characterization of the defect. The expert system has already been developed with the general purpose application expert system shell SUPER, which is briefly described. The knowledge base (SOCRATE) and the specific tools developed for this application are thoroughly described. The first results obtained with signals corresponding to real defects, that have been recorded in different places, are presented and discussed. The expert system is revealed efficient in all the studied cases, even with signals obtained in very noisy environments [fr

  17. Analysis of steam generator tube rupture as a severe accident using MELCOR 1.8.4

    International Nuclear Information System (INIS)

    Yang Hongrun; Hidaka, Akihide; Sugimoto, Jun

    1999-03-01

    This report presents the results from the MELCOR 1.8.4 calculations for Steam Generator Tube Rupture (SGTR) with stuck open of all the safety valves in faulted SG as a severe accident. The calculations are based on Surry nuclear power plant. After performed using the once-through primary system model alone by 1.0x10 5 s, the calculations were conducted with both of the once-through and the hot leg countercurrent natural circulation models. The results, including event sequences, processes and progressions of core degradation, radionuclides release from core and reactor cavity, and source terms to the environment are described in detail. It is concluded that the availability of High Pressure Safety Injection (HPSI) can significantly delay the progression of core heat-up and approximately 7% of cesium iodide (CsI) can be released to the environment directly through the stuck open safety valve. Comparisons between the results from the two models are also given in this report. The present analyses also showed that during SGTR accident, the hot leg countercurrent natural circulation flow cannot be established well and therefore it has little effect on the mitigation of the core degradation. (author)

  18. Analysis of Density Wave Oscillations in Helically Coiled Tube Once-Through Steam Generator

    Directory of Open Access Journals (Sweden)

    Junwei Hao

    2016-01-01

    Full Text Available Helically coiled tube Once-Through Steam Generator (H-OTSG is one of the key equipment types for small modular reactors. The flow instability of the secondary side of the H-OTSG is particularly serious, because the working condition is in the range of low and medium pressure. This paper presents research on density wave oscillations (DWO in a typical countercurrent H-OTSG. Based on the steady-state calculation, the mathematical model of single-channel system was established, and the transfer function was derived. Using Nyquist stability criterion of the single variable, the stability cases were studied with an in-house computer program. According to the analyses, the impact law of the geometrical parameters to the system stability was obtained. RELAP5/MOD3.2 code was also used to simulate DWO in H-OTSG. The theoretical analyses of the in-house program were compared to the simulation results of RELAP5. A correction factor was introduced to reduce the error of RELAP5 when modeling helical geometry. The comparison results agreed well which showed that the correction is effective.

  19. U-tube steam generator empirical model development and validation using neural networks

    International Nuclear Information System (INIS)

    Parlos, A.G.; Chong, K.T.; Atiya, A.

    1992-01-01

    Empirical modeling techniques that use model structures motivated from neural networks research have proven effective in identifying complex process dynamics. A recurrent multilayer perception (RMLP) network was developed as a nonlinear state-space model structure along with a static learning algorithm for estimating the parameter associated with it. The methods developed were demonstrated by identifying two submodels of a U-tube steam generator (UTSG), each valid around an operating power level. A significant drawback of this approach is the long off-line training times required for the development of even a simplified model of a UTSG. Subsequently, a dynamic gradient descent-based learning algorithm was developed as an accelerated alternative to train an RMLP network for use in empirical modeling of power plants. The two main advantages of this learning algorithm are its ability to consider past error gradient information for future use and the two forward passes associated with its implementation. The enhanced learning capabilities provided by the dynamic gradient descent-based learning algorithm were demonstrated via the case study of a simple steam boiler power plant. In this paper, the dynamic gradient descent-based learning algorithm is used for the development and validation of a complete UTSG empirical model

  20. A study on LMFBR steam generator design without tube failure propagation in water leak events

    International Nuclear Information System (INIS)

    Futagami, Satoshi; Hayafune, Hiroki; Fujimura, Ken; Sato, Mitsuru

    2009-01-01

    The major target performance of the SG for commercialized FBR is not only economic performance but also property protection performance. The candidate SG design will be selected at the end of JFY 2010. The straight double wall tube SG is one of the SG candidates for commercialized FBR, and other SG concepts were studied in this paper. In proposing an alternative SG, alternative technological measures with a double wall tube were investigated and included reinforcing the tube against wastage and quick detection of initial tube leaks. Alternative SG concept candidates for preventing tube failure propagation and mitigation of water leak accidents were proposed through a combination of technological measures. The candidates were then comparatively evaluated from the point of view of property protection performance, total weight, technological issues, and so on. A coated wall tube SG and protective wall tube SG were decided on as the alternative SGs because of superior property protection performance and with the technological issues. At the end of JFY 2010, the straight double wall tube SG will be decided upon as the result of R and D activities, and alternative SGs evaluated in feasibility studies. A plan for studying feasibility with the technological issues of the alternative SG was proposed. (author)

  1. Examination of a steam-generator tube section from the Zorita nuclear plant

    International Nuclear Information System (INIS)

    Stiegelmeyer, W.N.; Agrawal, A.K.

    1986-01-01

    One of the Zorita tubes (R18C43 HL) with the field EC indications on the ID was nondestructively and destructively examined. Also, deposits from the outside of the R18C53 HL were analyzed by several microchemical methods. The results of eddy current, radiography and dye penetrant examinations are given. Axial cracks, up to 80% thru wall, were found on the ID at the roll transition by dye penetrant exams. Similar OD tube sheet region intergranular attack (IGA), tube support plate region IGA and above tube sheet wastage were found as with tube R18C53 HL examined by Westinghouse. The microstructure of the tube R18C43 is shown. The results of a spark source mass spectroscopy (SSMS) analysis of deposits are given. This analysis indicates that the Na/P was >3 in the tube sheet crevice region sample and was <3 for the sample taken from the tube surface. An X-ray diffraction analysis performed on the core and surface samples showed that the deposits were primarily magnetite and copper with a mixture of phosphate compounds. The conclusions of this work are given. Free caustic was considered the likely cause of the OD surface IGA detected. High stresses in the roll transition were likely responsible for the ID axial cracks

  2. OTSGI--a program analysing two-phase flow instabilities in helical tubes of once-through steam generator

    International Nuclear Information System (INIS)

    Shi Shaoping; Zhou Fangde; Wang Maohua

    1998-01-01

    The author has studied the two-phases flow instabilities of the helical tubes of once-through steam generator. Using liner-frequency-domain analytical method, the authors have derived out a mathematical model and designed the program. In this model, the authors also have considered the thermal dynamic characteristics of the tube's wall. The program is used to calculate the threshold of the stability and the influences of some factors, such as entrance throttling coefficient, system pressure, entrance supercooling degree, et al. The outcomes are compared with other studies

  3. Improvement of ISI techniques by multi-frequency eddy current testing method for steam generator tube in PWR plant

    International Nuclear Information System (INIS)

    Endo, Takashi; Kamimura, Takeo; Nishihara, Masatoshi; Araki, Yasuo; Fukui, Shigetaka.

    1982-05-01

    Eddy current flaw detection techniques are applied to the in-service inspection (ISI) of steam generator tubes in pressurized water reactors (PWR) plant. To improve the reliability and operating efficiency of the plants, efforts are being made to develop eddy current testing methods of various kinds. Multi-frequency eddy current testing method, one of new method, has recently been applied to actual heat exchanger tubes, contributing to the improvement of the detectability and signal evaluation of the ISI. The outline of multi-frequency eddy current testing method and its effects on the improvement of flaw detecting and signal evaluation accuracy are described. (author)

  4. The accelerator tube of ions of the generator Van de Graaff of the CEA. Survey of development. First results

    International Nuclear Information System (INIS)

    Bruck, H.; Prevot, F.

    1953-01-01

    Rare are the Van de Graaff supplies whose tube doesn't collapse electrically to tensions and currents very lower to those that the generator can provide. We chose the general measurements: length and diameter, and put the accent on the survey of the individual element, so much to the mechanical viewpoint (installation, solidity, tightness and degassing), that to the electric viewpoint (to increase the electric rigidity of it). After modification the breakdown voltage as well as the performances of the tube have been improved greatly. (M.B.) [fr

  5. Tridimensional finite element stress analysis of the primary side and tube-sheet of a PWR steam generator

    International Nuclear Information System (INIS)

    Xu Dinggeng; Ye Wejuan; Wang Baisong

    1988-08-01

    The results of a tridimensional finite element stress analysis of primary side and tube-sheet of a PWR steam generator is presented. It is subjected to internal pressure load and external load at the safe-end of the nozzle. The interacted effect of different components, maximum peak stress and minimum ligament stress of tube-sheet were obtained. The results of this tridimensional calculation are compared with results of axisymmetrical finite element analysis. At major locations the results have been evaluated in compliance with stress limits of AMSE code section III

  6. Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Fynan, Douglas A.; Ahn, Kwang-Il, E-mail: kiahn@kaeri.re.kr

    2016-12-15

    Highlights: • Pressure drop-flow rate curves for superheated steam in U-tubes were generated. • Forward flow of hot steam is favored in the longer and taller U-tubes. • Reverse flow of cold steam is favored in short U-tubes. • Steam generator U-tube bundle geometry and tube diameter are important. • Need for correlation development for natural convention heat transfer coefficient. - Abstract: Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

  7. Testing the Tube Super-Dielectric Material Hypothesis: Increased Energy Density Using NaCl

    Science.gov (United States)

    Gandy, Jonathan; Cortes, Francisco Javier Quintero; Phillips, Jonathan

    2016-11-01

    The focus of the present work is the evaluation of the low-frequency dielectric performance of titanium dioxide nanotube arrays, created by anodization, filled with aqueous NaCl solutions. At low frequency (ca. capacitors made up of this so-called tube super-dielectric material were found to have extreme dielectric constants, greater than 1 billion. The same capacitors also registered unprecedented energy densities, nearly 400 J/cm3, better than that observed (<250 J/cm3) for the same type of anodized titania filled with an aqueous solution of NaNO3, and about an order of magnitude better than commercial supercapacitors. Sufficient data were collected to propose a correlation relating dielectric thickness and salt concentration to overall energy density.

  8. Ferrocyanide safety program: Final report on adiabatic calorimetry and tube propagation tests with synthetic ferrocyanide materials

    International Nuclear Information System (INIS)

    Fauske, H.F.; Meacham, J.E.; Cash, R.J.

    1995-01-01

    Based on Fauske and Associates, Inc. Reactive System Screening Tool tests, the onset or initiation temperature for a ferrocyanide-nitrate propagating reaction is about 250 degrees Celcius. This is at about 200 degrees Celcius higher than current waste temperatures in the highest temperature ferrocyanide tanks. Furthermore, for current ambient waste temperatures, the tube propagation tests show that a ferrocyanide concentration of 15.5 wt% or more is required to sustain a propagation reaction in the complete absence of free water. Ignoring the presence of free water, this finding rules out propagating reactions for all the Hanford flowsheet materials with the exception of the ferrocyanide waste produced by the original In Farm flowsheet

  9. Flow-induced vibration analysis of Three Mile Island Unit-2 once-through steam generator tubes. Volume 1. Final report

    International Nuclear Information System (INIS)

    Johnson, J.R.; Brown, J.C.; Harris, C.E.; McGuinn, E.J.; Simonis, J.C.; Thoren, D.E.

    1981-06-01

    Tube responses to flow-induced vibration were measured in the top two spans and the tenth span in the B once-through steam generator at Three Mile Island, Unit 2. This program evaluated the effects of flow-induced biration of OTSG tubes during steady-state and transient operation. Twenty-three tubes were instrumented with accelerometers and strain gages in tubes located along the open lane, in the bundle, and at the tenth span. Tube displacements, frequencies, dynamic strains, and mode shapes were determined during steady-state and transient operation. Pressure sensors were installed in the OTSG to measure pressure fluctuations and plant parameters, which were recorded for correlation with tube response. Data analysis results indicate that the steady-state tube response increases with increasing reactor power, with the maximum response (12 mils peak to peak at midspan) at the outer perimeter of the generator in the 16th span

  10. A study on integrity of LMFBR secondary cooling system to hypothetical tube failure propagation in the steam generator

    International Nuclear Information System (INIS)

    Yoshihisa Shindo; Kazuo Haga

    2005-01-01

    Full text of publication follows: A fundamental safety issue of liquid-metal-cooled fast breeder reactor (LMFBR) is to maintain the integrity of the secondary cooling system components against violent chemical sodium-water reaction caused by the water leak from the heat transfer tube of steam generators (SG). The produced sodium-water reaction jet would attack more severely surrounding tubes and would cause other tube failures (tube failure propagation), if it was assumed that the water leak was not detected by function-less detectors and proper operating actions to mitigate the tube failure propagation, such as isolations of the SG from the secondary cooling system and turbine water/steam system, and blowing water and steam inside tubes in the SG, were not taken. This study has been made focusing on the affection of large-scale water leak enlarged due to SG tube failure propagation to the structural integrity of the secondary cooling system because the generated pressure pulse caused by a large-scale sodium-water reaction might break heat transfer tubes of the intermediate heat exchanger (IHX). The present work has been made as one part of the study of probabilistic safety assessment (PSA) of LMFBR, because if the heat-transfer tubes of IHX were failed, the reactor core may be affected by the pressure pulse and/or by the sodium-water reaction products transported through the primary cooling system. As tools for PSA of the water leak incident of SG, we have developed QUARK-LP Version 4 code that mainly analyzes the high temperature rupture phenomena and estimates the number of failed tubes during the middle-scale water leak. The pressure pulse behavior generated by sodium-water reaction in the failure SG and the pressure propagation in the secondary cooling system are calculated by using the SWAAM-2 code developed by ANL. Furthermore, the quasi-steady state high pressure and temperature of the secondary cooling system in a long term is estimated by using the SWAAM

  11. Design And Construction of an Impedance Tube for Measuring Sound Absorptivity and Transmissibility of Materials Using Transfer Function Method

    Science.gov (United States)

    Gowda, Haarish Kapaninaikappa

    Noise is defined as unwanted sound, when perceived in excess can cause many harmful effects such as annoyance, interference with speech, and hearing loss, hence there is a need to control noise in practical situations. Noise can be controlled actively and/or passively, here we discuss the passive noise control techniques. Passive noise control involves using energy dissipating or reflecting materials such as absorbers or barriers respectively. Damping and isolating materials are also used in eliminating structure-borne noise. These materials exhibit properties such as reflection, absorption and transmission loss when incidence is by a sound source. Thus, there is a need to characterize the acoustical properties of these materials for practical use. The theoretical background of the random incident sound absorption with reverberation room and normal incident sound absorption using impedance tube are well documented. The Transfer Matrix method for measuring transmission loss and absorption coefficient using impedance tube is very attractive since it is rather inexpensive and fast. In this research, a low-cost Impedance Tube is constructed using transfer function method to measure both absorption and transmissibility of materials. Equipment and measurement instruments available in the laboratory were used in the construction of the tube, adhering to cost-effectiveness. Care has been taken for precise construction of tube to ensure better measurement results. Further various samples varying from hard non-porous to soft porous materials were tested for absorption and sound transmission loss. Absorption values were also compared with reverberation room method with the available samples further ensuring the reliability of the newly constructed tube for future measurements.

  12. Secondary side corrosion in steam generator tubes: lessons learned in France from the in-service inspection results

    International Nuclear Information System (INIS)

    Comby, R.

    1997-01-01

    Non-destructive testing (NDT) has proved to be very important in the maintenance of steam generator tubing. This is particularly true in the case of secondary side corrosion, because this type of degradation leads to various morphologies which are often complex (intergranular attack) (IGA), intergranular stress corrosion cracking (IGSCC), or a mixture of both. Their detection and characterization by the usual NDT techniques have been achieved through numerous laboratory studies, which were conducted in order to determine the performance and limitations of NDT. Pulled tube examination in a hot laboratory was very valuable, for both NDT and fracture mechanics aspects. The eddy current bobbin coil probe, used for multipurpose inspection of tubes, allows the detection of IGA-SCC at the tube support plate elevation. In France, the use of rotating probes is not required for that type of degradation, since the repair criterion is based on bobbin coil results only. The bobbin coil is also used for detection of IGSCC occurring in free spans, within sludge deposits. The eddy current rotating probe allows, in that case, characterization of main cracks. Concerning the outer diameter initiated circumferential cracks which occur at the top of the tube sheet, only the rotating probe is used. An ultrasonic (UT) inspection was performed several times, in order to obtain information on UT capabilities. The goal of tube inspection is obviously knowledge of the status of steam generators, but also to follow up degradations and to estimate their revolution, and to verify the beneficial effect of some corrective measures, e.g. boric acid injection. (orig.)

  13. STEAM GENERATOR TUBE INTEGRITY ANALYSIS OF A TOTAL LOSS OF ALL HEAT SINKS ACCIDENT FOR WOLSONG NPP UNIT 1

    Directory of Open Access Journals (Sweden)

    HEOK-SOON LIM

    2014-02-01

    Full Text Available A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS and the steam generator (SG secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

  14. Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo [Korea Htydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Kim, Seoungrae [Nuclear Engineering Service and Solution, Daejeon (Korea, Republic of)

    2014-02-15

    A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

  15. Integrity evaluation for steam generator tube of system integrated modular advanced reactor

    International Nuclear Information System (INIS)

    Kim, J. S.; Jin, T. E.; Jeong, M. J.; Choi, Y. H.; Jeo, J. C.

    2003-01-01

    In this study, the structural integrity for SG tube of system integrated modular advanced reactor, which is subjected to dominant external pressure as well as helical type, is evaluated using the commercial finite element package ABAQUS and the American petrochemical industry code API 579 Appendix B. First of all, the crack behavior under the assumption of local heating is assessed using ABAQUS. And, the buckling behavior of tube with 40% wall thinning is assessed using API 579 Appendix B. As a result, it is found that the crack closure phenomenon occurs under external pressure and the buckling doesn't occur even if 40% wall thinning exists in tube

  16. Overview of magnetic bias X-probe qualification and inspection of PNGS Monel 400 steam generator tubing

    International Nuclear Information System (INIS)

    Lepine, B.A.; Van Langen, J.; Obrutsky, L.

    2006-01-01

    This paper presents an overview of the X-probe MB 350, the qualification for detection of open OD axial crack-like flaws, and a selection of inspection results from the subsequent field inspections performed with this probe during the 2003 and 2004 period at Pickering Nuclear Generating Station A and B. Examples of the field indications to be presented are axial cracking, OD pitting at top of tubesheet location (TTS), and flow assisted corrosion (top hats). During the 2003 in-service eddy current inspection results of Pickering Nuclear Generating Station A (PNGS-A) Unit 2, a 13 mm (0.5 inch) long axial indication was detected by the CTR1 bobbin and CTR2-C4 array probes in Tube R25-C52 of Steam Generator (SG) 11 in the hot leg sludge pile region. An experimental magnetic bias X-probe, especially designed by Zetec for inspection of Monel 400 tubing, was deployed and the indication was characterized as a potential outer diameter (OD) axially oriented crack. Posterior tube pulling and destructive examination confirmed the presence of an Environmentally Assisted Crack (EAC), approximately 80% deep and 13 mm long. Due to the significance of this discovery, Ontario Power Generation (OPG) requested AECL to initiate a program for qualification of the X-Probe MB 350 for the detection of OD axial cracks in medium to high magnetic permeability (μ r ) Monel 400 PNGS-A and B steam generator tubing at different locations. The X-probe MB 350 subsequently has been deployed as a primary inspection probe for crack detection for PNGS steam generators. (author)

  17. Structural materials for the next generation of technologies

    CERN Document Server

    Van de Voorde, Marcel Hubert

    1996-01-01

    1. Overview of advanced technologies; i.e. aerospace-aeronautics; automobile; energy technology; accelerator engineering etc. and the need for new structural materials. 2. Familiarisation with polymers, metals and alloys, structural ceramics, composites and surface engineering. The study of modern materials processing, generation of a materials data base, engineering properties includind NDE, radiation damage etc. 3. Development of new materials for the next generation of technologies; including the spin-off of materials developed for space and military purposes to industrial applications. 4. Materials selection for modern accelerator engineering. 5. Materials research in Europe, USA and Japan. Material R & D programmes sponsored by the European Union and the collaboration of CERN in EU sponsored programmes.

  18. In service inspection of steam generator tubes with a multifrequence eddy current apparatus

    International Nuclear Information System (INIS)

    Pigeon, M.; Saglio, R.

    1976-01-01

    The described multifrequency device has been designed and developed for a complete testing of tubes. It is necessary to eliminate the signals given by plates, expanded parts, background noises and magnetic signals which can mask some possible defects [fr

  19. Thermal-hydraulic analysis of SMART steam generator tube rupture using TASS/SMR-S code

    International Nuclear Information System (INIS)

    Kim, Hee-Kyung; Kim, Soo Hyoung; Chung, Young-Jong; Kim, Hyeon-Soo

    2013-01-01

    Highlights: ► The analysis was performed from the viewpoint of primary coolant leakage. ► The thermal hydraulic responses and the maximum leakage have been identified. ► There is no direct release into the atmosphere caused by an SGTR accident. ► SMART safety system works well against an SGTR accident. - Abstract: A steam generator tube rupture (SGTR) accident analysis for SMART was performed using the TASS/SMR-S code. SMART with a rated thermal power of 330 MWt has been developed at the Korea Atomic Energy Research Institute. The TASS/SMR-S code can analyze the thermal hydraulic phenomena of SMART in a full range of reactor operating conditions. An SGTR is one of the most important accidents from a thermal hydraulic and radiological viewpoint. A conservative analysis against a SMART SGTR was performed. The major concern of this analysis is to find the thermal hydraulic responses and maximum leakage amount from a primary to a secondary side caused by an SGTR accident. A sensitivity study searching for the conservative thermal hydraulic conditions, break locations, reactivity and other conditions was performed. The dominant parameters related with the integral leak are the high RCS pressure, low core inlet coolant temperature and low break location of the SG cassette. The largest integral leak comes to 28 tons in the most conservative case during 1 h. But there is no direct release into the atmosphere because the secondary system pressure is maintained with a sufficient margin for the design pressure. All leaks go to the condenser. The analysis results show that the primary and secondary system pressures are maintained below the design pressure and the SMART safety system is working well against an SGTR accident

  20. Development of ultrasonic testing scanner for NPP steam generator tubes (I)

    Energy Technology Data Exchange (ETDEWEB)

    Shin, J. I.; Huh, H

    1998-12-01

    Testing tubes are designed and fabricated to investigate the optimum test conditions through the various experiments. The proto-type P/C-controlled automatic rotating scanner is fabricated to obtain the ultrasonic data automatically from test tubes. It was attempted to visualize the shape of defects presented inside the specimen using peak amplitude at each point. However, further research works will be needed to be applied at the plant site as a more reliable technology.

  1. Development of ultrasonic testing scanner for NPP steam generator tubes (I)

    International Nuclear Information System (INIS)

    Shin, J. I.; Huh, H.

    1998-12-01

    Testing tubes are designed and fabricated to investigate the optimum test conditions through the various experiments. The proto-type P/C-controlled automatic rotating scanner is fabricated to obtain the ultrasonic data automatically from test tubes. It was attempted to visualize the shape of defects presented inside the specimen using peak amplitude at each point. However, further research works will be needed to be applied at the plant site as a more reliable technology

  2. Effect of beta phase composition and surface machining on the oxidation behavior of Zr-2.5Nb pressure tube material

    International Nuclear Information System (INIS)

    Nouduru, S.K.; Kiran Kumar, M.; Kain, V.; Khanna, A.S.

    2015-01-01

    Zr-2.5Nb is commonly used as the pressure tube material in pressurized heavy water reactors. it is also the pressure tube material for Advanced Heavy Water Reactor (AHWR) being developed indigenously in India with light water as coolant and water chemistry similar to Boiling Water Reactors (BWR). Oxidation of the pressure tube depends on various factors like material composition, microstructure, fabrication route, and water chemistry. In the present research, the role of the composition and morphology of second phase β on the high temperature and pressure oxidation behavior of Zr-2.5Nb pressure tube material in steam was systematically studied. The as-received pressure tube material (fabricated through cold worked and stress relieved, CWSR route) was subjected to selective heat treatments to generate microstructures containing predominantly β(Zr) (∼ 20% Nb) and β(Nb) (∼ 80% Nb) phases. The presence of such phases was characterized by X-ray diffraction and transmission electron microscopy-energy dispersive spectroscopy. Subsequently both the heat treated materials were subjected to surface machining. The Zr-2.5Nb material in different microstructural conditions was subjected to accelerated oxidation exposures in steam at 400 C. degrees, and 10 MPa pressure up to 30 days. Raman spectroscopy was carried out on the oxide surfaces to observe the variation in tetragonal versus monoclinic phase fractions with oxidation duration. The microstructure consisting of predominantly β(Nb) showed a relatively improved oxidation resistance as compared to the one with predominantly β(Zr). The tetragonal phase fraction in the oxide film decreased with oxidation time in all microstructural conditions and was found to be the least in the microstructure containing β(Zr) after 10 days of exposures. The explanation for the observed higher oxidation resistance of β(Nb) microstructure lies in the context of depleted matrix Nb content in the case of β(Nb). Surface machining

  3. 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

  4. Development of tube rupture evaluation code for FBR steam generator (II). Modification of heat transfer model in sodium side

    International Nuclear Information System (INIS)

    Hamada, H.; Kurihara, A.

    2003-05-01

    The thermal effect of sodium-water reaction jet on neighboring heat transfer tubes was examined to rationally evaluate the structural integrity of the tube for overheating rupture under a water leak in an FBR steam generator. Then, the development of new heat transfer model and the application analysis were carried out. Main results in this paper are as follows. (1) The evaluation method of heat flux and heat transfer coefficient (HTC) on the tube exposed to reaction jet was developed. By using the method, it was confirmed that the heat flux could be realistically evaluated in comparison with the previous method. (2) The HTC between reaction jet and the tube was theoretically examined in the two-phase flow model, and new heat transfer model considering the effect of fluid temperature and cover gas pressure was developed. By applying the model, a tentative experimental correlation was conservatively obtained by using SWAT-1R test data. (3) The new model was incorporated to the Tube Rupture Evaluation Code (TRUE), and the conservatism of the model was confirmed by using sodium-water reaction data such as the SWAT-3 tests. (4) In the application analysis of the PFR large leak event, there was no significant difference of calculation results between the new model and previous one; the importance of depressurization in the tube was confirmed. (5) In the application analysis of the Monju evaporator, it was confirmed that the calculation result in the previous model would be more conservative than that in the new one and that the maximum cumulative damage of 25% could be reduced in the new model. (author)

  5. Experiment data report for Semiscale Mod-1 test S-28-3 (steam generator tube rupture test). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gillins, R.L.; Sackett, K.E.

    1977-10-01

    Recorded test data are presented for Test S-28-3 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-3 was conducted from initial conditions of 15621 kPa and 555 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Twelve steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg.

  6. Experiment data report for Semiscale Mod-1 Test S-28-1 (steam generator tube rupture test series)

    Energy Technology Data Exchange (ETDEWEB)

    Collins, B.L.; Coppin, C.E.; Sackett, K.E.

    1977-10-01

    Recorded test data are presented for Test S-28-1 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-1 was conducted from initial conditions of 15 767 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Sixty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg.

  7. Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

    Energy Technology Data Exchange (ETDEWEB)

    Esparza, V.; Sackett, K.E.

    1977-10-01

    Recorded test data are presented for Test S-28-4 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-4 was conducted from initial conditions of 15 646 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Thirty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg.

  8. Experiment data report for Semiscale Mod-1 test S-28-6 (steam generator tube rupture test)

    Energy Technology Data Exchange (ETDEWEB)

    Patton, M.L.; Sackett, K.E.; Coppin, C.E.

    1977-11-01

    Recorded test data are presented for Test S-28-6 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-6 was conducted from initial conditions of 15,770 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Sixteen steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg.

  9. Experiment data report for semiscale Mod-1 test S-28-2 (steam generator tube rupture test). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Patton, M.L.; Sackett, K.E.

    1977-10-01

    Recorded test data are presented for Test S-28-2 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-2 was conducted from initial conditions of 15 936 kPa and 558 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. For Test S-28-2, accumulator injection into the intact loop hot leg was provided to simulate simulate the rupture of six steam generator tubes.

  10. Experiment data report for Semiscale Mod-1 Test S-28-5 (steam generator tube rupture test). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    1977-11-01

    Recorded test data are presented for Test S-28-5 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-5 was conducted from initial conditions of 15,768 kPa and 556 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. For Test S-28-5, accumulator injection into the intact loop hot leg was provided to simulate the rupture of 20 steam generator tubes.

  11. Critical heat flux and transition boiling characteristics for a sodium-heated steam generator tube for LMFBR applications

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, S.; Holmes, D.H.

    1977-04-01

    An experimental program was conducted to characterize critical heat flux (CHF) in a sodium-heated steam generator tube model at a proposed PLBR steam generator design pressure of 7.2 MPa. Water was circulated vertically upward in the tube and the heating sodium was flowing counter-current downward. The experimental ranges were: mass flux, 110 to 1490 kg/s.m/sup 2/ (0.08 to 1.10 10/sup 6/ lbm/h.ft/sup 2/); critical heat flux, 0.16 to 1.86 MW/m/sup 2/ (0.05 to 0.59 10/sup 6/ Btu/h.ft/sup 2/); and critical quality, 0.48 to 1.0. The CHF phenomenon for the experimental conditions is determined to be dryout as opposed to departure from nucleate boiling (DNB). The data are divided into high- and low-mass flux regions.

  12. Review—Organic Materials for Thermoelectric Energy Generation

    KAUST Repository

    Cowen, Lewis M.

    2017-01-29

    Organic semiconductor materials have been promising alternatives to their inorganic counterparts in several electronic applications such as solar cells, light emitting diodes, field effect transistors as well as thermoelectric generators. Their low cost, light weight and flexibility make them appealing in future applications such as foldable electronics and wearable circuits using printing techniques. In this report, we present a mini-review on the organic materials that have been used for thermoelectric energy generation.

  13. Material challenges for the next generation of fission reactor systems

    International Nuclear Information System (INIS)

    Buckthorpe, Derek

    2010-01-01

    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO 2 emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  14. Material challenges for the next generation of fission reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Buckthorpe, Derek [AMEC, Knutsford, Cheshire (United Kingdom)

    2010-07-01

    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO{sub 2} emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  15. Analysis of Communication between Main Control Room Operators in Decision-making Process in Steam Generator Tube Rupture Accident

    International Nuclear Information System (INIS)

    Petkov, M.; Petkov, G.

    2006-01-01

    The paper presents an investigation results for Main Control Room operators' reliability by Performance Evaluation of Teamwork method, based on FSS-1000 training archives in KNPP in case of Steam Generator Tube Rupture accident. The advantages of operators' teamwork are shown: a) group decision-making vs. individual one: b) positive influence of crew initiated communication consisting of orders and reports that are required by instruction. (authors)

  16. Data analysis algorithms for flaw sizing based on eddy current rotating probe examination of steam generator tubes

    International Nuclear Information System (INIS)

    Bakhtiari, S.; Elmer, T.W.

    2009-01-01

    Computer-aided data analysis tools can help improve the efficiency and reliability of flaw sizing based on nondestructive examination data. They can further help produce more consistent results, which is important for both in-service inspection applications and for engineering assessments associated with steam generator tube integrity. Results of recent investigations at Argonne on the development of various algorithms for sizing of flaws in steam generator tubes based on eddy current rotating probe data are presented. The research was carried out as part of the activities under the International Steam Generator Tube Integrity Program (ISG-TIP) sponsored by the U.S. Nuclear Regulatory Commission. A computer-aided data analysis tool has been developed for off-line processing of eddy current inspection data. The main objectives of the work have been to a) allow all data processing stages to be performed under the same user interface, b) simplify modification and testing of signal processing and data analysis scripts, and c) allow independent evaluation of viable flaw sizing algorithms. The focus of most recent studies at Argonne has been on the processing of data acquired with the +Point probe, which is one of the more widely used eddy current rotating probes for steam generator tube examinations in the U.S. The probe employs a directional surface riding differential coil, which helps reduce the influence of tubing artifacts and in turn helps improve the signal-to-noise ratio. Various algorithms developed under the MATLAB environment for the conversion, segmentation, calibration, and analysis of data have been consolidated within a single user interface. Data acquired with a number of standard eddy current test equipment are automatically recognized and converted to a standard format for further processing. Because of its modular structure, the graphical user interface allows user-developed routines to be easily incorporated, modified, and tested independent of the

  17. One-Tube-Only Standardized Site-Directed Mutagenesis: An Alternative Approach to Generate Amino Acid Substitution Collections.

    Directory of Open Access Journals (Sweden)

    Janire Mingo

    Full Text Available Site-directed mutagenesis (SDM is a powerful tool to create defined collections of protein variants for experimental and clinical purposes, but effectiveness is compromised when a large number of mutations is required. We present here a one-tube-only standardized SDM approach that generates comprehensive collections of amino acid substitution variants, including scanning- and single site-multiple mutations. The approach combines unified mutagenic primer design with the mixing of multiple distinct primer pairs and/or plasmid templates to increase the yield of a single inverse-PCR mutagenesis reaction. Also, a user-friendly program for automatic design of standardized primers for Ala-scanning mutagenesis is made available. Experimental results were compared with a modeling approach together with stochastic simulation data. For single site-multiple mutagenesis purposes and for simultaneous mutagenesis in different plasmid backgrounds, combination of primer sets and/or plasmid templates in a single reaction tube yielded the distinct mutations in a stochastic fashion. For scanning mutagenesis, we found that a combination of overlapping primer sets in a single PCR reaction allowed the yield of different individual mutations, although this yield did not necessarily follow a stochastic trend. Double mutants were generated when the overlap of primer pairs was below 60%. Our results illustrate that one-tube-only SDM effectively reduces the number of reactions required in large-scale mutagenesis strategies, facilitating the generation of comprehensive collections of protein variants suitable for functional analysis.

  18. Fiber Fabry-Perot Force Sensor with Small Volume and High Performance for Assessing Fretting Damage of Steam Generator Tubes.

    Science.gov (United States)

    Huang, Peijian; Wang, Ning; Li, Junying; Zhu, Yong; Zhang, Jie

    2017-12-13

    Measuring the radial collision force between the steam generator tube (SGT) and the tube support plate (TSP) is essential to assess the fretting damage of the SGT. In order to measure the radial collision force, a novel miniaturized force sensor based on fiber Fabry-Perot (F-P) was designed, and the principle and characteristics of the sensor were analyzed in detail. Then, the F-P force sensor was successfully fabricated and calibrated, and the overall dimensions of the encapsulated fiber F-P sensor were 17 mm × 5 mm × 3 mm (L × W × H). The sensor works well in humid, high pressure (10 MPa), high temperature (350 °C), and vibration (40 kHz) environments. Finally, the F-P force sensors were installed in a 1:1 steam generator test loop, and the radial collision force signals between the SGT and the TSP were obtained. The experiments indicated that the F-P sensor with small volume and high performance could help in assessing the fretting damage of the steam generator tubes.

  19. Experience in ultrasonic gap measurement between calandria tubes and liquid injection shutdown systems nozzles in Bruce Nuclear Generating Station

    International Nuclear Information System (INIS)

    Abucay, R.C.; Mahil, K.S.; Goszczynski, J.J.

    1995-01-01

    The gaps between calandria tubes (CT) and Liquid Injection Shutdown System (LISS) nozzles at the Bruce Nuclear Generating Station ''A'' (Bruce A) are known to decrease with time due to radiation induced creep/sag of the calandria tubes. If this gap decreases to a point where the calandria tubes come into contact with the LISS nozzle, the calandria tubes could fail as a result of fretting damage. Proximity measurements were needed to verify the analytical models and ensure that CT/LISS nozzle contact does not occur earlier than predicted. The technique used was originally developed at Ontario Hydro Technologies (formerly Ontario Hydro Research Division) in the late seventies and put into practical use by Research and Productivity Council (RPC) of New Brunswick, who carried out similar measurements at Point Lepreau NGS in 1989 and 1991. The gap measurement was accomplished y inserting an inspection probe, containing four ultrasonic transducers (2 to measure gaps and 2 to check for probe tilt) and a Fredericks electrolytic potentiometer as a probe rotational sensor, inside LISS Nozzle number-sign 7. The ultrasonic measurements were fed to a system computer that was programmed to convert the readings into fully compensated gaps, taking into account moderator heavy water temperature and probe tilt. Since the measured gaps were found to be generally larger than predicted, the time to CT/LISS nozzle contact is now being re-evaluated and the planned LISS nozzle replacement will likely be deferred, resulting in considerable savings

  20. Development of an Acoustic Impedance Tube Testbed for Material Sample Testing

    Science.gov (United States)

    Doty, Benjamin J.; Kolaini, Ali R.

    2012-01-01

    Acoustic impedance tube method: uses Traveling wave amplitudes are measured on either side of a sample in a tube. Many acoustic properties of the sample can be calculated. It is Simple and inexpensive to set up, ideal for high volume optimization tests

  1. A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Kim, Sung Hyun; Moon, Chan Ki; Park, Sung Baek; Na, Man Gyun

    2013-01-01

    The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study

  2. An Experimental Research on Uniform Corrosion of Inconel 600 and 690 Tubing Material

    International Nuclear Information System (INIS)

    Yeom, Yu Sun; Hwang, Jung Lae; Jun, In Sub; Kim, Soong Pyung; Yoon, Jang Hee

    2006-01-01

    By executing corrosion experiment on Inconel 600, 690 used to material of S/G tube in domestic NPP, this paper show estimation of amount of product such as Co-58, Co-60, Cr-51, Mn-54, Fe-59 which are main exposure cause to the workers in NPP. Therefore, Making the 12 samples consisted of Inconel 600, 690, whole corrosion experiment was carried out for 60 days(each pH by 20 days). The conditions of those tests were similar or more harsh than actual conditions of domestic NPP. The Glow Discharge Spectrometer(GDS) was used for quantitative analysis of results. The results of using GDS, the Inconel 600 corrodes more than Inconel 690 at pH 7 and pH 9. However, it is observed that Inconel 690 corrodes more than Inconel 600 at pH 4. Those results is estimated that test sections had the effect of transient. The long terms of experiment is required to minimize and solve the problem.

  3. Viability of use of PVC tubes in solar collectors: an analysis of materials

    Directory of Open Access Journals (Sweden)

    Luiz Guillherme Meira de Souza

    2003-06-01

    Full Text Available This paper presents a study of the inherent degradations of PVC tubes due to the thermal effect and ultraviolet solar radiation. The approach relates its causes and its effect of use of the PVC tubes as elements to absorption, forming a coil, in solar collectors for water heating. It is demonstrated that such degradations can be burst through the use of an outflow and an appropriate regimen of work, as well as of a protective layer for the tubes, in this case black ink used to magnify its absorption. The results of the properties of tubes that had been exposed to the degradation effect for up to five years are presented. The viability of use of this type of collector is demonstrated through comparative analysis of tubes exposed and not exposed to the sun, concluding for the low cost, easy assembly and maintenance of the system.

  4. Improve the material absorption of light and enhance the laser tube bending process utilizing laser softening heat treatment

    Science.gov (United States)

    Imhan, Khalil Ibraheem; Baharudin, B. T. H. T.; Zakaria, Azmi; Ismail, Mohd Idris Shah B.; Alsabti, Naseer Mahdi Hadi; Ahmad, Ahmad Kamal

    2018-02-01

    Laser forming is a flexible control process that has a wide spectrum of applications; particularly, laser tube bending. It offers the perfect solution for many industrial fields, such as aerospace, engines, heat exchangers, and air conditioners. A high power pulsed Nd-YAG laser with a maximum average power of 300 W emitting at 1064 nm and fiber-coupled is used to irradiate stainless steel 304 (SS304) tubes of 12.7 mm diameter, 0.6 mm thickness and 70 mm length. Moreover, a motorized rotation stage with a computer controller is employed to hold and rotate the tube. In this paper, an experimental investigation is carried out to improve the laser tube bending process by enhancing the absorption coefficient of the material and the mechanical formability using laser softening heat treatment. The material surface is coated with an oxidization layer; hence, the material absorption of laser light is increased and the temperature rapidly rises. The processing speed is enhanced and the output bending angle is increased to 1.9° with an increment of 70% after the laser softening heat treatment.

  5. Improvement of Eddy Current testing methods of steam generator tubings due to field experience

    International Nuclear Information System (INIS)

    Comby, R.; Meurgey, P.; David, B.

    1985-01-01

    This paper presents the main stages of the long rotating probe developed by EDF, this probe detects stress corrosion cracks. The method has been validated by the examination of numerous cracked tubes that the probe detected before. Methods to better characterize the signals with regard to the defects are being improved to avoid a complementary examination of the rolling zone more particularly [fr

  6. YouTube as a Qualitative Research Asset: Reviewing User Generated Videos as Learning Resources

    Science.gov (United States)

    Chenail, Ronald J.

    2011-01-01

    YouTube, the video hosting service, offers students, teachers, and practitioners of qualitative researchers a unique reservoir of video clips introducing basic qualitative research concepts, sharing qualitative data from interviews and field observations, and presenting completed research studies. This web-based site also affords qualitative…

  7. Importance of crevices formed between tubes and tube plate for the operational behaviour of heat exchangers

    International Nuclear Information System (INIS)

    Achten, N.; Herbsleb, G.; Wieling, N.

    1986-01-01

    It must be guaranteed by construction and manufacture of heat exchangers that primary and secondary medium are completely separated from each other. When this requirement is fullfilled, the operational use of heat exchangers can be impaired by corrosion reactions within the crevice formed between tube and tube plate which may result in corrosion damage. The various techniques which are in use to connect tubes and tube plate and which are described in the present report, must be valued with respect to the tightness of the connection as well as to the formation of crevices between tubes and tube plate. Corrosion resistant copperbase alloys and stainless steels are the most important materials which are in use for the construction of heat exchangers. The mechanisms of crevice corrosion with unalloyed and low alloy carbon steels, stainless steels, and mixed connections between tube and tube plate with these materials are described in detail. Crevice corrosion may be caused also by the formation of galvanic cells between materials of differing electrochemical response. Furthermore, the concentration of aggressive media in crevices between tubes and tube plate can lead to corrosion damage of heat exchanger tubes. For the service operation of heat exchangers without any hazard of corrosion damage in crevices between tubes and tube plate, such crevices must be avoided by proper construction and manufacture. As a model for suitable measures to avoid crevices, the manufacture of steam generators for PWR's is described. (orig.) [de

  8. Thermomechanical Model and Bursting Tests to Evaluate the Risk of Swelling and Bursting of Modified 9Cr-1Mo Steel Steam Generator Tubes during a Sodium-Water Reaction Accident

    Directory of Open Access Journals (Sweden)

    C. Bertrand

    2014-01-01

    Full Text Available The MECTUB code was developed to evaluate the risk of swelling and bursting of Steam Generator (SG tubes. This code deals with the physic of intermediate steam-water leaks into sodium which induce a Sodium-Water Reaction (SWR. It is based on a one-dimensional calculation to describe the thermomechanical behavior of tubes under a high internal pressure and a fast external overheating. The mechanical model of MECTUB is strongly correlated with the kind of the material of the SG tubes. It has been developed and validated by using experiments performed on the alloy 800. A change to tubes made of Modified 9Cr-1Mo steel requires more knowledge of Modified 9Cr-1Mo steel behavior which influences the bursting time at high temperatures (up to 1200°C. Studies have been initiated to adapt the mechanical model and to qualify it for this material. The first part of this paper focuses on the mechanical law modelling (elasticity, plasticity, and creep for Modified 9Cr-1Mo steel and on overheating thermal data. In a second part, the results of bursting tests performed on Modified 9Cr-1Mo tubes in the SQUAT facility of CEA are used to validate the mechanical model of MECTUB for the Modified 9Cr-1Mo material.

  9. Portable Thermoelectric Power Generator Coupled with Phase Change Material

    Directory of Open Access Journals (Sweden)

    Lim Chong C.

    2014-07-01

    Full Text Available Solar is the intermittent source of renewable energy and all thermal solar systems having a setback on non-functioning during the night and cloudy environment. This paper presents alternative solution for power generation using thermoelectric which is the direct conversion of temperature gradient of hot side and cold side of thermoelectric material to electric voltage. Phase change material with latent heat effect would help to prolong the temperature gradient across thermoelectric material for power generation. Besides, the concept of portability will enable different power source like solar, wasted heat from air conditioner, refrigerator, stove etc, i.e. to create temperature different on thermoelectric material for power generation. Furthermore, thermoelectric will generate direct current which is used by all the gadgets like Smartphone, tablet, laptop etc. The portable concept of renewable energy will encourage the direct usage of renewable energy for portable gadgets. The working principle and design of portable thermoelectric power generator coupled with phase change material is presented in this paper.

  10. Numerical simulations of eddy current testing signals of steam generator tubes by 3-D finite element method

    International Nuclear Information System (INIS)

    Sakai, Takayuki; Soneda, Naoki

    1996-01-01

    In every inspection of Japanese PWR plants, all of steam generator tubes are inspected using Eddy Current Testing (ECT) method. However, the relationships between the ECT signals and the defect shapes are known only for the representative shapes of defects. In order to improve the reliability of inspections and the capability of ECT probes, development of numerical simulation technique of the ECT signals for arbitrarily shaped defects is essential. In this study, three-dimensional finite element code is developed to simulate the ECT signals for any kinds of defects in the SG tubes. The code is fully vectorized so that it runs on the supercomputers very efficiently. The simulation results agree very well with the experimental results. Sensitivity analyses are performed to investigate the relationships between the defect shapes and the ECT signals. (author)

  11. An analysis of signal characteristics due to coil-gap variation of ECT bobbin probe for steam generation tube

    International Nuclear Information System (INIS)

    Nam, Min Woo; Cho, Chan Hee; Jee, Dong Hyun; Jung, Jee Hong; Lee, Hee Jong

    2006-01-01

    The bobbin probe technique is basically one of the important ECT methods for the steam generator tube integrity assesment that is practiced during each plant outage. The bobbin probe is one of the essential components which consist of the whole ECT examination system, and provides us a decisive data for the evaluation of tube integrity in compliance with acceptance criteria described in specific procedures. The selection of examination probe is especially important because the quality of acquired ECT data is determined by the probe design characteristics, geometry and operation frequencies, and has an important effect on examination results. In this study, the relationship between electric characteristic changes and differential coil gap variation has been investigated to optimize the ECT signal characteristics of the bobbin probe. With the results from this study, we have elucidated that the optimum coil gap is 1.2 - 1.6 mm that give the best result for O.D. volumetric defects in ASME calibration standards.

  12. Consumer-generated Advertising on YouTube : A quantitative study examining the effects of endorser credibility and coupon proneness on brands

    OpenAIRE

    Jonsson Brajim, Rahel; Romanov, Tina

    2016-01-01

    Consumer-generated advertising on YouTube is a developing phenomenon that in the last years has grown exponentially. Recent research by Holt (2016) suggests that well-known content creators on YouTube have a large impact on brands, so great that regular firms are unable to compete with these well-known content creators. However, firms do have the opportunity to cooperate with well-known content creators on YouTube. Therefore, examining YouTube as a marketing tool is beneficial for marketing m...

  13. Materials-based process tolerances for neutron generator encapsulation.

    Energy Technology Data Exchange (ETDEWEB)

    Berry, Ryan S.; Adolf, Douglas Brian; Stavig, Mark Edwin

    2007-10-01

    Variations in the neutron generator encapsulation process can affect functionality. However, instead of following the historical path in which the effects of process variations are assessed directly through functional tests, this study examines how material properties key to generator functionality correlate with process variations. The results of this type of investigation will be applicable to all generators and can provide insight on the most profitable paths to process and material improvements. Surprisingly, the results at this point imply that the process is quite robust, and many of the current process tolerances are perhaps overly restrictive. The good news lies in the fact that our current process ensures reproducible material properties. The bad new lies in the fact that it would be difficult to solve functional problems by changes in the process.

  14. Advanced Material Strategies for Next-Generation Additive Manufacturing

    Science.gov (United States)

    Chang, Jinke; He, Jiankang; Zhou, Wenxing; Lei, Qi; Li, Xiao; Li, Dichen

    2018-01-01

    Additive manufacturing (AM) has drawn tremendous attention in various fields. In recent years, great efforts have been made to develop novel additive manufacturing processes such as micro-/nano-scale 3D printing, bioprinting, and 4D printing for the fabrication of complex 3D structures with high resolution, living components, and multimaterials. The development of advanced functional materials is important for the implementation of these novel additive manufacturing processes. Here, a state-of-the-art review on advanced material strategies for novel additive manufacturing processes is provided, mainly including conductive materials, biomaterials, and smart materials. The advantages, limitations, and future perspectives of these materials for additive manufacturing are discussed. It is believed that the innovations of material strategies in parallel with the evolution of additive manufacturing processes will provide numerous possibilities for the fabrication of complex smart constructs with multiple functions, which will significantly widen the application fields of next-generation additive manufacturing. PMID:29361754

  15. Advanced Material Strategies for Next-Generation Additive Manufacturing.

    Science.gov (United States)

    Chang, Jinke; He, Jiankang; Mao, Mao; Zhou, Wenxing; Lei, Qi; Li, Xiao; Li, Dichen; Chua, Chee-Kai; Zhao, Xin

    2018-01-22

    Additive manufacturing (AM) has drawn tremendous attention in various fields. In recent years, great efforts have been made to develop novel additive manufacturing processes such as micro-/nano-scale 3D printing, bioprinting, and 4D printing for the fabrication of complex 3D structures with high resolution, living components, and multimaterials. The development of advanced functional materials is important for the implementation of these novel additive manufacturing processes. Here, a state-of-the-art review on advanced material strategies for novel additive manufacturing processes is provided, mainly including conductive materials, biomaterials, and smart materials. The advantages, limitations, and future perspectives of these materials for additive manufacturing are discussed. It is believed that the innovations of material strategies in parallel with the evolution of additive manufacturing processes will provide numerous possibilities for the fabrication of complex smart constructs with multiple functions, which will significantly widen the application fields of next-generation additive manufacturing.

  16. Advanced Material Strategies for Next-Generation Additive Manufacturing

    Directory of Open Access Journals (Sweden)

    Jinke Chang

    2018-01-01

    Full Text Available Additive manufacturing (AM has drawn tremendous attention in various fields. In recent years, great efforts have been made to develop novel additive manufacturing processes such as micro-/nano-scale 3D printing, bioprinting, and 4D printing for the fabrication of complex 3D structures with high resolution, living components, and multimaterials. The development of advanced functional materials is important for the implementation of these novel additive manufacturing processes. Here, a state-of-the-art review on advanced material strategies for novel additive manufacturing processes is provided, mainly including conductive materials, biomaterials, and smart materials. The advantages, limitations, and future perspectives of these materials for additive manufacturing are discussed. It is believed that the innovations of material strategies in parallel with the evolution of additive manufacturing processes will provide numerous possibilities for the fabrication of complex smart constructs with multiple functions, which will significantly widen the application fields of next-generation additive manufacturing.

  17. Tube failures due to cooling process problem and foreign materials in power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, J. [Kapar Energy Ventures Sdn Bhd, Jalan Tok Muda, Kapar 42200 (Malaysia); Purbolaksono, J., E-mail: judha@uniten.edu.m [Department of Mechanical Engineering, Universiti Tenaga Nasional, Km 7 Jalan Kajang-Puchong, Kajang 43009, Selangor (Malaysia); Beng, L.C. [Kapar Energy Ventures Sdn Bhd, Jalan Tok Muda, Kapar 42200 (Malaysia)

    2010-07-15

    Cooling process which uses water for heat transfer is an essential factor in coal-fired and nuclear plants. Loss of cooling upset can force the plants to shut down. In particular, this paper reports visual inspections and metallurgical examinations on the failed SA210-A1 right-hand side (RHS) water wall tube of a coal-fired plant. The water wall tube showed the abnormal outer surface colour and has failed with wide-open ductile rupture and thin edges indicating typical signs of short-term overheating. Metallurgical examinations confirmed the failed tube experiencing higher temperature operation. Water flow starvation due to restriction inside the upstream tube is identified as the main root cause of failure. The findings are important to take failure mitigation actions in the future operation. Discussion on the typical problems related to the cooling process in nuclear power plants is also presented.

  18. Tube failures due to cooling process problem and foreign materials in power plants

    International Nuclear Information System (INIS)

    Ahmad, J.; Purbolaksono, J.; Beng, L.C.

    2010-01-01

    Cooling process which uses water for heat transfer is an essential factor in coal-fired and nuclear plants. Loss of cooling upset can force the plants to shut down. In particular, this paper reports visual inspections and metallurgical examinations on the failed SA210-A1 right-hand side (RHS) water wall tube of a coal-fired plant. The water wall tube showed the abnormal outer surface colour and has failed with wide-open ductile rupture and thin edges indicating typical signs of short-term overheating. Metallurgical examinations confirmed the failed tube experiencing higher temperature operation. Water flow starvation due to restriction inside the upstream tube is identified as the main root cause of failure. The findings are important to take failure mitigation actions in the future operation. Discussion on the typical problems related to the cooling process in nuclear power plants is also presented.

  19. ''Risk safety of high frequency fatigue rupture for the vapor generators tubes''; ''Prevention du risque de rupture par fatigue vibratoire des tubes de generateurs de vapeur''

    Energy Technology Data Exchange (ETDEWEB)

    Solgadi, E.; Le Duff, J.A. [FRAMATOME, 92 - Paris-La-Defense (France); Bussy, B. [Electricite de France, 75 - Paris (France). Service Etudes et Projets Thermiques et Nucleaires

    2001-07-01

    Among the different rupture ways identified since 1975 for the steam generators tubes, the fatigue damage occurred on four cases. Two of them are analyzed in this paper: the NORTH ANNA 1 and the MIHAMA 2. From these analysis, it appears that the fatigue crack happens with aggravating factors as the tube embedding, the anti-vibration bars or fretting corrosion. As a preventive, the number of anti-vibration bars has been increase for the vapor generators 1300 and a new system of damper has been developed and implemented on the vapor generator 900. (A.L.B.)

  20. Ultrasonic inspection of steam generator tubes in Superphenix F.B.R. Power plant

    International Nuclear Information System (INIS)

    Gondard, C.

    1991-01-01

    An ultrasonic method has been developed to test of the S.G's tubes of SPX fast breeder reactor. A new type of rotating probes for cracks and wall thickness measurements have been built up and successfully tested. The data acquisition and processing system SPARTACUS was used; it allows high frequency digitalization and powerful signal processings using frequency representations. The actual performances were tested on natural defects under representative operating conditions

  1. Hand-eye coordination of a robot for the automatic inspection of steam-generator tubes in nuclear power plants

    International Nuclear Information System (INIS)

    Choi, D.H.; Song, Y.C.; Kim, J.H.; Kim, J.G.

    2004-01-01

    The inspection of steam-generator tubes in nuclear power plants needs to collect test signals in a highly radiated region that is not accessible by humans. In general, a robot equipped with a camera and a test probe is used to handle such a dangerous environment. The robot moves the probe to right below a tube to be inspected and then the probe is inserted into the tube. The inspection signals are acquired while the probe is pulling back. Currently, an operator in a control room controls all the process remotely. To make a fully automatic inspection system, first of all, a control mechanism is needed to position the probe to the proper location. This is so called a hand-eye coordination problem. In this paper, a hand-eye coordination method for a robot has been presented. The proposed method consists of the two consecutive control modes: rough positioning and fine-tuning. The rough positioning controller tries to position its probe near a target place using kinematics information and the known environments, and then the fine-tuning controller tries to adjust the probe to the target using the image acquired by the camera attached to the robot. The usefulness of the proposed method has been tested and verified through experiments. (orig.)

  2. Steady Secondary Flows Generated by Periodic Compression and Expansion of an Ideal Gas in a Pulse Tube

    Science.gov (United States)

    Lee, Jeffrey M.

    1999-01-01

    This study establishes a consistent set of differential equations for use in describing the steady secondary flows generated by periodic compression and expansion of an ideal gas in pulse tubes. Also considered is heat transfer between the gas and the tube wall of finite thickness. A small-amplitude series expansion solution in the inverse Strouhal number is proposed for the two-dimensional axisymmetric mass, momentum and energy equations. The anelastic approach applies when shock and acoustic energies are small compared with the energy needed to compress and expand the gas. An analytic solution to the ordered series is obtained in the strong temperature limit where the zeroth-order temperature is constant. The solution shows steady velocities increase linearly for small Valensi number and can be of order I for large Valensi number. A conversion of steady work flow to heat flow occurs whenever temperature, velocity or phase angle gradients are present. Steady enthalpy flow is reduced by heat transfer and is scaled by the Prandtl times Valensi numbers. Particle velocities from a smoke-wire experiment were compared with predictions for the basic and orifice pulse tube configurations. The theory accurately predicted the observed steady streaming.

  3. Statistical analysis of entropy generation in longitudinally finned tube heat exchanger with shell side nanofluid by a single phase approach

    Directory of Open Access Journals (Sweden)

    Konchada Pavan Kumar

    2016-06-01

    Full Text Available The presence of nanoparticles in heat exchangers ascertained increment in heat transfer. The present work focuses on heat transfer in a longitudinal finned tube heat exchanger. Experimentation is done on longitudinal finned tube heat exchanger with pure water as working fluid and the outcome is compared numerically using computational fluid dynamics (CFD package based on finite volume method for different flow rates. Further 0.8% volume fraction of aluminum oxide (Al2O3 nanofluid is considered on shell side. The simulated nanofluid analysis has been carried out using single phase approach in CFD by updating the user-defined functions and expressions with thermophysical properties of the selected nanofluid. These results are thereafter compared against the results obtained for pure water as shell side fluid. Entropy generated due to heat transfer and fluid flow is calculated for the nanofluid. Analysis of entropy generation is carried out using the Taguchi technique. Analysis of variance (ANOVA results show that the inlet temperature on shell side has more pronounced effect on entropy generation.

  4. 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)

  5. Radiolytic gas generation in plutonium contaminated waste materials

    International Nuclear Information System (INIS)

    Kazanjian, A.R.

    1976-01-01

    Many plutonium contaminated waste materials decompose into gaseous products because of exposure to alpha radiation. The gases generated (usually hydrogen) over long-storage periods may create hazardous conditions. To determine the extent of such hazards, knowing the gas generation yields is necessary. These yields were measured by contacting some common Rocky Flats Plant waste materials with plutonium and monitoring the enclosed atmospheres for extensive periods of time. The materials were Plexiglas, polyvinyl chloride, glove-box gloves, machining oil, carbon tetrachloride, chlorothene VG solvent, Kimwipes (dry and wet), polyethylene, Dowex-1 resin, and surgeon's gloves. Both 239 Pu oxide and 238 Pu oxide were used as radiation sources. The gas analyses were made by mass spectrometry and the results obtained were the total gas generation, the hydrogen generation, the oxygen consumption rate, and the gas composition over the entire storage period. Hydrogen was the major gas produced in most of the materials. The total gas yields varied from 0.71 to 16 cm 3 (standard temperature pressure) per day per curie of plutonium. The oxygen consumption rates varied from 0.0088 to 0.070 millimoles per day per gram of plutonium oxide-239 and from 0.0014 to 0.0051 millimoles per day per milligram 238 Pu

  6. 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.

  7. Bismuth Telluride and Its Alloys as Materials for Thermoelectric Generation

    Directory of Open Access Journals (Sweden)

    H. Julian Goldsmid

    2014-03-01

    Full Text Available Bismuth telluride and its alloys are widely used as materials for thermoelectric refrigeration. They are also the best materials for use in thermoelectric generators when the temperature of the heat source is moderate. The dimensionless figure of merit, ZT, usually rises with temperature, as long as there is only one type of charge carrier. Eventually, though, minority carrier conduction becomes significant and ZT decreases above a certain temperature. There is also the possibility of chemical decomposition due to the vaporization of tellurium. Here we discuss the likely temperature dependence of the thermoelectric parameters and the means by which the composition may be optimized for applications above room temperature. The results of these theoretical predictions are compared with the observed properties of bismuth telluride-based thermoelements at elevated temperatures. Compositional changes are suggested for materials that are destined for generator modules.

  8. Material for electrodes of low temperature plasma generators

    Science.gov (United States)

    Caplan, Malcolm; Vinogradov, Sergel Evge'evich; Ribin, Valeri Vasil'evich; Shekalov, Valentin Ivanovich; Rutberg, Philip Grigor'evich; Safronov, Alexi Anatol'evich

    2008-12-09

    Material for electrodes of low temperature plasma generators. The material contains a porous metal matrix impregnated with a material emitting electrons. The material uses a mixture of copper and iron powders as a porous metal matrix and a Group IIIB metal component such as Y.sub.2O.sub.3 is used as a material emitting electrons at, for example, the proportion of the components, mass %: iron: 3-30; Y.sub.2O.sub.3:0.05-1; copper: the remainder. Copper provides a high level of heat conduction and electric conductance, iron decreases intensity of copper evaporation in the process of plasma creation providing increased strength and lifetime, Y.sub.2O.sub.3 provides decreasing of electronic work function and stability of arc burning. The material can be used for producing the electrodes of low temperature AC plasma generators used for destruction of liquid organic wastes, medical wastes, and municipal wastes as well as for decontamination of low level radioactive waste, the destruction of chemical weapons, warfare toxic agents, etc.

  9. On the possibility for laboratory simulation of generation of Alfven disturbances in magnetic tubes in the solar atmosphere

    Science.gov (United States)

    Prokopov, Pavel; Zaharov, Yuriy; Tishchenko, Vladimir; Boyarintsev, Eduard; Melehov, Aleksandr; Ponomarenko, Arnold; Posuh, Vitaliy; Shayhislamov, Ildar

    2016-03-01

    The paper deals with generation of Alfven plasma disturbances in magnetic flux tubes through exploding laser plasma in magnetized background plasma. Processes with similar effect of excitation of torsion-type waves seem to provide energy transfer from the solar photosphere to corona. The studies were carried out at experimental stand KI-1 represented a high-vacuum chamber of 1.2 m diameter, 5 m long, external magnetic field up to 500 Gs along the chamber axis, and up to 2×10^-6 Torr pressure in operating mode. Laser plasma was produced when focusing the CO2 laser pulse on a flat polyethylene target, and then the laser plasma propagated in θ-pinch background hydrogen (or helium) plasma. As a result, the magnetic flux tube of 15-20 cm radius was experimentally simulated along the chamber axis and the external magnetic field direction. Also, the plasma density distribution in the tube was measured. Alfven wave propagation along the magnetic field was registered from disturbance of the magnetic field transverse component B_ψ and field-aligned current J_z. The disturbances propagate at near-Alfven velocity of 70-90 km/s and they are of left-hand circular polarization of the transverse component of magnetic field. Presumably, Alfven wave is generated by the magnetic laminar mechanism of collisionless interaction between laser plasma cloud and background. The right-hand polarized high-frequency whistler predictor was registered which have been propagating before Alfven wave at 300 km/s velocity. The polarization direction changed with Alfven wave coming. Features of a slow magnetosonic wave as a sudden change in background plasma concentration along with simultaneous displacement of the external magnetic field were found. The disturbance propagates at ~20-30 km/s velocity, which is close to that of ion sound at low plasma beta value. From preliminary estimates, the disturbance transfers about 10 % of the original energy of laser plasma.

  10. Multi-region fuzzy logic controller with local PID controllers for U-tube steam generator in nuclear power plant

    Directory of Open Access Journals (Sweden)

    Puchalski Bartosz

    2015-12-01

    Full Text Available In the paper, analysis of multi-region fuzzy logic controller with local PID controllers for steam generator of pressurized water reactor (PWR working in wide range of thermal power changes is presented. The U-tube steam generator has a nonlinear dynamics depending on thermal power transferred from coolant of the primary loop of the PWR plant. Control of water level in the steam generator conducted by a traditional PID controller which is designed for nominal power level of the nuclear reactor operates insufficiently well in wide range of operational conditions, especially at the low thermal power level. Thus the steam generator is often controlled manually by operators. Incorrect water level in the steam generator may lead to accidental shutdown of the nuclear reactor and consequently financial losses. In the paper a comparison of proposed multi region fuzzy logic controller and traditional PID controllers designed only for nominal condition is presented. The gains of the local PID controllers have been derived by solving appropriate optimization tasks with the cost function in a form of integrated squared error (ISE criterion. In both cases, a model of steam generator which is readily available in literature was used for control algorithms synthesis purposes. The proposed multi-region fuzzy logic controller and traditional PID controller were subjected to broad-based simulation tests in rapid prototyping software - Matlab/Simulink. These tests proved the advantage of multi-region fuzzy logic controller with local PID controllers over its traditional counterpart.

  11. Nordic Nuclear Materials Forum for Generation IV Reactors

    International Nuclear Information System (INIS)

    Anghel, C.; Penttilae, S.

    2010-03-01

    A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

  12. Nordic Nuclear Materials Forum for Generation IV Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anghel, C. (Studsvik Nuclear AB, Nykoeping (Sweden)); Penttilae, S. (Technical Research Centre of Finland, VTT (Finland))

    2010-03-15

    A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

  13. Model-based ECT signal interpretation and experimental verification for the quantitative flaw characterization in steam generator tubes

    International Nuclear Information System (INIS)

    Song, Sung Jin; Kim, Young Hwan; Kim, Eui Lae; Chung, Tae Eon; Yim, Chang Jae

    2002-01-01

    The model-based inversion tools for eddy current signals have been developed by the novel combination of neural networks and finite element modeling for quantitative flaw characterization in steam generator tubes. In the present work, interpretation of experimental eddy current signals was carried out in order to validate the developed inversion tools. A database was constructed using the synthetic flaw signals generated by the finite element modeling. The hybrid neural networks of a PNN classifier and BPNN size estimators were trained using the synthetic signals. Experimental eddy current signals were obtained from axisymmetric artificial flaws. Interpretations of flaws were carried out by feeding experimental signals into the neural networks. The results of interpretations were excellent, so that the developed inversion tools would be applicable to the interpretation of experimental eddy current signals.

  14. Contribution to perfecting eddy current testing of steam generator tubes of sodium cooled breeders: description of the Monacault loop for the study of sodium deposit influence

    International Nuclear Information System (INIS)

    Lapicore, A.; Lemarquis, J.C.; Oberlin, C.; Pigeon, M.

    1981-12-01

    In the event of sodium-water reaction in the steam generator of a sodium cooled breeder reactor, it is essential to be able to monitor the local loss of thickness of the tubes located in the reaction area. A method for monitoring the tubes by an eddy current probe is being developed for Super Phenix. The sodium deposits on the outer wall of the tubes, as well as their prolonged contact with high temperature sodium are likely to bring about a change in the signals picked up. A test loop, Monacault, has been built in order to clarify the importance of these parameters (effect of sodium deposits, reproducibility of the wetting at different temperatures). It includes three test cells containing the sample tubes having a total of 61 standard defects to be tested. The first results on the wetting of tubes are given and discussed [fr

  15. Numerical simulation and experimental results of horizontal tube falling film generator working in a NH3-LiNO3 absorption refrigeration system

    International Nuclear Information System (INIS)

    Herrera, J.V.; Garcia-Valladares, O.; Gomez, V.H.; Best, R.

    2010-01-01

    This paper describes the work made at the Centro de Investigacion en Energia in the development of an absorption refrigeration system for cooling and refrigeration applications with a capacity of 10 kW. The single effect unit utilizes ammonia-lithium nitrate as working pair and it is air cooled. The generator is a falling film type with horizontal tubes where the heating oil flows inside the tube bank and the ammonia-lithium nitrate solution flows as a falling film on the tube outside, where it is heated and ammonia vapor is generated. The generator consists of tree columns and four rows per column of horizontal tubes. The system was tested at controlled conditions with heating oil obtained from an electric resistance heating loop. A numerical model of the horizontal falling film generator was developed that divided the system into three different thermal elements: the flow inside the tube, the heat conduction in the tube wall and the falling film solution flow. The mathematical model was tested and validated with experimental data and a study of the influence of the heat transfer coefficient for ammonia-lithium nitrate solution in the numerical model was carried out. A comparison between experimental and numerical data for the heat flux in the system and the temperature profiles in the oil and solution flows shown a good degree of correlation.

  16. Fast reactor steam generators with sodium on the tube side. Design and operational parameters

    International Nuclear Information System (INIS)

    1994-01-01

    A comparison of design and operational characteristics as well as analysis of experience gained during the long terms operation of the Micro Module Inverse Steam Generator and Module Inverse Steam Generator at BOR 60 reactor are main aims of this technical report. 20 refs, 47 figs, 14 tabs

  17. Neutron tubes

    Science.gov (United States)

    Leung, Ka-Ngo [Hercules, CA; Lou, Tak Pui [Berkeley, CA; Reijonen, Jani [Oakland, CA

    2008-03-11

    A neutron tube or generator is based on a RF driven plasma ion source having a quartz or other chamber surrounded by an external RF antenna. A deuterium or mixed deuterium/tritium (or even just a tritium) plasma is generated in the chamber and D or D/T (or T) ions are extracted from the plasma. A neutron generating target is positioned so that the ion beam is incident thereon and loads the target. Incident ions cause D-D or D-T (or T-T) reactions which generate neutrons. Various embodiments differ primarily in size of the chamber and position and shape of the neutron generating target. Some neutron generators are small enough for implantation in the body. The target may be at the end of a catheter-like drift tube. The target may have a tapered or conical surface to increase target surface area.

  18. Generation method of educational materials using qualitative reasoning

    International Nuclear Information System (INIS)

    Yoshimura, Seiichi; Yamada, Shigeo; Fujisawa, Noriyoshi.

    1992-01-01

    Central Research Institute of Electric Power Industry has developed a nuclear power plant educational system in which educational materials for several events are included. The system effectively teaches operators by tailoring the event explanations to their knowledge levels of understanding. The preparation of the educational materials, however, is laborious and this becomes one of the problems in the practical use of the system. Discussed in the present paper is a basic explanation generation method using qualitative reasoning. This has been developed to solve the problem. Qualitative equations describing a recirculation pumps trip were transformed into production rules. These were stored in the knowledge base of an event explanation generation system together with explanation sentences. When an operator selects a certain variable's time-interval in which he wants to know the reasons for a variable change, the inference engine searches for the rule which satisfies both the qualitative value and qualitative differential value concerned with this time-interval. Then the event explanation generation section provides explanations by combining the explanation sentences attached to the rules. This paper demonstrates that it is possible to apply qualitative reasoning to such complex reactor systems, and also that explanations can be generated using the simulation results from a transient analysis code. (author)

  19. Reduction of residual stresses in internal skin of transient zones of PWR steam generator expanded tubes: tests with a ''rotating brush''

    International Nuclear Information System (INIS)

    Vidal, P.

    1984-04-01

    A process aiming at preventing or suppressing cracks under stress corrosion on the primary side in the expanded zones of PWR steam generator tubes has been studied; it consists in hammering the internal skin of tubes in these zones what reduces the level of residual expanding stresses to lower values around 100-150 MPa without modifying the stress level in external skin. Tests in magnesium chloride to estimate the residual stresses of tubes in low carbon stainless austenitic steel 18% Cr-12% Ni with molybdene [fr

  20. Analysis of flow-induced vibration of heat exchanger and steam generator tube bundles using the AECL computer code PIPEAU-2

    International Nuclear Information System (INIS)

    Gorman, D.J.

    1983-12-01

    PIPEAU-2 is a computer code developed at the Chalk River Nuclear Laboratories for the flow-induced vibration analysis of heat exchanger and steam generator tube bundles. It can perform this analysis for straight and 'U' tubes. All the theoretical work underlying the code is analytical rather than numerical in nature. Highly accurate evaluation of the free vibration frequencies and mode shapes is therefore obtained. Using the latest experimentally determined parameters available, the free vibration analysis is followed by a forced vibration analysis. Tube response due to fluid turbulence and vortex shedding is determined, as well as critical fluid velocity associated with fluid-elastic instability

  1. An efficient miniature 120 Hz pulse tube cryocooler using high porosity regenerator material

    Science.gov (United States)

    Yu, Huiqin; Wu, Yinong; Ding, Lei; Jiang, Zhenhua; Liu, Shaoshuai

    2017-12-01

    A 1.22 kg coaxial miniature pulse tube cryocooler (MPTC) has been fabricated and tested in our laboratory to provide cooling for cryogenic applications demanding compactness, low mass and rapid cooling rate. The geometrical parameters of regenerator, pulse tube and phase shifter are optimized. The investigation demonstrates that using higher mesh number and thinner wire diameter of stainless steel screen (SSS) can promote the coefficient of performance (COP) when the MPTC operates at 120 Hz. In this study, the 604 mesh SSS with 17 μm diameter of mesh wire is constructed as filler of regenerator. The experimental results show the MPTC operating at 120 Hz achieves a no-load temperature of 53.5 K with 3.8 MPa charging pressure, and gets a cooling power of 2 W at 80 K with 55 W input electric power which has a relative Carnot efficiency of 9.68%.

  2. Extraction: a system for automatic eddy current diagnosis of steam generator tubes in nuclear power plants; Extracsion: un systeme de controle automatique par courants de Foucault des tubes de generateurs de vapeur de centrales nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Georgel, B.; Zorgati, R.

    1994-12-31

    Improving speed and quality of Eddy Current non-destructive testing of steam generator tubes leads to automatize all processes that contribute to diagnosis. This paper describes how we use signal processing, pattern recognition and artificial intelligence to build a software package that is able to automatically provide an efficient diagnosis. (authors). 2 figs., 5 refs.

  3. Theoretical Analysis of Effects of Wall Suction on Entropy Generation Rate in Laminar Condensate Layer on Horizontal Tube

    Directory of Open Access Journals (Sweden)

    Tong-Bou Chang

    2014-01-01

    Full Text Available The effects of wall suction on the entropy generation rate in a two-dimensional steady film condensation flow on a horizontal tube are investigated theoretically. In analyzing the liquid flow, the effects of both the gravitational force and the viscous force are taken into account. In addition, a film thickness reduction ratio, Sf, is introduced to evaluate the effect of wall suction on the thickness of the condensate layer. The analytical results show that, the entropy generation rate depends on the Jakob number Ja, the Rayleigh number Ra, the Brinkman number Br, the dimensionless temperature difference ψ, and the wall suction parameter Sw. In addition, it is shown that in the absence of wall suction, a closed-form correlation for the Nusselt number can be derived. Finally, it is shown that the dimensionless entropy generation due to heat transfer, NT, increases with an increasing suction parameter Sw, whereas the dimensionless entropy generation due to liquid film flow friction, NF, decreases.

  4. Steam generator assembly for pressurized water reactors with a straight tube bundle and a partial flow preheater traversible by pressurized water

    International Nuclear Information System (INIS)

    Michel, R.

    1976-01-01

    To reduce the temperature difference between a straight tube bundle and the housing surrounding the same in a steam generator assembly for pressurized water reactors, a preheater for feed water is provided, and part of the pressurized water, after it has flowed through the heat exchanger or steam generator proper, is used for heating the feedwater in the preheater. 3 claims, 1 drawing figure

  5. [A presumption calculating formula of the X-ray spectrum generated from a molybdenum target X-ray tube].

    Science.gov (United States)

    Kato, Hideki; Fujii, Shigehisa; Shirakawa, Seiji; Suzuki, Yusuke; Nishii, Yoshio

    2011-01-01

    A presumption calculating formula of the X-ray spectrum generated from a molybdenum target X-ray tube is presented. The calculation procedure is to add an amount of characteristic X-ray photons that corresponds to the ratio of characteristic photons and bremsstrahlung photons to the bremsstrahlung spectrum obtained using semiempirical calculation. The bremsstrahlung spectrum was calculated by using a corrected Tucker's formula. The corrected content was a formula for calculating the self-absorption length in the target that originated in the difference of the incident angle to the target of the electron and the mass stopping power data. The measured spectrum was separated into the bremsstrahlung component and the characteristic photon component, and the ratio of the characteristic photons and bremsstrahlung photons was obtained. The regression was derived from the function of the tube voltage. Based on this calculation procedure, computer software was constructed that can calculate an X-ray spectrum in arbitrary exposure conditions. The X-ray spectrum obtained from this presumption calculating formula and the measured X-ray spectrum corresponded well. This formula is very useful for analyzing various problems related to mammography by means of Monte Carlo simulations.

  6. Characterization of the Crystallographic Textures and Mechanical Anisotropy Factors in Two Modifications of Zr-2.5Nb Pressure Tube Materials

    Science.gov (United States)

    Fong, Randy; Vogel, Sven; Miller, Ron; Saari, Henry

    Zr-2.5Nb alloy is used for the pressure tubes in CANDU reactors. Current as-manufactured tubes are produced in a cold-worked and stress-relieved metallurgical condition. The tubes installed in reactors normally operate at 300°C. In a hypothetical loss-of-coolant-accident (LOCA), the pressure tube may be overheated to 1000°C. During the temperature transient, a phase transformation occurs that changes the microstructure and affects the material's high-temperature deformation behaviour. In this study, improvements to enhance the performance of pressure-tube materials are being explored by modifying the texture and microstructure of as-manufactured pressure tubes. Two modifications were carried out by high-temperature annealing, with or without subsequent cold-working. This paper presents the resulting modified textures as measured by neutron diffraction and their texture evolution during heating. The anisotropy factors calculated for the modified Zr-2.5Nb pressure-tube materials using the measured texture data are compared with those previously characterized for Zircaloy-4 fuel cladding. The resulting effect of these texture and microstructure modifications with regard to the material's response to anisotropic deformation during heating to high temperatures is also discussed in this paper.

  7. The generation of calandria tube (CT) inner diameter profiles from fuel channel (FC) inspection data

    Energy Technology Data Exchange (ETDEWEB)

    Sedran, P.J., E-mail: paul.sedran@amec.com [AMEC NSS, Toronto, ON (Canada); Rankin, B., E-mail: brankin@nbpower.com [NB Power, Fredericton, NB (Canada); Lemire, C., E-mail: Lemire.Christian@hydro.qc.ca [Hydro-Quebec, Montreal, QC (Canada)

    2015-07-01

    Studies of CT deformation at spacer locations, key to the development of FC deformation modelling, have been limited by the availability of gauging measurements from removed CTs. In [1], it was proposed that CT dimensional profiles could be generated using FC inspection data. Since then, the concept was investigated further by assessing: (1) the normalisation of gap measurements to the diameter of the spacer coil, (2) the validity of gap measurements from inspections of Point Lepreau and Gentilly-2, and the CT dimensional profiles generated from the inspection data.It was concluded, from the work presented in this paper, that the CT-PT gap data and the CT dimensional profiles generated using the data from the two subject inspections are reasonable. (author)

  8. Analogy for the effect of material and geometrical variables on energy-absorption capability of composite tubes

    Science.gov (United States)

    Farley, Gary L.; Jones, Robert M.

    1992-01-01

    Simplified procedures for determining the qualitative effect a variable has on structural response of a composite tube are very useful in both preliminary design as well as in providing insight into the general response. An analysis procedure is presented that can be used to determine the qualitative change in the sustained crushing load due to a change in specimen material properties or geometry. The analysis procedure is similar in form to the equation for the buckling load of a column on an elastic foundation.

  9. 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)

  10. Heat transfer and pressure drop characteristics of the tube bank fin heat exchanger with fin punched with flow redistributors and curved triangular vortex generators

    Science.gov (United States)

    Liu, Song; Jin, Hua; Song, KeWei; Wang, LiangChen; Wu, Xiang; Wang, LiangBi

    2017-10-01

    The heat transfer performance of the tube bank fin heat exchanger is limited by the air-side thermal resistance. Thus, enhancing the air-side heat transfer is an effective method to improve the performance of the heat exchanger. A new fin pattern with flow redistributors and curved triangular vortex generators is experimentally studied in this paper. The effects of the flow redistributors located in front of the tube stagnation point and the curved vortex generators located around the tube on the characteristics of heat transfer and pressure drop are discussed in detail. A performance comparison is also carried out between the fins with and without flow redistributors. The experimental results show that the flow redistributors stamped out from the fin in front of the tube stagnation points can decrease the friction factor at the cost of decreasing the heat transfer performance. Whether the combination of the flow redistributors and the curved vortex generators will present a better heat transfer performance depends on the size of the curved vortex generators. As for the studied two sizes of vortex generators, the heat transfer performance is promoted by the flow redistributors for the fin with larger size of vortex generators and the performance is suppressed by the flow redistributors for the fin with smaller vortex generators.

  11. Development boiling to sprinkled tube bundle

    Science.gov (United States)

    Kracík, Petr; Pospíšil, Jiří

    2016-03-01

    This paper presents results of a studied heat transfer coefficient at the surface of a sprinkled tube bundle where boiling occurs. Research in the area of sprinkled exchangers can be divided into two major parts. The first part is research on heat transfer and determination of the heat transfer coefficient at sprinkled tube bundles for various liquids, whether boiling or not. The second part is testing of sprinkle modes for various tube diameters, tube pitches and tube materials and determination of individual modes' interface. All results published so far for water as the falling film liquid apply to one to three tubes for which the mentioned relations studied are determined in rigid laboratory conditions defined strictly in advance. The sprinkled tubes were not viewed from the operational perspective where there are more tubes and various modes may occur in different parts with various heat transfer values. The article focuses on these processes. The tube is located in a low-pressure chamber where vacuum is generated using an exhauster via ejector. The tube consists of smooth copper tubes of 12 mm diameter placed horizontally one above another.

  12. Development boiling to sprinkled tube bundle

    Directory of Open Access Journals (Sweden)

    Kracík Petr

    2016-01-01

    Full Text Available This paper presents results of a studied heat transfer coefficient at the surface of a sprinkled tube bundle where boiling occurs. Research in the area of sprinkled exchangers can be divided into two major parts. The first part is research on heat transfer and determination of the heat transfer coefficient at sprinkled tube bundles for various liquids, whether boiling or not. The second part is testing of sprinkle modes for various tube diameters, tube pitches and tube materials and determination of individual modes’ interface. All results published so far for water as the falling film liquid apply to one to three tubes for which the mentioned relations studied are determined in rigid laboratory conditions defined strictly in advance. The sprinkled tubes were not viewed from the operational perspective where there are more tubes and various modes may occur in different parts with various heat transfer values. The article focuses on these processes. The tube is located in a low-pressure chamber where vacuum is generated using an exhauster via ejector. The tube consists of smooth copper tubes of 12 mm diameter placed horizontally one above another.

  13. Generation of nano roughness on fibrous materials by atmospheric plasma

    International Nuclear Information System (INIS)

    Kulyk, I; Scapinello, M; Stefan, M

    2012-01-01

    Atmospheric plasma technology finds novel applications in textile industry. It eliminates the usage of water and of hazard liquid chemicals, making production much more eco-friendly and economically convenient. Due to chemical effects of atmospheric plasma, it permits to optimize dyeing and laminating affinity of fabrics, as well as anti-microbial treatments. Other important applications such as increase of mechanical resistance of fiber sleeves and of yarns, anti-pilling properties of fabrics and anti-shrinking property of wool fabrics were studied in this work. These results could be attributed to the generation of nano roughness on fibers surface by atmospheric plasma. Nano roughness generation is extensively studied at different conditions. Alternative explanations for the important practical results on textile materials and discussed.

  14. Accretor: Generative Materiality in the Work of Driessens and Verstappen.

    Science.gov (United States)

    Whitelaw, Mitchell

    2015-01-01

    Accretor, by the Dutch artists Erwin Driessens and Maria Verstappen, is a generative artwork that adopts and adapts artificial life techniques to produce intricate three-dimensional forms. This article introduces and analyzes Accretor, considering the enigmatic quality of the generated objects and in particular the role of materiality in this highly computational work. Accretor demonstrates a tangled continuity between digital and physical domains, where the constraints and affordances of matter inform both formal processes and aesthetic interpretations. Drawing on Arp's notion of the concrete artwork and McCormack and Dorin's notion of the computational sublime, the article finally argues that Accretor demonstrates what might be called a processual sublime, evoking expansive processes that span both computational and non-computational systems.

  15. Generation of nano roughness on fibrous materials by atmospheric plasma

    Science.gov (United States)

    Kulyk, I.; Scapinello, M.; Stefan, M.

    2012-12-01

    Atmospheric plasma technology finds novel applications in textile industry. It eliminates the usage of water and of hazard liquid chemicals, making production much more eco-friendly and economically convenient. Due to chemical effects of atmospheric plasma, it permits to optimize dyeing and laminating affinity of fabrics, as well as anti-microbial treatments. Other important applications such as increase of mechanical resistance of fiber sleeves and of yarns, anti-pilling properties of fabrics and anti-shrinking property of wool fabrics were studied in this work. These results could be attributed to the generation of nano roughness on fibers surface by atmospheric plasma. Nano roughness generation is extensively studied at different conditions. Alternative explanations for the important practical results on textile materials and discussed.

  16. Next Generation Nuclear Plant Materials Research and Development Program Plan

    International Nuclear Information System (INIS)

    G.O. Hayner; R.L. Bratton; R.N. Wright

    2005-01-01

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. The NGNP Materials Research and Development (R and D) Program is responsible for performing R and D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. The NGNP Materials R and D Program includes the following elements: (1) Developing a specific approach, program plan and other project management

  17. Next Generation Nuclear Plant Materials Research and Development Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    G.O. Hayner; R.L. Bratton; R.N. Wright

    2005-09-01

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. The NGNP Materials R&D Program includes the following elements: (1) Developing a specific approach, program plan and other project management tools for

  18. Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

    1982-01-01

    Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

  19. Carbon-nanostructured materials for energy generation and storage applications

    Directory of Open Access Journals (Sweden)

    V. Linkov

    2010-01-01

    Full Text Available We have developed and refined a chemical vapour deposition method to synthesise nanotubes using liquid petroleum gasasthe carbonsource. The nanotubes were thoroughly characterised by scanning electron microscopy, transmission electron microscopy
    X-ray diffraction and thermogravimetric analysis. The protocol to grow nanotubes was then adapted to deposit nanotubes on the surface of different substrates, which were chosen based upon how
    the substrates could be applied in various hydrogen energyconver-sion systems. Carbon nanotubes area nanostructured material with an extremely wide range of application sinvariousenergy applications. The methods outlined demonstrate the complete
    development of carbon nanotube composite materials with direct applications in hydrogen energy generation, storage and conversion.

  20. Electrochemical selenium hydride generation with in situ trapping in graphite tube atomizers

    Czech Academy of Sciences Publication Activity Database

    Šíma, Jan; Rychlovský, P.

    2003-01-01

    Roč. 58, č. 5 (2003), s. 919-930 ISSN 0584-8547 R&D Projects: GA ČR GA203/98/0754; GA ČR GA203/01/0453 Institutional research plan: CEZ:AV0Z4031919 Keywords : hydride generation * electrothermal atomic absorption spectrometry * In situ trapping Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 2.361, year: 2003