WorldWideScience

Sample records for httr longitudinal welding

  1. Ultrasonic test results for the reactor pressure vessel of the HTTR. Longitudinal welding line of bottom dome

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Ohwada, Hiroyuki; Kato, Yasushi

    2008-06-01

    This paper describes the inspection method, the measured area, etc. of the ultrasonic test of the in-service inspection (ISI) for welding lines of the reactor pressure vessel of the HTTR and the inspection results of the longitudinal welding line of the bottom dome. The pre-service inspection (PSI) results for estimation of occurrence and progression of defects to compare the ISI results is described also. (author)

  2. Movement of liquid metal in welding bath during welding in longitudinal magnetic field

    International Nuclear Information System (INIS)

    Kovalev, I.M.; Rybakov, A.S.

    1977-01-01

    The specific features are considered of liquid metal flow in a bath during welding of steel 12Kh18N10T plates with a non-consumable electrode in argon under interaction of the arc and bath with a longitudinal constant magnetic field. In controlling the velocity field of metal flow, the longitudinal magnetic field permits to form a seam at automatic welding of horizontal joints on a vertical plane

  3. Influence of anisotropic hardening on longitudinal welding strains and stresses

    International Nuclear Information System (INIS)

    Gatovskij, K.M.; Revutskij, M.N.

    1981-01-01

    The algorithm and program for estimation of longitudinal welding strains and stresses with account of hardening and Bauschinger effect, which expand the possibilities of more complete description of stress change during thermodeformation welding cycles at bead surfacing on plate made of the 06Kh18N9T steel and AMg61 alloy. It is shown that for metals, deformation curves which are characterized by considerable yield moduli (Esub(T)/E>=0.05) hardening effect is considerable and its account leads to the decrease of stress level in the heataffected zone (down to 20%) [ru

  4. Analysis of the Mechanism of Longitudinal Bending Deformation Due to Welding in a Steel Plate by Using a Numerical Model

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Rae; Yan, Jieshen; Kim, Jae-Woong [Yeungnam Univ., Gyeongsan (Korea, Republic of); Song, Gyu Yeong [Gyeongbuk Hybrid Technology Institute, Yeongcheon (Korea, Republic of)

    2017-01-15

    Welding deformation is a permanent deformation that is caused in structures by welding heat. Welding distortion is the primary cause of reduced productivity, due to welded structural strength degradation, low dimensional accuracy, and appearance. As a result, research and numerous experiments are being carried out to control welding deformation. The aim of this study is to analyze the mechanism of longitudinal bending deformation due to welding. Welding experiments and numerical analyses were performed for this study. The welding experiments were performed on 4 mm and 8.5 mm thickness steel plates, and the numerical analysis was conducted on the welding deformation using the FE software MSC.marc.

  5. Longitudinal Weld Land Buckling in Compression-Loaded Orthogrid Cylinders

    Science.gov (United States)

    Thornburgh, Robert P.; Hilburger, Mark W.

    2010-01-01

    Large stiffened cylinders used in launch vehicles (LV), such as the Space Shuttle External Tank, are manufactured by welding multiple curved panel sections into complete cylinders. The effects of the axial weld lands between the panel sections on the buckling load were studied, along with the interaction between the acreage stiffener arrangement and the weld land geometry. This document contains the results of the studies.

  6. Deviation of longitudinal and shear waves in austenitic stainless steel weld metal

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Reimann, K.J.

    1980-01-01

    One of the difficulties associated with the ultrasonic inspection of stainless steel weld metal is the deviation of the ultrasonic beams. This can lead to errors in determining both the location and size of reflectors. The present paper compares experimental and theoretical data related to beam steering for longitudinal and shear waves in a sample of 308 SS weld metal. Agreement between predicted and measured beam deviations is generally good. Reasons for discrepancies are discussed

  7. Formation mechanisms of periodic longitudinal microstructure and texture patterns in friction stir welded magnesium AZ80

    Energy Technology Data Exchange (ETDEWEB)

    Hiscocks, J., E-mail: j.hiscocks@queensu.ca [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, Ontario (Canada); Diak, B.J. [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, Ontario (Canada); Gerlich, A.P. [Department of Mechanical and Mechatronics Engineering, Waterloo University, Waterloo, Ontario (Canada); Daymond, M.R. [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, Ontario (Canada)

    2016-12-15

    Many studies of friction stir welding have shown that periodicity of metal flow around the tool pin may result in the formation of periodic differences in microstructure and texture in the weld nugget area correlated with the weld pitch. The current work investigates the periodicity of magnesium weld microtexture in the nugget region and its association with material flow using optical and electron microscopy. Two welds created in AZ80 at different processing conditions are presented in detail, one illustrating periodic longitudinal texture change, and one showing for the first time that periodic variations in texture, grain size, or composition are not defining features of periodic nugget flow. While nugget texture is dominated by shear deformation, it was found here to be affected to a lesser degree by compaction of material behind the welding tool, which led to reduction in intensity of the shear texture fiber. The decreased tendency for magnesium based alloys to form periodic patterns as compared to aluminum based alloys is explained with reference to the shear textures. - Highlights: •It is shown here that periodic material flow in the nugget does not necessitate longitudinal texture patterns. •Longitudinal texture patterns are shown to be present or absent in Mg AZ80 based on processing conditions. •Texture in the nugget is mainly dictated by shear deformation, but has measurable effects from other deformation modes. •Explanation of why longitudinal texture change is frequently reported in aluminum but not magnesium alloys is provided. •A new vector visualization of material flow based on EBSD data analysis is shown.

  8. Predicting welding distortion in a panel structure with longitudinal stiffeners using inherent deformations obtained by inverse analysis method.

    Science.gov (United States)

    Liang, Wei; Murakawa, Hidekazu

    2014-01-01

    Welding-induced deformation not only negatively affects dimension accuracy but also degrades the performance of product. If welding deformation can be accurately predicted beforehand, the predictions will be helpful for finding effective methods to improve manufacturing accuracy. Till now, there are two kinds of finite element method (FEM) which can be used to simulate welding deformation. One is the thermal elastic plastic FEM and the other is elastic FEM based on inherent strain theory. The former only can be used to calculate welding deformation for small or medium scale welded structures due to the limitation of computing speed. On the other hand, the latter is an effective method to estimate the total welding distortion for large and complex welded structures even though it neglects the detailed welding process. When the elastic FEM is used to calculate the welding-induced deformation for a large structure, the inherent deformations in each typical joint should be obtained beforehand. In this paper, a new method based on inverse analysis was proposed to obtain the inherent deformations for weld joints. Through introducing the inherent deformations obtained by the proposed method into the elastic FEM based on inherent strain theory, we predicted the welding deformation of a panel structure with two longitudinal stiffeners. In addition, experiments were carried out to verify the simulation results.

  9. Water chemistry control in HTTR

    International Nuclear Information System (INIS)

    Sekita, Kenji; Furusawa, Takayuki; Emori, Koichi; Kuroha, Misao; Hayakawa, Masato; Ohuchi, Hiroshi; Ishii, Taro

    2008-08-01

    A carbon steel is used for the main material for the components and pipings of the pressurized water cooling system etc. that are the reactor cooling system of the HTTR. Water quality is managed by using the hydrazine in the coolant of the water cooling system to prevent corrosion of the components and deoxidize the coolant. Also, regular analysis is carried out for the confirmation of the water quality. The following results were obtained through the water quality analysis. (1) In the pressurized water cooling system, the coolant temperature rises higher due to the heat removal of the primary coolant. So, the ammonia was formed in the thermal decomposition of the hydrazine. The electric conductivity increased, while the concentration of the hydrazine decreased, there was no problem as the plan it. (2) Thermal decomposition of the hydrazine was not occurred in the auxiliary water cooling system and vessel cooling system because of the coolant temperature was low. (3) An indistinct procedure is clarified and procedure of water quality analysis was established in the HTTR. (4) It is assumed that the corrosion of the components in these water cooling system hardly occurred from measurement results of dissolved oxide and chloride ion. Thus, the water quality was managed enough. (author)

  10. SAFT and TOFD evaluation for ultrasonic testing of longitudinal welds in large-diameter pipes; SAFT- und TOFD-Auswertung fuer die Ultraschall-Schweissnahtpruefung von laengsnahtgeschweissten Grossrohren

    Energy Technology Data Exchange (ETDEWEB)

    Rieder, Hans; Dillhoefer, Alexander; Spies, Martin [Fraunhofer-Institut fuer Techno- und Wirtschaftsmathematik ITWM, Kaiserslautern (Germany); Graff, Alfred; Orth, Thomas [Salzgitter Mannesmann Forschung GmbH, Duisburg (Germany); Kersting, Thomas [Europipe GmbH, Muelheim an der Ruhr (Germany). Werk Muelheim

    2010-07-01

    The authors present a mobile multichannel system for testing of longitudinal welds in large-diameter tubes. The prototype was designed for in situ inspection of longitudinal welds using transversal and longitudinal waves impacting at an angle from both sides in combined SE and IE operation. The reconstruction and imaging software enables SAFT calculations for all surface curves and superposition of the SAFT reconstructions from both sides into a general image of the weld. This includes superposition of the TOFD test data with a suitable evaluation and assessment concept. Details are presented, as are test results on test pieces for validation and on longitudinal welds in large-diameter tubes. (orig.)

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

    Institute of Scientific and Technical Information of China (English)

    Li Bing

    2002-01-01

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

  12. Current status of HTTR project at JAERI

    International Nuclear Information System (INIS)

    Saito, Shinzo; Sudo, Yukio; Tanaka, Toshiyuki; Baba, Osamu

    1992-01-01

    The HTTR is a high temperature gas cooled test reactor with thermal output of 30 MW, outlet coolant temperatures of 850degC at rated operation and 950degC at high temperature test operation and primary coolant pressure of 4 MPa. The HTTR consists of a reactor pressure vessel with a prismatic core in it, a primary cooling loop with an intermediate helium-helium heat exchanger and a pressurized water cooler in parallel, an auxiliary cooling system, a reactor vessel cooling system and related components. The HTTR is utilized for establishing and upgrading the technology bases for advanced HTGRs including irradiation tests for fuels and materials, safety demonstration tests for HTGRs and nuclear heat application and for carrying out various kinds of innovative basic researches on high temperature technologies. Since 1969, the JAERI has carried out research and development works on block type fuel, high temperature materials, high temperature in-core instrumentations, high temperature components, reactor physics, heat transfer and fluid dynamics, plate-out of fission products etc., in order to construct the HTTR which can supply high temperature coolant of 950degC to the outside of the pressure vessel for the nuclear heat application, for the first time in the world. In November 1990, the installation permit was issued by the Government through about 20 month safety review by the Science and Technology Agency and Nuclear Safety Commission. The construction of the HTTR facility was initiated on the site in the Oarai Research Establishment, JAERI in March 1991. The excavation of ground is now finished at the HTTR site. It will take about six years for the completion of the HTTR facility and the first criticality will be attained in 1996. (author)

  13. Post irradiation examinations on HTTR materials

    International Nuclear Information System (INIS)

    Sakai, Haruyuki; Ohmi, Masao; Eto, Motokuni; Watanabe, Katsutoshi

    1995-01-01

    The HTTR (High Temperature engineering Test Reactor) is being constructed at Oarai Research Establishment of the Japan Atomic Energy Research Institute. In order to develop necessary materials for the HTTR, after irradiations in the JMTR, PIEs are being carried out on these materials in the JMTRHL (JMTR Hot Laboratory). Impact test, tensile test, fatigue test, creep test, metallography and so on were performed for irradiated 2 1/4Cr 1Mo steel as the pressure vessel material and Alloy 800H as the cladding material of the control rod. A fatigue testing machine and four creep testing machines newly designed were fabricated and installed in the steel cells in order to evaluate the integrity of the HTTR materials. The development process and PIE results obtained with these machines are given in this paper

  14. Status of the HTTR project in Japan

    International Nuclear Information System (INIS)

    Sanokawa, K.

    1989-01-01

    In June 1987, the Japanese Atomic Energy Commission issued the revision of the Long-Term Program for Development and Utilization of Nuclear Energy, recommending that Japan should proceed with the development of more advanced new technologies for the future, parallel to the existing nuclear systems. It also emphasizes that the HTGR is one of the most promising reactors with high efficiency and inherent safety, therefore it should be explored for the broader use of nuclear energy, not only for power production. Then the early construction of a High-Temperature engineering Test Reactor (HTTR) by JAERI was proposed. Based on this program, JAERI has changed the ''VHTR program'' for ''HTTR program'' to establish the HTGR technology basis and upgrade them. The permission of the HTTR construction was made by the Science and Technology Agency (STA) in February 1989. 3 figs, 2 tabs

  15. 49 CFR 178.61 - Specification 4BW welded steel cylinders with electric-arc welded longitudinal seam.

    Science.gov (United States)

    2010-10-01

    ... Transportation PIPELINE AND HAZARDOUS MATERIALS SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION HAZARDOUS... inspection must conform to the techniques and acceptability criteria set forth in CGA Pamphlet C-3. When... prescribed tests to be acceptable. Repair of welded seams by welding is authorized provided that all...

  16. Welding.

    Science.gov (United States)

    Cowan, Earl; And Others

    The curriculum guide for welding instruction contains 16 units presented in six sections. Each unit is divided into the following areas, each of which is color coded: terminal objectives, specific objectives, suggested activities, and instructional materials; information sheet; transparency masters; assignment sheet; test; and test answers. The…

  17. Track-Bridge Longitudinal Interaction of Continuous Welded Rails on Arch Bridge

    Directory of Open Access Journals (Sweden)

    Rong Chen

    2013-01-01

    Full Text Available Taking arch bridges, including deck, half-through, and through arch bridges (short for DAB, HTAB, and TAB as examples, mechanics analysis models of longitudinal interaction between continuously welded rails (short for CWRs and arch bridges are established. Based on the finite element method (FEM, the longitudinal interaction calculation software of CWR on arch bridges has been developed. Focusing on an HTAB, the tension, compression, and deflection conditions are calculated and analyzed. The results show that the mechanics analysis models of three types of arch bridges can truly reflect the real state of the structure; the calculation software can be used for systematic research of the CWR on arch bridge; as for HTAB, temperature difference of arch rib has a small effect on rail tension/compression, and arch bridge can be simplified as a continuous beam for rail tension/compression additional force calculation; in calculation of deflection conditions of HTAB, it is suggested that train loads are arranged on half span and full span and take the direction of load entering bridge into account. Additionally, the deflection additional force variation of CFST basket handle arch bridge is different from that of ordinary bridge.

  18. HTTR workshop (workshop on hydrogen production technology)

    International Nuclear Information System (INIS)

    Shiina, Yasuaki; Takizuka, Takakazu

    2004-12-01

    Various research and development efforts have been performed to solve the global energy and environmental problems caused by large consumption of fossil fuels. Research activities on advanced hydrogen production technology by the use of nuclear heat from high temperature gas cooled reactors, for example, have been flourished in universities, research institutes and companies in many countries. The Department of HTTR Project and the Department of Advanced Nuclear Heat Technology of JAERI held the HTTR Workshop (Workshop on Hydrogen Production Technology) on July 5 and 6, 2004 to grasp the present status of R and D about the technology of HTGR and the nuclear hydrogen production in the world and to discuss about necessity of the nuclear hydrogen production and technical problems for the future development of the technology. More than 110 participants attended the Workshop including foreign participants from USA, France, Korea, Germany, Canada and United Kingdom. In the Workshop, the presentations were made on such topics as R and D programs for nuclear energy and hydrogen production technologies by thermo-chemical or other processes. Also, the possibility of the nuclear hydrogen production in the future society was discussed. The workshop showed that the R and D for the hydrogen production by the thermo-chemical process has been performed in many countries. The workshop affirmed that nuclear hydrogen production could be one of the competitive supplier of hydrogen in the future. The second HTTR Workshop will be held in the autumn next year. (author)

  19. Characteristics of HTTR's startup physics tests

    International Nuclear Information System (INIS)

    Nojiri, N.; Nakano, M.; Takeuchi, M.; Pohl, P.; Yamashita, K.

    1997-01-01

    The High Temperature Engineering Test Reactor (HTTR) which is under construction by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium gas-cooled reactor with an outlet temperature of 950 deg. C and a thermal output of 30MW. The first criticality is expected at the end of October 1997. The start-up physics tests (SPTs) are planned in the period from mid 1997 to the end of 1998. Characteristic items of the SPTs are: 1) Criticality approach; 2) Tests on a preliminary annual core; 3) Measurement of scram reactivity; 4) Excess reactivity test; 5) Measurements along with a 2-step-scram reactor shutdown procedure. (author)

  20. Test results of HTTR control system

    International Nuclear Information System (INIS)

    Motegi, Toshihiro; Iigaki, Kazuhiko; Saito, Kenji; Sawahata, Hiroaki; Hirato, Yoji; Kondo, Makoto; Shibutani, Hideki; Ogawa, Satoru; Shinozaki, Masayuki; Mizushima, Toshihiko; Kawasaki, Kozo

    2006-06-01

    The plant control performance of the IHX helium flow rate control system, the PPWC helium flow rate control system, the secondary helium flow rate control system, the inlet temperature control system, the reactor power control system and the outlet temperature control system of the HTTR are obtained through function tests and power-up tests. As the test results, the control systems show stable control response under transient condition. Both of inlet temperature control system and reactor power control system shows stable operation from 30% to 100%, respectively. This report describes the outline of control systems and test results. (author)

  1. Characteristic test of initial HTTR core

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Shimakawa, Satoshi; Fujimoto, Nozomu; Goto, Minoru

    2004-01-01

    This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations

  2. HTTR plant dynamic simulation using a hybrid computer

    International Nuclear Information System (INIS)

    Shimazaki, Junya; Suzuki, Katsuo; Nabeshima, Kunihiko; Watanabe, Koichi; Shinohara, Yoshikuni; Nakagawa, Shigeaki.

    1990-01-01

    A plant dynamic simulation of High-Temperature Engineering Test Reactor has been made using a new-type hybrid computer. This report describes a dynamic simulation model of HTTR, a hybrid simulation method for SIMSTAR and some results obtained from dynamics analysis of HTTR simulation. It concludes that the hybrid plant simulation is useful for on-line simulation on account of its capability of computation at high speed, compared with that of all digital computer simulation. With sufficient accuracy, 40 times faster computation than real time was reached only by changing an analog time scale for HTTR simulation. (author)

  3. Innovative and basic researches for high temperature technologies at HTTR

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku

    1995-01-01

    The HTTR is the first HTGR which is under construction at JAERI. The objectives of the HTTR are to establish basic technologies for HTGRs, to upgrade technologies for HTGRs and to conduct innovative and basic researches for high temperature technologies. The first two are concerned with HTGR developments. The last one is not necessarily for HTGR developments, but for future innovative researches which are expected to be applied to various technologies. (author)

  4. Validation of JENDL-3.3 for the HTTR criticality

    International Nuclear Information System (INIS)

    Goto, Minoru; Nojiri, Naoki; Shimakawa, Satoshi

    2004-01-01

    Validation of JENDL-3.3 has been performed for the HTTR criticality using the MVP code with a ''lattice-cell'' of infinite models and a ''whole-core'' of finite models. It was found that the keff values calculated with JENDL-3.3 was decreased about 0.2-0.4%Δk from one with JENDL-3.2. The criticality prediction was closed to the experimental data in the critical approach situation of the HTTR. (author)

  5. Acceptance test for graphite components and construction status of HTTR

    International Nuclear Information System (INIS)

    Iyoku, T.; Ishihara, M.; Maruyama, S.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    In March, 1991, the Japan Atomic Energy Research Institute (JAERI) started to constructed the High Temperature engineering Test Reactor(HTTR) which is a 30-MW(thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on the core support graphite structures. Two types of graphite materials are used in the HTTR. One is the garde IG-110, isotropic fine grain graphite, another is the grade PGX, medium-to-fine grained molded graphite. These materials were selected on the basis of the appropriate properties required by the HTTR reactor design. Industry-wide standards for an acceptance test of graphite materials used as main components of a nuclear reactor had not been established. The acceptance standard for graphite components of the HTTR, therefore, was drafted by JAERI and reviewed by specialists outside JAERI. The acceptance standard consists of the material testing, non-destructive examination such as the ultrasonic and eddy current testings, dimensional and visual inspections and assembly test. Ultrasonic and eddy current testings are applied to graphite logs to detect an internal flaw and to graphite components to detect a surface flaw, respectively. The assembly test is performed at the works, prior to their installation in the reactor pressure vessel, to examine fabricating precision of each component and alignment of piled-up structures. The graphite components of the HTTR had been tested on the basis of the acceptance standard. It was confirmed that the graphite manufacturing process was well controlled and high quality graphite components were provided to the HTTR. All graphite components except for the fuel graphite blocks are to be installed in the reactor pressure vessel of the HTTR in September 1995. The paper describes the construction status of the HTTR focusing on the graphite components. The acceptance test results are also presented in this paper. (author). Figs

  6. Development program on HTTR heat application systems at JAERI

    International Nuclear Information System (INIS)

    Ogawa, M.; Inagaki, Y.; Nishihara, T.; Shimizu, S.; Shiozawa, S.; Miyamoto, Y.

    2000-01-01

    The High Temperature Engineering Test Reactor (HTTR), which is a Japanese High Temperature Gas-cooled Reactor (HTGR) with 30 MW thermal output at 950 deg. C of the coolant outlet temperature, was constructed at Oarai Research Establishment of Japan Atomic Energy Research Institute (JAERI). The HTTR has attained the first criticality on November 1998. In JAERI, a hydrogen production system was selected as a heat utilization system of the HTTR. The development program on the HTTR hydrogen production system consists of two parts: one is to establish technologies connecting the hydrogen production system with the HTTR, the other is to establish technologies producing hydrogen from water by using nuclear heat. Finally, hydrogen can be produced from water by using nuclear heat supplied by the HTTR. In the hydrogen production system connected to the HTTR at first, JAERI selected a steam reforming process because its technology had matured. The HTTR hydrogen production system adopting the steam reforming process is being designed to produce hydrogen of about 3800 Nm 3 /hr by using nuclear heat (10MW, 905 deg. C) supplied from the HTTR. The safety principle and criteria are also being investigated for the HTTR hydrogen production system. A facility for an out-of-pile test prior to the demonstration test with the HTTR hydrogen production system is under manufacturing to carry out tests of safety, controllability and performance. The out-of-pile test facility simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30. The tests will be started in 2001 and continued for 4 years or longer. In parallel to the tests, a hydrogen/tritium permeation test and a corrosion test of a catalyst tube of a steam reformer are being carried out to obtain data necessary for the design of the HTTR hydrogen production system. A kind of thermochemical method called IS process is under studying to produce hydrogen from water by

  7. Fatigue damage evaluation method for the longitudinal welded joint of a FBR main vessel in the vicinity of the sodium surface

    International Nuclear Information System (INIS)

    Tanigawa, Masayuki; Shimokoshi, Minoru; Negishi, Hitoshi; Nagata, Takashi

    1990-01-01

    Metallurgical discontinuities are dominant in the fatigue strength reductions at the welded joints of vessels whose surfaces could be finished. In the welded joints of SUS 304 with TYPE 308 weld metal fatigue strength reductions are caused by strain concentrations as the result of the softening of the weld metal. A combination model of two elastic fully plastic materials is applicable to the structures under thermal stresses where displacements are self-controlled. Metallurgical discontinuities are represented by the difference of the yield strength. The longitudinal welded joint of a large FBR main vessel in the vicinity of the sodium surface was analyzed using this model under various conditions related to the design. Strain concentrations at the welded joint could be evaluated using the elastic follow-up model. The maximum value of the elastic follow-up parameter was 3.0 if the yield stress ratio of the weld metal to the base metal was not less than 0.8. (author)

  8. WELDING METHOD

    Science.gov (United States)

    Cornell, A.A.; Dunbar, J.V.; Ruffner, J.H.

    1959-09-29

    A semi-automatic method is described for the weld joining of pipes and fittings which utilizes the inert gasshielded consumable electrode electric arc welding technique, comprising laying down the root pass at a first peripheral velocity and thereafter laying down the filler passes over the root pass necessary to complete the weld by revolving the pipes and fittings at a second peripheral velocity different from the first peripheral velocity, maintaining the welding head in a fixed position as to the specific direction of revolution, while the longitudinal axis of the welding head is disposed angularly in the direction of revolution at amounts between twenty minutas and about four degrees from the first position.

  9. Test programmes of HTTR for future generation HTGRs

    International Nuclear Information System (INIS)

    Kunitomi, K.; Tachibana, Y.; Takeda, T.; Saikusa, A.; Shiozawa, S.

    1997-01-01

    Test programs of the High Temperature Engineering Test Reactor (HTTR) for future generation High Temperature Gas-Cooled Reactors (HTGRs) have been established considering design and development status of HTGRs in Japan and the world. Test programs are divided into six categories, thermal hydraulics, fuel, safety, high temperature components, core physics and control-instrumentations. All programs are related to the technology of future generation HTGRs and will be submitted to a new Coordinated Research Program (CRP) so that all participants from the world in test programs of the HTTR can use measured data for their future generation.HTGRs. This paper describes test programs of the HTTR for the development of future generation HTGRs after explanation of a future generation HTGR in Japan. (author)

  10. Results of assembly test of HTTR reactor internals

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

  11. Research and development of HTTR hydrogen production systems

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku; Ogawa, Masuro; Inagaki, Yoshiyuki; Onuki, Kaoru; Takeda, Tetsuaki; Nishihara, Tetsuo; Hayashi, Koji; Kubo, Shinji; Inaba, Yoshitomo; Ohashi, Hirofumi

    2002-01-01

    The Japan Atomic Energy Research Institute (JAERI) has constructed the High Temperature Engineering Test Reactor (HTTR) with a thermal output of 30MW and a reactor out let coolant temper at ure of 950 .deg. C. There search and development (R and D) program on nuclear production of hydrogen was started on January in 1997 as a study consigned by Ministry of Education, Culture, Sports, Science and Technology. A hydrogen production system connected to the HTTR is being designed to be able to produce hydrogen of about 4000m 3 /h by steam reforming of natural gas, using a nuclear heat of 10MW supplied by the HTTR hydrogen production system. In order to confirm controllability, safety and performance of key components in the HTTR hydrogen production system, the facility for the out-of-pile test was constructed on the scale of approximately 1/30 of the HTTR hydrogen production system. In parallel to the out-of-pile test, the following tests as essential problem, a corrosion test of a reforming tube, a permeation test of hydrogen isotopes through heat exchanger and reforming tubes, and an integrity test of a high-temperature isolation valve are carried out to obtain detailed data for safety review and development of analytical codes. Other basis studies on the hydrogen production technology of thermochemical water splitting called an iodine sulfur (IS) process, has been carried out for more effective and various uses of nuclear heat. This paper describes the present status and a future plan on the R and D of the HTTR hydrogen production systems in JAERI

  12. A simplified 2D HTTR benchmark problem

    International Nuclear Information System (INIS)

    Zhang, Z.; Rahnema, F.; Pounders, J. M.; Zhang, D.; Ougouag, A.

    2009-01-01

    To access the accuracy of diffusion or transport methods for reactor calculations, it is desirable to create heterogeneous benchmark problems that are typical of relevant whole core configurations. In this paper we have created a numerical benchmark problem in 2D configuration typical of a high temperature gas cooled prismatic core. This problem was derived from the HTTR start-up experiment. For code-to-code verification, complex details of geometry and material specification of the physical experiments are not necessary. To this end, the benchmark problem presented here is derived by simplifications that remove the unnecessary details while retaining the heterogeneity and major physics properties from the neutronics viewpoint. Also included here is a six-group material (macroscopic) cross section library for the benchmark problem. This library was generated using the lattice depletion code HELIOS. Using this library, benchmark quality Monte Carlo solutions are provided for three different configurations (all-rods-in, partially-controlled and all-rods-out). The reference solutions include the core eigenvalue, block (assembly) averaged fuel pin fission density distributions, and absorption rate in absorbers (burnable poison and control rods). (authors)

  13. Fabrication of HTTR first loading fuel

    International Nuclear Information System (INIS)

    Kato, S.; Yoshimuta, S.; Hasumi, T.; Sato, K.; Sawa, K.; Suzuki, S.; Mogi, H.; Shiozawa, S.; Tanaka, T.

    2001-01-01

    This paper summarizes the fabrication of the first loading fuel for HTTR, High Temperature engineering Test Reactor constructed by JAERI, Japan Atomic Energy Research Institute. The fuel fabrication started at the HTR fuel facility of NFI, Nuclear Fuel Industries, Ltd., June 1995. 4,770 fuel rods were fabricated through the fuel kernel, coated fuel particle and fuel compaction process, then 150 fuel elements were assembled in the reactor building December 1997. Fabrication technology for the fuel was established through a lot of R and D activities and fabrication experience of irradiation examination samples spread over about 30 years. Most of all, very high quality and production efficiency of fuel were achieved by the development of the fuel kernel process using the vibration dropping technology, the continuous 4-layer coating process and the automatic compaction process. As for the inspection technology, the development of the automatic measurement equipment for coated layer thickness of a coated fuel particle and uranium content of a fuel compact contributed to the higher reliability and rationalization of the inspection process. The data processing system for the fabrication and quality control, which was originally developed by NFI, made possible not only quick feedback of statistical quality data to the fabrication processes, but also automatic document preparation, such as inspection certificates and accountability control reports. The quality of the first loading fuel fully satisfied the design specifications for the fuel. In particular, average bare uranium fraction and SiC defective fraction of fuel compacts were 2x10 -6 and 8x10 -5 , respectively. According to the preceding irradiation examinations being performed at JMTR, Japan Materials Testing Reactor of JAERI, the specimen sampled from the first loading fuel shows good irradiation performance. (author)

  14. Seismic scrammability of HTTR control rods

    International Nuclear Information System (INIS)

    Nishiguchi, I.; Iyoku, T.; Ito, N.; Watanabe, Y.; Araki, T.; Katagiri, S.

    1990-01-01

    Scrammability tests on HTTR (High-Temperature Engineering Test Reactor) control rods under seismic conditions have been carried out and seismic conditions influences on scram time as well as functional integrity were examined. A control rod drive located in a stand-pipe at the top of a reactor vessel, raises and lowers a pair of control rods by suspension cables. Each flexible control rod consists of 10 neutron absorber sections held together by a metal spine passing through the center. It falls into a hole in graphite blocks due to gravity at scram. In the tests, a full scale control rod drive and a pair of control rods were employed with a column of graphite blocks in which holes for rods were formed. Blocks misalignment and contact with the hole surface during earthquakes were considered as major causes of disturbance in scram time. Therefore, the following parameters were set up in the tests: excitation direction, combination or horizontal and vertical excitation, acceleration, frequency and block to block gaps. Main results obtained from tests are as follow. 1) Every scram time obtained under the design conditions was within 6 seconds. On the contrary, the scram times were 5.2 seconds when there were no vibration. Therefore, it was concluded that the seismic effects on scram time were not significant. 2) Scram time became longer with increase in both acceleration and horizontal excitation frequency, and control rods fell very smoothly without any jerkiness. This suggests that collision between control rods and hole surface is the main disturbing factor of falling motion. 3) Mechanical and functional integrity of control rod drive mechanism, control rods and graphite blocks was confirmed after 140 seismic scrammability tests. (author). 10 figs, 1 tab

  15. Production process and quality control for the HTTR fuel

    International Nuclear Information System (INIS)

    Yoshimuta, S.; Suzuki, N.; Kaneko, M.; Fukuda, K.

    1991-01-01

    Development of the production and inspection technology for High Temperature Engineering Test Reactor (HTTR) fuel has been carried out by cooperative work between Japan Atomic Energy Research Institute (JAERI) and Nuclear Fuel Industries, Ltd (NFI). The performance and the quality level of the developed fuel are well established to meet the design requirements of the HTTR. For the commercial scale production of the fuel, statistical quality control and quality assurance must be carefully considered in order to assure the safety of the HTTR. It is also important to produce the fuel under well controlled process condition. To meet these requirements in the production of the HTTR fuel, a new production process and quality control system is to be introduced in the new facilities. The main feature of the system is a computer integrated control system. Process control data at each production stage of products and semi-products are all gathered by terminal computers and processed by a host computer. The processed information is effectively used for the production, quality and accountancy control. With the aid of this system, all the products will be easily traceable from starting materials to final stages and the statistical evaluation of the quality of products becomes more reliable. (author). 8 figs

  16. HTTR demonstration test plan for industrial utilization of nuclear heat

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Yan, Xing L.; Kubo, Shinji; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki

    2014-09-01

    Japan Atomic Energy Agency has been conducting research and development with a central focus on the utilization of High Temperature engineering Test Reactor (HTTR), the first High Temperature Gas-cooled Reactor (HTGR) in Japan, towards the realization of industrial use of nuclear heat. Several studies have made on the integration of the HTTR with thermochemical iodine-sulfur process and steam methane reforming hydrogen production plant (H 2 plant) as well as helium gas turbine power conversion system. In addition, safety standards for coupling a H 2 plant to a nuclear facility has been investigated. Based on the past design information, the present study identified test items to be validated in the HTTR demonstration test to accomplish a formulation of safety requirement and design consideration for coupling a H 2 plant to a nuclear facility as well as confirmation of overall performance of helium gas turbine system. In addition, plant concepts for the heat utilization system to be connected with the HTTR are investigated. (author)

  17. HTTR demonstration program for nuclear cogeneration of hydrogen and electricity

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Sumita, Junya; Terada, Atsuhiko; Ohashi, Hirofumi; Yan, Xing L.; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki

    2015-01-01

    Japan Atomic Energy Agency initiated a High Temperature Engineering Test Reactor (HTTR) demonstration program in accordance with recommendations of a task force established by Ministry of Education, Culture, Sports, Science and Technology according to the Strategic Energy Plan as of April 2014. The demonstration program is designed to complete helium gas turbine and hydrogen production system technologies aiming at commercial plant deployment in 2030s. The program begins with coupling a helium gas turbine in the secondary loop of the HTTR and expands by adding the H 2 plant to a tertiary loop to enable hydrogen cogeneration. Safety standards for coupling the helium gas turbine and H 2 plant to the nuclear reactor will be established through safety review in licensing. A system design and its control method are planned to be validated with a series of test operations using the HTTR-GT/H 2 plant. This paper explains the outline of HTTR demonstration program with a plant concept of the heat application system directed at establishing an HTGR cogeneration system with 950°C reactor outlet temperature for production of power and hydrogen as recommended by the task force. Commercial deployment strategy including a development plan for the helium gas turbine is also presented. (author)

  18. Pulsed ultrasonic stir welding system

    Science.gov (United States)

    Ding, R. Jeffrey (Inventor)

    2013-01-01

    An ultrasonic stir welding system includes a welding head assembly having a plate and a rod passing through the plate. The rod is rotatable about a longitudinal axis thereof. During a welding operation, ultrasonic pulses are applied to the rod as it rotates about its longitudinal axis. The ultrasonic pulses are applied in such a way that they propagate parallel to the longitudinal axis of the rod.

  19. Characteristics of first loaded IG-110 graphite in HTTR core

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Sawa, Kazuhiro; Hanawa, Satoshi; Ishihara, Masahiro

    2006-10-01

    IG-110 graphite is a fine-grained isotropic and nuclear-grade graphite with excellent resistivity on both irradiation and corrosion and with high strength. The IG-110 graphite is used for the graphite components of High Temperature Engineering Test Reactor (HTTR) such as fuel and control rod guide blocks and support posts. In order to design and fabricate the graphite components in the HTTR, the Japan Atomic Energy Research Institute (the Japan Atomic Energy Agency at present) had established the graphite structural design code and design data on the basis of former research results. After the design code establishment, the IG-110 graphite components were fabricated and loaded in the HTTR core. This report summarized the characteristics of the first loaded IG-110 graphite as basic data for surveillance test, measuring material characteristics changed by neutron irradiation and oxidation. By comparing the design data, it was shown that the first loaded IG-110 graphite had excellent strength properties and enough safety margins to the stress limits in the design code. (author)

  20. Conceptual design of the HTTR-IS hydrogen production system

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Sato, Hiroyuki; Hara, Teruo; Kato, Ryoma; Ohashi, Kazutaka; Nishihara, Tetsuo; Kunitomi, Kazuhiko

    2007-08-01

    Since hydrogen produced by nuclear should be economically competitive compared with other methods in a hydrogen society, it is important to build hydrogen production system to be coupled with the reactor as a conventional chemical plant. Japan Atomic Energy Agency started the safety study to establish a new safety philosophy to meet safety requirements for non-nuclear grade hydrogen production system. Also, structural concepts with integrating functions for the Bunsen reactor and sulphuric acid decomposer were proposed to reduce construction cost of the IS process hydrogen production system. In addition, HI decomposer which enables the process condition to be eased consisting of conventional materials and technologies was studied. Moreover, technical feasibility of the HTTR-IS system in which the hydrogen production rate of 1,000 Nm 3 /h by using the supplied heat of 10 MW from the intermediate heat exchanger of the HTTR was confirmed. This paper describes the conceptual design of the HTTR-IS hydrogen production system. (author)

  1. Design of high temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

    1994-09-01

    Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

  2. Study on control characteristics for HTTR hydrogen production system with mock-up test facility

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo; Ohashi, Hirofumi; Nishihara, Tetsuo; Sato, Hiroyuki; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Hayashi, Koji; Takada, Shoji

    2005-01-01

    The Japan Atomic Energy Research Institute has a demonstration test plan of a hydrogen production system by steam reforming of methane coupling with the High-Temperature Engineering Test Reactor (HTTR). Prior to the coupling of a hydrogen production plant with the HTTR, simulation tests with a mock-up test facility of the HTTR hydrogen production system (HTTR-H2) is underway. The test facility is a 1/30-scale of the HTTR-H2 and simulates key components downstream from an intermediate heat exchanger of the HTTR. The main objective of the simulation tests is the establishment and demonstration of control technology, focusing on the mitigation of a thermal disturbance to the reactor by a steam generator (SG) and on the controllability of the pressure difference between the helium and process gases at the reaction tube in a steam reformer (SR). It was confirmed that the fluctuation of the outlet helium gas temperature at the SG and the pressure difference in the SR can be controlled within the allowable range for the HTTR-H2 in the case of the system controllability test for the fluctuation of chemical reaction. In addition, a dynamic simulation code for the HTTR-H2 was verified with the obtained test data

  3. 46 CFR 154.660 - Pipe welding.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Pipe welding. 154.660 Section 154.660 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY STANDARDS FOR... § 154.660 Pipe welding. (a) Pipe welding must meet Part 57 of this chapter. (b) Longitudinal butt welds...

  4. 49 CFR 179.300-9 - Welding.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Welding. 179.300-9 Section 179.300-9... Specifications for Multi-Unit Tank Car Tanks (Classes DOT-106A and 110AW) § 179.300-9 Welding. (a) Longitudinal... fusion welded on class DOT-110A tanks. Welding procedures, welders and fabricators must be approved in...

  5. Pulsed ultrasonic stir welding method

    Science.gov (United States)

    Ding, R. Jeffrey (Inventor)

    2013-01-01

    A method of performing ultrasonic stir welding uses a welding head assembly to include a plate and a rod passing through the plate. The rod is rotatable about a longitudinal axis thereof. In the method, the rod is rotated about its longitudinal axis during a welding operation. During the welding operation, a series of on-off ultrasonic pulses are applied to the rod such that they propagate parallel to the rod's longitudinal axis. At least a pulse rate associated with the on-off ultrasonic pulses is controlled.

  6. Welding technologies for nuclear machinery and equipment

    International Nuclear Information System (INIS)

    Kobayashi, Masahiro; Yokono, Tomomi.

    1991-01-01

    The main welding methods applied to nuclear machinery and equipment are shielded metal arc welding, submerged arc welding, MAG welding and TIG welding. But in the last 10 years, in order to improve the reliability required for the welding of nuclear machinery and equipment, the welding technologies aiming at the reduction of heat input, the decrease of the number of welding pass and the automatic control of welding factors have been applied for the main purpose of bettering the quality and excluding human errors. The merits and the technology of narrow gap, pulsed MAG welding and melt-through welding are explained. As the automation of TIG welding, image processing type narrow gap, hot wire TIG welding and remote control type automatic TIG welding are described. For the longitudinal welding of active metal sheet products, plasma key-hole welding is applied. Since the concentration of its arc is good, high speed welding with low heat input can be done. For the stainless steel cladding by welding, electroslag welding has become to be employed in place of conventional submerged arc welding. Arc is not generated in the electroslag welding, and the penetration into base metal is small. (K.I.)

  7. Prospects of HTGR process heat application and role of HTTR

    International Nuclear Information System (INIS)

    Shiozawa, S.; Miyamoto, Y.

    2000-01-01

    At Japan Atomic Energy Research Institute, an effort on development of process heat application with high temperature gas cooled reactor (HTGR) has been continued for providing a future clean alternative to the burning of fossil energy for the production of industrial process heat. The project is named 'HTTR Heat Utilization Project', which includes a demonstration of hydrogen production using the first Japanese HTGR of High Temperature Engineering Test Reactor (HTTR). In the meantime, some countries, such as China, Indonesia, Russia and South Africa are trying to explore the HTGR process heat application for industrial use. One of the key issues for this application is economy. It has been recognized for a long time and still now that the HTGR heat application system is not economically competitive to the current fossil ones, because of the high cost of the HTGR itself. However, the recent movement on the HTGR development, as represented by South Africa Pebble Beds Modular Reactor (SA-PBMR) Project, has revealed that the HTGRs are well economically competitive in electricity production to fossil fuel energy supply under a certain condition. This suggests that the HTGR process heat application will also possibly get economical in the near future. In the present paper, following a brief introduction describing the necessity of the HTGRs for the future process heat application, Japanese activities and prospect of the development on the process heat application with the HTGRs are described in relation with the HTTR Project. In conclusion, the process heat application system with HTGRs is thought technically and economically to be one of the most promising applications to solve the global environmental issues and energy shortage which may happen in the future. However, the commercialization for the hydrogen production system from water, which is the final goal of the HTGR process heat application, must await the technology development to be completed in 2030's at the

  8. Analytical method and result of radiation exposure for depressurization accident of HTTR

    International Nuclear Information System (INIS)

    Sawa, K.; Shiozawa, S.; Mikami, H.

    1990-01-01

    The Japan Atomic Energy Research Institute (JAERI) is now proceeding with the construction design of the High Temperature Engineering Test Reactor (HTTR). Since the HTTR has some characteristics different from LWRs, analytical method of radiation exposure in accidents provided for LWRs can not be applied directly. This paper describes the analytical method of radiation exposure developed by JAERI for the depressurization accident, which is the severest accident in respect to radiation exposure among the design basis accidents of the HTTR. The result is also described in this paper

  9. Test plan using the HTTR for commercialization of GTHTR300C

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Sakaba, Nariaki; Ohashi, Hirofumi; Sato, Hiroyuki; Ueta, Shohei; Aihara, Jun; Goto, Minoru; Sumita, Junya; Shibata, Taiju; Takamatsu, Kuniyoshi; Inagaki, Yoshiyuki; Kunitomi, Kazuhiko; Nishihara, Tetsuo; Hamamoto, Shimpei; Iyoku, Tatsuo

    2010-02-01

    The High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor (HTGR) built at the Oarai Research and Development Center of JAEA with thermal power of 30 MW and the maximum reactor outlet coolant temperature of 950degC. The HTTR achieved the first criticality in 1998, the reactor outlet coolant temperature of 950degC in 2004, and 30 days continuous operation in 2007. Since 2002, safety demonstration tests including reactivity insertion tests and coolant flow reduction tests have been conducted to show inherent safety features of the HTGRs by using the HTTR. This report describes full scope of the future feasible test plan mainly using the HTTR. The test items cover fuel performance and radionuclide transport, core physics, reactor thermal hydraulics and plant dynamics, and reactor operations, maintenance, control, etc. The test results will be utilized for realization of Japan's commercial Very High Temperature Reactor (VHTR) system, GTHTR300C. (author)

  10. Evaluation of hydrogen production system coupling with HTTR using dynamic analysis code

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo; Hayashi, Koji; Inagaki, Yoshiyuki

    2006-01-01

    The Japan Atomic Energy Agency (JAEA) was entrusted 'Development of Nuclear Heat Utilization Technology' by Ministry of Education, Culture, Sports, Science and Technology. In this development, the JAEA investigated the system integration technology to couple the hydrogen production system by steam reforming with the High Temperature Engineering Test Reactor (HTTR). Prior to the construction of the hydrogen production system coupling with the HTTR, a dynamic analysis code had to be developed to evaluate the system transient behaviour of the hydrogen production system because there are no examples of chemical facilities coupled with nuclear reactor in the world. This report describes the evaluation of the hydrogen production system coupling with HTTR using analysis code, N-HYPAC, which can estimate transient behaviour of the hydrogen production system by steam reforming. The results of this investigation provide that the influence of the thermal disturbance caused by the hydrogen production system on the HTTR can be estimated well. (author)

  11. Preliminary safety analysis of the HTTR-IS nuclear hydrogen production system

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Tachibana, Yukio; Sakaba, Nariaki

    2010-06-01

    Japan Atomic Energy Agency is planning to demonstrate hydrogen production by thermochemical water-splitting IS process utilizing heat from the high-temperature gas-cooled reactor HTTR (HTTR-IS system). The previous study identified that the HTTR modification due to the coupling of hydrogen production plant requires an additional safety review since the scenario and quantitative values of the evaluation items would be altered from the original HTTR safety review. Hence, preliminary safety analyses are conducted by using the system analysis code. Calculation results showed that evaluation items such as a coolant pressure, temperatures of heat transfer tubes at the pressure boundary, etc., did not exceed allowable values. Also, the peak fuel temperature did not exceed allowable value and therefore the reactor core was not damaged and cooled sufficiently. This report compiles calculation conditions, event scenarios and the calculation results of the preliminary safety analysis. (author)

  12. Uncertainties in calculations of nuclear design code system for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Shindo, R.; Yamashita, K.; Murata, I.

    1991-01-01

    The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs

  13. Nondestructive testing on graphite structures for high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Kambe, Mamoru; Tsuji, Nobumasa.

    1994-01-01

    The application of ultrasonic (for internal defects) and eddy current testing (for surface defects) were investigated on the structures of nuclear-grade IG-110 and PGX graphite for the HTTR. The equipment were developed in order to detect the specific configuration of graphite blocks and the testing conditions were defined as the practical testing methods. The established testing methods are being used for the acceptance tests of graphite structures in the HTTR. (author)

  14. Out-of-pile demonstration test of hydrogen production system coupling with HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Nishihara, Tetsuo; Takeda, Tetsuaki; Hada, Kazuhiko; Hayashi, Koji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1999-07-01

    In Japan Atomic Energy Research Institute, a hydrogen production system is being designed to produce hydrogen by means of a steam reforming process of natural gas using nuclear heat (10 MW, 905degC) supplied by the High Temperature Engineering Test Reactor (HTTR). The safety principle and criteria are also being investigated in the HTTR hydrogen production system. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile demonstration test was planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 has a hydrogen production capacity of 110 Nm{sup 3}/h using an electric heater as a reactor substitute. The test facility is under manufacturing aiming at completion in 2000 and followed by the test till 2004. In parallel to this, a hydrogen permeation test and a corrosion test of a catalyst tube of a steam reformer are being carried out to obtain data necessary for the design of the system. This report describes outline of the out-of-pile hydrogen production facility and demonstration test program for the HTTR hydrogen production system at present status. (author)

  15. Development of irradiation rig in HTTR and dosimetry method. I-I type irradiation equipment

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Taiju; Kikuchi, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi; Ogura, Kazutomo [Japan Atomic Power Co., Tokyo (Japan)

    2002-12-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium gas-cooled test reactor with a maximum power of 30 MW. The HTTR aims not only to establish and upgrade the technological basis for the HTGRs but also to perform the innovative basic research on high temperature engineering with high temperature irradiation fields. It is planned that the HTTR is used to perform various engineering tests such as the safety demonstration test, high temperature test operation and irradiation test with large irradiation fields at high temperatures. This paper describes the design of the I-I type irradiation equipment developed as the first irradiation rig for the HTTR and does the planned dosimetry method at the first irradiation test. It was developed to perform in-pile creep test on a stainless steel with large standard size specimens in the HTTR. It can give great loads on the specimens stably and can control the irradiation temperature precisely. The in-core creep properties on the specimens are measured by newly developed differential transformers and the irradiation condition in the core is monitored by thermocouples and self-powered neutron detectors (SPNDs), continuously. The irradiated neutron fluence is assessed by neutron fluence monitors of small metallic wires after the irradiation. The obtained data at the first irradiation test can strongly be contributed to upgrade the technological basis for the HTGRs, since it is the first direct measurement of the in-core irradiation environments of the HTTR. (author)

  16. Radiation monitoring for the HTTR rise-to-power test (1) and (2)'

    Energy Technology Data Exchange (ETDEWEB)

    Nakazawa, Takashi; Yoshino, Toshiaki; Yasu, Katsuji; Ashikagaya, Yoshinobu; Kikuchi, Toshiki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-02-01

    The High Temperature Engineering Test Reactor (HTTR) is the first high temperature gas-cooled research reactor in Japan. This reactor is a helium-gas-cooled and graphite-moderated reactor with a thermal output of 30 MW. The rated operation temperature of the outlet coolant is 850degC. (During high temperature test operation, this reaches 950degC). The first criticality of the HTTR was attained in November 1998. The single loaded, parallel loaded operation with a thermal output of 9 MW (called the HTTR Rise-to-Power Test (1)) was completed between September 16, 1999 and July 8, 2000. The single loaded, parallel loaded continuous operation with a thermal output of 20 MW (called the HTTR Rise-to-Power Test (2)) has also been carried out, but it was shutdown at the halfway stage by a single from the reactor, when the thermal output was 16.5 MW and the reactor outlet coolant temperature was 500degC. This report describes the radiation monitoring carried out during the HTTR Rise-to-Power Tests (1) and (2)'. The data measured by the various radiation monitors is also reported. These data will be used for the estimation of radiation levels (such as the radiation dose equivalent rate, the radioactive concentration in effluents, etc.) for the next HTTR Rise-to-Power Test, and for periodic inspections. (author)

  17. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

    International Nuclear Information System (INIS)

    2014-12-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30MW in December 2001 and achieved the 950degC of outlet coolant temperature at the outside the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2013, we started to prepare the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 when the Pacific coast of Tohoku Earthquake (2011.3.11) occurred. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2013. (author)

  18. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2014

    International Nuclear Information System (INIS)

    2016-02-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30 MW in December 2001 and achieved the 950degC of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2014, we started to apply the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 by the Pacific coast of Tohoku Earthquake. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2014. (author)

  19. Out-of-pile demonstration test of hydrogen production system coupling with HTTR

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Nishihara, Tetsuo; Takeda, Tetsuaki; Hada, Kazuhiko; Hayashi, Koji

    1999-01-01

    In Japan Atomic Energy Research Institute, a hydrogen production system is being designed to produce hydrogen by means of a steam reforming process of natural gas using nuclear heat (10 MW, 905degC) supplied by the High Temperature Engineering Test Reactor (HTTR). The safety principle and criteria are also being investigated in the HTTR hydrogen production system. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile demonstration test was planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 has a hydrogen production capacity of 110 Nm 3 /h using an electric heater as a reactor substitute. The test facility is under manufacturing aiming at completion in 2000 and followed by the test till 2004. In parallel to this, a hydrogen permeation test and a corrosion test of a catalyst tube of a steam reformer are being carried out to obtain data necessary for the design of the system. This report describes outline of the out-of-pile hydrogen production facility and demonstration test program for the HTTR hydrogen production system at present status. (author)

  20. The Assessment Of High Temperature Reactor Fuel (Characteristics Of HTTR Fuel)

    International Nuclear Information System (INIS)

    Dewita, Erlan; Tuka, Veronica; Gunandjar

    1996-01-01

    HTTR is one of the reactor type with Helium coolant and outlet coolant temperature of 950 o C. One possibility of HTTR application is the coo generation of steam in high temperature and electric power for supply energy to industry in the future. Considering to the high operating temperature of HTTR, therefore it is needed the reactor fuel which have good mechanical, chemical and physical stability to the high temperature, and stable to the influence of fission fragment and neutron during irradiation. This assessment of the HTTR fuel characteristic based on the experiment data to find information of HTTR operation feasibility. Result of the assessment indicated that fission gas release at burn-up of 3.6 % FIMA which was the same as the maximum burn up in the HTTR design was fairly lower than the maximum release estimated in the design (5 x 10 - 4), which is R/B from the fuel fabricated by the prismatic block fuel method would be low (between 10 - 9 dan 10 - 8)

  1. Design of a steam reforming system to be connected to the HTTR

    International Nuclear Information System (INIS)

    Hada, K.; Nishihara, T.; Shibata, T.; Shiozawa, S.

    1996-01-01

    Top priority objective for developing the first heat utilization system to be connected to the HTTR is to demonstrate technical feasibility of a nuclear process heat utilization system for production of hydrogen for the first time in the world. Major issues to be resolved for coupling the heat utilization system to the HTTR are 1)to develop safety philosophy for reasonably and reliably ensuring safety of the nuclear reactor, 2)to develop control design concept for the total system of the nuclear reactor and heat utilization system because thermal dynamics of endothermic chemical reactor to be heated by nuclear heat is much different from the nuclear reactor, 3)to develop helium-heated components and 4)to develop enhanced hydrogen production technologies for achieving competitiveness to a fossil-fired plant. A steam reforming hydrogen production system was studied as one of the first priority candidates for an HTTR-heat utilization system due to matured technology in fossil-fired plants and since technical solutions demonstrated by the coupling of the steam reforming system to the HTTR will contribute to all other hydrogen production systems. Basic design philosophy for the HTTR-steam reforming system is that the steam reforming plant downstream of an intermediate secondary helium loop is designed at the same safety level as fossil-fired plants and therefore the secondary helium loop was selected as a safety barrier to the HTTR nuclear reactor. (J.P.N.)

  2. Status of HTTR and Technology Developments for Near Term Deployment of Nuclear Process Heat Applications in Japan

    International Nuclear Information System (INIS)

    Yan, Xing L.

    2017-01-01

    JAEA's High Temperature Engineering Test Reactor: Main features: Thermal power 30 MWt; Fuel SiC TRISO UO2 coated particle fuel, pin in block; Design type Prismatic core; Coolant Helium; Temperature 950 °C (Max.); Pressure 4 MPa. Proven quality of HTTR coated particle fuel - 50-day 950°C operation test in HTTR (2009): • Proved fuel production quality in commercial-scale facility - Fabricated 2 core loads of fuel for HTTR (2 t of U used); - Failure fraction: 2 orders of magnitude < licensed limits • Demonstrated fuel integrity in HTTR operation and extended burnup test: an order of magnitude < licensed limit

  3. Improvement in oil seal performance of gas compressor in HTTR

    International Nuclear Information System (INIS)

    Oyama, Sunao; Hamamoto, Shimpei; Nemoto, Takahiro; Sekita, Kenji; Isozaki, Minoru; Emori, Kouichi; Ohta, Yukimaru; Mizushima, Toshihiko; Kaneshiro, Noriyuki; Ito, Yoshiteru; Yamamoto, Hideo

    2007-08-01

    High-Temperature engineering Test Reactor (HTTR) built by Japan Atomic Energy Agency (JAEA) has reciprocating compressor commonly used to extract and discharge helium gas into primary/secondary coolant helium loop from helium purification system. Piston rod seal of the compressor consist of several components to prevent coolant leak. However, rod seal system has weak reliability during long term operation due to repeated leakage of seal oil in operation. As a result of investigations, leakage's root is found in that seal were used in a range beyond limit sliding properties of seal material. For this reason, a lip of the seal was worn and transformed itself and was not able to sustain a seal function. Therefore, through tests using facility actual equipment for endurance of candidate materials, one seal material were chosen for long term operation. (author)

  4. Structural integrity of graphite core support structures of HTTR

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Iyoku, Tatsuo; Toyota, Junji; Sato, Sadao; Shiozawa, Shusaku

    1990-02-01

    The graphite core support structures (GCSSs) of the HTTR (High Temperature Engineering Test Reactor) are an arrangement of graphite blocks and posts that support the core and provide a lower plenum and a hot-leg path for the primary coolant. The GCSSs are designed not to be replaced by new items during plant life time (about twenty years). To maintain structural integrity of the GCSSs, conservative design has been made sufficiently on the basis of structural tests. The present study confirmed that reactor safety was still maintained even if failure and destruction of the GCSSs is supposed to occur. The GCSSs are fabricated under strict quality control and the observation and surveillance programs are planed to examine the structual integrity of the GCSSs during an operation. This paper describes the concept of design and quality control and summarizes structural tests, observation and surveillance programs. (author)

  5. Construction of the HTTR in-core components

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Jinza, K.; Miki, T.

    1996-01-01

    The reactor internals of HTTR consist of graphite and metallic core support structures and shielding blocks and are designed to support core elements and to shield neutron fluence. They also have functions to restrict by-pass flow for ensuring the core cooling performance and to maintain the temperature of metallic core support structures within their design limits. The detailed design of the HTTR core support structure was approved by the government through safety review, 1990-1991. Machining of all graphite components, which consist of about 150 large blocks, was finished in September 1994 successfully. Machining and fabricating of the metallic components were also finished in September. Prior to their installation in the reactor pressure vessel (RPV), the assembly test of actual reactor internals was performed at the works to confirm above mentioned functions. The assembly test was conducted by examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the RPV and reactor internals as well as under the core support plates with respect to structural integrity, and measuring by-pass flow rate through gaps between graphite components which may degrade core performance. The another purpose of the assembly test was to confirm the installation procedure of those components. All components were assembled at the works according to the planned procedure, and the tests were executed while assembling. As a result of the tests, measured level difference and gap width between reactor internals were negligible from core thermal and hydraulic performance point of view. Coolant flows uniformly in circumferential direction at any axial level in the RPV. By-pass flow rate was found to be suppressed sufficiently and far less than the design limit. (J.P.N.)

  6. System analysis for HTTR-GT/H2 plant. Safety analysis of HTTR for coupling helium gas turbine and H2 plant

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Yan, Xing L.; Ohashi, Hirofumi

    2017-08-01

    High Temperature Gas-cooled Reactor (HTGR) is expected to extend the use of nuclear heat to a wider spectrum of industrial applications because of the high temperature heat supply capability and inherently safe characteristics. Japan Atomic Energy Agency initiated a nuclear cogeneration demonstration project with helium gas turbine power generation and thermochemical hydrogen production utilizing the High Temperature engineering Test Reactor (HTTR), the first HTGR in Japan. This study carries out safety evaluation for the HTTR gas turbine hydrogen cogeneration test plant (HTTR-GT/H 2 plant). The evaluation was conducted for the events newly identified corresponding to the coupling of helium gas turbine and hydrogen production plant to the HTTR. The results showed that loss of load event does not have impact on temperature of fuel and reactor coolant pressure boundary. In addition, reactor coolant pressure does not exceed the evaluation criteria. Furthermore, it was shown that reactor operation can be maintained against temperature transients induced by abnormal events in hydrogen production plant. (author)

  7. Performance test results of mock-up test facility of HTTR hydrogen production system

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo

    2004-01-01

    For the purpose to demonstrate effectiveness of high-temperature nuclear heat utilization, Japan Atomic Energy Research Institute has been developing a hydrogen production system and has planned to connect the hydrogen production system to High Temperature Engineering Test Reactor (HTTR). Prior to construction of a HTTR hydrogen production system, a mock-up test facility was constructed to investigate transient behavior of the hydrogen production system and to establish system controllability. The Mock-up test facility with a full-scale reaction tube is an approximately 1/30-scale model of the HTTR hydrogen production system and an electric heater is used as a heat source instead of a reactor. After its construction, a performance test of the test facility was carried out in the same pressure and temperature conditions as those of the HTTR hydrogen production system to investigate its performance such as hydrogen production ability, controllability and so on. It was confirmed that hydrogen was stably produced with a hot helium gas about 120m 3 /h, which satisfy the design value, and thermal disturbance of helium gas during the start-up could be mitigated within the design value by using a steam generator. The mock-up test of the HTTR hydrogen production system using this facility will continue until 2004. (author)

  8. Preliminary test results for post irradiation examination on the HTTR fuel

    International Nuclear Information System (INIS)

    Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

    2007-01-01

    The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

  9. Development of control technology for HTTR hydrogen production system with mock-up test facility

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo; Takeda, Tetsuaki; Hayashi, Koji; Takada, Shoji; Inagaki, Yoshiyuki

    2006-01-01

    The Japan Atomic Energy Agency has been planning the demonstration test of hydrogen production with the High Temperature Engineering Test Reactor (HTTR). In a HTTR hydrogen production system (HTTR-H2), it is required to control a primary helium temperature within an allowable value at a reactor inlet to prevent a reactor scram. A cooling system for a secondary helium with a steam generator (SG) and a radiator is installed at the downstream of a chemical rector in a secondary helium loop in order to mitigate the thermal disturbance caused by the hydrogen production system. Prior to HTTR-H2, the simulation test with a mock-up test facility has been carried out to establish the controllability on the helium temperature using the cooling system against the loss of chemical reaction. It was confirmed that the fluctuations of the helium temperature at chemical reactor outlet, more than 200 K, at the loss of chemical reaction could be successfully mitigated within the target of ±10 K at SG outlet. A dynamic simulation code of the cooling system for HTTR-H2 was verified with the obtained test data

  10. Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

    1990-01-01

    The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

  11. Present status of high-temperature engineering test reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1994-01-01

    The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950degC at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test using HTTR. (author)

  12. Present status of High-Temperature engineering Test Reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1993-01-01

    The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950 deg C at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test plan using HTTR. (author)

  13. Application of new design methodologies to very high-temperature metallic components of the HTTR

    International Nuclear Information System (INIS)

    Hada, Kazuhiko; Ohkubo, Minoru; Baba, Osamu

    1991-01-01

    The high-temperature piping and helium-to-helium intermediate heat exchanger of the High-Temperature Engineering Test Reactor (HTTR) are designed to be operating at very high temperatures of about 900deg C among the class 1 components of the HTTR. At such a high temperature, mechanical strength of heat-resistant metallic materials is very low and thermal expansions of structural members are large. Therefore, innovative design methodologies are needed to reduce both mechanical and thermal loads acting on these components. To the HTTR, the design methodologies which can separate the heat-resistant function from the pressure-retaining functions and allow them to expand freely are applied to reduce pressure and thermal loads. Since these design methodologies need to verify their applicability, the Japan Atomic Energy Research Institute (JAERI) has been performing many design and research works on their verifications. The details of the design methodologies and their verifications are given in this paper. (orig.)

  14. Evaluation of local power distribution with fine-mesh core model for the HTTR

    International Nuclear Information System (INIS)

    Murata, Isao; Yamashita, Kiyonobu; Maruyama, So; Shindo, Ryuichi; Fujimoto, Nozomu; Sudo, Yukio; Nakata, Tetsuo.

    1991-01-01

    An evaluation method of the local power distribution was developed considering the radial and axial heterogeneity caused by fuel rods, BP rods and block end graphite for the High Temperature Engineering Test Reactor (HTTR) in Japan Atomic Energy Research Institute (JAERI). The evaluation method was verified through the analyses of critical assembly experiments. A good agreement was obtained between calculations and measurements and the difference was less than 3 % on the power distribution. This method was applied to the core design for the HTTR to evaluate the maximum fuel temperature. From these results, it was confirmed that this evaluation method has an enough accuracy and is able to predict the detailed power distribution of the HTTR. (author)

  15. Design and evaluation of heat utilization systems for the HTTR through international cooperation

    Energy Technology Data Exchange (ETDEWEB)

    Lewkowicz, I. [International Atomic Energy Agency, Vienna (Austria)

    1996-07-01

    The International Atomic Energy Agency (IAEA) has the statutory function to `foster the exchange of scientific and technical information`, and `encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world`. The IAEA Co-ordinated Research Programmes (CRPs) are effective vehicles for implementing the above. The CRP on Design and Evaluation of Heat Utilization Systems for HTTR has started in September 1994 and is aimed at promoting international co-operation to identify the most promising heat utilization system(s) to be demonstrated at the HTTR, for the benefit of current operators and future designers and constructors of HTGRs. Participating Member States are collaborating by exchanging existing technical information on the technology of heat utilization systems, by developing design concepts and by performing evaluations of candidate systems for potential demonstration with the HTTR. In this report, the systems are reviewed. (J.P.N.)

  16. Development of visual inspection technology for HTTR core support graphite structure

    International Nuclear Information System (INIS)

    Maruyama, So; Iyoku, Tatsuo; Inagaki, Yoshiyuki; Shiozawa, Shusaku; Masuma, Yoshitaka; Miki, Toshiya.

    1996-01-01

    The Japan Atomic Energy Research Institute is now constructing the High Temperature Engineering Test Reactor (HTTR), which employs a visual inspection of core support graphite structure, as an inservice inspection (ISI). In this inspection, TV camera will be used to investigate the alignment and integrity of the structure. Therefore, the ISI system, a combination of radiation tolerant TV camera and graphic processing system, is developed and examined its detectability and viewing angles using a simulated hot plenum of HTTR, which has artificial defects. As a result of a series of tests, it was confirmed that this system satisfied the requirements and was quite applicable for the ISI system of HTTR core support graphite structure. In addition, further improvement of the system, like a remote control procedure, will be investigated. (author)

  17. Assessment of damage domains of the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Flores, Alain; Izquierdo, José María; Tuček, Kamil; Gallego, Eduardo

    2014-01-01

    Highlights: • We developed an adequate model for the identification of damage domains of the HTTR. • We analysed an anticipated operational transient, using the HTTR5+/GASTEMP code. • We simulated several transients of the same sequence. • We identified the corresponding damage domains using two methods. • We calculated exceedance frequency using the two methods. - Abstract: This paper presents an assessment analysis of damage domains of the 30 MW th prototype High-Temperature Engineering Test Reactor (HTTR) operated by the Japan Atomic Energy Agency (JAEA). For this purpose, an in-house deterministic risk assessment computational tool was developed based on the Theory of Stimulated Dynamics (TSD). To illustrate the methodology and applicability of the developed modelling approach, assessment results of a control rod (CR) withdrawal accident during subcritical conditions are presented and compared with those obtained by the JAEA

  18. Design and evaluation of heat utilization systems for the HTTR through international cooperation

    International Nuclear Information System (INIS)

    Lewkowicz, I.

    1996-01-01

    The International Atomic Energy Agency (IAEA) has the statutory function to 'foster the exchange of scientific and technical information', and 'encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world'. The IAEA Co-ordinated Research Programmes (CRPs) are effective vehicles for implementing the above. The CRP on Design and Evaluation of Heat Utilization Systems for HTTR has started in September 1994 and is aimed at promoting international co-operation to identify the most promising heat utilization system(s) to be demonstrated at the HTTR, for the benefit of current operators and future designers and constructors of HTGRs. Participating Member States are collaborating by exchanging existing technical information on the technology of heat utilization systems, by developing design concepts and by performing evaluations of candidate systems for potential demonstration with the HTTR. In this report, the systems are reviewed. (J.P.N.)

  19. Evaluation of hot spot factors for thermal and hydraulic design of HTTR

    International Nuclear Information System (INIS)

    Maruyama, So; Yamashita, Kiyonobu; Fujimoto, Nozomu; Murata, Isao; Sudo, Yukio; Murakami, Tomoyuki; Fujii, Sadao.

    1993-01-01

    High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950degC in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly. (author)

  20. Welding hazards

    International Nuclear Information System (INIS)

    Khan, M.A.

    1992-01-01

    Welding technology is advancing rapidly in the developed countries and has converted into a science. Welding involving the use of electricity include resistance welding. Welding shops are opened in residential area, which was causing safety hazards, particularly the teenagers and children who eagerly see the welding arc with their naked eyes. There are radiation hazards from ultra violet rays which irritate the skin, eye irritation. Welding arc light of such intensity could damage the eyes. (Orig./A.B.)

  1. Performance tests of the reactor containment structures of HTTR

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Iigaki, Kazuhiko; Kawaji, Satoshi; Iyoku, Tatsuo

    1998-03-01

    The containment structures of the HTTR consist of the reactor containment vessel (CV), service area (SA) and emergency air purification system, which minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. The CV is designed to withstand the temperature and pressure transients and to be leak-tight within the specified leakage limit even in the case of a rupture of the primary concentric hot gas duct. The pressure of inside of the SA should be maintained slightly lower than that of atmosphere by the emergency air purification system. The radioactive materials are released from the stack to environment via the emergency air purification system under the accident condition. Then the emergency air purification system should remove airborne radio-activities and should maintain proper pressure in the SA. We established the method to measure leak rate of the CV with closed reactor coolant pressure boundary although it is normally measured under opened reactor coolant pressure boundary as employed in LWRs. The CV leak rate test was carried out by the newly developed method and the expected performance was obtained. The SA and emergency air purification system were also confirmed by the performance test. We concluded that the reactor containment structures were fabricated to minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. (author)

  2. Demonstration of inherent safety features of HTGRs using the HTTR

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo

    2004-01-01

    Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) quantitatively as well as providing the core and plant transient data for validation of HTGR analysis codes for safety evaluation. The safety demonstration test are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase test simulating reactivity insertion events and coolant flow reduction events stared in FY 2002. Post-test analyses have been conducted to reproduced the test results by using the core and plant dynamics analysis code, ACCORD and Monte Carlo code, MVP. The analysis results agreed fairly well with the test results of a control rod withdrawal test simulating reactivity insertion, and gas circulators trip test simulating coolant flow reduction, at power levels of 50% and 30% of the rated power, respectively. It is shown that improvement of the ACCORD code by taking into consideration vertical and horizontal temperature distribution gives better analysis results in the control rod withdrawal test. The fist phase safety demonstration tests will continue until FY 2005, and the second phase tests are planned to be started in FY 2006. (author)

  3. Simplified two and three dimensional HTTR benchmark problems

    International Nuclear Information System (INIS)

    Zhang Zhan; Rahnema, Farzad; Zhang Dingkang; Pounders, Justin M.; Ougouag, Abderrafi M.

    2011-01-01

    To assess the accuracy of diffusion or transport methods for reactor calculations, it is desirable to create heterogeneous benchmark problems that are typical of whole core configurations. In this paper we have created two and three dimensional numerical benchmark problems typical of high temperature gas cooled prismatic cores. Additionally, a single cell and single block benchmark problems are also included. These problems were derived from the HTTR start-up experiment. Since the primary utility of the benchmark problems is in code-to-code verification, minor details regarding geometry and material specification of the original experiment have been simplified while retaining the heterogeneity and the major physics properties of the core from a neutronics viewpoint. A six-group material (macroscopic) cross section library has been generated for the benchmark problems using the lattice depletion code HELIOS. Using this library, Monte Carlo solutions are presented for three configurations (all-rods-in, partially-controlled and all-rods-out) for both the 2D and 3D problems. These solutions include the core eigenvalues, the block (assembly) averaged fission densities, local peaking factors, the absorption densities in the burnable poison and control rods, and pin fission density distribution for selected blocks. Also included are the solutions for the single cell and single block problems.

  4. Helium leak and chemical impurities control technology in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Shimizu, Atsushi; Hamamoto, Shimpei; Sakaba, Nariaki

    2014-01-01

    Japan Atomic Energy Agency (JAEA) has designed and developed high-temperature gas-cooled reactor (HTGR) hydrogen cogeneration system named gas turbine high-temperature reactor (GTHTR300C) as a commercial HTGR. Helium gas is used as the primary coolant in HTGR. Helium gas is easy to leak, and the primary helium leakage should be controlled tightly from the viewpoint of preventing the release of radioactive materials to the environment. Moreover from the viewpoint of preventing the oxidization of graphite and metallic material, the helium coolant chemistry should be controlled tightly. The primary helium leakage and the helium coolant chemistry during the operation is the major factor in the HTGR for commercialization of HTGR system. This paper shows the design concept and the obtained operational experience on the primary helium leakage control and primary helium impurity control in the high-temperature engineering test reactor (HTTR) of JAEA. Moreover, the future plan to obtain operational experience of these controls for commercialization of HTGR system is shown. (author)

  5. Plant data evaluation of performance confirmation test in HTTR after Tohoku-Pacific Ocean Earthquake

    International Nuclear Information System (INIS)

    Ono, Masato; Tochio, Daisuke; Shinohara, Masanori; Shimazaki, Yosuke; Yanagi, Shunki; Iigaki, Kazuhiko

    2012-03-01

    Tohoku-Pacific Ocean Earthquake occurred on March 11th 2011 and the earthquake intensity of an upper 5 on the Japanese scale was observed in Oarai town. HTTR conducted the confirmation test on cold state in order to ensure the facilities/instruments of reactor building operate normally. In this test, the plant data in the facilities/instruments start-up phase and continue steady operation phase were measured and compared with the previous operation data, and the soundness of facilities/instruments is evaluated. As a result, in after the earthquake, the facilities/instruments operate normally and the reactor cooling function of the HTTR were ensured. (author)

  6. Plant data evaluation of performance confirmation test in HTTR after Tohoku-Pacific Ocean Earthquake

    International Nuclear Information System (INIS)

    Ono, Masato; Tochio, Daisuke; Shinohara, Masanori; Shimazaki, Yosuke; Yanagi, Shunki; Iigaki, Kazuhiko

    2012-01-01

    Tohoku-Pacific Ocean Earthquake occurred on March 11th 2011 and the earthquake intensity of an upper 5 on the Japanese scale was observed in Oarai town. HTTR conducted the confirmation test on cold state in order to ensure the facilities/instruments of reactor building operate normally. In this test, the plant data in the facilities/instruments start-up phase and continue steady operation phase were measured and compared with the previous operation data, and the soundness of facilities/instruments is evaluated. As a result, in after the earthquake, the facilities/instruments operate normally and the reactor cooling function of the HTTR were ensured. (author)

  7. Present status and prospects of high-temperature engineering test reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1995-01-01

    It is essentially important in Japan, which has limited amount of natural resources, to make efforts to obtain more reliable and stable energy supply by extended use of nuclear energy including high temperature heat from nuclear reactors. Hence, efforts are to be continuously devoted to establish and upgrade High Temperature Gas-cooled Reactor (HTGR) technologies and to make much of research resources accumulated so far. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then, the construction of High Temperature engineering Test Reactor (HTTR), which is an HTGR with a maximum helium coolant temperature of 950degC at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and is now under way by the Japan Atomic Energy Research Institute (JAERI). The construction of the HTTR started in March 1991, with first criticality in 1998 to be followed after commissioning testing. At present the HTTR reactor building and its containment vessel have been nearly completed and its main components, such as a reactor pressure vessel, an intermediate heat exchanger, hot gas pipings and core support structures, have been manufactured at their factories and delivered to the Oarai Research Establishment of the JAERI for their installation in the middle of 1994. Fuel fabrication will be started as well. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. The IAEA Coordinated Research Programme on Design and Evaluation of Heat Utilization Systems for the HTTR, such as steam reforming of methane and thermochemical water splitting for hydrogen production, was launched successfully in January 1994. Some heat utilization system is planned to be connected to the HTTR and demonstrated at the former stage of the second core. At present, steam-reforming of methane is the first candidate. The JAERI also plans to conduct material

  8. Application of local vacuum slide sealing electron beam welding procedure

    International Nuclear Information System (INIS)

    Sato, Shozo; Takano, Genta; Minami, Masaharu; Enami, Koji; Uchikawa, Takashi; Kuri, Shuhei

    1982-01-01

    Electron beam welding process is efficient and is superior in workmanship and its application to the welding of large plate structures is eagerly awaited. However, since electron beam welding is generally performed with the object of welding entirely put in a vacuum chamber, high welding cost becomes a problem. In response to this demand, two kinds of local vacuum slide sealing type electron beam welding machines have been developed. These welding machines are designed to perform welding with only the neighborhood of the weld line put in vacuum, one of which is for longitudinal joints and the other for circumferential joints. The welding machine for circumferential joints has been put to practical use for the welding of nucear fusion reactor vacuum vessels (outside diameter 3.5 m, inside diameter 1.7 m), showing that it is applicable to the welding of large structures. (author)

  9. Welding Curriculum.

    Science.gov (United States)

    Alaska State Dept. of Education, Juneau. Div. of Adult and Vocational Education.

    This competency-based curriculum guide is a handbook for the development of welding trade programs. Based on a survey of Alaskan welding employers, it includes all competencies a student should acquire in such a welding program. The handbook stresses the importance of understanding the principles associated with the various elements of welding.…

  10. State of Art Report for the OECD-NEA Loss-of-Forced Cooling (LOFC) Test Project using HTTR Reactor

    International Nuclear Information System (INIS)

    Jun, Ji Su

    2011-05-01

    The OECD/NEA Project is planned to perform the LOFC (Loss Of Forced Cooling) test using the HTTR (High Temperature engineering Test Reactor) in Japan from 31 March 2011 to 31 March 2013 in order to obtain the data for the code validation of the VHTR safety analysis. Based on the Project Agreement Document, this report gives a description of the HTTR-LOFC test, HTTR test facility, project schedule and deliverable items as the technical state art of the project, and appends the full translation of the project agreement articles on the project management

  11. Study on system layout and component design in the HTTR hydrogen production system. Contract research

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo; Shimizu, Akira; Uchida, Shoji

    2003-01-01

    The global warming becomes a significant issue in the world so that it needs to reduce the CO 2 emission. It is expected that hydrogen is in place of the fossil fuels such as coal and oil, and plays the important role to resolve the global warming. There are several hydrogen making processes such as water electrolysis and steam reforming of hydrocarbon. Steam reforming of hydrocarbon is a major hydrogen making process because of economy in industry. It utilizes the fossil fuels as process heat for chemical reaction and results in a large CO 2 emission. New steam reforming system without fossil fuel can contribute to resolve the global warming. High temperature gas-cooled reactor (HTGR) has a unique feature to be able to supply a hot helium gas whose temperature is approximately 950degC at the reactor outlet. This makes HTGR possible to utilize for not only power generation but also process heat utilization. JAERI constructed the high temperature engineering test reactor (HTTR) that is a sort of HTGR in Oarai establishment and starts operation. Nuclear heat utilization is one of the R and D items of the HTTR. The steam reforming system coupling to the HTTR for hydrogen production has been designed. This report represents the system layout and design specification of key components in HTTR steam reforming system. (author)

  12. Study on system layout and component design in the HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Nishihara, Tetsuo; Shimizu, Akira [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tanihira, Masanori [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Uchida, Shoji [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

    2003-01-01

    The global warming becomes a significant issue in the world so that it needs to reduce the CO{sub 2} emission. It is expected that hydrogen is in place of the fossil fuels such as coal and oil, and plays the important role to resolve the global warming. There are several hydrogen making processes such as water electrolysis and steam reforming of hydrocarbon. Steam reforming of hydrocarbon is a major hydrogen making process because of economy in industry. It utilizes the fossil fuels as process heat for chemical reaction and results in a large CO{sub 2} emission. New steam reforming system without fossil fuel can contribute to resolve the global warming. High temperature gas-cooled reactor (HTGR) has a unique feature to be able to supply a hot helium gas whose temperature is approximately 950degC at the reactor outlet. This makes HTGR possible to utilize for not only power generation but also process heat utilization. JAERI constructed the high temperature engineering test reactor (HTTR) that is a sort of HTGR in Oarai establishment and starts operation. Nuclear heat utilization is one of the R and D items of the HTTR. The steam reforming system coupling to the HTTR for hydrogen production has been designed. This report represents the system layout and design specification of key components in HTTR steam reforming system. (author)

  13. Japanese HTTR program for demonstration of high temperature applications of nuclear energy

    International Nuclear Information System (INIS)

    Nishihara, T.; Hada, K.; Shiozawa, S.

    1997-01-01

    Construction works of the HTTR started in March 1991 in order to establish and upgrade the HTGR technology basis, to carry out innovative basic researches on high temperature engineering and to demonstrate high temperature heat utilization and application of nuclear heat. This report describes the demonstration program of high temperature heat utilization and application. (author). 2 refs, 4 figs, 3 tabs

  14. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY2003

    International Nuclear Information System (INIS)

    2005-03-01

    The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research Establishment of The Japan Atomic Energy Research Institute (JAERI) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power. Coolant of helium-gas circulates under the pressure of about 4Mpa, and the reactor inlet and outlet temperature are 395degC and 950degC (maximum), respectively coated particle fuel is used as fuel, and the HTTR core is composed of graphite prismatic blocks. The full power operation of 30MW was attained in December, 2001, and then JAERI received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2003 before the high temperature test operation of 950degC. (author)

  15. Full integrated system of real-time monitoring based on distributed architecture for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Subekti, Muhammad; Ohno, Tomio; Kudo, Kazuhiko; Takamatsu, Kuniyoshi; Nabeshima, Kunihiko

    2005-01-01

    A new monitoring system scheme based on distributed architecture for the High Temperature Engineering Test Reactor (HTTR) is proposed to assure consistency of the real-time process of expanded system. A distributed monitoring task on client PCs as an alternative architecture maximizes the throughput and capabilities of the system even if the monitoring tasks suffer a shortage of bandwidth. The prototype of the on-line monitoring system has been developed successfully and will be tested at the actual HTTR site. (author)

  16. Advanced Welding Concepts

    Science.gov (United States)

    Ding, Robert J.

    2010-01-01

    Four advanced welding techniques and their use in NASA are briefly reviewed in this poster presentation. The welding techniques reviewed are: Solid State Welding, Friction Stir Welding (FSW), Thermal Stir Welding (TSW) and Ultrasonic Stir Welding.

  17. Resistance seam welding

    International Nuclear Information System (INIS)

    Schueler, A.W.

    1977-01-01

    The advantages and disadvantages of the resistance seam welding process are presented. Types of seam welds, types of seam welding machines, seam welding power supplies, resistance seam welding parameters and seam welding characteristics of various metals

  18. Welding Technician

    Science.gov (United States)

    Smith, Ken

    2009-01-01

    About 95% of all manufactured goods in this country are welded or joined in some way. These welded products range in nature from bicycle handlebars and skyscrapers to bridges and race cars. The author discusses what students need to know about careers for welding technicians--wages, responsibilities, skills needed, career advancement…

  19. Investigation of water content in primary upper shield of high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Sumita, Junya; Sawa, Kazuhiro; Mogi, Haruyoshi; Itahashi, Shuuji; Kitami, Toshiyuki; Akutu, Youichi; Fuchita, Yasuhiro; Kawaguchi, Toru; Moriya, Masahiro

    1999-09-01

    A primary upper shield of the High Temperature Engineering Test Reactor (HTTR) is composed of concrete (grout) which is packed into iron frames. The main function of the primary upper shield is to attenuate neutron and gamma ray from the core, that leads to satisfy dose equivalent rate limit of operating floor and stand-pipe room. Water content in the concrete is one of the most important things because it strongly affects neutron-shielding ability. Then, we carried out out-of-pile experiments to investigate relationship between temperature and water content in the concrete. Based on the experimental results, a hydrolysis-diffusion model was developed to investigate water release behavior from the concrete. The model showed that water content used for shielding design in the primary upper shield of the HTTR will be maintained if temperature during operating life is under 110degC. (author)

  20. Steady state HTTR model in the RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Scari, Maria E.; Reis, Patrícia A. L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A. F., E-mail: melizabethscari@yahoo.com, E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPQ (Brazil)

    2017-07-01

    The HTTR (High Temperature Engineering Test Reactor) is a high temperature gas-cooled reactor (HTGR). It is a helium-cooled and graphite-moderated with a thermal power of 30 MW, developed and operated by the Japan Atomic Energy Research Institute (JAERI). In a previous work, a thermal model for the HTTR has been developed using the RELAP5-3D code. In the present work, additional thermal analyses have been presented. Studies about the core neutronic simulation were also initiated and are described here. The code WIMSD-5B was used to generate the cross sections and the group constants necessary to the NESTLE neutron kinetic code incorporated in the RELAP5-3D. (author)

  1. Fundamental philosophy on the safety design of the HTTR-IS hydrogen production system

    International Nuclear Information System (INIS)

    Ohashi, Kazutaka; Nishihara, Tetsuo; Kunitomi, Kazuhiko

    2007-01-01

    Japan Atomic Energy Agency (JAEA) has been conducting an R and D work on the VHTR reactor system and IS hydrogen production system to realize hydrogen production using nuclear heat. As a part of this activity, JAEA is planning to connect an IS test system to the High Temperature Engineering Test Reactor (HTTR) to demonstrate its technical feasibility. This paper proposes a fundamental philosophy on the safety design of the HTTR-IS hydrogen production system including the methodology to select postulated abnormal events and its event sequences and to define safety functions of the IS system to ensure the reactor safety. Also the measure to clarify the IS system as non-reactor system is proposed. (author)

  2. Preliminary analyses for HTTR`s start-up physics tests by Monte Carlo code MVP

    Energy Technology Data Exchange (ETDEWEB)

    Nojiri, Naoki [Science and Technology Agency, Tokyo (Japan); Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  3. Preliminary analyses for HTTR's start-up physics tests by Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  4. Resistance welding

    DEFF Research Database (Denmark)

    Bay, Niels; Zhang, Wenqi; Rasmussen, Mogens H.

    2003-01-01

    Resistance welding comprises not only the well known spot welding process but also more complex projection welding operations, where excessive plastic deformation of the weld point may occur. This enables the production of complex geometries and material combinations, which are often not possible...... to weld by traditional spot welding operations. Such joining processes are, however, not simple to develop due to the large number of parameters involved. Development has traditionally been carried out by large experimental investigations, but the development of a numerical programme system has changed...... this enabling prediction of the welding performance in details. The paper describes the programme in short and gives examples on industrial applications. Finally investigations of causes for failure in a complex industrial joint of two dissimilar metals are carried out combining numerical modelling...

  5. Friction welding method

    International Nuclear Information System (INIS)

    Ishida, Ryuichi; Hatanaka, Tatsuo.

    1969-01-01

    A friction welding method for forming a lattice-shaped base and tie plate supporter for fuel elements is disclosed in which a plate formed with a concavity along its edge is pressure welded to a rotating member such as a boss by longitudinally contacting the projecting surfaces remaining on either side of the concavity with the rotating member during the high speed rotation thereof in the presence of an inert gas. Since only the two projecting surfaces of the plate are fused by friction to the rotary member, heat expansion is absorbed by the concavity to prevent distortion; moreover, a two point contact surface assures a stable fitting and promotes the construction of a rigid lattice in which a number of the abovementioned plates are friction welded between rotating members to form any desired complex arrangement. The inert has serves to protect the material quality of the contacting surfaces from air during the welding step. The present invention thus provides a method in which even Zircaloy may be friction welded in place of casting stainless steel in the construction of supporting lattices to thereby enhance neutron economy. (K. J. Owens)

  6. Hydrogen production system coupled with high-temperature gas-cooled reactor (HTTR)

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku

    2003-01-01

    On the HTTR program, R and D on nuclear reactor technology and R and D on thermal application technology such as hydrogen production and so on, are advanced. When carrying out power generation and thermal application such as hydrogen production and so on, it is, at first, necessary to supply nuclear heat safely, stably and in low cost, JAERI carries out some R and Ds on nuclear reactor technology using HTTR. In parallel to this, JAERI also carries out R and D for jointing nuclear reactor system with thermal application systems because of no experience in the world on high temperature heat of about 1,000 centigrade supplied by nuclear reactor except power generation, and R and D on thermochemical decomposition method IS process for producing hydrogen from water without exhaust of carbon dioxide. Here were described summaries on R and D on nuclear reactor technology, R and D on jointing technology using HTTR hydrogen production system, R and D on IS process hydrogen production, and comparison hydrogen production with other processes. (G.K.)

  7. Inherent safety features of the HTTR revealed in the accident condition

    International Nuclear Information System (INIS)

    Kunitomi, K.; Shinozaki, M.; Baba, O.; Saito, S.

    1992-01-01

    The High Temperature Engineering Test Reactor (HTTR) being constructed by JAERI (Japan Atomic Energy Research Institute) is a graphite-moderated and helium-cooled reactor with an outlet gas temperature of 950degC. The inherent safety characteristics in the HTTR prevent temperature increase of reactor fuels and fission product release from the reactor core in postulated accident conditions. The reactor core can be cooled by a Vessel Cooling System (VCS) indirectly, even in the case that no forced cooling is expected during the accident such as primary pipe break. The VCS consists of independent water cooling loop and cooling panel around the reactor pressure vessel. The cooling panel whose temperature of 60-90degC cools the reactor pressure vessel by radiation and removes the decay heat from the core indirectly. Furthermore, even if failure of VCS is assumed during this accident as a severe accident, the reactor core is remained safe despite the temperature increase of biological concrete shield around the reactor pressure vessel. This paper describes the inherent safety features of the HTTR specially focused on the accident condition without forced cooling. The detailed analytical results of such an accident are described together with clarifying the role of the VCS. (author)

  8. Control rod position and temperature coefficients in HTTR power-rise tests. Interim report

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji; Saito, Kenji; Kobayashi, Shoichi; Sawahata, Hiroaki; Kokusen, Sigeru

    2001-03-01

    Power-rise tests of the High Temperature Engineering Test Reactor (HTTR) have been carried out aiming to achieve 100% power. So far, 50% of power operation and many tests have been carried out. In the HTTR, temperature change in core is so large to achieve the outlet coolant temperature of 950degC. To improve the calculation accuracy of the HTTR reactor physics characteristics, control rod positions at criticality and temperature coefficients were measured at each step to achieve 50% power level. The calculations were carried out using Monte Carlo code and diffusion theory with temperature distributions in the core obtained by reciprocal calculation of thermo-hydraulic code and diffusion theory. Control rod positions and temperature coefficients were calculated by diffusion theory and Monte Carlo method. The test results were compared to calculation results. The control rod positions at criticality showed good agreement with calculation results by Monte Carlo method with error of 50 mm. The control position at criticality at 100% was predicted around 2900mm. Temperature coefficients showed good agreement with calculation results by diffusion theory. The improvement of calculation will be carried out comparing the measured results up to 100% power level. (author)

  9. HTTR operation monitoring with neural network in 30 days operation at 850degC

    International Nuclear Information System (INIS)

    Shimizu, Atsushi; Nabeshima, Kunihiko; Nakagawa, Shigeaki

    2009-01-01

    The High temperature engineering test reactor (HTTR) executed the rated power operation for 30days of the first time (850degC in temperature of the nuclear reactor outlet coolant) until March, 27th through April, 26th, 2007. In this operation, HTTR was observed according to the operation monitoring model with the neural network, and the detection performance of neural network was verified during slight changes of reactor state at rated power. The neural network used for the operation monitoring was an auto-associative network, where 31 input 31 outputs and the hidden layers were connected with 20 units by the hierarchy of three layer structure. Back-propagation algorithm was used for study rule. The operation monitoring model in initial study was constructed by using the power up data between 30% and rated power, which were randomly studied. The adjustment study during the operation monitoring changes the internal structure of the initial study model to follow the changes of reactor status, such as the burn-up of the nuclear fuel for the rated power operation. As a monitoring result, slight changes of reactor state by the control system operation were correctly detected, and the on-line application to an early anomaly diagnosis for HTTR facilities will be expected. (author)

  10. Operation, test, research and development of the high temperature engineering test reactor (HTTR). (FY2005)

    International Nuclear Information System (INIS)

    2007-03-01

    The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power. The full power operation of 30 MW was attained in December, 2001, and then JAERI (JAEA) received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. In fiscal 2005 year, periodical inspection and overhaul of reactivity control system were conducted, and safety demonstration tests were promoted. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2005. (author)

  11. Fuel temperature analysis method for channel-blockage accident in HTTR

    International Nuclear Information System (INIS)

    Maruyama, So; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro; Hayakawa, Hitoshi

    1994-01-01

    During operation of the High Temperature Engineering Test Reactor (HTTR), coolability must be maintained without core damage under all postulated accident conditions. Channel blockage of a fuel element was selected as one of the design-basis accidents in the safety evaluation of the reactor. The maximum fuel temperature for such a scenario has been evaluated in the safety analysis and is compared to the core damage limits.For the design of the HTTR, an in-core thermal and hydraulic analysis code ppercase[flownet/trump] was developed. This code calculates fuel temperature distribution, not only for a channel blockage accident but also for transient conditions. The validation of ppercase[flownet/trump] code was made by comparison of the analytical results with the results of thermal and hydraulic tests by the Helium Engineering Demonstration Loop (HENDEL) multi-channel test rig (T 1-M ), which simulated one fuel column in the core. The analytical results agreed well with the experiments in which the HTTR operating conditions were simulated.The maximum fuel temperature during a channel blockage accident is 1653 C. Therefore, it is confirmed that the integrity of the core is maintained during a channel blockage accident. ((orig.))

  12. Development of operation and maintenance technology for HTGRs by using HTTR (High Temperature engineering Test Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Atsushi, E-mail: shimizu.atsushi35@jaea.go.jp [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Kawamoto, Taiki [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Tochio, Daisuke [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Saito, Kenji; Sawahata, Hiroaki; Honma, Fumitaka; Furusawa, Takayuki; Saikusa, Akio [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Takada, Shoji [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Shinozaki, Masayuki [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan)

    2014-05-01

    To establish the technical basis of HTGR (High Temperature Gas cooled Reactor), the long term high temperature operation using HTTR was carried out in the high temperature test operation mode during 50-day since January till March, 2010. It is necessary to establish the technical basis of operation and maintenance by demonstrating the stability of plant during long-term operation and the reliability of components and facilities special to HTGRs, in order to attain the stable supply of the high temperature heat to the planned heat utilization system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, it was confirmed that the temperatures and flow rate of primary and secondary coolant were well controlled within sufficiently small deviation against the disturbance by the atmospheric temperature variation in daily. Stability and reliability of the components and facility special to HTGRs was demonstrated through the long term high temperature operation by evaluating the heat transfer performance of high temperature components, the stability performance of pressure control to compensate helium gas leak, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the long term high temperature operation of HTTR, the technical basis for the operation and maintenance technology of HTGRs was established.

  13. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  14. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    International Nuclear Information System (INIS)

    Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

    2014-01-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

  15. Development of operation and maintenance technology for HTGRs by using HTTR (High Temperature engineering Test Reactor)

    International Nuclear Information System (INIS)

    Shimizu, Atsushi; Kawamoto, Taiki; Tochio, Daisuke; Saito, Kenji; Sawahata, Hiroaki; Honma, Fumitaka; Furusawa, Takayuki; Saikusa, Akio; Takada, Shoji; Shinozaki, Masayuki

    2014-01-01

    To establish the technical basis of HTGR (High Temperature Gas cooled Reactor), the long term high temperature operation using HTTR was carried out in the high temperature test operation mode during 50-day since January till March, 2010. It is necessary to establish the technical basis of operation and maintenance by demonstrating the stability of plant during long-term operation and the reliability of components and facilities special to HTGRs, in order to attain the stable supply of the high temperature heat to the planned heat utilization system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, it was confirmed that the temperatures and flow rate of primary and secondary coolant were well controlled within sufficiently small deviation against the disturbance by the atmospheric temperature variation in daily. Stability and reliability of the components and facility special to HTGRs was demonstrated through the long term high temperature operation by evaluating the heat transfer performance of high temperature components, the stability performance of pressure control to compensate helium gas leak, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the long term high temperature operation of HTTR, the technical basis for the operation and maintenance technology of HTGRs was established

  16. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

  17. Towards predictive control of extrusion weld seams: an integrated approach

    NARCIS (Netherlands)

    Bakker, A.J. den; Werkhoven, R.J.; Sillekens, W.H.; Katgerman, L.

    2010-01-01

    Longitudinal weld seams are an intrinsic feature in hollow extrusions produced with porthole dies. The formation of longitudinal weld seams is a solid bonding process, controlled by the local conditions in the extrusion die. Being the weakest areas within the extrusion cross section, it is desirable

  18. Soldadura (Welding). Spanish Translations for Welding.

    Science.gov (United States)

    Hohhertz, Durwin

    Thirty transparency masters with Spanish subtitles for key words are provided for a welding/general mechanical repair course. The transparency masters are on such topics as oxyacetylene welding; oxyacetylene welding equipment; welding safety; different types of welds; braze welding; cutting torches; cutting with a torch; protective equipment; arc…

  19. WELDING TORCH

    Science.gov (United States)

    Correy, T.B.

    1961-10-01

    A welding torch into which water and inert gas are piped separately for cooling and for providing a suitable gaseous atmosphere is described. A welding electrode is clamped in the torch by a removable collet sleeve and a removable collet head. Replacement of the sleeve and head with larger or smaller sleeve and head permits a larger or smaller welding electrode to be substituted on the torch. (AEC)

  20. Welding stresses

    International Nuclear Information System (INIS)

    Poirier, J.; Barbe, B.; Jolly, N.

    1976-01-01

    The aim is to show how internal stresses are generated and to fix the orders of magnitude. A realistic case, the vertical welding of thick plates free to move one against the other, is described and the deformations and stresses are analyzed. The mathematical model UEDA, which accounts for the elastic modulus, the yield strength and the expansion coefficient of the metal with temperature, is presented. The hypotheses and results given apply only to the instantaneous welding of a welded plate and to a plate welded by a moving electrode [fr

  1. Basic data for surveillance test on core support graphite structures for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Kikuchi, Takayuki; Iyoku, Tatsuo; Fujimoto, Nozomu; Ishihara, Masahiro; Sawa, Kazuhiro

    2007-02-01

    Both of the visual inspection by a TV camera and the measurement of material properties by surveillance test on core support graphite structures are planned for the High Temperature Engineering Test Reactor (HTTR) to confirm their structural integrity and characteristics. The surveillance test is aimed to investigate the change of material properties by aging effects such as fast neutron irradiation and oxidation. The obtained data will be used not only for evaluating the structural integrity of the core support graphite structures of the HTTR but also for design of advanced Very High Temperature Reactor (VHTR) discussed at generation IV international forum. This report describes the initial material properties of surveillance specimens before installation and installed position of surveillance specimens in the HTTR. (author)

  2. Temperature analysis of the control rods at the scram shutdown of the HTTR. Evaluation by using measurement data at scram test of HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Eiji; Fujimoto, Nozomu; Nakagawa, Shigeaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Matsuda, Atsuko [Toshiba Co., Tokyo (Japan)

    2003-03-01

    In the High Temperature Engineering Test Reactor (HTTR), since the primary coolant temperature become 950 degrees centigrade at the high temperature test operation, the special alloy Alloy800H is used for cladding tubes and spines of the control rods to endure the high temperature. The temperature limitation of control rod is 900 degrees centigrade according to the strength data of Alloy800H. The scram shutdown by loss of off-site electric power at the high temperature test operation was assumed as an transient of the temperature of the control rods cladding might exceed 900 degrees centigrade. In this report, the temperature of the control rods is analyzed by using the measurement data of the rise-to-power test. From the result of this analysis, it was confirmed that the control rod temperature does not exceed the limit even at the transient of the loss of off-site electric power from the high temperature test operation. (author)

  3. Prediction of Fission Product Release during the LOFC Experiments at the HTTR

    International Nuclear Information System (INIS)

    Shi, D.; Xhonneux, A.; Verfondern, K.; Ueta, S.; Allelein, H.-J.

    2014-01-01

    Demonstration tests were conducted using the High Temperature Engineering Test Reactor (HTTR) in Oarai, Japan, to confirm the safety of HTGR technologies and assure the expected physical phenomena to occur under given conditions. As part of the OECD directed LOFC (“loss of forced cooling”) project, a series of three tests at the HTTR has been planned with tripping of all gas circulators while deactivating all reactor reactivity control to disallow reactor scram due to abnormal reduction of primary coolant flow rate. The tests fall into anticipated transient without scram (ATWS) with occurrence of reactor recriticality. They serve the important purpose to provide a valuable data base for the validation of computer models regarding neutronics, heat transfer and fluid dynamics, fuel performance and fission product transport and release behavior in HTGRs. The Source Term Analysis Code System (STACY) is a new code development at the Research Center Jülich encompassing the original verified and validated computer models for simulating fission product transport and release. For verification of the modernized and extended version, it was assured that results obtained with the original tools could be reproduced. One of the new features of STACY is its ability to also treat fuel compacts of (full) cylindrical or annular shape and a complete prismatic block reactor core, respectively, supposed sufficient input data be available. The paper will describe the new STACY tool and present the results of fission product behavior in the HTTR core under the LOFC test conditions. Calculations are based on time-dependent neutronics and fluid dynamics results obtained with the Serpent and MGT models. (author)

  4. Soil-structure interaction analysis of HTTR building by a simplified model

    International Nuclear Information System (INIS)

    Yagishita, F.; Suzuki, H.; Yamagishi, Y.

    1990-01-01

    For the evaluation of the design seismic forces of the embedded High-Temperature-Testing-Reactor (HTTR) structure, a sway-rocking model considering the embedment of the structure is used. As for the composition of this model; the structure is modeled into beams with lumped masses, and the soil into the horizontal side springs and the horizontal and rotational bottom springs. At the same time, the input motion to the structure which has the form of multiple excitation is calculated based on one dimensional wave propagation theory. This paper presents the concept of this modelling and evaluated results. (author). 9 refs, 11 figs

  5. Study on practical of eddy current testing of core and core support graphite components in HTTR

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Iyoku, Tatsuo; Ooka, Norikazu; Shindo, Yoshihisa; Kawae, Hidetoshi; Hayashi, Motomitsu; Kambe, Mamoru; Takahashi, Masaaki; Ide, Akira.

    1994-01-01

    Core and core support graphite components in the HTTR (High Temperature Engineering Test Reactor) are mainly made of nuclear-grade IG-110 and PGX graphites. Nondestructive inspection with Eddy Current Testing (ECT) is planned to be applied to these components. The method of ECT has been already established for metallic components, however, cannot be applied directly to the graphite ones, because the characteristics of graphite are quite different in micro-structure from those of metals. Therefore, ECT method and condition were studied for the application of the ECT to the graphite components. This paper describes the study on practical method and conditions of ECT for above mentioned graphite structures. (author)

  6. Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

    2007-01-01

    The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

  7. MCNP qualification on the HTR critical configurations: HTTR, HTR10 and PROTEUS results

    Energy Technology Data Exchange (ETDEWEB)

    TRAKAS, Christos; STOVEN, Gilles [AREVA NP, Tour Areva, 92084 Paris La Defence Cedex (France)

    2008-07-01

    Recent critical experiments, including PROTEUS, HTTR and HTR-10 provide a reliable qualification base for HTR criticality predictions. The fuel tested in these experiments, be it hexagonal block or pebble type, is irradiated in a spectrum comparable to that of the HTR planned by AREVA NP. The neutron spectrum is comparable in all three cases; the mean C/M value for all critical cases is less than +350 pcm (JEF2.2), +250 pcm (JEFF3.1) and +60 pcm (ENDF BVI). The C/M obtained for the rods worth, the reaction rates and the isothermal coefficient are very satisfactory. (authors)

  8. Determination of hot spot factors for calculation of the maximum fuel temperatures in the core thermal and hydraulic design of HTTR

    International Nuclear Information System (INIS)

    Maruyama, Soh; Yamashita, Kiyonobu; Fujimoto, Nozomu; Murata, Isao; Shindo, Ryuichi; Sudo, Yukio

    1988-12-01

    The Japan Atomic Energy Research Institute (JAERI) has been designing the High Temperature Engineering Test Reactor (HTTR), which is 30 MW in thermal power, 950deg C in reactor outlet coolant temperature and 40 kg/cm 2 G in primary coolant pressure. This report summarizes the hot spot factors and their estimated values used in the evaluation of the maximum fuel temperature which is one of the major items in the core thermal and hydraulic design of the HTTR. The hot spot factors consist of systematic factors and random factors. They were identified and their values adopted in the thermal and hydraulic design were determined considering the features of the HTTR. (author)

  9. Design of helium-gas supplying facility of out-of-pile demonstration test for HTTR heat utilization system

    Energy Technology Data Exchange (ETDEWEB)

    Hino, Ryutaro; Fujisaki, Katsuo; Kobayashi, Toshiaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    1996-09-01

    One of the objectives of the High-Temperature Engineering Test Reactor (HTTR) is to demonstrate effectiveness of high-temperature heat utilization. Prior to connect a heat utilization system to the HTTR, a series of out-of-pile demonstration test is indispensable to improve components` performance, to demonstrate operation, control and safety technologies and to verify analysis codes for design and safety evaluation. After critical review and discussion on the out-of-pile demonstration test, a test facility have been designed. In this report, a helium-gas supplying facility simulated the HTTR system was described in detail, which supplies High-temperature helium-gas of 900degC to a steam reforming facility mocking-up the HTTR heat utilization system. Components of the Helium Engineering Demonstration Loop (HENDEL) were selected to reuse in the helium-gas supplying facility in order to decrease construction cost. Structures and specifications of new components such as a high-temperature heater and a preheater were decided after evaluation of thermal and hydraulic performance and strength. (author)

  10. Welding process

    International Nuclear Information System (INIS)

    Abdul Nassir Ibrahim; Azali Muhammad; Ab. Razak Hamzah; Abd. Aziz Mohamed; Mohamad Pauzi Ismail

    2008-01-01

    For the final chapter of this book, there is basic introduction on welding process. The good radiography must know somehow on welding process so that they can know what kind of welding that must rejected or not. All of the exposure technique that mention in earlier chapter almost applicable in this field because welding process is critical problem if there is no inspection will be done. So, for this chapter, all the discontinuity that usually appeared will be discussed and there is another discontinuity maybe not to important and do not give big impact if found it, do not described here. On top of that, the decision to accept or reject based on code, standard and specification that agreed by both to make sure that decision that agreed is corrected and more meaningful.

  11. HTTR hydrogen production system. Structure and main specifications of mock-up test facility (Contract research)

    International Nuclear Information System (INIS)

    Kato, Michio; Aita, Hideki; Inagaki, Yoshiyuki; Hayashi, Koji; Ohashi, Hirofumi; Sato, Hiroyuki; Iwatsuki, Jin; Takada, Shoji; Inaba, Yoshitomo

    2007-03-01

    The mock-up test facility was fabricated to investigate performance of the steam generator for mitigation of the temperature fluctuation of helium gas and transient behavior of the hydrogen production system for HTTR and to obtain experimental data for verification of a dynamic analysis code. The test facility has an approximate hydrogen production capacity of 120Nm 3 /h and the steam reforming process of methane; CH 4 +H 2 O=3H 2 +CO, was used for hydrogen production of the test facility. An electric heater was used as a heat source instead of the reactor in order to heat helium gas up to 880degC (4MPa) at the chemical reactor inlet which is the same temperature as the HTTR hydrogen production system. Fabrication of the test facility was completed in February in 2002, and seven cycle operations were carried out from March in 2002 to December in 2004. This report describes the structure and main specifications of the test facility. (author)

  12. Analysis of the European results on the HTTR's core physics benchmarks

    International Nuclear Information System (INIS)

    Raepsaet, X.; Damian, F.; Ohlig, U.A.; Brockmann, H.J.; Haas, J.B.M. de; Wallerboss, E.M.

    2002-01-01

    Within the frame of the European contract HTR-N1 calculations are performed on the benchmark problems of the HTTR's start-up core physics experiments initially proposed by the IAEA in a Co-ordinated Research Programme. Three European partners, the FZJ in Germany, NRG and IRI in the Netherlands, and CEA in France, have joined this work package with the aim to validate their calculational methods. Pre-test and post-test calculational results, obtained by the partners, are compared with each other and with the experiment. Parts of the discrepancies between experiment and pre-test predictions are analysed and tackled by different treatments. In the case of the Monte Carlo code TRIPOLI4, used by CEA, the discrepancy between measurement and calculation at the first criticality is reduced to Δk/k∼0.85%, when considering the revised data of the HTTR benchmark. In the case of the diffusion codes, this discrepancy is reduced to: Δk/k∼0.8% (FZJ) and 2.7 or 1.8% (CEA). (author)

  13. The HTTR project as the world leader of HTGR research and development

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku; Komori, Yoshihiro; Ogawa, Masuro

    2005-01-01

    As a next generation type nuclear system which will expand nuclear energy use area with high temperature nuclear heat utilization and improve economic competitiveness greatly, High Temperature Gas-cooled Reactor (HTGR) has become the R and D item of prime importance at home as well as abroad to establish hydrogen society to cope with global environmental problems. JAERI has conducted R and D on HTGR as the world leader such as to achieve a reactor outlet coolant temperature of 950 degC in the HTTR (High Temperature Engineering Test Reactor) in April 2004 as the world's first and also to succeed in continuous hydrogen production with a bench-scale apparatus of closed cycle iodine-sulfur (IS) process for six and half hours in August 2003 as the world's first. Overview and present status of HTTR program were presented in details with background and main R and D results as well as international trend of HTGR development and future program on pilot tests facilities for hydrogen production demonstration in Japan. (T. Tanaka)

  14. Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

    2007-03-01

    The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

  15. Study on flow-induced vibration of the fuel rod in HTTR

    International Nuclear Information System (INIS)

    Takase, Kazuyuki

    1988-03-01

    This study was performed in order to investigate flow-induced vibration characteristics of a fuel rod in HTTR (High Temperature engineering Test Reactor) from both an experiment and a numerical simulation. Two kinds of fuel rods were used in this experiment: one was a graphite rod which simulated a specification of the HTTR's fuel rod and the other was an aluminum rod whose weight was a half of the graphite one. The experiment was carried out up to Re = 31000 using air at room temperature and pressure. Air flowed downstream in an annular passage which consisted of the fuel rod and the graphite channel. Numerical simulations by fluid and frequency equations were also carried out. Numerical and experimental results were then compared. The following conclusions were drived: (1) The fuel rod amplitudes increase with the flow rate and with a decrease of the fuel rod weight. (2) The fuel rod amplitudes are obtained by δ/De = 2.22 x 10 -10 Re 1.43 , 9000 ≤ Re ≤ 31000, where δ is a vibration amplitude, De is a hydraulic diameter and Reis Reynolds number. (3) The fuel rod frequencies shift from lower natural frequency to higher as the flow rate increases. (4) The flow-induced vibration behavior of the fuel rod can simulate well by simultaneous equations which used the turbulence model for fluid and the mass model for vibration of the fuel rod. (author)

  16. Effect of Trailing Intensive Cooling on Residual Stress and Welding Distortion of Friction Stir Welded 2060 Al-Li Alloy

    Science.gov (United States)

    Ji, Shude; Yang, Zhanpeng; Wen, Quan; Yue, Yumei; Zhang, Liguo

    2018-04-01

    Trailing intensive cooling with liquid nitrogen has successfully applied to friction stir welding of 2 mm thick 2060 Al-Li alloy. Welding temperature, plastic strain, residual stress and distortion of 2060 Al-Li alloy butt-joint are compared and discussed between conventional cooling and trailing intensive cooling using experimental and numerical simulation methods. The results reveal that trailing intensive cooling is beneficial to shrink high temperature area, reduce peak temperature and decrease plastic strain during friction stir welding process. In addition, the reduction degree of plastic strain outside weld is smaller than that inside weld. Welding distortion presents an anti-saddle shape. Compared with conventional cooling, the reductions of welding distortion and longitudinal residual stresses of welding joint under intense cooling reach 47.7 % and 23.8 %, respectively.

  17. Syllabus in Trade Welding.

    Science.gov (United States)

    New York State Education Dept., Albany. Bureau of Secondary Curriculum Development.

    The syllabus outlines material for a course two academic years in length (minimum two and one-half hours daily experience) leading to entry-level occupational ability in several welding trade areas. Fourteen units covering are welding, gas welding, oxyacetylene welding, cutting, nonfusion processes, inert gas shielded-arc welding, welding cast…

  18. Case study on chemical plant accidents for flow-sheet design of the HTTR-IS system

    International Nuclear Information System (INIS)

    Homma, Hiroyuki; Sato, Hiroyuki; Kasahara, Seiji; Hara, Teruo; Kato, Ryoma; Sakaba, Nariaki; Ohashi, Hirofumi

    2007-02-01

    At the present time, we are alarmed by depletion of fossil energy and adverse effect of rapid increase in fossil fuel burning on environment such as climate changes and acid rain, because our lives depend still heavily upon fossil energy. It is thus widely recognized that hydrogen is one of important future energy carriers in which it is used without emission of carbon dioxide greenhouse gas and atmospheric pollutants and that hydrogen demand will increase greatly as fuel cells are developed and applied widely in the near future. To meet massive demand of hydrogen, hydrogen production from water utilizing nuclear, especially by thermochemical water-splitting Iodine-Sulphur (IS) process utilizing heat from High-Temperature Gas-cooled Reactors (HTGRs), offers one of the most attractive zero-emission energy strategies and the only one practical on a substantial scale. However, to establish a technology based for the HTGR hydrogen production by the IS process, we should close several technology gaps through R and D with the High-Temperature Engineering Test Reactor (HTTR), which is the only Japanese HTGR built and operated at the Oarai Research and Development Centre of Japan Atomic Energy Agency (JAEA). We have launched design studies of the IS process hydrogen production system coupled with the HTTR (HTTR-IS system) to demonstrate HTGR hydrogen production. In designing the HTTR-IS system, it is necessary to consider preventive and breakdown maintenance against accidents occurred in the IS process as a chemical plant. This report describes case study on chemical plant accidents relating to the IS process plant and shows a proposal of accident protection measures based on above case study, which is necessary for flow-sheet design of the HTTR-IS system. (author)

  19. Longitudinal wave ultrasonic inspection of austenitic weldments

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  20. Friction stir welding process to repair voids in aluminum alloys

    Science.gov (United States)

    Rosen, Charles D. (Inventor); Litwinski, Edward (Inventor); Valdez, Juan M. (Inventor)

    1999-01-01

    The present invention provides an in-process method to repair voids in an aluminum alloy, particularly a friction stir weld in an aluminum alloy. For repairing a circular void or an in-process exit hole in a weld, the method includes the steps of fabricating filler material of the same composition or compatible with the parent material into a plug form to be fitted into the void, positioning the plug in the void, and friction stir welding over and through the plug. For repairing a longitudinal void (30), the method includes machining the void area to provide a trough (34) that subsumes the void, fabricating filler metal into a strip form (36) to be fitted into the trough, positioning the strip in the trough, and rewelding the void area by traversing a friction stir welding tool longitudinally through the strip. The method is also applicable for repairing welds made by a fusing welding process or voids in aluminum alloy workpieces themselves.

  1. A study on laser welding deformation of 304 stainless steel

    International Nuclear Information System (INIS)

    Kitagawa, Akikazu; Maehara, Kenji; Takeda, Shinnosuke; Matsunawa, Akira

    2002-01-01

    In heavy industries, 304 austenitic stainless steel is the most popular material which is used for nuclear equipment, chemical vessels, vacuum vessels and so on. On the fabrication, not only a joint quality but also severe dimensional accuracy is required. To keep dimensional accuracy, considerable cost and efforts are requested, because the welding deformation of austenitic stainless steel is deeply depended on the physical properties of material itself. To decrease welding deformation, big jigs or water cooling method are commonly used which lead to the high cost. In general, the fusion welding by high energy density heat source results in less distortion. Today, laser welding technology has grown up to the stage that enables to weld thick plate with small deformation. The researches of welding deformation have been conducted intensively, but they are mainly concerned for arc welding, and studies for laser welding are very few. In this report, the authors will show the test results of deformation behavior in laser welding of 304 stainless steel. Also, they will discuss the deformation behavior comparing to that in arc welding. The main results of this study are as follows. 1. The angular distortion of laser welding can be unified by heat input parameter (Hp) which is used for arc welding deformation. 2. The angular distortion are same under the condition of Hp 3 in spite of different welding method, however under the condition of Hp>6-9 J/mm 3 the angular distortion is quite different depending on the power density of welding method. 3. Pure angular distortion seemed to complete just after welding, but following longitudinal distortion took place for long period. 4. The critical value of longitudinal distortion can be estimated from heat input parameter. The transverse deformation can be also estimated by heat input parameter. (author)

  2. Welding template

    International Nuclear Information System (INIS)

    Ben Venue, R.J. of.

    1976-01-01

    A welding template is described which is used to weld strip material into a cellular grid structure for the accommodation of fuel elements in a nuclear reactor. On a base plate the template carries a multitude of cylindrical pins whose upper half is narrower than the bottom half and only one of which is attached to the base plate. The others are arrested in a hexagonal array by oblong webs clamped together by chuck jaws which can be secured by means of screws. The parts are ground very accurately. The template according to the invention is very easy to make. (UWI) [de

  3. Analytical estimation of control rod shadowing effect for excess reactivity measurement of HTTR

    International Nuclear Information System (INIS)

    Nakano, Masaaki; Fujimoto, Nozomu; Yamashita, Kiyonobu

    1999-01-01

    The fuel addition method is generally used for the excess reactivity measurement of the initial core. The control rod shadowing effect for the excess reactivity measurement has been estimated analytically for High Temperature Engineering Test Reactor (HTTR). 3-dimensional whole core analyses were carried out. The movements of control rods in measurements were simulated in the calculation. It was made clear that the value of excess reactivity strongly depend on combinations of measuring control rods and compensating control rods. The differences in excess reactivity between combinations come from the control rod shadowing effect. The shadowing effect is reduced by the use of plural number of measuring and compensating control rods to prevent deep insertion of them into the core. The measured excess reactivity in the experiments is, however, smaller than the estimated value with shadowing effect. (author)

  4. Test program for NIS calibration to reactor thermal output in HTTR

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Shinozaki, Masayuki; Tachibana, Yukio; Kunitomi, Kazuhiko

    2000-03-01

    Rise-to-power test program for reactor thermal output measurement has been established to calibrate a neutron instrumentation system taking account of the characteristics of the High Temperature Engineering Test Reactor (HTTR). An error of reactor thermal output measurement was evaluated taking account of a configuration of instrumentation system. And the expected dispersion of measurement in the full power operation was evaluated from non-nuclear heat-up of primary coolant up to 213degC. From the evaluation, it was found that an error of reactor thermal output measurement would be less than ±2.0% at the rated power. This report presents the detailed program of rise-to-power test for reactor thermal output measurement and discusses its measurement error. (author)

  5. The approaches of safety design and safety evaluation at HTTR (High Temperature Engineering Test Reactor)

    International Nuclear Information System (INIS)

    Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Honma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu

    2006-06-01

    Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment. (author)

  6. Evaluation of heat exchange performance for the auxiliary component cooling water system cooling tower in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Kameyama, Yasuhiko; Shimizu, Atsushi; Inoi, Hiroyuki; Yamazaki, Kazunori; Shimizu, Yasunori; Aragaki, Etsushi; Ota, Yukimaru; Fujimoto, Nozomu

    2006-09-01

    The auxiliary component cooling water system (ACCWS) is one of the cooling system in High Temperature Engineering Test Reactor (HTTR). The ACCWS has main two features, many facilities cooling, and heat sink of the vessel cooling system which is one of the engineering safety features. Therefore, the ACCWS is required to satisfy the design criteria of heat removal performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the ACCWS cooling tower was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that ACCWS cooling tower has the required heat exchange performance in the design. (author)

  7. Evaluation of heat exchange performance for primary pressurized water cooler in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Nakagawa, Shigeaki

    2006-01-01

    In High Temperature Engineering Test Reactor (HTTR), the rated thermal power of 30 MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary cooling system, such as the primary pressurized water cooler (PPWC) and the intermediate heat exchanger (IHX). The heat exchangers in the primary cooling system are required the heat exchange performance to remove reactor generated heat 30 MW under the condition of reactor coolant outlet temperature 850degC/950degC. Therefore, the heat exchanges are required to satisfy the design criteria of heat exchange performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the PPWC in the main cooling system was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that PPWC has the required heat exchange performance in the design. (author)

  8. Evaluation of heat exchange performance for secondary pressurized water cooler in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Syuji; Saikusa, Akio; Oyama, Sunao; Nemoto, Takahiro; Hamamoto, Shinpei; Shinohara, Masanori; Isozaki, Minoru; Nakagawa, Shigeaki

    2006-02-01

    In High Temperature Engineering Test Reactor (HTTR), the rated thermal power of 30MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary cooling system, such as the intermediate heat exchanger (IHX) and the secondary pressurized water cooler (SPWC). The heat exchangers in the main cooling system are required the heat exchange performance to remove the reactor-generated-heat of 30MW under the condition of reactor coolant outlet temperature of 850degC/950degC. Therefore, the heat exchanges are required to satisfy the design criteria of heat exchange performance. In this report, heat exchange performance of the SPWC in the main cooling system was evaluated with the rise-to-power-up test and the in-service operation data. Moreover, evaluated value is compared with designed one, it is confirmed that the SPWC has required heat exchange performance. (author)

  9. Nondestructive control of residual stresses during welding and recharge processes

    International Nuclear Information System (INIS)

    Suarez, J.C.; Fernandez, L.M.; Cruz, C.; Merino, F.; Aragon, B.

    1993-01-01

    In this work, the stress state of material during welding and recharge processes is controlled with the help of Barkhausen effect. The changes, occurred in the longitudinal and transversal stress profile are show during deposition of welding rings. It is proved that the stress state of the base-material depends on the amount of recharge layers, deposited on it

  10. Development of in-service inspection system for core support graphite structures in the high temperature engineering test reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Hanawa, Satoshi; Kikuchi, Takayuki; Ishihara, Masahiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Visual inspection of core support graphite structures using TV camera as in-service inspection and measurement of material characteristics using surveillance test specimens are planned in the High Temperature Engineering Test Reactor (HTTR) to confirm structural integrity of the core support graphite structures. For the visual inspection, in-service inspection system developed from September 1996 to June 1998, and pre-service inspection using the system was carried out. As the result of the pre-service inspection, it was validated that high quality of visual inspection with TV camera can be carried out, and also structural integrity of the core support graphite structures at the initial stage of the HTTR operation was confirmed. (author)

  11. Radiation monitoring data on the power-up test of HTTR. Results up to 20 MW operation

    International Nuclear Information System (INIS)

    Ashikagaya, Yoshinobu; Nakazawa, Takashi; Yoshino, Toshiaki; Yasu, Katsuji

    2002-01-01

    The High Temperature Engineering Test Reactor (HTTR) have completed the Power-up test of 9 MW (the single and parallel loaded operation) in the rated operation mode. After that the Power-up test in the rated operation mode and the high-temperature test operation mode with a thermal output of 20 MW (the single and parallel loaded operation) were performed between January 16, 2001 and June 10, 2001. This report describes the radiation monitoring data carried out during the HTTR Power-up test in the rated operation mode and the high-temperature test operation mode with a thermal output of 20 MW. The followings were concluded from these radiation monitoring data. The monitoring of radioactive gaseous effluents and the radiation protection for the works will be easy to do and the exposure dose of the workers will be kept the low level. (author)

  12. A preliminary study on radiation damage effect in ceramics composite materials as innovative basic research using the HTTR

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Baba, Shinichi; Aihara, Jun; Arai, T.; Hayashi, K.; Ishino, S.

    1999-01-01

    An innovative basic research concerning with the basic science and applied technology is planned using the High Temperature Engineering Test Reactor (HTTR), which provides the advantage of not only a high temperature irradiation field above 400degC but also a large irradiation space. The first irradiation experiment is to be performed in 2001. Many research themes with a wide variety of scientific and technological interests are proposed as the innovative basic research. For the purpose of demonstration of scientific feasibility and advantages in the HTTR irradiation, several research themes have been being conducted as the preliminary studies. In this paper the outline of the innovative basic research is described, and the preliminary study on the radiation damage mechanism of ceramic composite materials is presented. (author)

  13. Out-of-pile demonstration test of HTTR hydrogen production system structure and fabrication technology of steam reformer. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Ouchi, Yoshihiro; Fujisaki, Katsuo; Kato, Michio; Uno, Hisao; Hayashi, Koji; Aita, Hideki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1999-10-01

    A hydrogen production system by steam reforming of natural gas, chemical reaction; CH{sub 4}+H{sub 2}O = 3H{sub 2}+CO, is to be the first heat utilization system of the HTTR. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test facility is presently under construction in order to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility, using an electric heater as a reactor substitute, simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 with a hydrogen production rate of 110 Nm{sup 3}/h. A steam reformer (SR) is a key component to produce hydrogen by steam reforming of natural gas. A bayonet-type catalyst tube was applied to the SR of the out-of-pile test facility in order to enhance the heat utilization rate. Also to promote heat transfer, the thickness of the catalyst tube should be decreased to 10 mm while augmenting heat transfer by fins formed on the outer surface of the catalyst tube. Therefore, the catalyst tube was designed on the basis of pressure difference between helium and process gases instead of total pressure of them. This design method was authorized for the first time in Japan. Furthermore, a function of explosion proof was applied to the SR because it contains inflammable gas and electric heater. This report describes the structure of the SR as well as the authorization both of the design method of the catalyst tube and the explosion proof function of the SR. (author)

  14. HTTR criticality calculations with SCALE6: Studies of various geometric and unit-cell options in modeling

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J. Y.; Chiang, M. H.; Sheu, R. J.; Liu, Y. W. H. [Inst. of Nuclear Engineering and Science, National Tsing Hua Univ., Hsinchu 30013, Taiwan (China)

    2012-07-01

    The fuel element of the High Temperature Engineering Test Reactor (HTTR) presents a doubly heterogeneous geometry, where tiny TRISO fuel particles dispersed in a graphite matrix form the fuel region of a cylindrical fuel rod, and a number of fuel rods together with moderator or reflector then constitute the lattice design of the core. In this study, a series of full-core HTTR criticality calculations were performed with the SCALE6 code system using various geometric and unit-cell options in order to systematically investigate their effects on neutronic analysis. Two geometric descriptions (ARRAY or HOLE) in SCALE6 can be used to construct a complicated and repeated model. The result shows that eliminating the use of HOLE in the HTTR geometric model can save the computation time by a factor of 4. Four unit-cell treatments for resonance self-shielding corrections in SCALE6 were tested to create problem-specific multigroup cross sections for the HTTR core model. Based on the same ENDF/B-VII cross-section library, their results were evaluated by comparing with continuous-energy calculations. The comparison indicates that the INFHOMMEDIUM result overestimates the system multiplication factor (k{sub eff}) by 55 mk, whereas the LATTICECELL and MULTIREGION treatments predict the k{sub eff} values with similar biases of approximately 10 mk overestimation. The DOUBLEHET result shows a more satisfactory agreement, about 4.2 mk underestimation in the k{sub eff} value. In addition, using cell-weighted cross sections instead of an explicit modeling of TRISO particles in fuel region can further reduce the computation time by a factor of 5 without sacrificing accuracy. (authors)

  15. Out-of-pile demonstration test of HTTR hydrogen production system structure and fabrication technology of steam reformer. Contract research

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Ouchi, Yoshihiro; Fujisaki, Katsuo; Kato, Michio; Uno, Hisao; Hayashi, Koji; Aita, Hideki

    1999-10-01

    A hydrogen production system by steam reforming of natural gas, chemical reaction; CH 4 +H 2 O = 3H 2 +CO, is to be the first heat utilization system of the HTTR. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test facility is presently under construction in order to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility, using an electric heater as a reactor substitute, simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 with a hydrogen production rate of 110 Nm 3 /h. A steam reformer (SR) is a key component to produce hydrogen by steam reforming of natural gas. A bayonet-type catalyst tube was applied to the SR of the out-of-pile test facility in order to enhance the heat utilization rate. Also to promote heat transfer, the thickness of the catalyst tube should be decreased to 10 mm while augmenting heat transfer by fins formed on the outer surface of the catalyst tube. Therefore, the catalyst tube was designed on the basis of pressure difference between helium and process gases instead of total pressure of them. This design method was authorized for the first time in Japan. Furthermore, a function of explosion proof was applied to the SR because it contains inflammable gas and electric heater. This report describes the structure of the SR as well as the authorization both of the design method of the catalyst tube and the explosion proof function of the SR. (author)

  16. WELDABILITY, WELDING METALLURGY, WELDING CHEMISTRY

    OpenAIRE

    Sarjito Jokosisworo

    2012-01-01

    Sambungan las merupakan bagian penting dari stuktur/bangunan yang dilas, dan kunci dari logam induk yang baik adalah kemampuan las (weld ability). Kemampuan las yang baik dan kemudahan dalam fabrikasi dari suatu logam merupakan pertimbangan dalam memilih suatu logam untuk konstruksi.

  17. PIE technology on mechanical tests for HTTR core component and structural materials developed at Research Hot Laboratory

    International Nuclear Information System (INIS)

    Kizaki, Minoru; Honda, Junichi; Usami, Kouji; Ouchi, Asao; Oeda, Etsuro; Matsumoto, Seiichiro

    2001-02-01

    The high temperature engineering test reactor (HTTR) with the target operation temperature of 950degC established the first criticality on November, 1998 based on a large amount of R and D results on fuel and materials. In such R and D works, the development of reactor materials are one of the key issues from the view point of reactor environments such as extremely high temperature, neutron irradiation and so on for the HTTR. The Research Hot Laboratory (RHL) had carried out much kind of post irradiation examinations (PIEs) on core component and pressure vessel materials for during more than a quarter century. And obtained data played an important role in development, characterization and licensing of those materials for the HTTR. This paper describes the PIE technology developed at RHL and typical results on mechanical tests such as elevated temperature tensile and creep rupture tests for Hasteloy-X, Incolloy 800H and so on, and Charpy impact, J IC fracture toughness, K Id fracture toughness and small punch tests for normalized and tempered 2 1/4Cr-1Mo steel from historical view. In addition, an electrochemical test technique established for investigating the irradiation embrittlement mechanism on 2 1/4Cr-1Mo steel is also mentioned. (author)

  18. Preliminary design of steam reformer in out-pile demonstration test facility for HTTR heat utilization system

    Energy Technology Data Exchange (ETDEWEB)

    Haga, Katsuhiro; Hino, Ryutaro; Inagaki, Yosiyuki; Hata, Kazuhiko; Aita, Hideki; Sekita, Kenji; Nishihara, Tetsuo; Sudo, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Yamada, Seiya

    1996-11-01

    One of the key objectives of HTTR is to demonstrate effectiveness of high-temperature nuclear heat utilization system. Prior to connecting a heat utilization system to HTTR, an out-pile demonstration test is indispensable for the development of experimental apparatuses, operational control and safety technology, and verification of the analysis code of safety assessment. For the first heat utilization system of HTTR, design of the hydrogen production system by steam reforming is going on. We have proposed the out-pile demonstration test plan of the heat utilization system and conducted preliminary design of the test facility. In this report, design of the steam reformer, which is the principal component of the test facility, is described. In the course of the design, two types of reformers are considered. The one reformer contains three reactor tubes and the other contains one reactor tube to reduce the construction cost of the test facility. We have selected the steam reformer operational conditions and structural specifications by analyzing the steam reforming characteristics and component structural strength for each type of reformer. (author)

  19. Structural integrity assessment of intermediate heat exchanger in the HTTR. Based on results of rise-to-power test

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Tachibana, Yukio; Nakagawa, Shigeaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-12-01

    A helium/helium intermediate heat exchanger (IHX) in the high temperature engineering test reactor (HTTR) is an essential component for demonstration of future nuclear process heat utilization of high temperature gas-cooled reactor (HTGR). The IHX with a heat capacity of 10 MW has 96 helically-coiled heat transfer tubes. Structural design for the IHX had been conducted through elastic-creep analysis of superalloy Hastelloy XR components such as heat transfer tubes and center pipe. In the HTTR rise-to-power test, it was clarified that temperature of the coolant in the IHX at the reactor scrams changes more rapidly than expected in the design. Effects of the IHX coolant temperature change, at anticipated reactor scram from the full power of 30 MW at high temperature test operation, on structural integrity of the heat transfer tubes and the lower reducer of the center pipe were investigated analytically based on the coolant temperature data obtained from the rise-to-power test. As results of the assessment, it was confirmed that cumulative principal creep strain, cumulative creep and fatigue damage factor of the IHX components during 10{sup 5} h of the HTTR lifetime should be below the allowable limits, which are established in the high-temperature structural design code for the HTGR Class 1 components. (author)

  20. High temperature continuous operation in the HTTR (HP-11). Summary of the test results in the high temperature operation mode

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Ueta, Shohei; Sumita, Junya; Goto, Minoru; Nakagawa, Shigeaki; Hamamoto, Shimpei; Tochio, Daisuke

    2010-11-01

    A high temperature (950 degrees C) continuous operation has been performed for 50 days on the HTTR from January to March in 2010, and the potential to supply stable heat of high temperature for hydrogen production for a long time was demonstrated for the first time in the world. JAEA has evaluated the experimental data obtained by this operation and past rated continuous one, and built the database necessary for commercial HTGRs. According to the results, the concentration of FP released from the fuels in the HTTR was a single through triple-digit lower than that in the foreign HTGRs. It became apparent that the fuels used in the HTTR are the best quality in the world. This successful operation could establish technological basis of HTGRs and show potential of nuclear energy as heat source for innovative thermo-chemical-based hydrogen production, emitting greenhouse gases on a 'low-carbon path' for the first time in the world. We have a plan to progress R and D for practical use of hydrogen production system with HTGRs in the future. (author)

  1. Non destructive evaluation of residual stresses in welding and hard-surfacing processes

    International Nuclear Information System (INIS)

    Suarez, J.C.; Fernandez, L.M.; Cruz, C.; Aragon, B.; Merino, F.

    1995-01-01

    In this paper transversal and longitudinal stress profiles in welding and hard-surfacing by welding processes are presented. The stresses were measured by RMS of Barkhausen signal. In this work it is shown that in each case the level of stresses is strongly dependent on the number of weld beads of surfacing layers deposited. The subsequent deposition of new weld beads or surfacing layers produces a stress-relieving effect

  2. WELDING PROCESS

    Science.gov (United States)

    Zambrow, J.; Hausner, H.

    1957-09-24

    A method of joining metal parts for the preparation of relatively long, thin fuel element cores of uranium or alloys thereof for nuclear reactors is described. The process includes the steps of cleaning the surfaces to be jointed, placing the sunfaces together, and providing between and in contact with them, a layer of a compound in finely divided form that is decomposable to metal by heat. The fuel element members are then heated at the contact zone and maintained under pressure during the heating to decompose the compound to metal and sinter the members and reduced metal together producing a weld. The preferred class of decomposable compounds are the metal hydrides such as uranium hydride, which release hydrogen thus providing a reducing atmosphere in the vicinity of the welding operation.

  3. Creep behavior evaluation of welded joint

    International Nuclear Information System (INIS)

    Susei, Shuzo; Matsui, Shigetomo; Mori, Eisuke; Shimizu, Shigeki; Satoh, Keisuke.

    1980-01-01

    In the creep design of high temperature structural elements, it is necessary to grasp the creep performance of joints as a whole, paying attention to the essential lack of uniformity between the material qualities of parent metals and welds. In this study, the factors controlling the creep performance of butt welded joints were investigated theoretically, when they were subjected to lateral tension and longitudinal tension. It was clarified that the rupture time in the case of laterally pulled joints was determined by the ratio of the creep rupture times of weld metals and parent metals, and the rupture time in the case of longitudinally pulled joints was determined by the ratio of the creep rupture times and the ratio of the creep strain rates of weld metals and parent metals. Moreover, when the joints of the former ratio less than 1 and the latter ratio larger than 1 were investigated experimentally, the rupture time in the case of laterally pulled joints was affected by the relative thickness, and when the relative thickness was large, the theoretical and the experimental values coincided, but the relative thickness was small, the theoretical values gave the evaluation on safe side as compared with the experimental values due to the effect of restricting deformation. In the case of longitudinally pulled joints, the theoretical and the experimental values coincided relatively well. The diagram of classifying the creep performance of welded joints was proposed. (Kako, I.)

  4. PHISICS/RELAP5-3D Adaptive Time-Step Method Demonstrated for the HTTR LOFC#1 Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Baker, Robin Ivey [Idaho National Lab. (INL), Idaho Falls, ID (United States); Balestra, Paolo [Univ. of Rome (Italy); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-05-01

    A collaborative effort between Japan Atomic Energy Agency (JAEA) and Idaho National Laboratory (INL) as part of the Civil Nuclear Energy Working Group is underway to model the high temperature engineering test reactor (HTTR) loss of forced cooling (LOFC) transient that was performed in December 2010. The coupled version of RELAP5-3D, a thermal fluids code, and PHISICS, a neutronics code, were used to model the transient. The focus of this report is to summarize the changes made to the PHISICS-RELAP5-3D code for implementing an adaptive time step methodology into the code for the first time, and to test it using the full HTTR PHISICS/RELAP5-3D model developed by JAEA and INL and the LOFC simulation. Various adaptive schemes are available based on flux or power convergence criteria that allow significantly larger time steps to be taken by the neutronics module. The report includes a description of the HTTR and the associated PHISICS/RELAP5-3D model test results as well as the University of Rome sub-contractor report documenting the adaptive time step theory and methodology implemented in PHISICS/RELAP5-3D. Two versions of the HTTR model were tested using 8 and 26 energy groups. It was found that most of the new adaptive methods lead to significant improvements in the LOFC simulation time required without significant accuracy penalties in the prediction of the fission power and the fuel temperature. In the best performing 8 group model scenarios, a LOFC simulation of 20 hours could be completed in real-time, or even less than real-time, compared with the previous version of the code that completed the same transient 3-8 times slower than real-time. A few of the user choice combinations between the methodologies available and the tolerance settings did however result in unacceptably high errors or insignificant gains in simulation time. The study is concluded with recommendations on which methods to use for this HTTR model. An important caveat is that these findings

  5. Residual stresses in zircaloy welds

    International Nuclear Information System (INIS)

    Santisteban, J. R.; Fernandez, L; Vizcaino, P.; Banchik, A.D.; Samper, R; Martinez, R. L; Almer, J; Motta, A.T.; Colas, K.B; Kerr, M.; Daymond, M.R

    2009-01-01

    Welds in Zirconium-based alloys are susceptible to hydrogen embrittlement, as H enters the material due to dissociation of water. The yield strain for hydride cracking has a complex dependence on H concentration, stress state and texture. The large thermal gradients produced by the applied heat; drastically changes the texture of the material in the heat affected zone, enhancing the susceptibility to delayed hydride cracking. Normally hydrides tend to form as platelets that are parallel to the normal direction, but when welding plates, hydride platelets may form on cooling with their planes parallel to the weld and through the thickness of the plates. If, in addition to this there are significant tensile stresses, the susceptibility of the heat affected zone to delayed hydride cracking will be increased. Here we have measured the macroscopic and microscopic residual stressed that appear after PLASMA welding of two 6mm thick Zircaloy-4 plates. The measurements were based on neutron and synchrotron diffraction experiments performed at the Isis Facility, UK, and at Advanced Photon Source, USA, respectively. The experiments allowed assessing the effect of a post-weld heat treatment consisting of a steady increase in temperature from room temperature to 450oC over a period of 4.5 hours; followed by cooling with an equivalent cooling rate. Peak tensile stresses of (175± 10) MPa along the longitudinal direction were found in the as-welded specimen, which were moderately reduced to (150±10) MPa after the heat-treatment. The parent material showed intergranular stresses of (56±4) MPa, which disappeared on entering the heat-affected zone. In-situ experiments during themal cyclong of the material showed that these intergranular stresses result from the anisotropy of the thermal expansion coefficient of the hexagonal crystal lattice. [es

  6. Welding processes handbook

    CERN Document Server

    Weman, Klas

    2011-01-01

    Offers an introduction to the range of available welding technologies. This title includes chapters on individual techniques that cover principles, equipment, consumables and key quality issues. It includes material on such topics as the basics of electricity in welding, arc physics, and distortion, and the weldability of particular metals.$bThe first edition of Welding processes handbook established itself as a standard introduction and guide to the main welding technologies and their applications. This new edition has been substantially revised and extended to reflect the latest developments. After an initial introduction, the book first reviews gas welding before discussing the fundamentals of arc welding, including arc physics and power sources. It then discusses the range of arc welding techniques including TIG, plasma, MIG/MAG, MMA and submerged arc welding. Further chapters cover a range of other important welding technologies such as resistance and laser welding, as well as the use of welding techniqu...

  7. Friction Stir Welding

    Science.gov (United States)

    Nunes, Arthur C., Jr.

    2008-01-01

    Friction stir welding (FSW) is a solid state welding process invented in 1991 at The Welding Institute in the United Kingdom. A weld is made in the FSW process by translating a rotating pin along a weld seam so as to stir the sides of the seam together. FSW avoids deleterious effects inherent in melting and promises to be an important welding process for any industries where welds of optimal quality are demanded. This article provides an introduction to the FSW process. The chief concern is the physical effect of the tool on the weld metal: how weld seam bonding takes place, what kind of weld structure is generated, potential problems, possible defects for example, and implications for process parameters and tool design. Weld properties are determined by structure, and the structure of friction stir welds is determined by the weld metal flow field in the vicinity of the weld tool. Metal flow in the vicinity of the weld tool is explained through a simple kinematic flow model that decomposes the flow field into three basic component flows: a uniform translation, a rotating solid cylinder, and a ring vortex encircling the tool. The flow components, superposed to construct the flow model, can be related to particular aspects of weld process parameters and tool design; they provide a bridge to an understanding of a complex-at-first-glance weld structure. Torques and forces are also discussed. Some simple mathematical models of structural aspects, torques, and forces are included.

  8. Ultrasonic testing of austenitic stainless steel welds

    International Nuclear Information System (INIS)

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

    1982-05-01

    Ultrasonic testing of austenitic stainless steel welds has been considered difficult because of the high noise level and remarkable attenuation of ultrasonic waves. To improve flaw detectability in this kind of steel, various inspection techniques have been studied. A series of tests indicated: (1) The longitudinal angle beam transducers newly developed during this study can detect 4.8 mm dia. side drilled holes in dissimilar metal welds (refraction angle: 55 0 from SUS side, 45 0 from CS side) and in cast stainless steel welds (refraction angle: 45 0 , inspection frequency: 1 MHz). (2) Cracks more than 5% t in depth in the heat affected zones of fine-grain stainless steel pipe welds can be detected by the 45 0 shear wave angle beam method (inspection frequency: 2 MHz). (3) The pattern recognition method using frequency analysis technology was presumed useful for discriminating crack signals from spurious echoes. (author)

  9. Tensile properties of four types of austenitic stainless steel welded joints

    International Nuclear Information System (INIS)

    Balladon, P.

    1990-01-01

    In the field of an LMFBR research programme on austenitic stainless steel welds in a Shared Cost Action Safety, Research Area 8, coordinated by JRC-Ispra, four cooperating laboratories (ECN, IKE/MPA, the Welding Institute and UNIREC) have been involved in the fabrication and extensive characterization of welded joints made from one plate of ICL 167 stainless steel. The materials included parent metal, four vacuum electron beam welds, one non vacuum electron beam weld, one submerged arc weld, one gas metal arc weld and one manual metal arc weld. This report summarizes the 106 tensile tests performed at room temperature and 550 0 C, including the influence of strain rate, specimen orientation and welding procedure. Main results are that electron beam welds have tensile properties close to those of parent metal with higher values of yield strength in longitudinal orientation and lower values of total elongation in transverse orientation but with a similar reduction of area, that filler metal welds own the highest values of yield strength and lowest values of ductility. Most of the welds properties are higher than the minimum specified for parent metal, except for some values of total elongation, mainly in transverse orientation. In view of using electron beam welding for production of components used in LMFBR, results obtained show that tensile properties of electron beam welds compare well to those of classical welds. (author)

  10. Electron beam welding of dissimilar metals

    International Nuclear Information System (INIS)

    Metzger, G.; Lison, R.

    1976-01-01

    Thirty-three two-memeber combinations of dissimilar metals were electron beam welded as square-groove butt joints in 0.08 and 0.12 in. sheet material. Many joints were ''braze welded'' by offsetting the electron beam about 0.02 in. from the butt joint to achieve fusion of the lower melting point metal, but no significant fusion of the other member of the pair. The welds were evaluated by visual and metallographic examination, transverse tensile tests, and bend tests. The welds Ag/Al, Ag/Ni15Cr7Fe, Cu/Ni15Cr7Fe, Cu/V, Cu20Ni/Ni15Cr7Fe, Fe18Cr8Ni/Ni, Fe18Cr8Ni/Ni15Cr7Fe, Nb/Ti, Nb/V, Ni/Ni15Cr7Fe, and Cb/V10Ti were readily welded and weld properties were excellent. Others which had only minor defects included the Ag/Cu20Ni, Ag/Ti, Ag/V, Cu/Fe18Cr8Ni, Cu/V10Ti, Cu20Ni/Fe18Cr8Ni, and Ti/Zr2Sn welds. The Cu/Ni weld had deep undercut, but was in other respects excellent. The mechanical properties of the Ag/Fe18Cr8Ni weld were poor, but the defect could probably be corrected. Difficulty with cracking was experienced with the Al/Ni and Fe18Cr8Ni/V welds, but sound welds had excellent mechanical properties. The remaining welds Al-Cu, Al/Cu20Ni, Al/Fe18Cr8Ni, Al/Ni15Cr7Fe, Cu20Ni/V, Cu20Ni/V10Ti, Cb/Zr2Sn, Ni/Ti, Ni15Cr7Fe/V, Ni15Cr7Fe/V10Ti, and Ti/V were unsuccessful, due to brittle phases, primarily at the weld metal-base metal interface. In addition to the two-member specimens, several joints were made by buttering. Longitudinal weld specimens of the three-member combination Al/Ni/Fe18Cr8Ni and the five member combination Fe18Cr8Ni/V/Cb/Ti/Zr2Sn showed good tensile strength and satisfactory elongation. 6 tables, 16 figures

  11. Ultrasonically-assisted Thermal Stir Welding System

    Science.gov (United States)

    Ding, R. Jeffrey (Inventor)

    2014-01-01

    A welding head assembly has a work piece disposed between its containment plates' opposing surfaces with the work piece being maintained in a plastic state thereof at least in a vicinity of the welding head assembly's stir rod as the rod is rotated about its longitudinal axis. The welding head assembly and the work piece experience relative movement there between in a direction perpendicular to the rod's longitudinal axis as the work piece is subjected to a compressive force applied by the containment plates. A first source coupled to the first containment plate applies a first ultrasonic wave thereto such that the first ultrasonic wave propagates parallel to the direction of relative movement. A second source coupled to the second containment plate applies a second ultrasonic wave thereto such that the second ultrasonic wave propagates parallel to the direction of relative movement.propagates parallel to the direction of relative movement.

  12. Ship construction and welding

    CERN Document Server

    Mandal, Nisith R

    2017-01-01

    This book addresses various aspects of ship construction, from ship types and construction materials, to welding technologies and accuracy control. The contents of the book are logically organized and divided into twenty-one chapters. The book covers structural arrangement with longitudinal and transverse framing systems based on the service load, and explains basic structural elements like hatch side girders, hatch end beams, stringers, etc. along with structural subassemblies like floors, bulkheads, inner bottom, decks and shells. It presents in detail double bottom construction, wing tanks & duct keels, fore & aft end structures, etc., together with necessary illustrations. The midship sections of various ship types are introduced, together with structural continuity and alignment in ship structures. With regard to construction materials, the book discusses steel, aluminum alloys and fiber reinforced composites. Various methods of steel material preparation are discussed, and plate cutting and form...

  13. Determination of the temperature coefficients and the kinetic parameters for the HTTR safety analysis

    International Nuclear Information System (INIS)

    Tokuhara, K.; Nakata, T.; Murata, I.; Yamashita, K.; Shindo, R.

    1991-01-01

    This report describes the calculational methods which were employed to determine the temperature coefficients and the kinetic parameters for the safety analysis in the HTTR (High Temperature Engineering Test Reactor). The temperature coefficients (doppler, moderator temperature) and the kinetic parameters (prompt neutron life time; l, effective delayed neutron fraction; β eff) are important for the point model core dynamic analysis and should be evaluated properly. The temperature coefficients were calculated by the whole core model. Doppler coefficient was evaluated under the conditions of all control rods withdrawn and the uniform change of fuel temperature. The minimum and the maximum value of the evaluated doppler coefficients in a burnup cycle are -4.6x10 -5 and -1.5x10 -5 ΔK/K/deg. C respectively. The moderator temperature coefficient was evaluated under the conditions of all control rods withdrawn and the uniform change of moderator temperature. The minimum and the maximum value of the evaluated moderator temperature coefficients in a burnup cycle are -17.1x10 -5 and 0.99x10 -5 ΔK/K/deg. C respectively. In spite of positive moderator temperature coefficient, the power coefficient is always negative. Therefore the HTTR possesses inherent power-suppressing feed back characteristic in all operating condition. We surveyed the effects of the Xe existence, the control rods existence, the fuel temperature and the region in which the temperature was changed on the moderator temperature coefficients. The kinetic parameters were calculated by the perturbation method with the whole core model. The minimum and the maximum value of the evaluated effective delayed neutron fraction (β eff) are 0.0047 and 0.0065 respectively. These of the evaluated prompt neutron life time (l) are 0.67 and 0.78 ms respectively. We have surveyed the effects of the fuel depletion and the core power level on these parameters, and considered these effects on the kinetic parameters. From

  14. Introduction to Welding.

    Science.gov (United States)

    Fortney, Clarence; Gregory, Mike

    This curriculum guide provides six units of instruction on basic welding. Addressed in the individual units of instruction are the following topics: employment opportunities for welders, welding safety and first aid, welding tools and equipment, basic metals and metallurgy, basic math and measuring, and procedures for applying for a welding job.…

  15. Distortion Control during Welding

    NARCIS (Netherlands)

    Akbari Pazooki, A.M.

    2014-01-01

    The local material expansion and contraction involved in welding result in permanent deformations or instability i.e., welding distortion. Considerable efforts have been made in controlling welding distortion prior to, during or after welding. Thermal Tensioning (TT) describes a group of in-situ

  16. Welding and cutting

    International Nuclear Information System (INIS)

    Drews, P.; Schulze Frielinghaus, W.

    1978-01-01

    This is a survey, with 198 literature references, of the papers published in the fields of welding and cutting within the last three years. The subjects dealt with are: weldability of the materials - Welding methods - Thermal cutting - Shaping and calculation of welded joints - Environmental protection in welding and cutting. (orig.) [de

  17. Influence of Welding Strength Matching Coefficient and Cold Stretching on Welding Residual Stress in Austenitic Stainless Steel

    Science.gov (United States)

    Lu, Yaqing; Hui, Hu; Gong, Jianguo

    2018-05-01

    Austenitic stainless steel is widely used in pressure vessels for the storage and transportation of liquid gases such as liquid nitrogen, liquid oxygen, and liquid hydrogen. Cryogenic pressure vessel manufacturing uses cold stretching technology, which relies heavily on welding joint performance, to construct lightweight and thin-walled vessels. Residual stress from welding is a primary factor in cases of austenitic stainless steel pressure vessel failure. In this paper, on the basis of Visual Environment 10.0 finite element simulation technology, the residual stress resulting from different welding strength matching coefficients (0.8, 1, 1.2, 1.4) for two S30408 plates welded with three-pass butt welds is calculated according to thermal elastoplastic theory. In addition, the stress field was calculated under a loading as high as 410 MPa and after the load was released. Path 1 was set to analyze stress along the welding line, and path 2 was set to analyze stress normal to the welding line. The welding strength matching coefficient strongly affected both the longitudinal residual stress (center of path 1) and the transverse residual stress (both ends of path 1) after the welding was completed. However, the coefficient had little effect on the longitudinal and transverse residual stress of path 2. Under the loading of 410 MPa, the longitudinal and transverse stress decreased and the stress distribution, with different welding strength matching coefficients, was less diverse. After the load was released, longitudinal and transverse stress distribution for both path 1 and path 2 decreased to a low level. Cold stretching could reduce the effect of residual stress to various degrees. Transverse strain along the stretching direction was also taken into consideration. The experimental results validated the reliability of the partial simulation.

  18. Advanced Welding Applications

    Science.gov (United States)

    Ding, Robert J.

    2010-01-01

    Some of the applications of advanced welding techniques are shown in this poster presentation. Included are brief explanations of the use on the Ares I and Ares V launch vehicle and on the Space Shuttle Launch vehicle. Also included are microstructural views from four advanced welding techniques: Variable Polarity Plasma Arc (VPPA) weld (fusion), self-reacting friction stir welding (SR-FSW), conventional FSW, and Tube Socket Weld (TSW) on aluminum.

  19. Research on fatigue behavior and residual stress of large-scale cruciform welding joint with groove

    International Nuclear Information System (INIS)

    Zhao, Xiaohui; Liu, Yu; Liu, Yong; Gao, Yuan

    2014-01-01

    Highlights: • The fatigue behavior of the large-scale cruciform welding joint with groove was studied. • The longitudinal residual stress of the large-scale cruciform welding joint was tested by contour method. • The fatigue fracture mechanism of the large-scale cruciform welding joint with groove was analyzed. - Abstract: Fatigue fracture behavior of the 30 mm thick Q460C-Z steel cruciform welded joint with groove was investigated. The fatigue test results indicated that fatigue strength of 30 mm thick Q460C-Z steel cruciform welded joint with groove can reach fatigue level of 80 MPa (FAT80). Fatigue crack source of the failure specimen initiated from weld toe. Meanwhile, the microcrack was also found in the fusion zones of the fatigue failure specimen, which was caused by weld quality and weld metal integrity resulting from the multi-pass welds. Two-dimensional map of the longitudinal residual stress of 30 mm thick Q460C-Z steel cruciform welded joint with groove was obtained by using the contour method. The stress nephogram of Two-dimensional map indicated that longitudinal residual stress in the welding center is the largest

  20. Residual stresses determination in an 8 mm Incoloy 800H weld via neutron diffraction

    International Nuclear Information System (INIS)

    Chen, Xizhang; Zhang, Shu Yan; Wang, Jingjun; Kelleher, Joe F.

    2015-01-01

    Highlights: • Stress through thickness at 5 mm from weld centerline indicates a “U” distribution. • Declining of tensile stress through thickness occurred at weld centerline. • Residual stress between layers is the lowest. - Abstract: To investigate the distribution of residual stresses, the 8 mm 800H alloy was joined by multi-layer butt TIG process. Residual stresses in the longitudinal, transverse and normal directions were measured via neutron diffraction. These residual stress measurements were taken at a series of points 2 mm below the top surface, covering the fusion zone, heat affected zone (HAZ) and base metal. In addition, two lines of longitudinal residual stress values at the weld centerline and 5 mm from weld centerline through thickness were measured. Results show that both the longitudinal and transverse stresses from the weld centerline to base metal are mainly tensile stresses. The longitudinal residual stress is the largest, with a maximum value of 330 MPa. As for the normal residual stress, the weld zone shows tensile stress, while the HAZ shows compressive stress. The middle of the thickness shows compressive residual stress along the thickness direction. The longitudinal stress at weld centerline through thickness reveals the interlayer heat treat effects leads to a declining of tensile stress. While the stress at 5 mm from weld centerline indicates a “U” distribution due to the mixed microstructure close to fusion line. With the increasing distance from weld seam, the residual stress decreases gradually

  1. Leak-tightness characteristics concerning the containment structures of the HTTR

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Iigaki, Kazuhiko; Kondo, Masaaki; Emori, Koichi

    2004-01-01

    The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents, which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot-gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radioactivity and will maintain a correct pressure in the service area. The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept in a way that the design pressure becomes well below its allowable limitation by the emergency air purification system, which filters efficiency of particle removal and iodine removal well over the limited values. The obtained data demonstrate that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities

  2. Soundness confirmation through cold test of the system equipment of HTTR

    International Nuclear Information System (INIS)

    Ono, Masato; Shinohara, Masanori; Iigaki, Kazuhiko; Tochio, Daisuke; Nakagawa, Shigeaki; Shimazaki, Yosuke

    2014-01-01

    HTTR was established at the Oarai Research and Development Center of Japan Atomic Energy Agency, for the purpose of the establishment and upgrading of high-temperature gas-cooled reactor technology infrastructure. Currently, it performs a safety demonstration test in order to demonstrate the safety inherent in high-temperature gas-cooled reactor. After the Great East Japan Earthquake, it conducted confirmation test for the purpose of soundness survey of facilities and equipment, and it confirmed that the soundness of the equipment was maintained. After two years from the confirmation test, it has not been confirmed whether the function of dynamic equipment and the soundness such as the airtightness of pipes and containers are maintained after receiving the influence of damage or deterioration caused by aftershocks generated during two years or aging. To confirm the soundness of these facilities, operation under cold state was conducted, and the obtained plant data was compared with confirmation test data to evaluate the presence of abnormality. In addition, in order to confirm through cold test the damage due to aftershocks and degradation due to aging, the plant data to compare was supposed to be the confirmation test data, and the evaluation on abnormality of the plant data of machine starting time and normal operation data was performed. (A.O.)

  3. Safety demonstration test (SR-1/S1C-1) plan of HTTR (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Safety demonstration tests in the HTTR (High Temperature Engineering Test Reactor) will be carried out in order to verify inherent safety features of the HTGR (High Temperature Gas-cooled Reactor). The first phase of the safety demonstration tests includes the reactivity insertion test by the control rod withdrawal and the coolant flow reduction test by the circulator trip. In the second phase, accident simulation tests will be conducted. By comparison of their experimental and analytical results, the prediction capability of the safety evaluation codes such as the core and the plant dynamics codes will be improved and verified, which will contribute to establish the safety design and the safety evaluation technologies of the HTGRs. The results obtained through its safety demonstration tests will be also utilised for the establishment of the safety design guideline, the safety evaluation guideline, etc. This paper describes the test program of the overall safety demonstration tests and the test method, the test conditions and the results of the pre-test analysis of the reactivity insertion test and the partial gas circulator trip test planned in March 2003. (author)

  4. Cause and countermeasure for heat up of HTTR core support plate at power rise tests

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Nozomu; Takada, Eiji; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-01-01

    HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate reached higher than expected at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in the core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses. (author)

  5. Construction of the HTTR and its testing program for advanced HTGR development

    International Nuclear Information System (INIS)

    Tanaka, T.; Baba, O.; Shiozawa, S.; Okubo, M.; Kunitomi, K.

    1996-01-01

    Concerning about global warming due to emission of greenhouse effect gas like CO 2 , it is essentially important to make efforts to obtain more reliable and stable energy supply by extended use of nuclear energy including high temperature heat from nuclear reactors, because it can supply a large amount of energy and its plants emit only little amount of CO 2 during their lifetime. Hence, efforts are to be continuously devoted to establish and upgrade technologies of High Temperature Gas-cooled Reactor (HTGR) which can supply high-temperature heat with high thermal efficiency as well as high heat-utilizing efficiency. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then, the construction of High Temperature engineering Test Reactor (HTTR), which is an HTGR with a maximum helium coolant temperature of 950 deg. C at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and is now under way by the Japan Atomic Energy Research Institute (JAERI). 2 refs, 2 figs, 1 tab., 2 photos

  6. Measures ensuring safety of the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    1998-04-01

    JAERI has conducted research and development of an HTGR type reactor since 1969 under the project of the multi-purpose high-temperate gas-cooled experimental reactor, whose design was changed to the HTTR in 1985. The reactor license was granted by the Government in 1990 and the construction started next year. Various functions and performances have been tested since 1996 and the initial criticality achieved in 1998. This document consists of six chapters, describing safety matters examined in several development phases. The first chapter deals with succession of the multi-purpose experimental reactor technology and its exchange between JAERI and domestic industries. Chapter 2 reviews new technical findings after the licensing which were reflected to the current safety assessment. These technical items are given in the table form of extensive pages. Chapter 3 and 4 describe the performance tests and the criticality access, respectively. Chapter 5 and 6 deal with the detection of fuel failures and helium gas leaks, respectively. (H.Y.)

  7. Welding Residual Stress Analysis and Fatigue Strength Assessment at Elevated Temperature for Multi-pass Dissimilar Material Weld Between Alloy 617 and P92 Steel

    Science.gov (United States)

    Lee, Juhwa; Hwang, Jeongho; Bae, Dongho

    2018-03-01

    In this paper, welding residual stress analysis and fatigue strength assessment were performed at elevated temperature for multi-pass dissimilar material weld between Alloy 617 and P92 steel, which are used in thermal power plant. Multi-pass welding between Alloy 617 and P92 steel was performed under optimized welding condition determined from repeated pre-test welding. In particular, for improving dissimilar material weld-ability, the buttering welding technique was applied on the P92 steel side before multi-pass welding. Welding residual stress distribution at the dissimilar material weld joint was numerically analyzed by using the finite element method, and compared with experimental results which were obtained by the hole-drilling method. Additionally, fatigue strength of dissimilar material weld joint was assessed at the room temperature (R.T), 300, 500, and 700 °C. In finite element analysis results, numerical peak values; longitudinal (410 MPa), transverse (345 MPa) were higher than those of experiments; longitudinal (298 MPa), transverse (245 MPa). There are quantitatively big differences between numerical and experimental results, due to some assumption about the thermal conductivity, specific heat, effects of enforced convection of the molten pool, dilution, and volume change during phase transformation caused by actual shield gas. The low fatigue limit at R.T, 300 °C, 500 °C and 700 °C was assessed to be 368, 276, 173 and 137 MPa respectively.

  8. Welding Residual Stress Analysis and Fatigue Strength Assessment at Elevated Temperature for Multi-pass Dissimilar Material Weld Between Alloy 617 and P92 Steel

    Science.gov (United States)

    Lee, Juhwa; Hwang, Jeongho; Bae, Dongho

    2018-07-01

    In this paper, welding residual stress analysis and fatigue strength assessment were performed at elevated temperature for multi-pass dissimilar material weld between Alloy 617 and P92 steel, which are used in thermal power plant. Multi-pass welding between Alloy 617 and P92 steel was performed under optimized welding condition determined from repeated pre-test welding. In particular, for improving dissimilar material weld-ability, the buttering welding technique was applied on the P92 steel side before multi-pass welding. Welding residual stress distribution at the dissimilar material weld joint was numerically analyzed by using the finite element method, and compared with experimental results which were obtained by the hole-drilling method. Additionally, fatigue strength of dissimilar material weld joint was assessed at the room temperature (R.T), 300, 500, and 700 °C. In finite element analysis results, numerical peak values; longitudinal (410 MPa), transverse (345 MPa) were higher than those of experiments; longitudinal (298 MPa), transverse (245 MPa). There are quantitatively big differences between numerical and experimental results, due to some assumption about the thermal conductivity, specific heat, effects of enforced convection of the molten pool, dilution, and volume change during phase transformation caused by actual shield gas. The low fatigue limit at R.T, 300 °C, 500 °C and 700 °C was assessed to be 368, 276, 173 and 137 MPa respectively.

  9. Selected Welding Techniques, Part 2

    National Research Council Canada - National Science Library

    1964-01-01

    Partial contents: CONVENTIONAL WELD JOINTS VERSUS BUTT JOINTS IN 1-INCH ALUMINUM PLATE, SPECIAL WELD JOINT PREPARATION, UPSET METAL EDGES FOR INCREASED WELD JOINT STRENGTH, OUT-OF-POSITION WELDING OF HEAVY GAGE...

  10. Development of In-Service Inspection system for heat transfer tubes in the primary pressurized water cooler in the HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Shinozaki, Masayuki; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Wada, Shigeyuki

    1999-08-01

    The ISI (In-Service Inspection) system has been developed so as to maintain the structural integrity of heat transfer tubes in the primary pressurized water cooler in the HTTR (High Temperature Engineering Test Reactor). This system consists of eddy current probes, ultra-sonic probes, insertion and extraction units, positioning unit and so on. Verification and performance tests of the developed ISI system were carried out using mock-up heat transfer tubes in the primary pressurized water cooler. The constitution of the system, R and D results of the inspection probes, and verification and performance test results of the ISI system for heat transfer tubes are described in this paper. (author)

  11. Structure of Ti-6Al-4V nanostructured titanium alloy joint obtained by resistance spot welding

    Energy Technology Data Exchange (ETDEWEB)

    Klimenov, V. A., E-mail: klimenov@tpu.ru [Tomsk State University of Architecture and Building, 2 Solyanaya Sq, Tomsk, 634003 (Russian Federation); National Research Tomsk Polytechnic University, 30 Lenin Av., Tomsk, 634050 (Russian Federation); Kurgan, K. A., E-mail: kirill-k2.777@mail.ru [Tomsk State University of Architecture and Building, 2 Solyanaya Sq, Tomsk, 634003 (Russian Federation); Chumaevskii, A. V., E-mail: tch7av@gmail.com [Institute of Strength Physics and Materials Science, Siberian Branch of the Russian Academy of Sciences, 2/4 Akademicheskii pr., Tomsk, 634021 (Russian Federation); Klopotov, A. A., E-mail: klopotovaa@tsuab.ru [Tomsk State University of Architecture and Building, 2 Solyanaya Sq, Tomsk, 634003 (Russian Federation); National Research Tomsk State University, 36 Lenin Ave., Tomsk, 634050 (Russian Federation); Gnyusov, S. F., E-mail: gnusov@rambler.ru [National Research Tomsk Polytechnic University, 30 Lenin Av., Tomsk, 634050 (Russian Federation)

    2016-01-15

    The structure of weld joints of the titanium alloy Ti-6Al-4V in the initial ultrafine-grained state, obtained by resistance spot welding, is studied using the optical and scanning electron microscopy method and the X-ray structure analysis. The carried out studies show the relationship of the metal structure in the weld zone with main joint zones. The structure in the core zone and the heat affected zone is represented by finely dispersed grains of needle-shaped martensite, differently oriented in these zones. The change in the microhardness in the longitudinal section of the weld joint clearly correlates with structural changes during welding.

  12. Automatic welding of fuel elements

    International Nuclear Information System (INIS)

    Briola, J.

    1958-01-01

    The welding process depends on the type of fuel element, the can material and the number of cartridges to be welded: - inert-gas welding (used for G2 and the 1. set of EL3), - inert atmosphere arc welding (used for welding uranium and zirconium), - electronic welding (used for the 2. set of EL3 and the tank of Proserpine). (author) [fr

  13. Detecting flaws in welds

    International Nuclear Information System (INIS)

    Woodacre, A.; Lawton, H.

    1979-01-01

    An apparatus and a method for detecting flaws in welds in a workpiece, the portion of the workpiece containing the weld is maintained at a constant temperature and the weld is scanned by an infra red detector. The weld is then scanned again with the workpiece in contact with a cooling probe to produce a steeper temperature gradient across the weld. Comparison of the signals produced by each scan reveals the existence of defects in the welds. The signals may be displayed on an oscilloscope and the display may be observed by a TV camera and recorded on videotape. (UK)

  14. Fusion welding process

    Science.gov (United States)

    Thomas, Kenneth C.; Jones, Eric D.; McBride, Marvin A.

    1983-01-01

    A process for the fusion welding of nickel alloy steel members wherein a ferrite containing pellet is inserted into a cavity in one member and melted by a welding torch. The resulting weld nugget, a fusion of the nickel containing alloy from the members to be welded and the pellet, has a composition which is sufficiently low in nickel content such that ferrite phases occur within the weld nugget, resulting in improved weld properties. The steel alloys encompassed also include alloys containing carbon and manganese, considered nickel equivalents.

  15. Weld controller for automated nuclear service welding

    International Nuclear Information System (INIS)

    Barfield, K.L.; Strubhar, P.M.; Green, D.I.

    1995-01-01

    B and W Nuclear Technologies (BWNT) uses many different types of weld heads for automated welding in the commercial nuclear service industry. Some weld heads are purchased as standard items, while others are custom designed and fabricated by BWNT requiring synchronized multiaxis motion control. BWNT recently completed a development program to build a common weld controller that interfaces to all types of weld heads used by BWNT. Their goal was to construct a system that had the flexibility to add different modules to increase the capability of the controller as different application needs become necessary. The benefits from having a common controller are listed. This presentation explains the weld controller system and the types of applications to which it has been applied

  16. Improvement of calculation method for temperature coefficient of HTTR by neutronics calculation code based on diffusion theory. Analysis for temperature coefficient by SRAC code system

    International Nuclear Information System (INIS)

    Goto, Minoru; Takamatsu, Kuniyoshi

    2007-03-01

    The HTTR temperature coefficients required for the core dynamics calculations had been calculated from the HTTR core calculation results by the diffusion code with which the corrections had been performed using the core calculation results by the Monte-Carlo code MVP. This calculation method for the temperature coefficients was considered to have some issues to be improved. Then, the calculation method was improved to obtain the temperature coefficients in which the corrections by the Monte-Carlo code were not required. Specifically, from the point of view of neutron spectrum calculated by lattice calculations, the lattice model was revised which had been used for the calculations of the temperature coefficients. The HTTR core calculations were performed by the diffusion code with the group constants which were generated by the lattice calculations with the improved lattice model. The core calculations and the lattice calculations were performed by the SRAC code system. The HTTR core dynamics calculation was performed with the temperature coefficient obtained from the core calculation results. In consequence, the core dynamics calculation result showed good agreement with the experimental data and the valid temperature coefficient could be calculated only by the diffusion code without the corrections by Monte-Carlo code. (author)

  17. Handbook of Plastic Welding

    DEFF Research Database (Denmark)

    Islam, Aminul

    The purpose of this document is to summarize the information about the laser welding of plastic. Laser welding is a matured process nevertheless laser welding of micro dimensional plastic parts is still a big challenge. This report collects the latest information about the laser welding of plastic...... materials and provides an extensive knowhow on the industrial plastic welding process. The objectives of the report include: - Provide the general knowhow of laser welding for the beginners - Summarize the state-of-the-art information on the laser welding of plastics - Find the technological limits in terms...... of design, materials and process - Find the best technology, process and machines adaptive to Sonion’s components - Provide the skills to Sonion’s Design Engineers for successful design of the of the plastic components suitable for the laser welding The ultimate goal of this report is to serve...

  18. Investigation of the loss of forced cooling test by using the high temperature engineering test reactor (HTTR) (Contract research)

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Inaba, Yoshitomo; Goto, Minoru; Tochio, Daisuke

    2007-09-01

    The three gas circulators trip test and the vessel cooling system stop test as the safety demonstration test by using the High Temperature engineering Test Reactor (HTTR) are under planning to demonstrate inherent safety features of High Temperature Gas-cooled Reactor. All three gas circulators to circulate the helium gas as the coolant are stopped to simulate the loss of forced cooling in the three gas circulators trip test. The stop of the vessel cooling system located outside the reactor pressure vessel to remove the residual heat of the reactor core follows the stop of all three gas circulators in the vessel cooling system stop test. The analysis of the reactor transient for such tests and abnormal events postulated during the test was performed. From the result of analysis, it was confirmed that the three gas circulators trip test and the vessel cooling system stop test can be performed within the region of the normal operation in the HTTR and the safety of the reactor facility is ensured even if the abnormal events would occur. (author)

  19. Valuation of the safety concept of the combined nuclear/chemical complex for hydrogen production with HTTR

    International Nuclear Information System (INIS)

    Verfondern, K.; Nishihara, T.

    2004-06-01

    The high-temperature engineering test reactor (HTTR) in Oarai, Japan, will be worldwide the first plant to demonstrate the production of hydrogen by applying the steam reforming process and using nuclear process heat as primary energy. Particular safety aspects for such a combined nuclear/chemical complex have to be investigated to further detail. One of these special aspects is the fire and explosion hazard associated with the presence of flammable gases including a large LNG storage tank in close vicinity to the reactor building. A special focus is laid upon the conceivable development of a detonation pressure wave and its damaging effect on the reactor building. A literature study has shown that methane is a comparatively slow reacting gas and that a methane vapor cloud in the open atmosphere or partially obstructed areas is highly unlikely to result in a detonation if inadvertently released and ignited. Various theoretical assessments and experimental studies, which have been conducted in the past and which are of significance for the HTTR-steam reforming system, include the spreading and combustion behavior of cryogenic liquids and flammable gas mixtures providing the basis of a comprehensive safety analysis of the combined nuclear/chemical facility. (orig.)

  20. Nondestructive testing: welding industry

    International Nuclear Information System (INIS)

    Raj, Baldev; Subramanian, C.V.

    1992-01-01

    This chapter highlights various conventional and advanced nondestructive testing (NDT) techniques that have been used for weld evaluation. Welding Codes and Standards of International and National organisations that have been followed in India for various weld evaluation purposes are also included. The chapter also emphasises the importance of NDT by way of a few case studies that have been carried out on important critical welded components. (author). 12 refs., 17 figs., 1 appendix

  1. Instructional Guidelines. Welding.

    Science.gov (United States)

    Fordyce, H. L.; Doshier, Dale

    Using the standards of the American Welding Society and the American Society of Mechanical Engineers, this welding instructional guidelines manual presents a course of study in accordance with the current practices in industry. Intended for use in welding programs now practiced within the Federal Prison System, the phases of the program are…

  2. Welding Course Curriculum.

    Science.gov (United States)

    Genits, Joseph C.

    This guide is intended for use in helping students gain a fundamental background on the major aspects of the welding trade. The course emphasis is on mastery of the manipulative skills necessary to develop successful welding techniques and on acquisition of an understanding of the specialized tools and equipment used in welding. The first part…

  3. Underwater welding of steel

    International Nuclear Information System (INIS)

    Ibarra, S.; Olson, D.L.

    1992-01-01

    A fundamental basis to understand the behavior of wet underwater welding of steel is introduced. Both the pyrometallurgical and physical metallurgy concepts are discussed. Modifications of welding consumables and practice are suggested. This chapter promotes further contributions of meatllurgical research to improve and promote wet underwater welding. (orig.)

  4. Welding Over Paint Primer

    National Research Council Canada - National Science Library

    Johnson, Kevin S; Liu, Stephen; Olson, David L

    1998-01-01

    .... According to the hydrogen-oxygen and }hydrogen-fluorine equilibrium considerations, an increase in the partial pressure of oxygen or fluorine could decrease the partial pressure of hydrogen within the welding arc. Consequently, a welding consumable that contains chemical ingredients of high oxygen and fluorine potential would be capable of minimizing hydrogen pick-up in the weld pool.

  5. Vibration stress relief of DH 36 rectangle welded plates

    Science.gov (United States)

    Li, Shuqi; Fang, Hongyuan

    2018-03-01

    Vibration stress relief (VSR) is widely used in reducing residual stress in welded structures. However, the effectiveness of this method is still instable in some circumstance. In this study, a covert negative treatment phenomenon was investigated, i.e. natural frequency of welded structures decreased after VSR but residual stress in one direction increased. When the alteration of natural frequency after VSR is significant, the residual stresses in both the longitudinal and transversal directions shall decrease. Otherwise, residual stresses may increase on one direction. Thus, sufficient power shall be applied to the welded structures to avoid negative results.

  6. Orbital welding technique

    International Nuclear Information System (INIS)

    Hoeschen, W.

    2003-01-01

    The TIG (Tungsten-inert gas) orbital welding technique is applied in all areas of pipe welding. The process is mainly used for austenitic and ferritic materials but also for materials like aluminium, nickel, and titanium alloys are commonly welded according to this technique. Thin-walled as well as thick-walled pipes are welded economically. The application of orbital welding is of particular interest in the area of maintenance of thick-walled pipes that is described in this article. (orig.) [de

  7. Homogeneous weldings of copper

    International Nuclear Information System (INIS)

    Campurri, C.; Lopez, M.; Fernandez, R.; Osorio, V.

    1995-01-01

    This research explored the metallurgical and mechanical properties of arc welding of copper related with influence of Argon, Helium and mixtures of them. Copper plates of 6 mm thickness were welded with different mixtures of the mentioned gases. The radiography of welded specimens with 100% He and 100% Ar does not show show any porosity. On the other hand, the copper plates welded different gas mixtures presented uniform porosity in the welded zone. The metallographies show recrystallized grain in the heat affected zone, while the welding zone showed a dendritic structure. The results of the tensile strength vary between a maximum of 227 MPa for 100% He and a minimum of 174 MOa for the mixture of 60% He and 40% Ar. For the elongation after fracture the best values, about 36%, were obtained for pure gases. As a main conclusion, we can say that arc welding of copper is possible without loosing the mechanical and metallurgical properties of base metal. 6 refs

  8. Effect of welding processes and joint configuration on the residual stresses and distortion in type 316 LN stainless steel weld joints

    International Nuclear Information System (INIS)

    Vasantharaja, P.; Vasudevan, M.; Palanichamy, P.

    2012-01-01

    Fabrication by welding introduces significant residual stresses in the welded structure/component due to non-uniform heat distribution during heating and cooling cycle. To control, reduce, or beneficially redistribute the residual stresses in weld joints, the stress distribution needs to be known. In the present study, weld joints of 16 mm thick 316LN stainless steel were made by multi-pass TIG, A-TIG welding and combination of TIG and A-TIG welding processes with various joint configurations. While V-groove edge preparation was required for making multi-pass TIG weld joint, square-edge preparation was sufficient for making A-TIG weld joint. Ultrasonic nondestructive technique based on the critically refracted longitudinal waves (LCR waves) has been used for the quantitative surface/sub-surface residual stress measurements in the weld joints. Distortion measurements were carried out before and after welding using height gauge. A-TIG weld joint was found to exhibit significant reduction in tensile residual stresses and distortion in comparison to that of other joints. (author)

  9. Analysis of the HTTR with Monte-Carlo and diffusion theory. An IRI-ECN intercomparison

    International Nuclear Information System (INIS)

    De Haas, J.B.M.; Wallerbos, E.J.M.

    2000-09-01

    In the framework of the IAEA Co-ordinated Research Program (CRP) 'Evaluation of HTGR Performance' for the start-up core physics benchmark of the High Temperature Engineering Test Reactor (HTTR) two-group cross section data for a fuel compact lattice and for a two-dimensional R-Z model have been generated for comparison purposes. For this comparison, 5.2% enriched uranium was selected. Furthermore, a simplified core configuration utilising only the selected type of fuel has been analysed with both the Monte Carlo code KENO and with the diffusion theory codes BOLD VENTURE and PANTHER. With a very detailed KENO model of this simplified core, k eff was calculated to be 1.1278±0.0005. Homogenisation of the core region was seen to increase k eff by 0.0340 which can be attributed to streaming of neutrons in the detailed model. The difference in k eff between the homogenised models of KENO and BOLD VENTURE amounts then only,Δk =0.0025. The PANTHER result for this core is k eff = 1. 1251, which is in good agreement with the KENO result. The fully loaded core configuration, with a range of enrichments, has also been analysed with both KENO and BOLD VENTURE. In this case the homogenisation was seen to increase k eff by 0.0375 (streaming effect). In BOLD VENTURE the critical state could be reached by the insertion of the control rods through adding an effective 10 B density over the insertion depth while the streaming of neutrons was accounted for by adjustment of the diffusion coefficient. The generation time and the effective fraction of delayed neutrons in the critical state have been calculated to be 1.11 ms and 0.705 %, respectively. This yields a prompt decay constant at critical of 6.9 s -1 . The analysis with PANTHER resulted in a k eff =1.1595 and a critical control rod setting of 244.5 cm compared to the detailed KENO results of: k eff = 1.1600 and 234.5 cm, again an excellent agreement. 5 refs

  10. Analysis of the HTTR with Monte-Carlo and diffusion theory. An IRI-ECN intercomparison

    Energy Technology Data Exchange (ETDEWEB)

    De Haas, J.B.M. [Nuclear Research and Consultancy Group NRG, Petten (Netherlands); Wallerbos, E.J.M. [Interfaculty Reactor Institute IRI, Delft University of Technology, Delft (Netherlands)

    2000-09-01

    In the framework of the IAEA Co-ordinated Research Program (CRP) 'Evaluation of HTGR Performance' for the start-up core physics benchmark of the High Temperature Engineering Test Reactor (HTTR) two-group cross section data for a fuel compact lattice and for a two-dimensional R-Z model have been generated for comparison purposes. For this comparison, 5.2% enriched uranium was selected. Furthermore, a simplified core configuration utilising only the selected type of fuel has been analysed with both the Monte Carlo code KENO and with the diffusion theory codes BOLD VENTURE and PANTHER. With a very detailed KENO model of this simplified core, k{sub eff} was calculated to be 1.1278{+-}0.0005. Homogenisation of the core region was seen to increase k{sub eff} by 0.0340 which can be attributed to streaming of neutrons in the detailed model. The difference in k{sub eff} between the homogenised models of KENO and BOLD VENTURE amounts then only,{delta}k =0.0025. The PANTHER result for this core is k{sub eff} = 1. 1251, which is in good agreement with the KENO result. The fully loaded core configuration, with a range of enrichments, has also been analysed with both KENO and BOLD VENTURE. In this case the homogenisation was seen to increase k{sub eff} by 0.0375 (streaming effect). In BOLD VENTURE the critical state could be reached by the insertion of the control rods through adding an effective {sup 10}B density over the insertion depth while the streaming of neutrons was accounted for by adjustment of the diffusion coefficient. The generation time and the effective fraction of delayed neutrons in the critical state have been calculated to be 1.11 ms and 0.705 %, respectively. This yields a prompt decay constant at critical of 6.9 s{sup -1}. The analysis with PANTHER resulted in a k{sub eff} =1.1595 and a critical control rod setting of 244.5 cm compared to the detailed KENO results of: k{sub eff} = 1.1600 and 234.5 cm, again an excellent agreement. 5 refs.

  11. Development of monitoring system using acoustic emission for detection of helium gas leakage for primary cooling system and flow-induced vibration for heat transfer tube of heat exchangers for the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Kunitomi, Kazuhiko; Furusawa, Takayuki; Shinozaki, Masayuki; Satoh, Yoshiyuki; Yanagibashi, Minoru

    1998-10-01

    The High Temperature Engineering Test Reactor (HTTR) uses helium gas for its primary coolant, whose leakage inside reactor containment vessel is considered in design of the HTTR. It is necessary to detect leakage of helium gas at an early stage so that total amount of the leakage should be as small as possible. On the other hand, heat transfer tubes of heat exchangers of the HTTR are designed not to vibrate at normal operation, but the flow-induced vibration is to be monitored to provide against an emergency. Thus monitoring system of acoustic emission for detection of primary coolant leakage and vibration of heat transfer tubes was developed and applied to the HTTR. Before the application to the HTTR, leakage detection test was performed using 1/4 scaled model of outer tube of primary concentric hot gas duct. Result of the test covers detectable minimum leakage rate and effect of difference in gas, pressure, shape of leakage path and distance from the leaking point. Detectable minimum leakage rate was about 5 Ncc/sec. The monitoring system is promising in leakage detection, though countermeasure to noise is to be needed after the HTTR starts operating. (author)

  12. X-Ray diffraction technique applied to study of residual stresses after welding of duplex stainless steel plates

    International Nuclear Information System (INIS)

    Monin, Vladimir Ivanovitch; Assis, Joaquim Teixeira de; Lopes, Ricardo Tadeu; Turibus, Sergio Noleto; Payao Filho, Joao C.

    2014-01-01

    Duplex stainless steel is an example of composite material with approximately equal amounts of austenite and ferrite phases. Difference of physical and mechanical properties of component is additional factor that contributes appearance of residual stresses after welding of duplex steel plates. Measurements of stress distributions in weld region were made by X-ray diffraction method both in ferrite and austenite phases. Duplex Steel plates were joined by GTAW (Gas Tungsten Arc Welding) technology. There were studied longitudinal and transverse stress components in welded butt joint, in heat affected zone (HAZ) and in points of base metal 10 mm from the weld. Residual stresses measured in duplex steel plates jointed by welding are caused by temperature gradients between weld zone and base metal and by difference of thermal expansion coefficients of ferrite and austenite phases. Proposed analytical model allows evaluating of residual stress distribution over the cross section in the weld region. (author)

  13. Evaluation of parameters effect on the maximum fuel temperature in the core thermal and hydraulic design of HTTR

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Maruyama, Soh; Sudo, Yukio; Fujii, Sadao; Niguma, Yoshinori.

    1988-10-01

    This report presents the results of quantitative evaluation on the effects of the dominant parameters on the maximum fuel temperature in the core thermal hydraulic design of the High Temperature Engineering Test Reactor(HTTR) of 30 MW in thermal power, 950 deg C in reactor outlet coolant temperature and 40 kg/cm 2 G in coolant pressure. The dominant parameters investigated are 1) Gap conductance. 2) Effect of eccertricity of fuel compacts in graphite sleeve. 3) Effect of spacer ribs on heat transfer coefficients. 4) Contact probability of fuel compact and graphite sleeve. 5) Validity of uniform radial power density in the fuel compacts. 6) Effect of impurity gas on gap conductance. 7) Effect of FP gas on gap conductance. The effects of these items on the maximum fuel temperature were quantitalively identified as hot spot factors. A probability of the appearance of the maximum fuel temperature was also evaluated in this report. (author)

  14. Outlines and verifications of the codes used in the safety analysis of High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Shiina, Yasuaki; Kunitomi, Kazuhiko; Maruyama, Soh; Fujita, Shigeki; Nakagawa, Shigeaki; Iyoku, Tatsuo; Shindoh, Masami; Sudo, Yukio; Hirano, Masashi.

    1990-03-01

    This paper presents brief description of the computer codes used in the safety analysis of High Temperature Engineering Test Reactor. The list of the codes is: 1. BLOOST-J2 2. THYDE-HTGR 3. TAC-NC 4. RATSAM6 5. COMPARE-MOD1 6. GRACE 7. OXIDE-3F 8. FLOWNET/TRUMP. Of described above, 1, 3, 4, 5, 6 and 7 were developed for the multi-hole type gas cooled reactor and improved for HTTR and 2 was originated by THYDE-codes which were developed to treat the transient thermo-hydraulics during LOCA of LWR. Each code adopted the models and properties which yield conservative analytical results. Adequacy of each code was verified by the comparison with the experimental results and/or the analytical results obtained from the other codes which were already proven. (author)

  15. Artificial neural networks for dynamic monitoring of simulated-operating parameters of high temperature gas cooled engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Seker, Serhat; Tuerkcan, Erdinc; Ayaz, Emine; Barutcu, Burak

    2003-01-01

    This paper addresses to the problem of utilisation of the artificial neural networks (ANNs) for detecting anomalies as well as physical parameters of a nuclear power plant during power operation in real time. Three different types of neural network algorithms were used namely, feed-forward neural network (back-propagation, BP) and two types of recurrent neural networks (RNN). The data used in this paper were gathered from the simulation of the power operation of the Japan's High Temperature Engineering Testing Reactor (HTTR). For the wide range of power operation, 56 signals were generated by the reactor dynamic simulation code for several hours of normal power operation at different power ramps between 30 and 100% nominal power. Paper will compare the outcomes of different neural networks and presents the neural network system and the determination of physical parameters from the simulated operating data

  16. Evaluation of aseismic integrity in HTTR core-bottom structure. Pt. 1. Aseismic test for core-bottom structure

    International Nuclear Information System (INIS)

    Iyoku, T.; Futakawa, M.; Ishihara, M.

    1994-01-01

    The aseismic tests were carried out using (1)/(5)-scale and (1)/(3)-scale models of the core-bottom structure of the HTTR to quantitatively evaluate the response of acceleration, strain, impact load etc. The following conclusions are obtained. (i) The frequency response of the keyway strain is correlative with that of the impact acceleration on the hot plenum block. (ii) It was confirmed through (1)/(5)-scale and (1)/(3)-scale model tests that the applied similarity law is valid to evaluate the seismic response characteristics of the core-bottom structure. (ii) The stress of graphite components estimated from the scale model test using S 2 -earthquake excitation was sufficiently lower than the allowable stress used as the design criterion. ((orig.))

  17. Evaluation for the models of neutron diffusion theory in terms of power density distributions of the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Shimakawa, Satoshi; Nojiri, Naoki; Fujimoto, Nozomu

    2003-10-01

    In the case of evaluations for the highest temperature of the fuels in the HTTR, it is very important to expect the power density distributions accurately; therefore, it is necessary to improve the analytical model with the neutron diffusion and the burn-up theory. The power density distributions are analyzed in terms of two models, the one mixing the fuels and the burnable poisons homogeneously and the other modeling them heterogeneously. Moreover these analytical power density distributions are compared with the ones derived from the gross gamma-ray measurements and the Monte Carlo calculational code with continuous energy. As a result the homogeneous mixed model isn't enough to expect the power density distributions of the core in the axial direction; on the other hand, the heterogeneous model improves the accuracy. (author)

  18. Eddy current testing on structures of nuclear-grade IG-110 graphite for acceptance test in HTTR

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Saikusa, Akio; Iyoku, Tatsuo

    1993-09-01

    Core and core support graphite structures in the HTTR are mainly made of IG-110 graphite which is fine-grained isotropic and nuclear-grade. Nondestructive inspection with eddy current testing is planned to be applied to these graphite structures. Eddy current testing is widely applied to metallic structures and its testing method has been already established. On the other hand, the characteristics of graphite are quite different in micro-structure from these of metals. Therefore, the eddy current testing method provided for metallic structures can not be applied directly to graphite structures. Thus the eddy current testing method and condition were established for the graphite structures made of IG-110 graphite. (author)

  19. Welding skate with computerized controls

    Science.gov (United States)

    Wall, W. A., Jr.

    1968-01-01

    New welding skate concept for automatic TIG welding of contoured or double-contoured parts combines lightweight welding apparatus with electrical circuitry which computes the desired torch angle and positions a torch and cold-wire guide angle manipulator.

  20. Challenges to Resistance Welding

    DEFF Research Database (Denmark)

    Song, Quanfeng

    This report originates from the compulsory defense during my Ph.D. study at the Technical University of Denmark. Resistance welding is an old and well-proven technology. Yet the emergence of more and more new materials, new designs, invention off new joining techniques, and more stringent...... requirement in quality have imposed challenges to the resistance welding. More some research and development have to be done to adapt the old technology to the manufacturing industry of the 21st century. In the 1st part of the report, the challenging factors to the resistance welding are reviewed. Numerical...... simulation of resistance welding has been under development for many years. Yet it is no easy to make simulation results reliable and accurate because of the complexity of resistance welding process. In the 2nd part of the report numerical modeling of resistance welding is reviewed, some critical factors...

  1. Ultrasonic Stir Welding

    Science.gov (United States)

    Nabors, Sammy

    2015-01-01

    NASA Marshall Space Flight Center (MSFC) developed Ultrasonic Stir Welding (USW) to join large pieces of very high-strength metals such as titanium and Inconel. USW, a solid-state weld process, improves current thermal stir welding processes by adding high-power ultrasonic (HPU) energy at 20 kHz frequency. The addition of ultrasonic energy significantly reduces axial, frictional, and shear forces; increases travel rates; and reduces wear on the stir rod, which results in extended stir rod life. The USW process decouples the heating, stirring, and forging elements found in the friction stir welding process allowing for independent control of each process element and, ultimately, greater process control and repeatability. Because of the independent control of USW process elements, closed-loop temperature control can be integrated into the system so that a constant weld nugget temperature can be maintained during welding.

  2. Automatization of welding

    International Nuclear Information System (INIS)

    Iwabuchi, Masashi; Tomita, Jinji; Nishihara, Katsunori.

    1978-01-01

    Automatization of welding is one of the effective measures for securing high degree of quality of nuclear power equipment, as well as for correspondence to the environment at the site of plant. As the latest ones of the automatic welders practically used for welding of nuclear power apparatuses in factories of Toshiba and IHI, those for pipes and lining tanks are described here. The pipe welder performs the battering welding on the inside of pipe end as the so-called IGSCC countermeasure and the succeeding butt welding through the same controller. The lining tank welder is able to perform simultaneous welding of two parallel weld lines on a large thin plate lining tank. Both types of the welders are demonstrating excellent performance at the shops as well as at the plant site. (author)

  3. Dual wire welding torch and method

    Science.gov (United States)

    Diez, Fernando Martinez; Stump, Kevin S.; Ludewig, Howard W.; Kilty, Alan L.; Robinson, Matthew M.; Egland, Keith M.

    2009-04-28

    A welding torch includes a nozzle with a first welding wire guide configured to orient a first welding wire in a first welding wire orientation, and a second welding wire guide configured to orient a second welding wire in a second welding wire orientation that is non-coplanar and divergent with respect to the first welding wire orientation. A method of welding includes moving a welding torch with respect to a workpiece joint to be welded. During moving the welding torch, a first welding wire is fed through a first welding wire guide defining a first welding wire orientation and a second welding wire is fed through a second welding wire guide defining a second welding wire orientation that is divergent and non-coplanar with respect to the first welding wire orientation.

  4. Electric arc welding gun

    Science.gov (United States)

    Luttrell, Edward; Turner, Paul W.

    1978-01-01

    This invention relates to improved apparatus for arc welding an interior joint formed by intersecting tubular members. As an example, the invention is well suited for applications where many similar small-diameter vertical lines are to be welded to a long horizontal header. The improved apparatus includes an arc welding gun having a specially designed welding head which is not only very compact but also produces welds that are essentially free from rolled-over solidified metal. The welding head consists of the upper end of the barrel and a reversely extending electrode holder, or tip, which defines an acute angle with the barrel. As used in the above-mentioned example, the gun is positioned to extend upwardly through the vertical member and the joint to be welded, with its welding head disposed within the horizontal header. Depending on the design of the welding head, the barrel then is either rotated or revolved about the axis of the vertical member to cause the electrode to track the joint.

  5. Electron beam welding

    International Nuclear Information System (INIS)

    Schwartz, M.M.

    1974-01-01

    Electron-beam equipment is considered along with fixed and mobile electron-beam guns, questions of weld environment, medium and nonvacuum welding, weld-joint designs, tooling, the economics of electron-beam job shops, aspects of safety, quality assurance, and repair. The application of the process in the case of individual materials is discussed, giving attention to aluminum, beryllium, copper, niobium, magnesium, molybdenum, tantalum, titanium, metal alloys, superalloys, and various types of steel. Mechanical-property test results are examined along with the areas of application of electron-beam welding

  6. Robot welding process control

    Science.gov (United States)

    Romine, Peter L.

    1991-01-01

    This final report documents the development and installation of software and hardware for Robotic Welding Process Control. Primary emphasis is on serial communications between the CYRO 750 robotic welder, Heurikon minicomputer running Hunter & Ready VRTX, and an IBM PC/AT, for offline programming and control and closed-loop welding control. The requirements for completion of the implementation of the Rocketdyne weld tracking control are discussed. The procedure for downloading programs from the Intergraph, over the network, is discussed. Conclusions are made on the results of this task, and recommendations are made for efficient implementation of communications, weld process control development, and advanced process control procedures using the Heurikon.

  7. Use of servo controlled weld head for end closure welding

    Energy Technology Data Exchange (ETDEWEB)

    Pathak, S.K.; Setty, D.S.; Rameswara Rao, A.; Hemantha Rao, G.V.S.; Jayaraj, R.N. [Nuclear Fuel Complex, Dept. of Atomic Energy, Hyderabad (India)

    2010-07-01

    In the PHWR fuel fabrication line resistance welding processes are used for joining various zirconium based alloy components to fuel tube of similar material. The quality requirement of these welding processes is very stringent and has to meet all the product requirements. At present these welding processes are being carried out by using standard resistance welding machines. In the resistance welding process in addition to current and time, force is one of the critical and important parameter, which influences the weld quality. At present advanced feed back type fast response medium frequency weld controllers are being used. This has upslope/down slope, constant and repetitive weld pattern selection features makes this critical welding process more reliable. Compared to weld controllers, squeeze force application devices are limited and normally standard high response pneumatic cylinders are used in the welding process. With this type of devices the force is constant during welding process and cannot be varied during welding process as per the material deformation characteristics. Similarly due to non-availability of feed back systems in the squeeze force application systems restricts the accuracy and quality of the welding process. In the present paper the influence of squeeze force pattern on the weld quality using advanced feed back type servo based force control system was studied. Different squeeze forces were used during pre and post weld heat periods along with constant force and compared with the weld quality. (author)

  8. Survey report on high temperature irradiation experiment programs for new ceramic materials in the HTTR (High Temperature Engineering Test Reactor). 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-02-01

    A survey research on status of research activities on new ceramic materials in Japan was carried out under contract between Japan Atomic Energy Research Institute and Atomic Energy Society of Japan. The purpose of the survey is to provide information to prioritize prospective experiments and tests in the HTTR. The HTTR as a high temperature gas cooled reactor has a unique and superior capability to irradiate large-volumed specimen at high temperature up to approximately 800degC. The survey was focused on mainly the activities of functional ceramics and heat resisting ceramics as a kind of structural ceramics. As the result, the report recommends that the irradiation experiment of functional ceramics is feasible to date. (K. Itami)

  9. Investigation on cause of outage of Wide Range Monitor (WRM) in High Temperature engineering Test Reactor (HTTR). Transport operation toward investigation for cause of outage

    International Nuclear Information System (INIS)

    Shinohara, Masanori; Sawahata, Hiroaki; Kawamoto, Taiki; Saito, Kenji; Takada, Shoji; Yoshida, Naoaki; Isozaki, Ryosuke; Katsuyama, Kozo; Motegi, Toshihiro

    2012-08-01

    An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, a post irradiation examination was planned to specify the damaged part causing the event in the WRM was also planned. For the investigation, the X-ray computed tomography scanner in Fuels Monitoring Facility (FMF). This report describes the preliminary investigation on the cause of outage of the WRM. The results of study for transportation method of the irradiated WRM from HTTR to FMF is also reported with the record to complete the transport operation. (author)

  10. Evaluation on materials performance of Hastelloy Alloy XR for HTTR uses-5 (Creep properties of base metal and weldment in air)

    International Nuclear Information System (INIS)

    Watanabe, Katsutoshi; Nakajima, Hajime; Koikegami, Hajime; Higuchi, Makoto; Nakanishi, Tsuneo; Saitoh, Teiichiro; Takatsu, Tamao.

    1994-01-01

    Creep properties of weldment made from Hastelloy Alloy XR base metals and filler metals for the High Temperature Engineering Test Reactor (HTTR) components were examined by means of creep and creep rupture tests at 900 and 950degC in air. The results obtained are as follows: creep rupture strength was nearly equal or higher than that of Hastelloy Alloy XR master curve and was much higher than design creep rupture strength [S R ]. Furthermore, creep rupture strength and ductility of the present filler metal was in the data band in comparison with those of the previous filler metals. It is concluded from these reasons that this filler metal has fully favorable properties for HTTR uses. (author)

  11. Fatigue strength of welds and welded materials of high-temperature steels resistant to pressurized hydrogen of the type 2.25% Cr/1% Mo

    International Nuclear Information System (INIS)

    Burlat, J.; Cheviet, A.; Million, A.

    1986-01-01

    The aim of the study is to examine systematically the creep strength of welded joints (base material, heat influence zone and welded seam) and of pure welding materials of the type 2 1/4-3% Cr/1% Mo. According to the AD standard rules, the rule which stipulates that the creep strength of welded seams under full stress be calculated with the strength characteristic value reduced by 20% applies to all heat-resistant steels, if no rupture stress values for the welded joints are available. Manufacturers of steel and weld fillers together with the Union of Technical Control Associations (VdTUeV) have prepared a test programme according to which on the one hand welded joints are tested at right angles to their seams, and on the other pure welding material is tested with respect to its creep strength. The development of the testes and their results have been described. The first results are available as VdTUeV material performance sheets, for 2 materials, and as provisional VdTUeV specification sheets, for 3 weld fillers. With the tested materials, it becomes practically feasible to reduce the creep strength of longitudinally welded pressure-bearing components by about 20% of wall thickness. (orig.) [de

  12. Hydrogen production by high-temperature gas-cooled reactor. Conceptual design of advanced process heat exchangers of the HTTR-IS hydrogen production system

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Ohashi, Hirofumi; Sato, Hiroyuki; Hara, Teruo; Kato, Ryoma; Kunitomi, Kazuhiko

    2008-01-01

    Nuclear hydrogen production is necessary in an anticipated hydrogen society that demands a massive quantity of hydrogen without economic disadvantage. Japan Atomic Energy Agency (JAEA) has launched the conceptual design study of a hydrogen production system with a near-term plan to connect it to Japan's first high-temperature gas-cooled reactor HTTR. The candidate hydrogen production system is based on the thermochemical water-splitting iodine sulphur (IS) process.The heat of 10 MWth at approximately 900degC, which can be provided by the secondary helium from the intermediate heat exchanger of the HTTR, is the energy input to the hydrogen production system. In this paper, we describe the recent progresses made in the conceptual design of advanced process heat exchangers of the HTTR-IS hydrogen production system. A new concept of sulphuric acid decomposer is proposed. This involves the integration of three separate functions of sulphuric acid decomposer, sulphur trioxide decomposer, and process heat exchanger. A new mixer-settler type of Bunsen reactor is also designed. This integrates three separate functions of Bunsen reactor, phase separator, and pump. The new concepts are expected to result in improved economics through construction and operation cost reductions because the number of process equipment and complicated connections between the equipment has been substantially reduced. (author)

  13. Outline of operation and control system and analytical investigation of transient behavior of an out-of-pile hydrogen production system for HTTR heat utilization system

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Hada, Kazuhiko; Nishihara, Tetsuo; Takeda, Tetsuaki; Haga, Katsuhiro; Hino, Ryutaro.

    1997-10-01

    The hydrogen production system by steam reforming of natural gas is to be constructed to demonstrate effectiveness of high-temperature nuclear heat utilization systems with the HTTR. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test system is planned to investigate the system characteristics, to develop high-temperature components such as a reformer, a high-temperature isolation valve and so on, and to verify operation and control technologies and safety technology at accidents. This paper presents outline of operation and control systems and analytical review of transient behavior of the out-of-pile hydrogen production system. Main function of the operation and control systems is made not to give disturbance to the HTTR at transient state under start-up and stop operations. The operation modes are separated into two ones, namely normal and accident operation modes, and operation sequences are made for each operation mode. The normal operation sequence includes start-up, steady operation and stop of the out-of-pile system. The accident one deals with accident conditions at which supply of feed gas is stopped and helium gas is cooled passively by the steam generator. Transient behavior of the out-of-pile system was analyzed numerically according as the operation sequences. As the results, it was confirmed that the designed operation and control systems are adequate to the out-of-pile system. (author)

  14. Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2003-03-01

    Research on a hydrogen production system by steam reforming of methane, chemical reaction; CH{sub 4} + H{sub 2}O {yields} 3H{sub 2}O + CO, has been carried out to couple with the HTTR for establishment of high-temperature nuclear heat utilization technology and contribution to hydrogen energy society in future. The mock-up test facility with a full-scale reaction tube test facility, a model simulating one reaction tube of a steam reformer of the HTTR hydrogen production system in full scale, was fabricated to perform tests on controllability, hydrogen production performance etc. under the same pressure and temperature conditions as those of the HTTR hydrogen production system. The design and fabrication of the test facility started from 1997, and the all components were installed until September in 2001. In a performance test conducted from October in 2001 to February in 2002, performance of each component was examined and hydrogen of 120m{sup 3}{sub N}/h was successfully produced with high-temperature helium gas. This report describes the performance test results on components performance, hydrogen production characteristics etc., and main troubles and countermeasures. (author)

  15. Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system. Contract research

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio

    2003-03-01

    Research on a hydrogen production system by steam reforming of methane, chemical reaction; CH 4 + H 2 O → 3H 2 O + CO, has been carried out to couple with the HTTR for establishment of high-temperature nuclear heat utilization technology and contribution to hydrogen energy society in future. The mock-up test facility with a full-scale reaction tube test facility, a model simulating one reaction tube of a steam reformer of the HTTR hydrogen production system in full scale, was fabricated to perform tests on controllability, hydrogen production performance etc. under the same pressure and temperature conditions as those of the HTTR hydrogen production system. The design and fabrication of the test facility started from 1997, and the all components were installed until September in 2001. In a performance test conducted from October in 2001 to February in 2002, performance of each component was examined and hydrogen of 120m 3 N /h was successfully produced with high-temperature helium gas. This report describes the performance test results on components performance, hydrogen production characteristics etc., and main troubles and countermeasures. (author)

  16. Comparison of welding induced residual stresses austenitic and ferritic steel weld joints

    International Nuclear Information System (INIS)

    Rajkumar, K.V.; Arun Kumar, S.; Mahadevan, S.; Manojkumar, R.; Rao, B. Purna Chandra; Albert, Shaju K.; Murugan, S.

    2015-01-01

    X-ray diffraction (XRD) is a well established technique for measurement of residual stresses in components and is being widely used. In XRD technique, the distance between the crystallographic planes (d spacing) is measured from peak position (2è) at various ø angles, where ø is the angle between the normal to the sample and the bisector of the incident and diffracted beam. From the slope of sin2ø vs. d spacing plot, the residual stresses are arrived by assuming a plane stress model. Welding induced residual stresses is of high importance as it is a major cause of failure in components. Surface compressive stresses improve the fatigue strength, whereas tensile residual stresses tend to decrease the fatigue strength. The present study compares the residual stresses that develop in 3 mm thick SS 316 and P91 TIG weld joints using the XRD technique. This study is aimed at understanding the influence of shrinkage during cooling and the effect of phase transformation induced volume changes on residual stress development in these two steels. While the first effect is predominant in the SS 316 weld, both the effects are present in the P91 welds. Stress measurements on SS 316 and P91 were carried out using Cr Kâ (λ-2.0840 Å) and Cr Ká (λ-2.2896 Å) radiations respectively. Typical 'M' type stress profile was observed across the weld centre line in both the welds. The variation and similarities between the longitudinal stress profiles observed in these two weld joints would be discussed. (author)

  17. Measuring weld heat to evaluate weld integrity

    Energy Technology Data Exchange (ETDEWEB)

    Schauder, V., E-mail: schauder@hks-prozesstechnik.de [HKS-Prozesstechnik GmbH, Halle (Germany)

    2015-11-15

    Eddy current and ultrasonic testing are suitable for tube and pipe mills and have been used for weld seam flaw detection for decades, but a new process, thermography, is an alternative. By measuring the heat signature of the weld seam as it cools, it provides information about weld integrity at and below the surface. The thermal processes used to join metals, such as plasma, induction, laser, and gas tungsten arc welding (GTAW), have improved since they were developed, and they get better with each passing year. However, no industrial process is perfect, so companies that conduct research in flaw detection likewise continue to develop and improve the technologies used to verify weld integrity: ultrasonic testing (UT), eddy current testing (ET), hydrostatic, X-ray, magnetic particle, and liquid penetrant are among the most common. Two of these are used for verifying the integrity of the continuous welds such as those used on pipe and tube mills: UT and ET. Each uses a transmitter to send waves of ultrasonic energy or electrical current through the material and a receiver (probe) to detect disturbances in the flow. The two processes often are combined to capitalize on the strengths of each. While ET is good at detecting flaws at or near the surface, UT penetrates the material, detecting subsurface flaws. One drawback is that sound waves and electrical current waves have a specific direction of travel, or an alignment. A linear defect that runs parallel to the direction of travel of the ultrasonic sound wave or a flaw that is parallel to the coil winding direction of the ET probe can go undetected. A second drawback is that they don't detect cold welds. An alternative process, thermography, works in a different fashion: It monitors the heat of the material as the weld cools. Although it measures the heat at the surface, the heat signature provides clues about cooling activity deep in the material, resulting in a thorough assessment of the weld's integrity It

  18. TIG welding method and TIG welding device

    International Nuclear Information System (INIS)

    Yoneda, Eishi

    1998-01-01

    The present invention provides a method of TIG welding for members having different heat capacities including a cladding tube and an end plug of a fuel rod to be used, for example, in a reactor, and a device therefor. Namely, in the TIG welding method, the flow rate of a sealed gas to the side of a member having smaller heat capacity is made greater than that on the side of the member having greater heat capacity bordered on the top end of a welding electrode. Since the sealed gas is jetted being localized relative to the welding electrode, arc is restricted in a region of the member having smaller heat capacity and is increased at a region having a larger heat capacity. As a result, the arc is localized, so that the heat input amount to the region having a large heat capacity is increased, and then a plurality of members at the abutting portion are melted uniformly thereby capable of obtaining a uniform molten pool. A bead is formed at the abutting portion thereby capable of obtaining a welded portion with less unevenness and having large strength. (I.S.)

  19. Explosion metal welding

    International Nuclear Information System (INIS)

    Popoff, A.A.

    1976-01-01

    Process parameters pertaining to welding similar and dissimilar metals using explosives are reviewed. The discussion centers on the interrelationship of physical parameters which play a part in achieving desirable metallurgical results. Present activities in explosion metal welding at LASL are presented and shown how they related to the interests of the ERDA community

  20. Electron beam welding

    International Nuclear Information System (INIS)

    Gabbay, M.

    1972-01-01

    The bead characteristics and the possible mechanisms of the electron beam penetration are presented. The different welding techniques are exposed and the main parts of an electron beam welding equipment are described. Some applications to nuclear, spatial and other industries are cited [fr

  1. Residual strains and microstructure development in single and sequential double sided friction stir welds in RQT-701 steel

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, S.J. [School of Materials, University of Manchester, Grosvenor Street, Manchester M1 7HS (United Kingdom)], E-mail: simon.barnes-2@manchester.ac.uk; Steuwer, A. [FaME38, ILL ESRF, 6 rue J.Horowitz, 38042 Grenoble, Cedex (France); University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom); Mahawish, S. [School of Materials, University of Manchester, Grosvenor Street, Manchester M1 7HS (United Kingdom); Johnson, R. [TWI Yorkshire, Wallis Way, Catcliffe, Rotherham S60 5TZ (United Kingdom); Withers, P.J. [School of Materials, University of Manchester, Grosvenor Street, Manchester M1 7HS (United Kingdom)

    2008-09-25

    Single and double sided partial penetration friction stir butt welds, in a rolled, quenched and tempered steel (RQT-701), were produced at The Welding Institute (TWI) under controlled process conditions. The residual strain distributions in the longitudinal and transverse directions have been measured using energy dispersive synchrotron X-ray diffraction. The measured strains were indicative of longitudinal tensile residual stresses at levels greater than the 0.2% yield stress of the parent metal in both the single and double pass welds. In both cases, the maximum tensile strain was found in the parent metal at the boundary of the heat affected zone (HAZ). Microstructural analysis of the welds was carried out using optical microscopy and hardness variations were also mapped across the weld-plate cross-section. The maximum hardness was observed in the mixed bainite/martensite structure of the weld nugget on the advancing side of the stir zone. The minimum hardness was observed in the HAZ.

  2. Residual strains and microstructure development in single and sequential double sided friction stir welds in RQT-701 steel

    International Nuclear Information System (INIS)

    Barnes, S.J.; Steuwer, A.; Mahawish, S.; Johnson, R.; Withers, P.J.

    2008-01-01

    Single and double sided partial penetration friction stir butt welds, in a rolled, quenched and tempered steel (RQT-701), were produced at The Welding Institute (TWI) under controlled process conditions. The residual strain distributions in the longitudinal and transverse directions have been measured using energy dispersive synchrotron X-ray diffraction. The measured strains were indicative of longitudinal tensile residual stresses at levels greater than the 0.2% yield stress of the parent metal in both the single and double pass welds. In both cases, the maximum tensile strain was found in the parent metal at the boundary of the heat affected zone (HAZ). Microstructural analysis of the welds was carried out using optical microscopy and hardness variations were also mapped across the weld-plate cross-section. The maximum hardness was observed in the mixed bainite/martensite structure of the weld nugget on the advancing side of the stir zone. The minimum hardness was observed in the HAZ

  3. Welding problems in nuclear power engineering

    International Nuclear Information System (INIS)

    Zubchenko, A.S.

    1986-01-01

    The problems of welding industry in nuclear power plant engineering, mainly related to the improvement of molten bath protection, are considered. Development of new materials for welding electrodes, for cladding and welding fluxes, is pointed out. Production of the following equipment is brought to a commercial level: welding heads and welding machines for branch pipe welding, anticorrosion cladding, zonal thermal treatment, electron beam welding facilities for the welding and maintenance of turbineblades, equipment for nondestructive testing of welded joints

  4. Method for welding beryllium

    Science.gov (United States)

    Dixon, R.D.; Smith, F.M.; O`Leary, R.F.

    1997-04-01

    A method is provided for joining beryllium pieces which comprises: depositing aluminum alloy on at least one beryllium surface; contacting that beryllium surface with at least one other beryllium surface; and welding the aluminum alloy coated beryllium surfaces together. The aluminum alloy may be deposited on the beryllium using gas metal arc welding. The aluminum alloy coated beryllium surfaces may be subjected to elevated temperatures and pressures to reduce porosity before welding the pieces together. The aluminum alloy coated beryllium surfaces may be machined into a desired welding joint configuration before welding. The beryllium may be an alloy of beryllium or a beryllium compound. The aluminum alloy may comprise aluminum and silicon. 9 figs.

  5. Method for welding beryllium

    International Nuclear Information System (INIS)

    Dixon, R.D.; Smith, F.M.; O'Leary, R.F.

    1997-01-01

    A method is provided for joining beryllium pieces which comprises: depositing aluminum alloy on at least one beryllium surface; contacting that beryllium surface with at least one other beryllium surface; and welding the aluminum alloy coated beryllium surfaces together. The aluminum alloy may be deposited on the beryllium using gas metal arc welding. The aluminum alloy coated beryllium surfaces may be subjected to elevated temperatures and pressures to reduce porosity before welding the pieces together. The aluminum alloy coated beryllium surfaces may be machined into a desired welding joint configuration before welding. The beryllium may be an alloy of beryllium or a beryllium compound. The aluminum alloy may comprise aluminum and silicon. 9 figs

  6. Transition welds in welding of two-ply steels

    International Nuclear Information System (INIS)

    Fartushnyj, V.G.; Evsyukov, Yu.G.

    1977-01-01

    Studied were physico-mechanical properties of welds made by various welding wires of chromium-nickel and nickel-chromium steels in submerged arc welding of double-layer steels with main layer of the VSt.3sp. carbon steel. It is shown that service-reliable structures welded of two-layer steels are obtained by providing the content from 11 to 20 % Ni in the automatically welded transition layer

  7. Grinding Parts For Automatic Welding

    Science.gov (United States)

    Burley, Richard K.; Hoult, William S.

    1989-01-01

    Rollers guide grinding tool along prospective welding path. Skatelike fixture holds rotary grinder or file for machining large-diameter rings or ring segments in preparation for welding. Operator grasps handles to push rolling fixture along part. Rollers maintain precise dimensional relationship so grinding wheel cuts precise depth. Fixture-mounted grinder machines surface to quality sufficient for automatic welding; manual welding with attendant variations and distortion not necessary. Developed to enable automatic welding of parts, manual welding of which resulted in weld bead permeated with microscopic fissures.

  8. Prediction of Weld Residual Stress of Narrow Gap Welds

    International Nuclear Information System (INIS)

    Yang, Jun Seog; Huh, Nam Su

    2010-01-01

    The conventional welding technique such as shield metal arc welding has been mostly applied to the piping system of the nuclear power plants. It is well known that this welding technique causes the overheating and welding defects due to the large groove angle of weld. On the other hand, the narrow gap welding(NGW) technique has many merits, for instance, the reduction of welding time, the shrinkage of weld and the small deformation of the weld due to the small groove angle and welding bead width comparing with the conventional welds. These characteristics of NGW affect the deformation behavior and the distribution of welding residual stress of NGW, thus it is believed that the residual stress results obtained from conventional welding procedure may not be applied to structural integrity evaluation of NGW. In this paper, the welding residual stress of NGW was predicted using the nonlinear finite element analysis to simulate the thermal and mechanical effects of the NGW. The present results can be used as the important information to perform the flaw evaluation and to improve the weld procedure of NGW

  9. Capabilities of infrared weld monitor

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, P.G.; Keske, J.S.; Leong, K.H.; Kornecki, G.

    1997-11-01

    A non-obtrusive pre-aligned, solid-state device has been developed to monitor the primary infrared emissions during laser welding. The weld monitor output is a 100-1000 mV signal that depends on the beam power and weld characteristics. The DC level of this signal is related to weld penetration, while AC portions of the output can be correlated with surface irregularities and part misalignment or contamination. Changes in DC behavior are also noted for both full and deep penetration welds. Full penetration welds are signified by an abrupt reduction in the weld monitor output. Bead on plate welds were made on steel, aluminum, and magnesium with both a CW CO{sub 2} laser and a pulsed Nd:YAG laser to explore the relationships between the weld characteristics and the weld monitor output.

  10. Welding method, and welding device for use therein, and method of analysis for evaluating welds

    NARCIS (Netherlands)

    Aendenroomer, A.J.; Den Ouden, G.; Xiao, Y.H.; Brabander, W.A.J.

    1995-01-01

    Described is a method of automatically welding pipes, comprising welding with a pulsation welding current and monitoring, by means of a sensor, the variations occurring in the arc voltage caused by weld pool oscillations. The occurrence of voltage variations with only frequency components below 100

  11. Investigation of mixing and diffusion processes in hybrid spot laser-MIG keyhole welding

    International Nuclear Information System (INIS)

    Zhou, J; Tsai, H L

    2009-01-01

    In hybrid laser-MIG keyhole welding, anti-crack elements can be added into the weld pool through a filler metal in anticipation of compensating mass loss, preventing porosity formation and improving compositional and mechanical properties of the welds. Understanding the mixing and diffusion of the filler metal in the molten pool is vital to achieve these desired objectives. In this study, mathematical models and associated numerical techniques have been developed to investigate the mixing and diffusion processes in hybrid laser-MIG keyhole welding. The transient interactions between droplets and weld pool and dynamics of the melt flow are studied. The effects of key process parameters, such as droplet size (wire diameter), droplet generation frequency (wire feed speed) and droplet impinging speed, on mixing/diffusion are systematically investigated. It was found that compositional homogeneity of the weld pool is determined by the competition between the mixing rate and the solidification rate. A small-size filler droplet together with high generation frequency can increase the latitudinal diffusion of the filler metal into the weld pool, while the large-size droplet along with the low generation frequency helps to get more uniform longitudinal diffusion. Increasing the impinging velocity of the filler droplet can improve the latitudinal diffusion of the filler metal. However, a high impinging velocity can cause a lower diffusion zone in the upper part of the welds. This study provides a good foundation for optimizing the hybrid laser-MIG keyhole welding process to achieve quality welds with desired properties.

  12. Advances in submerged arc, narrow-gap welding with strip electrodes and thin, dual-wire electrodes

    International Nuclear Information System (INIS)

    Nies, H.

    1990-01-01

    Container and tank construction for nuclear installations traditionally is one of the major applications of narrow-gap welding with the submerged arc technique. This type of welding presents one problem, namely to completely and reliably remove the welding slag from the deep and narrow gap. The research report in hand explains the variants of welding techniques that have been tested and describes the results obtained, which primarily are reduced occurrence of faults, i.e. enhanced reliability, and better welding economy. As an alternative to welding with thick wire electrodes, which is the standard method for the applications under review, a new technique has been conceived and extensively tested, which uses thin strip electrodes at longitudinal position in the gap. This submerged arc, dual-wire technique with thin electrodes is characterised by a significantly higher thermal efficiency compared to welding with thick wires, so that the same energy input yields better efficiency of metal deposition. (orig./MM) [de

  13. Weld analysis and control system

    Science.gov (United States)

    Kennedy, Larry Z. (Inventor); Rodgers, Michael H. (Inventor); Powell, Bradley W. (Inventor); Burroughs, Ivan A. (Inventor); Goode, K. Wayne (Inventor)

    1994-01-01

    The invention is a Weld Analysis and Control System developed for active weld system control through real time weld data acquisition. Closed-loop control is based on analysis of weld system parameters and weld geometry. The system is adapted for use with automated welding apparatus having a weld controller which is capable of active electronic control of all aspects of a welding operation. Enhanced graphics and data displays are provided for post-weld analysis. The system provides parameter acquisition, including seam location which is acquired for active torch cross-seam positioning. Torch stand-off is also monitored for control. Weld bead and parent surface geometrical parameters are acquired as an indication of weld quality. These parameters include mismatch, peaking, undercut, underfill, crown height, weld width, puddle diameter, and other measurable information about the weld puddle regions, such as puddle symmetry, etc. These parameters provide a basis for active control as well as post-weld quality analysis and verification. Weld system parameters, such as voltage, current and wire feed rate, are also monitored and archived for correlation with quality parameters.

  14. Welding processes handbook

    CERN Document Server

    Weman, Klas

    2003-01-01

    Deals with the main commercially significant and commonly used welding processes. This title takes the student or novice welder through the individual steps involved in each process in an easily understood way. It covers many of the requirements referred to in European Standards including EN719, EN 729, EN 729 and EN 287.$bWelding processes handbook is a concise, explanatory guide to the main commercially significant and commonly-used welding processes. It takes the novice welder or student through the individual steps involved in each process in a clear and easily understood way. It is intended to provide an up-to-date reference to the major applications of welding as they are used in industry. The contents have been arranged so that it can be used as a textbook for European welding courses in accordance with guidelines from the European Welding Federation. Welding processes and equipment necessary for each process are described so that they can be applied to all instruction levels required by the EWF and th...

  15. Thermal stir welding process

    Science.gov (United States)

    Ding, R. Jeffrey (Inventor)

    2012-01-01

    A welding method is provided for forming a weld joint between first and second elements of a workpiece. The method includes heating the first and second elements to form an interface of material in a plasticized or melted state interface between the elements. The interface material is then allowed to cool to a plasticized state if previously in a melted state. The interface material, while in the plasticized state, is then mixed, for example, using a grinding/extruding process, to remove any dendritic-type weld microstructures introduced into the interface material during the heating process.

  16. Thermal stir welding apparatus

    Science.gov (United States)

    Ding, R. Jeffrey (Inventor)

    2011-01-01

    A welding method and apparatus are provided for forming a weld joint between first and second elements of a workpiece. The method includes heating the first and second elements to form an interface of material in a plasticized or melted state interface between the elements. The interface material is then allowed to cool to a plasticized state if previously in a melted state. The interface material, while in the plasticized state, is then mixed, for example, using a grinding/extruding process, to remove any dendritic-type weld microstructures introduced into the interface material during the heating process.

  17. Review of Welding Terminology

    Directory of Open Access Journals (Sweden)

    Angelika Petrėtienė

    2011-04-01

    Full Text Available The paper discusses welding terms in accordance with the Lithuanian standard LST EN 1792 „Welding. The multilingual list of welding terms and similar processes”, „The Russian–Lithuanian dictionary of the terms of mechanical engineering technology and welding“ and the examples from postgraduates‘ final works. It analyses the infringement of lexical, word-building and morphological rules. First-year students should already be familiar with the standardized terms of their speciality. More active propagation of the terms should help to avoid terminology mistakes in various scientific spheres.

  18. Half bead welding technique

    International Nuclear Information System (INIS)

    Canonico, D.A.; Holz, P.P.

    1978-05-01

    The ORNL has employed the Section XI half-bead procedure for six repair welds. Table 2 identifies the repairs and the components upon which they were accomplished. The weld repairs were performed to permit us to evaluate material properties, residual stresses, weld repair procedures, and structural behavior of repaired pressure vessels. As a consequence of our study we concluded that when the half bead procedure is correctly applied: (1) there is no metallurgical degradation of the base material, (2) residual stresses of yield point magnitude will be present, and (3) the structural integrity of the pressure vessel is not impaired at Charpy V-notch upper shelf temperatures

  19. Head waves in ultrasonic testing. Physical principle and application to welded joint testing

    International Nuclear Information System (INIS)

    Wustenberg, H.; Erhard, A.

    1984-01-01

    A head wave sensor is developed from distinct emitter and receiver sensors using longitudinal waves under a 70 0 incidence. These heat wave sensors present a high sensitivity for underlying cracks and are not influenced by surface accidents like liquid drops or welding projection. They are multi mode sensors emitting simultaneously longitudinal head waves, a main longitudinal lobe and a transverse wave with a maximum at about 38 0 . This wave combination can be used for automatic testing of welded joints even with austenitic materials for defect detection near internal or external surfaces. This process can substitute or complete liquid penetrant inspection or magnetic inspection for testing pipes (13 references are given) [fr

  20. Characterization and Strain-Hardening Behavior of Friction Stir-Welded Ferritic Stainless Steel

    Science.gov (United States)

    Sharma, Gaurav; Dwivedi, Dheerendra Kumar; Jain, Pramod Kumar

    2017-12-01

    In this study, friction stir-welded joint of 3-mm-thick plates of 409 ferritic stainless steel (FSS) was characterized in light of microstructure, x-ray diffraction analysis, hardness, tensile strength, ductility, corrosion and work hardening properties. The FSW joint made of ferritic stainless steel comprises of three distinct regions including the base metal. In stir zone highly refined ferrite grains with martensite and some carbide precipitates at the grain boundaries were observed. X-ray diffraction analysis also revealed precipitation of Cr23C6 and martensite formation in heat-affected zone and stir zone. In tensile testing of the transverse weld samples, the failure eventuated within the gauge length of the specimen from the base metal region having tensile properties overmatched to the as-received base metal. The tensile strength and elongation of the longitudinal (all weld) sample were found to be 1014 MPa and 9.47%, respectively. However, in potentiodynamic polarization test, the corrosion current density of the stir zone was highest among all the three zones. The strain-hardening exponent for base metal, transverse and longitudinal (all weld) weld samples was calculated using various equations. Both the transverse and longitudinal weld samples exhibited higher strain-hardening exponents as compared to the as-received base metal. In Kocks-Mecking plots for the base metal and weld samples at least two stages of strain hardening were observed.

  1. Recent developments in pipeline welding practice

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    Fourteen chapters are included: overview of pipeline welding systems and quality assurance, CRC automatic welding system, H.C. Price Co. automatic welding system, semi-automatic MIG-welding process, partial penetration welding of steel pipes for gas distribution, construction procedures and quality control in offshore pipeline construction, welding in repair and maintenance of gas transmission pipelines, British Gas studies of welding on pressurized gas transmission pipelines, hot tapping pipelines, underwater welding for offshore pipelines and associated equipment, radial friction welding, material composition vs weld properties, review of NDT of pipeline welds, and safety assurance in pipeline construction. A bibliography of approximately 150 references is included, arranged according to subject and year.

  2. Hybrid laser-arc welding

    DEFF Research Database (Denmark)

    Hybrid laser-arc welding (HLAW) is a combination of laser welding with arc welding that overcomes many of the shortfalls of both processes. This important book gives a comprehensive account of hybrid laser-arc welding technology and applications. The first part of the book reviews...... the characteristics of the process, including the properties of joints produced by hybrid laser-arc welding and ways of assessing weld quality. Part II discusses applications of the process to such metals as magnesium alloys, aluminium and steel as well as the use of hybrid laser-arc welding in such sectors as ship...... building and the automotive industry. With its distinguished editor and international team of contributors, Hybrid laser-arc welding, will be a valuable source of reference for all those using this important welding technology. Professor Flemming Ove Olsen works in the Department of Manufacturing...

  3. Residual stress by repair welds

    International Nuclear Information System (INIS)

    Mochizuki, Masahito; Toyoda, Masao

    2003-01-01

    Residual stress by repair welds is computed using the thermal elastic-plastic analysis with phase-transformation effect. Coupling phenomena of temperature, microstructure, and stress-strain fields are simulated in the finite-element analysis. Weld bond of a plate butt-welded joint is gouged and then deposited by weld metal in repair process. Heat source is synchronously moved with the deposition of the finite-element as the weld deposition. Microstructure is considered by using CCT diagram and the transformation behavior in the repair weld is also simulated. The effects of initial stress, heat input, and weld length on residual stress distribution are studied from the organic results of numerical analysis. Initial residual stress before repair weld has no influence on the residual stress after repair treatment near weld metal, because the initial stress near weld metal releases due to high temperature of repair weld and then stress by repair weld regenerates. Heat input has an effect for residual stress distribution, for not its magnitude but distribution zone. Weld length should be considered reducing the magnitude of residual stress in the edge of weld bead; short bead induces high tensile residual stress. (author)

  4. Welding of refractory alloys

    International Nuclear Information System (INIS)

    Lessmann, G.G.

    1984-01-01

    This review primarily summarizes welding evaluations supported by NASA-Lewis Research Center in the 1960s. A literature search run in preparation for this review indicates that more recent work is modest by comparison. Hence, this review restates these accomplishments briefly and addresses opportunities which have evolved in welding technology (such as lasers) in the intervening decade. Emphasis in this review is given to tantalum- and niobium-base alloys. Considerable work was also done to assure that a consistent comparison was made with tungsten. A wide variety of candidate alloys derived primarily from developments directed at aircraft propulsion applications were available. Early efforts by NASA were directed at screening studies to select promising structural alloys for the space power application. This objective required fine tuning of welding procedures, e.g., the demonstration of stringent standards for control of welding atmosphere to assure good corrosion resistance in liquid alkali metals. 16 figures, 6 tables

  5. Multispot fiber laser welding

    DEFF Research Database (Denmark)

    Schutt Hansen, Klaus

    This dissertation presents work and results achieved in the field of multi beam fiber laser welding. The project has had a practical approach, in which simulations and modelling have been kept at a minimum. Different methods to produce spot patterns with high power single mode fiber lasers have...... been examined and evaluated. It is found that both diamond turned DOE’s in zinc sulphide and multilevel etched DOE’s (Diffractive Optical Elements) in fused silica have a good performance. Welding with multiple beams in a butt joint configuration has been tested. Results are presented, showing it has...... been possible to control the welding width in incremental steps by adding more beams in a row. The laser power was used to independently control the keyhole and consequently the depth of fusion. An example of inline repair of a laser weld in butt joint configuration was examined. Zinc powder was placed...

  6. Friction stir welding tool

    Science.gov (United States)

    Tolle,; Charles R. , Clark; Denis E. , Barnes; Timothy, A [Ammon, ID

    2008-04-15

    A friction stir welding tool is described and which includes a shank portion; a shoulder portion which is releasably engageable with the shank portion; and a pin which is releasably engageable with the shoulder portion.

  7. Electron beam welding in the fabrication of thick-walled large-size pipes of C-Mn steels. Final report; Elektronenstrahlschweissen bei der Fertigung von dickwandigen Grossrohren aus C-Mn-Staehlen. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Woeste, K

    2001-11-01

    This research project investigates electron beam welding as a method of fabrication of large-size pipes with longitudinal welds. The effects of the welding speed on the mechanical and technological properties of the weld are investigated. From the economic view, electron beam welding is much more favourable than submerged-arc welding. [German] Dieses Forschungsprojekt soll dazu beitragen, das Elektronenstrahlschweissen als Fertigungsverfahren fuer laengsnahtgeschweisste Grossrohre zu qualifizieren. Dabei wird der Einfluss der Schweissgeschwindigkeit auf die mechanisch-technologischen Eigenschaften der Schweissung untersucht. Im Wirtschaftlichkeitsvergleich schneidet Elektronenstrahlschweissverfahren gegenueber dem Unterpulverschweissverfahren eindeutig besser ab.

  8. Development of carbon/carbon composite control rod for HTTR. 2. Concept, specifications and mechanical test of materials

    International Nuclear Information System (INIS)

    Eto, Motokuni; Ishiyama, Shintaro; Fukaya, Kiyoshi; Saito, Tamotsu; Ishihara, Masahiro; Hanawa, Satoshi.

    1998-01-01

    A concept and specifications of carbon/carbon composite (C/C) control rod were proposed, aiming at the application of the material to the HTTR. The outer diameter and length of the control rod were kept as the same as those of the present control rod, i.e., 113 mm and 3094 mm, respectively. According to the concept, the rod consists of ten units which are connected in series using bolts. Then, the stresses generated by dead loads in the control rod elements were estimated and compared with the design strengths which were derived from the results of measurements of tensile, compressive, bending and shear strengths of two candidate materials, AC250 (Across Co.) and CX-270 (Toyo Tanso Co.). Design strength was preliminarily determined as one-third or one-fifth of the mean strength. Ratio of the design strength to generated stress for the AC250 (2D) was : Tensile stress in the outer sleeve tube, 66, tensile and shear stresses in the M16 bolt, 8.8 and 8.5, shear stress in the plug support bolt M8, 2.43. These results are believed to indicate the mechanical integrity of the control rod structure. Data available on the candidate materials were also compiled in the Appendix. (author)

  9. Underwater Welding Techniques

    OpenAIRE

    Esam F. Alajmi; Ahmad A. Alqenaei

    2017-01-01

    Welding demand in offshore and marine applications is increased with the increasing in oil and gas activities as well as increasing in the marine transportation and industrial applications. Applications of underwater welding well be increased in Kuwait in the coming years due to the strategic directive of the country toward starting the offshore oil and gas exploration and production, and the increase in marine transportation projects. Therefore, there is a need to understand the concept of u...

  10. Apparatus and process for ultrasonic seam welding stainless steel foils

    Science.gov (United States)

    Leigh, Richard W.

    1992-01-01

    An ultrasonic seam welding apparatus having a head which is rotated to form contact, preferably rolling contact, between a metallurgically inert coated surface of the head and an outside foil of a plurality of layered foils or work materials. The head is vibrated at an ultrasonic frequency, preferably along a longitudinal axis of the head. The head is constructed to transmit vibration through a contacting surface of the head into each of the layered foils. The contacting surface of the head is preferably coated with aluminum oxide to prevent the head from becoming welded to layered stainless steel foils.

  11. Automatic welding machine for piping

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  12. Increased lung function decline in blue-collar workers exposed to welding fumes.

    Science.gov (United States)

    Thaon, Isabelle; Demange, Valérie; Herin, Fabrice; Touranchet, Annie; Paris, Christophe

    2012-07-01

    There is no consensus at the present time about the effect of welding on lung function decline. This study compared lung function decline between blue-collar workers exposed and not exposed to welding fumes in a French longitudinal cohort of 21,238 subjects aged 37 to 52 years at inclusion. Medical data, occupation, sector of activity, and spirometry were recorded twice by occupational physicians in 1990 and 1995. A job-exposure matrix was used to identify 503 male blue-collar workers exposed to welding fumes and 709 control subjects and to define the weekly duration of exposure to welding fumes. Baseline lung function parameters were higher in workers exposed to welding fumes than in control subjects. After a 5-year follow-up, welding-fume exposure was associated with a nonsignificant decline in FVC (P = .06) and FEV(1) (P = .07) after adjustment for age, pack-years, BMI, and baseline value of the parameter. A significant accelerated decline in FEV(1) (P = .046) was also observed in never smokers exposed to welding fumes. An “exposure-response” relationship was observed between FEV(1) decline and weekly duration of exposure to welding fumes in nonsmokers but not in smokers. Blue-collar workers exposed to welding fumes showed accelerated decline in lung function, which, in nonsmokers, was related to weekly duration of exposure.

  13. Different finite element techniques to predict welding residual stresses in aluminum alloy plates

    International Nuclear Information System (INIS)

    Moein, Hadi; Sattari-Far, Iradj

    2014-01-01

    This study is a 3D thermomechanical finite element (FE) analysis of a single-pass and butt-welded work-hardened aluminum (Al) 5456 plates. It aims to validate the use of FE welding simulations to predict residual stress states in assessing the integrity of welded components. The predicted final residual stresses in the plate from the FE simulations are verified through comparison with experimental measurements. Three techniques are used to simulate the welding process. In the first two approaches, welding deposition is applied by using element birth and interaction techniques. In the third approach, the entire weld zone is simultaneously deposited. Results show a value at approximately the yield strength for longitudinal residual stresses of the welded center of the butt-welded Al alloy plates with a thickness of 2 mm. Considering the application of a comprehensive heat source, along with heat loss modeling and the temperature dependent properties of the material, the approach without deposition predicts a reasonable distribution of residual stresses. However, the element birth and interaction techniques, compared with the no-deposit technique, provide more accurate results in calculating residual stresses. Furthermore, the element interaction technique, compared with the element birth technique, exhibits higher efficiency and flexibility in modeling the deposition of welded metals as well as less modeling cost.

  14. Three Transducer Ultrasonic Examination of Nuclear Fuel Rod Flush Welds At ENUSA

    International Nuclear Information System (INIS)

    Domingo, A.; Jimenez, J.M.

    1998-01-01

    From 1991 ENUSA are using UT microscope examination of flush weld with longitudinal wave (perpendicular to weld) looking for welding defects as porosity, weld thickness and penetration. In 1994 we included a new transducer off set placed, with shear waves to control cracks in welding. In 1997 we incorporated a new shear transducer, 30 degree centigrade angle, in order to control different orientation of cracks or Grain Boundary Separation (GBS) and to improve software capabilities. Then actual UT microscope equipment used to inspect rod welds consist, mainly in three transducer of 50 MHz mounted over a rotatory head. UT system is electronically synchronized and obtains 750 data points per revolution by transducer. A set of seven images of approx 100.000 data points is obtained for each weld. Thickness, pore size, pore depth, two of crack size and two of crack depth are presented and evaluated evaluated by the computer to obtain weld disposition. Resolution of 0,05 mm pore size is achieved by this equipment, thickness and penetration precision should be in the order of 0,005 mm and 0,05 mm respectively. Crack detection depend basically on its orientation, nevertheless position of transducer assures a high capability detection of cracks which should be formed at these welds. (Author)

  15. Surface residual stress evaluation in double-electrode butt welded steel plates

    International Nuclear Information System (INIS)

    Estefen, S.F.; Gurova, T.; Castello, X.; Leontiev, A.

    2010-01-01

    Surface residual stress evaluation for double-electrode welding was studied. The stresses were monitored after each operational step: positioning, implementing of constraints, welding and constraints removal. The measurements were performed at the deposited metal, heat affected zone, base metal close to the weld joint and along the plate using the X-ray diffraction method. It was observed differences in the stress evaluations for double-electrode welding which resulted in lower bending distortions and higher values of surface residual stresses, compared with single-electrode welding. This behavior is associated with the stress distribution just after the welding processes in both heat affected zone and base metal close to the fillet for double-electrode welding. The main results from the laboratorial tests indicated lower values of the bending distortions for double-electrode welding compared with the single-electrode. In relation to the residual stress, the double-electrode welding generated, in general, higher stress values in both longitudinal and transversal directions.

  16. Alternate Welding Processes for In-Service Welding

    Science.gov (United States)

    2009-04-24

    Conducting weld repairs and attaching hot tap tees onto pressurized pipes has the advantage of avoiding loss of service and revenue. However, the risks involved with in-service welding need to be managed by ensuring that welding is performed in a rep...

  17. Certification of a weld produced by friction stir welding

    Science.gov (United States)

    Obaditch, Chris; Grant, Glenn J

    2013-10-01

    Methods, devices, and systems for providing certification of friction stir welds are disclosed. A sensor is used to collect information related to a friction stir weld. Data from the sensor is compared to threshold values provided by an extrinsic standard setting organizations using a certification engine. The certification engine subsequently produces a report on the certification status of the weld.

  18. Research and development for the high-temperature helium-leak detection system (Joint research). Part 2. Development of temperature sensors using optical fibre for the HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Sakaba, Nariaki; Nakazawa, Toshio; Kawasaki, Kozo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Urakami, Masao; Saisyu, Sadanori [Japan Atomic Power Co., Tokyo (Japan)

    2003-03-01

    In the second stage of the research and development for a high-temperature helium-leak detection system, the temperature sensor using optical fibres was studied. The sensor detects the helium leakage by the temperature increase surrounded optical fibre with or without heat insulator. Moreover, the applicability of high temperature equipments as the HTTR system was studied. With the sensor we detected 5.0-20.0 cm{sup 3}/s helium leakages within 60 minutes. Also it was possible to detect earlier when the leakage level is at 20.0 cm {sup 3}/s. (author)

  19. Ultrasonic inspection of austenitic welds

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  20. Arc-weld pool interactions

    International Nuclear Information System (INIS)

    Glickstein, S.S.

    1978-08-01

    The mechanisms involved in arc-weld pool interactions are extremely complex and no complete theory is presently available to describe much of the phenomena observed during welding. For the past several years, experimental and analytical studies have been undertaken at the Bettis Atomic Power Laboratory to increase basic understanding of the gas tungsten arc welding process. These studies have included experimental spectral analysis of the arc in order to determine arc temperature and analytical modeling of the arc and weld puddle. The investigations have been directed toward determining the cause and effects of variations in the energy distribution incident upon the weldment. In addition, the effect of weld puddle distortion on weld penetration was investigated, and experimental and analytical studies of weld process variables have been undertaken to determine the effects of the variables upon weld penetration and configuration. A review of the results and analysis of these studies are presented

  1. Resistance Spot Welding of dissimilar Steels

    Directory of Open Access Journals (Sweden)

    Ladislav Kolařík

    2012-01-01

    Full Text Available This paper presents an analysis of the properties of resistance spot welds between low carbon steel and austenitic CrNi stainless steel. The thickness of the welded dissimilar materials was 2 mm. A DeltaSpot welding gun with a process tape was used for welding the dissimilar steels. Resistance spot welds were produced with various welding parameters (welding currents ranging from 7 to 8 kA. Light microscopy, microhardness measurements across the welded joints, and EDX analysis were used to evaluate the quality of the resistance spot welds. The results confirm the applicability of DeltaSpot welding for this combination of materials.

  2. Multipass autogenous electron beam welding

    International Nuclear Information System (INIS)

    Murphy, J.L.; Mustaleski, T.M. Jr.; Watson, L.C.

    1986-01-01

    A multipass, autogenous welding procedure was developed for 7.6 mm (0.3 in.) wall thickness Type 304L stainless steel cylinders. The joint geometry has a 1.5 mm (0.06 in.) root-face width and a rectangular stepped groove that is 0.762 mm (0.03 in.) wide at the top of the root face and extends 1.5 mm in height, terminating into a groove width of 1.27 mm which extends to the outside of the 1.27 mm high weld-boss. One weld pass is made on the root, three passes on the 0.762 mm wide groove and three passes to complete the weld. Multipass, autogenous, electron beam welds maintain the characteristic high depth-to-width ratios and low heat input of single-pass, electron beam welds. The increased part distortion (which is still much less than from arc processes) in multipass weldments is corrected by a preweld machined compensation. Mechanical properties of multipass welds compare well with single-pass welds. The yield strength of welds in aluminum alloy 5083 is approximately the same for single-pass or multipass electron beam and gas, metal-arc welds. The incidence and size of porosity is less in multipass electron beam welding of aluminum as compared to gas, metal-arc welds. The multipass, autogenous, electron beam welding method has proven to be a reliable way to make some difficult welds in multilayer parts or in an instance where inside part temperature or weld underbead must be controlled and weld discontinuities must be minimized

  3. The influence of low dose irradiation on the creep properties of type 316 welds

    International Nuclear Information System (INIS)

    Marshall, P.; Steeds, J.W.; Lin, Y.P.; Finlan, G.T.

    1987-01-01

    Fully instrumented creep and stress rupture tests have been performed at 873K for times up to 20,000h on a series of type 316 steel/17Cr 8Ni 2Mo weld metal specimens in the unirradiated and thermal neutron irradiated conditions. The specimens tested included all weld metal longitudinal and transverse composites in the as-welded condition and following a stress relief heat treatment of 10h at 1075K. Simulated heat affected zone (HAZ) specimens were also tested. Analysis of the creep results combined with metallography, autoradiography and TEM established that the decrease in properties of irradiated samples is caused by an increasing secondary strain rate due to enhanced helium induced grain boundary fracture of the simulated HAZ and enhanced interdendritic fracture in the weld metal. Implications of strength reductions on the design of welded structures subjected to thermal irradiation are briefly assessed. (author)

  4. Laser welding engineering

    International Nuclear Information System (INIS)

    Bhieh, N. M.; El Eesawi, M. E.; Hashkel, A. E.

    2007-01-01

    Laser welding was in its early life used mainly for unusual applications where no other welding process would be suitable that was twenty five years ago. Today, laser welding is a fully developed part of the metal working industry, routinely producing welds for common items such as cigarette lighters, which springs, motor/transformer lamination, hermetic seals, battery and pacemaker cans and hybrid circuit packages. Yet very few manufacturing engineering have seriously considers employing lasers in their own operations. Why? There are many reasons, but a main one must be not acquainted with the operation and capabilities of a laser system. Other reasons, such as a relatively high initial cost and a concern about using lasers in the manufacturing environment, also are frequently cited, and the complexity of the component and flexibility of the light delivery system. Laser welding could be used in place of many different standard processes, such as resistance (spot or seam), submerged arc, RF induction, high-frequency resistance, ultrasonic and electronic and electron-beam. while each of these techniques has established an independent function in the manufacturing world, the flexible laser welding approach will operate efficiently and economically in many different applications. Its flexibility will even permit the welding system to be used for other machining function, such as drilling, scribing, sealing and serializing. In this article, we will look at how laser welding works and what benefits it can offer to manufacturing engineers. Some industry observers state that there are already 2,000 laser machine tools being used for cutting, welding and drilling and that the number could reach 30,000 over the next 15 years as manufacturing engineers become more aware of the capabilities of lasers [1). While most laser applications are dedicated to one product or process that involves high-volume, long-run manufacturing, the flexibility of a laser to supply energy to hard

  5. Welding. Performance Objectives. Intermediate Course.

    Science.gov (United States)

    Vincent, Kenneth

    Several intermediate performance objectives and corresponding criterion measures are listed for each of nine terminal objectives for an intermediate welding course. The materials were developed for a 36-week (3 hours daily) course designed to prepare the student for employment in the field of welding. Electric welding and specialized (TIG & MIG)…

  6. Welding. Performance Objectives. Basic Course.

    Science.gov (United States)

    Vincent, Kenneth

    Several intermediate performance objectives and corresponding criterion measures are listed for each of eight terminal objectives for a basic welding course. The materials were developed for a 36-week (2 hours daily) course developed to teach the fundamentals of welding shop work, to become familiar with the operation of the welding shop…

  7. Pulsed TIG welding of pipes

    International Nuclear Information System (INIS)

    Killing, U.

    1989-01-01

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

  8. Residual stress investigation of copper plate and canister EB-Welds Complementary Results

    International Nuclear Information System (INIS)

    Gripenberg, H.

    2009-03-01

    distribution in Step II of the EB-weld differed from this behaviour by following the trend of the BM results. The CM presented weld longitudinal stress maps over the whole cross-section of the EB-welded plates. Tensile stresses of 50 - 70 MPa were observed in the welds and widely outside the HAZ, almost 10 times the width of the weld. (orig.)

  9. Friction stir welding tool and process for welding dissimilar materials

    Science.gov (United States)

    Hovanski, Yuri; Grant, Glenn J; Jana, Saumyadeep; Mattlin, Karl F

    2013-05-07

    A friction stir welding tool and process for lap welding dissimilar materials are detailed. The invention includes a cutter scribe that penetrates and extrudes a first material of a lap weld stack to a preselected depth and further cuts a second material to provide a beneficial geometry defined by a plurality of mechanically interlocking features. The tool backfills the interlocking features generating a lap weld across the length of the interface between the dissimilar materials that enhances the shear strength of the lap weld.

  10. Plasma Processes of Cutting and Welding

    Science.gov (United States)

    1976-02-01

    TIG process. 2.2.2 Keyhole Welding In plasma arc welding , the term...Cutting 3 3 4 4 4 2.2 Plasma Arc Welding 5 2.2.1 Needle Arc Welding 2.2.2 Keyhole Welding 5 6 3. Applications 8 93.1 Economics 4. Environmental Aspects of...Arc Lengths III. Needle Arc Welding Conditions IV. Keyhole Welding Conditions v. Chemical Analyses of Plates Used - vii - 1. 2. 3. 4. 5. 6. 7. 8.

  11. Friction Stir Welding Process: A Green Technology

    OpenAIRE

    Esther T. Akinlabi; Stephen A. Akinlabi

    2012-01-01

    Friction Stir Welding (FSW) is a solid state welding process invented and patented by The Welding Institute (TWI) in the United Kingdom in 1991 for butt and lap welding of metals and plastics. This paper highlights the benefits of friction stir welding process as an energy efficient and a green technology process in the field of welding. Compared to the other conventional welding processes, its benefits, typical applications and its use in joining similar and dissimilar materia...

  12. EB-welding of the copper canister for the nuclear waste disposal. Final report of the development programme 1994-1997

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, H. [Outokumpu Oy Poricopper, Pori (Finland)

    1998-10-01

    During 1994-1997 Posiva Oy and Outokumpu Poricopper Oy had a joint project Development of EB-welding method for massive copper canister manufacturing. The project was part of the national technology program `Weld 2000` and it was supported financially by Technology Development Centre (TEKES). The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper canisters and placed deep into the bedrock. The thick copper layer of the canister provides a long time corrosion resistance and prevents deposited nuclear fuel from contact with water. The quality requirements of the copper components are high because of the designed long lifetime of the canister. The EB-welding technology has proved to be applicable method for the production of the copper canisters and the EB-welding technique is needed at least when the lids of the copper canister will be closed. There are a number of parameters in EB-welding which affect weldability. However, the effect of the welding parameters and their optimization has not been extensively studied in welding of thick copper sections using conventional high vacuum EB-welding. One aim of this development work was to extensively study effect of welding parameters on weld quality. The final objective was to minimise welding defects in the main weld and optimize slope out procedure in thick copper EB-welding. Welding of 50 mm thick copper sections was optimized using vertical and horizontal EB-welding techniques. As a result two full scale copper lids were welded to a short cylinder successfully. The resulting weld quality with optimised welding parameters was reasonable good. The optimised welding parameters for horizontal and vertical beam can be applied to the longitudinal body welds of the canister. The optimal slope out procedure for the lid closure needs some additional development work. In addition of extensive EB-welding program ultrasonic inspection and creep strength of the weld were studied. According

  13. EB-welding of the copper canister for the nuclear waste disposal. Final report of the development programme 1994-1997

    International Nuclear Information System (INIS)

    Aalto, H.

    1998-10-01

    During 1994-1997 Posiva Oy and Outokumpu Poricopper Oy had a joint project Development of EB-welding method for massive copper canister manufacturing. The project was part of the national technology program 'Weld 2000' and it was supported financially by Technology Development Centre (TEKES). The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper canisters and placed deep into the bedrock. The thick copper layer of the canister provides a long time corrosion resistance and prevents deposited nuclear fuel from contact with water. The quality requirements of the copper components are high because of the designed long lifetime of the canister. The EB-welding technology has proved to be applicable method for the production of the copper canisters and the EB-welding technique is needed at least when the lids of the copper canister will be closed. There are a number of parameters in EB-welding which affect weldability. However, the effect of the welding parameters and their optimization has not been extensively studied in welding of thick copper sections using conventional high vacuum EB-welding. One aim of this development work was to extensively study effect of welding parameters on weld quality. The final objective was to minimise welding defects in the main weld and optimize slope out procedure in thick copper EB-welding. Welding of 50 mm thick copper sections was optimized using vertical and horizontal EB-welding techniques. As a result two full scale copper lids were welded to a short cylinder successfully. The resulting weld quality with optimised welding parameters was reasonable good. The optimised welding parameters for horizontal and vertical beam can be applied to the longitudinal body welds of the canister. The optimal slope out procedure for the lid closure needs some additional development work. In addition of extensive EB-welding program ultrasonic inspection and creep strength of the weld were studied. According

  14. Weld Nugget Temperature Control in Thermal Stir Welding

    Science.gov (United States)

    Ding, R. Jeffrey (Inventor)

    2014-01-01

    A control system for a thermal stir welding system is provided. The control system includes a sensor and a controller. The sensor is coupled to the welding system's containment plate assembly and generates signals indicative of temperature of a region adjacent and parallel to the welding system's stir rod. The controller is coupled to the sensor and generates at least one control signal using the sensor signals indicative of temperature. The controller is also coupled to the welding system such that at least one of rotational speed of the stir rod, heat supplied by the welding system's induction heater, and feed speed of the welding system's weld material feeder are controlled based on the control signal(s).

  15. Thermal Stir Welding: A New Solid State Welding Process

    Science.gov (United States)

    Ding, R. Jeffrey

    2003-01-01

    Thermal stir welding is a new welding process developed at NASA's Marshall Space Flight Center in Huntsville, AL. Thermal stir welding is similar to friction stir welding in that it joins similar or dissimilar materials without melting the parent material. However, unlike friction stir welding, the heating, stirring and forging elements of the process are all independent of each other and are separately controlled. Furthermore, the heating element of the process can be either a solid-state process (such as a thermal blanket, induction type process, etc), or, a fusion process (YG laser, plasma torch, etc.) The separation of the heating, stirring, forging elements of the process allows more degrees of freedom for greater process control. This paper introduces the mechanics of the thermal stir welding process. In addition, weld mechanical property data is presented for selected alloys as well as metallurgical analysis.

  16. Numerical simulation of welding

    DEFF Research Database (Denmark)

    Hansen, Jan Langkjær; Thorborg, Jesper

    Aim of project:To analyse and model the transient thermal field from arc welding (SMAW, V-shaped buttweld in 15mm plate) and to some extend the mechanical response due to the thermal field. - To implement this model in a general purpose finite element program such as ABAQUS.The simulation...... stress is also taken into account.Work carried out:With few means it is possible to define a thermal model which describes the thermal field from the welding process in reasonable agreement with reality. Identical results are found with ABAQUS and Rosenthal’s analytical solution of the governing heat...... transfer equation under same conditions. It is relative easy tointroduce boundary conditions such as convection and radiation where not surprisingly the radiation has the greatest influence especially from the high temperature regions in the weld pool and the heat affected zone.Due to the large temperature...

  17. Milestones in welding technology

    Science.gov (United States)

    Dolby, Richard E.

    2013-09-01

    Sir Alan's PhD thesis describes his research into cracking during arc welding of armour steels. Throughout his career, he had a strong interest in defects of all types, how they formed in metallic structures and how the larger ones could be detected and sized by non-destructive techniques. He was also vitally concerned with how defects impacted on the engineering integrity of welded structures, particularly the risk of fracture in nuclear plant. This study presents a view of some of the major milestones in global welding technology that took place over the 60 or more years of Sir Alan's career and highlights those where he had a personal and direct involvement.

  18. Design and development of weld inspection manipulator for reactor pressure vessel of TAPS-1

    International Nuclear Information System (INIS)

    Chatterjee, H.; Singh, J.P.; Ranjon, R.; Kulkarni, M.P.; Patel, R.J.

    2013-01-01

    The reactor pressure vessel (RPV) of TAPS-1 BWR contains six longitudinal and four circumferential welds. Periodical in-service inspection of these weld joints has been a regulatory issue pending for long. In the 22 nd refuelling outage in July 2012 the inspection of L1-1, L1-2 longitudinal welds as well as their junctions with C1 circumferential weld were proposed to be done using ultrasonic technique. Approaching these welds from OD side of the RPV is a difficult and tedious task. Therefore it was decided to examine these welds from ID side of the RPV by filling the cavity with water and approaching the RPV from top. No technology was locally available to take the probes at a depth of 10-12 m under water. NPCIL approached RTD, BARC to develop an underwater manipulator to accomplish this task. RTD took up this work as a challenge and came out with the design of manipulator. The weld inspection manipulator (WIM) was fabricated on a war foot basis, tested and successfully implemented in the reactor for the first time in TAPS history. The entire activity was completed in three months time. This article gives the details of design, manufacturing, performance testing, qualification trials and implementation of WIM in the reactor. Ultrasonic testing techniques were developed by QAD, BARC which are not covered in this article. (author)

  19. Study on the stress and strain during welding of plate-to-pipe joint

    Energy Technology Data Exchange (ETDEWEB)

    Na, S.J.; Kim, H.W.

    1986-09-01

    In manufacturing of pipe walls for boiler units, distortion can result in pipe-to-pipe joints from the nonuniform expansion and contraction of the weld metal and the adjacent base metal during heating and cooling cycle of the welding process. In this study, the stresses and strains during longitudinal welding of the plate-to-pipe joint were investigated. Using the method of sucessive elastic solution, longitudinal stresses and strains during and after welding were calculated from the information of temperature distributions obtained by Rosenthal's equations. In order to confirm the validity of the numerical results, the temperature and residual stress distributions were measured and compared with the calculated results. In spite of some assumptions, the one-dimensional analytical results of residual stresses were in fairly good agreement with the experimental ones. The residual stresses due to welding of plate-to-pipe joints are tensile near the weld line and compressive in the base metal as in the welding of plates. The amount and distribution of residual stresses were deeply dependent on the heat input ratio of the plate and pipe.

  20. Influence of repair length on residual stress in the repair weld of a clad plate

    International Nuclear Information System (INIS)

    Jiang Wenchun; Xu, X.P.; Gong, J.M.; Tu, S.T.

    2012-01-01

    Highlights: ► Residual stress in the repair weld of a stainless steel clad plate is investigated. ► The effect of repair length on residual stress has been studied. ► Large tensile residual stress is generated in the repair weld and heat affected zone. ► With the increase of repair length, transverse stress is decreased. ► Repair length has little effect on longitudinal stress. - Abstract: A 3-D sequential coupling finite element simulation is performed to investigate the temperature field and residual stress in the repair weld of a stainless steel clad plate. The effect of repair length on residual stress has been studied, aiming to provide a reference for repairing the cracked clad plate. The results show that large tensile residual stresses are generated in the repair weld and heat affected zone (HAZ), and then decrease gradually away from the weld and HAZ. The residual stresses through thickness in the clad layer are relative uniform, while they are non-uniform in the base metal. A discontinuous stress distribution is generated across the interface between weld metal and base metal. The repair length has a great effect on transverse stress. With the increase of repair length, the transverse stress is decreased. When the repair length is increased to 14 cm, the peak of transverse stress has been decreased below yield strength, and the transverse stress in the weld and HAZ has also been greatly decreased. But the repair length has little effect on longitudinal stress.

  1. Mechanized hyperbaric welding by robots

    International Nuclear Information System (INIS)

    Aust, E.; Santos, J.F. dos; Bohm, K.H.; Hensel, H.D.

    1988-01-01

    At the GKSS-Forschungszentrum investigations are carried out on mechanized welded test plates produced under working pressure between 10 to 110 bar in breathable TRIMIX-5-atmosphere. The welds are performed by a modified industrial robot, which was adapted in its components to withstand these severe conditions. Variations on the welding parameters were made to maintain a stable arc as well as to provide on indication of the effect of the variables on the mechanical properties of the welded joint. During all tests the robot showed a very good function. Good reliable welds were achieved meeting the requirements according API II04 or BS 4515-1984. (orig.) [de

  2. Spot Welding of Honeycomb Structures

    Science.gov (United States)

    Cohal, V.

    2017-08-01

    Honeycomb structures are used to prepare meals water jet cutting machines for textile. These honeycomb structures are made of stainless steel sheet thickness of 0.1-0.2 mm. Corrugated sheet metal strips are between two gears with special tooth profile. Hexagonal cells for obtaining these strips are welded points between them. Spot welding device is three electrodes in the upper part, which carries three welding points across the width of the strip of corrugated sheet metal. Spot welding device filled with press and advance mechanisms. The paper presents the values of the regime for spot welding.

  3. Understanding Friction Stir Welding

    Science.gov (United States)

    Nunes, A. C., Jr.

    2018-01-01

    This Technical Memorandum explains the friction stir welding process in terms of two basic concepts: the concentration of deformation in a shear surface enveloping the tool and the composition of the overall plastic flow field around the tool from simple flow field components. It is demonstrated how weld structure may be understood and torque, drag, and lateral tool forces may be estimated using these concepts. Some discrepancies between computations and accompanying empirical data are discussed in the text. This work is intended to be helpful to engineers in diagnosing problems and advancing technology.

  4. Characteristics of DC electrical braking method of the gas circulator to limit the temperature rise at the heat transfer pipes in the HTTR

    International Nuclear Information System (INIS)

    Kawasaki, K.; Saito, K.; Iyoku, T.

    2001-01-01

    In the safety evaluation of a High Temperature Engineering Test Reactor (HTTR), it must be confirmed that the core has no chance to be damaged and the barrier against the FP release is designed properly not to be affecting the influence of radiation around the reactor site. Especially the maximum temperature of the reactor pressure boundary such as the heat transfer pipes of pressurized water cooler (PWC) must not exceed the permissible values under an anticipated accident such as pipe of rupture in PWC. A requirement for the gas circulator which circulates helium gas in the primary cooling line and the secondary cooling line, is to be braked within 10 seconds by an electrical braking method after the HTTR reactor has scrammed under the accident in PWC. The reason is that the temperature rise of the heat transfer pipe at PWC has to be suppressed when the gas circulator has stopped, the revolution of the gas circulator decreases like the free coast down so that it takes about 90 seconds to be zero and the temperature rise of the pipe in the PWC exceeds the permissible value. By braking within 10 secs., the temperature of the pipe in the PWC reaches about 368 deg. C, less than the permissible value. Using a simplified equivalent circuit of an induction motor, braking time analysis was performed with obtained electrical resistance and inductance. The obtained braking time is about 10 secs., showing close agreement with analysis values. (author)

  5. Welding method by remote handling

    International Nuclear Information System (INIS)

    Hashinokuchi, Minoru.

    1994-01-01

    Water is charged into a pit (or a water reservoir) and an article to be welded is placed on a support in the pit by remote handling. A steel plate is disposed so as to cover the article to be welded by remote handling. The welding device is positioned to the portion to be welded and fixed in a state where the article to be welded is shielded from radiation by water and the steel plate. Water in the pit is drained till the portion to be welded is exposed to the atmosphere. Then, welding is conducted. After completion of the welding, water is charged again to the pit and the welding device and fixing jigs are decomposed in a state where the article to be welded is shielded again from radiation by water and the steel plate. Subsequently, the steel plate is removed by remote handling. Then, the article to be welded is returned from the pit to a temporary placing pool by remote handling. This can reduce operator's exposure. Further, since the amount of the shielding materials can be minimized, the amount of radioactive wastes can be decreased. (I.N.)

  6. Ultrasonic inspection of austenitic welds

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-11-01

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

  7. Visualization of Spot- welding Resistance

    Directory of Open Access Journals (Sweden)

    Michal Černý

    2016-01-01

    Full Text Available This contribution devotes to monitoring of processes running during joining of steel sheets by incadescent so called point welding using non-destructive trial method – acoustic emission (AE. The joining process is detailed described within experimental measuring from the point of view of metallurgic effects runnig during weld creation (records obtained by means of AE method. It takes into consideration quality of joined steels within welding data of steel producer. Steel welding (determined by chemical composition during mechanical verification and firmness of welds consider results of measurement AE and fracture effect of point joints. The measurement also demonstrates conclusion about connection of metallurgic processes with material wave effects (AE measurement and their impact on firmness of joint at steel with guaranteed welding, difficult welding and at their potential combination.

  8. Recent Corrosion Research Trends in Weld Joints

    International Nuclear Information System (INIS)

    Kim, Hwan Tae; Kil, Sang Cheol; Hwang, Woon Suk

    2007-01-01

    The increasing interest in the corrosion properties of weld joints in the corrosive environment is placing stringent demands on the manufacturing techniques and performance requirements, and the manufacture employs the high quality and efficiency welding process to produce welds. Welding plays an important role in the fabrication of chemical plants, nuclear power plant, ship construction, and this has led to an increasing attention to the corrosion resistant weld joints. This paper covers recent technical trends of welding technologies for corrosion resistance properties including the COMPENDEX DB analysis of welding materials, welding process, and welding fabrications

  9. Repair welding of fusion reactor components. Final technical report

    International Nuclear Information System (INIS)

    Chin, B.A.; Wang, C.A.

    1997-01-01

    The exposure of metallic materials, such as structural components of the first wall and blanket of a fusion reactor, to neutron irradiation will induce changes in both the material composition and microstructure. Along with these changes can come a corresponding deterioration in mechanical properties resulting in premature failure. It is, therefore, essential to expect that the repair and replacement of the degraded components will be necessary. Such repairs may require the joining of irradiated materials through the use of fusion welding processes. The present ITER (International Thermonuclear Experimental Reactor) conceptual design is anticipated to have about 5 km of longitudinal welds and ten thousand pipe butt welds in the blanket structure. A recent study by Buende et al. predict that a failure is most likely to occur in a weld. The study is based on data from other large structures, particularly nuclear reactors. The data used also appear to be consistent with the operating experience of the Fast Flux Test Facility (FFTF). This reactor has a fuel pin area comparable with the area of the ITER first wall and has experienced one unanticipated fuel pin failure after two years of operation. The repair of irradiated structures using fusion welding will be difficult due to the entrapped helium. Due to its extremely low solubility in metals, helium will diffuse and agglomerate to form helium bubbles after being trapped at point defects, dislocations, and grain boundaries. Welding of neutron-irradiated type 304 stainless steels has been reported with varying degree of heat-affected zone cracking (HAZ). The objectives of this study were to determine the threshold helium concentrations required to cause HAZ cracking and to investigate techniques that might be used to eliminate the HAZ cracking in welding of helium-containing materials

  10. Quantitative metal magnetic memory reliability modeling for welded joints

    Science.gov (United States)

    Xing, Haiyan; Dang, Yongbin; Wang, Ben; Leng, Jiancheng

    2016-03-01

    Metal magnetic memory(MMM) testing has been widely used to detect welded joints. However, load levels, environmental magnetic field, and measurement noises make the MMM data dispersive and bring difficulty to quantitative evaluation. In order to promote the development of quantitative MMM reliability assessment, a new MMM model is presented for welded joints. Steel Q235 welded specimens are tested along the longitudinal and horizontal lines by TSC-2M-8 instrument in the tensile fatigue experiments. The X-ray testing is carried out synchronously to verify the MMM results. It is found that MMM testing can detect the hidden crack earlier than X-ray testing. Moreover, the MMM gradient vector sum K vs is sensitive to the damage degree, especially at early and hidden damage stages. Considering the dispersion of MMM data, the K vs statistical law is investigated, which shows that K vs obeys Gaussian distribution. So K vs is the suitable MMM parameter to establish reliability model of welded joints. At last, the original quantitative MMM reliability model is first presented based on the improved stress strength interference theory. It is shown that the reliability degree R gradually decreases with the decreasing of the residual life ratio T, and the maximal error between prediction reliability degree R 1 and verification reliability degree R 2 is 9.15%. This presented method provides a novel tool of reliability testing and evaluating in practical engineering for welded joints.

  11. Welding technology transfer task/laser based weld joint tracking system for compressor girth welds

    Science.gov (United States)

    Looney, Alan

    1991-01-01

    Sensors to control and monitor welding operations are currently being developed at Marshall Space Flight Center. The laser based weld bead profiler/torch rotation sensor was modified to provide a weld joint tracking system for compressor girth welds. The tracking system features a precision laser based vision sensor, automated two-axis machine motion, and an industrial PC controller. The system benefits are elimination of weld repairs caused by joint tracking errors which reduces manufacturing costs and increases production output, simplification of tooling, and free costly manufacturing floor space.

  12. The origin of weld seam defects related to metal flow in the hot extrusion of aluminium alloys en AW-6060 and en AW-6082

    NARCIS (Netherlands)

    Bakker, A.J. den; Werkhoven, R.J.; Sillekens, W.H.; Katgerman, L.

    2014-01-01

    Longitudinal weld seams are an intrinsic feature in hollow extrusions produced with porthole dies. As these joins occur along the entire extruded length, it is desirable that these weld seams have a minimal impact on the structural integrity of the extrudate. In particular, defects associated with

  13. 49 CFR 192.225 - Welding procedures.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Welding procedures. 192.225 Section 192.225... BY PIPELINE: MINIMUM FEDERAL SAFETY STANDARDS Welding of Steel in Pipelines § 192.225 Welding procedures. (a) Welding must be performed by a qualified welder in accordance with welding procedures...

  14. Integrated sensors for robotic laser welding

    NARCIS (Netherlands)

    Iakovou, D.; Aarts, Ronald G.K.M.; Meijer, J.; Beyer, E.; Dausinger, F; Ostendorf, A; Otto, A.

    2005-01-01

    A welding head is under development with integrated sensory systems for robotic laser welding applications. Robotic laser welding requires sensory systems that are capable to accurately guide the welding head over a seam in three-dimensional space and provide information about the welding process as

  15. ICT diagnostic method of beryllium welding quality

    International Nuclear Information System (INIS)

    Sun Lingxia; Wei Kentang; Ye Yunchang

    2002-01-01

    To avoid the interference of high density material for the quality assay of beryllium welding line, a slice by slice scanning method was proposed based upon the research results of the Industrial Computerized Tomography (ICT) diagnostics for weld penetration, weld width, off-centered deviation and weld defects of beryllium-ring welding seam with high density material inside

  16. Sensor integration for robotic laser welding processes

    NARCIS (Netherlands)

    Iakovou, D.; Aarts, Ronald G.K.M.; Meijer, J.; Ostendorf, A; Hoult, A.; Lu, Y.

    2005-01-01

    The use of robotic laser welding is increasing among industrial applications, because of its ability to weld objects in three dimensions. Robotic laser welding involves three sub-processes: seam detection and tracking, welding process control, and weld seam inspection. Usually, for each sub-process,

  17. Numerical analysis of weld pool oscillation in laser welding

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jung Ho [Chungbuk National University, Cheongju (Korea, Republic of); Farson, Dave F [The Ohio State University, Columbus (United States); Hollis, Kendall; Milewski, John O. [Los Alamos National Laboratory, Los Alamos (United States)

    2015-04-15

    Volume of fluid (VOF) numerical simulation was used to investigate melt flow and volumetric oscillation of conduction-mode pulsed laser weld pools. The result is compared to high speed video stream of titanium laser spot welding experiment. The total simulation time is 10ms with the first 5 ms being heating and melting under constant laser irradiation and the remaining 5 ms corresponding to resolidification of the weld pool. During the melting process, the liquid pool did not exhibit periodic oscillation but was continually depressed by the evaporation recoil pressure. After the laser pulse, the weld pool was excited into volumetric oscillation by the release of pressure on its surface and oscillation of the weld pool surface was analyzed. The simulation model suggested adjusting thermal diffusivity to match cooling rate and puddle diameter during solidification which is distinguishable from previous weld pool simulation. The frequency continuously increased from several thousand cycles per second to tens of thousands of cycles per second as the weld pool solidified and its diameter decreased. The result is the first trial of investigation of small weld pool oscillation in laser welding although there have been several reports about arc welding.

  18. Deconvoluting the Friction Stir Weld Process for Optimizing Welds

    Science.gov (United States)

    Schneider, Judy; Nunes, Arthur C.

    2008-01-01

    In the friction stir welding process, the rotating surfaces of the pin and shoulder contact the weld metal and force a rotational flow within the weld metal. Heat, generated by the metal deformation as well as frictional slippage with the contact surface, softens the metal and makes it easier to deform. As in any thermo-mechanical processing of metal, the flow conditions are critical to the quality of the weld. For example, extrusion of metal from under the shoulder of an excessively hot weld may relax local pressure and result in wormhole defects. The trace of the weld joint in the wake of the weld may vary geometrically depending upon the flow streamlines around the tool with some geometry more vulnerable to loss of strength from joint contamination than others. The material flow path around the tool cannot be seen in real time during the weld. By using analytical "tools" based upon the principles of mathematics and physics, a weld model can be created to compute features that can be observed. By comparing the computed observations with actual data, the weld model can be validated or adjusted to get better agreement. Inputs to the model to predict weld structures and properties include: hot working properties ofthe metal, pin tool geometry, travel rate, rotation and plunge force. Since metals record their prior hot working history, the hot working conditions imparted during FSW can be quantified by interpreting the final microstructure. Variations in texture and grain size result from variations in the strain accommodated at a given strain rate and temperature. Microstructural data from a variety of FSWs has been correlated with prior marker studies to contribute to our understanding of the FSW process. Once this stage is reached, the weld modeling process can save significant development costs by reducing costly trial-and-error approaches to obtaining quality welds.

  19. Ultrasonic Spot and Torsion Welding of Aluminum to Titanium Alloys: Process, Properties and Interfacial Microstructure

    Science.gov (United States)

    Balle, Frank; Magin, Jens

    Hybrid lightweight structures shape the development of future vehicles in traffic engineering and the aerospace industry. For multi-material concepts made out of aluminum and titanium alloys, the ultrasonic welding technique is an alternative effective joining technology. The overlapped structures can be welded in the solid state, even without gas shielding. In this paper the conventional ultrasonic spot welding with longitudinal oscillation mode is compared to the recent ultrasonic torsion welding with a torsional mode at 20 kHz working frequency. For each technique the process parameters welding force, welding energy and oscillation amplitude were optimized for the hybrid joints using design of experiments. Relationships between the process parameters, mechanical properties and related welding zone should be understood. Central aspects of the research project are microscopic studies of the joining zone in cross section and extensive fracture surface analysis. Detailed electron microscopy and spectroscopy of the hybrid interface help to understand the interfacial formation during ultrasonic welding as well as to transfer the gained knowledge for further multi-metal joints.

  20. Welding induced residual stress evaluation using laser-generated Rayleigh waves

    Science.gov (United States)

    Ye, Chong; Zhou, Yuanlai; Reddy, Vishnu V. B.; Mebane, Aaron; Ume, I. Charles

    2018-04-01

    Welding induced residual stress could affect the dimensional stability, fatigue life, and chemical resistance of the weld joints. Ultrasonic method serves as an important non-destructive tool for the residual stress evaluation due to its easy implementation, low cost and wide application to different materials. Residual stress would result in the ultrasonic wave velocity variation, which is the so called acoustoelastic effect. In this paper, Laser/EMAT ultrasonic technique was proposed to experimentally study the relative velocity variation ΔV/V of Rayleigh wave, which has the potential to evaluate surface/subsurface longitudinal residual stress developed during the Gas Metal Arc Welding process. Broad band ultrasonic waves were excited by pulsed Q-Switched Nd: YAG laser. An electromagnetic acoustic transducer (EMAT) attached to the welded plates was used to capture the Rayleigh wave signals propagating along the weld seam direction. Different time of flight measurements were conducted by varying the distance between the weld seam and Rayleigh wave propagating path in the range of 0 to 45 mm. The maximum relative velocity difference was found on the weld seam. With the increasing distance away from the weld seam, the relative velocity difference sharply decreased to negative value. With further increase in distance, the relative velocity difference slowly increased and approached zero. The distribution of relative velocity variations indicates that tensile stress appears in the melted zone as it becomes compressive near the heat-affected zone.

  1. Welding in nuclear engineering

    International Nuclear Information System (INIS)

    1982-01-01

    The 3rd international conference 'Welding in nuclear engineering', organized in 1978 by the Deutscher Verband fuer Schweisstechnik e.V., was, like the two foregoing conferences in 1970 and 1974, an absolute success. The noteworthy echo to this meeting in the international technical world - the number of 650 participants from 26 countries is self-evidence - and this fact, was for the Deutscher Verband fuer Schweisstechnik e.V. occasion and at the same time an obligation now to follow in the same way, the meeting that was started 12 years ago, by organizing the international conference 'Welding in nuclear engineering'. The conference this year offers in addition to the two plenary session lectures, 34 short reports and a further 28 single contributions in the form of two poster-sessions. Unfortunately, it was again not possible to accept all the papers submitted because the conference was limited to 2 days only. Nevertheless, the papers will offer a representative cross-section through the total range of welding engineering. In particular, the poster session, which take place for the first time within the scope of a meeting organized by the Working Group 'Welding in Nuclear Engineering', should contribute to the aim that this time again the discussions will form the main point of the conference. (orig./RW) [de

  2. Welding. Student Learning Guide.

    Science.gov (United States)

    Palm Beach County Board of Public Instruction, West Palm Beach, FL.

    This student learning guide contains 30 modules for completing a course in welding. It is designed especially for use in secondary schools in Palm Beach County, Florida. Each module covers one task, and consists of a purpose, performance objective, enabling objectives, learning activities keyed to resources, information sheets, student self-check…

  3. Thermal Stresses in Welding

    DEFF Research Database (Denmark)

    Hansen, Jan Langkjær

    1998-01-01

    Studies of the transient temperature fields and the hereby induced deformations and stressses in a butt-welded mild steel plate modelledrespectively in 2D plane stress state (as well as plane strain state) and in full 3D have been done. The model has been implemented in the generalpurpose FE...

  4. State Skill Standards: Welding

    Science.gov (United States)

    Pointer, Mike; Naylor, Randy; Warden, John; Senek, Gene; Shirley, Charles; Lefcourt, Lew; Munson, Justin; Johnson, Art

    2005-01-01

    The Department of Education has undertaken an ambitious effort to develop statewide occupational skill standards. The standards in this document are for welding programs and are designed to clearly state what the student should know and be able to do upon completion of an advanced high-school program. The writing team determined that any statewide…

  5. Elementary TIG Welding Skills.

    Science.gov (United States)

    Pierson, John E., III

    The text was prepared to help deaf students develop the skills needed by an employed welder. It uses simplified language and illustrations to present concepts which should be reinforced by practical experience with welding skills. Each of the 12 lessons contains: (1) an information section with many illustrations which presents a concept or…

  6. Vibration stress relief treatment in welded samples of ST-3 steel

    International Nuclear Information System (INIS)

    Suarez, J.C.; Fernandez, L.M.; Echevarria, J.F.; Estevez, A.; Perez, A.; Aragon, B.

    1996-01-01

    The presented work is aimed to find the optimal vibration frequency and treatment duration for ST-3 steel welded test pieces. In the experiment transversal stresses were not virtually relieved by the application of vibrations at the three natural frecuencies. With regard to the optimal frequency for our system, the firths natural frequency appears to be most effective one, wherewith a maximum 35-70 % longitudinal stress relief was obtained. The influence of the propagation direction (transversal or longitudinal) of vibrations on stress relief in a welded joint was confirmed

  7. Welding, Bonding and Fastening, 1984

    Science.gov (United States)

    Buckley, J. D. (Editor); Stein, B. A. (Editor)

    1985-01-01

    A compilation of papers presented in a joint NASA, American Society for Metals, The George Washington University, American Welding Soceity, and Society of Manufacturing Engineers conference on Welding, Bonding, and Fastening at Langley Research Center, Hampton, VA, on October 23 to 25, 1984 is given. Papers were presented on technology developed in current research programs relevant to welding, bonding, and fastening of structural materials required in fabricating structures and mechanical systems used in the aerospace, hydrospace, and automotive industries. Topics covered in the conference included equipment, hardware and materials used when welding, brazing, and soldering, mechanical fastening, explosive welding, use of unique selected joining techniques, adhesives bonding, and nondestructive evaluation. A concept of the factory of the future was presented, followed by advanced welding techniques, automated equipment for welding, welding in a cryogenic atmosphere, blind fastening, stress corrosion resistant fasteners, fastening equipment, explosive welding of different configurations and materials, solid-state bonding, electron beam welding, new adhesives, effects of cryogenics on adhesives, and new techniques and equipment for adhesive bonding.

  8. Galvanic corrosion of beryllium welds

    International Nuclear Information System (INIS)

    Hill, M.A.; Butt, D.P.; Lillard, R.S.

    1997-01-01

    Beryllium is difficult to weld because it is highly susceptible to cracking. The most commonly used filler metal in beryllium welds is Al-12 wt.% Si. Beryllium has been successfully welded using Al-Si filler metal with more than 30 wt.% Al. This filler creates an aluminum-rich fusion zone with a low melting point that tends to backfill cracks. Drawbacks to adding a filler metal include a reduction in service temperature, a lowering of the tensile strength of the weld, and the possibility for galvanic corrosion to occur at the weld. To evaluate the degree of interaction between Be and Al-Si in an actual weld, sections from a mock beryllium weldment were exposed to 0.1 M Cl - solution. Results indicate that the galvanic couple between Be and the Al-Si weld material results in the cathodic protection of the weld and of the anodic dissolution of the bulk Be material. While the cathodic protection of Al is generally inefficient, the high anodic dissolution rate of the bulk Be during pitting corrosion combined with the insulating properties of the Be oxide afford some protection of the Al-Si weld material. Although dissolution of the Be precipitate in the weld material does occur, no corrosion of the Al-Si matrix was observed

  9. In-field Welding and Coating Protocols

    Science.gov (United States)

    2009-05-12

    Gas Technology Institute (GTI) and Edison Welding Institute (EWI) created both laboratory and infield girth weld samples to evaluate the effects of weld geometry and hydrogen off-gassing on the performance of protective coatings. Laboratory made plat...

  10. Perspectives of special welding methods. 1

    International Nuclear Information System (INIS)

    Herden, G.; Buness, G.; Wiesner, P.

    1976-01-01

    Laser, electron, ion, and light beam welding as well as plasma arc welding are considered to be special fusion welding methods. The stage of development and possible future applications of these methods are described. (author)

  11. Improvements in and relating to welding

    International Nuclear Information System (INIS)

    Taylor, B.D.

    1979-01-01

    This invention concerns apparatus for use in welding, particularly welding which must be effected in a predetermined, for example, inert atmosphere, e.g. the welding of reactive materials such as zircaloy, titanium, magnesium, aluminium, etc. (U.K.)

  12. Welding for the CRBRP steam generators

    International Nuclear Information System (INIS)

    Spalaris, C.N.; Ring, P.J.; Durand, R.E.; Wright, E.A.

    1979-01-01

    The rationale for selecting weld design, welding procedures and inspection methods was based upon the desire to obtain the highest reliability welds for the CRBRP steam generators. To assure the highest weld reliability, heavy emphasis was placed on the control of material cleanliness and composition substantially exceeding the requirements of the ASME Code for 2-1/4Cr--1Mo. The high tube/tubesheet weld quality was achieved through close material control, an extensive weld development program and the selection of high reliability welding equipment. Shell and nozzle weld fabrication using TIG, MIG, and submerged arc procedures are also being controlled through precise specifications, including preheat and postheat programs, together with radiography and ultrasonic inspection to ascertain the weld quality desired. Details of the tube/tubesheet welding and shell welding are described and results from the weld testing program are discussed

  13. Development of carbon/carbon composite control rod for HTTR. 1. Preparation of elements and their fracture tests

    International Nuclear Information System (INIS)

    Eto, Motokuni; Ishiyama, Shintaro; Ugachi, Hirokazu

    1996-08-01

    For the High Temperature Engineering Test Reactor(HTTR) the control rod sleeve is made of Alloy 800H for which a particular process is imposed when the reactor needs to be scrammed. The less restricted operation of the reactor would be attained if there would be the control rod more resistant to high temperature and neutron irradiation. This report summarizes the results which have been obtained as of March 1996 in the course of the development of the C/C composite control rod. Materials used were pitch- or PAN-based fiber-reinforced 2-dimensional carbon composites, from which preforms of the elements of a control rod were fabricated. The preforms were carbonized at 1000degC after being impregnated with pitch. Then they were graphitized at 3000degC, followed by a purification treatment with halogen. The elements included the pellet holder, lace truck and pin. The pin was fabricated by the fiber laminating technique. A control rod is to consist of pellet holders which are connected by the lace trucks with pins. Various strength tests were carried out on these elements. An irradiation of the elements made of PAN-based material was performed in JRR-3 at 900±50degC to a neutron fluence of 1x10 25 n/m 2 (E>29fJ). As for the strength tests on the elements, there were some differences between PAN- and pitch-based composites: In general, elements made of PAN-based composite showed the more plastic behavior before they fractured, whereas those of pitch-based material behaved in the more brittle manner. Fracture tests of the irradiated elements showed that fracture load and fracture displacement enough for assuring the integrity of the control rod structure were maintained even after the irradiation. It was also found that if the applied load was parallel to the fiber felt plane both fracture load and strain increased, whereas the load increase and strain decrease were observed for the applied load against the plane. (J.P.N.)

  14. Development of carbon/carbon composite control rod for HTTR. 1. Preparation of elements and their fracture tests

    Energy Technology Data Exchange (ETDEWEB)

    Eto, Motokuni; Ishiyama, Shintaro; Ugachi, Hirokazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-08-01

    For the High Temperature Engineering Test Reactor(HTTR) the control rod sleeve is made of Alloy 800H for which a particular process is imposed when the reactor needs to be scrammed. The less restricted operation of the reactor would be attained if there would be the control rod more resistant to high temperature and neutron irradiation. This report summarizes the results which have been obtained as of March 1996 in the course of the development of the C/C composite control rod. Materials used were pitch- or PAN-based fiber-reinforced 2-dimensional carbon composites, from which preforms of the elements of a control rod were fabricated. The preforms were carbonized at 1000degC after being impregnated with pitch. Then they were graphitized at 3000degC, followed by a purification treatment with halogen. The elements included the pellet holder, lace truck and pin. The pin was fabricated by the fiber laminating technique. A control rod is to consist of pellet holders which are connected by the lace trucks with pins. Various strength tests were carried out on these elements. An irradiation of the elements made of PAN-based material was performed in JRR-3 at 900{+-}50degC to a neutron fluence of 1x10{sup 25} n/m{sup 2} (E>29fJ). As for the strength tests on the elements, there were some differences between PAN- and pitch-based composites: In general, elements made of PAN-based composite showed the more plastic behavior before they fractured, whereas those of pitch-based material behaved in the more brittle manner. Fracture tests of the irradiated elements showed that fracture load and fracture displacement enough for assuring the integrity of the control rod structure were maintained even after the irradiation. It was also found that if the applied load was parallel to the fiber felt plane both fracture load and strain increased, whereas the load increase and strain decrease were observed for the applied load against the plane. (J.P.N.)

  15. Developments in welding and joining methods of metallic materials

    International Nuclear Information System (INIS)

    Pilarczyk, J.

    2007-01-01

    The impact of the welding technology on the economy development. The welding and joining methods review. The particular role of the laser welding and its interesting applications: with filler metal, twin spot laser welding, hybrid welding process, remote welding. The fiber lasers. The high intensity electron beams applications for surface modification. The TIG welding with the use of the active flux. Friction welding, friction stir welding and friction linear welding. (author)

  16. Welding wire pressure sensor assembly

    Science.gov (United States)

    Morris, Timothy B. (Inventor); Milly, Peter F., Sr. (Inventor); White, J. Kevin (Inventor)

    1994-01-01

    The present invention relates to a device which is used to monitor the position of a filler wire relative to a base material being welded as the filler wire is added to a welding pool. The device is applicable to automated welding systems wherein nonconsumable electrode arc welding processes are utilized in conjunction with a filler wire which is added to a weld pool created by the electrode arc. The invention senses pressure deviations from a predetermined pressure between the filler wire and the base material, and provides electrical signals responsive to the deviations for actuating control mechanisms in an automatic welding apparatus so as to minimize the pressure deviation and to prevent disengagement of the contact between the filler wire and the base material.

  17. Fundamental Laser Welding Process Investigations

    DEFF Research Database (Denmark)

    Bagger, Claus; Olsen, Flemming Ove

    1998-01-01

    In a number of systematic laboratory investigations the fundamental behavior of the laser welding process was analyzed by the use of normal video (30 Hz), high speed video (100 and 400 Hz) and photo diodes. Sensors were positioned to monitor the welding process from both the top side and the rear...... side of the specimen.Special attention has been given to the dynamic nature of the laser welding process, especially during unstable welding conditions. In one series of experiments, the stability of the process has been varied by changing the gap distance in lap welding. In another series...... video pictures (400 Hz), a clear impact on the seam characteristics has been identified when a hump occurs.Finally, a clear correlation between the position of the focus point, the resultant process type and the corresponding signal intensity and signal variation has been found for sheets welded...

  18. INERT GAS SHIELD FOR WELDING

    Science.gov (United States)

    Jones, S.O.; Daly, F.V.

    1958-10-14

    S>An inert gas shield is presented for arc-welding materials such as zirconium that tend to oxidize rapidly in air. The device comprises a rectangular metal box into which the welding electrode is introduced through a rubber diaphragm to provide flexibility. The front of the box is provided with a wlndow having a small hole through which flller metal is introduced. The box is supplied with an inert gas to exclude the atmosphere, and with cooling water to promote the solidification of the weld while in tbe inert atmosphere. A separate water-cooled copper backing bar is provided underneath the joint to be welded to contain the melt-through at the root of the joint, shielding the root of the joint with its own supply of inert gas and cooling the deposited weld metal. This device facilitates the welding of large workpieces of zirconium frequently encountered in reactor construction.

  19. Metals welding by using laser

    International Nuclear Information System (INIS)

    Al-Qaisy, R.A.W.

    1991-01-01

    In the present work, same welding ''conduction limited type'' under atmospheric conditions was performed using pulsed Ng:YAG laser to weld; low carbon steel (LCS), stainless steel (304) (SUS304), stainless steel (303) (SUS303), and brass. Microstructure of welded zone, heat affected zone (HAZ), and the laser energy on penetration depth and effective diameter were studied. Tensile test, micro-hardness, and surface roughness of welded and parent metals were also dealt with. Melting efficiency was worked out and an under vacuum seam welding of low carbon steel has been accomplished. Finally spot welding of aluminium tungsten, and platinium wires were employed using different layer energies. 34 tabs.; 82 figs.; 51 refs.; 1 app

  20. Corrosion of carbon steel welds

    International Nuclear Information System (INIS)

    Daniel, B.

    1988-09-01

    This report assesses the factors which cause preferential attack to occur in carbon steel fusion welds. It was concluded that the main factors were: the inclusion content of the weld metal, the potential of the weld metal being less noble than that of the parent, and the presence of low-temperature transformation products in the heat-affected zone of the weld. These factors should be minimized or eliminated as appropriate so that the corrosion allowances determined for carbon steel waste drums is also adequate for the welds. An experimental/theoretical approach is recommended to evaluate the relative corrosion resistance of welds prepared from BS 4360 grade 43A steel to that of the parent material. (author)

  1. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  2. Investigation on mechanical properties of welded material under different types of welding filler (shielded metal arc welding)

    Science.gov (United States)

    Tahir, Abdullah Mohd; Lair, Noor Ajian Mohd; Wei, Foo Jun

    2018-05-01

    The Shielded Metal Arc Welding (SMAW) is (or the Stick welding) defined as a welding process, which melts and joins metals with an arc between a welding filler (electrode rod) and the workpieces. The main objective was to study the mechanical properties of welded metal under different types of welding fillers and current for SMAW. This project utilized the Design of Experiment (DOE) by adopting the Full Factorial Design. The independent variables were the types of welding filler and welding current, whereas the other welding parameters were fixed at the optimum value. The levels for types of welding filler were by the models of welding filler (E6013, E7016 and E7018) used and the levels for welding current were 80A and 90A. The responses were the mechanical properties of welded material, which include tensile strength and hardness. The experiment was analyzed using the two way ANOVA. The results prove that there are significant effects of welding filler types and current levels on the tensile strength and hardness of the welded metal. At the same time, the ANOVA results and interaction plot indicate that there are significant interactions between the welding filler types and the welding current on both the hardness and tensile strength of the welded metals, which has never been reported before. This project found that when the amount of heat input with increase, the mechanical properties such as tensile strength and hardness decrease. The optimum tensile strength for welded metal is produced by the welding filler E7016 and the optimum of hardness of welded metal is produced by the welding filler E7018 at welding current of 80A.

  3. Viewing Welds By Computer Tomography

    Science.gov (United States)

    Pascua, Antonio G.; Roy, Jagatjit

    1990-01-01

    Computer tomography system used to inspect welds for root penetration. Source illuminates rotating welded part with fan-shaped beam of x rays or gamma rays. Detectors in circular array on opposite side of part intercept beam and convert it into electrical signals. Computer processes signals into image of cross section of weld. Image displayed on video monitor. System offers only nondestructive way to check penetration from outside when inner surfaces inaccessible.

  4. Welding facilities for NPP assembling

    International Nuclear Information System (INIS)

    Rojtenberg, S.S.

    1987-01-01

    Recommendations concerning the choice of equipment for welding in pre-assembling work shops, in the enlarging assembling shops and at the assembling site, are given. Advanced production automatic welders and semiautomatic machines, applied during the NPP equipment assembling as well as automatic machines specially produced for welding the main reactor components and pipelines are described. Automatic and semiautomatic machine and manual welding post supply sources are considered

  5. Residual stress reduction in beam welded joints by means of stress redistribution using defocused electron or laser beams; Eigenspannungsreduktion in strahlgeschweissten Naehten mittels Spannungsumlagerung durch den Einsatz defokussierter Elektronen- bzw. Laserstrahlen

    Energy Technology Data Exchange (ETDEWEB)

    Toelle, Florian

    2013-08-01

    Among the multiple advantages of beam welding processes the high longitudinal residual stresses in beam welds ranging till the local yield stress are one disadvantage. These high stresses can influence the service life of the welded components. The residual stresses in other welding processes exist in an equal high level but primarily in the transverse direction to the weld. To mitigate the high residual stresses a couple of methods were developed for these welding processes in the last decades. However these methods need large contact surfaces next to the welds for the installation of matched heating and cooling elements and other additional equipment. Furthermore, the previous developed stress mitigating processes offer a low efficiency for the small beam welds. The stress reduction by using the welding source after the welding process for a remote heat treatment of the welded components afford a flexible tool for the stress mitigation in beam welds. This method does not need any additional equipment and it is applicable for complex welding and component geometries. During this post welding heat treatment the material next to the weld is heated by the defocused electron or by the defocused laser beam, respectively, to temperatures of some hundreds degree Celsius. Hereby low plastic deformations in these regions are generated. While cooling down due to the thermal shrinkage the material between the weld and the heat treated region is compressed in longitudinal direction to the weld. This intermediate material zone constrained the shrinkage of the weld while cooling down from the melting temperature and leads to the high longitudinal residual stresses in the weld. In consequence of the compression of this intermediate zones by the heat treated zones the resistance to the shrinkage of the weld is lowered and the longitudinal stresses in the weld are reduced. In the process the quantity of the stress reduction is controlled by the selection of the process parameters

  6. Socket welds in nuclear facilities

    International Nuclear Information System (INIS)

    Anderson, P.A.; Torres, L.L.

    1995-01-01

    Socket welds are easier and faster to make than are butt welds. However, they are often not used in nuclear facilities because the crevices between the pipes and the socket sleeves may be subject to crevice corrosion. If socket welds can be qualified for wider use in facilities that process nuclear materials, the radiation exposures to welders can be significantly reduced. The current tests at the Idaho Chemical Processing Plant (ICPP) are designed to determine if socket welds can be qualified for use in the waste processing system at a nuclear fuel processing plant

  7. Programmable Automated Welding System (PAWS)

    Science.gov (United States)

    Kline, Martin D.

    1994-01-01

    An ambitious project to develop an advanced, automated welding system is being funded as part of the Navy Joining Center with Babcock & Wilcox as the prime integrator. This program, the Programmable Automated Welding System (PAWS), involves the integration of both planning and real-time control activities. Planning functions include the development of a graphical decision support system within a standard, portable environment. Real-time control functions include the development of a modular, intelligent, real-time control system and the integration of a number of welding process sensors. This paper presents each of these components of the PAWS and discusses how they can be utilized to automate the welding operation.

  8. Comparison of Welding Residual Stresses of Hybrid Laser-Arc Welding and Submerged Arc Welding in Offshore Steel Structures

    DEFF Research Database (Denmark)

    Andreassen, Michael Joachim; Yu, Zhenzhen; Liu, Stephen

    2016-01-01

    In the offshore industry, welding-induced distortion and tensile residual stresses have become a major concern in relation to the structural integrity of a welded structure. Particularly, the continuous increase in size of welded plates and joints needs special attention concerning welding induced...... residual stresses. These stresses have a negative impact on the integrity of the welded joint as they promote distortion, reduce fatigue life, and contribute to corrosion cracking and premature failure in the weld components. This paper deals with the influence and impact of welding method on the welding...... induced residual stresses. It is also investigated whether the assumption of residual stresses up to yield strength magnitude are present in welded structures as stated in the design guidelines. The fatigue strength for welded joints is based on this assumption. The two welding methods investigated...

  9. Steels and welding nuclear

    International Nuclear Information System (INIS)

    Sessa, M.; Milella, P.P.

    1987-01-01

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

  10. Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR. Contract research

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji; Tochio, Daisuke; Ohwada, Hiroyuki

    2005-03-01

    Safety demonstration tests using the HTTR are to be conducted from the FY2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR that is one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3, S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was verified that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram. (author)

  11. Verification of in-core thermal and hydraulic analysis code FLOWNET/TRUMP for the high temperature engineering test reactor (HTTR) at JAERI

    International Nuclear Information System (INIS)

    Maruyama, Soh; Sudo, Yukio; Saito, Shinzo; Kiso, Yoshihiro; Hayakawa, Hitoshi

    1989-01-01

    The FLOWNET/TRUMP code consists of a flow network analysis code 'FLOWNET' for calculations of coolant flow distribution and coolant temperature distribution in the core with a thermal conduction analysis code 'TRUMP' for calculation of temperature distribution in solid structures. The verification of FLOWNET/TRUMP was made by the comparison of the analytical results with the results of steady state experiments by the HENDEL multichannel test rig, T1-M, which consisted of twelve simulated fuel rods heated electrically and eleven hexagonal graphite fuel blocks. The T1-M simulated the one fuel column in the core. The analytical results agreed well with the results of the experiment in which the HTTR operating conditions were simulated. (orig.)

  12. Improvement works report on mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system (Contract research)

    International Nuclear Information System (INIS)

    Sakaki, Akihiro; Kato, Michio; Hayashi, Koji; Fujisaki, Katsuo; Aita, Hideki; Ohashi, Hirofumi; Takada, Shoji; Shimizu, Akira; Morisaki, Norihiro; Maeda, Yukimasa; Sato, Hiroyuki; Hanawa, Hiromi; Yonekawa, Hideo; Inagaki, Yoshiyuki

    2005-04-01

    In order to establish the system integration technology to connect a hydrogen production system to a high temperature gas cooled reactor; the mock-up test facility with a full-scale reaction tube for the steam reforming HTTR hydrogen production system was constructed in fiscal year 2001 and its functional test operation was performed in the year. Seven experimental test operations were performed from fiscal year 2001 to 2004. On a period of each test operation, there happened some troubles. For each trouble, the cause was investigated and the countermeasures and the improvement works were performed to succeed the experiments. The tests were successfully achieved according to plan. This report describes the improvement works on the test facility performed from fiscal year 2001 to 2004. (author)

  13. Gas Metal Arc Welding. Welding Module 5. Instructor's Guide.

    Science.gov (United States)

    Missouri Univ., Columbia. Instructional Materials Lab.

    This guide is intended to assist vocational educators in teaching an eight-unit module in gas metal arc welding. The module is part of a welding curriculum that has been designed to be totally integrated with Missouri's Vocational Instruction Management System. The following topics are covered in the module: safety and testing, gas metal arc…

  14. Laser Welding Test Results with Gas Atmospheres in Welding Chamber

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seo-Yun; Yang, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The weld beads of specimens welded under identical conditions in the helium and argon gas were cleaner, more regular, and steadier than those in a vacuum. The penetration depth of the FZ in the vacuum was much deeper than those in the helium and argon gas. To measure the irradiation properties of nuclear fuel in a test reactor, a nuclear fuel test rod instrumented with various sensors must be fabricated with assembly processes. A laser welding system to assemble the nuclear fuel test rod was designed and fabricated to develop various welding technologies of the fuel test rods to joint between a cladding tube and end-caps. It is an air-cooling optical fiber type and its emission modes are a continuous (CW) mode of which the laser generates continuous emission, and pulse (QCW) mode in which the laser internally generates sequences of pulses. We considered the system welding a sample in a chamber that can weld a specimen in a vacuum and inert gas atmosphere, and the chamber was installed on the working plate of the laser welding system. In the chamber, the laser welding process should be conducted to have no defects on the sealing area between a cladding tube and an end-cap.

  15. Sustainability of Welding Process through Bobbin Friction Stir Welding

    Science.gov (United States)

    Sued, M. K.; Samsuri, S. S. M.; Kassim, M. K. A. M.; Nasir, S. N. N. M.

    2018-03-01

    Welding process is in high demand, which required a competitive technology to be adopted. This is important for sustaining the needs of the joining industries without ignoring the impact of the process to the environment. Friction stir welding (FSW) is stated to be benefitting the environment through low energy consumption, which cannot be achieved through traditional arc welding. However, this is not well documented, especially for bobbin friction stir welding (BFSW). Therefore, an investigation is conducted by measuring current consumption of the machine during the BFSW process. From the measurement, different phases of BFSW welding process and its electrical demand are presented. It is found that in general total energy in BFSW is about 130kW inclusive of all identified process phases. The phase that utilise for joint formation is in weld phase that used the highest total energy of 120kWs. The recorded total energy is still far below the traditional welding technology and the conventional friction stir welding (CFSW) energy demand. This indicates that BFSW technology with its vast benefit able to sustain the joining technology in near future.

  16. Performance test results of helium gas circulator of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Akira; Kato, Michio; Hayashi, Koji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2003-03-01

    Hydrogen production system by steam reforming of methane will be connected to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) against development of nuclear heat utilization system. To obtain design and safety database of the HTTR hydrogen production system, mock-up test facility with full-scale reaction was constructed in FY 2001 and hydrogen of 120m{sup 3}N{sub /}h was successfully produced in overall performance test. This report describes performance test results of a helium gas circulator in this facility. The circulator performance curves regarding to pressure-rise, input power and adiabatic thermal efficiency at standard revolution number were made based on the measured flow-rate, temperature and pressure data in overall performance test. The circulator performance prediction code was made based on these performance curves. The code can calculate revolution number, electric power and temperature-rise of the circulator using flow-rate, inlet temperature, inlet pressure and pressure-rise data. The verification of the code was carried out with the test data in FY 2002. Total pressure loss of the helium gas circulation loop was also evaluated. The circulator should be operated in conditions such as pressure from 2.7MPa to 4.0MPa and flow-rate from 250g/s to 400g/s and at maximum pressure-rise of 250 kPa in test operation. It was confirmed in above verification and evaluations that the circulator had performance to satisfy above conditions within operation limitation of the circulator such as maximum input-power of 150 kW and maximum revolution number of 12,000 rpm. (author)

  17. Tensometry technique for X-ray diffraction in applied analysis of welding

    International Nuclear Information System (INIS)

    Turibus, S.N.; Caldas, F.C.M.; Miranda, D.M.; Monine, V.I.; Assis, J.T.

    2010-01-01

    This paper presents the analysis of residual stress introduced in welding process. As the stress in a material can induce damages, it is necessary to have a method to identify this residual stress state. For this it was used the non-destructive X-ray diffraction technique to analyze two plates from A36 steel jointed by metal inert gas (MIG) welding. The stress measurements were made by the sin 2 ψ method in weld region of steel plates including analysis of longitudinal and transverse residual stresses in fusion zone, heat affected zone (HAZ) and base metal. To determine the stress distribution along the depth of the welded material it was used removing of superficial layers made by electropolishing. (author)

  18. [New welding processes and health effects of welding].

    Science.gov (United States)

    La Vecchia, G Marina; Maestrelli, Piero

    2011-01-01

    This paper describes some of the recent developments in the control technology to enhance capability of Pulse Gas Metal Arc Welding. Friction Stir Welding (FSW) processing has been also considered. FSW is a new solid-state joining technique. Heat generated by friction at the rotating tool softens the material being welded. FSW can be considered a green and energy-efficient technique without deleterious fumes, gas, radiation, and noise. Application of new welding processes is limited and studies on health effects in exposed workers are lacking. Acute and chronic health effects of conventional welding have been described. Metal fume fever and cross-shift decline of lung function are the main acute respiratory effects. Skin and eyes may be affected by heat, electricity and UV radiations. Chronic effects on respiratory system include chronic bronchitis, a benign pneumoconiosis (siderosis), asthma, and a possible increase in the incidence of lung cancer. Pulmonary infections are increased in terms of severity, duration, and frequency among welders.

  19. Fundamental Study of Electron Beam Welding of AA6061-T6 Aluminum Alloy for Nuclear Fuel Plate Assembly (II)

    International Nuclear Information System (INIS)

    Kim, Soosung; Lee, Haein; Lee, Donbae; Park, Jongman; Lee, Yoonsang

    2013-01-01

    Certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes posses the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the nuclear fuel plate fabrication and assembly, a fundamental EBW experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the welding process, and satisfy the requirements of the weld quality, EBW apparatus using a electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. In this experiment, a feasibility test was carried out by tensile tester, bead-on-plate welding and metallographic examination to comply with the aluminum welding procedure. The EB weld quality of AA6061-T6 aluminum alloy for the fuel plate assembly has been also studied by the mechanical testing and microstructure examinations. This study was carried out to determine the suitable welding process and to investigate tensile strength of AA6061-T6 aluminum alloy. In the present experiment, satisfactory EBW of the square butt weld specimens was developed. In comparison with the rolling directions of test specimens, the tensile strengths were no difference between the longitudinal and transverse welds. Based on this fundamental study, fabrication and assembly of the nuclear fuel plates will be provided for the future Kijang research reactor project

  20. Characterization of Mechanical Properties and Residual Stress in API 5L X80 Steel Welded Joints

    Science.gov (United States)

    de Sousa Lins, Amilton; de Souza, Luís Felipe Guimarães; Fonseca, Maria Cindra

    2018-01-01

    The use of high-strength and low-alloy steels, high design factors and increasingly stringent safety requirements have increased the operating pressure levels and, consequently, the need for further studies to avoid and prevent premature pipe failure. To evaluate the possibility of improving productivity in manual arc welding of this type of steel, this work characterizes the mechanical properties and residual stresses in API 5L X80 steel welded joints using the SMAW and FCAW processes. The residual stresses were analyzed using x-ray diffraction with the sin2 ψ method at the top and root of the welded joints in the longitudinal and transverse directions of the weld bead. The mechanical properties of the welded joints by both processes were characterized in terms of tensile strength, impact toughness and Vickers microhardness in the welded and shot peening conditions. A predominantly compressive residual stress was found, and shot peening increased the tensile strength and impact toughness in both welded joints.

  1. A control system for uniform bead in fillet arc welding on tack welds

    International Nuclear Information System (INIS)

    Kim, Jae Woong; Lee, Jun Young

    2008-01-01

    Positioning a workpiece accurately and preventing weld distortion, tack welding is often adopted before main welding in the construction of welded structures. However, this tack weld deteriorates the final weld bead profile, so that the grinding process is usually performed for a uniform weld bead profile. In this study, a control system for uniform weld bead is proposed for the fillet arc welding on tack welds. The system consists of GMA welding machine, torch manipulator, laser vision sensor for measuring the tack weld size and the database for optimal welding conditions. Experiments have been performed for constructing the database and for evaluating the control capability of the system. It has been shown that the system has the capability to smooth the bead at the high level of quality

  2. Numerical investigation on the variation of welding stresses after material removal from a thick titanium alloy plate joined by electron beam welding

    International Nuclear Information System (INIS)

    Liu, Chuan; Zhang, Jianxun; Wu, Bing; Gong, Shuili

    2012-01-01

    Highlights: → After less materials removal from the top, stresses on the bottom remain unchanged. → The transverse stress within the weld decreases significantly with material removal. → Local material removal does not influence the longitudinal stress significantly. -- Abstract: The stress modification after material removal from a 50 mm thick titanium alloy plate jointed by electron beam welding (EBW) was investigated through the finite element method (FEM). The welding experiment and milling process were carried out to experimentally determine the stresses induced by EBW and their modification after local material removal. The modification of as-welded stresses due to the local material removal method and the whole layer removal method was discussed with the finite element analysis. Investigated results showed that with less materials removal from the top, the stresses on the bottom surface remain almost unchanged; after material removal from the top and bottom part, the transverse stress on the newly-formed surface decreases significantly as compared to the as-welded stresses at the same locations; however, the stress modification only occurs at the material removal region in the case of local region removal method; the longitudinal stress decreases with the whole layer removal method while remains almost unchanged with the local region removal method.

  3. Effects of Fusion Tack Welds on Self-Reacting Friction Stir Welds

    Science.gov (United States)

    Nunes, A. C., Jr.; Pendleton, M. L.; Brooke, S. A.; Russell, C. K.

    2012-01-01

    In order to know whether fusion tack welds would affect the strength of self-reacting friction stir seam welds in 2195-T87 aluminum alloy, the fracture stresses of 144 tensile test coupons cut from 24 welded panels containing segments of friction stir welds were measured. Each of the panels was welded under unique processing conditions. A measure of the effect of the tack welds for each panel was devised. An analysis of the measures of the tack weld effect supported the hypothesis that fusion tack welds do not affect the strength of self-reacting friction stir welds to a 5% level of confidence.

  4. Friction Welding For Cladding Applications: Processing, Microstructure and Mechanical Properties of Inertia Friction Welds of Stainless Steel to Low Carbon Steel and Evaluation of Wrought and Welded Austenitic Stainless Steels for Cladding Applications in Acidchloride Service

    Science.gov (United States)

    Switzner, Nathan

    Friction welding, a solid-state joining method, is presented as a novel alternative process step for lining mild steel pipe and forged components internally with a corrosion resistant (CR) metal alloy for petrochemical applications. Currently, fusion welding is commonly used for stainless steel overlay cladding, but this method is costly, time-consuming, and can lead to disbonding in service due to a hard martensite layer that forms at the interface due to partial mixing at the interface between the stainless steel CR metal and the mild steel base. Firstly, the process parameter space was explored for inertia friction butt welding using AISI type 304L stainless steel and AISI 1018 steel to determine the microstructure and mechanical properties effects. A conceptual model for heat flux density versus radial location at the faying surface was developed with consideration for non-uniform pressure distribution due to frictional forces. An existing 1 D analytical model for longitudinal transient temperature distribution was modified for the dissimilar metals case and to account for material lost to the flash. Microstructural results from the experimental dissimilar friction welds of 304L stainless steel to 1018 steel were used to discuss model validity. Secondly, the microstructure and mechanical property implications were considered for replacing the current fusion weld cladding processes with friction welding. The nominal friction weld exhibited a smaller heat softened zone in the 1018 steel than the fusion cladding. As determined by longitudinal tensile tests across the bond line, the nominal friction weld had higher strength, but lower apparent ductility, than the fusion welds due to the geometric requirements for neck formation adjacent to a rigid interface. Martensite was identified at the dissimilar friction weld interface, but the thickness was smaller than that of the fusion welds, and the morphology was discontinuous due to formation by a mechanism of solid

  5. Residual stress reduction in the penetration nozzle weld joint by overlay welding

    International Nuclear Information System (INIS)

    Jiang, Wenchun; Luo, Yun; Wang, B.Y.; Tu, S.T.; Gong, J.M.

    2014-01-01

    Highlights: • Residual stress reduction in penetration weld nozzle by overlay welding was studied. • The overlay weld can decrease the residual stress in the weld root. • Long overlay welding is proposed in the actual welding. • Overlay weld to decrease residual stress is more suitable for thin nozzle. - Abstract: Stress corrosion cracking (SCC) in the penetration nozzle weld joint endangers the structural reliability of pressure vessels in nuclear and chemical industries. How to decrease the residual stress is very critical to ensure the structure integrity. In this paper, a new method, which uses overlay welding on the inner surface of nozzle, is proposed to decrease the residual stresses in the penetration joint. Finite element simulation is used to study the change of weld residual stresses before and after overlay welding. It reveals that this method can mainly decrease the residual stress in the weld root. Before overlay welding, large tensile residual stresses are generated in the weld root. After overlay weld, the tensile hoop stress in weld root has been decreased about 45%, and the radial stress has been decreased to compressive stress, which is helpful to decrease the susceptibility to SCC. With the increase of overlay welding length, the residual stress in weld root has been greatly decreased, and thus the long overlay welding is proposed in the actual welding. It also finds that this method is more suitable for thin nozzle rather than thick nozzle

  6. Ultrasonic inspection of austenitic welds

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  7. Metal Working and Welding Operations.

    Science.gov (United States)

    Marine Corps Inst., Washington, DC.

    This student guide, one of a series of correspondence training courses designed to improve the job performance of members of the Marine Corps, deals with the skills needed by metal workers and welders. Addressed in the six individual units of the course are the following topics: weldable metals and their alloys, arc welding, gas welding,…

  8. Welding abilities of UFG metals

    Science.gov (United States)

    Morawiński, Łukasz; Chmielewski, Tomasz; Olejnik, Lech; Buffa, Gianluca; Campanella, Davide; Fratini, Livan

    2018-05-01

    Ultrafine Grained (UFG) metals are characterized by an average grain size of welded joints with similar properties to the base of UFG material are crucial for the production of finished engineering components. Conventional welding methods based on local melting of the joined edges cannot be used due to the UFG microstructure degradation caused by the heat occurrence in the heat affected zone. Therefore, the possibility of obtaining UFG materials joints with different shearing plane (SP) positions by means of friction welded processes, which do not exceed the melting temperature during the process, should be investigated. The article focuses on the Linear Friction Welding (LFW) method, which belongs to innovative welding processes based on mixing of the friction-heated material in the solid state. LFW is a welding process used to joint bulk components. In the process, the friction forces work due to the high frequency oscillation and the pressure between the specimens is converted in thermal energy. Character and range of recrystallization can be controlled by changing LFW parameters. Experimental study on the welded UFG 1070 aluminum alloy by means of FLW method, indicates the possibility of reducing the UFG structure degradation in the obtained joint. A laboratory designed LFW machine has been used to weld the specimens with different contact pressure and oscillation frequency.

  9. Arc modeling for welding analysis

    International Nuclear Information System (INIS)

    Glickstein, S.S.

    1978-04-01

    A one-dimensional model of the welding arc that considers heat generation by the Joule effect and heat losses by radiation and conduction has been used to study the effects of various gases and gas mixtures currently employed for welding applications. Minor additions of low ionization potential impurities to these gases are shown to significantly perturb the electrical properties of the parent gas causing gross changes in the radial temperature distribution of the arc discharge. Such changes are reflected in the current density distribution and ultimately in the input energy distribution to the weldment. The result is observed as a variation in weld penetration. Recently published experiments and analyses of welding arcs are also evaluated and shown to contain erroneous data and results. Contrary to previous beliefs, the inclusion of a radiation loss term in the basic energy balance equation is important and cannot be considered as negligible in an argon arc at temperatures as low as 10,000 0 K. The one-dimensional analysis of the welding arc as well as the evaluation of these earlier published reports helps to explain the effects of various gases used for welding, improves our understanding of the physics of the welding arc, and provides a stepping stone for a more elaborate model which can be applied to help optimize welding parameters

  10. Physical simulation method for the investigation of weld seam formation during the extrusion of aluminum alloys

    NARCIS (Netherlands)

    Fang, G; Zhou, J.

    2017-01-01

    Extrusion through the porthole die is a predominant forming process used in the production of hollow aluminum alloy profiles across the aluminum extrusion industry. Longitudinal weld seams formed during the process may negatively influence the quality of extruded profiles. It is therefore of

  11. SCC testing of steam generator tubes repaired by welded sleeves

    International Nuclear Information System (INIS)

    Pierson, E.; Stubbe, J.

    1993-01-01

    One way to repair steam generator tubing is to introduce a sleeve inside the tube so that it spans the corroded area and to seal it at both ends. This technique has been studied at Laborelec with a particular attention paid to the occurrence of new SCC cracks at the upper joint. Tube segments coming from the same lot of mill annealed alloy 600 were sent to six manufacturers to be sleeved by their own procedure (including TIG, laser or kinetic welding, followed or not by a stress relief heat treatment), and then tested at Laborelecin 10% NaOH at 350 degrees C. The tests were performed with and without differential pressure i.e. in capsules (Δ = 9 and 19 MPa) and in autoclave (Δp = 0). Nearly all the not stress relieved mock-ups developed through cracks in several hundred hours in auto-clave. The cracks were circumferential and situated near the weld. At 9 and 19 MPa, the time to failure decreased and longitudinal cracks appeared near the weld and at the transition zone of expanded areas. Cracks were never observed in the alloy 690 sleeve, except in the weld bead. Reference capsules (roll expaned tubes) made of the same lot of alloy 600 were tested in the same environment

  12. Laser welding of tailored blanks

    International Nuclear Information System (INIS)

    Pecas, P.; Gouveia, H.; Quintino, L.; Olsen, F.O.

    1998-01-01

    Laser welding has an increasing role in the automotive industry, namely on the sub-assemblies manufacturing. Several sheet-shape parts are laser welded, on a dissimilar combination of thicknesses and materials, and are afterwards formed (stamped) being transformed in a vehicle body component. In this paper low carbon CO 2 laser welding, on the thicknesses of 1,25 and 0.75 mm, formability investigation is described. There will be a description of how the laser welded blanks behave in different forming tests, and the influence of misalignment and undercut on the formability. The quality is evaluated by measuring the limit strain and limit effective strain for the laser welded sheets and the base material, which will be presented in a forming limit diagram. (Author) 14 refs

  13. Reduction of Biomechanical and Welding Fume Exposures in Stud Welding.

    Science.gov (United States)

    Fethke, Nathan B; Peters, Thomas M; Leonard, Stephanie; Metwali, Mahmoud; Mudunkotuwa, Imali A

    2016-04-01

    The welding of shear stud connectors to structural steel in construction requires a prolonged stooped posture that exposes ironworkers to biomechanical and welding fume hazards. In this study, biomechanical and welding fume exposures during stud welding using conventional methods were compared to exposures associated with use of a prototype system that allowed participants to weld from an upright position. The effect of base material (i.e. bare structural beam versus galvanized decking) on welding fume concentration (particle number and mass), particle size distribution, and particle composition was also explored. Thirty participants completed a series of stud welding simulations in a local apprenticeship training facility. Use of the upright system was associated with substantial reductions in trunk inclination and the activity levels of several muscle groups. Inhalable mass concentrations of welding fume (averaged over ~18 min) when using conventional methods were high (18.2 mg m(-3) for bare beam; 65.7 mg m(-3) for through deck), with estimated mass concentrations of iron (7.8 mg m(-3) for bare beam; 15.8 mg m(-3) for through deck), zinc (0.2 mg m(-3) for bare beam; 15.8 mg m(-3) for through deck), and manganese (0.9 mg m(-3) for bare beam; 1.5 mg m(-3) for through deck) often exceeding the American Conference of Governmental Industrial Hygienists Threshold Limit Values (TLVs). Number and mass concentrations were substantially reduced when using the upright system, although the total inhalable mass concentration remained above the TLV when welding through decking. The average diameters of the welding fume particles for both bare beam (31±17 nm) through deck conditions (34±34 nm) and the chemical composition of the particles indicated the presence of metallic nanoparticles. Stud welding exposes ironworkers to potentially high levels of biomechanical loading (primarily to the low back) and welding fume. The upright system used in this study improved exposure

  14. 49 CFR 195.214 - Welding procedures.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Welding procedures. 195.214 Section 195.214... PIPELINE Construction § 195.214 Welding procedures. (a) Welding must be performed by a qualified welder in accordance with welding procedures qualified under Section 5 of API 1104 or Section IX of the ASME Boiler and...

  15. 29 CFR 1910.255 - Resistance welding.

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 5 2010-07-01 2010-07-01 false Resistance welding. 1910.255 Section 1910.255 Labor... OCCUPATIONAL SAFETY AND HEALTH STANDARDS Welding, Cutting and Brazing § 1910.255 Resistance welding. (a.... Ignitron tubes used in resistance welding equipment shall be equipped with a thermal protection switch. (3...

  16. METHOD AND SYSTEM FOR LASER WELDING

    DEFF Research Database (Denmark)

    2008-01-01

    The invention relates to laser welding of at least two adjacent, abutting or overlapping work pieces in a welding direction using multiple laser beams guided to a welding region, wherein at least two of the multiple laser beams are coupled into the welding region so as to form a melt and at least...

  17. Tailor-welded blanks and their production

    Science.gov (United States)

    Yan, Qi

    2005-01-01

    Tailor welded blanks had been widely used in the automobile industry. A tailor welded blank consists of several flat sheets that were laser welded together before stamping. A combination of different materials, thickness, and coatings could be welded together to form a blank for stamping car body panels. As for the material for automobile industry, this technology was one of the development trend for automobile industry because of its weight reduction, safety improvement and economical use of materials. In this paper, the characters and production of tailor welded blanks in the market were discussed in detail. There had two major methods to produce tailor welded blanks. Laser welding would replace mesh seam welding for the production of tailor welded blanks in the future. The requirements on the edge preparation of unwelded blanks for tailor welded blanks were higher than the other steel processing technology. In order to produce the laser welded blank, there had the other process before the laser welding in the factory. In the world, there had three kinds of patterns for the large volume production of tailor welded blanks. In China, steel factory played the important role in the promotion of the application of tailor welded blanks. The competition for the supply of tailor welded blanks to the automobile industry would become fierce in the near future. As a result, the demand for the quality control on the production of tailor welded blanks would be the first priority concern for the factory.

  18. Clamp and Gas Nozzle for TIG Welding

    Science.gov (United States)

    Gue, G. B.; Goller, H. L.

    1982-01-01

    Tool that combines clamp with gas nozzle is aid to tungsten/inert-gas (TIG) welding in hard-to-reach spots. Tool holds work to be welded while directing a stream of argon gas at weld joint, providing an oxygen-free environment for tungsten-arc welding.

  19. Electron beam welding fundamentals and applications

    International Nuclear Information System (INIS)

    Mara, G.L.; Armstrong, R.E.

    1975-01-01

    The electron beam welding process is described and the unique mode of operation and penetration explained by a description of the forces operating within the weld pool. This penetration model is demonstrated by high speed cinematography of the weld pool on several materials. The conditions under which weld defects are formed are discussed and examples are presented. (auth)

  20. Electron beam welding of aluminium components

    International Nuclear Information System (INIS)

    Maajid, Ali; Vadali, S.K.; Maury, D.K.

    2015-01-01

    Aluminium is one of the most widely used materials in industries like transportation, shipbuilding, manufacturing, aerospace, nuclear, etc. The challenges in joining of aluminium are distortion, cleanliness and quality. Main difficulties faced during fusion welding of aluminium components are removal of surface oxide layer, weld porosity, high heat input requirement, distortion, hot cracking, etc. Physical properties of aluminium such as its high thermal conductivity, high coefficient of thermal expansion, no change in colour at high temperature, large difference in the melting points of the metal and its oxide (∼ 1400 °C) compound the difficulties faced during welding. Gas Tungsten Arc Welding (GTAW), Gas Metal Arc Welding (GMAW), Plasma Arc Welding (PAW), etc are generally used in industries for fusion welding of aluminium alloys. However in case of thicker jobs the above processes are not suitable due to requirements of elaborate edge preparation, preheating of jobs, fixturing to prevent distortion, etc. Moreover, precise control over the heat input during welding and weld bead penetration is not possible with above processes. Further, if heat sensitive parts are located near the weld joint then high energy density beam welding process like Electron Beam Welding (EBW) is the best possible choice for aluminium welding.This paper discusses EB welding of aluminium components, typical geometry of components, selection/optimization of welding parameters, problems faced during standardization of welding and process parameters and their remedies etc.

  1. Performance-based improvement of the leakage rate test program for the reactor containment of HTTR. Adoption of revised test programs containing 'Type A, Type B and Type C tests'

    International Nuclear Information System (INIS)

    Kondo, Masaaki; Emori, Koichi; Sekita, Kenji; Furusawa, Takayuki; Hayakawa, Masato; Kozawa, Takayuki; Aono, Tetsuya; Kuroha, Misao; Ouchi, Hiroshi; Kimishima, Satoru

    2008-10-01

    The reactor containment of HTTR is periodically tested to confirm leak-tight integrity by conducting overall integrated leakage rate tests, so-called 'Type A tests,' in accordance with a standard testing method provided in Japan Electric Association Code (JEAC) 4203. 'Type A test' is identified as a basic one for measuring whole leakage rates for reactor containments, it takes, however, much of cost and time of preparation, implementation and restoration of itself. Therefore, in order to upgrade the maintenance technology of HTTR, the containment leakage rate test program for HTTR was revised by adopting efficient and economical alternatives including Type B and Type C tests' which intend to measure leakage rates for containment penetrations and isolation valves, respectively. In JEAC4203-2004, following requirements are specified for adopting an alternative program: upward trend of the overall integrated leakage rate due to aging affection should not be recognized; performance criterion for combined leakage rate, that is a summation of local leakage rates evaluated by Type B and Type C tests and converted to whole leakage rates, should be established; the criterion of the combined leakage rate should be satisfied as well as of the overall integrated leakage rate; correlation between the overall integrated and combined leakage rates should be recognized. Considering the historical performances, policies of conforming to the forgoing requirements and of carrying out the revised test program were developed, which were accepted by the regulatory agency. This report presents an outline of the leakage rate tests for the reactor containment of HTTR, identifies practical issues of conventional Type A tests, and describes the conforming and implementing policies mentioned above. (author)

  2. Machine for welding solar cell connections

    Energy Technology Data Exchange (ETDEWEB)

    Lorans, D.Y.

    1977-08-09

    A machine for welding a connection wire over a solar cell electrode is described which comprises a base, a welding mount for the solar cell which is supported on the base, means for holding the solar cell on the welding mount, welding electrodes, means to lower the welding electrodes over the solar cell and the connection wire superimposed thereon, means for applying electric current pulses to said welding electrodes. It is characterized by the fact that it further comprises means for imparting to said mount an alternating transverse movement in relation to said base before and during the welding operation.

  3. Spot Welding Characterizations With Time Variable

    International Nuclear Information System (INIS)

    Abdul Hafid; Pinitoyo, A.; History; Paidjo, Andryansyah; Sagino, Sudarmin; Tamzil, M.

    2001-01-01

    For obtain spot welding used effective data, this research is made, so that time operational of machine increasing. Welding parameters are material classification, electrical current, and weld time. All of the factors are determined welding quality. If the plate more thick, the time must be longer when the current constant. Another factor as determined welding quality are surface condition of electrode, surface condition of weld material, and material classifications. In this research, the weld machine type IP32A2 VI (110 V), Rivoira trademark is characterized

  4. Development of automatic laser welding system

    International Nuclear Information System (INIS)

    Ohwaki, Katsura

    2002-01-01

    Laser are a new production tool for high speed and low distortion welding and applications to automatic welding lines are increasing. IHI has long experience of laser processing for the preservation of nuclear power plants, welding of airplane engines and so on. Moreover, YAG laser oscillators and various kinds of hardware have been developed for laser welding and automation. Combining these welding technologies and laser hardware technologies produce the automatic laser welding system. In this paper, the component technologies are described, including combined optics intended to improve welding stability, laser oscillators, monitoring system, seam tracking system and so on. (author)

  5. Research on the Effects of Technical Parameters on the Molding of the Weld by A-TIG Welding

    OpenAIRE

    Shi, Kai; Pan, Wu

    2012-01-01

    The effects of welding parameters on the molding of weld by A-TIG welding of a 4mm thickness mild steel plate is studied in the present paper. The results obtained show that: as welding current increases A-TIG welding penetration gets deeper than TIG welding; size and shape of HAZ has remarkable change; A-TIG welding has the narrower weld pool width than TIG welding.

  6. Plasma spot welding of ferritic stainless steels

    International Nuclear Information System (INIS)

    Lesnjak, A.; Tusek, J.

    2002-01-01

    Plasma spot wedding of ferritic stainless steels studied. The study was focused on welding parameters, plasma and shieldings and the optimum welding equipment. Plasma-spot welded overlap joints on a 0.8 mm thick ferritic stainless steel sheet were subjected to a visual examination and mechanical testing in terms of tension-shear strength. Several macro specimens were prepared Plasma spot welding is suitable to use the same gas as shielding gas and as plasma gas , i. e. a 98% Ar/2% H 2 gas mixture. Tension-shear strength of plasma-spot welded joint was compared to that of resistance sport welded joints. It was found that the resistance welded joints withstand a somewhat stronger load than the plasma welded joints due to a large weld sport diameter of the former. Strength of both types of welded joints is approximately the same. (Author) 32 refs

  7. Automatic welding of stainless steel tubing

    Science.gov (United States)

    Clautice, W. E.

    1978-01-01

    The use of automatic welding for making girth welds in stainless steel tubing was investigated as well as the reduction in fabrication costs resulting from the elimination of radiographic inspection. Test methodology, materials, and techniques are discussed, and data sheets for individual tests are included. Process variables studied include welding amperes, revolutions per minute, and shielding gas flow. Strip chart recordings, as a definitive method of insuring weld quality, are studied. Test results, determined by both radiographic and visual inspection, are presented and indicate that once optimum welding procedures for specific sizes of tubing are established, and the welding machine operations are certified, then the automatic tube welding process produces good quality welds repeatedly, with a high degree of reliability. Revised specifications for welding tubing using the automatic process and weld visual inspection requirements at the Kennedy Space Center are enumerated.

  8. Torque Measurement of Welding of Endplug-Endplate using Multi-pin Remote Welding System

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Dae-Seo; Kim, Soo-Sung; Park, Geun-Il; Lee, Jung-Won; Song, Kee-Chan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    As fuel bundles in PHWR irradiates, inner pressure in claddings of fuel rods increases owing to outer pressure and fission products of nuclear fissions. Because of leak possibility of welding between cladding and end plug, this welding part connects with safety of nuclear fuel rods. Because of importance of this welding part, weldability of end plug-cladding of nuclear fuel rods is continually researched. Welding method for research and commercialization is classified as melting, solid type welding or resistance welding. End plug cladding welding of nuclear fuel rods in PHWR takes advantage of resistance upset butt welding using multicycle mode. This method makes weld flash and shapes re-entrant corner owing to welding heat due to resistivity, contact resistance of cladding-end plug, and inelasticity deformation due to pressure. Welding part between cladding and end plug receives stresses and makes small cracks. In this study, remote welding system for multi-pin assembly was designed, fabricated and welding specimens of end plug-endplate were made using electrical resistance method. The torques of welding between end plug and endplate were measured. These results on welding current, pressure of main electrode and pressure of branch electrode were analyzed. Weldability between end plug and endplate was confirmed through metallographic examinations. In the future, optimal welding examinations due to welding current, welding pressure and welding time will be performed to improve weldability of end plug-endplate.

  9. Welding Penetration Control of Fixed Pipe in TIG Welding Using Fuzzy Inference System

    Science.gov (United States)

    Baskoro, Ario Sunar; Kabutomori, Masashi; Suga, Yasuo

    This paper presents a study on welding penetration control of fixed pipe in Tungsten Inert Gas (TIG) welding using fuzzy inference system. The welding penetration control is essential to the production quality welds with a specified geometry. For pipe welding using constant arc current and welding speed, the bead width becomes wider as the circumferential welding of small diameter pipes progresses. Having welded pipe in fixed position, obviously, the excessive arc current yields burn through of metals; in contrary, insufficient arc current produces imperfect welding. In order to avoid these errors and to obtain the uniform weld bead over the entire circumference of the pipe, the welding conditions should be controlled as the welding proceeds. This research studies the intelligent welding process of aluminum alloy pipe 6063S-T5 in fixed position using the AC welding machine. The monitoring system used a charge-coupled device (CCD) camera to monitor backside image of molten pool. The captured image was processed to recognize the edge of molten pool by image processing algorithm. Simulation of welding control using fuzzy inference system was constructed to simulate the welding control process. The simulation result shows that fuzzy controller was suitable for controlling the welding speed and appropriate to be implemented into the welding system. A series of experiments was conducted to evaluate the performance of the fuzzy controller. The experimental results show the effectiveness of the control system that is confirmed by sound welds.

  10. Versatile Friction Stir Welding/Friction Plug Welding System

    Science.gov (United States)

    Carter, Robert

    2006-01-01

    A proposed system of tooling, machinery, and control equipment would be capable of performing any of several friction stir welding (FSW) and friction plug welding (FPW) operations. These operations would include the following: Basic FSW; FSW with automated manipulation of the length of the pin tool in real time [the so-called auto-adjustable pin-tool (APT) capability]; Self-reacting FSW (SRFSW); SR-FSW with APT capability and/or real-time adjustment of the distance between the front and back shoulders; and Friction plug welding (FPW) [more specifically, friction push plug welding] or friction pull plug welding (FPPW) to close out the keyhole of, or to repair, an FSW or SR-FSW weld. Prior FSW and FPW systems have been capable of performing one or two of these operations, but none has thus far been capable of performing all of them. The proposed system would include a common tool that would have APT capability for both basic FSW and SR-FSW. Such a tool was described in Tool for Two Types of Friction Stir Welding (MFS- 31647-1), NASA Tech Briefs, Vol. 30, No. 10 (October 2006), page 70. Going beyond what was reported in the cited previous article, the common tool could be used in conjunction with a plug welding head to perform FPW or FPPW. Alternatively, the plug welding head could be integrated, along with the common tool, into a FSW head that would be capable of all of the aforementioned FSW and FPW operations. Any FSW or FPW operation could be performed under any combination of position and/or force control.

  11. Corrosion Properties of Laser Welded Stainless Steel

    DEFF Research Database (Denmark)

    Weldingh, Jakob; Olsen, Flemmming Ove

    1997-01-01

    In this paper the corrosion properties of laser welded AISI 316L stainless steel are examined. A number of different welds has been performed to test the influence of the weld parameters of the resulting corrosion properties. It has been chosen to use the potential independent critical pitting...... temperature (CPT) test as corrosion test. The following welding parameters are varied: Welding speed, lsser power, focus point position and laser operation mode (CW or pulsed)....

  12. Heat Source - Materials Interactions during Fusion Welding.

    Science.gov (United States)

    1982-04-30

    the capabilities of ultrasonic weld pool measurement, and to address questions of applications to active pool size control. -- mom- 44 TIG welding ...preparation. The fraction of absorbed power increases dramatically upon formation of a keyhole . As a result, welds made with sharply beveled edge...laser end electron beam welding processes characteristically produce a deel,, narrow weld bead. This bead is formed by a keyhole mode of operation in

  13. Real time computer controlled weld skate

    Science.gov (United States)

    Wall, W. A., Jr.

    1977-01-01

    A real time, adaptive control, automatic welding system was developed. This system utilizes the general case geometrical relationships between a weldment and a weld skate to precisely maintain constant weld speed and torch angle along a contoured workplace. The system is compatible with the gas tungsten arc weld process or can be adapted to other weld processes. Heli-arc cutting and machine tool routing operations are possible applications.

  14. Collection of arc welding process data

    OpenAIRE

    K. Luksa; Z. Rymarski

    2006-01-01

    Purpose: The aim of the research was to examine the possibility of detecting welding imperfections by recording the instant values of welding parameters. The microprocessor controlled system for real-time collection and display of welding parameters was designed, implemented and tested.Design/methodology/approach: The system records up to 4 digital or analog signals collected from welding process and displays their run on the LCD display. To disturb the welding process artificial disturbances...

  15. Analysis And Control System For Automated Welding

    Science.gov (United States)

    Powell, Bradley W.; Burroughs, Ivan A.; Kennedy, Larry Z.; Rodgers, Michael H.; Goode, K. Wayne

    1994-01-01

    Automated variable-polarity plasma arc (VPPA) welding apparatus operates under electronic supervision by welding analysis and control system. System performs all major monitoring and controlling functions. It acquires, analyzes, and displays weld-quality data in real time and adjusts process parameters accordingly. Also records pertinent data for use in post-weld analysis and documentation of quality. System includes optoelectronic sensors and data processors that provide feedback control of welding process.

  16. Friction Welding of Titanium and Carbon Steel

    OpenAIRE

    Atsushi, HASUI; Yoichi, KIRA; Faculty of Science and Technology, Keio University; Ishikawajima-Harima Heavy Industries, Co., Ltd.

    1985-01-01

    Titanium-steel is a combination of dissimilar materials, which are difficult to weld in general, owing to inevitable formation of brittle intermetallic compounds. A prominent feature of friction welding process is ability to weld dissimilar materials in many kinds of combinations. This report deals with friction weldabilily of pure titanium and S25C steel, which are 12 mm in diameter. Main results are summarized as follows; (1) Suitable welding conditions to obtain a sound weld, which has a j...

  17. Recent advances in the TIG welding process and the application of the welding of nuclear components

    International Nuclear Information System (INIS)

    Lucas, W.; Males, B.O.

    1982-01-01

    Recent advances in the field of precision arc welding techniques and infacilities for production of nuclear power plant components arc presented. Of the precision welding techniques, pulsed TIG welding, pulsed plasma arc welding, hot-wire TIG welding, and pulsed inert-gas metal-arc welding. In the field of weld cladding, GMA plasma welding is cited as an alternative to submerged-arc welding with a strip electrode. Transistors and computer-controlled welding systems get a special mention. Applications of TIG welding in the UK are cited, e.g. welding of components for the AGR nuclear power plant and construction of equipment for repair work in feedwater pipes of the MAGNOX reactor. (orig.) [de

  18. Investigation into Variations of Welding Residual Stresses and Redistribution Behaviors for Different Repair Welding Widths

    International Nuclear Information System (INIS)

    Park, Chiyong; Lee, Hweesueng; Huh, Namsu

    2014-01-01

    In this study, we investigated the variations in welding residual stresses in dissimilar metal butt weld due to width of repair welding and re-distribution behaviors resulting from similar metal welding (SMW) and mechanical loading. To this end, detailed two-dimensional axi-symmetric finite element (FE) analyses were performed considering five different repair welding widths. Based on the FE results, we first evaluated the welding residual stress distributions in repair welding. We then investigated the re-distribution behaviors of the residual stresses due to SMW and mechanical loads. It is revealed that large tensile welding residual stresses take place in the inner surface and that its distribution is affected, provided repair welding width is larger than certain value. The welding residual stresses resulting from repair welding are remarkably reduced due to SMW and mechanical loading, regardless of the width of the repair welding

  19. 10,170 flawless welds

    CERN Multimedia

    Antonella Del Rosso

    2014-01-01

    The welding of tubes containing the principal current-carrying busbars in the LHC magnets was one of the main activities of the SMACC project. After a year of preparation and another of intense activity in the tunnel, the last weld was completed on Wednesday 14 May. Over 10,170 welds have been inspected and not a single fault has been found.    The welder (above) creates the weld using an orbital welding machine (below) specifically designed for CERN. Each of the eight sectors of the LHC contains around 210 interconnects between the superconducting magnets. Consolidating these interconnections was the SMACC project’s primary objective. One of the last jobs before closing the interconnects is the welding of the M lines: each has a 104 mm diameter and a radial clearance of just 45 mm. In total: 10,170 welds carried out in a single year of activities. A true challenge, which was carried out by a team of 30 highly specialised welders, working under the supervision o...

  20. Deformation During Friction Stir Welding

    Science.gov (United States)

    White, Henry J.

    2002-01-01

    Friction Stir Welding (FSW) is a solid state welding process that exhibits characteristics similar to traditional metal cutting processes. The plastic deformation that occurs during friction stir welding is due to the superposition of three flow fields: a primary rotation of a radially symmetric solid plug of metal surrounding the pin tool, a secondary uniform translation, and a tertiary ring vortex flow (smoke rings) surrounding the tool. If the metal sticks to the tool, the plug surface extends down into the metal from the outer edge of the tool shoulder, decreases in diameter like a funnel, and closes up beneath the pin. Since its invention, ten years have gone by and still very little is known about the physics of the friction stir welding process. In this experiment, an H13 steel weld tool (shoulder diameter, 0.797 in; pin diameter, 0.312 in; and pin length, 0.2506 in) was used to weld three 0.255 in thick plates. The deformation behavior during friction stir welding was investigated by metallographically preparing a plan view sections of the weldment and taking Vickers hardness test in the key-hole region.

  1. LASER WELDING WITH MICRO-JET COOLING FOR TRUCK FRAME WELDING

    OpenAIRE

    Jan PIWNIK; Bożena SZCZUCKA-LASOTA; Tomasz WĘGRZYN; Wojciech MAJEWSKI

    2017-01-01

    The aim of this paper is to analyse the mechanical properties of the weld steel structure of car body truck frames after laser welding. The best welding conditions involve the use of proper materials and alloy elements in steel and filer materials, in addition to welding technology, state of stress and temperature of exploitation. We present for the first time the properties of steel track structures after laser welding with micro-jet cooling. Therefore, good selection of both welding paramet...

  2. A study of weld quality in ultrasonic spot welding of similar and dissimilar metals

    International Nuclear Information System (INIS)

    Al-Sarraf, Z; Lucas, M

    2012-01-01

    Several difficulties are faced in joining thinner sheets of similar and dissimilar materials from fusion welding processes such as resistance welding and laser welding. Ultrasonic metal welding overcomes many of these difficulties by using high frequency vibration and applied pressure to create a solid-state weld. Ultrasonic metal welding is an effective technique in joining small components, such as in wire bonding, but is also capable of joining thicker sheet, depending on the control of welding conditions. This study presents the design, characterisation and test of a lateral-drive ultrasonic metal welding device. The ultrasonic welding horn is modelled using finite element analysis and its vibration behaviour is characterised experimentally to ensure ultrasonic energy is delivered to the weld coupon. The welding stack and fixtures are then designed and mounted on a test machine to allow a series of experiments to be conducted for various welding and ultrasonic parameters. Weld strength is subsequently analysed using tensile-shear tests. Control of the vibration amplitude profile through the weld cycle is used to enhance weld strength and quality, providing an opportunity to reduce part marking. Optical microscopic examination and scanning electron microscopy (SEM) were employed to investigate the weld quality. The results show how the weld quality is particularly sensitive to the combination of clamping force and vibration amplitude of the welding tip.

  3. The technology and welding joint properties of hybrid laser-tig welding on thick plate

    Science.gov (United States)

    Shenghai, Zhang; Yifu, Shen; Huijuan, Qiu

    2013-06-01

    The technologies of autogenous laser welding and hybrid laser-TIG welding are used on thick plate of high strength lower alloy structural steel 10CrNiMnMoV in this article. The unique advantages of hybrid laser-TIG welding is summarized by comparing and analyzing the process parameters and welding joints of autogenous laser welding laser welding and hybrid laser-TIG welding. With the optimal process parameters of hybrid welding, the good welding joint without visible flaws can be obtained and its mechanical properties are tested according to industry standards. The results show that the hybrid welding technology has certain advantages and possibility in welding thick plates. It can reduce the demands of laser power, and it is significant for lowering the aspect ratio of weld during hybrid welding, so the gas in the molten pool can rise and escape easily while welding thick plates. Therefore, the pores forming tendency decreases. At the same time, hybrid welding enhances welding speed, and optimizes the energy input. The transition and grain size of the microstructure of hybrid welding joint is better and its hardness is higher than base material. Furthermore, its tensile strength and impact toughness is as good as base material. Consequently, the hybrid welding joint can meet the industry needs completely.

  4. Plasticity Theory of Fillet Welds

    DEFF Research Database (Denmark)

    Hansen, Thomas

    2005-01-01

    a safe and statically admissible stress distribution is established. The plasticity solutions are compared with tests carried out at the Engineering Academy of Denmark, Lyngby, in the early nineties, and old fillet weld tests. The new failure conditions are in very good agreement with the yield load......This paper deals with simple methods for calculation of fillet welds based on the theory of plasticity. In developing the solutions the lower-bound theorem is used. The welding material and parts of the base material are subdivided into triangular regions with homogeneous stress fields; thereby...... tests, but not so good agreement with the old failure load tests....

  5. Review of laser hybrid welding

    DEFF Research Database (Denmark)

    Bagger, Claus

    2004-01-01

    In this artucle an overview og the hybrid welding process is given. After a short historic overview, a review of the fundamental phenomenon taking place when a laser (CO2 or Nd:YAG) interacts in the same molten pool as a more conventional source of energy, e.g. tungsten in-active gas, plasma......, or metal inactive gas/metal active gas.This is followed by reports of how the many process parameters governing the hybrid welding process can be set and how the choice of secondary energy source, shielding gas, etc. can affect the overall welding process....

  6. Novel Process Revolutionizes Welding Industry

    Science.gov (United States)

    2008-01-01

    Glenn Research Center, Delphi Corporation, and the Michigan Research Institute entered into a research project to study the use of Deformation Resistance Welding (DRW) in the construction and repair of stationary structures with multiple geometries and dissimilar materials, such as those NASA might use on the Moon or Mars. Traditional welding technologies are burdened by significant business and engineering challenges, including high costs of equipment and labor, heat-affected zones, limited automation, and inconsistent quality. DRW addresses each of those issues, while drastically reducing welding, manufacturing, and maintenance costs.

  7. Waste canister closure welding using the inertia friction welding process

    International Nuclear Information System (INIS)

    Klein, R.F.; Siemens, D.H.; Kuruzar, D.L.

    1986-02-01

    Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it in a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive seal weld, the properties and thickness of which are at least equal to those of the canister material. This paper describes the inertia friction welding process and a proposed equipment design concept that will provide a positive, reliable, inspectable, and full thickness seal weld while providing easily maintainable equipment, even though the weld is made in a highly contaminated hot cell. All studies and tests performed have shown the concept to be highly feasible. 2 refs., 6 figs

  8. Dictionary: Welding, cutting and allied processes. Pt. 2

    International Nuclear Information System (INIS)

    Kleiber, A.W.

    1987-01-01

    The dictionary contains approximately 40 000 entries covering all aspects of welding technology. It is based on the evaluation of numerous English, American and German sources. This comprehensive and up to date dictionary will be a reliable and helpful aid in evaluation and translating. The dictionary covers the following areas: Welding: gas welding, arc welding, gas shielded welding, resistance welding, welding of plastics, special welding processes; Cutting: flame cutting, arc cutting and special thermal cutting processes; Soldering: brazing and soldering; Other topics: thermal spraying, metal to metal adhesion, welding filler materials and other consumables, test methods, plant and equipment, accessories, automation, welding trade, general welding terminology. (orig./HP) [de

  9. Automatic welding and cladding in heavy fabrication

    International Nuclear Information System (INIS)

    Altamer, A. de

    1980-01-01

    A description is given of the automatic welding processes used by an Italian fabricator of pressure vessels for petrochemical and nuclear plant. The automatic submerged arc welding, submerged arc strip cladding, pulsed TIG, hot wire TIG and MIG welding processes have proved satisfactory in terms of process reliability, metal deposition rate, and cost effectiveness for low alloy and carbon steels. An example shows sequences required during automatic butt welding, including heat treatments. Factors which govern satisfactory automatic welding include automatic anti-drift rotator device, electrode guidance and bead programming system, the capability of single and dual head operation, flux recovery and slag removal systems, operator environment and controls, maintaining continuity of welding and automatic reverse side grinding. Automatic welding is used for: joining vessel sections; joining tubes to tubeplate; cladding of vessel rings and tubes, dished ends and extruded nozzles; nozzle to shell and butt welds, including narrow gap welding. (author)

  10. Materials participation in welded joints manufacturing

    Science.gov (United States)

    Ghenghea, L. D.

    2016-08-01

    Management of materials dilution to form a joint with higher features asked by complex metallic structures is a problem that took attention and efforts of welding processes researchers and this communication will give a little contribution presenting some scientific and experimental results of dilution processes studied by Welding Research Group from Iasi, Romania, TCM Department. Liquid state welding processes have a strong dependence related to dilution of base and filler materials, the most important are for automatic joining using welding. The paper presents a review of some scientific works already published and their contributions, results of dilution coefficient evaluation using weighing, graphics and software applied for shielded metal arc welding process. Paper results could be used for welders’ qualification, welding procedure specification and other welding processes researchers’ activities. The results of Welding Research Group from Iasi, Romania, TCM Department, show dilution coefficient values between 20-30 % of base material and 70-80 % of filler material for studied welding process.

  11. Gas Shielding Technology for Welding and Brazing

    Science.gov (United States)

    Nunes, Arthur J.; Gradl, Paul R.

    2012-01-01

    Welding is a common method that allows two metallic materials to be joined together with high structural integrity. When joints need to be leak-tight, light-weight, or free of contaminant-trapping seams or surface asperities, welding tends to be specified. There are many welding techniques, each with its own advantages and disadvantages. Some of these techniques include Forge Welding, Gas Tungsten Arc Welding, Friction Stir Welding, and Laser Beam Welding to name a few. Whichever technique is used, the objective is a structural joint that meets the requirements of a particular component or assembly. A key practice in producing quality welds is the use of shielding gas. This article discusses various weld techniques, quality of the welds, and importance of shielding gas in each of those techniques. Metallic bonds, or joints, are produced when metals are put into intimate contact. In the solid-state "blacksmith welding" process, now called Forge Welding (FOW), the site to be joined is pounded into intimate contact. The surfaces to be joined usually need to be heated to make it easier to deform the metal. The surfaces are sprinkled with a flux to melt surface oxides and given a concave shape so that surface contamination can be squeezed out of the joint as the surfaces are pounded together; otherwise the surface contamination would be trapped in the joint and would weaken the weld. In solid-state welding processes surface oxides or other contamination are typically squeezed out of the joint in "flash."

  12. Method for enhanced control of welding processes

    Science.gov (United States)

    Sheaffer, Donald A.; Renzi, Ronald F.; Tung, David M.; Schroder, Kevin

    2000-01-01

    Method and system for producing high quality welds in welding processes, in general, and gas tungsten arc (GTA) welding, in particular by controlling weld penetration. Light emitted from a weld pool is collected from the backside of a workpiece by optical means during welding and transmitted to a digital video camera for further processing, after the emitted light is first passed through a short wavelength pass filter to remove infrared radiation. By filtering out the infrared component of the light emitted from the backside weld pool image, the present invention provides for the accurate determination of the weld pool boundary. Data from the digital camera is fed to an imaging board which focuses on a 100.times.100 pixel portion of the image. The board performs a thresholding operation and provides this information to a digital signal processor to compute the backside weld pool dimensions and area. This information is used by a control system, in a dynamic feedback mode, to automatically adjust appropriate parameters of a welding system, such as the welding current, to control weld penetration and thus, create a uniform weld bead and high quality weld.

  13. MFDC - technological improvement in resistance welding controls

    Energy Technology Data Exchange (ETDEWEB)

    Somani, A.K.; Naga Bhaskar, V.; Chandramouli, J.; Rameshwara Rao, A. [Nuclear Fuel Complex, Dept. of Atomic Energy, Hyderabad (India)

    2008-07-01

    Among the various Resistance Welding operations carried out in the production line of a fuel bundle end plug welding is the most critical operation. Welding controllers play a very vital role in obtaining consistent weld quality by regulating and controlling the weld current. Conventional mains synchronized welding controllers are at best capable of controlling the weld current at a maximum speed of the mains frequency. In view of the very short welding durations involved in the various stages of a fuel bundle fabrication, a need was felt for superior welding controllers. Medium Frequency Welding Controllers offer a solution to these limitations in addition to offering other advantages. Medium Frequency power sources offer precise welding current control as they regulate and correct the welding current faster, typically twenty times faster when operated at 1000Hz. An MFDC was employed on one of the welding machines and its performance was studied. This paper discusses about the various advantages of MFDCs with other controllers employed at NFC to end plug welding operation. (author)

  14. Welding-induced local maximum residual stress in heat affected zone of low-carbon austenitic stainless steel with machined surface layer and its influential factors

    International Nuclear Information System (INIS)

    Okano, Shigetaka; Ihara, Ryohei; Kanamaru, Daisuke; Mochizuki, Masahito

    2015-01-01

    In this study, the effects of work-hardening and pre-existing stress in the machined surface layer of low-carbon austenitic stainless steel on the welding-induced residual stress were experimentally investigated through the use of weld specimens with three different surface layers; as-cutout, mechanically-polished and electrolytically-polished. The high tensile and compressive stresses exist in the work-hardened surface layer of the as-cutout and mechanically-polished specimens, respectively. Meanwhile, no stress and work-hardened surface layer exist in the electrolytically-polished specimen. TIG bead-on-plate welding under the same welding heat input conditions was performed to introduce the residual stress into these specimens. Using these welded specimens, the distributions of welding-induced residual stress were measured by the X-ray diffraction method. Similarly, the distributions of hardness in welds were estimated by the Vickers hardness test. And then, these distributions were compared with one another. Based on the results, the residual stress in the weld metal (WM) is completely unaffected by the machined surface layer because the work-hardened surface layer disappears through the processes of melting and solidification during welding. The local maximum longitudinal tensile residual stress in the heat affected zone (HAZ) depends on the work-hardening but not on the existing stress, regardless of whether tensile or compressive, in the machined surface layer before welding. At the base metal far from WM and HAZ, the residual stress is formed by the addition of the welding-induced residual stress to the pre-existing stress in the machined surface layer before welding. The features of the welding-induced residual stress in low-carbon austenitic stainless steel with the machined surface layer and their influential factors were thus clarified. (author)

  15. Cu-Fe welding techniques by electromagnetic and electron beam welding processes

    International Nuclear Information System (INIS)

    Kumar, Satendra; Saroj, P.C.; Kulkarni, M.R.; Sharma, A.; Rajawat, R.K.; Saha, T.K.

    2015-01-01

    Electromagnetic welding being a solid state welding process has been found suitable for welding Copper and Iron which are conventionally very tricky. Owing to good electrical conductivity of both copper and iron, they are best suited combination for EM welding. For the experimental conditions presented above, 1.0 mm wall thickness of Cu tube was lap welded to Fe disc. A heavy duty four disc stainless steel coil was used for electromagnetic welding of samples. MSLD of the welded samples indicated leak proof joints. Metallographic examination of the welds also revealed defect free interfaces. Electron beam welding is also a non-conventional welding process used for joining dissimilar materials. Autogenous welding of the above specimen was carried out by EBW method for the sake of comparison. A characterization analysis of the above mentioned joining processes will be discussed in the paper. (author)

  16. Effect of Welding Parameters on Dilution and Weld Bead Geometry in Cladding

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    The effect of pulsed gas metal arc welding (GMAW) variables on the dilution and weld bead geometry in cladding X65 pipeline steel with 316L stainless steel was studied. Using a full factorial method, a series of experiments were carried out to know the effect of wire feed rate, welding speed, distance between gas nozzle and plate, and the vertical angle of welding on dilution and weld bead geometry. The findings indicate that the dilution of weld metal and its dimension i.e. width, height and depth increase with the feed rate, but the contact angle of the bead decreases first and then increases. Meantime, welding speed has an opposite effect except for dilution. There is an interaction effect between welding parameters at the contact angle. The results also show forehand welding or decreasing electrode extension decrease the angle of contact. Finally,a mathematical model is contrived to highlight the relationship between welding variables with dilution and weld bead geometry.

  17. Weld bonding of stainless steel

    DEFF Research Database (Denmark)

    Santos, I. O.; Zhang, Wenqi; Goncalves, V.M.

    2004-01-01

    . The overall assessment of the weld bonding process is made using several commercial adhesives with varying working times under different surface conditions. The quality of the resulting joints is evaluated by means of macroetching observations, tension-shear tests and peel tests. The theoretical investigation......This paper presents a comprehensive theoretical and experimental investigation of the weld bonding process with the purpose of evaluating its relative performance in case of joining stainless steel parts, against alternative solutions based on structural adhesives or conventional spot-welding...... of the process consists of numerical predictions based on the commercial finite element program SORPAS with the purpose of establishing the most favourable parameters that allow spot-welding through the adhesives....

  18. Welding and Production Metallurgy Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This 6000 square foot facility represents the only welding laboratory of its kind within DA. It is capable of conducting investigations associated with solid state...

  19. Thermomechanical Modelling of Resistance Welding

    DEFF Research Database (Denmark)

    Bay, Niels; Zhang, Wenqi

    2007-01-01

    The present paper describes a generic programme for analysis, optimization and development of resistance spot and projection welding. The programme includes an electrical model determining electric current and voltage distribution as well as heat generation, a thermal model calculating heat...

  20. Welding process modelling and control

    Science.gov (United States)

    Romine, Peter L.; Adenwala, Jinen A.

    1993-01-01

    The research and analysis performed, and software developed, and hardware/software recommendations made during 1992 in development of the PC-based data acquisition system for support of Welding Process Modeling and Control is reported. A need was identified by the Metals Processing Branch of NASA Marshall Space Flight Center, for a mobile data aquisition and analysis system, customized for welding measurement and calibration. Several hardware configurations were evaluated and a PC-based system was chosen. The Welding Measurement System (WMS) is a dedicated instrument, strictly for the use of data aquisition and analysis. Although the WMS supports many of the functions associated with the process control, it is not the intention for this system to be used for welding process control.

  1. Welding development for LMFBR applications

    International Nuclear Information System (INIS)

    Slaughter, G.M.; Edmonds, D.P.; Goodwin, G.M.; King, J.F.; Moorhead, A.J.

    1976-01-01

    High-quality welds with suitable properties for long-time elevated-temperature nuclear service are among the most critical needs in today's welding technology. Safe, reliable, and economic generation of future power depends on welded construction in systems such as Liquid Metal Fast Breeder Reactors (LMFBRs). Rapid thermal transients in LMFBR systems at coolant temperatures around 590 to 650 0 C (1000 to 1200 0 F) could cause creep and creep-fatigue damage that is not encountered in lower temperature reactor systems. The undesirable consequences of interaction between the two working fluids - sodium and steam - in the steam generators are also of major concern. Thus sound welds that have excellent reliability over a 30-year service life are essential. Several programs are actively underway at ORNL to satisfy this critical need and selected portions of three of these programs are discussed briefly

  2. Mehanical Properties of Electron Beam Welded Joints in Thick Gage CA6NM Stainless Steel

    Science.gov (United States)

    Sarafan, Sheida; Wanjara, Priti; Gholipour, Javad; Champliaud, Henri; Mathieu, Louis

    2017-10-01

    Design of hydroelectric turbine components requires high integrity welds (without detectable volumetric defects) in heavy gage sections of stainless steel materials, such as ASTM A743 grade CA6NM—a low carbon 13% Cr-4% Ni martensitic stainless steel that is manufactured in cast form. In this work, 90-mm-thick plates of CA6NM were joined using a single-pass autogenous electron beam (EB) welding process and the mechanical properties were evaluated in the as-welded condition to characterize the performance of the joints. The static tensile properties that were evaluated in two directions—transverse and longitudinal to the EB weld seam—demonstrated conformance of the joints with the requirements of the ASME Section IX standard. The Charpy impact energies of the EB welds—measured at -18 °C on samples with V-notch roots located in the fusion and heat-affected zones—met the minimum requirements of 27 J specified in ASME Section VIII standard. In addition, bend tests that were conducted on the entire weld cross section displayed no discontinuities on the tension side of the bent joints. Hence, the developed EB welding process was demonstrated to render high-performance joints and promises key advantages for industrialization, such as cost savings through reductions in consumable material, production time and labor intensity.

  3. Modelling of microstructural creep damage in welded joints of 316L stainless steel

    International Nuclear Information System (INIS)

    Bouche, G.

    2000-01-01

    Welded joints of 316L stainless steel under service conditions at elevated temperature are known to be preferential sites of creep damage, as compared to the base material. This damage results in the formation of cavities and the development of creep cracks which can lead to a premature failure of welded components. The complex two-phase microstructure of 316L welds was simulated by manually filling a mould with longitudinal deposited weld beads. The moulded material was then aged during 2000 hours at 600 deg. C. High resolution Scanning Electron Microscopy was largely used to examine the microstructure of the simulated material before and after ageing. Smooth and notched creep specimens were cut from the mould and tested at 600 deg. C under various stress levels. A comparison of the lifetime versus nominal stress curves for the base and welded materials shows a greater dependence of the welded material to creep phenomena. Observation and EBSD analysis show that damage is preferentially located along the austenite grain boundaries. The stress and strain fields in the notched specimens were calculated by finite element method. A correlation of this field to the observed damage was made in order to propose a predictive law relating the creep damage to the mechanical conditions applied locally. Further mechanical tests and simulation on CT specimens and mode II tubular specimens allowed validating the model under various multiaxial loading conditions. (author)

  4. Laser welding of aluminium alloys

    OpenAIRE

    Forsman, Tomas

    2000-01-01

    This thesis treats laser welding of aluminium alloys from a practical perspective with elements of mathematical analysis. The theoretical work has in all cases been verified experimentally. The aluminium alloys studied are from the 5xxx and 6xxx groups which are common for example in the automotive industry. Aluminium has many unique physical properties. The properties which more than others have been shown to influence the welding process is its high reflection, high thermal conductivity, lo...

  5. Austenitic stainless steel weld inspection

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  6. Laser based spot weld characterization

    Science.gov (United States)

    Jonietz, Florian; Myrach, Philipp; Rethmeier, Michael; Suwala, Hubert; Ziegler, Mathias

    2016-02-01

    Spot welding is one of the most important joining technologies, especially in the automotive industry. Hitherto, the quality of spot welded joints is tested mainly by random destructive tests. A nondestructive testing technique offers the benefit of cost reduction of the testing procedure and optimization of the fabrication process, because every joint could be examined. This would lead to a reduced number of spot welded joints, as redundancies could be avoided. In the procedure described here, the spot welded joint between two zinc-coated steel sheets (HX340LAD+Z100MB or HC340LA+ZE 50/50) is heated optically on one side. Laser radiation and flash light are used as heat sources. The melted zone, the so called "weld nugget" provides the mechanical stability of the connection, but also constitutes a thermal bridge between the sheets. Due to the better thermal contact, the spot welded joint reveals a thermal behavior different from the surrounding material, where the heat transfer between the two sheets is much lower. The difference in the transient thermal behavior is measured with time resolved thermography. Hence, the size of the thermal contact between the two sheets is determined, which is directly correlated to the size of the weld nugget, indicating the quality of the spot weld. The method performs well in transmission with laser radiation and flash light. With laser radiation, it works even in reflection geometry, thus offering the possibility of testing with just one-sided accessibility. By using heating with collimated laser radiation, not only contact-free, but also remote testing is feasible. A further convenience compared to similar thermographic approaches is the applicability on bare steel sheets without any optical coating for emissivity correction. For this purpose, a proper way of emissivity correction was established.

  7. Welding of Prosthetic Alloys

    Directory of Open Access Journals (Sweden)

    Wojciechowska M.

    2015-04-01

    Full Text Available This paper presents the techniques of joining metal denture elements, used in prosthetic dentistry: the traditional soldering technique with a gas burner and a new technique of welding with a laser beam; the aim of the study was to make a comparative assessment of the quality of the joints in view of the possibility of applying them in prosthetic structures. Fractographic examinations were conducted along with tensile strength and impact strength tests, and the quality of the joints was assessed compared to the solid metal. The experiments have shown that the metal elements used to make dentures, joined by the technique which employs a laser beam, have better strength properties than those achieved with a gas burner.

  8. Modeling of fuel performance and fission product release behavior during HTTR normal operation. A comparative study of the FZJ and JAERI modeling approach

    International Nuclear Information System (INIS)

    Verfondern, Karl; Sumita, Junya; Ueta, Shohei; Sawa, Kazuhiro

    2001-03-01

    For the prediction of fuel performance and fission product release behavior in the High Temperature Engineering Test Reactor, HTTR of the Japan Atomic Energy Research Institute(JAERI), during its normal operation, calculation tools were applied as have been used at the Research Center Juelich (FZJ) in safety analyses for pebble-bed HTGR designs. Calculations were made assuming the HTTR operation with a nominal operation time of 660 efpd including a 110 efpd period with elevated fuel temperatures. Fuel performance calculations by the PANAMA code with given fuel temperature distribution in the core have shown that the additional failure level of about 5x10 -6 is expected which is about twice as much as the as-fabricated through-coatings failure level. Under the extreme safety design conditions, the predicted particle failure fraction in the core increases to about 1x10 -3 in maximum. The diffusive release of metallic fission products from the fuel primarily occurs in the core layer with the maximum fuel temperature (layer 3) whereas there is hardly any contribution from layer 1 except for the recoil fraction. Silver most easily escapes the fuel; the predicted release fractions from the fuel compacts are 10% (expected) and 50% (safety design). The figures for strontium (expected: 1.5x10 -3 ), safety design: 3.1x10 -2 ) and cesium (5.6x10 -4 , 2.9x10 -2 ) reveal as well a significant fraction to originate already from intact particles. Comparison with the calculation based on JAERI's diffusion model for cesium shows a good agreement for the release behavior from the particles. The differences in the results can be explained mainly by the different diffusion coefficients applied. The release into the coolant can not modelled because of the influence of the gap between compact and graphite sleeve lowering the release by a factor of 3 to 10. For the prediction of performance and fission product release behavior of advanced ZrC TRISO particles, more experimental work is

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

    Science.gov (United States)

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

    2018-04-01

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

  10. Hybrid laser-TIG welding, laser beam welding and gas tungsten arc welding of AZ31B magnesium alloy

    International Nuclear Information System (INIS)

    Liu Liming; Wang Jifeng; Song Gang

    2004-01-01

    Welding of AZ31B magnesium alloy was carried out using hybrid laser-TIG (LATIG) welding, laser beam welding (LBW) and gas tungsten arc (TIG) welding. The weldability and microstructure of magnesium AZ31B alloy welded using LATIG, LBW and TIG were investigated by OM and EMPA. The experimental results showed that the welding speed of LATIG was higher than that of TIG, which was caught up with LBW. Besides, the penetration of LATIG doubles that of TIG, and was four times that of LBW. In addition, arc stability was improved in hybrid of laser-TIG welding compared with using the TIG welding alone, especially at high welding speed and under low TIG current. It was found that the heat affect zone of joint was only observed in TIG welding, and the size of grains in it was evidently coarse. In fusion zone, the equiaxed grains exist, whose size was the smallest welded by LBW, and was the largest by TIG welding. It was also found that Mg concentration of the fusion zone was lower than that of the base one by EPMA in three welding processes

  11. Effects of post weld heat treatment and weld overlay on the residual stress and mechanical properties in dissimilar metal weld

    Energy Technology Data Exchange (ETDEWEB)

    Campos, Wagner R.C.; Ribeiro, Vladimir S.; Vilela, Alisson H.F.; Almeida, Camila R.O.; Rabello, Emerson G., E-mail: wrcc@cdtn.br, E-mail: camilarezende.cr@gmail.com, E-mail: egr@cdtn.br, E-mail: vladimirsoler@hotmail.com, E-mail: ahfv02@outlook.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The object of this work is a dissimilar metal weld (DMW) pipe joint between carbon steel (A-106 Gr B) and stainless steel (A-312 TP316L) pipes and filler metals of Nickel alloy (82/182), which find wide application in the field of chemical, oil, petroleum industries, fossil fuel and nuclear power plant. A lot of the failures that have occurred in dissimilar metal welded are affected greatly by residual stresses. Residual stress is often a cause of premature failure of critical components under normal operation of welded components. Several methods have been tested and developed for removing the tensile residual stresses. The aim of the methods is to reduce the tensile stress state or to create compressive stresses at a predefined area, such as the inner surface of a welded pipe joint. Post weld heat treatment (PWHT) and weld overlay (WOL) are two of the residual stress mitigation methods which reduce the tensile residual stress, create compressive stresses and arrest crack initiation and crack growth. The technique used to substantially minimized or eliminated this failure development in the root weld is the post weld heat treatments (stress relief heat treatment) or the weld overlay. In this work was studied the effectiveness in reducing internal residual stress in dissimilar metal welded pipe joints subjected to post weld heat treatment and weld overlay, measurement by hole-drilling strain-gage method of stress relaxation. Also held was mechanical characterization of the welded pipe joint itself. (author)

  12. Effects of post weld heat treatment and weld overlay on the residual stress and mechanical properties in dissimilar metal weld

    International Nuclear Information System (INIS)

    Campos, Wagner R.C.; Ribeiro, Vladimir S.; Vilela, Alisson H.F.; Almeida, Camila R.O.; Rabello, Emerson G.

    2017-01-01

    The object of this work is a dissimilar metal weld (DMW) pipe joint between carbon steel (A-106 Gr B) and stainless steel (A-312 TP316L) pipes and filler metals of Nickel alloy (82/182), which find wide application in the field of chemical, oil, petroleum industries, fossil fuel and nuclear power plant. A lot of the failures that have occurred in dissimilar metal welded are affected greatly by residual stresses. Residual stress is often a cause of premature failure of critical components under normal operation of welded components. Several methods have been tested and developed for removing the tensile residual stresses. The aim of the methods is to reduce the tensile stress state or to create compressive stresses at a predefined area, such as the inner surface of a welded pipe joint. Post weld heat treatment (PWHT) and weld overlay (WOL) are two of the residual stress mitigation methods which reduce the tensile residual stress, create compressive stresses and arrest crack initiation and crack growth. The technique used to substantially minimized or eliminated this failure development in the root weld is the post weld heat treatments (stress relief heat treatment) or the weld overlay. In this work was studied the effectiveness in reducing internal residual stress in dissimilar metal welded pipe joints subjected to post weld heat treatment and weld overlay, measurement by hole-drilling strain-gage method of stress relaxation. Also held was mechanical characterization of the welded pipe joint itself. (author)

  13. Influence of Welding Process and Post Weld Heat Treatment on Microstructure and Pitting Corrosion Behavior of Dissimilar Aluminium Alloy Welds

    Science.gov (United States)

    Venkata Ramana, V. S. N.; Mohammed, Raffi; Madhusudhan Reddy, G.; Srinivasa Rao, K.

    2018-03-01

    Welding of dissimilar Aluminum alloy welds is becoming important in aerospace, shipbuilding and defence applications. In the present work, an attempt has been made to weld dissimilar aluminium alloys using conventional gas tungsten arc welding (GTAW) and friction stir welding (FSW) processes. An attempt was also made to study the effect of post weld heat treatment (T4 condition) on microstructure and pitting corrosion behaviour of these welds. Results of the present investigation established the differences in microstructures of the base metals in T4 condition and in annealed conditions. It is evident that the thickness of the PMZ is relatively more on AA2014 side than that of AA6061 side. In FS welds, lamellar like shear bands are well noticed on the top of the stir zone. The concentration profile of dissimilar friction stir weld in T4 condition revealed that no diffusion has taken place at the interface. Poor Hardness is observed in all regions of FS welds compared to that of GTA welds. Pitting corrosion resistance of the dissimilar FS welds in all regions was improved by post weld heat treatment.

  14. Study on Laser Welding Process Monitoring Method

    OpenAIRE

    Knag , Heeshin

    2017-01-01

    International audience; In this paper, a study of quality monitoring technology for the laser welding was conducted. The laser welding and the industrial robotic systems were used with robot-based laser welding systems. The laser system used in this study was 1.6 kW fiber laser, while the robot system was Industrial robot (pay-load : 130 kg). The robot-based laser welding system was equipped with a laser scanner system for remote laser welding. The welding joints of steel plate and steel plat...

  15. Study on Laser Welding Process Monitoring Method

    OpenAIRE

    Heeshin Knag

    2016-01-01

    In this paper, a study of quality monitoring technology for the laser welding was conducted. The laser welding and the industrial robotic systems were used with robot-based laser welding systems. The laser system used in this study was 1.6 kW fiber laser, while the robot system was Industrial robot (pay-load : 130 kg). The robot-based laser welding system was equipped with a laser scanner system for remote laser welding. The welding joints of steel plate and steel plate coated with zinc were ...

  16. Automatic Control Of Length Of Welding Arc

    Science.gov (United States)

    Iceland, William F.

    1991-01-01

    Nonlinear relationships among current, voltage, and length stored in electronic memory. Conceptual microprocessor-based control subsystem maintains constant length of welding arc in gas/tungsten arc-welding system, even when welding current varied. Uses feedback of current and voltage from welding arc. Directs motor to set position of torch according to previously measured relationships among current, voltage, and length of arc. Signal paths marked "calibration" or "welding" used during those processes only. Other signal paths used during both processes. Control subsystem added to existing manual or automatic welding system equipped with automatic voltage control.

  17. Nuclear fuel rod end plug weld inspection

    International Nuclear Information System (INIS)

    Parker, M. A.; Patrick, S. S.; Rice, G. F.

    1985-01-01

    Apparatus and method for testing TIG (tungsten inert gas) welds of end plugs on a sealed nuclear reactor fuel rod. An X-ray fluorescent spectrograph testing unit detects tungsten inclusion weld defects in the top end plug's seal weld. Separate ultrasonic weld inspection system testing units test the top end plug's seal and girth welds and test the bottom end plug's girth weld for penetration, porosity and wall thinning defects. The nuclear fuel rod is automatically moved into and out from each testing unit and is automatically transported between the testing units by rod handling devices. A controller supervises the operation of the testing units and the rod handling devices

  18. Welding wires for high-tensile steels

    International Nuclear Information System (INIS)

    Laz'ko, V.E.; Starova, L.L.; Koval'chuk, V.G.; Maksimovich, T.L.; Labzina, I.E.; Yadrov, V.M.

    1993-01-01

    Strength of welded joints in arc welding of high-tensile steels of mean and high thickness by welding wires is equal to approximately 1300 MPa in thermohardened state and approximately 600 MPa without heat treatment. Sv-15Kh2NMTsRA-VI (EhK44-VI) -Sv-30Kh2NMTsRA-VI (EkK47-VI) welding wires are suggested for welding of medium-carbon alloyed steels. These wires provide monotonous growth of ultimate strength of weld metal in 1250-1900 MPa range with increase of C content in heat-treated state

  19. The Application of the Method of Continuous Casting for Manufacturing of Welding Wire AMg6

    International Nuclear Information System (INIS)

    Azhazha, V.M.; Sverdlov, V.Ya.; Kondratov, A.A.; Rudycheva, T.Yu.

    2007-01-01

    The method of manufacturing semifinished item of high alloyed of aluminum, silver and copper alloys has been investigated on the basis of the continuous casting method. The sample of aluminum alloy AMg6 consist of small grains with the vios-cut dimension ∼ 15 mkm and which are stretched in the direction of longitudinal axis of the sample Such microstructure is favourable for plastic deformation of the sample. Welding wire which meets the demands of standards of commercial welding wires of this brand has been produced by the drawing from the sample

  20. Upgraded HFIR Fuel Element Welding System

    International Nuclear Information System (INIS)

    Sease, John D.

    2010-01-01

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.

  1. Development of technique for laser welding of biological tissues using laser welding device and nanocomposite solder.

    Science.gov (United States)

    Gerasimenko, A; Ichcitidze, L; Podgaetsky, V; Ryabkin, D; Pyankov, E; Saveliev, M; Selishchev, S

    2015-08-01

    The laser device for welding of biological tissues has been developed involving quality control and temperature stabilization of weld seam. Laser nanocomposite solder applied onto a wound to be weld has been used. Physicochemical properties of the nanocomposite solder have been elucidated. The nature of the tissue-organizing nanoscaffold has been analyzed at the site of biotissue welding.

  2. Welding of Thin Steel Plates by Hybrid Welding Process Combined TIG Arc with YAG Laser

    Science.gov (United States)

    Kim, Taewon; Suga, Yasuo; Koike, Takashi

    TIG arc welding and laser welding are used widely in the world. However, these welding processes have some advantages and problems respectively. In order to improve problems and make use of advantages of the arc welding and the laser welding processes, hybrid welding process combined the TIG arc with the YAG laser was studied. Especially, the suitable welding conditions for thin steel plate welding were investigated to obtain sound weld with beautiful surface and back beads but without weld defects. As a result, it was confirmed that the shot position of the laser beam is very important to obtain sound welds in hybrid welding. Therefore, a new intelligent system to monitor the welding area using vision sensor is constructed. Furthermore, control system to shot the laser beam to a selected position in molten pool, which is formed by TIG arc, is constructed. As a result of welding experiments using these systems, it is confirmed that the hybrid welding process and the control system are effective on the stable welding of thin stainless steel plates.

  3. TECHNOLOGICAL ISSUES IN MECHANISED FEED WIG/TIG WELDING SURFACING OF WELDING

    Directory of Open Access Journals (Sweden)

    BURCA Mircea

    2016-09-01

    manual welding tests in the light of using the process for welding surfacing being known that in such applications mechanised operations are recommended whenever possible given the latter strengths i.e. increased productivity and quality deposits. The research also aims at achieving a comparative a study between wire mechanised feed based WIG manual welding and the manual rod entry based manual welding in terms of geometry deposits, deposits aesthetics, operating technique, productivity, etc . In this regard deposits were made by means of two welding procedures, and subsequently welding surfacing was made with the optimum values of the welding parameters in this case.

  4. Nitrogen And Oxygen Amount In Weld After Welding With Micro-Jet Cooling

    Directory of Open Access Journals (Sweden)

    Węgrzyn T.

    2015-06-01

    Full Text Available Micro-jet cooling after welding was tested only for MIG welding process with argon, helium and nitrogen as a shielded gases. A paper presents a piece of information about nitrogen and oxygen in weld after micro-jet cooling. There are put down information about gases that could be chosen both for MIG/MAG welding and for micro-jet process. There were given main information about influence of various micro-jet gases on metallographic structure of steel welds. Mechanical properties of weld was presented in terms of nitrogen and oxygen amount in WMD (weld metal deposit.

  5. Mechanical and electrochemical characteristics with welding materials in robotic MIG welding of dissimilar Al alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong Jong; Han, Min Su; Woo, Yong Bin [Mokpo Maritime Univ., Mokpo (Korea, Republic of)

    2013-05-15

    In this study, mechanical and electrochemical characteristics with welding material in MIG welded with ROBOT for dissimilar Al alloys were investigated using various experiment methods. The MIG welding by ROBOT with ER5183 and ER5556 for the 5456-H116 and 6061-T6 Al alloy were carried out. The hardness of welding zone was lower than that of base metal. In electrochemical experiment, ER5183 welding material presented excellent characteristics. The yield strength and maximum tensile strength in welding with welding material of ER5183 presented lower value than those of ER5556. The elongation and time-to-fracture showed the opposite results.

  6. Improvement of the abnormal diagnosis technology by the development of an abnormal parts assignment system for the engineered safety features actuating system of the HTTR

    International Nuclear Information System (INIS)

    Hirato, Yoji; Kozawa, Takayuki; Saito, Kenji

    2015-01-01

    The safety protection sequence panel of HTTR is a control panel to actuate an engineering safety system for protecting the reactor core, reactor coolant pressure boundary, and containment vessel boundary at the time of an accident of the nuclear reactor facilities. The safety code stipulates that the control panel should receive safety check at a frequency of once a month during reactor operation. When abnormality has been found, it is required to eliminate its causes and restore normal operation as soon as possible. However, since this control panel is composed of a complex control circuit, the cause check during abnormality requires the confirmation by a knowledgeable person spending quite a lot of time for chart checking, which leads to a delay of restoration. To achieve a rapid restoration, the abnormal part assignment system (APAS), which can specify abnormality instantaneously even by a common operator, was developed. It has been confirmed that with this system, rapid initial response and prompt restoration can be effectively made. (A.O.)

  7. Heat transfer characteristics evaluation of heat exchangers of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system (Contract research)

    International Nuclear Information System (INIS)

    Shimizu, Akira; Ohashi, Hirofumi; Kato, Michio; Hayashi, Koji; Aita, Hideki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Morisaki, Norihiro; Sakaki, Akihiro; Maeda, Yukimasa; Sato, Hiroyuki; Inagaki, Yoshiyuki; Hanawa, Hiromi; Fujisaki, Katsuo; Yonekawa, Hideo

    2005-06-01

    Connection of hydrogen production system by steam reforming of methane to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been surveyed until now. Mock-up test facility of this steam reforming system with full-scale reaction tube was constructed in FY 2001 and hydrogen of 120 Nm 3 /h was successfully produced in overall performance test. Totally 7 times operational tests were performed from March 2002 to December 2004. A lot of operational test data on heat exchanges were obtained in these tests. In this report specifications and structures of steam reformer, steam superheater, steam generator, condenser, helium gas cooler, feed gas heater and feed gas superheater were described. Heat transfer correlation equations for inside and outside tube were chosen from references. Spreadsheet programs were newly made to evaluate heat transfer characteristics from measured test data such as inlet and outlet temperature pressure and flow-rate. Overall heat-transfer coefficients obtained from the experimental data were compared and evaluated with the calculated values with heat transfer correlation equation. As a result, actual measurement values of all heat exchangers gave close agreement with the calculated values with correlation equations. Thermal efficiencies of the heat exchangers were adequate as they were well accorded with design value. (author)

  8. Investigation on cause of outage of Wide Range Monitor (WRM) in High Temperature engineering Test Reactor (HTTR). Post Irradiation Examination (PIE) toward investigation of the cause

    International Nuclear Information System (INIS)

    Shinohara, Masanori; Saito, Kenji; Takada, Shoji; Ishimi, Akihiro; Katsuyama, Kozo; Motegi, Toshihiro

    2012-08-01

    An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, two experimental investigations were carried out to reveal the cause of the outage by specifying the damaged part causing the event in the WRM. The one is a post irradiation examination using the X-ray computed tomography scanner in Fuels Monitoring Facility (FMF) to specify the damaged part in the WRM. The other is an experiment using a mock-up simulating the WRM fabricated by the fabricator. The characteristic impedance of the damaged WRM was measured by Time Domain Reflectometry, which was compared with that of the mock-up, which could narrow down the damaged part in the WRM. This report summarized the results of the PIE and the experimental investigation using the mock-up to reveal the cause of outage of WRM. (author)

  9. Investigation on cause of malfunction of wide range monitor (WRM) in high temperature engineering test reactor (HTTR). Sample tests and destructive tests

    International Nuclear Information System (INIS)

    Shinohara, Masanori; Saito, Kenji; Haga, Hiroyuki; Sasaki, Shinji; Katsuyama, Kozo; Motegi, Toshihiro; Takada, Kiyoshi; Higashimura, Keisuke; Fujii, Junichi; Ukai, Takayuki; Moriguchi, Yusuke

    2012-11-01

    An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, two experimental investigations were carried out to reveal the cause of the malfunction by specifying the damaged part causing the event in the WRM. One is an experiment using a mock-up sample test which strength degradation at assembly process and heat cycle to specify the damaged part in the WRM. The other is a destructive test in Fuels Monitoring Facility (FMF) to specify the damaged part in the WRM. This report summarized the results of the destructive test and the experimental investigation using the mock-up to reveal the cause of malfunction of WRM. (author)

  10. Analysis of irradiation-induced stresses in coating layers of coated fuel particles for the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Kikuchi, Teruo; Fukuda, Kousaku; Sato, Sadao; Toyota, Junji; Shiozawa, Shusaku; Sawa, Kazuhiro; Kashimura, Satoru.

    1991-07-01

    Irradiation-induced stresses in coating layers of coated fuel particles were analyzed by the MICROS-2 code for the fuels of the High Temperature Engineering Test Reactor (HTTR) under its operating conditions. The analyses were made on the standard core fuel (A-type) and the test fuels comprising the advanced SiC-coated particle fuel (B-1 type) and the ZrC-coated particle fuel (B-2 type). For the B-1 type fuel, the stresses were relieved due to the thicker buffer and SiC layers than for the A type fuel. The slightly decreased thickness of the fourth layer for the B-1 type than for the A type fuel had no significant effect on the stresses. As for the B-2 type fuel, almost the same results as for the B-1 type were obtained under an assumption that the ZrC layer as well as the SiC layer undergoes negligible dimension change within the analysis conditions. The obtained results indicated that the B-1 and B-2 type fuels are better than the A type fuel in terms of integrity against the irradiation-induced stresses. Finally, research subjects for development of the analysis code on the fuel behavior are discussed. (author)

  11. Evaluation of thermal displacement behavior of high temperature piping system in power-up test of HTTR. No. 1 results up to 20 MW operation

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Kojima, Takao; Sumita, Junya; Tachibana, Yukio

    2002-03-01

    Temperature of the primary cooling system of the High Temperature Engineering Test Reactor, HTTR, becomes very high because the coolant temperature at the reactor outlet reaches 950degC, and 400degC at inlet of the reactor. Therefore, it is important to confirm the thermal displacement behavior of the high temperature piping system in the primary cooling system from the viewpoint of the structural integrity. Moreover, newly designed 3-dimensional floating support system is adopted to the cooling system, it is meaningful to verify the thermal displacement behavior of the piping system applied the 3-dimensional floating support system. In the power-up test (up to 20 MW operation), thermal displacement behavior of the high temperature piping system was measured. This paper describes the experimental and analytical results of thermal displacement characteristics of the high temperature piping system. The results showed that the resistance force induced from the supporting system effects to the thermal displacement behavior of cooling system, and the analytical results have a good agreement with the experimental results by optimizing the resistant force of the floating support system. Additionally, structural integrity at the 30 MW operation was confirmed by the analysis. (author)

  12. HTTR(高温工学試験研究炉)の試験・運転と技術開発; 2010年度

    OpenAIRE

    高温工学試験研究炉部

    2011-01-01

    日本原子力研究開発機構大洗研究開発センターのHTTR(高温工学試験研究炉)は、熱出力30MWの黒鉛減速ヘリウムガス冷却型原子炉で、我が国初の高温ガス炉である。平成22年度は、熱出力9MWにおける安全性実証試験を行い、安全性に関するデータを取得・評価した。本報告書は、平成22年度(2010年)のHTTRの運転と保守及び各種技術開発の状況等について紹介する。

  13. Experimental and simulated strength of spot welds

    DEFF Research Database (Denmark)

    Nielsen, Chris Valentin; Bennedbæk, Rune A.K.; Larsen, Morten B.

    2014-01-01

    Weld strength testing of single spots in DP600 steel is presented for the three typical testing procedures, i.e. tensile-shear, cross-tension and peel testing. Spot welds are performed at two sets of welding parameters and strength testing under these conditions is presented by load......-elongation curves revealing the maximum load and the elongation at break. Welding and strength testing is simulated by SORPAS® 3D, which allows the two processes to be prepared in a combined simulation, such that the simulated welding properties are naturally applied to the simulation of strength testing. Besides...... the size and shape of the weld nugget, these properties include the new strength of the material in the weld and the heat affected zone based on the predicted hardness resulting from microstructural phase changes simulated during cooling of the weld before strength testing. Comparisons between overall...

  14. Automatic monitoring of vibration welding equipment

    Science.gov (United States)

    Spicer, John Patrick; Chakraborty, Debejyo; Wincek, Michael Anthony; Wang, Hui; Abell, Jeffrey A; Bracey, Jennifer; Cai, Wayne W

    2014-10-14

    A vibration welding system includes vibration welding equipment having a welding horn and anvil, a host device, a check station, and a robot. The robot moves the horn and anvil via an arm to the check station. Sensors, e.g., temperature sensors, are positioned with respect to the welding equipment. Additional sensors are positioned with respect to the check station, including a pressure-sensitive array. The host device, which monitors a condition of the welding equipment, measures signals via the sensors positioned with respect to the welding equipment when the horn is actively forming a weld. The robot moves the horn and anvil to the check station, activates the check station sensors at the check station, and determines a condition of the welding equipment by processing the received signals. Acoustic, force, temperature, displacement, amplitude, and/or attitude/gyroscopic sensors may be used.

  15. Advances in welding science and technology

    International Nuclear Information System (INIS)

    David, S.A.; Babu, S.S.; Vitek, J.M.

    1995-01-01

    Over the years, welding has been more of an art than a science, but in the last few decades major advances have taken place in welding science and technology. With the development of new methodologies at the crossroads of basic and applied sciences, enormous opportunities and potential exist to develop a science-based design of composition, structure, and properties of welds with intelligent control and automation of the welding processes. In the last several decades, welding has evolved as an interdisciplinary activity requiring synthesis of knowledge from various disciplines and incorporating the most advanced tools of various basic applied sciences. A series of international conferences and other publications have covered the issues, current trends and directions in welding science and technology. In the last few decades, major progress has been made in (i) understanding physical processes in welding, (ii) characterization of microstructure and properties, and (iii) intelligent control and automation of welding. This paper describes some of these developments

  16. Physical bases for diffusion welding processes optimization

    International Nuclear Information System (INIS)

    Bulygina, S.M.; Berber, N.N.; Mukhambetov, D.G.

    1999-01-01

    One of wide-spread method of different materials joint is diffusion welding. It has being brought off at the expense of mutual diffusion of atoms of contacting surfaces under long-duration curing at its heating and compression. Welding regime in dependence from properties of welding details is defining of three parameters: temperature, pressure, time. Problem of diffusion welding optimization concludes in determination less values of these parameters, complying with requirements for quality of welded joint. In the work experiments on diffusion welding for calculated temperature and for given surface's roughness were carried out. Tests conduct on samples of iron and iron-nickel alloy with size 1·1·1 cm 3 . Optimal regime of diffusion welding of examined samples in vacuum is defined. It includes compression of welding samples, heating, isothermal holding at temperature 650 deg C during 0.5 h and affords the required homogeneity of joint

  17. Laser-GMA Hybrid Pipe Welding System

    National Research Council Canada - National Science Library

    Reutzel, Edward W; Kern, Ludwig; Sullivan, Michael J; Tressler, Jay F; Avalos, Juan

    2007-01-01

    The combination of laser welding with conventional gas metal arc welding technology offers substantial increases in production rate of joining pipe through single-pass joining compared to multi-pass...

  18. Repair welding of fusion reactor components

    International Nuclear Information System (INIS)

    Chin, B.A.

    1993-01-01

    Experiments have shown that irradiated Type 316 stainless steel is susceptible to heat-affected-zone (HAZ) cracking upon cooling when welded using the gas tungsten arc (GTA) process under lateral constraint. The cracking has been hypothesized to be caused by stress-assisted helium bubble growth and rupture at grain boundaries. This study utilized an experimental welding setup which enabled different compressive stresses to be applied to the plates during welding. Autogenous GTA welds were produced in Type 316 stainless steel doped with 256 appm helium. The application of a compressive stress, 55 MPa, during welding suppressed the previously observed catastrophic cracking. Detailed examinations conducted after welding showed a dramatic change in helium bubble morphology. Grain boundary bubble growth along directions parallel to the weld was suppressed. Results suggest that stress-modified welding techniques may be used to suppress or eliminate helium-induced cracking during joining of irradiated materials

  19. Interpretation of aluminum-alloy weld radiography

    Science.gov (United States)

    Duren, P. C.; Risch, E. R.

    1971-01-01

    Report proposes radiographic terminology standardization which allows scientific interpretation of radiographic films to replace dependence on individual judgement and experience. Report includes over 50 photographic pages where radiographs of aluminum welds with defects are compared with prepared weld sections photomacrographs.

  20. Nickel-base alloy overlay weld with improved ultrasonic flaw detection by magnetic stirring welding

    International Nuclear Information System (INIS)

    Takashi, Hirano; Kenji, Hirano; Masayuki, Watando; Takahiro, Arakawa; Minoru, Maeda

    2001-01-01

    Ultrasonic flaw detection is more difficult in Nickel-base alloy welds containing dendrites owing to the decrease ultrasonic transmissibility they cause. The present paper discusses application of magnetic stirring welding as a means for reducing dendrite growth with consequent improvement in ultrasonic transmissibility. Single pass and multi-pass welding tests were conducted to determine optimal welding conditions. By PT and macro observation subsequent to welding was carried out, optimal operation conditions were clarified. Overlay welding tests and UT clearly indicated ultrasonic beam transmissibility in overlay welds to be improved and detection capacity to be greater through application of magnetic stirring welding. Optimal operation conditions were determined based on examination of temper bead effects in the heat affected zone of low alloy steel by application of magnetic stirring welding to the butt welded joints between low alloy and stainless steel. Hardness in this zone of low alloy steel after the fourth layer was less than 350 HV. (author)

  1. Modeling of welded bead profile for rapid prototyping by robotic MAG welding

    Institute of Scientific and Technical Information of China (English)

    CAO Yong; ZHU Sheng; WANG Tao; WANG Wanglong

    2009-01-01

    As a deposition technology, robotic metal active gas(MAG) welding has shown new promise for rapid prototyping (RP) of metallic parts. During the process of metal forming using robotic MAG welding, sectional profile of single-pass welded bead is critical to formed accuracy and quality of metal pans. In this paper, the experiments of single-pass welded bead for rapid prototyping using robotic MAG welding were carried out. The effect of some edge detectors on the cross-sectional edge of welded bead was discussed and curve fitting was applied using leat square fitting. Consequently, the mathematical model of welded bead profile was developed. The experimental results show that good shape could be obtained under suitable welding parameters. Canny operawr is suitable to edge detection of welded bead profile, and the mathematical model of welded bead profile developed is approximately parabola.

  2. The effects of welded joint characteristics on its properties in HDPE thermal fusion welding

    Science.gov (United States)

    Dai, Hongbin; Peng, Jun

    2017-05-01

    In this paper, PE100 pipes with the diameter of 200 mm and the thickness of 11.9 mm were used as material. The welded joints were obtained in different welding pressures with the optimal welding temperature of 220∘C. Reheating process on the welded joints with the temperature of 130∘C was carried out. The joints exhibited X-type, and the cause of X-type joints was discussed. The temperature field in the forming process of welded joints was measured, and tensile and bending tests on welded joints were carried out. The fracture surface of welded joints was observed by scanning electron microscopy (SEM), and crystallinity calculation was taken by X-ray diffraction (XRD). The mechanism of X-type weld profile effects on welded joints properties was analyzed. It was concluded that the mechanical properties of welded joints decrease with the reduced X distance between lines.

  3. Influence of use of ultrasound on the mechanical properties of plated pieces by welding in ultrasonic field

    Directory of Open Access Journals (Sweden)

    Gh. Amza

    2015-07-01

    Full Text Available Plating by welding in an ultrasonic field represents a technological solution to increase resistance to corrosion and / or wear of pieces of the machinery industry. Research has been carried out for two types of parts, namely a piece of flange end type and bonnet type made of AISI 4130 steel, and as filler material for plating was used Inconel 625 Fe developed as electrode wire with a diameter of ø 1.2 / mm. The plating was done by depositing a single layer by welding in ultrasonic field, welding process in Ar 100/ % environment non-consumable tungsten electrode, WIG process, and when using ultrasonic activation it was used a longitudinal and a transverse wave with a frequency of 15 / kHz. For pieces plated by welding there have been made attempts of the hardness and tensile and bend shock.

  4. Effects of Thermocapillary Forces during Welding of 316L-Type Wrought, Cast and Powder Metallurgy Austenitic Stainless Steels

    CERN Document Server

    Sgobba, Stefano

    2003-01-01

    The Large Hadron Collider (LHC) is now under construction at the European Organization for Nuclear Research (CERN). This 27 km long accelerator requires 1248 superconducting dipole magnets operating at 1.9 K. The cold mass of the dipole magnets is closed by a shrinking cylinder with two longitudinal welds and two end covers at both extremities of the cylinder. The end covers, for which fabrication by welding, casting or Powder Metallurgy (PM) was considered, are dished-heads equipped with a number of protruding nozzles for the passage of the different cryogenic lines. Structural materials and welds must retain high strength and toughness at cryogenic temperature. AISI 316L-type austenitic stainless steel grades have been selected because of their mechanical properties, ductility, weldability and stability of the austenitic phase against low-temperature spontaneous martensitic transformation. 316LN is chosen for the fabrication of the end covers, while the interconnection components to be welded on the protrud...

  5. Fusion welding of thin metal foils

    International Nuclear Information System (INIS)

    Casey, H.

    1975-01-01

    Aspects of fusion welding of thin metal foils are reviewed and the current techniques employed at LASL to join foils are described. Techniques for fusion welding approximately 0.025-mm-thick foils of copper, aluminum, and stainless steels have been developed using both electron beam and laser welding equipment. These techniques, together with the related aspects of joint design, tooling and fixturing, joint preparation, and modifications to the commercially available welding equipment, are included in the review. (auth)

  6. Characterization of duplex stainless steel weld metals obtained by hybrid plasma-gas metal arc welding

    OpenAIRE

    Yurtisik,Koray; Tirkes,Suha; Dykhno,Igor; Gur,C. Hakan; Gurbuz,Riza

    2013-01-01

    Despite its high efficiency, autogenous keyhole welding is not well-accepted for duplex stainless steels because it causes excessive ferrite in as-welded duplex microstructure, which leads to a degradation in toughness and corrosion properties of the material. Combining the deep penetration characteristics of plasma arc welding in keyhole mode and metal deposition capability of gas metal arc welding, hybrid plasma - gas metal arc welding process has considered for providing a proper duplex mi...

  7. Corrosion Behavior of Arc Weld and Friction Stir Weld in Al 6061-T6 Alloys

    International Nuclear Information System (INIS)

    Yoon, Byoung Hyun; Kim, Heung Ju; Chang, Woong Seong; Kweon, Young Gak

    2006-01-01

    For the evaluation of corrosion resistance of Al 6061-T6 Alloy, Tafel method and immersion test was performed with Friction Stir Weld(FSW) and Gas Metal Arc Weld(GMAW). The Tafel and immersion test results indicated that GMA weld was severely attacked compared with those of friction stir weld. It may be mainly due to the galvanic corrosion mechanism act on the GMA weld

  8. Internato Longitudinal

    Directory of Open Access Journals (Sweden)

    Marcelo Marcos Piva Demarzo

    Full Text Available O internato médico tem gerado recorrente debate frente às transformações curriculares em andamento no País. A despeito das discussões, um modelo de internato consonante com essas mudanças ainda não foi consistentemente delineado. Neste ensaio, trazemos uma proposta de matriz estruturante para o internato médico. Propomos que o internato médico seja realizado durante os seis anos do curso, de forma longitudinal, tendo como eixo estruturante a clínica da Atenção Básica (AB. Esse modelo de "internato longitudinal" prevê a introdução progressiva na prática clínica, iniciando-se pela AB nos dois primeiros anos, acrescentando-se progressivamente os ambulatórios de especialidades, os estágios hospitalares e demais atividades práticas, alcançando-se, dessa forma, o rol de diversidade e complexidade previsto para o egresso da escola médica.

  9. Formation Mechanisms for Entry and Exit Defects in Bobbin Friction Stir Welding

    Directory of Open Access Journals (Sweden)

    Abbas Tamadon

    2018-01-01

    Full Text Available Bobbin friction stir welding (BFSW is an innovative variant for the solid state welding process whereby a rotating symmetrical tool causes a fully penetrated bond. Despite the process development, there are still unknown variables in the characterization of the process parameters which can cause uncontrolled weld defects. The entry zone and the exit zone consist of two discontinuity-defects and removing them is one of the current challenges for improving the weld quality. In the present research, the characteristic features of the entry and exit defects in the weld structure and formation mechanism of them during the BFSW processing was investigated. Using stacked layers of multi-colour plasticine the material flow, analogous to metal flow, can be visualised. By using different colours as the path markers of the analogue model, the streamline flow can be easily delineated in the discontinuity defects compared with the metal welds. AA6082-T6 aluminium plates and multi-layered plasticine slabs were employed to replicate the entry-exit defects in the metal weld and analogue samples. The fixed-bobbin tool utilized for this research was optimized by adding a thread feature and tri-flat geometry to the pin and closed-end spiral scrolls on both shoulder surfaces. Samples were processed at different rotating and longitudinal speeds to show the degree of dependency on the welding parameters for the defects. The analogue models showed that the entry zone and the exit zone of the BFSW are affected by the inhomogeneity of the material flow regime which causes the ejection or disruption of the plastic flow in the gap between the bobbin shoulders. The trial aluminium welds showed that the elimination of entry-exit defects in the weld body is not completely possible but the size of the defects can be minimized by modification of the welding parameters. For the entry zone, the flow pattern evolution suggested formation mechanisms for a sprayed tail, island zone

  10. 46 CFR 154.665 - Welding procedures.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Welding procedures. 154.665 Section 154.665 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY STANDARDS... Construction § 154.665 Welding procedures. Welding procedure tests for cargo tanks for a design temperature...

  11. 49 CFR 195.224 - Welding: Weather.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Welding: Weather. 195.224 Section 195.224 Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND HAZARDOUS MATERIALS SAFETY... PIPELINE Construction § 195.224 Welding: Weather. Welding must be protected from weather conditions that...

  12. 49 CFR 179.100-9 - Welding.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Welding. 179.100-9 Section 179.100-9... Specifications for Pressure Tank Car Tanks (Classes DOT-105, 109, 112, 114 and 120) § 179.100-9 Welding. (a) All..., appendix W (IBR, see § 171.7 of this subchapter). Welding procedures, welders and fabricators shall be...

  13. 49 CFR 179.400-11 - Welding.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Welding. 179.400-11 Section 179.400-11...-11 Welding. (a) Except for closure of openings and a maximum of two circumferential closing joints in... subchapter). (d) Each welding procedure, welder, and fabricator must be approved. [Amdt. 179-32, 48 FR 27708...

  14. 49 CFR 179.200-10 - Welding.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Welding. 179.200-10 Section 179.200-10... Specifications for Non-Pressure Tank Car Tanks (Classes DOT-111AW and 115AW) § 179.200-10 Welding. (a) All joints... W (IBR, see § 171.7 of this subchapter). Welding procedures, welders and fabricators shall be...

  15. 49 CFR 179.11 - Welding certification.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Welding certification. 179.11 Section 179.11 Transportation Other Regulations Relating to Transportation PIPELINE AND HAZARDOUS MATERIALS SAFETY... Design Requirements § 179.11 Welding certification. (a) Welding procedures, welders and fabricators shall...

  16. 30 CFR 75.1729 - Welding operations.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Welding operations. 75.1729 Section 75.1729 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Miscellaneous § 75.1729 Welding operations. Welding...

  17. 30 CFR 77.408 - Welding operations.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Welding operations. 77.408 Section 77.408 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH... for Mechanical Equipment § 77.408 Welding operations. Welding operations shall be shielded and the...

  18. 49 CFR 179.220-10 - Welding.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Welding. 179.220-10 Section 179.220-10... Specifications for Non-Pressure Tank Car Tanks (Classes DOT-111AW and 115AW) § 179.220-10 Welding. (a) All joints... of this subchapter). Welding procedures, welders, and fabricators shall be approved. (b) Radioscopy...

  19. Welding metallurgy of austenitic stainless steels

    International Nuclear Information System (INIS)

    Ibrahim, A.N.

    1983-01-01

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

  20. Low Speed Control for Automatic Welding

    Science.gov (United States)

    Iceland, W. E.

    1982-01-01

    Amplifier module allows rotating positioner of automatic welding machine to operate at speeds below normal range. Low speeds are precisely regulated by a servomechanism as are normal-range speeds. Addition of module to standard welding machine makes it unnecessary to purchase new equipment for low-speed welding.

  1. New process for weld metal reliability

    International Nuclear Information System (INIS)

    Hebel, A.G.

    1985-01-01

    The industry-wide nature of weld cracking alerts one to the possibility that there is a fundamental law being overlooked. And in overlooking this law, industry is unable to counteract it. That law mandates that restraint during welding causes internal stress; internal stress causes weld metal to crack. Component restraint during welding, according to the welding standard, is the major cause of weld metal failures. When the metal working industry accepts this fact and begins to counter the effects of restraint, the number of weld failures experienced fall dramatically. Bonal Technologies, inc., of Detroit, has developed the first consistently effective non-thermal process to relieve stress caused by restraint during welding. Bonal's patented Mets-Lax sub-resonant stress relief acts as a restraint neutralizer when used during welding. Meta-Lax weld conditioning produces a finer more uniform weld grain structure. A finer, more uniform grain structure is a clear metallurgical indication of improved mechanical weld properties. Other benefits like less internal stress, and less warpage are also achieved

  2. Resistance welding equipment manufacturing capability for exports

    Energy Technology Data Exchange (ETDEWEB)

    Sastry, V.S.; Raju, Y.S.; Somani, A.K.; Setty, D.S.; Rameswara Raw, A.; Hermantha Rao, G.V.S.; Jayaraj, R.N. [Nuclear Fuel Complex, Dept. of Atomic Energy, Hyderbad (India)

    2010-07-01

    Indian Pressurised Heavy Water Reactor (PHWR) fuel bundle is fully welded and is unique in its design. Appendage welding, end closure welding, and end plate welding is carried out using resistance welding technique. Out of many joining processes available, resistance-welding process is reliable, environment friendly and best suitable for mass production applications. Nuclear Fuel Complex (NFC), an industrial unit is established in Hyderabad, under the aegis of the Dept of Atomic Energy to manufacture fuel for Pressurised Heavy Water Reactors. From inception, NFC has given importance for self-reliance and indigenization with respect to manufacturing process and equipment. Sintering furnaces, centreless grinders, appendage-welding machines, end-closure welding equipment and end-plate welding equipments, which were initially imported, are either indigenized or designed and manufactured in house. NFC has designed, manufactured a new appendage-welding machine for manufacturing 37 element fuel bundles. Recently NFC has bagged an order from IAEA through international bidding for design, manufacture, supply, erection and commissioning of end-closure welding equipment. The paper gives in detail the salient features of these welding equipment. (author)

  3. Fracture toughness of stainless steel welds

    International Nuclear Information System (INIS)

    Mills, W.J.

    1985-11-01

    The effects of temperature, composition and weld-process variations on the fracture toughness behavior for Types 308 and 16-8-2 stainless steel (SS) welds were examined using the multiple-specimen J/sub R/-curve procedure. Fracture characteristics were found to be dependent on temperature and weld process but not on filler material. Gas-tungsten-arc (GTA) welds exhibited the highest fracture toughness, a shielded metal-arc (SMA) weld exhibited an intermediate toughness and submerged-arc (SA) welds yielded the lowest toughness. Minimum-expected fracture properties were defined from lower-bound J/sub c/ and tearing modulus values generated here and in previous studies. Fractographic examination revealed that microvoid coalescence was the operative fracture mechanism for all welds. Second phase particles of manganese silicide were found to be detrimental to the ductile fracture behavior because they separated from the matrix during the initial stages of plastic straining. In SA welds, the high density of inclusions resulting from silicon pickup from the flux promoted premature dimple rupture. The weld produced by the SMA process contained substantially less manganese silicide, while GTA welds contained no silicide inclusions. Delta ferrite particles present in all welds were substantially more resistant to local failure than the silicide phase. In welds containing little or no manganese silicide, delta ferrite particles initiated microvoid coalescence but only after extensive plastic straining

  4. Multipass welding of nuclear reactor components - computations

    International Nuclear Information System (INIS)

    Hedblom, E.

    2002-01-01

    The finite element method is used to compare different welding procedures. The simulations are compared with measurements. Two different geometries and two different welding procedures are evaluated. It is found that a narrow gap weld gives smaller tensile residual axial stresses on the inside of the pipe. This is believed to reduce the risk for intergranular stress corrosion cracking

  5. Performance of mesh seam welds in tailor welded blanks; Terado blank yo mash seam yosetsubu no tokusei

    Energy Technology Data Exchange (ETDEWEB)

    Uchihara, M; Takahashi, M; Kurita, M; Hirose, Y; Fukui, K [Sumitomo Metal Industries, Ltd., Osaka (Japan)

    1997-10-01

    Formability, fatigue properties and corrosion behavior of mash seam welded steel sheets were investigated and the results were compared with laser weld. The stretch formability of mash seam weld and laser weld were same level. Mash seam weld however, showed slightly smaller formability in hole expansion test. The fatigue strength of mash seam welds was lower than that of laser welds in case of differential thickness joints. Corrosion was apt to initiate at weld in both mash seam and laser weld with E-coat. The corrosion resistance of welds was improved by using zinc coated steel. 3 refs., 14 figs., 2 tabs.

  6. Advances in automatic welding control

    International Nuclear Information System (INIS)

    White, D.; Woodacre, A.; Taylor, A.F.

    1972-01-01

    The development at the Reactor Fuel Element Laboratories, UKAEA Springfields, of a computer-based welding process control system, was aimed initially at the TIG welding of the end seals of nuclear fuel elements. The system provides for mixed multi-station operation with on-line real-time capability and can be used either as a research tool or for production requirements at competitive costs. The operation of the control system, the form of power source, and the servo motor control units are described. Typically, continuous or pulse-arc welding sequences can be digitally programmed on 0.1 sec increments, with current in 0.5 A increments up to a maximum of 256 A; up to three servo motors can be operated with speeds selected in 0.1 percent increments of their maximum. Up to six welding parameters can be monitored digitally at speeds from once every 10 msec. Some applications are described and it is shown that the equipment has wider uses outside the nuclear fuel element field. High quality industrial welding requirements can also be met and the system is not limited to the TIG process

  7. Advances in automatic welding control

    International Nuclear Information System (INIS)

    White, D.; Woodacre, A.; Taylor, A.F.

    1972-01-01

    The development at the Reactor Fuel Element Laboratories, UKAEA Springfields, of a computer-based welding process control system, was aimed initially at the TIG welding of the end seals of nuclear fuel elements. The system provides for mixed multi-station operation with on-line real-time capability and can be used either as a research tool or for production requirements at competitive costs. The operation of the control system, the form of power source and servo motor control units are described. Typically, continuous or pulse-arc welding sequences can be digitally programmed on 0.1 sec increments, with current in 0.5 A increments up to a maximum of 256 A; up to three servo motors can be operated with speeds selected in 0.1% increments of their maximum. Up to six welding parameters can be monitored digitally at speeds from once every 10 msec. Some applications are described and it is shown that the equipment has wider uses outside the nuclear fuel element field. High quality industrial welding requirements can also be met and the system is not limited to the TIG process. (author)

  8. Laser weld process monitoring and control using chromatic filtering of thermal radiation from a weld pool

    International Nuclear Information System (INIS)

    Kim, Cheol Jung; Kim, Min Suk; Baik, Sung Hoon; Chung, Chin Man

    2000-06-01

    The application of high power Nd: YAG lasers for precision welding in industry has been growing quite fast these days in diverse areas such as the automobile, the electronics and the aerospace industries. These diverse applications also require the new developments for the precise control and the reliable process monitoring. Due to the hostile environment in laser welding, a remote monitoring is required. The present development relates in general to weld process monitoring techniques, and more particularly to improved methods and apparatus for real-time monitoring of thermal radiation of a weld pool to monitor a size variation and a focus shift of the weld pool for weld process control, utilizing the chromatic aberration of focusing lens or lenses. The monitoring technique of the size variation and the focus shift of a weld pool is developed by using the chromatic filtering of the thermal radiation from a weld pool. The monitoring of weld pool size variation can also be used to monitor the weld depth in a laser welding. Furthermore, the monitoring of the size variation of a weld pool is independent of the focus shift of a weld pool and the monitoring of the focus shift of a weld pool is independent of the size variation of a weld pool

  9. Laser weld process monitoring and control using chromatic filtering of thermal radiation from a weld pool

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Cheol Jung; Kim, Min Suk; Baik, Sung Hoon; Chung, Chin Man

    2000-06-01

    The application of high power Nd: YAG lasers for precision welding in industry has been growing quite fast these days in diverse areas such as the automobile, the electronics and the aerospace industries. These diverse applications also require the new developments for the precise control and the reliable process monitoring. Due to the hostile environment in laser welding, a remote monitoring is required. The present development relates in general to weld process monitoring techniques, and more particularly to improved methods and apparatus for real-time monitoring of thermal radiation of a weld pool to monitor a size variation and a focus shift of the weld pool for weld process control, utilizing the chromatic aberration of focusing lens or lenses. The monitoring technique of the size variation and the focus shift of a weld pool is developed by using the chromatic filtering of the thermal radiation from a weld pool. The monitoring of weld pool size variation can also be used to monitor the weld depth in a laser welding. Furthermore, the monitoring of the size variation of a weld pool is independent of the focus shift of a weld pool and the monitoring of the focus shift of a weld pool is independent of the size variation of a weld pool.

  10. Material flow in butt friction stir welds in AA2024-T3

    International Nuclear Information System (INIS)

    Schmidt, H.N.B.; Dickerson, T.L.; Hattel, J.H.

    2006-01-01

    The properties of a workpiece joined by friction stir welding (FSW) are directly related to the material flow around the tool. In the present work, the material flow is investigated by traditional metallography as well as X-ray and computer tomography (CT). By introducing a thin copper strip in the workpiece and welding through it, thus, acting as a marker material, detailed information about the flow field is gathered. The two- and three-dimensional CT images are used in parallel with micrographs for visualization of the flow field. Two procedures for estimating the average velocities for material flowing through the shear layer are presented. The procedures depend on the configuration of marker material relative to the welding direction, i.e. longitudinal and transverse. As such, the present work constitutes the first attempt in the literature to estimate flow velocities in FSW based on thorough experimental investigations

  11. Estimation of weld nugget temperature by thermography method in resistance projection welding process

    International Nuclear Information System (INIS)

    Setty, D.S.; Rameswara Roa, A.; Hemantha Rao, G.V.S.; Jaya Raj, R.N.

    2008-01-01

    In the Pressurized Heavy Water Reactor (PHWR) fuel manufacturing, zirconium alloy appendages like spacer and bearing pads are welded to the thin wall zirconium alloy fuel tubes by using resistance projection welding process. Out of many joining processes available, resistance-welding process is reliable, environment friendly and best suitable for mass production applications. In the fuel assembly, spacer pads are used to get the required inter-element spacing and Bearing pads are used to get the required load-bearing surface for the fuel assembly. Performance of the fuel assembly in the reactor is greatly influenced by these weld joint's quality. Phase transformation from α to β phase is not acceptable while welding these tiny appendages. At present only destructive metallography test is available for this purpose. This can also be achieved by measuring weld nugget temperature where in the phase transformation temperature for zirconium alloy material is 853 o C. The temperature distribution during resistance welding of tiny parts cannot be measured by conventional methods due to very small space and short weld times involved in the process. Shear strength, dimensional accuracy and weld microstructures are some of the key parameters used to measure the quality of appendage weld joints. Weld parameters were optimized with the help of industrial experimentation methodology. Individual projection welding by split electrode concept, and during welding on empty tube firm support is achieved on inner side of the tube by using expandable pneumatic mandrel. In the present paper, an attempt was made to measure the weld nugget temperature by thermography technique and is correlated with standard microstructures of zirconium alloy material. The temperature profiles in the welding process are presented for different welding conditions. This technique has helped in measuring the weld nugget temperature more accurately. It was observed that in the present appendage welding

  12. Effect of weld spacing on microstructure and mechanical properties of CLAM electron beam welding joints

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Yutao; Huang, Bo, E-mail: aufa0007@163.com; Zhang, Junyu; Zhang, Baoren; Liu, Shaojun; Huang, Qunying

    2016-11-15

    Highlights: • The welded joints of CLAM steel with different weld spacings have been fabricated with electron beam welding, and a simplified model of CLAM sheet was proposed. • The microstructure and mechanical properties such as microhardness, impact and tensile were investigated at different welding spacing for both conditions of as-welded and post weld heat treatment (PWHT). • The effect of the welding thermal cycle was significantly when the weld spacings were smaller than 4 mm. • When the weld spacing was small enough, the original microstructures would be fragmented with the high heat input. - Abstract: China low activation martensitic (CLAM) steel has been chosen as the primary structural material in the designs of dual function lithium-lead (DFLL) blanket for fusion reactors, China helium cooled ceramic breeder (HCCB) test blanket module (TBM) for ITER and China fusion engineering test reactor (CFETR) blanket. The cooling components of the blankets are designed with high density cooling channels (HDCCs) to remove the high nuclear thermal effectively. Hence, the welding spacing among the channels are small. In this paper, the welded joints of CLAM steel with different weld spacings have been fabricated with electron beam welding (EBW). The weld spacing was designed to be 2 mm, 3 mm, 4 mm, 6 mm and 8 mm. The microstructure and mechanical properties such as microhardness, impact and tensile were investigated at different welding spacing for both conditions of as-welded and post weld heat treatment (PWHT). The PWHT is tempering at 740 °C for 120 min. The results showed that the grain size in the heat affected zone (HAZ) increased with the increasing weld spacing, and the joint with small weld spacing had a better performance after PWHT. This work would give useful guidance to improve the preparation of the cooling components of blanket.

  13. Repair welding and online radiography

    International Nuclear Information System (INIS)

    Nuding, W.; Grimm, R.; Link, R.; Schroeder, P.; Schroeder, G.

    1990-01-01

    The status of a joint project is reported, which is to develop a computerized testing and welding system for repair work in turbine blades. An X-ray radiographic testing device consisting of microfocus tube, manipulator and image processing system, is modified for this purpose so as to offer a greater number of image points scanned for image processing, and to thus achieve a better resolution for reliable detection of even very small defects. The consistency of the X-ray tube performance, which is a pre-requisite for automation, is to be achieved by a wa tercooled, high-duty tube head. The recording of defect coordinates in the repair zone is done for input into a welding robot to be developed by other partners in the project, so as to allow automated welding work. (orig.) [de

  14. Characterization of Friction Stir Welded Tubes by Means of Tube Bulge Test

    International Nuclear Information System (INIS)

    D'Urso, G.; Longo, M.; Giardini, C.

    2011-01-01

    Mechanical properties of friction stir welded joints are generally evaluated by means of conventional tensile test. This testing method might provide insufficient information because maximum strain obtained in tensile test before necking is small; moreover, the application of tensile test is limited when the joint path is not linear or even when the welds are executed on curved surfaces. Therefore, in some cases, it would be preferable to obtain the joints properties from other testing methods. Tube bulge test can be a valid solution for testing circumferential or longitudinal welds executed on tubular workpieces. The present work investigates the mechanical properties and the formability of friction stir welded tubes by means of tube bulge tests. The experimental campaign was performed on tubular specimens having a thickness of 3 mm and an external diameter of 40 mm, obtained starting from two semi-tubes longitudinally friction stir welded. The first step, regarding the fabrication of tubes, was performed combining a conventional forming process and friction stir welding. Sheets in Al-Mg-Si-Cu alloy AA6060 T6 were adopted for this purpose. Plates having a dimension of 225x60 mm were bent (with a bending axis parallel to the main dimension) in order to obtain semi-tubes. A particular care was devoted to the fabrication of forming devices (punch and die) in order to minimize the springback effects. Semi-tubes were then friction stir welded by means of a CNC machine tool. Some preliminary tests were carried out by varying the welding parameters, namely feed rate and rotational speed. A very simple tool having flat shoulder and cylindrical pin was used. The second step of the research was based on testing the welded tubes by means of tube bulge test. A specific equipment having axial actuators with a conical shape was adopted for this study. Some analyses were carried out on the tubes bulged up to a certain pressure level. In particular, the burst pressure and the

  15. Annual report on experimental operation of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2001 fiscal year (Contract research)

    International Nuclear Information System (INIS)

    Hayashi, Koji; Inagaki, Yoshiyuki; Kato, Michio; Fujisaki, Katsuo; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Ohashi, Hirofumi; Katanishi, Shoji; Takada, Shoji; Shimizu, Akira; Morisaki, Norihiro; Sakaki, Akihiro; Maeda, Yukimasa; Sato, Hiroyuki

    2005-06-01

    This is an annual report on the experimental operation of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2001 fiscal year. The first experimental operation was performed during two weeks from March 1, 2002 to March 13, 2002 to test on the thermal hydraulic performance of the steam reformer and also to train the operators. The thermal hydraulic performance test of the steam reformer was performed to evaluate the heat transfer characteristics between helium gas and process gas in the steam reformer. This report is summarized with an overview of the test, the results and its operation records. (author)

  16. Fatigue assessment of a double submerged arc welded gas pipeline

    Energy Technology Data Exchange (ETDEWEB)

    Fazzini, Pablo; Otegui, Jose Luis [Universidad Nacional Mar del Plata, Mar del Plata (Argentina). Instituto de Ciencia y Tecnologia de Materiales (INTEMA); Teutonico, Mauricio; Manfredi, Carlos [GIE S.A., Mar del Plata (Argentina)

    2005-07-01

    An uncommon blowout in a 24'' diameter, 7 mm thick API 5L X52 gas pipeline was due to fracture at the longitudinal double submerged arc weld. Oddly enough for gas pipelines, it was found that fatigue cracks had propagated from a large embedded weld defect of lack of fusion resulting from severe geometrical mismatch between inner and outer weld passes. What makes this failure particularly interesting is that: previous in line inspections failed to detect any defect, no evidence of third party damage was found, and very few large pressure cycles had been recorded during the last 5 years of service, which were believed to be representative of the entire service life of the pipeline. Fatigue tests were carried out to characterize propagation of fatigue cracks in weld metal, it was found that a large Paris exponent made the few large amplitude cycles most contributing to crack propagation. Crack growth path and striation patterns were studied. Fatigue growth was modelled by integrating experimental results and by extrapolating striation spacing in the fracture surface of the failed pipe. Crack growth path and striation patterns were studied. It was found that microstructure discontinuities govern propagation at low {delta}K, but one striation per cycle was produced at large {delta}K, due to a mostly ductile propagation mode. Fatigue growth was modelled by integrating experimental results and by extrapolating striation spacing in the fracture surface of the failed pipe. It was found that in the early life of the line many more large pressure cycles than expected had occurred. Good correspondence between predicted and actual fatigue lives was in this way obtained (author)

  17. Fatigue assessment of a double submerged arc welded gas pipeline

    Energy Technology Data Exchange (ETDEWEB)

    Fazzini, Pablo; Otegui, Jose Luis [Universidad Nacional Mar del Plata, Mar del Plata (Argentina). Instituto de Ciencia y Tecnologia de Materiales (INTEMA); Teutonico, Mauricio; Manfredi, Carlos [GIE S.A., Mar del Plata (Argentina)

    2005-07-01

    An uncommon blowout in a 24'' diameter, 7 mm thick API 5L X52 gas pipeline was due to fracture at the longitudinal double submerged arc weld. Oddly enough for gas pipelines, it was found that fatigue cracks had propagated from a large embedded weld defect of lack of fusion resulting from severe geometrical mismatch between inner and outer weld passes. What makes this failure particularly interesting is that: previous in line inspections failed to detect any defect, no evidence of third party damage was found, and very few large pressure cycles had been recorded during the last 5 years of service, which were believed to be representative of the entire service life of the pipeline. Fatigue tests were carried out to characterize propagation of fatigue cracks in weld metal, it was found that a large Paris exponent made the few large amplitude cycles most contributing to crack propagation. Crack growth path and striation patterns were studied. Fatigue growth was modelled by integrating experimental results and by extrapolating striation spacing in the fracture surface of the failed pipe. Crack growth path and striation patterns were studied. It was found that microstructure discontinuities govern propagation at low {delta}K, but one striation per cycle was produced at large {delta}K, due to a mostly ductile propagation mode. Fatigue growth was modelled by integrating experimental results and by extrapolating striation spacing in the fracture surface of the failed pipe. It was found that in the early life of the line many more large pressure cycles than expected had occurred. Good correspondence between predicted and actual fatigue lives was in this way obtained (author)

  18. Method and device for weld deposit cladding

    International Nuclear Information System (INIS)

    Barger, J.J.

    1977-01-01

    In order to get weld beads of good quality, uniform thickness and faultless transition regions between neighboring beads in weld deposit cladding of metallic workpoeces, it is proposed to use a device in which the electromagnets are arranged adjacent to th zone of molten welding powder and molten metal besides having got suitable supplies for applying the welding powder, the polarity of the magnets being chosen in such a way that the lines of flux between the poles are counteracting the lines of flux surrounding the electrode band because of the welding current. Several variants of arranging the electrodes are presented in detail. (UWI) [de

  19. Grain refinement control in TIG arc welding

    Science.gov (United States)

    Iceland, W. F.; Whiffen, E. L. (Inventor)

    1975-01-01

    A method for controlling grain size and weld puddle agitation in a tungsten electrode inert gas welding system to produce fine, even grain size and distribution is disclosed. In the method the frequency of dc welding voltage pulses supplied to the welding electrode is varied over a preselected frequency range and the arc gas voltage is monitored. At some frequency in the preselected range the arc gas voltage will pass through a maximum. By maintaining the operating frequency of the system at this value, maximum weld puddle agitation and fine grain structure are produced.

  20. Advances in welding science - a perspective

    International Nuclear Information System (INIS)

    David, S.A.; Vitek, J.M.; Babu, S.S.; DebRoy, T.

    1995-01-01

    The ultimate goal of welding technology is to improve the joint integrity and increase productivity. Over the years, welding has been more of an art than a science, but in the last few decades major advances have taken place in welding science and technology. With the development of new methodologies at the crossroads of basic and applied sciences, enormous opportunities and potential exist to develop a science-based tailoring of composition, structure, and properties of welds with intelligent control and automation of the welding processes