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Sample records for tube reactor connected

  1. Pressure tube type reactors

    International Nuclear Information System (INIS)

    Komada, Masaoki.

    1981-01-01

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

  2. Pressure tube reactor

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  3. Pressure tube reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Kumasaka, Katsuyuki.

    1988-01-01

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

  4. Pressure tube reactor

    International Nuclear Information System (INIS)

    Matsumoto, Tomoyuki; Fujino, Michihira.

    1980-01-01

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

  5. Pressure tube reactors

    International Nuclear Information System (INIS)

    Natori, Hisahide.

    1981-01-01

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

  6. Method of repairing pressure tube type reactors

    International Nuclear Information System (INIS)

    Asada, Takashi.

    1983-01-01

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

  7. Pressure tube reactor

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro; Kaneto, Kunikazu.

    1979-01-01

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

  8. Reactor scram device using fluid poison tubes

    International Nuclear Information System (INIS)

    Iwasaki, Toshio; Hasegawa, Koji.

    1979-01-01

    Purpose: To improve the response function in the reactor scram with no wide space by injecting poisons in soluble poison guide tubes to such a liquid level as giving no effect on usual reactor operation. Constitution: Soluble poison guide tubes in a reactor are connected at their upper ends to a buffer tank and at their lower ends to a pressurizer by way of a header and an injection valve. The header is connected by way of a valve with a level meter, one end of which is connected to the buffer tank. During reactor operation, the injection valve is closed and the soluble poisons in the pressurizer vessel is maintained at a pressurized state and, while on the other hand, soluble poisons are injected by way of the header to the lower end of the soluble poison guide tubes by the opening of a valve, which is thereafter closed. Upon scram, a valve is closed to protect the level meter and pressurized poisons are rapidly filled in the guide tubes by the release of the injection valve. (Kawakami, Y.)

  9. N Reactor pressure tube 1350 postirradiation examination

    International Nuclear Information System (INIS)

    Cook, D.J.

    1977-01-01

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

  10. N Reactor pressure tube 2566 postirradiation examination

    International Nuclear Information System (INIS)

    Scott, K.V.

    1978-01-01

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

  11. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1976-01-01

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

  12. MAPLE research reactor beam-tube performance

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Gillespie, G.E.

    1989-05-01

    Atomic Energy of Canada Limited (AECL) has been developing the MAPLE (Multipurpose Applied Physics Lattice Experimental) reactor concept as a medium-flux neutron source to meet contemporary research reactor applications. This paper gives a brief description of the MAPLE reactor and presents some results of computer simulations used to analyze the neutronic performance. The computer simulations were performed to identify how the MAPLE reactor may be adapted to beam-tube applications such as neutron radiography

  13. Bottom nozzle to guide tube connection

    International Nuclear Information System (INIS)

    Bryan, W.J.

    1991-01-01

    This patent describes a nuclear fuel assembly which includes an upper end fitting and a lower end fitting spaced therefrom and connected thereto by elongated guide tubes of one alloy having an open upper end and a closed lower end with spaced fuel element retaining grids mounted on the guide tubes therebetween, the closed lower ends of the guide tubes including a threaded central passageway and the attachment of the guide tubes to the lower end fitting of another alloy. It comprises: an externally threaded bolt with a first end threadably received in the threaded central passageway of the lower end of the guide tube and a head at the other end of the side of the lower end fitting opposite the guide tube; an interruption in the external threads of the bolt which forms a groove which communicates the interior of the guide tube with the side of the lower end fitting opposite the guide tube and enhances its frictional engagement with the threaded central passageway, thereby to hold and attach the guide tube and lower end fitting firmly together, even through a series of temperature cycles

  14. Improving the calandria tubes for CANDU reactors

    International Nuclear Information System (INIS)

    Coleman, C.E.; Fong, R.W.L.; Doubt, G.L.

    1997-01-01

    CANDU calandria tubes are made from annealed Zircaloy-2 sheet formed into a cylinder and welded along its length to make the tube. The current calandria tubes have given exemplary service for many years. With more stringent regulations and the need to accommodate warm cooling water in tropical countries, we started a development program to increase the margins for failure during postulated accidents. These improvements involve increasing the tube strength and optimising the heat-transfer from an excessively hot fuel channel to the cool moderator. If the postulated accident involves a pressure tube break, it would be desirable if the calandria tube withstood the full pressure of the heat-transport system. The weakest link in current calandria tubes is the weld. Thickening the weld can increase the strength by 20% while seamless tubes can be 45% stronger than current tubes. The latter tubes can hold full system pressure for many hours without failure. If during the postulated accident the fuel and pressure tube become excessively hot but do not touch the calandria tube, the radiant heat loss must be maximised. Current calandria tubes have an absorptivity (emissivity) of about 0.2. To protect the fuel and the fuel channel we have devised a finish to the inside surface of the calandria tube that increases the emissivity to 0.7. If during the postulated accident the hot pressure tube touches the cool calandria tube, the contact conductance and the critical heat flux must be optimised to ensure nucleate boiling of the moderator at the outside surface of the calandria tube and therefore efficient exploitation of the moderator as a heat sink. In laboratory tests small ridges on the inside surface and roughening of the outside surface have been shown to increase the margins against failure and increase the possible moderator temperatures thus providing the opportunity to decrease the cost of the moderator heat-exchange system and remove restrictions on reactor operation in

  15. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1975-01-01

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

  16. Control rod guide tube of nuclear reactor

    International Nuclear Information System (INIS)

    Suda, Yoshitaka; Ito, Kenji; Matsumoto, Kunio.

    1994-01-01

    Zr having a residual tensile stress of 3 to 10kg/mm 2 in a circumferential direction is used for the main ingredient of a control guide tube of a nuclear reactor. For this purpose, an appropriate correction method such as a roll-correction, tension-correction and press-correction method is applied to an existent Zr-base alloy tube with no substantial residual stress. If the residual tensile stress in the circumferential direction is smaller than 3kg/mm 2 , an effect sufficient to suppress irradiation growth is not obtainable, if it exceeds 10kg/mm 2 , dimensional changes, cracks or the like occurs locally since the wall thickness of the control rod guide tube is small and, accordingly, this often results in failed products as the control guide tube. (N.H.)

  17. Device for removing a spent reactor core instrument tube

    International Nuclear Information System (INIS)

    Watanabe, Shigeru; Tsuji, Teruaki.

    1980-01-01

    Purpose: To easily and exactly execute works for removing a used reactor core instrument tube to be mounted in a reactor core from the lattice space of the core or for charging the tube into the lattice of the core. Constitution: When fuel assembly is pulled out of a reactor core and a spent reactor core instrument tube is then bent and removed from the core at periodical inspection time, a lower gripping unit integral with an upper gripping unit and a bending unit is provided at the lower end of a hanging rope of a winch, and lowered to the reactor core. Then, the spent reactor core instrument tube is gripped by the upper and lower gripping units, the bending unit is operated, the spent reactor core instrument tube is bent, and the tube is then pulled upwardly by the winch to remove the tube. (Aizawa, K.)

  18. Calandria cooling structure in pressure tube reactor

    International Nuclear Information System (INIS)

    Hyugaji, Takenori; Sasada, Yasuhiro.

    1976-01-01

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

  19. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  20. Feedwater heater tube-to-tubesheet connections

    International Nuclear Information System (INIS)

    Yokell, S.

    1993-01-01

    This paper discusses some practical aspects of expanded, welded, and welded-and-expanded feedwater heater tube-to-tubesheet joints. It outlines elastic-plastic tube expanding theory. It examines uniform-pressure-expanded tube joint strength and correlating roller-expanded joint strength with wall reduction and rolling torque. For materials subject to stress-corrosion cracking (SCC), it recommends heat treating tube ends before expanding. For materials subject to fatigue and tube-end cracking, it advocates two-stage expanding: (1) expanding enough to create firm tube-hole contact over the full tubesheet thickness; and (2) re-expanding at full pressure or torque. The paper emphasizes the desirability of segregating heats of tubing, mapping the tube-heat locations and making the heat map a permanent part of the heater maintenance file. It recommends when to provide TEMA/HEI Power Plant Standard annular grooves for roller-expanding and provides an equation for determining optimum groove width for uniform-pressure expanding. The paper also reviews welding requirements for welds of tubes to tubesheets. The review covers front-face welding before and after expanding and the reasons for welding first. It outlines current thinking about definitions of strength- and seal-welds of front-face welded joint in terms of their functions and load-carrying abilities. It presents a proposal for determining the required size of strength welds for use in Section VIII of the ASME Boiler and Pressure Vessel Code (the Code). It shows why welded-and-expanded feedwater heater tube-to-tubesheet joints should be full-strength and full-depth expanded. It makes recommendations for pressure- and leak-testing. This work also proposes the industry consider butt welding the tubes to the steam-side face of the tubesheet as a regular method of tube joining. The results of a survey of manufacturers practices are appended. 30 refs., 14 figs

  1. Releasing method of connection of control rod and its drive mechanism in a reactor

    International Nuclear Information System (INIS)

    Ishida, Kazuo; Futatsugi, Masao.

    1976-01-01

    Object: To disengage a control rod from a control rod drive device in a boiling water reactor with a minimal failure of the device, when connection there between cannot be released in a normal manner. Structure: First, a part of a piston tube in the control rod drive device is withdrawn externally of a control rod housing and cut. Next, a discharge tool, which is designed to be connected with the cut piston tube, is connected to the remainder of the piston tube within the housing and the aforesaid piston tube is pushed into the index tube. The index tube is then cut by the discharge tool. Thus, the control rod drive device and the control rod may be separated. Thereafter, the control rod may be removed from the top of the reactor container whereas the control rod drive device removed from the bottom thereof. (Ikeda, J.)

  2. Device and method for shortening reactor process tubes

    Science.gov (United States)

    Frantz, Charles E.; Alexander, William K.; Lander, Walter E. B.

    1980-01-01

    This disclosure describes a device and method for in situ shortening of nuclear reactor zirconium alloy process tubes which have grown as a result of radiation exposure. An upsetting technique is utilized which involves inductively heating a short band of a process tube with simultaneous application of an axial load sufficient to cause upsetting with an attendant decrease in length of the process tube.

  3. Steam generator tubing development for commercial fast breeder reactors

    International Nuclear Information System (INIS)

    Sessions, C.E.; Uber, C.F.

    1981-01-01

    The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs

  4. Apparatus for securing structural tubes in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Kerry, J.S.

    1987-01-01

    This patent describes a nuclear reactor fuel assembly having a structural tube with a predetermined inside diameter, a generally cylindrical insert of an axial length substantially smaller than the axial length of the structural tube and having a generally cylindrical passageway of a predetermined diameter smaller than the predetermined inside diameter for providing an effectively reduced inside diameter for the structural tube. The insert comprises: means, having an outside diameter approximately equal to the predetermined inside diameter, for coaxially centering the insert within the structural tube; forming lobes, operable when expanded to locally deform against the structural tube thereby locking the insert within the structural tube

  5. Assessment of beam tube performance for the maple research reactor

    International Nuclear Information System (INIS)

    Lee, A.G.

    1986-06-01

    The MAPLE research reactor is a versatile new research facility that can be adapted to meet the requirements of a variety of reactor applications. A particular group of reactor applications involves the use of beams of radiation extracted from the reactor core via tubes that penetrate through the biological shield and terminate in the reflector surrounding the fuelled core. An assessment is given of the neutron and gamma radiation fields entering beam tubes that are located radially or tangentially with respect to the core

  6. Development of austenitic stainless steel tubes for nuclear reactor cladding

    International Nuclear Information System (INIS)

    Padilha, A.F.; Ferreira, P.I.; Andrade, P.I.; Andrade, A.H.P. de; Meyerhof, S.; Mauricio, J.

    1984-01-01

    In the development of thin wall tubes for nuclear reactor fuel cladding applications, a great number of activities, related to the fabrication process as the qualification are involved. A test program was envisaged to verify the quality of seam welded stainless steel tubes (AISI 304), obtained as a result of an effort by the IPEN-CNEN/SP and the brazilian industry. The relevant aspects involved in the preparation of the tubes and some preliminary test results are presented. (Author) [pt

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

    International Nuclear Information System (INIS)

    Prodea, Iosif; Catana, Alexandru

    2010-01-01

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

  8. Improving the calandria tubes for CANDU reactors

    International Nuclear Information System (INIS)

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

    1997-07-01

    The Zircaloy-2 calandria tube has been improved to guard against abnormal operating conditions. It has been strengthened by either thickening or eliminating the weld to withstand the consequences of a pressure tube rupture. To exploit the moderator as a heat sink, both surfaces have been roughened and the inside surface ridged to maximise heat-transfer from an over-heated fuel channel during a postulated loss of coolant accident. (author)

  9. Visual beam tube inspection at the TRIGA reactor Vienna

    International Nuclear Information System (INIS)

    Boeck, H.; Musilek, A.; Villa, M.

    2006-01-01

    Of the four TRIGA beam tubes two have been visually inspected in 1985. Prior to the inspection the reactor was shut down for 3 weeks. The fuel elements around the beam tubes were removed. Stainless steel dummy elements were inserted in the fuel positions to shield the core radiation. The active part of the Fast Rabbit Tube was removed into the beam tube loading device and transferred to an interim storage: Front dose rate was ∼ 50 mSv/h. Generally the beam tube was very clean, after the last inspection about 30 years ago. A1 cm cut was observed at the beam tube front end. A rigid endoscope was used to check the beam tube's inner surface using a 90 degree deflection objective and photo- and video equipment. The direct dose rate in front of the beam tube was about 30 mSv/h. The beam tube was vacuum cleaned. A corroded shielding tank containing boric acid has leaked. A wooden collimator partially disintegrating due to extreme temperature was removed from beam tube D. Documentation of the inspection for visible defects is produced for later comparison

  10. Conceptual design of a pressure tube light water reactor with variable moderator control

    International Nuclear Information System (INIS)

    Rachamin, R.; Fridman, E.; Galperin, A.

    2012-01-01

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  11. Holographic NDE of pressure tubes for Cirene nuclear reactor

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  12. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  13. Fabrication characteristics of zircaloy tubes for nuclear reactors

    International Nuclear Information System (INIS)

    Haydt, H.M.

    1980-11-01

    The production sequence for zircaloy cladding tubes to be used in nuclear reactors is described, with emphasis on the texture after reduction and on the variation in the hydrides orientation. The qualities requested for the cladding tubes are presented and reference is made to the quality control applied in the process. The destructive tests as well as the final inspection to which those tubes are subjected are related. A Fabrication Quality Project is requested from the manufacturers by reason of what Quality Control Plans are submitted to be clients. At last an evaluation of the quality to be obtained and of the control performed is mentioned. (Author) [pt

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

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

  16. Process and device for change of catalyst in tube reactors

    International Nuclear Information System (INIS)

    Fedders, H.; Cremer, P.; Erben, R.

    1985-01-01

    The change of catalyst in narrow reactor tubes with a height: diameter ratio of at least 30:1 is done by the catalyst filling being driven out against the force of gravity using a pulsating liquid flow. Pauses in the flow of between 0.1 to 1 sec between flow periods of 2 to 20 secs are useful. (orig./PW) [de

  17. Analysis of Reactor Vessel Lower Head Penetration Tube Failure

    International Nuclear Information System (INIS)

    Stempniewicz, Marek

    1999-01-01

    This paper presents results of two studies, performed to investigate the behavior of the reactor vessel penetration tubes in case of relocation of molten material into the tubes. The first study is on the CORVIS drain line experiment 03/1. Results of pre-test calculations are presented, and compared to the later obtained experimental data. The timing of the drain line melting and the velocity of the debris flowing inside the drain line were predicted correctly, but the penetration depth was clearly underestimated. If the calculations are done using different correlation for the melt-to-wall convective heat transfer, the results are closer to the experiment. It cannot however be concluded that the alternative correlation is more appropriate until other uncertainties are clarified. The second study presents calculations performed for GKN Dodewaard CRD, instrument tubes and drain line. Calculations were performed to estimate whether the tubes have a chance to withstand the first attack of the melt and thus postpone vessel failure until the water in the lower plenum evaporates. Calculations were performed assuming that the melt can move into the tubes without any resistance, e.g. presence of water in the tubes was not taken into account. The results indicate that the critical penetration of the GKN vessel, which is most likely to fail, is the drain line. Results also indicate that external flooding should prevent early tube failure, at least in case of low vessel pressure. (author)

  18. Low in reactor creep Zr-base alloy tubes

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  19. Instruments for non-destructive evaluation of advanced test reactor inpile tubes

    International Nuclear Information System (INIS)

    Livingston, R.A.; Beller, L.S.; Edgett, S.M.

    1986-01-01

    The Advanced Test Reactor is a 250 MW LWR used primarily for irradiation testing of materials contained in inpile tubes that pass through the reactor core. These tubes provided the high pressure and temperature water environment required for the test specimens. The reactor cooling water surrounding the inpile tubes is at much lower pressure and temperature. The structural integrity of the inpile tubes is monitored by routine surveillance to ensure against unplanned reactor shutdowns to replace defective inpile tubes. The improved instruments developed for inpile tube surveillance include a bore profilometer, ultrasonic flaw detetion system and bore diameter gauges. The design and function of these improved instruments is presented

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

    International Nuclear Information System (INIS)

    Ueda, Makoto; Aoki, Katsutada.

    1975-01-01

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

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

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

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

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

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

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

  3. Electric arc apparatus for severing split-pin assemblies of guide tubes of nuclear reactors

    International Nuclear Information System (INIS)

    Burns, D.C.; Kauric, C.E.; Persang, J.C.

    1987-01-01

    This patent describes an apparatus for use in the replacement of an old split-pin assembly of a guide tube of a nuclear reactor by a new split-pin assembly, the old split-pin assembly including an old split pin and an old nut securing the old split pin to the guide tube, the old split-pin assembly and the guide tube being radioactive. The apparatus includes a metal disintegration machining tool, the tool having an electrode, means for mounting the tool submerged in a pool of water in engagement with the guide tube and with the old split-pin assembly secured to the guide tube, the tool being so mounted with the electrode in position to coact electrically with the last-named old split-pin assembly but not with the guide tube, and means, connected to the tool, for firing a disintegrating arc between the electrode and the assembly to disintegrate the assembly into readily removable fragments

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

    International Nuclear Information System (INIS)

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

    1988-12-01

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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  6. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-04-15

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

  7. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    International Nuclear Information System (INIS)

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

    1961-04-01

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

  8. Method of reactivity control in pressure tube reactor

    International Nuclear Information System (INIS)

    Fukumura, Nobuo.

    1988-01-01

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

  9. Guide tube insert assembly for use in a nuclear reactor

    International Nuclear Information System (INIS)

    Hopkins, R.J.; Land, J.T.

    1992-01-01

    This patent describes an internals assembly for a nuclear reactor of the type including an upper support plate and an upper core plate, each having apertures for conducting control rod assemblies into an out of fuel assemblies with the apertures of the upper support plate being aligned with the apertures of the upper core plate, a guide tube insert assembly comprising: an elongated tubular body extending between at least one of the aligned apertures formed in the upper support plate and the upper core plate; guide plates within the elongated tubular body, each of the guide plates having a planar surface extending substantially perpendicular to an axial direction of the tubular body; at least one interconnecting means for interconnecting the guide plates into a guide tube insert assembly such that the guide plates are simultaneously mountable within and removable from the elongated body, and the periphery of each of the guide plates is spaced apart from the inner walls of the elongated tubular body at every point when the insert assembly is mounted within the tubular body, and a stabilizing means for securing the lowermost guide plate of the guide tube insert assembly within the elongated tubular body to prevent rotational and lateral movement between the guide tube insert assembly and the tubular body

  10. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1985-01-01

    An expandable antivibration bar for use in stabilizing the U-bend portion of heat transfer tubes in a pressurized water reactor steam generator comprises two adjustable rods connected together by an arcuate connector. The two adjustable rods preferably comprise two mating rod sections having complementary angular sliding surfaces thereon, with means provided to move the rod sections relative to each other along the sliding surfaces so as to expand the rods from a first mated cross-sectional width to a second larger cross-sectional width. The ends of the rod sections have means for aligning the two rod sections and maintaining them in alignment during expansion. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  12. Automated Determination of Oxygen-Dependent Enzyme Kinetics in a Tube-in-Tube Flow Reactor.

    Science.gov (United States)

    Ringborg, Rolf H; Toftgaard Pedersen, Asbjørn; Woodley, John M

    2017-09-08

    Enzyme-mediated oxidation is of particular interest to synthetic organic chemists. However, the implementation of such systems demands knowledge of enzyme kinetics. Conventionally collecting kinetic data for biocatalytic oxidations is fraught with difficulties such as low oxygen solubility in water and limited oxygen supply. Here, we present a novel method for the collection of such kinetic data using a pressurized tube-in-tube reactor, operated in the low-dispersed flow regime to generate time-series data, with minimal material consumption. Experimental development and validation of the instrument revealed not only the high degree of accuracy of the kinetic data obtained, but also the necessity of making measurements in this way to enable the accurate evaluation of high K MO enzyme systems. For the first time, this paves the way to integrate kinetic data into the protein engineering cycle.

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

    Science.gov (United States)

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

    2018-04-01

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

  14. Flow chemistry: intelligent processing of gas-liquid transformations using a tube-in-tube reactor.

    Science.gov (United States)

    Brzozowski, Martin; O'Brien, Matthew; Ley, Steven V; Polyzos, Anastasios

    2015-02-17

    reactive gas in a given reaction mixture. We have developed a tube-in-tube reactor device consisting of a pair of concentric capillaries in which pressurized gas permeates through an inner Teflon AF-2400 tube and reacts with dissolved substrate within a liquid phase that flows within a second gas impermeable tube. This Account examines our efforts toward the development of a simple, unified methodology for the processing of gaseous reagents in flow by way of development of a tube-in-tube reactor device and applications to key C-C, C-N, and C-O bond forming and hydrogenation reactions. We further describe the application to multistep reactions using solid-supported reagents and extend the technology to processes utilizing multiple gas reagents. A key feature of our work is the development of computer-aided imaging techniques to allow automated in-line monitoring of gas concentration and stoichiometry in real time. We anticipate that this Account will illustrate the convenience and benefits of membrane tube-in-tube reactor technology to improve and concomitantly broaden the scope of gas/liquid/solid reactions in organic synthesis.

  15. Detachable connection for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Christiansen, D.W.; Karnesky, R.A.

    1986-01-01

    This patent describes a connection for releasably attaching a socket member to a tube. It consists of: a tube defined by a wall provided with circumferentially spaced openings adjacent one end; a housing having an internal thread formation and laterally spaced fingers at one end. The fingers have pins engagable within the tube openings, a socket member threadably received within the housing and have a tapered portion at one end for urging the fingers radially outwardly. It also contains a split retaining ring interposed between the housing and the socket member and completely encapsulated

  16. Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

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

    1980-01-01

    The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

  17. Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II

    International Nuclear Information System (INIS)

    2010-01-01

    The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  19. Seismic behavior of conxl connections in concrete filled steel Tube ...

    African Journals Online (AJOL)

    This connection consists of collar flange, collar corner, and collar web extension. In this paper, the seismic behavior of these types of connections is investigated using the numerical method. For this purpose, three samples of ConXL connections without concrete filler, with concrete filler and with concrete filler and stiffener ...

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

    International Nuclear Information System (INIS)

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

    2000-09-01

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

  1. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1983 and 1984

    International Nuclear Information System (INIS)

    Tatone, O.S.; Meindl, P.; Taylor, G.F.

    1986-06-01

    A review of the performance of steam generator tubes in water-cooled nuclear power reactors showed that tubes were plugged at 47 (35.6%) of the reactors in 1983 and at 63 (42.6%) of the reactors during 1984. In 1983 and 1984 3291 and 3335 tubes, respectively, were removed from service, about the same as in 1982. The leading causes assigned to tube failure were stress corrosion cracking from the primary side and stress corrosion cracking or intergranular attack from the secondary side. In addition 5668 tubes were repaired for further service by installation of internal sleeves. Most of these were believed to have deteriorated by one of the above mechanisms or by pitting. There is a continuing trend towards high-integrity condenser tube materials at sites cooled by brackish or sea water. 31 refs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-09-01

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

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

    International Nuclear Information System (INIS)

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

    1980-02-01

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

  4. Evaluation of tube to collector connection by hydraulic expansion method in PGV-1000 steam generators

    International Nuclear Information System (INIS)

    Dashti, H.G.; Hashemi, B.; Jahromi, S.A.

    2011-01-01

    Research highlights: → The produced residual stresses in the collector body due to hydraulic expansion method have been compared with explosive method. → The residual stresses were obtained using two methods of FEM and strain gauging tests. → The effect of clearance between tube and collector on the residual stresses was investigated. → The contact stresses between the tube and collector interface were modeled and the required connection strength between tube and collector is estimated based on ASME rules and compared with FE results. - Abstract: Investigations on steam generators failure due to cracking in collector ligaments at perforated parts determined that connection process of the tubes to collector could be one of the main breakdown causes. The stability and strength of tube to collector joint is dependent to the geometry of tube and collector, the joining process and the operational conditions. In this research hydraulic expansion method has been considered as connection method of tube to collector. The Finite Element Method (FEM) was used to simulate the hydraulic expansion process and determine stress condition of the joints. The contact stresses between the tube and collector interface were modeled using contact elements of ANSYS program. Furthermore, the effect of clearance between tube and collector on the residual stresses around of joints was investigated. Some specimens from collector and tube materials were tested at various temperatures and their results were used at rate-independent multi-linear Mises plasticity model for FE analysis. Required connection strength between tube and collector is estimated based on ASME rules and compared with FE results. The results show that the residual tensile stresses could be greatly increased by decreasing of initial clearance. The highest value of residual stresses was observed around of collector holes nevertheless it was considerably lesser than obtained residual stresses in explosive method. The

  5. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1977

    International Nuclear Information System (INIS)

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

    1979-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1977. Failures were reported in 34 of the 79 reactors surveyed. Causes of these failures and inspection and repair procedures designed to deal with them are presented. Although corrosion remained the leading cause of tube failures, specific mechanisms have been identified and methods of dealing with them developed. These methods are being applied and should lead to a reduction of corrosion failures in future. (author)

  6. Thin-walled large-diameter zirconium alloy tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Price, E.G.; Richinson, P.J.

    1978-08-01

    The requirements of the thin-walled large-diameter Zircaloy-2 tubing used in CANDU reactors are reviewed. Strength, residual stress patterns, texture and prior deformation contribute to the stability of these tubes. The extent to which the present manufacturing route meets these requirements is discussed. (author)

  7. On the heat exchange tube failures in steam generators at NPPs with WWER reactors

    International Nuclear Information System (INIS)

    Titov, V.F.; Banyuk, G.F.; Brykov, S.I.

    1992-01-01

    Data on dynamics of failed heat exchanging tube closing in steam generators of NPPs with WWER type reactors for the whole period of their operation are presented. It is shown that the main cause of the tube failures consists in their corrosion cracking under stresses. The effect of chlorine ions on tubes is intensified by the presence of porous sediments on heat exchaning surfaces in quantities exceeding 150 g/m 2

  8. How to secure the connection between thoracostomy tube and drainage system?

    Science.gov (United States)

    Li, Ka Ki Pat; Wong, Kit Shing John; Wong, Yau Hang Henry; Cheng, Ka Lok; So, Fung Ling; Lau, Chu Leung; Kam, Chak Wah

    2014-01-01

    Thoracostomy tube insertion is one of the common bedside procedures in emergency medicine and many acute specialties. Dislodgement of thoracostomy tube from the connection tube of chest drainage system is an important problem with potential complications such as contamination, infection and pneumothorax. Besides, mere loosening can also lead to malfunction. It is a common practice to tape the connection of the system. This study aimed to evaluate the materials and methods of connection of chest drain system to minimize drainage dislodgement. We conducted an experimental study to assess the tightness of the connection with various taping materials and methods. We selected three commonly used adhesive materials (3M™ Transpore™ Medical tape, 3M™ Micropore™ Medical tape, 3M™ Soft Cloth Tape on Liner) and three different methods (cross method, straight method, nylon band) to secure the junction between the thoracostomy tube and the bi-conical adaptor in the drainage system. The measured outcome was the weight causing visible loosening of the junction between thoracotomy tube and the adaptor. For each taping material and taping method, 10 trials were performed. The median weight required to disconnect the junction is 26.22 lb for Transpore™, 31.29 lb for Micropore™ and 32.44 lb for Soft Cloth Tape on Liner. A smaller force was required to disconnect if Transpore™ is used (Ptube to the chest drainage system. Transpore™ is not a recommended material for thoracostomy tube taping.

  9. Connection of leak detectors for duplex tube plates of modular steam generator

    International Nuclear Information System (INIS)

    Banovec, J.; Konarik, M.; Vytopil, O.

    1985-01-01

    The sensors are connected to common line and column conductors. This connection significantly reduces the number of evaluation points and thus also the required number of evaluation unit channels. The reliability of the instrument is increased by each sensor being connected to two separate group conductors. Even upon failure of one of the conductors, the group of tube plates can be identified where a leak occurred. (J.B.)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  11. Development of a helical-coil double wall tube steam generator for 4S reactor

    International Nuclear Information System (INIS)

    Kitajima, Yuko; Maruyama, Shigeki; Jimbo, Noboru; Hino, Takehisa; Sato, Katsuhiko

    2011-01-01

    The 4S, Super-Safe Small and Simple, is a small-sized sodium-cooled fast reactor. A fast reactor usually uses sodium as a coolant to transfer heat from core to turbine/generator system. The heat of the intermediate heat transport system and that of the water stream systems are exchanged by the steam generator (SG) tubes. If the tube failure occurs, a sodium/water reaction could be occurred. To prevent the reaction and enhance safety, a helical-coil-type double wall tube with wire mesh interlayer and continuous monitoring systems of tube failure are applied to the SG of the 4S. The development and general features of this type double wall tube were described in Ref. 1) and Ref. 2). Those paper summarized following results; The tubes studied in these references were straight type. To establish this SG, development of manufacturing method of helical-coil-type double wall tube and validation of the tube failure monitoring system are needed. In this study, three demonstration tests have been performed; welding test of the double wall tube to manufacture the tubes with 70-80m length, assembling test of the helical-coil tube, and confirmation test of the tube processing system using the fabricated helical-coil tubes. As a result, following technologies have been successfully established. (1) Development of the welding techniques for manufacturing of the helical-coil-type double wall tube with wire mesh interlayer. (2) The confirmation test for manufacturing the helical coil tube of the SG. (author)

  12. Infill Panels and the tube connection in timber frames

    NARCIS (Netherlands)

    Leijten, A.J.M.; Jorissen, A.J.M.; Hoenderkamp, J.C.D.

    2012-01-01

    In recent years timber infill panels have been proposed for multi-story column-beam frame structures with the aim to substitute the stabilizing function of column-beam moment connections. The preliminary study reported in this paper considers a column-beam timber frame where stability is assured by

  13. The cracking of pressure tubes in the Pickering reactor

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.

    1978-01-01

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

  14. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

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

    1980-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

  15. Erythorbic acid promoted formation of CdS QDs in a tube-in-tube micro-channel reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Yan; Tan, Jiawei; Wang, Jiexin; Chen, Jianfeng [State Key Laboratory of Organic–Inorganic Composites, Beijing University of Chemical Technology, Beijing 100029 (China); Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, Beijing 100029 (China); Sun, Baochang, E-mail: sunbc@mail.buct.edu.cn [State Key Laboratory of Organic–Inorganic Composites, Beijing University of Chemical Technology, Beijing 100029 (China); Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, Beijing 100029 (China); Shao, Lei, E-mail: shaol@mail.buct.edu.cn [State Key Laboratory of Organic–Inorganic Composites, Beijing University of Chemical Technology, Beijing 100029 (China); Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, Beijing 100029 (China)

    2014-12-15

    Erythorbic acid assistant synthesis of CdS quantum dots (QDs) was conducted by homogeneous mixing of two continuous liquids in a high-throughput microporous tube-in-tube micro-channel reactor (MTMCR) at room temperature. The effects of the micropore size of the MTMCR, liquid flow rate, mixing time and reactant concentration on the size and size distribution of CdS QDs were investigated. It was found that the size and size distribution of CdS QDs could be tuned in the MTMCR. A combination of erythorbic acid promoted formation technique with the MTMCR may be a promising pathway for controllable mass production of QDs.

  16. In-situ inspection of grooves in reactor tube sheet using a remotely operated cast impression taking device

    International Nuclear Information System (INIS)

    Rajendran, S.; Ramakumar, M.S.

    1996-01-01

    Utmost importance is given to the in-service inspection of critical components of a reactor to ensure its reliable performance during the reactor operation. This paper describes a cast taking device using cold setting resin to take impression of the grooves being made in the tube sheet for sparger tube installation in pressurised heavy water reactor. (author)

  17. Flooding in a loop with a vertical and a horizontal tube connected by an elbow

    International Nuclear Information System (INIS)

    Yan Changqi

    1994-01-01

    The experimental research of flooding and flow-reverse in a test loop which a vertical and a horizontal tube connected by an elbow is introduced. According to the experimental results, the effects of the elbow on flooding and flow-reverse is analyzed. The experimental results is compared with the results obtained in vertical tubes. The effect of horizontal tube length and hysteresis in de-flooding are analyzed. Dimensionless parameters was used in data process. The correlations for predicting the flooding point, de-flooding point, completed carry up and flow reverse points are given

  18. Steam generator tube performance: world experience with water-cooled nuclear power reactors during 1979

    International Nuclear Information System (INIS)

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

    1981-01-01

    The performance of steam generator tubes in water-cooled nuclear power reactors is reviewed for 1979. Tube failures occurred at 38 of the 93 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The defect rate, although higher than that in 1978, was still lower than the rates of the two previous years. Methods being employed to detect defects include the increased use of multifrequency eddy-current testing and a trend to full-length inspection of all tubes. To reduce the incidence of tube failure by corrosion, plant operators are turning to full-flow condensate demineralization and more leak-resistant condenser tubes. 10 tables

  19. Method of fabricating a poision tube for reactor control rods

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Yasuhiko; Yoshida, Toshimi; Masaoka, Isao; Naruse, Akisuke

    1983-04-28

    A method to unify the neutron absorbing performance, enhance the workability in the insertion of neutron absorber tube and further decrease the stresses acting on the neutron absorber coating tube is described. The neutron absorber coated rod comprising neutron absorbing substance and a metal pipe is fabricated by compressing a metal pipe filled with the neutron absorber. Specifically, neutron absorbing substance such as boron carbide powder or the like is filled in a metal pipe such as made of stainless steel tube by way of vibration packing or the like. Then, after heating the metal pipe, it is applied with compression working such as swaging into a fine tube to increase the packing density of the absorbing substance filled in the pipe to greater than 60% of the theoretical density and completely contacted closely to the inner wall of the pipe. The neutron absorber coated rod thus fabricated can be inserted to an external coating tube with ease at a predetermined gap.

  20. Evolution of processing of GE fuel clad tubing for corrosion resistance in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Williams, C.D. [GE Nuclear Energy, Wilmington, NC (United States); Adamson, R.B. [GE Nuclear Energy, Wilmington, NC (United States); Marlowe, M.O. [GE Nuclear Energy, Wilmington, NC (United States); Plaza-Meyer, E. [GE Nuclear Energy, Wilmington, NC (United States); Proebstle, R.A. [GE Nuclear Energy, Wilmington, NC (United States); White, D.W. [GE Nuclear Energy, Wilmington, NC (United States)

    1996-05-01

    The current modification of the primary GE in-process solution-quench heat treatment, an (alpha+beta) solution-quench carried out at a tube diameter requiring only two subsequent reduction and anneal cycles, is applicable to Zr barrier fuel clad tubing, to non-barrier fuel clad tubing, and to the TRICLAD tubing product. A combination of good in-reactor corrosion performance and degradation resistance is anticipated for these products, based on knowledge of metallurgical characteristics and supported by the demonstrated performance capability of the Zircaloy-2 materials used. (orig.)

  1. Steam-generator tube failures: world experience in water-cooled nuclear power reactors during 1972

    International Nuclear Information System (INIS)

    Stevens-Guille, P.D.

    1975-01-01

    During 1972, approximately one in three operating reactors with steam generators incurred tube failures, predominantly near the tube sheet and in the bend region. Various forms of corrosion were the most frequent cause of failure. Eddy-current inspection was the preferred method for locating and investigating the cause of failure. Extensive use was made of both mechanical and explosive plugs for repair. As a class, steam generators with Monel 400 tubes had the lowest failure rates, and those with Inconel 600 tubes had the highest. (U.S.)

  2. Analysis of defect tubes of fast reactor heat exchanger

    International Nuclear Information System (INIS)

    Rukhlyada, N.Ya.

    2014-01-01

    The experimental Auger electron spectroscopy and X-ray diffraction microanalysis data of laboratory investigations of defect tubes of heat exchanger with sodium coolant are presented. Element distribution through depth of corrosion layers form on the side of coolant (sodium) and on the surface contacting with steam in heat exchanger tube is studied. Sodium presence through all thickness of the tube is determined. It is shown that treatment of 12Cr18N9 steel surface by plasma pulses decreases intergranular corrosion susceptibility. It is related with structural changes of surface layer (∼ 20 μm), its enrichment by chromium and formation of chromium oxide protective film [ru

  3. Method of cooling a pressure tube type reactor

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro.

    1983-01-01

    Purpose: To improve the operation efficiency of a nuclear reactor by carrying out cooling depending on the power distribution in the reactor core. Constitution: Reactor core channels are divided into a plurality of channel groups depending on the reactor power, and a water drum and a pump are disposed to each of the channel groups so as to increase the amount of coolants in response to the magnitude of the power from each of the channel groups. In this way, the minimum limiting power ratio can be increased. (Seki, T.)

  4. Visual inspection technology of the narrow and small confined area for monitoring feederpipe support of pressure tube in calandria reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Wan; Lee, Nam Ho; Choi, Young Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    There are 760 feederpipes, which they are connected to inlet/outlet of the 380 pressure tube channels on the front of the calandria, in CANDU-type Reactor of Wolsung Nuclear Power Plant. As an ISI(In-Service Inspection) and PSI (Post-Service Inspection) requirements, maintenance activities of measuring the thickness of curvilinear part of feederpipe and inspecting the feederpipe support area within calandria are needed to ensure continued reliable operation of nuclear power plant. And ultrasonic probe is used to measure the thickness of curvilinear part of feederpipe, however workers are exposed to radioactivity irradiation during the measurement period. But, it is exposed to radioactivity irradiation during the measurement period. But, it is impossible to inspect feederpipe support area thoroughly because of narrow and confined accessibility, that is , an inspection space between the pressure tube channels is less than 100 mm and pipes in feederpipe support area are congested. And also, workers involved in inspecting feederpipe support area are under the jeopardy of high-level radiation exposure. Concerns about sliding home, which make the move of feederpipe connected to pressure tube channel smooth as pressure tube expands and contracts in its axial direction, stuck to feedeerpipe support and some of the structural components have made necessary the development of video inspection probe system with narrow and confined accessibility to observe and inspect feederpipe support area more close. Using video inspection probe system, it is possible to inspect and repair abnormality of feederpipe support connected to pressure tube channels of the calandria more accurate and quantative than naked eye. Therefore, that will do much for ensuring safety of CANDU-type nuclear power plant. 45 figs.,31 tabs. (Author)

  5. On the calculation of hydrostatic pressures with SIMMER-II and the problem of connected tubes

    International Nuclear Information System (INIS)

    Schmuck, P.; Kleinheins, S.

    1988-07-01

    SIMMER simulations of flows in connected tubes exhibited non-physical movement of liquid under gravity. Analytical expressions are developed to evaluate the reasons for this effect and are used as a basis for proposals to keep the error small in SIMMER calculations. The effectiveness of the proposals is demonstrated by SIMMER code simulations of some numerical examples. (orig.) [de

  6. Fuel cladding tube and fuel rod for BWR type reactor

    International Nuclear Information System (INIS)

    Urata, Megumu; Mitani, Shinji.

    1995-01-01

    A fuel cladding tube has grooves fabricated, on the surface thereof, with a predetermined difference between crest and bottom (depth of the groove) in the circumferential direction. The cross sectional shape thereof is sinusoidal. The distribution of the grain size of iron crud particles in coolants is within a range about from 2μm to 12μm. If the surface roughness of the fuel cladding tube (depth of the groove) is determined greater than 1.6μm and less than 12.5, iron cruds in coolants can be positively deposited on the surface of the fuel cladding tube. In addition, once deposited iron cruds can be prevented from peeling from the surface of the fuel cladding tube. With such procedures, iron cruds deposited and radioactivated on the fuel cladding tube can be prevented from peeling, to prevent and reduce the increase of radiation dose on the surface of the pipelines without providing any additional device. (I.N.)

  7. Tube sheet structural analysis of intermediate heat exchanger for fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    Nakagawa, Y.; Ueno, T.; Fukuda, Y.; Ichimiya, M.

    1983-01-01

    The Prototype Fast Breeder Reactor 'Monju' is the first power generating fast breeder reactor in Japan. We have been designing the components of the plant for manufacturing. Among these is the intermediate heat exchanger (IHX) which exchanges heat between primary and secondary sodium loop. The tube sheet of IHX (shell to ligament junction) is a difficult area from the view point of structural strength design under elevated temperature. To validate the structural integrity of tube sheet we performed the series of inelastic analysis and tube sheet thermal shock test using test pieces and half scale model of actual design. The results of inelastic analyses showed there is little progressive deformation around shell to ligament structural discontinuous junction. Furthermore, thermal shock tests showed no increase of an accumulative deformation. By these analyses and experiments, structural reliability of tube sheet could be shown. (author)

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

    International Nuclear Information System (INIS)

    Foulquier, H.; Clara, P.

    1964-01-01

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

  9. Cladding tube of fuel rod for a BWR type reactor

    International Nuclear Information System (INIS)

    Nakayama, Hitoshi; Fujie, Kunio; Kuwahara, Heikichi; Hirai, Tadamasa; Kakizaki, Kimio.

    1976-01-01

    Object: To form a cladding tube wall with tunnels in communication with the exterior through a number of small-diameter openings to rapidly disperse a large quantity of heat thereby providing high density of the fuel rod. Structure: Tunnels adjacent to each other are provided under the skin in contact with cooling liquid of a cladding tube, and a number of openings through which said tunnels and the periphery of the cladding tube are placed in communication are formed, said openings each having its section smaller than that of said tunnel. With this arrangement, the cooling water entered the tunnel through some of small diameter openings absorbs heat of the fuel rod to be vaporized, which is flown out into the cooling water through the other small diameter openings and formed into vapor bubbles which move up for release of heat. (Taniai, N.)

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

    International Nuclear Information System (INIS)

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

    1978-02-01

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

  11. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  12. Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

    International Nuclear Information System (INIS)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement

  13. Studies on the instrumentation of a beam-tube medium flux reactor

    International Nuclear Information System (INIS)

    Axmann, A.; Pollet, J.L.; Queudot, J.

    1979-01-01

    In the years 1977/78, the ad hoc commitee for medium-flux reactor development of the Federal Ministry for Research and Technology developed constructional concepts for a medium-flux reactor to be utilized by beam tube experiments. The HMI has elaborated contributions for discussions of the subject of instrumentation, in particular for experiments in solid state physics. These contributions are contained in the report. (orig./RW) [de

  14. Calculation of fuel element temperature TRIGA 2000 reactor in sipping test tubes using CFD

    International Nuclear Information System (INIS)

    Sudjatmi KA

    2013-01-01

    It has been calculated the fuel element temperature in the sipping test of Bandung TRIGA 2000 reactor. The calculation needs to be done to ascertain that the fuel element temperatures are below or at the limit of the allowable temperature fuel elements during reactor operation. ensuring that the implementation of the test by using this device, the temperature is still within safety limits. The calculation is done by making a model sipping test tubes containing a fuel element surrounded by 9 fuel elements. according to the position sipping test tubes in the reactor core. by using Gambit. Dimensional model adapted to the dimensions of the tube and the fuel element in the reactor core of Bandung TRIGA 2000 reactor. Sipping test Operation for each fuel element performed for 30 minutes at 300 kW power. Calculations were performed using CFD software and as input adjusted parameters of TRIGA 2000 reactor. Simulations carried out on the operation of the 30, 60, 90, 120, 150, 180 and 210 minutes. The calculation result shows that the temperature of the fuel in tubes sipping test of 236.06 °C, while the temperature of the wall is 87.58 °C. The maximum temperature in the fuel center of TRIGA 2000 reactor in normal operation is 650 °C. and the boiling is not allowed in the reactor. So it can be concluded that the operation of the sipping test device are is very safe because the fuel center temperature is below the temperature limits the allowable fuel under normal operating conditions as well as the fuel element wall temperature is below the boiling temperature of water. (author)

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

    International Nuclear Information System (INIS)

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

    1978-07-01

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

  16. The shock tube as wave reactor for kinetic studies and material systems

    Energy Technology Data Exchange (ETDEWEB)

    Bhaskaran, K.A. [Indian Institute of Technology, Chennai (India). Department of Mechanical Engineering; Roth, P. [Gerhard Mercator Universitat, Duisberg (Germany). Institut fur Verbrennung und Gasdynamik

    2002-07-01

    Several important reviews of shock tube kinetics have appeared earlier, prominent among them being 'Shock Tube Technique in Chemical Kinetics' by Belford and Strehlow (Ann Rev Phys Chem 20 (1969) 247), 'Chemical Reaction of Shock Waves' by Wagner (Proceedings of the Eighth International Shock Tube Symposium (1971) 4/1), 'Shock Tube and Shock Wave Research' by Bauer and Lewis (Proceedings of the 11th International Symposium on Shock Tubes and Waves (1977) 269), 'Shock Waves in Chemistry' edited by Assa Lifshitz (Shock Waves in Chemistry, 1981) and 'Shock Tube Techniques in Chemical Kinetics' by Wing Tsang and Assa Lifshitz (Annu Rev Phys Chem 41 (1990) 559). A critical analysis of the different shock tube techniques, their limitations and suggestions to improve the accuracy of the data produced are contained in these reviews. The purpose of this article is to present the current status of kinetic research with emphasis on the diagnostic techniques. Selected studies on homogeneous and dispersed systems are presented to bring out the versatility of the shock tube technique. The use of the shock tube as high temperature wave reactor for gas phase material synthesis is also highlighted. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-02-15

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

  18. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1987-01-01

    This patent describes a pressurized water reactor steam generator having spaced rows of heat transfer tubes through which primary coolant from the reactor flows, the tubes being of a U-shaped design, with the U-bend portions of the U-shaped tubes stabilized by antivibration bars. The improvement described here comprises expandable antivibration bars for stabilizing the U-bend portions of the U-shaped tubes, the expandable bars having a pair of adjustable rods, formed from a pair of rod sections affixed to a connector, one rod section of each of the pair of rod sections having a plurality of protrusions. Each of the protrusions has slidable surfaces thereon. The other rod section of each of the pair of rod sections has indentations, each of the indentations having slidable surfaces thereon complementary to the sliding surfaces of the protrusions, such that the rods are expandable from a first cross-sectional width less than the spacing between two adjacent rows of the tubes, to a second cross-sectional width greater than the first cross-sectional width. The expanded rods are adapted to contact tubes of the two adjacent rows of the tubes

  19. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

  20. Utilization of Stop-flow Micro-tubing Reactors for the Development of Organic Transformations.

    Science.gov (United States)

    Toh, Ren Wei; Li, Jie Sheng; Wu, Jie

    2018-01-04

    A new reaction screening technology for organic synthesis was recently demonstrated by combining elements from both continuous micro-flow and conventional batch reactors, coined stop-flow micro-tubing (SFMT) reactors. In SFMT, chemical reactions that require high pressure can be screened in parallel through a safer and convenient way. Cross-contamination, which is a common problem in reaction screening for continuous flow reactors, is avoided in SFMT. Moreover, the commercially available light-permeable micro-tubing can be incorporated into SFMT, serving as an excellent choice for light-mediated reactions due to a more effective uniform light exposure, compared to batch reactors. Overall, the SFMT reactor system is similar to continuous flow reactors and more superior than batch reactors for reactions that incorporate gas reagents and/or require light-illumination, which enables a simple but highly efficient reaction screening system. Furthermore, any successfully developed reaction in the SFMT reactor system can be conveniently translated to continuous-flow synthesis for large scale production.

  1. Design and Computational Fluid Dynamics Optimization of the Tube End Effector for Reactor Pressure Vessel Head Type VVER-1000

    International Nuclear Information System (INIS)

    Novosel, D.

    2006-01-01

    In this paper is presented development and optimization of the tube end effector design which should consist of 4 ultrasonic transducers, 4 Eddy Current's transducers and Radiation Proof Dot Camera. Basically, designing was conducted by main input requests, such as: inner diameter of a tested reactor pressure vessel head penetration tube, dimensions of a transducers and maximum allowable vertical movement of a manipulator connection rod in order to cover all inner tube surface. As is obvious, for ultrasonic testing should be provided the thin layer of liquid material (in our case water was chosen) which is necessary to make physical contact between transducer surface and investigated inner tube surface. By help of Computational Fluid Dynamics, determined were parameters of geometry, as the most important factor of transducer housing, hydraulically parameters for water supply and primary drain together implemented into this housing, movement of the end effectors (vertical and cylindrical) and finally, necessary equipment which has to provide all hydraulically and pneumatic requirements. As the cylindrical surface of the inner tube diameter was liquefied and contact between transducer housing and tested tube wasn't ideally covered, water leakage could occur in downstream direction. To reduce water leakage, which is highly contaminated, developed was second water drain by diffuser assembly which is driven by Venturi pipe, commercially called vacuum generator. Using the Computational Fluid Dynamic, obtained was optimized geometry of diffuser control volume with the highest efficiency, in other words, unobstructed fluid flux. Afterwards, the end effectors system was synchronized to the existing operable system for NDT methods all invented and designed by INETEC. (author)

  2. CFD in supercritical water-cooled nuclear reactor (SCWR) with horizontal tube bundles

    Energy Technology Data Exchange (ETDEWEB)

    Zhi Shang, E-mail: zhi.shang@stfc.ac.uk [Science and Technology Facilities Council, Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Lo, Simon, E-mail: simon.lo@uk.cd-adapco.com [CD-adapco, Trident House, Basil Hill Road, Didcot OX11 7HJ (United Kingdom)

    2011-11-15

    The commercial CFD code STAR-CD 4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round tubes and tube bundles. Reactors with vertical or horizontal flow in the core can be found. In a vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in a horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal tubes and tube bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. From the study of single round tubes, the Speziale quadratic non-linear high-Re k-{epsilon} turbulence model with the two-layer model for near wall treatment is found to produce the best results in comparison with experimental data. In tube bundle simulations, it is found that the temperature is higher in the top half of the bundle and the highest tube wall temperature is located at the outside tubes where the flow rate is the lowest. The secondary flows across the bundle are highly complex. Their main effect is to even out the temperature over the area within each individual recirculation region. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.

  3. CFD in supercritical water-cooled nuclear reactor (SCWR) with horizontal tube bundles

    International Nuclear Information System (INIS)

    Shang, Zhi; Lo, Simon

    2009-01-01

    The commercial CFD code STAR-CD 4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round tubes and tube bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal tubes and tube bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. From the study of single round tubes, the Speziale quadratic non-linear high-Re k-ε turbulence model with the two-layer model for near wall treatment is found to produce the best results in comparison with experimental data. In tube bundle simulations, it is found that the temperature is higher in the top half of the bundle and the highest tube wall temperature is located at the outside tubes where the flow rate is the lowest. The secondary flows across the bundle are highly complex. Their main effect is to even out the temperature over the area within each individual recirculating region. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR. (author)

  4. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  5. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L.

    1997-01-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis

  6. Development and fabrication of seamless Aluminium finned clad tubes for metallic uranium fuel rods for research reactor

    International Nuclear Information System (INIS)

    Singh, A.K.; Hussain, M.M.; Jayachandran, N.K.; Abdulla, K.K.

    2012-01-01

    Natural uranium metal or its alloy is used as fuel in nuclear reactors. Usually fuel is clad with compatible material to prevent its direct contact with coolant which prevents spread of activity. One of the methods of producing fuel for nuclear reactor is by co-drawing finished uranium rods with aluminum clad tube to develop intimate contact for effective heat removal during reactor operation. Presently seam welded Aluminium tubes are used as clad for Research Reactor fuel. The paper will highlight entire fabrication process followed for the fabrication of seamless Aluminium finned tubes along with relevant characterisation results

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  8. Steam-generator tube failures: world experience in water-cooled nuclear power reactors in 1974

    International Nuclear Information System (INIS)

    Hare, M.G.

    1976-01-01

    Steam-generator tube failures were reported at 25 of 59 water-cooled nuclear power reactors surveyed in 1974, compared to 11 of 49 in 1973. A summary is presented of these failures, most of which, where the cause is known, were the result of corrosion. Water chemistry control, inspection and repair procedures, and failure rates are discussed

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  11. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock

    2013-01-01

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube

  12. The management of severe accidents in modern pressure tube reactors

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Blahnik, C.; Snell, V.G.; Duffey, R.B.

    2007-01-01

    Advanced new reactor designs resist severe accidents through a balance between prevention and mitigation. This balance is achieved by designing to ensure that such accidents are very rare; and by limiting core damage progression and releases from the plant in the event of such rare accidents. These design objectives are supported by a suitable combination of probabilistic safety analysis, engineering judgment and experimental and analytical study. This paper describes the approach used for the Advanced CANDU Reactor TM -1000 (ACR-1000) design, which includes provisions to both prevent and mitigate severe accidents. The paper describes the use of PSA as a 'design assist' tool; the analysis of core damage progression pathways; the definition of the core damage states; the capability of the mitigating systems to stop and control severe accident events; and the severe accident management opportunities for consequence reduction. (author)

  13. High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

    International Nuclear Information System (INIS)

    Chrysochoides, N.G.; Cundy, M.R.; Von der Hardt, P.; Husmann, K.; Swanenburg de Veye, R.J.; Tas, A.

    1985-01-01

    The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

  14. New north beam tube for the neutron radiography reactor

    International Nuclear Information System (INIS)

    Pruett, D.P.; Richards, W.J.; Heidel, C.C.

    1982-01-01

    Neutron radiography of the fuel undergoing examination in the argon cell is performed in the NRAD Facility and is one of many examinations performed on the fuel. The reactor and examination procedure are described. The new radiography system, developed to expand the present radiography capabilities to radiograph both irradiated and unirradiated specimens and to provide for the development of new radiography techniques without interfering with the argon cell production schedule is presented

  15. Simulation of Reforming Reactor Tube: Quantifying Catalyst Pellet's Effectiveness Factor

    OpenAIRE

    Da Cruz, Flavio Eduardo

    2016-01-01

    In this work, a consistent mathematical model to simulate a spherical catalytic pellet and a Packed-Bed Reactor (PBR) is develop. The Dusty Gas Model (DGM) is applied to the calculation of the diffusive fluxes in the porous media. Simulations are executed considering hydrogen production from steam methane reforming. Species’ diffusivities are calculated using data from literature as well as the values for tortuosity and porosity. The pellet simulation is performed considering mass, species, m...

  16. Carbon nanotubes: from nano test tube to nano-reactor.

    Science.gov (United States)

    Khlobystov, Andrei N

    2011-12-27

    Confinement of molecules and atoms inside carbon nanotubes provides a powerful strategy for studying structures and chemical properties of individual molecules at the nanoscale. In this issue of ACS Nano, Allen et al. explore the nanotube as a template leading to the formation of unusual supramolecular and covalent structures. The potential of carbon nanotubes as reactors for synthesis on the nano- and macroscales is discussed in light of recent studies.

  17. Testing of a 7-tube palladium membrane reactor for potential use in TEP

    International Nuclear Information System (INIS)

    Carlson, Bryan J.; Trujillo, Stephen; Willms, R. Scott

    2010-01-01

    A Palladium Membrane Reactor (PMR) consists of a palladium/silver membrane permeator filled with catalyst (catalyst may be inside or outside the membrane tubes). The PMR is designed to recover tritium from the methane, water, and other impurities present in fusion reactor effluent. A key feature of a PMR is that the total hydrogen isotope content of a stream is significantly reduced as (1) methane-steam reforming and/or water-gas shift reactions proceed on the catalyst bed and (2) hydrogen isotopes are removed via permeation through the membrane. With a PMR design matched to processing requirements, nearly complete hydrogen isotope removals can be achieved. A 3-tube PMR study was recently completed. From the results presented in this study, it was possible to conclude that a PMR is appropriate for TEP, perforated metal tube protectors function well, platinum on aluminum (PtA) catalyst performs the best, conditioning with air is probably required to properly condition the Pd/Ag tubes, and that CO/CO 2 ratios maybe an indicator of coking. The 3-tube PMR had a permeator membrane area of 0.0247 m 2 and a catalyst volume to membrane area ratio of 4.63 cc/cm 2 (with the catalyst on the outside of the membrane tubes and the catalyst only covering the membrane tube length). A PMR for TEP will require a larger membrane area (perhaps 0.35 m 2 ). With this in mind, an intermediate sized PMR was constructed. This PMR has 7 permeator tubes and a total membrane area of 0.0851 m 2 . The catalyst volume to membrane area ratio for the 7-tube PMR was 5.18 cc/cm 2 . The total membrane area of the 7-tube PMR (0.0851 m 2 ) is 3.45 times larger than total membrane area of the 3-tube PMR (0.0247 m 2 ). The following objectives were identified for the 7-tube PMR tests: (1) Refine test measurements, especially humidity and flow; (2) Refine maintenance procedures for Pd/Ag tube conditioning; (3) Evaluate baseline PMR operating conditions; (4) Determine PMR scaling method; (5) Evaluate PMR

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

    International Nuclear Information System (INIS)

    Yang Fanglin; Yang Yingli; Wang Guangshen

    1989-10-01

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

  19. Nuclear reactor incorporating locking device for threaded bolt connections

    International Nuclear Information System (INIS)

    Blaushild, R.M.

    1987-01-01

    A nuclear reactor having a pressure vessel and a first element is described comprising a core barrel situated within the pressure vessel. The core barrel has a baffle former secured in and to the core barrel by bolted connections, and a second element comprising baffle plates secured to the inner surface of the baffle former by bolted connections, with a locking device to prevent loosening of bolted connections between the baffle former and at least one of the elements. The baffle former and at least one element are held together by a headed, threaded bolt engaged in a bore coaxially extending in the baffle former and at least one element and threadedly engaged in a threaded section in at least the baffle former. The threaded section has first threaded of a first direction, with the head of the bolt engaged with a shoulder about the bore in at least one element to hold the baffle formed and at least one element together, the head of the bolt having a first diameter and a cavity, having an unsymmetrical wall thereabout, in the end surface thereof. It comprises a recess in at least one element coaxial with the bore forming a wall thereabout and extending inwardly from the outer surface of at least one element, the recess having a second diameter greater than the first diameter, with at least one element having second threads in the wall of a direction opposite the direction of the first threads of the threaded bore; a locking nut having a base with a downwardly depending cylindrical wall thereabout

  20. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2011-07-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  1. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

    2011-01-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  2. Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR [High Flux Isotope Reactor] Reactor

    International Nuclear Information System (INIS)

    Childs, R.L.; Rhoades, W.A.; Williams, L.R.

    1988-01-01

    The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs

  3. Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

    International Nuclear Information System (INIS)

    Diamond, W.T.

    2010-01-01

    One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

  4. Thermal hydraulic stability in a pressure tube nuclear reactor

    International Nuclear Information System (INIS)

    Villani, A.; Ravetta, R.; Mansani, L.

    1986-01-01

    The CIRENE plant which will undergo preoperational tests in the near future is equipped with a 40 MW(e) Heavy Water moderated Boiling Light Water cooled Reactor (HWBLWR); at the start-up and up to about 30 % of nominal power, the necessary low coolant density is obtained injecting into the core a mixture of liquid and steam. To verify the thermal-hydraulic stability of the plant in this situation, tests have been carried out in a facility simulating two full scale power channels; the system stability has been confirmed in the reference conditions, and is not reduced by even a significant reduction of the liquid flowrate, where a decrease in liquid temperature has some negative effect and steam flowrate has a small influence. (author)

  5. Heat exchanging tube behaviour in steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Pastor, D.; Oertel, K.

    1979-01-01

    Based on a comprehensive failure statistics, materials corrosion chemistry and thermohydraulics problems of the tubings of steam generators are considered. A historical review of failures in the tubings of steam generators in pressurized water reactors reflects the often successless measures by designers, manufacturers and operating organizations for preventing failures, especially with regard to materials selection and water regime. It is stated that laboratory tests could not give sufficient information about safe and stable operation of nuclear steam generators unless real constructive, hydrodynamic, thermodynamical and chemical conditions of operation had been taken into account. (author)

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

    Science.gov (United States)

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  7. SOLAR REFRIGERATING UNIT WITH AN ADSORPTION REACTOR AND EVACUATED TUBE COLLECTORS

    Directory of Open Access Journals (Sweden)

    M.E. Vieira

    1997-09-01

    Full Text Available This work presents the principles of operation of a solar refrigerator with the following basic components: a reactor, a set of evacuated tube solar collectors, a condenser, a heat exchanger, and an evaporator. During the heating phase, solar radiation is collected and transferred to the reactor for desorption by a vapor thermal siphon loop. During the cooling phase, heat from the reactor is released to the ambient by a second water vapor loop. Ambient data collected daily during a period of 18 years were divided into hourly values and used to simulate the temperatures of the reactor, which uses salt impregnated with graphite and ammonia, during the adsorption / desorption processes. The results show that the refrigerator operates well in Fortaleza and that better results are expected for the countryside of the state of Ceara. It is concluded that only a high efficiency collector set can be used in the system

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  9. Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

    International Nuclear Information System (INIS)

    Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

    1983-01-01

    A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

  10. On the anisotropy of in-reactor creep of Zr-2.5Nb tubes

    International Nuclear Information System (INIS)

    Causey, A.R.; Holt, R.A.

    1993-06-01

    Creep specimens made from cold-worked Zr-2.5Nb tubes, fabricated with two different microstructures and crystallographic textures, were irradiated in the Osiris reactor in France in a fast-neutron flux of about 1.8 x 10 18 n.m -2 .s -1 , E > MeV, at 553 and 585 K. The hoop stresses from internal Fluences, up to 4 x 10 25 n.m -2 , more than double those achieved an any other creep test on cold-worked Zr-2.5Nb in which both axial and transverse strain were measured. Creep rates were obtained from strain versus fluence plots, and creep compliances were obtained from plots of the strain rates against hoop stress for each material at each temperature. The ratio of creep rates at 583 K to those at 553 K was ∼ 1.36, a little higher than that extrapolated from stress relaxation results at temperatures between 523 and 568 K. The ratio of the biaxial creep compliance in the axial direction to that in the transverse directions is different for the two test materials: 0.0 to -0.1 for the fuel sheathing texture and 0.5 to 0.6 for the pressure tube texture. The results were analysed using a self-consistent model developed to account for the contributions to the creep anisotropy of the three microstructure parameters involved and to account for the grain interaction effects. The model, which was normalized to test reactor and power reactor creep data for cold-worked Zr-2.5Nb tubes, predicted the ratio of the creep compliancies to be -0.26 and 0.63, respectively. Thus the creep anisotropy of Zr-2.5Nb tubes with pressure-tube-like crystallographic texture can be adequately predicted. (author). 18 refs., 4 tabs., 13 figs

  11. Remote-controlled television for locating leaking tubes in pressurized-water reactor steam generators

    International Nuclear Information System (INIS)

    Cormault, P.; Denis, J.

    1978-01-01

    The Scarabee system is designed for observation of the tubes in water boxes of pressurized-water reactor nuclear-power-station steam generators. It consists essentially of a camera and a projector used as a marker, both of which swivel freely. The whole unit is housed in a water-tight container which can easily be decontaminated. Remote control of camera and marker movement is carried out from a console. (author)

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

    International Nuclear Information System (INIS)

    Kapoor, K.; Prahlad, B.

    2012-01-01

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

  13. Design and use of the ORNL HFIR [High Flux Isotope Reactor] pneumatic tube irradiation systems

    International Nuclear Information System (INIS)

    Dyer, F.F.; Emery, J.F.; Robinson, L.; Teasley, N.A.

    1987-01-01

    A second pneumatic tube that was recently installed in the High Flux Isotope Reactor for neutron activation analysis is described. Although not yet tested, the system is expected to have a thermal neutron flux of about 1.5 x 10 14 cm -2 s -1 . A delayed neutron counter is an integral part of the pneumatic tube, and all of the hardware is present to enable automated use of the counter. The system is operated with a Gould programmable controller that is programmed with an IBM personal computer. Automation of any mode of operation, including the delayed neutron counter, will only require a nominal amount of software development. Except for the lack of a hot cell, the irradiation facility has all of the advantageous features of an older pneumatic tube that has been in operation for 17 years. The design of the system and some applications and methods of operation are described

  14. The calculation of hydrostatic pressures in SIMMER-II and the problem of connected tubes

    International Nuclear Information System (INIS)

    Schmuck, P.; Kleinheins, S.

    1986-03-01

    Some SIMMER-II calculations had shown that the hydrostatic equilibrium in connected tubes had not been simulated correctly under certain conditions. The present report documents the results of the investigations, which had been performed to understand these inconsistencies. The reasons are presented in analytical form and possibilities to avoid the mistakes or to keep their consequences as small as possible are given. For the verification of the analytical results some SIMMER-II, version 9 have been performed and their results are presented

  15. An approach to estimate the reactivity worth of R-5 poison tube system and experimental verification in ZERLINA reactor

    International Nuclear Information System (INIS)

    Khosla, S.K.; Paul, O.P.K.; Sengupta, S.N.

    1976-01-01

    It is proposed to employ a liquid poison injection system as an emergency shut down device for R-5 reactor. The liquid poison consists of gadolinium nitrate solution, which is injected into twenty poison tubes made of zircaloy that are located in between the regular lattice positions in R-5 reactor. The calculational model adopted to estimate the reactivity worth of the poison tubes so as to hold the reactor subcritical by 50 mk at full tank, is described. Similar reactivity estimates have also been carried out for R-5 poison tubes installed in Zerlina reactor in order to assess the adequacy of the calculational mode. The results of the calculations are compared with experimental values for single poison tubes. (author)

  16. Non-destructive evaluation of stream generator tubes and pressure tubes from the PHWR reactors, using the rotating magnetic field method

    International Nuclear Information System (INIS)

    Premel, D.; Placko, D.; Grimberg, R.; Savin, A.

    2001-01-01

    This work presents a new type of eddy current transducer with a rotating magnetic field devoted to the inspection of steam generator tubes and pressure tubes from the PHWR reactors. A theoretical model has been developed that permits the calculations of the emf induced in the reception coils in the presence of the copper or magnetite deposits, anti-vibration railing and garter springs. (authors)

  17. Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

    1997-07-01

    Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

  18. Some engineering aspects of the investigation into the cracking of pressure tubes in the Pickering reactors

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Towgood, G.R.; Hunter, T.A.

    1976-01-01

    In August 1974, Pickering Unit 3 (514 MWe) was shutdown for a period of 8 months because of cracks in 17 of the 390 pressure tubes. The cracks were a result of incorrect installation procedures during construction. Improper positioning of the rolling tool used to join the Zr-2.5 wt% Nb pressure tube to the end fitting produced very high residual tensile stresses. High stresses in combination with periods with the tubes cold caused the cracking. Crack propagation was by fracture of hydrides which are brittle when cold. Subsequent investigation confirmed that properly rolled joints are not susceptible to such cracking. The resources of Canadian industry, Ontario Hydro and Atomic Energy of Canada were coordinated to find engineering solutions to the crack program. The defective tubes were removed from reactor, thoroughly examined to identify the cause of the cracks, and thoroughly tested to prove safety. Non-destructive techniques were quickly adopted for inspection of tubes in Pickering. Tools and procedures for retubing the 17 channels were prepared and Pickering Unit 3 was returned to service at the end of March 1975. (author)

  19. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1983-08-01

    Heating tests using 53 wt % U 3 O 8 -Al pellets show that an exothermic reaction occurs between 875 and 1000 0 C and takes 10 to 20 seconds to reach maximum temperature. The maximum temperature is a function of particle size of the U 3 O 8 with large particles exhibiting lower peak temperatures. The calculated energy release was 123 cal/g of U 3 O 8 -aluminum fuel. Tests using aluminum clad outer fuel tube sections gave lower peak temperatures than for pellets. No violent reactions occurred. The results are reasonably consistent with recent reported data indicating that the exothermic U 3 O 8 -Al reaction is not an important energy source. The compressive and tensile strengths of U 3 O 8 tubes above 660 0 C are low. In compression, sections with 2 psi average axial stress failed at 917 0 C, while sections with 7 psi failed at 669 0 C. Tubes with U-Al alloy cores failed at about 670 0 C with no applied load. The stresses in fuel tubes during a reactor transient may range up to several hundred psi and are less than 7 psi only in the upper part of the fuel tube

  20. Do existing research reactors teach us all about beam tube optimisation?

    International Nuclear Information System (INIS)

    Roegler, Hans-Joachim; Feltes, Wolfgang

    1998-01-01

    The contribution makes the attempt to analyse the data base available in the literature and in Siemens' own projects and to find out potential systematics from the existing research reactors with beam tubes, separated into reactors with different reflectors and distinguished for tangential and radial tubes and cold neutron sources, respectively some generic calculations serve as gauging data. The contribution is not meant as critics on any design. The results might serve supporting designers and operators when evaluating the pros and cons of existing or planned design in terms of the optimum beam tubes. Existing lacks of systematics are evaluated in view of suitable explanations and constraints, which do not allow optimisation. Examples of such constraints are the different material layers between fuel zone and reflector zone which have various reasons. The limited data in the literature plus the numerous lacks of precision of the representation of those data should be an incentive to improve the performed analysis by collecting more exact data and re-doing the evaluation before answering the title question really

  1. Do existing research reactors teach us all about beam tube optimization?

    International Nuclear Information System (INIS)

    Roegler, Hans Joachim; Feltes, Wolfgang

    1998-01-01

    The contribution makes the attempt to analyse the data base available in the literature and in Siemens' own projects and to find out potential systematics from the existing research reactor with beam tubes, separated into reactors with different reflectors and distinguished for tangential and radial tubes and cold neutron sources, resp. Some generic calculations serve as gauging data. The contribution is not meant as critics on any design.The results might serve supporting designers and operators when evaluating the pros and cons of existing or planned design in terms of the optimum beam tubes. Existing lacks of systematics are evaluated in view of suitable explanations and constraints, which do not allow optimisation. Examples pf such constraints are the different material layers between fuel zone and reflector zone which have various reasons. The limited data in the literature plus the numerous lacks of precision of the representation of those data should be an incentive to improve the performed analysis by collecting more exact data and re-doing the evaluation before answering the title-question really. (author)

  2. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    Science.gov (United States)

    Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

    2008-12-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

  3. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

    2008-01-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

  4. Statistical analysis and modelling of in-reactor diametral creep of Zr-2.5Nb pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jyrkama, Mikko I., E-mail: mjyrkama@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada); Bickel, Grant A., E-mail: grant.bickel@cnl.ca [Canadian Nuclear Laboratories, Chalk River Laboratories, Chalk River, ON, Canada K0J 1J0 (Canada); Pandey, Mahesh D., E-mail: mdpandey@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada)

    2016-04-15

    Highlights: • New and simple statistical model of pressure tube diametral creep. • Based on surveillance data of 328 pressure tubes from eight different CANDU reactors. • Uses weighted least squares (WLS) to regress out operating conditions. • The shape of the diametral creep profiles are predicted very well. • Provides insight and relative ranking of strain behaviour of in-service tubes. - Abstract: This paper presents the development of a simplified regression approach for modelling the diametral creep over time in Zr-2.5 wt% Nb pressure tubes used in CANDU reactors. The model is based on a large dataset of in-service inspection data of 328 different pressure tubes from eight different CANDU reactor units. The proposed weighted least squares (WLS) regression model is linear in time as a function of flux and temperature, with a temperature-dependent variance function. The model predicts the shape of the observed diametral creep profiles very well, and is useful not merely for prediction, but also for assessing tube-to-tube variability and manufacturing properties among the inspected tubes.

  5. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P. C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  6. Evaluation of maintenance strategies for steam generator tubes in pressurized waster reactors. 2. Cost and profitability analyses

    International Nuclear Information System (INIS)

    Isobe, Y.; Sagisaka, M.; Yoshimura, S.; Yagawa, G.

    2000-01-01

    As an application of probabilistic fracture mechanics (PFM), risk-benefit analysis was carried out to evaluate maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). The analysis was conducted for SG tubes made of Inconel 600, and Inconel 690 as well assuming its crack initiation and crack propagation law based on Inconel 600 data. The following results were drawn from the analysis. Improvement of inspection accuracy reduces the maintenance costs significantly and is preferable from the viewpoint of profitability due to reduction of SG tube leakage and rupture. There is a certain region of SCC properties of SG tubes where sampling inspection is effective. (author)

  7. Analysing The Thermalhydraulic Parameters Of VVER-1000 Reactor For The Accident Of Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Luu Nam Hai; Truong Cong Thang

    2011-01-01

    To ensure the safety operation of nuclear power plant (NPP), a lot of postulated accident scenarios were considered and analysed. This research chose and analysed the accident of steam generator tube rupture (SGTR) under the actual plant conditions by using the simulation program PCTRAN. The SGTR accident is happen when the NPP is under operation with the steady state condition (power of 3000 MWth, primary pressure of 157 bar and secondary pressure of 63 bar). The accident is initiated by creating a break with equivalent diameter of 100 mm in the area of lower row heat exchanging tubes. The result of analysis is compared with the calculation of the Shariz University, Iran using the thermal hydraulics code RELAP5/mod3.2 and the report in the PSAR data of VVER-1000. This comparison shows that it is possible for using PCTRAN to analyse accidents of VVER-1000 reactor. (author)

  8. Preliminary evaluation of steam generator tube rupture (SGTR) accident in lead cooled reactor

    International Nuclear Information System (INIS)

    Frano, R. Lo; Forasassi, G.

    2009-01-01

    In this paper some contributions are provided to the development of a European Lead-cooled System, known as the ELSY project (within EU-6 Framework Project); that will constitute a possible reference system for a large lead-cooled reactor of GEN IV. Steam generator (SG) tubing of this system type might be subject to a variety of degradation processes, such as cracking, wall thinning and potential leakage or rupture, eventually leading to the failure of one or more SG tubes that constitute a steam generator tube rupture (SGTR) accident with possible consequences for the safety of the primary systems. It is therefore of interest for the designer to know how the SG itself, as well as the vessel and internals structures, behave under impulsive loading conditions (in form of a rapid and strong increase of pressure) that can arise as consequences of the interaction between the primary and secondary coolants (lead-water interaction). The analysed initiator event, as already mentioned, is a large break (up to a double ended guillotine break) of one (or more) SG cooling tubes that may become severe enough to determine dangerous effects on the interested structures. In order to better simulate and perform the mentioned postulated SGTR accident sequence analyses, an appropriate numerical model with the available computing resources (FEM codes) was set up at the DIMNP of Pisa University. That model was used to evaluate the effects of the propagation of the blast pressure waves inside the SG structures, taking into account also the sloshing phenomenon that could be induced by the lead primary coolant motions. Therefore the SGTR effects study may be considered as a transient and non linear problem the solution of which provides the 'time histories' of hydrodynamic pressures and stresses on the reactor pressure vessel and internals walls. (author)

  9. Diaphragm flange and method for lowering particle beam impedance at connected beam tubes of a particle accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Biallas, George Herman

    2017-07-04

    A diaphragm flange for connecting the tubes in a particle accelerator while minimizing beamline impedance. The diaphragm flange includes an outer flange and a thin diaphragm integral with the outer flange. Bolt holes in the outer flange provide a means for bolting the diaphragm flange to an adjacent flange or beam tube having a mating bolt-hole pattern. The diaphragm flange includes a first surface for connection to the tube of a particle accelerator beamline and a second surface for connection to a CF flange. The second surface includes a recessed surface therein and a knife-edge on the recessed surface. The diaphragm includes a thickness that enables flexing of the integral diaphragm during assembly of beamline components. The knife-edge enables compression of a soft metal gasket to provide a leak-tight seal.

  10. Advanced nondestructive examination of the reactor vessel head penetration tube welds

    International Nuclear Information System (INIS)

    Cvitanovic, M.; Zado, V.

    1996-01-01

    Beside a referent code examination requirements, appearance of the service induced flaws on the Reactor Vessel Head (RVH) penetration tube welds forced development of remotely operated examination tools and techniques. Several systems were developed for examination of RVH PWR type while only one system for examination of VVER - 440 type RVH has been developed by Inetec. In this article the most advanced RVH VVER - 440 type examination techniques such as ultrasonic, eddy current and visual testing techniques as well as remotely operated tool are described. (author)

  11. Factors in the selection of broiler tube materials for a civil fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tyzack, C; Chitty, A

    1975-07-01

    This paper briefly considers some of the factors which must be balanced in the selection of a boiler tube material for a Civil Fast Reactor. The merits and possible demerits of low alloy ferritic steels and the austenitic Alloy 800 are compared with respect to waterside corrosion resistance, mechanical properties, fabrication and weldability and possible effects of exposure to the sodium environment under normal and fault conditions. It is pointed out that although there is operational experience of most of the materials in boiler superheater applications there is little or none in evaporative regimes. (author)

  12. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  13. GERDA test facility for pressurized water reactors with straight tube steam generators

    International Nuclear Information System (INIS)

    Ahrens, G.; Haury, G.; Lahner, K.; Schatz, A.

    1983-01-01

    A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.) [de

  14. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    International Nuclear Information System (INIS)

    Bromley, B.P.; Edwards, G.W.R.; Sambavalingam, P.

    2016-01-01

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  15. Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

    1989-01-01

    Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm 2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

  16. A connection-disconnection device for the nuclear reactor control rods

    International Nuclear Information System (INIS)

    Drean, H.; Breant, C.

    1992-01-01

    An automatic and relatively light system is developed to connect and disconnect the control rods in a nuclear reactor; it is designed to apply, in a controlled manner, constraints on the rods and to detect a possible misalignment of the connection-disconnection means in the corresponding butt

  17. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  18. Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

    International Nuclear Information System (INIS)

    Jain, Vikas; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Highlights: → We have highlighted the problem of drum level dynamics in a multiple loop type NC system using RELAP5 code. → The need of interconnections in steam and liquid spaces close to drum is established. → The steam space interconnections equalize pressure and liquid space interconnections equalize level. → With this scheme, the system can withstand anomalous conditions. → However, the controller is found to be inevitable for inventory balance. - Abstract: Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam-water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam-water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

  19. Pressure tube reactors and a sustainable energy future: the ultra development path

    International Nuclear Information System (INIS)

    Duffey, R.

    2008-01-01

    Nuclear energy must be made available, freely and readily, to help meet world energy needs, concerns over energy price and security of supply, and alleviating the uncertainties over potential climate change. The perspective offered here is a model for others to consider, adopting and adapting using whatever elements fit their own strategies and needs. The underlying philosophy is to retain flexibility in the reactor development, deployment and fuel cycle, while ensuring the principle that customer, energy market, safety, non-proliferation and sustainability needs are all addressed. Canada is the world's largest exporter of uranium, providing about one-third of the world supply for nuclear power reactors. Pressure tube reactors (PTRs), of which CANDU is a prime example, have a major role to play in a sustainable energy future. The inherent fuel cycle flexibility of the PTR offers many technical, resource and sustainability, and economic advantages over other reactor technologies and is the subject of this paper. The design evolution and development intent is to be consistent with improved or enhanced safety, licensing and operating limits, global proliferation concerns, and waste stream reduction, thus enabling sustainable energy futures. The limits are simply those placed by safety, economics and resource availability. (author)

  20. The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

    International Nuclear Information System (INIS)

    Zheng Youqi; Zu Tiejun; Wu Hongchun; Cao Liangzhi; Yang Chao

    2012-01-01

    In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.

  1. Pressure tube reactors and a sustainable energy future: the ultra development path

    International Nuclear Information System (INIS)

    Duffey, R.

    2008-01-01

    Nuclear energy must be made available, freely and readily, to help meet world energy needs, concerns over energy price and security of supply, and alleviating the uncertainties over potential climate change. The perspective offered here is a model for others to consider, adopting and adapting using whatever elements fit their own strategies and needs. The underlying philosophy is to retain flexibility in the reactor development, deployment and fuel cycle, while ensuring the principle that customer, energy market, safety, non-proliferation and sustainability needs are all addressed. Canada is the world's largest exporter of uranium, providing about one-third of the world supply for nuclear power reactors. Pressure tube reactors (PTRs), of which CANDU, is a prime example, have a major role to play in a sustainable energy future. The inherent fuel cycle flexibility of the PTR offers many technical, resource and sustainability and economic advantages over other reactor technologies and is the subject of this paper. The design evolution and development intent is to be consistent with improved or enhanced safety, licensing and operating limits, global proliferation concerns, and waste stream reduction, thus enabling sustainable energy futures. The limits are simply those placed by safety, economics and resource availability. (author)

  2. Influence of connection tubing in modern size exclusion chromatography and its impact on the characterization of mAbs.

    Science.gov (United States)

    Fekete, Szabolcs; Guillarme, Davy

    2018-02-05

    The goal of the study was to evaluate the impact of connection tubing in modern size exclusion chromatography (SEC), since it may strongly impact the apparent column efficiency, as the compounds are not retained in SEC. For this purpose, a reference SEC column of 150×4.6mm, 1.8μm was considered, and various proteins were tested as model compounds. Different tube geometries (lengths and internal diameters) and materials (stainless steel and PEEK) were evaluated in a systematic way. Large proteins always showed larger tube dispersion vs. small molecules, especially when the residence time in the tube was long (at low flow rate). This confirms the need to drastically reduce the tube volume (using the shortest and narrowest connector tubing) to attain the full benefits of UHPSEC columns. In addition, PEEK tubing were found to be more appropriate than stainless steel tubing, since adsorption of proteins was less pronounced, and higher plate count can be obtained. Finally, after a careful system optimization, up to 40% increase of apparent column efficiency can be achieved compared to a regular UHPLC system, when using a 150×4.6mm UHPSEC columns packed with sub-3μm particles. Copyright © 2017 Elsevier B.V. All rights reserved.

  3. Loading and Contact Stress Analysis on the Thread Teeth in Tubing and Casing Premium Threaded Connection

    Directory of Open Access Journals (Sweden)

    Honglin Xu

    2014-01-01

    Full Text Available Loading and contact stress distribution on the thread teeth in tubing and casing premium threaded connections are of great importance for design optimization, pretightening force control, and thread failure prevention. This paper proposes an analytical method based on the elastic mechanics. This is quite different from other papers, which mainly rely on finite element analysis. The differential equation of load distribution on the thread teeth was established according to equal pitch of the engaged thread after deformation and solved by finite difference method. Furthermore, the relation between load acting on each engaged thread and mean contact stress on its load flank is set up based on the geometric description of thread surface. By comparison, this new analytical method with the finite element analysis for a modified API 177.8 mm premium threaded connection is approved. Comparison of the contact stress on the last engaged thread between analytical model and FEM shows that the accuracy of analytical model will decline with the increase of pretightening force after the material enters into plastic deformation. However, the analytical method can meet the needs of engineering to some extent because its relative error is about 6.2%~18.1% for the in-service level of pretightening force.

  4. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1984-01-01

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U 3 O 8 -aluminum cermets. Above the aluminum melting point, U 3 O 8 and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and therefore prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube sections with 53 wt % U 3 O 8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 900 0 C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U 3 O 8 -aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660 0 C. In tension, sections failed at about the aluminum melting point. In compression with 2 psi average axial stress, failure occurred at 917 0 C, while 7 psi average axial stress produced failure at 669 0 C. (author)

  5. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1983-01-01

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U 3 O 8 -aluminum cermets. Above the aluminum melting point, U 3 O 8 and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube-sections with 53 wt % U 3 O 8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 900 0 C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U 3 O 8 -aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660 0 C. In tension, sections failed at about the aluminum melting point. In compression with 2-psi average axial stress, failure occurred at 917 0 C, while 7 psi average axial stress produced failure at 669 0 C

  6. Evaluation of wrapper tube temperatures of fast neutron reactors using the TRANSCOEUR-2 code

    Energy Technology Data Exchange (ETDEWEB)

    Valentin, B.; Brun P. [CEA/DRN/DEC/SECA/LHC CEN, St Paul Lez Durance (France); Chaigne, G. [FRAMATOME/NOVATOME, Lyon (France)

    1995-09-01

    This paper deals with the thermal loading estimation of wrapper tubes using the TRANSCOEUR-2 code. This estimation requires a knowledge of two temperature fields: the first involves the peripheral sub-channel temperatures of each sub-assembly calculated by the design code CADET, and the second, outside the sub-assemblies, is the inter-wrapper flow temperature field calculated by the thermal-hydraulic code TRIO-VF with boundary conditions taken from CADET. Theoretical models of the three codes are presented as well as the first TRANSCOEUR-2 wrapper tube temperature calculation performed on the European Fast Reactor (EFR) Core Design 6/91 (CD 6/91) under nominal power conditions. The results show a temperature variation of 115{degrees}C between the bottom of the lower blanket and the top of the upper blanket fuel sub-assemblies in the center of the core and 95{degrees}C at the core periphery. The wrapper tube temperatures are higher in the center than in the external core.

  7. Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

  8. Raw materials problems in connection with fast breeder type reactors

    International Nuclear Information System (INIS)

    Hirsch, H.; Kreusch, J.

    1981-01-01

    The power supply by the FBR type reactors which depends upon the availability of essential raw materials such as Cr and Mo for structural and special steels is supposed to be less ensured than supply by fossil-fueled power plants. This contribution tries to verify this statement by means of estimates of the annual Cr and Mo demand, of the resources, production and consumption as well as by a study of the possibilities of recycling and substituting Cr and Mo. The only realistic alternative to the fast breeder type reactor is supposed to be a soft path of development according to the principle of decentralization, utilization of renewable energy sources regard to environmental protection, and use of less sophisticated technology. (DG) [de

  9. Power control device for nuclear reactors

    International Nuclear Information System (INIS)

    Kagawa, Tatsuo

    1984-01-01

    Purpose: To eliminate for requirement of control rods and movable portions, as well as ensure the safety and reliability of the operation. Constitution: A plurality of control tubes are disposed within a reactor core instead of control rods. Tubes are connected from below the reactor core to the control tubes for supplying liquid poisons such as aqueous boric acid to the inside of the control tubes. Further, tubes are connected to the upper portion of the control tubes for guiding the liquid poisons from the reactor core to the outside. The tubes for supplying and discharging the liquid poisons are introduced externally through the flange disposed at the upper portion of a pressure vessel. At the outside of the pressure vessel, are disposed a liquid poison tank, a pressurizing source, a pressure control valve, a liquid level meter and the like. The control for the reactor power is conducted by controlling the level of the liquid poisons in the control tubes. (Ikeda, J.)

  10. Method of experimental and theoretical modeling for multiple pressure tube rupture for RBMK reactor

    International Nuclear Information System (INIS)

    Medvedeva, N.Y.; Goldstein, R.V.; Burrows, J.A.

    2001-01-01

    The rupture of single RBMK reactor channels has occurred at a number of stations with a variety of initiating events. It is assumed in RBMK Safety Cases that the force of the escaping fluid will not cause neighbouring channels to break. This assumption has not been justified. A chain reaction of tube breaks could over-pressurise the reactor cavity leading to catastrophic failure of the containment. To validate the claims of the RBMK Safety Cases the Electrogorsk Research and Engineering Centre, in participation with experts from the Institute of Mechanics of RAS, has developed the method of interacting multiscale physical and mathematical modelling for coupled thermophysical, hydrogasodynamic processes and deformation and break processes causing and (or) accompanying potential failures, design and beyond the design RBMK reactor accidents. To realise the method the set of rigs, physical and mathematical models and specialized computer codes are under creation. This article sets out an experimental philosophy and programme for achieving this objective to solve the problem of credibility or non-credibility for multiple fuel channel rupture in RBMK.(author)

  11. Highly Selective Continuous Flow Hydrogenation of Cinnamaldehyde to Cinnamyl Alcohol in a Pt/SiO2 Coated Tube Reactor

    Directory of Open Access Journals (Sweden)

    Yang Bai

    2018-02-01

    Full Text Available A novel continuous flow process for selective hydrogenation of α, β-unsaturated aldehyde (cinnamaldehyde, CAL to the unsaturated alcohol (cinnamyl alcohol, COL has been reported in a tube reactor coated with a Pt/SiO2 catalyst. A 90% selectivity towards the unsaturated alcohol was obtained at the aldehyde conversion of 98.8%. This is a six-fold improvement in the selectivity compared to a batch process where acetals were the main reaction products. The increased selectivity in the tube reactor was caused by the suppression of acid sites responsible for the acetal formation after a short period on stream in the continuous process. In a fixed bed reactor, it had a similar acetal suppression phenomenon but showed lower product selectivity of about 47–72% due to mass transfer limitations. A minor change in selectivity and conversion caused by product inhibition was observed during the 110 h on stream with a turnover number (TON reaching 3000 and an alcohol production throughput of 0.36 kg gPt−1 day−1 in the single tube reactor. The catalysts performance after eight reaction cycles was fully restored by calcination in air at 400 °C. The tube reactors provide an opportunity for process intensification by increasing the reaction rates by a factor of 2.5 at the reaction temperature of 150 °C compared to 90 °C with no detrimental effects on catalyst stability or product selectivity.

  12. Method and apparatus for testing closed-end tubes in heat exchangers of nuclear reactors and the like

    International Nuclear Information System (INIS)

    Seyd, G.; Bergbauer, A.; Paulsen, U.

    1975-01-01

    A description is given of a test stopper which is insertable into a tube closed at one end for testing the tightness of the tube with a fluid under pressure, the tube being in a heat exchanger of a nuclear reactor or the like. The test stopper includes a tubular outer jacket that is expandable outwardly to tightly seat the stopper in the tube. The stopper also has front and back end-face members joined to the ends of the outer jacket to define a closed space within the jacket. With the stopper inserted into the tube, the front end-face member and the closed end portion of the tube define a closed inner region of the tube. An inner tubular member, disposed within the outer jacket, partitions the closed space within the jacket into an annular outer chamber and a cylindrical inner chamber. A pressure-fluid supply selectively supplies fluid to the chambers. The outer jacket expands in response to fluid admitted to the annular chamber and the front end-face member has a through bore to admit fluid under pressure to the inner region of the tube. A method of testing of such a tube with a fluid under pressure includes inserting the test stopper into the tube and then expanding the outer jacket of the stopper to seat the stopper firmly in the tube. A fluid under pressure is directed through the stopper and into the closed region defined by the front end-face member of the stopper and the closed end portion of the tube. The pressure of the fluid introduced into this closed region is monitored for detecting a leak in the closed-end tube

  13. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem mirror fusion reactor

    International Nuclear Information System (INIS)

    Werner, R.W.

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  14. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem-mirror fusion reactor

    International Nuclear Information System (INIS)

    Werner, R.W.; Hoffman, M.A.; Johnson, G.L.

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  15. X-ray diffraction device comprising cooling medium connections provided on the x-ray tube

    NARCIS (Netherlands)

    1996-01-01

    An X-ray diffraction device comprises a water-cooled X-ray tube which exhibits a line focus as well as, after rotation through 90 DEG , a point focus. Contrary to customary X-ray tubes, the cooling water is not supplied via the housing (12) in which the X-ray tube is mounted, but the cooling water

  16. Combustion, cofiring and emissions characteristics of torrefied biomass in a drop tube reactor

    International Nuclear Information System (INIS)

    Ndibe, Collins; Maier, Jörg; Scheffknecht, Günter

    2015-01-01

    The study investigates cofiring characteristics of torrefied biomass fuels at 50% thermal shares with coals and 100% combustion cases. Experiments were carried out in a 20 kW, electrically heated, drop-tube reactor. Fuels used include a range of torrefied biomass fuels, non-thermally treated white wood pellets, a high volatile bituminous coal and a lignite coal. The reactor was maintained at 1200 °C while the overall stoichiometric ratio was kept constant at 1.15 for all combustion cases. Measurements were performed to evaluate combustion reactivity, emissions and burn-out. Torrefied biomass fuels in comparison to non-thermally treated wood contain a lower amount of volatiles. For the tests performed at a similar particle size distribution, the reduced volatile content did not impact combustion reactivity significantly. Delay in combustion was only observed for test fuel with a lower amount of fine particles. The particle size distribution of the pulverised grinds therefore impacts combustion reactivity more. Sulphur and nitrogen contents of woody biomass fuels are low. Blending woody biomass with coal lowers the emissions of SO 2 mainly as a result of dilution. NO X emissions have a more complex dependency on the nitrogen content. Factors such as volatile content of the fuels, fuel type, furnace and burner configurations also impact the final NO X emissions. In comparison to unstaged combustion, the nitrogen conversion to NO X declined from 34% to 9% for air-staged co-combustion of torrefied biomass and hard coal. For the air-staged mono-combustion cases, nitrogen conversion to NO X declined from between 42% and 48% to about 10%–14%. - Highlights: • Impact of torrefaction on cofiring was studied at high heating rates in a drop tube. • Cofiring of torrefied biomasses at high thermal shares (50% and higher) is feasible. • Particle size impacts biomass combustion reactivity more than torrefaction. • In a drop tube reactor, torrefaction has no negative

  17. Selectivity of benzene sulphonation in three gas—liquid reactors with different mass transfer characteristics: II: Mass transfer and selectivity in a cyclone reactor and in a tube reactor

    NARCIS (Netherlands)

    Beenackers, Antonie A.C.M.; van Swaaij, Willibrordus Petrus Maria

    1978-01-01

    Liquid benzene was sulphonated with gaseous sulphur trioxide in a tube reactor and in a new gas—liquid cyclone reactor. The products are benzenesulphonic acid and diphenyl sulphone (byproduct). The observed selectivity depends on the conversion, the initial benzene concentration and the mass

  18. Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

    1978-12-01

    This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

  19. In-service inspection of sub-coating defects in PWR reactor vessel tubes

    International Nuclear Information System (INIS)

    Birac, A.; Frappier, J.C.; Saglio, Robert.

    1982-08-01

    Since the presence of cracks under the coating of the tubes of certain PWR reactor vessels were noted during manufacture, the need emerged to develop a nondestructive testing method to guarantee the detection of existing cracks and to determine their potential evolution. An ultrasonic testing method was developed for the purpose. In Part 1, the choice of ultrasonic transducers is justified from the theoretical and practical standpoints. In Part 2, the results obtained on test specimens containing artificial defects are presented in accordance with the different parameters involved. In Part 3, covering parts with a large number of real defects, the results of real defect/recorded signal correlations are given, with respect to both detection and dimensions. Examples of automatic data processing are analyzed [fr

  20. β-85 wt % Nb precipitates: the effect on in-reactor diametric creep of pressure tubes

    International Nuclear Information System (INIS)

    Sarce, Alicia

    2006-01-01

    By linking the microstructure evolution of an α-Zr crystal with the macroscopic behaviour, the deformation of an in-service reactor pressure tube is calculated. Microstructure evolution is considered through rate theory modelling of the interaction between point defects and sinks. Different densities of β-85 wt % Nb precipitates are proposed to be distributed inside the α grains and act as point defect sinks, doing a screening effect on the grain boundaries. From the interatomic pair potential which is used to describe the material, positive tangential deformation rates (on the other hand negatives) are obtained when these densities in the c-crystal direction are bigger than a minimum value. (author) [es

  1. Steam-generator-tube-rupture transients for pressurized-water reactors

    International Nuclear Information System (INIS)

    Dobranich, D.; Henninger, R.J.; DeMuth, N.S.

    1982-01-01

    Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures

  2. Preloading of bolted connections in nuclear reactor component supports

    Energy Technology Data Exchange (ETDEWEB)

    Yahr, G T

    1984-10-01

    A number of failures of threaded fasteners in nuclear reactor component supports have been reported. Many of those failures were attributed to stress corrosion cracking. This report discusses how stress corrosion cracking can be avoided in bolting by controlling the maximum bolt preloads so that the sustained stresses in the bolts are below the level required to cause stress corrosion cracking. This is a basic departure from ordinary bolted joint design where the only limits on preload are on the minimum preload. Emphasis is placed on the importance of detailed analysis to determine the acceptable range of preload and the selection of a method for measuring the preload that is sufficiently accurate to ensure that the preload is actually within the acceptable range. Procedures for determining acceptable preload range are given, and the accuracy of various methods of measuring preload is discussed.

  3. Preloading of bolted connections in nuclear reactor component supports

    International Nuclear Information System (INIS)

    Yahr, G.T.

    1984-10-01

    A number of failures of threaded fasteners in nuclear reactor component supports have been reported. Many of those failures were attributed to stress corrosion cracking. This report discusses how stress corrosion cracking can be avoided in bolting by controlling the maximum bolt preloads so that the sustained stresses in the bolts are below the level required to cause stress corrosion cracking. This is a basic departure from ordinary bolted joint design where the only limits on preload are on the minimum preload. Emphasis is placed on the importance of detailed analysis to determine the acceptable range of preload and the selection of a method for measuring the preload that is sufficiently accurate to ensure that the preload is actually within the acceptable range. Procedures for determining acceptable preload range are given, and the accuracy of various methods of measuring preload is discussed

  4. Expansion connection of socket in flow distributed cabin of heavy water research reactor inner shell

    International Nuclear Information System (INIS)

    Jiang Zhiliang; Li Yanshui

    1995-01-01

    Expansion connection of aluminium alloy LT21 socket in flow distributed cabin of Heavy Water Research Reactor (HWRR) inner shell is described systematically. The expansion connection technology parameters of products are determined through tests. They are as following: bounce value of inner diameter after expansion, expansion degree, space between socket and plate hole, device for expanding pipes, selection of tools for enlarging or reaming holes, manufacture for socket inner hole and cleaning after expansion

  5. Physical aspects of liquid-impelled loop reactors

    NARCIS (Netherlands)

    Sonsbeek, van H.

    1992-01-01

    The liquid-impelled loop reactor (LLR) is a reactor that consists of two parts : the main tube and the circulation tube. Both parts are in open connection at the bottom and at the top. The reactor is filled with a liquid phase: the continuous phase. Another liquid phase is injected in the

  6. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    International Nuclear Information System (INIS)

    Schulz, K.C.

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K Q due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail

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

    International Nuclear Information System (INIS)

    Muscaloiu, C.

    1997-01-01

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

  8. Subcadmic and epicadmic flow in the dry tube of the TRIGA Mark III reactor of the Nuclear Center of Mexico

    International Nuclear Information System (INIS)

    Delfin L, A.; Mazon R, R.; Nava R, B.

    1991-04-01

    The mensuration of the thermal and fast flows of the irradiation facilities of the core of the reactor is important, since allow us to determine the optimum time of irradiation of the samples in the reactor. The Dry Tube especially, is an irradiation installation that it was designed in the I.N.I.N. to supply the pneumatic irradiation system of capsules with durations bigger than 15 minutes and it can be used for exposures until a maximum of three hours. The main users are the Nuclear Chemistry Department and the Neutron activation analysis. In this report the neutron flux sub cadmic and epi cadmic obtained in an experimental way in the Dry Tube for the reactor operating in stationary state to powers of 100 Kw, 300 Kw and 1000 Kw are reported and with these values it is interpolated for other powers. (Author)

  9. Savannah River reactor process water heat exchanger tube structural integrity margin Task Number 92-005-1

    International Nuclear Information System (INIS)

    Mertz, G.E.; Barnes, D.M.; Sindelar, R.L.

    1992-02-01

    Twelve process water heat exchangers are designed to remove heat generated in the reactor tank. Each heat exchanger has approximately 9000, 1/2 inch diameter x 0.049 inches thick tubes. Minimum structural tubing requirements and the leak rate through postulated tubing defects are developed in this report A comparison of the structural requirements and the defect size calculated to produce leak rates of 0.5 lbs./day demonstrate adequate structural margins against gross tube rupture. Commercial nuclear experience with pressurized water reactor (PWR) steam generator plugging criteria are used for guidance in performing this analysis. It is important to note that the SRS reactors are low energy systems with normal operating pressures of 203 psig at 130 degree F while the PWR is a high energy system with operating pressures near 2200 psig at 600 degree F. Clearly the PVM steam generator has loadings which are more severe than the SRS heat exchangers. Consistent with the Regulatory Guide 1.121 criteria both wastage (wall thinning) and cracking are addressed. Structural limits on wall thinning and crack size are developed to preclude gross rupture. ASME Section XI criteria, with the factors of safety recommended by Regulatory Guide 1.121 are used to develop the allowable crack size criteria. Normal operating conditions (pressure, dead weight, and hydraulic drag) are considered with seismic and water hammer accident conditions. Both the wall thinning and crack size criteria are developed for the end-of-evaluation period. Allowances for corrosion, wear, or crack growth have not been included in this analysis Structurally, the tubing is over designed and can tolerate large defects with adequate margins against gross rupture. The structural margins of heat exchanger tubing are evident by contrasting the tubing's structural capacity, per the ASME Code, with its operating conditions/configuration

  10. Intercrystalline Stress Corrosion of Inconel 600 Inspection Tubes in the Aagesta Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Groenwall, B; Ljungberg, L; Huebner, W; Stuart, W

    1966-08-15

    Intercrystalline stress corrosion cracking has occurred in the Aagesta reactor in three so-called inspection tubes made of Inconel 600. The tubes had been exposed to 217 deg C light water, containing 1-4 ppm LiOH (later KOH) but only small amounts of oxygen, chloride and other impurities. Some of the circumferential cracks developed in or at crevices on the outside surface. At these positions constituents dissolved in the water may have concentrated. The crevices are likely to have contained a gas phase, mainly nitrogen. Local boiling in the crevices may also have occurred. Some few cracks were also found outside the crevice region. Irradiation effects can be neglected. No surface contamination could be detected except for a very minor fluoride content (1 {mu}g/cm{sup 2}). The failed tubes had been subjected to high stresses, partly remaining from milling, partly induced by welding operations. The possibility that stresses slightly above the 0.2 per cent offset yield strength have occurred at the operating temperature cannot be excluded. The cracked tube material contained a large amount of carbide particles and other precipitates, both at grain boundaries and in the interior of grains. The particles appeared as stringers in circumferential zones. Zones depleted in precipitates were found along grain boundaries. The failed tube turned out to have an unusually high mechanical strength, likely due to a combination of some kind of ageing process and cold work (1.0 - 1.3 per cent plastic strain). Laboratory exposures of stressed surplus material in high purity water and in 1 M LiOH at 220 deg C showed some pitting but no cracking after 6800 h and 5900 h respectively. Though the encountered failures may have developed because of influence of some few or several of the above-mentioned detrimental factors, the actual cause cannot be stated with certainty. In the literature information is given concerning intercrystalline stress corrosion cracking of Inconel 600 both in

  11. Intercrystalline Stress Corrosion of Inconel 600 Inspection Tubes in the Aagesta Reactor

    International Nuclear Information System (INIS)

    Groenwall, B.; Ljungberg, L.; Huebner, W.; Stuart, W.

    1966-08-01

    Intercrystalline stress corrosion cracking has occurred in the Aagesta reactor in three so-called inspection tubes made of Inconel 600. The tubes had been exposed to 217 deg C light water, containing 1-4 ppm LiOH (later KOH) but only small amounts of oxygen, chloride and other impurities. Some of the circumferential cracks developed in or at crevices on the outside surface. At these positions constituents dissolved in the water may have concentrated. The crevices are likely to have contained a gas phase, mainly nitrogen. Local boiling in the crevices may also have occurred. Some few cracks were also found outside the crevice region. Irradiation effects can be neglected. No surface contamination could be detected except for a very minor fluoride content (1 μg/cm 2 ). The failed tubes had been subjected to high stresses, partly remaining from milling, partly induced by welding operations. The possibility that stresses slightly above the 0.2 per cent offset yield strength have occurred at the operating temperature cannot be excluded. The cracked tube material contained a large amount of carbide particles and other precipitates, both at grain boundaries and in the interior of grains. The particles appeared as stringers in circumferential zones. Zones depleted in precipitates were found along grain boundaries. The failed tube turned out to have an unusually high mechanical strength, likely due to a combination of some kind of ageing process and cold work (1.0 - 1.3 per cent plastic strain). Laboratory exposures of stressed surplus material in high purity water and in 1 M LiOH at 220 deg C showed some pitting but no cracking after 6800 h and 5900 h respectively. Though the encountered failures may have developed because of influence of some few or several of the above-mentioned detrimental factors, the actual cause cannot be stated with certainty. In the literature information is given concerning intercrystalline stress corrosion cracking of Inconel 600 both in caustic

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  13. Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II; Desgaste placas tubos guia barras de control interno superior vasija del reactor C.N. Vandellos II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-01

    The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

  14. Method of reactor fueling

    International Nuclear Information System (INIS)

    Saito, Toshiro.

    1983-01-01

    Purpose: To decrease the cost and shorten the working time by saving fueling neutron detectors and their components. Method: Incore drive tubes for the neutron source range monitor (SRM) and intermediate range monitor (IRM) are disposed respectively within in a reactor core and a SRM detector assembly is inserted to the IRM incore drive tube which is most nearest to the neutron source upon reactor fueling. The reactor core reactivity is monitored by the SRM detector assembly. The SRM detector asesembly inserted into the IRM drive tube is extracted at the time of charging fuels up to the frame connecting the SRM and, thereafter, IRM detection assembly is inserted into the IRM drive tube and the SRM detector assembly is inserted into the SRM drive tube respectively for monitoring the reactor core. (Sekiya, K.)

  15. Integration of a photocatalytic multi-tube reactor for indoor air purification in HVAC systems: a feasibility study.

    Science.gov (United States)

    van Walsem, Jeroen; Roegiers, Jelle; Modde, Bart; Lenaerts, Silvia; Denys, Siegfried

    2018-04-24

    This work is focused on an in-depth experimental characterization of multi-tube reactors for indoor air purification integrated in ventilation systems. Glass tubes were selected as an excellent photocatalyst substrate to meet the challenging requirements of the operating conditions in a ventilation system in which high flow rates are typical. Glass tubes show a low-pressure drop which reduces the energy demand of the ventilator, and additionally, they provide a large exposed surface area to allow interaction between indoor air contaminants and the photocatalyst. Furthermore, the performance of a range of P25-loaded sol-gel coatings was investigated, based on their adhesion properties and photocatalytic activities. Moreover, the UV light transmission and photocatalytic reactor performance under various operating conditions were studied. These results provide vital insights for the further development and scaling up of multi-tube reactors in ventilation systems which can provide a better comfort, improved air quality in indoor environments, and reduced human exposure to harmful pollutants.

  16. Experimental investigation of the vibration response of a flexible tube due to simulated reactor core, cross and annular exit flows

    International Nuclear Information System (INIS)

    Haslinger, K.H.; Martin, M.L.; Higgins, W.H.; Rossano, F.V.

    1989-01-01

    Instrumentation tubes in pressurized nuclear reactors have experienced wear due to excessive flow-induced vibrations. Experiments to identify the predominant flow excitation mechanism at a particular plant, and to develop a sleeve design to remedy the wear problem are reported. An instrumented flow visualization model enabled simulation of a wide range of individual or combined reactor core flow, cross flow and thimble flow conditions. The instrumentation scheme adopted for these experiments used proximity displacement transducers and a force transducer to measure respectively tube motion and contact/impact forces at the wear region. Extensive testing of the original, in-plant configuration identified the normal core flow as the primary source of excitation. Shielding the In-Core-Instrumentation thimble tube from the normal core flow curtailed vibration amplitudes; however, thimble flow excitation then became more pronounced. Various outlet nozzle configurations were investigated. An internal cavity combined with radial outlet slots became the optimum solution for the problem. The paper presents typical test data in the form of orbital tube motion, spectrum analysis and time history collages. The effectiveness of shielding the instrumentation tube from the flow is demonstrated. (author)

  17. Methods of evaluation of accuracy with multiple essential parameters for eddy current measurement of pressure tube to calandria tube gap in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shokralla, S., E-mail: shaddy.shokralla@opg.com [Ontario Power Generation, IMS NDE Projects, Ajax, Ontario (Canada); Krause, T.W., E-mail: thomas.krause@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2014-01-15

    The purpose of inspection qualification of a particular inspection system is to show that it meets applicable inspection specification requirements. Often a requirement of the inspection system is that it meets a particular accuracy. In the case of a system with multiple inputs accompanied by additional influential parameters, calculation of the system's output accuracy can be formidable. Measurement of pressure-tube to calandria tube gap in CANDU reactors using an eddy current based technique is presented as a particular example of a system where multiple essential parameters combine to generate a final uncertainty for the inspection system. This paper outlines two possible methods of calculating such a system's accuracy, and discusses the advantages and disadvantages of each. (author)

  18. High-definition radiography of tube-to-tubesheet welds of steam generator of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Venkatraman, B.; Sethi, V.K.; Jayakumar, T.; Raj, B.

    1995-01-01

    In the steam generator of the Prototype Fast Breeder Reactor (PFBR), steam is generated by the transfer of heat from secondary sodium to water. Due to the inherent dangers of sodium-water reaction, the integrity of weld joints separating sodium and water/steam is of paramount importance. This is particularly true and very important for the tube-to-tubesheet joints. This paper discusses the use of projective magnification technique by microfocal radiography for the quality evaluation and optimisation of the welding parameters of such small tube-to-tubesheet welds of the steam generator of PFBR. (author)

  19. Catalytic Chan–Lam coupling using a ‘tube-in-tube’ reactor to deliver molecular oxygen as an oxidant

    Directory of Open Access Journals (Sweden)

    Carl J. Mallia

    2016-07-01

    Full Text Available A flow system to perform Chan–Lam coupling reactions of various amines and arylboronic acids has been realised employing molecular oxygen as an oxidant for the re-oxidation of the copper catalyst enabling a catalytic process. A tube-in-tube gas reactor has been used to simplify the delivery of the oxygen accelerating the optimisation phase and allowing easy access to elevated pressures. A small exemplification library of heteroaromatic products has been prepared and the process has been shown to be robust over extended reaction times.

  20. New control strategy for grid connecting of wind turbine inverter without converter reactor

    DEFF Research Database (Denmark)

    Rasmussen, Tonny Wederberg; Sørensen, Kasper B.; Bjørneboe, Daniel

    2013-01-01

    Wind turbines and the belonging converters increase in size and price. A reduction in the number of main components is desirable while it reduces need of space, investments and increases the efficiency. A wind turbine contains both a converter reactor and a step up transformer. The paper presents...... theory and laboratory measurements for a new control strategy which make it possibly to connect a wind turbine converter to the utility grid without using the converter reactor or make measurements at the high voltage side of the transformer. The capability to control the DC voltage and reactive power...

  1. Predicting diametral creep of the pressure tubes in CANDU reactors using fuzzy neural networks

    International Nuclear Information System (INIS)

    Lee, Jae Yong; Na, Man Gyun; Park, Jong Ho

    2011-01-01

    Pressure tube (PT) creep is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS) in Canada deuterium uranium reactors. PT diametral creep affects the thermal hydraulic characteristics of coolant channels and the critical heat flux (CHF). CHF is a key parameter in determining the critical channel power, which is used in the trip setpoint calculations of regional overpower protection systems. This paper aims to predict PT diametral creep using the measured signals of the HTS by applying fuzzy neural networks (FNNs) according to operating conditions. The FNN model was optimized in terms of its fuzzy rules and parameters by a genetic algorithm combined with the least-squares method. Informative data that demonstrate the system's characteristic behavior were selected to train the FNN model using the subtractive clustering method. The proposed FNN model for predicting PT diametral creep was verified using the operating data of the Wolsong Unit 1 nuclear power plant in Korea. It was known that the FNN could predict the PT diametral creep accurately. Statistical and analytical uncertainty analysis methods were applied to the models and their uncertainties were evaluated using 60 sampled training and optimization data sets, as well as two fixed test data sets. (author)

  2. Neutron spectrum measurements from a neutron guide tube facility at the ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maayouf, R M.A.; El-Sayed, L A.A.; El-Kady, A S.I. [Reactor and Neutron Physics Dept., NRC, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The present work deals with measurements of the neutron spectrum emitted from a neutron guide tube (NGT) recently installed at one of the ETRR-1 reactor horizontal channels designed to deliver thermal neutrons, free from fast neutrons and gamma ray background, to a fourier reverse-time-of-flight (RTOF) diffractometer. The measurements were performed using a {sup 6} Li glass scintillation detector combined with a multichannel analyzer set at channel width 4 M sec and installed at 3.4 m from a disc Fermi chopper. Also a theoretical model was specially developed for the neutron spectrum calculations. According to the model developed, the spectrum calculated was found to be in good agreement with the measured one. It was found, both from measurements and calculations, that the spectrum emitted from the NGT covers, after transmission through a fourier chopper, neutron wavelengths from 1-4 A adequate for neutron diffraction measurements at D values between 0.71-2.9 A respectively. 6 FIGS.

  3. Results from integral tests of single reformer tubes under simulated nuclear reactor conditions

    International Nuclear Information System (INIS)

    Decken, C.B. von der; Fedders, H.; Harth, R.; Hoehlein, B.; Riensche, E.

    1980-01-01

    The possibility of supplying high temperature heat from a HTGR for process application is being investigated at some places in the world. In all programmes or projects existing with respect to this application, the endothermic steam reforming of methane is one main step in the transmission of heat produced by nuclear fission to different chemical processes. The KFA is involved in the two German projects PNP - Prototypanlage Nukleare Prozesswaerme (Prototype-plant Nuclear Process-heat), and NFE -Nukleare Fernenergie (Long Distance Energy Transport). In a HTGR, helium generally serves as reactor coolant. It transports the heat from the core to the different components which take over this heat for various purposes. In case of arranging a steam reformer in the helium circuit, it is necessary for economic reasons to reach very high temperatures. In the two German projects mentioned above, the helium temperature at HTGR core outlet is determined to 950 0 C. Thus the main design data for a steam reformer supplied by heat from a HTGR are maximum helium temperature 950 0 C, helium pressure 40 bar. By an extensive utilization of the available advanced conventional steam reforming technology, the helium heated steam reformer design is using normal steam reforming tubes arranged in compact bundles

  4. Control rod guide tube wear in operating reactors; operating experience report. Technical report December 1977-December 1979

    International Nuclear Information System (INIS)

    Riggs, R.

    1980-04-01

    Evidence of control rod guide tube wear has been observed in operating pressurized water reactors. The cause of this wear is identified as flow-induced vibration of the control rods. This report describes the measures being taken by both the industry and the NRC to deal with this matter. The staff also presents its technical positions and requirements to support continued operation of the plants as of December 1979 pending completion of this generic effort

  5. Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 3000C

    International Nuclear Information System (INIS)

    Pandarinathan, P.R.; Vasudevan, P.

    1980-01-01

    Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300 0 C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300 0 C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300 0 C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-01

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

  7. Hexagonal tube behaviour in fuel assemblies under neutron flux in a French fast neutron reactor core

    International Nuclear Information System (INIS)

    Bernard, A.; Ammann, P.

    This paper presents what is obtained in the field of the interpretation by calculation of the post irradiation examination of hexagonal tubes, and in the field of prevision by calculation of the behaviour of hexagonal tubes under fast flux [fr

  8. Reduction in degree of absorber-cladding mechanical interaction by shroud tube in control rods for the fast reactor

    International Nuclear Information System (INIS)

    Donomae, Takako; Katsuyama, Kozo; Tachi, Yoshiaki; Maeda, Koji; Yamamoto, Masaya; Soga, Tomonori

    2011-01-01

    Research and development of a long-life control rod for fast reactors is being conducted at Joyo. One of the challenges in developing a long-life control rod is the restraint of absorber-cladding mechanical interaction (ACMI). First, a helium-bonding rod was selected as a control rod for the experimental fast reactor Joyo, which is the first liquid metal fast reactor in Japan. Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B 4 C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B 4 C pellets and the cladding tube. However, once B 4 C pellets cracked and broke into small fragments, relocation occurred. After this, the narrow gap closed immediately as the degree of B 4 C pellet swelling increased. To solve this problem, the gap was widened during design, and sodium was selected as the bonding material instead of helium to restrain the increase in pellet temperature. Irradiation testing of the modified sodium-bonding control rod confirmed that ACMI would be restrained by the shroud tube regardless of the occurrence of B 4 C pellet relocation. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of postirradiation examination are reported. (author)

  9. Nondestructive inspection of the tubes of TRIGA IPR-R1 reactor heat exchanger by eddy current testing

    International Nuclear Information System (INIS)

    Silva Junior, Silverio F.; Silva, Roger F.; Oliveira, Paulo F.; Barreto, Erika S.; Ribeiro, Isabela G.; Fraiz, Felipe C.

    2013-01-01

    The IPR-R1 TRIGA MARK 1 reactor is an open pool type reactor, cooled light water. It is used for research activities, personnel training and radioisotopes production, in operation since 1960 at the Nuclear Technology Development Center - CDTN/CNEN. It operates at a maximum thermal power of 100 kW and usually, the fuel cooling is done by natural circulation. If necessary, an external auxiliary cooling system, with a shell-and-tube type heat exchanger, can be used to improve the water heat removal. As part of the ageing management program of the reactor, a nondestructive evaluation of their heat exchanger stainless steel tubes will be performed, in order to verify its integrity. The examinations will be performed using the eddy current test method, which allows the detection and characterization of structural discontinuities in the wall of the tubes, if existing. For this purpose, probes and reference standards were designed and manufactured at CDTN facilities and test procedures were established and validated. In this paper, a description of the proposed infrastructure as well as the test methodology to be used in the examinations are presented and discussed. (author)

  10. Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high

  11. Numerical Study on the Behaviour of Reduced Beam Section Presence in Rectangular Concrete Filled Tubes Connection

    Science.gov (United States)

    Amalia, A. R.; Suswanto, B.; Kristijanto, H.; Irawan, D.

    2018-01-01

    This paper discusses about the behaviour of two types of RCFT column connections with steel beams due to cyclic loads using software based on finite element method ABAQUS 6.14. This comparison involves modelling RCFT connections with rigid connection that do not allow any deformation and rotation in the joint. There are two types of model to be compared: BB and BRBS which include RCFT connections to ordinary beam without RBS (BB) and to Reduce Beam Section Beam (BRBS). The models behaviour can be discussed in this study are stress value, von misses stress pattern and rotational degree of each model. From the von misses stress pattern value, it found that the highest regions of stress occurs in vicinity of beam flange near column face for connection without RBS (BB). For earthquake resistant building, that behaviour needs to be avoided because sudden collapse often happen in that joint connection. Moreover, the connection with the presence of RBS (BRBS), the highest regions of stress occurs in reduced beam section of the beam, it means that the failure might be happen as proposed plan. The ultimate force that can be restrained by BB model (402 kN) is higher than BRBS model (257,18 kN) because of reducing of flange area. BRBS model has higher rotation angle (0,057 rad) than BB model (0,045 rad). The analysis results also observed that cyclic performances of the moment connection with RBS (BRBS) were more ductile than the connection with ordinary beam (BB).

  12. Connecting Music Education and Virtual Performance Practices from YouTube

    Science.gov (United States)

    Cayari, Christopher

    2018-01-01

    The Internet has inspired musicians to explore technologies to produce recorded music performances. Social media sites like YouTube provide spaces for musicians to share their works, and the advances of technologies that afford venues and opportunities for performers to share their crafts. As amateur Internet musicians develop practices to create…

  13. Measuring of tube expansion

    International Nuclear Information System (INIS)

    Vogeleer, J. P.

    1985-01-01

    The expansion of the primary tubes or sleeves of the steam generator of a nuclear reactor plant are measured while the tubes or sleeves are being expanded. A primary tube or sleeve is expanded by high pressure of water which flows through a channel in an expander body. The water is supplied through an elongated conductor and is introduced through a connector on the shank connected to the conductor at its outer end. A wire extends through the mandrel and through the conductor to the end of the connector. At its inner end the wire is connected to a tapered pin which is subject to counteracting forces produced by the pressure of the water. The force on the side where the wire is connected to the conductor is smaller than on the opposite side. The tapered pin is moved in the direction of the higher force and extrudes the wire outwardly of the conductor. The tapered surface of the tapered pin engages transverse captive plungers which are maintained in engagement with the expanding tube or sleeve as they are moved outwardly by the tapered pin. The wire and the connector extend out of the generator and, at its outer end, the wire is connected to an indicator which measures the extent to which the wire is moved by the tapered pin, thus measuring the expansion of the tube or sleeve as it progresses

  14. Process Intensification of Alkynol Semihydrogenation in a Tube Reactor Coated with a Pd/ZnO Catalyst

    Directory of Open Access Journals (Sweden)

    Nikolay Cherkasov

    2017-11-01

    Full Text Available Semihydrogenation of 2-methyl-3-butyn-2-ol (MBY was studied in a 5 m tube reactor wall-coated with a 5 wt% Pd/ZnO catalyst. The system allowed for the excellent selectivity towards the intermediate alkene of 97.8 ± 0.2% at an ambient H2 pressure and a MBY conversion below 90%. The maximum alkene yield reached 94.6% under solvent-free conditions and 96.0% in a 30 vol % MBY aqueous solution. The reactor stability was studied for 80 h on stream with a deactivation rate of only 0.07% per hour. Such a low deactivation rate provides a continuous operation of one month with only a two-fold decrease in catalyst activity and a metal leaching below 1 parts per billion (ppb. The excellent turn-over numbers (TON of above 105 illustrates a very efficient utilisation of the noble metal inside catalyst-coated tube reactors. When compared to batch operation at 70 °C, the reaction rate in flow reactor can be increased by eight times at a higher reaction temperature, keeping the same product decomposition of about 1% in both cases.

  15. Predicting the in-reactor mechanical behavior of Zr-2.5Nb pressure tubes from postirradiation microstructural examination data

    International Nuclear Information System (INIS)

    Griffiths, M.; Davies, P.H.; Davies, W.G.; Sagat, S.

    2002-01-01

    Postirradiation microstructure examinations of Zr-2.5Nb pressure tubes removed from service in CANDU reactors have shown clear trends in the dislocation structure and the state of the β-phase, as a function of operating temperature, neutron flux, and time. These microstructural parameters correlate well with changes in the mechanical properties. For example, the rapid increase in dislocation loop density in the early stages of irradiation corresponds with a rapid increase in tensile strength and DHC velocity, and a corresponding decrease in fracture toughness. There is also a strong negative correlation between the degree of β-phase decomposition and DHC velocity. In addition to the effects of microstructure evolution on the mechanical properties, changes in the a-type and c-component dislocation loop densities also affect irradiation deformation (creep and growth). Statistical analyses of the irradiation microstructure data have been used to derive empirical relationships between dislocation densities and β-phase structure with temperature, flux, and time. The relationships thus derived are useful in predicting where the mechanical properties are most affected by the in-reactor operating conditions. The predictions are compared with mechanical test data for samples from various axial and circumferential locations of 42 pressure tubes removed from operating CANDU reactors. The results are discussed in terms of the mechanisms controlling tensile strength, fracture, delayed-hydride-cracking, and in-reactor deformation. (author)

  16. Serpentine tube heat transfer characteristic under accident condition in gas cooled reactors

    International Nuclear Information System (INIS)

    Abouhadra, D.S.; Byrne, J.E.

    2004-01-01

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behavior of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. The Thermal Hydraulic Experimental Research Assembly was designed to operate with pressures up to 180 bar and temperatures of 450degC. The geometry and dimensions of this test section were similar to part of a gas cooled reactor boiler of the Hinkley Point design. Blowdown from a pressure of 60 bar with subcoolings of 5degC, 50degC, 100degC formed the main part of the programme. A set of tests was conducted using discharge orifices of different sizes to produce depressurization times from 30 s to 10 mins, and in a few cases, the duration of blowdown approached 1 hour. These times were defined using the criterion of blowdown end as a final pressure of 10% of the initial pressure. Pressures, wall and fluid temperatures were all measured at average time intervals of 1.1s during the excursion and an inventory of the remaining water content in the serpentine was taken when the blowdown ended. Some tests were also conducted at an initial pressure of 30 bar. The results obtained show interesting stratification effects for the relatively fast discharge, with substantial wall circumferential temperature variations. For these tests, a relatively small water inventory remained after blowdown. The discharge characteristics of the serpentine in terms of orifice size have been mapped, and tests at 30 bar show the equivalence in terms of orifice size have been mapped

  17. FEM analysis of mechanical behaviour of coil support connections in Wendelstein 7-X fusion reactor

    International Nuclear Information System (INIS)

    Krzesinski, G.; Zagrajek, T.; Marek, P.; Dobosz, R.; Czarkowski, P.; Kurzydlowski, K.J.

    2006-01-01

    The objective of Wendelstein 7-X project is the stellarator-type fusion reactor. In this device plasma channel is under control of magnetic field coming from magnet system of very complicated shape, made of 70 superconducting coils symmetrically arranged in 5 identical sections. Every coil is connected to central ring with two extensions which transfer loads resulting from electromagnetic field and gravity. The aim of this work was to analyse mechanical behaviour of the bolted connections using detailed 3D finite element models. All simulations were performed assuming elasto-plastic behaviour of the materials, assembly stresses and friction contacts between different parts of the connections. Stress distributions, displacements, forces acting on the bolts and welds were studied using standard and submodeling routines. The results were subsequently used to optimize the design of critical central support elements. (author)

  18. Steam generator tube failures: world experience in water-cooled nuclear power reactors in 1975

    International Nuclear Information System (INIS)

    Hare, M.G.

    1976-11-01

    Steam generator tube failures were reported in 22 out of 62 water-cooled nuclear power plants surveyed in 1975. This was less than in 1974, and the number of the tubes affected was noticeably less. This report summarizes these failures, most of which were due to corrosion. Secondary-water chemistry control, procedures for inspection and repair, tube materials, and failure rates are discussed. (author)

  19. Removal of portions of tubes from steam generator of nuclear reactor

    International Nuclear Information System (INIS)

    Wilkins, R.L.; Williams, C.F.

    1983-01-01

    After the tube portion to be removed is severed from the remainder of the U-tube and its weld to the header is machined off, the internal surface of the portion is engaged internally by an ID gripper and pulled out of the header. Then the external surface is engaged by an OD gripper and pulled further out of the header. The first tube length is pulled out as far as the space under the header permits and is then cut off. Successive lengths are likewise pulled out and cut off. The apparatus for accomplishing this object includes a base secured to the header by expanded mandrel mechanisms. A carriage is suspended from the base on screws which are driven by a motor to move the carriage away from and towards the base. An OD gripper assembly is suspended from the carriage and is movable by fluoroactuated piston rods away from and towards the carriage. An ID gripper assembly extends through the OD gripper assembly. The gripper of the ID assembly is actuable to engage the internal surface of the tube portion. With its gripper so engaged the ID assembly is engaged by the gripper of the OD assembly and the engaged tube portion is pulled out of the header by the OD assembly. The ID gripper is then disengaged and the OD gripper is engaged with the tube portion in the same way that it engages the ID assembly and the tube portion is pulled out further. The apparatus also includes a tube cutter having an abrasive wheel. The wheel cuts the lengths of the tube portion at an angle so that for examination and testing the tube lengths can be matched and the orientation of any defect with respect to the plate in the steam generator which separates the inlet and outlet ends of the tubes and the U-tube supports can be identified

  20. Corrugated thimble tube for controlling control rod descent in nuclear reactor

    International Nuclear Information System (INIS)

    Luetzow, H.J.

    1981-01-01

    A thimble tube construction is described which will provide a controlled descent for a control rod while minimizing the reaction forces which must be absorbed by the thimble tube and reducing the possibility that a foreign particle could interfere with the free descent of a control rod. A thimble tube is formed with helically-corrugate internal walls which cooperate with a control rod contained in the tube in an emergency situation to provide a progressively-increasing hydraulic restraining force as each adjacent corrugation is encountered

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

    International Nuclear Information System (INIS)

    2009-03-01

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

  2. The second eddy current testing of zircaloy tube samples from the OECD Halden reactor project at Reactor Fuel Examination Facility, Tokai, JAERI

    International Nuclear Information System (INIS)

    Ohwada, Isao; Nishino, Yasuharu

    1986-07-01

    The Reactor Fuel Examination Facility in Tokai/JAERI (Japan Atomic Energy Research Institute) joined to the second round robin programme on eddy current test of the Halden/IFE. In the programme, two zircaloy tube samples with some artificial defects were provided for measurements. To clarify the locations in axial and azimuthal directions, types and dimensions of the provided artificial defects, measured signals from eddy current test were analysed in comparison with the known defects on the calibration tube. As a result, fourteen defects were determined from the measurements. Then, the location, the type and the relative dimension of them were also revealed. The results of those eddy current test are described in this paper. (author)

  3. A new method to evaluate the sealing reliability of the flanged connections for Molten Salt Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Li, Qiming, E-mail: liqiming@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Tian, Jian; Zhou, Chong [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Wang, Naxiu, E-mail: wangnaxiu@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China)

    2015-06-15

    Highlights: • We novelly valuate the sealing reliability of the flanged connections for MSRs. • We focus on the passive decrease of the leak impetus in flanged connections. • The modified flanged connections are acquired a sealing ability of self-adjustment. • Effects of redesigned flange configurations on molten salt leakage are discussed. - Abstract: The Thorium based Molten Salt Reactor (TMSR) project is a future Generation IV nuclear reactor system proposed by the Chinese Academy of Sciences with the strategic goal of meeting the growing energy needs in the Chinese economic development and social progress. It is based on liquid salts served as both fuel and primary coolant and consequently great challenges are brought into the sealing of the flanged connections. In this study, an improved prototype flange assembly is performed on the strength of the Freeze-Flange initially developed by Oak Ridge National Laboratory (ORNL). The calculation results of the finite element model established to analyze the temperature profile of the Freeze-Flange agree well with the experimental data, which indicates that the numerical simulation method is credible. For further consideration, the ideal-gas thermodynamic model, together with the mathematical approximation, is novelly borrowed to theoretically evaluate the sealing performance of the modified Freeze-Flange and the traditional double gaskets bolted flange joint. This study focuses on the passive decrease of the leak driving force due to multiple gaskets introduced in flanged connections for MSR. The effects of the redesigned flange configuration on molten salt leakage resistance are discussed in detail.

  4. A new method to evaluate the sealing reliability of the flanged connections for Molten Salt Reactors

    International Nuclear Information System (INIS)

    Li, Qiming; Tian, Jian; Zhou, Chong; Wang, Naxiu

    2015-01-01

    Highlights: • We novelly valuate the sealing reliability of the flanged connections for MSRs. • We focus on the passive decrease of the leak impetus in flanged connections. • The modified flanged connections are acquired a sealing ability of self-adjustment. • Effects of redesigned flange configurations on molten salt leakage are discussed. - Abstract: The Thorium based Molten Salt Reactor (TMSR) project is a future Generation IV nuclear reactor system proposed by the Chinese Academy of Sciences with the strategic goal of meeting the growing energy needs in the Chinese economic development and social progress. It is based on liquid salts served as both fuel and primary coolant and consequently great challenges are brought into the sealing of the flanged connections. In this study, an improved prototype flange assembly is performed on the strength of the Freeze-Flange initially developed by Oak Ridge National Laboratory (ORNL). The calculation results of the finite element model established to analyze the temperature profile of the Freeze-Flange agree well with the experimental data, which indicates that the numerical simulation method is credible. For further consideration, the ideal-gas thermodynamic model, together with the mathematical approximation, is novelly borrowed to theoretically evaluate the sealing performance of the modified Freeze-Flange and the traditional double gaskets bolted flange joint. This study focuses on the passive decrease of the leak driving force due to multiple gaskets introduced in flanged connections for MSR. The effects of the redesigned flange configuration on molten salt leakage resistance are discussed in detail

  5. Application of automatic inspection system to nondestructive test of heat transfer tubes of primary pressurized water cooler in the high temperature engineering test reactor. Joint research

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Furusawa, Takayuki

    2001-07-01

    Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of the PPWC, were different from those obtained for the artificially defected tube. In UT echo amplitude of the PPWC tubes inspected was lower than 20% of distance-amplitude calibration curve. Thus, it was confirmed that there was no defect in depth, which was more than the detecting standard of the probes, on the outer surface of the heat transfer tubes of the PPWC inspected. (author)

  6. A flow-through amperometric sensor based on dialysis tubing and free enzyme reactors

    NARCIS (Netherlands)

    Bohm, S.; Pijanowska, D.G.; Pijanowska, D.; Olthuis, Wouter; Bergveld, Piet

    2001-01-01

    A generic flow-through amperometric microenzyme sensor is described, which is based on semi-permeable dialysis tubing carrying the sample to be analyzed. This tubing (300 μm OD) is led through a small cavity, containing the working and reference electrode. By filling this cavity with a few μl of an

  7. Proposals for investigating instrument tube line breaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Charlton, T.R.; Loomis, G.G.; Hall, D.G.; Cozzuol, J.M.

    1985-11-01

    Questions posed by the NRC pertaining to instrument tube critical flow and applicability of the Semiscale experimental facility are evaluated. A program is recommended to investigate the issue of generic PWR safety following hypothetical rupture of instrument tubes due to consequences of seismic events

  8. Fabrication of a pressurized water reactor fuel element prototype with Zy-control rod guide tubes

    International Nuclear Information System (INIS)

    Bezold, H.; Romeiser, H.J.

    1978-10-01

    A prototype fuel assembly with zircaloy guide was fabricated by mass production methods. The fastening of the Inconel spacer grids to the guide tubes and the transition joint for fixing the tubes to the stainless stell upper end-fitting of the assembly were investigated. Tools and welding devices were developed for the construction of the skeleton. (orig.) [de

  9. Corrosion study of heat exchanger tubes in pressurized water cooled nuclear reactors by conversion electron Moessbauer spectroscopy

    International Nuclear Information System (INIS)

    Homonnay, Z.; Kuzmann, E.; Varga, K.; Nemeth, Z.; Szabo, A.; Rado, K.; Schunk, J.; Tilky, P.; Patek, G.

    2005-01-01

    Nuclear energy production tends to return into the focus of interest because of the constantly increasing energy need of the world and the green house effect problems of the strongest competitor oil or gas based power plants. In addition to the construction of new nuclear power plants, lifetime extension of the existing ones is the most cost effective investment in the energy business. However, feasibility and safety issues become very important at this point, and corrosion of the construction materials should be carefully investigated before decision on a potential lifetime extension of a reactor. 57 Fe-Conversion Electron Moessbauer Spectroscopy (CEMS) is a sensitive tool to analyze the phase composition of corrosion products on the surface of stainless steel. The upper ∼300 nm can be investigated due to the penetration range of conversion electrons. The corrosion state of heat exchanger tubes from the four reactor units of the Paks Nuclear Power Plant, Hungary, were analyzed by several methods including CEMS. The primary circuit side of the tubes was studied on selected samples cut out from the heat exchangers during regular maintenance. Cr- and Ni-substituted magnetite, sometimes hematite, amorphous Fe-oxides/oxyhydroxides as well as the signal of bulk austenitic steel of the tubes were detected. The level of Cr- and Ni-substitution in the magnetite phase could be estimated from the Moessbauer spectra. Correlation between earlier decontamination cycles and the corrosion state of the heat exchangers was sought. In combination with other methods, a hybrid structure of the surface oxide layer of several microns was established. It is suggested that previous AP-CITROX decontamination cycles can be responsible for this structure which makes the oxide layer mobile. This mobility may be responsible for unwanted corrosion product transport into the reactor vessel by the primary coolant.

  10. Leak-before-break concept for evaluation of flows detected in pressure tubes in a Candu type reactor

    International Nuclear Information System (INIS)

    Crespi, J.C.

    1992-01-01

    This paper reviews the role of the Leak-Before-Break concept for evaluation of flaws detected in cold-worked Zr 2.5% Nb pressure tubes in a CANDU type reactors. The acceptance criteria are intended to prevent failure by brittle fracture, plastic collapse of the ligament and delayed hydride cracking. The methodology developed here was of great help in order to establish the operative conditions for fuel channel garter springs repositioning by means of the SLA Rette tool at Embalse Nuclear Generating Station, Cordoba, Argentina. (author)

  11. Spring/dimple instrument tube restraint

    International Nuclear Information System (INIS)

    DeMario, E.E.; Lawson, C.N.

    1993-01-01

    A nuclear fuel assembly for a pressurized water nuclear reactor has a spring and dimple structure formed in a non-radioactive insert tube placed in the top of a sensor receiving instrumentation tube thimble disposed in the fuel assembly and attached at a top nozzle, a bottom nozzle, and intermediate grids. The instrumentation tube thimble is open at the top, where the sensor or its connection extends through the cooling water for coupling to a sensor signal processor. The spring and dimple insert tube is mounted within the instrumentation tube thimble and extends downwardly adjacent the top. The springs and dimples restrain the sensor and its connections against lateral displacement causing impact with the instrumentation tube thimble due to the strong axial flow of cooling water. The instrumentation tube has a stainless steel outer sleeve and a zirconium alloy inner sleeve below the insert tube adjacent the top. The insert tube is relatively non-radioactivated inconel alloy. The opposed springs and dimples are formed on diametrically opposite inner walls of the insert tube, the springs being formed as spaced axial cuts in the insert tube, with a web of the insert tube between the cuts bowed radially inwardly for forming the spring, and the dimples being formed as radially inward protrusions opposed to the springs. 7 figures

  12. Simulated nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Berta, V.T.

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end

  13. Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

    1982-01-01

    Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

  14. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  15. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    Energy Technology Data Exchange (ETDEWEB)

    Zamani, M. [National Radiation Protection Department - NRPD, Atomic Energy Organization of Iran - AEOI, Tehran (Iran, Islamic Republic of); End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Kasesaz, Y.; Khalafi, H.; Shayesteh, M. [Radiation Application School, Nuclear Science and Technology Research Institute, AEOI, Tehran (Iran, Islamic Republic of)

    2015-07-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  16. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    International Nuclear Information System (INIS)

    Zamani, M.; Kasesaz, Y.; Khalafi, H.; Shayesteh, M.

    2015-01-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  17. Pressure-tube reactors as a part of Russian nuclear fleet

    International Nuclear Information System (INIS)

    Gmyrko, V.E.; Grozdov, I.I.; Nikitin, Yu.M.; Petrov, A.A.; Potapov, A.A.; Finyakin, A.F.

    2007-01-01

    The place and role of channel reactors in nuclear power in our country and the main measures for upgrading and improving the power generating units of nuclear power plants with RBMK reactors are described. It is shown that the risk indicators for serious damage to the core of power generating units with RBMK reactors are lower after upgrading and the corresponding IAEA criterion established for operating nuclear power plants. Upgrading and implementation of a service life extension program has made it possible to obtain licenses for continuing operation of power generating units with first-generation RBMK reactors and predicting a service life increase to 45 years. The characteristics of nuclear power plants with channel reactors with more highly developed internal and natural safety properties are shown in evolutionary designs of the power generating units MCER-860,-1000, and-1500, which have protective shells and which meet all requirements for power generating units built today. It is shown that innovative solutions for the channel reactor concept can be implemented on the basis of the designs of power generating units with nuclear superheating of steam or on the basis of designs for developing reactors with supercritical parameters [ru

  18. Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

    Energy Technology Data Exchange (ETDEWEB)

    Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  20. In reactor measurements, modeling and assessments to predict liquid injection shutdown system nozzle to Calandria tube time to contact

    International Nuclear Information System (INIS)

    Kirstein, K.; Kalenchuk, D.

    2011-01-01

    Over the past few years there has been an expanding effort to assess the potential for Calandria Tubes (CTs) coming into contact with Liquid Injection Shutdown System (LISS) Nozzles to ensure continued contact-free operation as required by CSA N285.4. LISS Nozzles (LINs), which run perpendicular to and between rows of fuel channels, sag at a slower rate than the fuel channels. As a result certain LINs may come in contact with CTs above them. The CT/LIN gaps can be predicted from calculated CT sag, LIN sag and a number of component and installation tolerances. This method however results in very conservative predictions when compared to measurements, confirmed with the in reactor measurements initiated in 2000, when gaps were successfully measured the first time using images obtained from a camera-assisted measurement tool inserted into the calandria. To reduce the conservatism of the CT/LIN gap predictions, statistical CT/LIN gap models are used instead. They are derived from a comparison between calculated gaps based on nominal dimensions and the visual image based measured gaps. These reactor specific (typically 95% confidence level) CT/LIN gap models account for all uncertainties and deviations from nominal values. Prediction error margins reduce as more in-reactor gap measurements become available. Each year more measurements are being made using this standardized visual CT/LIN proximity method. The subsequently prepared reactor-specific models have been used to provide time to contact for every channel above the LINs at these stations. In a number of cases it has been used to demonstrate that the reactor can be operated to its end of life before refurbishment with no predicted contact, or specific at-risk channels have been identified for which appropriate remedial actions could be implemented in a planned manner. (author)

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

    International Nuclear Information System (INIS)

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

    1980-10-01

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

  2. Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

    International Nuclear Information System (INIS)

    Jiang, J.T.; Luxat, J.C.

    2008-01-01

    Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

  3. Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J.T.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)

    2008-07-01

    Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

  4. Ultrasonic testing of canning tubes in stainless steel of the EL 4 reactor

    International Nuclear Information System (INIS)

    Prot, A.; Monnier, P.

    1964-01-01

    From all the methods possible for controlling thin cans the one chosen, for numerous reasons, vas that making use of ultrasonic techniques. A method has been developed which should make it possible to carry out a rapid and efficient industrial control of canning tubes, The reasons for the choice of the ultrasonic method are given in detail, together with the principles of the method and the actual control parameters. In the present state of our research, it should be possible to control at least 50 000 tubes a year. Improvements brought about in the details of the control technique itself should make it possible to increase this rate considerably. (authors) [fr

  5. Testing external surface of fuel element tubes for power nuclear reactors

    International Nuclear Information System (INIS)

    Naugol'nykh, O.G.; Nelyubin, Yu.V.

    1987-01-01

    Optical methods are regarded perspective for discovery and detection of flaws of external surfaces of fuel element tubes. The TV method has highest information content among them. Two mock-ups of facilities based on the TV method using a ''dissector'' type TV device and a TV tube with charge accumulation (vidikon) have been developed. It is concluded that complex testing - combination of ultrasonic, photoelectric and TV methods in a facility is necessary for discovery and analysis of the whole variety of flaws, though sensitivity of the TV method is enough for disclosure of all the main defects

  6. Realistic analysis of steam generator tube rupture accident in Angra-1 reactor

    International Nuclear Information System (INIS)

    Fontes, S.W.F.

    1989-01-01

    This paper presents the analysis of different scenarios for a Steam Generator Tube Rupture accident (SGTR) in Angra-1 NPP. The results and conclusions will be used as support in the preparation of the emergency situation programs for the plant. For the analysis a SGTR simulation was performed with RETRAN-02 code. The results indicated that the core integrity and the plant itself will not affect by small ruptures in SG tubes. For large ruptures the analysis demonstrated that the accident may have harmful consequences if the operator do not actuate effectively since the initial moments of the accidents. (author) [pt

  7. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  8. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  9. In-pile Creep Tests of Zircaloy Tubing in the Studsvik R2 Reactor. Final Report

    International Nuclear Information System (INIS)

    Tomani, Hans; Lindeloew, Ulf

    2000-12-01

    In this report are presented the findings of a prototype creep test on Zr4 guide tube specimens exposed in-pile and out-of-pile and stressed by constant bending moments. The calculated initial deflection curvature caused by the applied bending moment agrees very well with the measured initial values. Furthermore, the measurement results show excellent consistency. The dominant impact of neutron irradiation is clearly demonstrated. After 3 cycles (∼1300 hours) the irradiation creep is 4 times as large as the thermal creep. This is the case at least when fresh tube material is used. Irradiation creep progresses steadily, but the creep rate is not quite constant during the 3 irradiation cycles. The thermal creep, on the other hand, quickly saturates and there is hardly any further deflection after the second cycle for the specimen situated above the core. A limitation with the rig has been that the tube deflection became limited by the rig carrier body of the rig in the neutron flux (core) that disqualified the results of a fourth irradiation cycle actually performed in the fall of 1998. The test method appears to be suitable for testing the bending creep of different guide tube materials or designs under PWR conditions

  10. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Additional information

    International Nuclear Information System (INIS)

    2009-03-01

    The reports from Argentina, Canada, India, Korea and Romania are presented concerning the projects carried out under the Coordinated Research Program (CRP) I3.30.10 of the International Agency for Atomic Energy - Vienna related to 'Intercomparison of Techniques for Pressure Tube Inspection and Diagnostics'

  11. Apparatus for connecting an element attached to a metallic wire to an interior tube in a core drill

    Energy Technology Data Exchange (ETDEWEB)

    1970-04-17

    A locking device consists of at least one lever pivotable around an axis transversal to the axis of the core drill and mounted on the upper end of the core tube. A first arm oriented upward is intended to pivot during lowering of the wire line and to engage on the core tube, so that the core tube can be lifted by the wire line. A second arm prevents lifting of the core tube during drilling. The second arm also releases the core tube, after the wire line has been firmly attached and locked onto the core tube.

  12. Investigation of in-plane moment connections of I-beams to square concrete-filled steel tube columns under gravity loads

    Directory of Open Access Journals (Sweden)

    Abdelrahim K. Dessouki

    2015-04-01

    Full Text Available This paper focuses on experimental and analytical behavior of the ultimate moment of the connections of steel I-beams to square concrete-filled steel tube columns. External stiffeners around the columns are used at the beam flange levels. Five specimens are tested monotonically. The test parameters are the column stiffener dimensions and filling the steel tube column with concrete. Two types of failure modes are observed; beam flange failure and stiffener failure. The experimental results show that the ultimate moment of the connection is increased by increasing stiffener’s dimensions and filling the steel tube column with concrete. ANSYS finite element program is used to simulate the behavior, taking into account both geometric and material nonlinearities. Analytical results that are in fair agreement with the experimental ones are then used to discuss the influence of the main geometric parameters on the connection behavior. The parameters are the stiffener and column dimensions as well as filling the steel tube column with concrete. Different square column cross sections are chosen to cover the three classes of section classifications according to Egyptian code of practice, which are: compact, non compact or slender. The increase in the ultimate moment of the connections is based upon both column cross sections’ compactness and stiffener dimensions while the maximum advantages occur with slender columns.

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

    International Nuclear Information System (INIS)

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

    1965-11-01

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

  14. Cooling system for the connecting rings of a fast neutron reactor vessel

    International Nuclear Information System (INIS)

    Martin, J.-P.; Malaval, Claude

    1974-01-01

    A description is given of a cooling system for the vessel connecting rings of a fast neutron nuclear reactor, particularly of a main vessel containing the core of the reactor and a volume of liquid metal coolant at high temperature and a safety vessel around the main vessel, both vessels being suspended to a rigid upper slab kept at a lower temperature. It is mounted in the annular space between the two vessels and includes a neutral gas circuit set up between the wall of the main vessel to be cooled and that of the safety vessel itself cooled from outer. The neutral gas system comprises a plurality of ventilators fitted in holes made through the thickness of the upper slab and opening on to the space between the two vessels. It also includes two envelopes lining the walls of these vessels, establishing with them small section channels for the circulation of the neutral gas cooled against the safety vessel and heated against the main vessel [fr

  15. Distance alimentation device for an automatic machine such a plugger for nuclear reactor steam generators tubes

    International Nuclear Information System (INIS)

    Cartry, J.P.; Schlaudecker, D.

    1987-01-01

    This distance alimentation device for a plugger is made by a plug feeder connected to a source of compressed air, a flexible hose connecting the feeder to a mobile tool mounted on a support and a rotating arm with a housing and presenting the plug to the tool [fr

  16. Underclad crack development of steam generators tube sheets and reactor vessels nozzles in PWR plants

    International Nuclear Information System (INIS)

    Faure, F.; Bocquet, P.; Boudot, R.; Zacharie, G.

    1985-01-01

    Defects formed, before stress relieving treatment, under the coating of tube plates of steam generators and vessel pipes are cold cracks formed in the segregation zone during surface coating without pre- and postheating of the 2nd layers and eventually of the following coating layers. To solve this problem, the conditions of pre- and post-heating are reinforced and applied to all the coating layers. 13 refs [fr

  17. Fabrication and inspection of stainless-steel-clad tubes for fast reactors

    International Nuclear Information System (INIS)

    Spriet, M.

    1975-01-01

    The production of cladding tubes requires a selection of the raw material, particular core taken during the cold and hot processes, special surface preparations, heat treatments, and intermediate control during the principal steps of fabrication. The inspection is made in two stages: acceptance tests at Vallourec (Eddy current and ultrasonic tests, metrology of internal and external diameter and thickness, metallography, analyses, tensile tests) and ultrasonic tests, metrology of external diameter and thickness, metallography, analyses, mechanical tests at high temperature) [fr

  18. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1992-03-01

    A survey of steam generator operating experience for 1986 has been carried out for 184 pressurized water and pressurized heavy-water reactors, and 1 water-cooled, graphite-moderated reactor. Tubes were plugged at 75 of the reactors (40.5%). In 1986, 3737 tubes were plugged (0.14% of those in service) and 3148 tubes were repaired by sleeving. A small number of reactors accounted for the bulk of the plugged tubes, a phenomenon consistent with previous years. For 1986, the available tubesheet sludge data for 38 reactors has been compiled into tabular form, and sludge/deposit data will be incorporated into all future surveys

  19. Derivation of Elastic Stress Concentration Factor Equations for Debris Fretting Flaws in Pressure Tubes of Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Jong Sung; Oh, Young Jin

    2014-01-01

    If volumetric flaws such as bearing pad fretting flaws and debris fretting flaws are detected in the pressure tubes of pressurized heavy water reactors during in-service inspection, the initiation of fatigue cracks and delayed hydrogen cracking from the detected volumetric flaws shall be assessed by using elastic stress concentration factors in accordance with CSA N285.8-05. The CSA N285.8-05 presents only an approximate formula based on linear elastic fracture mechanics for the debris fretting flaw. In this study, an engineering formula considering the geometric characteristics of the debris fretting flaw in detail was derived using two-dimensional finite element analysis and Kinectrics, Inc.'s engineering procedure with slight modifications. Comparing the application results obtained using the derived formula with the three-dimensional finite element analysis results, it is found that the results obtained using the derived formula agree well with the results of the finite element analysis

  20. Denting of Inconel 600 steam generator tubes in pressurized water reactors

    International Nuclear Information System (INIS)

    Van Rooyen, D.; Weeks, J.R.

    1976-10-01

    Rapid, localized corrosion of carbon steel tube support plates (TSP) has led to cases of denting of steam generator tubes, due to the pressure of corrosion products formed in crevices between the tubes and TSP holes. The corrosion product is mainly magnetite (Fe 3 O 4 ), formed in ''run-away'' fashion as a result of local chemistry changes when an extended operation with phosphate (PO 4 ) treatment of the secondary coolant is followed by an all volatile treatment (AVT). The rate of the ''run-away'' magnetite formation, and therefore, the extent of damage will probably vary with the amounts of the harmful chemicals present and with temperature. Leaky condensers are felt to be responsible for the presence of Cl - ions, and for the observation that denting is more extensive in plants with salt water cooled condensers. It is possible that thermal cycles assist the denting process, both by mechanical and chemical ratchetting mechanisms. Recommendations are presented concerning the continued operation of plants with observed denting

  1. Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor. Influence of flow collector of backup CR guide tube

    International Nuclear Information System (INIS)

    Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki

    2016-01-01

    Design study of an advanced large-scale sodium-cooled fast reactor (SFR) has been conducted in JAEA. In the region between the bottom of the Upper Internal Structure (UIS) and the core outlet, the hot sodium from the fuel subassembly mixes with the cold sodium from the neighbor control rod (CR) channel. Therefore, temperature fluctuation due to mixing fluids at different temperatures may cause high cycle thermal fatigue at the bottom of the UIS. In the advanced design, installation of a flow guide structure named Flow-Collector (FC) to the backup control rod (BCR) guide tube is considered to enhance reliable operation of self-actuated shutdown system (SASS) and to ensure reactor shutdown operation. Previously, water experiments without the FC model had been examined in JAEA to investigate effective countermeasures to the significant temperature fluctuation generation at the bottom of the UIS. Since the FC may affect the thermal mixing behavior at the bottom of the UIS, influence of the FC on characteristics of the temperature fluctuation around the BCR channels was investigated using a water experimental facility with structure model of the FC. Through the experiment, small influence of the FC on the temperature fluctuation distribution at the bottom of the UIS was indicated. (author)

  2. Fast reactor steam generators with sodium on the tube side. Design and operational parameters

    International Nuclear Information System (INIS)

    1994-01-01

    A comparison of design and operational characteristics as well as analysis of experience gained during the long terms operation of the Micro Module Inverse Steam Generator and Module Inverse Steam Generator at BOR 60 reactor are main aims of this technical report. 20 refs, 47 figs, 14 tabs

  3. Problems of evaluation of nuclear reactor active zone tubes during pre-irradiation tests

    International Nuclear Information System (INIS)

    Lezinskaya, E.Ya.; Buryak, T.N.

    2004-01-01

    An analysis of standard methods of graine size estimation of basic indexes of austenitic steel and alloys of active area of atomic reactors. It is shown insolvency of standard methods of grain size estimation in the real wares. The suggested method of computer simulation of structures of pipes-shells raped for working aut of modes of heat treatment

  4. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  5. Effects of Relative SG Tube Pitches on the Performance Characteristics of a Small Modular Reactor driven by Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Youngjin; Yi, Kunwoo; Lee, Byungjin [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)

    2016-10-15

    In this research, the capacity and basic dimensions for SMRs driven by a natural circulation are preliminarily assumed to determine the SMR configuration for the conceptual design, and each of the pre-set values is explained below. Firstly, the PZR configuration is not considered because it is not included to the main flow of the primary coolant. One of the SMR requirements is that SMR shall carry on the road. Hence, the vehicle geometrical limits are 15 meters for the length, and 3.5 meters for the height, approximately. With these limits for the dimensions of the SMR, RV length is assumed about 13.8 meters and RV diameter about 2.5 meters. In IAEA definition for SMRs, the capacity of electric power is no more than 300 MWe. If the efficiency of SMR power plant is assumed to 33% compared to the commercial power plant, the core power is below 1,000 MWth. In this research, the core power is assumed to 200 MWth arbitrarily during normal operation. The primary coolant passes through the outside of tubes, and the heat is transfer to the secondary feedwater. The secondary feedwater passes through the inside of tubes, and the heat from the primary coolant is received to generate the superheated steam. The present work carries out numerical simulations to get an insight for the effects of the diameters of the reactor vessel and riser using the parameters such as the steam generator tube pitches. To sum up, the calculation results show a good agreement with the theoretical equation and the uniform diameter loop has a more uniform temperature distribution and larger mass flow rate.

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

    International Nuclear Information System (INIS)

    2006-05-01

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

  7. Statistical analysis of failure time in stress corrosion cracking of fuel tube in light water reactor

    International Nuclear Information System (INIS)

    Hirao, Keiichi; Yamane, Toshimi; Minamino, Yoritoshi

    1991-01-01

    This report is to show how the life due to stress corrosion cracking breakdown of fuel cladding tubes is evaluated by applying the statistical techniques to that examined by a few testing methods. The statistical distribution of the limiting values of constant load stress corrosion cracking life, the statistical analysis by making the probabilistic interpretation of constant load stress corrosion cracking life, and the statistical analysis of stress corrosion cracking life by the slow strain rate test (SSRT) method are described. (K.I.)

  8. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    Subbotin, V.; Alexeev, G.; Peskov, O.; Sapankevic, A.

    1976-01-01

    The conditions are formulated under which the results of the experimental research of the boilino. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented. (F.M.)

  9. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, V; Alexeev, G; Peskov, O; Sapankevic, A

    1976-08-01

    The conditions are formulated under which the results of the experimental research of the boiling. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented.

  10. Feasibility study for LEU conversion of the WWR-K reactor at the Institute of Nuclear Physics in Kazakhstan using a 5-tube fuel assembly

    International Nuclear Information System (INIS)

    Hanan, N.A.; Liaw, J.R.; Matos, J.E.

    2005-01-01

    A feasibility study by the RERTR program for possible LEU conversion of the 6 MW WWR-K reactor concludes that conversion is feasible using an LEU 5-tube Russian fuel assembly design. This 5-tube design is one of several LEU fuel assembly designs being studied (Ref. 1) for possible use in this reactor. The 5-tube assembly contains 200 g 235 U with an enrichment of 19.7% in four cylindrical inner tubes and an outer hexagonal tube with the same external dimensions as the current HEU (36%) 5-tube fuel assembly, which contains 112.5 g 235 U. The fuel meat material, LEU UO 2 -Al dispersion fuel with ∼ 2.5 g U/cm 3 , has been extensively irradiation tested in a number of reactors with uranium enrichments of 36% and 19.7%. Since the 235 U loading of the LEU assemblies is much larger than the HEU assemblies, a smaller LEU core with five rows of fuel assemblies is possible (instead of six rows of fuel assemblies in the HEU core). This smaller LEU core would consume about 60% as many fuel assemblies per year as the current HEU core and provide thermal neutron fluxes in the inner irradiation channels that are ∼ 17% larger than with the present HEU core. The current 21 day cycle length would be maintained and the average discharge burnup would be ∼ 42%. Neutron fluxes in the five outer irradiation channels would be smaller in the LEU core unless these channels can be moved closer to the LEU fuel assemblies. Results show that the smaller LEU core would meet the reactor's shutdown margin requirements and would have an adequate thermal-hydraulic safety margin to onset of nucleate boiling. (author)

  11. Development of a running robot in super high speed tube. Aiming at realization of in-tube inspection for primary cooler and so forth of nuclear reactor

    International Nuclear Information System (INIS)

    Kato, Shigeo

    2000-01-01

    Authors have carried out a study on an in-tube running robot in living body on a base of laying stretching of bellows at a means of running by thinking application of in-tube inspection in living body such as large and small bowels. As a result, an in-tube running robot with about 20 mm in inner diameter capable of running in soft small bowel as well as in hard running tube was developed successfully. After an accident of the Tsuruga nuclear power plant, inspection of a large diameter tube with 76 mm in inner diameter was found to be much important, to begin development of an in-tube running robot for 50 mm class diameter tube. As a result, an in-tube running robot capable of enough holding a micro video camera with about 20 g in mass and showing 4.6 N in tension at more than ten times higher speed of 248 mm/s in no loading state, could be made in trial. Here was reported on a foothold realizable on an in-tube running robot for the 76 mm class large diameter tube to be investigated in future. (G.K.)

  12. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  13. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2008-01-01

    Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

  14. Reactor fuel cladding tube with excellent corrosion resistance and method of manufacturing the same

    International Nuclear Information System (INIS)

    Okuda, Takanari; Kanehara, Mitsuo; Abe, Katsuhiro; Nishimura, Takashi.

    1995-01-01

    The present invention provides a fuel cladding tube having an excellent corrosion resistance and thus a long life, and a suitable manufacturing method therefor. Namely, in the fuel cladding tube, the outer circumference of an inner layer made of a zirconium base alloy is coated with an outer layer made of a metal more corrosion resistant than the zirconium base alloy. Ti or a titanium alloy is suitable for the corrosion resistant metal. In addition, the outer layer can be coated by a method such as vapor deposition or plating, not limited to joining of the inner layer material and the outer layer material. Specifically, a composite material having an inner layer made of a zirconium alloy coated by the outer material made of a titanium alloy is applied with hot fabrication at a temperature within a range of from 500 to 850degC and at a fabrication rate of not less than 5%. The fabrication method includes any of extrusion, rolling, drawing, and casting. As the titanium-base alloy, a Ti-Al alloy or a Ti-Nb alloy containing Al of not more than 20wt%, or Nb of not more than 20wt% is preferred. (I.S.)

  15. Denting of Inconel steam generator tubes in pressurized water reactors. Third informal report

    International Nuclear Information System (INIS)

    van Rooyen, D.; Weeks, J.R.

    1977-08-01

    The recent plant operating experience and laboratory test results on the phenomenon of denting in recirculating PWR steam generators is reviewed. Although denting was first reported only in plants that were converted from phosphate to AVT, it has now also been observed in plants still on phosphate, as well as in some that started on AVT. In some units, slightly abnormal eddy current signals have been observed at the top of the tube sheets. The degree of denting in operating steam generators may be related to the levels and duration of chloride inleakage. Chloride, however, is not the only active ingredient, and does not seem to give denting until local acid conditions arise; consequently, it may be necessary for soluble copper and/or nickel ions to be present to promote the denting reaction. Chloride concentrations in actively corroding crevices can increase by several orders of magnitude over the bulk coolant. It is thus difficult to develop a basis for Cl - specifications for secondary water. Maintaining Cl - low enough to prevent denting may be unmanageable without full flow condensate demineralization in coastal plants with copper alloy condensors and feedwater lines. Cathodic depolarization by oxidizing species are thought to promote the formation of acid chlorides in crevices and trigger the denting reactions; some ions may also catalyze the rapid formation of magnetite. These, and other mechanistic aspects of denting are discussed. The implications of the Inconel 600 tube defects at Ginna in non-dented areas, originating from the primary side, are also discussed

  16. Measurement of mechanical properties of a reactor operated Zr–2.5Nb pressure tube using an in situ cyclic ball indentation system

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

    Highlights: • Measurement of mechanical properties of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situProperty Measurement System has been designed in house. • The tool head is capable to carry out in situ ball indentation trials inside pressure tube. • The paper describes the theory and results of the trials conducted on irradiated pressure tube. - Abstract: Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situProperty Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated

  17. Measurement of mechanical properties of a reactor operated Zr–2.5Nb pressure tube using an in situ cyclic ball indentation system

    International Nuclear Information System (INIS)

    Chatterjee, S.; Panwar, Sanjay; Madhusoodanan, K.

    2015-01-01

    Highlights: • Measurement of mechanical properties of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situProperty Measurement System has been designed in house. • The tool head is capable to carry out in situ ball indentation trials inside pressure tube. • The paper describes the theory and results of the trials conducted on irradiated pressure tube. - Abstract: Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situProperty Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated

  18. Flow-Tube Reactor Experiments on the High Temperature Oxidation of Carbon Weaves

    Science.gov (United States)

    Panerai, Francesco; White, Jason D.; Robertson, Robert; Borner, Arnaud; Ferguson, Joseph C.; Mansour, Nagi N.

    2017-01-01

    Under entry conditions carbon weaves used in thermal protection systems (TPS) decompose via oxidation. Modeling this phenomenon is challenging due to the different regimes encountered along a flight trajectory. Approaches using equilibrium chemistry may lead to over-estimated mass loss and recession at certain conditions. Concurrently, there is a shortcoming of experimental data on carbon weaves to enable development of improved models. In this work, a flow-tube test facility was used to measure the oxidation of carbon weaves at temperatures up to 1500 K. The material tested was the 3D carbon weave used for the heat shield of the NASA Adaptive Deployable Entry and Placement Technology, ADEPT. Oxidation was characterized by quantifying decomposition gases (CO and CO2), by mass measurements, and by microscale surface analysis. The current set of measurements contributes to the development of finite rate chemistry models for carbon fabrics used in woven TPS materials.

  19. Numerical study on pressure drop and heat transfer for designing sodium-to-air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kang, Hie-Chan; Eoh, Jae-Hyuk; Cha, Jae-Eun; Kim, Seong-O.

    2013-01-01

    Highlights: ► Numerical simulation for the heat flow characteristic of the sodium-to-air heat exchanger (AHX) and tube banks. ► Parallelogram tube banks showed almost similar thermal and hydraulic characteristics to the rectangular tube banks. ► Pressure drop and heat transfer of the staggered and rectangular tube banks compared with Zhukauskas’ correlation. ► AHX was modeled as porous media and suggested design guide to enhance the performance. - Abstract: A numerical study is performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX are modeled as porous media and simulated heat and momentum transfer by a commercial program. Two-dimensional flow characteristic appears differently at the inlet region of the AHX annulus, and the required length of the inlet region is shorter for an inlet having a 45 degree chamber or a round shape than for one with a perpendicular corner. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX are evaluated and discussed. Pressure drop and heat transfer shows similar trends and underestimated values, respectively, when compared with Zhukauskas empirical correlations. The parallelogram tube bank shows similar results to the rectangular arrangement.

  20. A flow test for calibrating 177 core tubes of 1/5-scale reactor flow model for Yonggwang nuclear units 3 and 4

    International Nuclear Information System (INIS)

    Lee, Byung Jin; Jang, Ho Cheol; Cheong, Jong Sik; Kuh, Jung Eui

    1990-01-01

    A flow test was performed to find out the hydraulic characteristics of every one of 177 core tubes, representing a fuel assembly respectively, as a preparatory step of 1/5 scale reactor flow model test for Yonggwang Nuclear Units (hereafter YGN) 3 and 4. The axial hydraulic resistance of the fuel assembly was simulated in the square core tube with six orifice plates positioned along the tube length; core support structure below each fuel assembly was done in the core upstream geometry section of the test loop. For each core tube the pressure differentials across the inlet, exit orifice plate and overall tube length were measured, along with the flow rates and temperatures of the test fluid. The measured pressure drops were converted to pressure loss or flow metering coefficients. The metering coefficient of the inlet orifice plate was sensitive to the configuration and location of the upstream geometry. The hydraulic resistance of the core tubes were reasonably coincided with a target value and consistent. The polynomial curve fits of the calibrated coefficients for the 177 core tubes were obtained with reasonable data scatters

  1. High pressure thimble/guide tube seal fitting with built-in low pressure seal especially suitable for facilitated and more efficient nuclear reactor refueling service

    International Nuclear Information System (INIS)

    Bhatt, P.N.; Blaushield, R.M.

    1991-01-01

    This patent describes a HP/LP seal arrangement for an elongated guide tube and an elongated thimble disposed therein. The guide tube and thimble extending outwardly from the core of a nuclear reactor to a seal table where the guide tube is welded to the seal table to provide a high pressure seal relative thereto. It comprises: a tubular seal fitting disposed in alignment with the guide tube with the thimble extending therethrough on the low pressure side of the seal table; first high pressure sealing means coupling one end of the fitting to an end of the guide tube to prevent leakage from within the guide tube; inwardly facing thread means disposed adjacent the other and outer end of the seal fitting; a nut having an opening through which the thimble extends and further having outwardly facing threading in mating engagement with the fitting thread means; the fitting having a seal seat spaced longitudinally inwardly from the thread means and facing the fitting outer end and further disposed annularly about the inner surface of the fitting; deformable ring seal means; second releasable high pressure sealing means coupling the thimble to the outer end portion of the guide tube

  2. A ceramic breeder in a poloidal tube blanket for a tokamak reactor

    International Nuclear Information System (INIS)

    Amici, A.; Anzidei, L.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.; Zampaglione, V.; Petrizzi, L.

    1989-01-01

    A conceptual study of a helium-cooled solid breeder blanket for a tokamak reactor is presented. Tritium breeding capability together with system reliability are taken as the main design criteria. The blanket consists of tubular poloidal modules made of a central bundle of ceramic rods (γLiAlO 2 ) with a coaxial distribution of the inlet/outlet coolant flow (He) surrounded by a multiplier material (Be) in the form of bored bricks. The Be to γLiAlO 2 volume ratio is 4/1. The He inlet and outlet branches are cooling Be and γLiAlO 2 , respectively. A purge He flow running through small central holes of the ceramic rods is derived from the main flow. Under the typical conditions of a tokamak reactor (neutron wall load=2 MW/m 2 ), a full coverage tritium breeding ratio of 1.47 is achieved for the following design and operating parameters: outlet He temperature=570 0 C; inlet He temperature=250 0 ; total extracted power=2700 MW; He pumping power percentage=2%; minimum/maximum γLiAlO 2 temperature=400/900 0 C; maximum structural temperature=475 0 C; and maximum Be temperature=525 0 C. (orig.)

  3. Decommissioning of the research nuclear reactor IRT-M and problems connected with radioactive waste

    International Nuclear Information System (INIS)

    Abramidze, S.P.; Katamadze, N.M.; Kiknadze, G.G.; Saralidze, Z.K.

    2000-01-01

    The nuclear research reactor IRT-2000 is described, along with modifications and upgrades made over the past three decades. Considerations are outlined which followed a decision to shut-down the reactor and to dismantle it. (author)

  4. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor

    International Nuclear Information System (INIS)

    Hareux, F.; Roche, R.; Vrillon, B.

    1964-01-01

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [fr

  5. The study on water ingress mass in the steam generator heat-exchange tube rupture accident of modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Yan; Shi Lei; Li Fu; Zheng Yanhua

    2012-01-01

    The steam generator heat-exchange tube rupture (SGTR) accident is an important and particular accident which will result in water ingress to the primary loop of reactor. Water ingress will result in chemical reaction of graphite fuel and structure with water, which may cause overpressure due to generation of explosive gaseous in large quantity. The study on the water ingress accident is significant for the verification of the inherent characteristics of high temperature gas-cooled reactor. The previous research shows that the amount of water ingress mass is the dominant key factor on the severity of the accident consequence. The 200 MWe high temperature gas-cooled reactor (HTR-PM), which is the first modular pebble-bed high temperature gas-cooled reactor in China designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is selected to be analyzed in this paper. The different DBA accident scenarios of double-ended break of single heat-exchange tube are simulated respectively by the thermal-hydraulic analysis code RETRAN-02. The results show the water ingress mass through the broken heat-exchange tube is related to the break location. The amount of water ingress mass is affected obviously by the capacity of the emptier system. With the balance of safety and economical efficiency, the amount of water ingress mass from the secondary side of steam generator into the primary coolant loop will be reduced by increasing properly the diameter of the draining lines. (authors)

  6. Tool for cutting locking cups from guide tube mounting screws in a nuclear reactor

    International Nuclear Information System (INIS)

    Nee, J.D.; Hahn, J.J.

    1987-01-01

    This patent describes an apparatus for freeing a socket-head screw from a locking cup therefor in a reactor cavity, wherein the locking cup includes a fixed cylindrical side wall encircling the side surface of the screw head and an annular end wall overlying the outer end surface of the screw head. The apparatus consists of: frame means, cylindrical cutter means having a longitudinal axis and having a frustoconical cutting surface with an inner diameter less than the inner diameter of the locking cup side wall and with an outer diameter greater than the outer diameter of the locking cup side wall, and drive means carried by the frame means and coupled to the cutter means for effecting rotation thereof about the axis, the rotating cutter means are operable for severing the locking cup end wall from the locking cup side wall at the junction therebetween when the cutter means is moved against the locking cup substantially coaxially therewith

  7. Pressure drop and heat transfer in the sodium to air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kang, H.; Eoh, J.; Cha, J.; Kim, S.

    2011-01-01

    A numerical study was performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX were modeled as porous media and simulated heat and momentum transfer. Two-dimensional flow characteristic appeared at the most region of AHX annulus. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX were evaluated and compared with Zhukauskas empirical correlations. (author)

  8. Characterisation of Oxides Formed on the Internal Surface of Steam Generator Tubes in Alloy 690 Corroded in the Primary Environment of Pressurised Water Reactors

    International Nuclear Information System (INIS)

    Carrette, Florence; Leclercq, Stephanie; Legras, Laurent

    2012-09-01

    Since the end of the 1990s, EDF R and D has been studying the phenomenon of corrosion product release from Steam Generator tubes in order to minimize the Source Term of the contamination and radiation exposure during operation and maintenance of Pressurised Water Reactors. With the BOREAL loop, release tests in primary water at 325 deg. C were performed on various Steam Generator tubes made of alloy 690. The experimental conditions of these tests (chemistry, temperature and hydraulics) were the same for all the tests but the results showed various behaviours towards release. For some tubes, the release was weak whereas for others, it was higher; the release rate of the tubes decreased more or less quickly with time. In order to explain these results, the internal surface of the tubes was characterised before and after the tests. Before the tests, various parameters were studied; the main parameters were the roughness, the impurities, the grain size and the cold work. The results demonstrated that it was not easy to quantify the influence of each parameter on release and to differentiate the tubes. A new parameter was proposed to characterise the internal extreme surface of SG tubes: the surface nano-hardness by nano-indentation measurements. The tubes were also observed and analysed by SEM, (X)TEM. Data obtained by (X)TEM revealed differences of the surface state (layer of perturbed microstructure, density of dislocations, grain size, impurities, initial oxide,...). After the tests, the oxides formed on the internal surface and the underlying material of the samples were characterised by SEM, (X)TEM and SIMS. The examinations showed various types of oxides. For some tubes, a duplex oxide scale was identified, for the others, only one oxide scale was observed. For equivalent durations of corrosion, the thickness of the enriched - chromium oxide layer can vary from 5 nm to 100 nm and the chemical composition can be different. The examinations of the underlying

  9. The probable types, sizes, positions and orientations of the defects which may appear in connection with manufacture of reactor vessels

    International Nuclear Information System (INIS)

    Bergh, S.

    1980-02-01

    An review of welding technology in manufacture of reactor vessels is made. An inventory of principal defects appearing in connection with manual ARC-welding and coated electrodes is presented. Some important welded joints of BWR reactor vessels are scrutinized. Reheating cracks may appear during stress relief annealing beneath the cladding, and this problem is discussed in the third part. The interest is focussed towards the defects which depend on the conditions during the welding. Slag and incomplete fusion might be found. The review can serve for the guidance of nondestructive testing. The defects are estimated to have the size of a few MM with a maximum to approx. 10 MM right across the weld, possibly with exception for the electroslag welds of the OKG-1 reactor vessel. (GBn)

  10. The Texts of the Instruments connected with the Agency's Assistance to Finland in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-01-24

    The texts of the Supply Agreement between the Agency, the Government of Finland and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Finland, in connection with the Agency's assistance to the Government of Finland in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These agreements entered into force on 30 December 1960.

  11. The Texts of the Instruments connected with the Agency's Assistance to Pakistan in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1962-03-22

    The texts of the Supply Agreement between the Agency, the Government of Pakistan and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Pakistan, in connection with the Agency's assistance to the Government of Pakistan-in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These Agreements entered into force on 5 March 1962.

  12. The Texts of the Instruments connected with the Agency's Assistance to Yugoslavia in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-11-24

    The texts of the Supply Agreement between the Agency, the Government of Yugoslavia and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Yugoslavia, in connection with the Agency's assistance to the Government of Yugoslavia in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These Agreements entered into force on 4 October 1961.

  13. The Texts of the Instruments connected with the Agency's Assistance to Mexico in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-01-30

    The texts of the Supply Agreement between the Agency and the Governments of Mexico and the United States of America, and of the Project Agreement between the Agency and the Government of Mexico, in connection with the Agency's assistance to that Government in establishing a research reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 18 December 1963.

  14. The Texts of the Instruments connected with the Agency's Assistance to Iran in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1967-08-31

    The texts of the Supply Agreement between the Agency and the Governments of Iran and the United States of America, and of the Project Agreement between the Agency and the Government of Iran, connected with the Agency's assistance to the latter Government in establishing a research reactor project, are reproduced in this document for the information of all Members. The Agreements entered into force on 7 June and 10 May 1967 respectively.

  15. The Texts of the Instruments connected with the Agency's assistance to Uruguay in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1965-01-01

    The texts of the Supply Agreement between the Agency and the Governments of the United States of America and Uruguay, and of the Project Agreement between the Agency and the Government of Uruguay, in connection with the Agency's assistance to the latter Government in establishing a research reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 24 September 1965

  16. The Texts of the Instruments connected with the Agency's assistance to Uruguay in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-09

    The texts of the Supply Agreement between the Agency and the Governments of the United States of America and Uruguay, and of the Project Agreement between the Agency and the Government of Uruguay, in connection with the Agency's assistance to the latter Government in establishing a research reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 24 September 1965.

  17. The Texts of the Instruments connected with the Agency's Assistance to Pakistan in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1962-01-01

    The texts of the Supply Agreement between the Agency, the Government of Pakistan and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Pakistan, in connection with the Agency's assistance to the Government of Pakistan-in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These Agreements entered into force on 5 March 1962

  18. The Texts of the Instruments connected with the Agency's Assistance to Finland in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1961-01-01

    The texts of the Supply Agreement between the Agency, the Government of Finland and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Finland, in connection with the Agency's assistance to the Government of Finland in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These agreements entered into force on 30 December 1960

  19. The Texts of the Instruments connected with the Agency's Assistance to Iran in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1967-01-01

    The texts of the Supply Agreement between the Agency and the Governments of Iran and the United States of America, and of the Project Agreement between the Agency and the Government of Iran, connected with the Agency's assistance to the latter Government in establishing a research reactor project, are reproduced in this document for the information of all Members. The Agreements entered into force on 7 June and 10 May 1967 respectively

  20. The Texts of the Instruments connected with the Agency's Assistance to Yugoslavia in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1961-01-01

    The texts of the Supply Agreement between the Agency, the Government of Yugoslavia and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Yugoslavia, in connection with the Agency's assistance to the Government of Yugoslavia in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These Agreements entered into force on 4 October 1961

  1. The Texts of the Instruments connected with the Agency's Assistance to Mexico in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1964-01-01

    The texts of the Supply Agreement between the Agency and the Governments of Mexico and the United States of America, and of the Project Agreement between the Agency and the Government of Mexico, in connection with the Agency's assistance to that Government in establishing a research reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 18 December 1963

  2. Horizontal beam tubes in FRM-II

    International Nuclear Information System (INIS)

    Coors, D.; Vanvor, D.

    2001-01-01

    The new research reactor in Garching FRM-II is equipped with 10 leak tight horizontal beam tubes (BT1 - BT10), each of them consisting of a beam tube structure taking an insert with neutron channels. The design of all beam tube structures is similar whereas the inserts are adapted to the special requirements of the using of each beam tube. Inside the reflector tank the beam tube structures are shaped by the inner cones which are made of Al-alloy with circular and rectangular cross sections. They are located in the region of maximum neutron flux (exception BT10), they are directly connected to the flanges of the reflector tank, their lengths are about 1.5 m (exception BT10) and their axes are directed tagentially to the core centre thus contributing to a low γ-noise at the experiments. (orig.)

  3. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  4. Current status of materials development of nuclear fuel cladding tubes for light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Zhengang, E-mail: duan_zg@imr.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Yang, Huilong [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Satoh, Yuhki [Institute for Materials Research, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Abe, Hiroaki, E-mail: abe.hiroaki@n.t.u-tokyo.ac.jp [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan)

    2017-05-15

    Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

  5. Current status of materials development of nuclear fuel cladding tubes for light water reactors

    International Nuclear Information System (INIS)

    Duan, Zhengang; Yang, Huilong; Satoh, Yuhki; Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie; Abe, Hiroaki

    2017-01-01

    Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

  6. Development of Bundle Position-Wise Linear Model for Predicting the Pressure Tube Diametral Creep in CANDU Reactors

    International Nuclear Information System (INIS)

    Lee, Jae Yong; Na, Man Gyun

    2011-01-01

    Diametral creep of the pressure tube (PT) is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of a heat transport system. PT diametral creep leads to diametral expansion that affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux. Therefore, it is essential to predict the PT diametral creep in CANDU reactors, which is caused mainly by fast neutron irradiation, reactor coolant temperature and so forth. The currently used PT diametral creep prediction model considers the complex interactions between the effects of temperature and fast neutron flux on the deformation of PT zirconium alloys. The model assumes that long-term steady-state deformation consists of separable, additive components from thermal creep, irradiation creep and irradiation growth. This is a mechanistic model based on measured data. However, this model has high prediction uncertainty. Recently, a statistical error modeling method was developed using plant inspection data from the Bruce B CANDU reactor. The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. There are twelve bundles in a fuel channel and for each bundle, a linear model was developed by using the dependent variables, such as the fast neutron fluxes and the bundle temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3 and 4 were used to develop the BPLM models. The remaining 10 channels' data were used to test the developed BPLM models. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from the Units 2,3 and 4 in Korea. Two error components for the BPLM, which are the epistemic

  7. Feasibility study of chabazite absorber tube utilization in online absorption of released gaseous fission products and substitution of burnable absorber rods with chabazite absorber tubes in VVER-1000 reactor series

    International Nuclear Information System (INIS)

    Rahmani, Yashar

    2017-01-01

    Highlights: • Chabazite tubes are used for online removal of the released gaseous fission products. • The feasibility of using chabazite tubes instead of burnable absorber rods was studied. • A computational cycle was designed using the WIMSD5-B, CITATION-LDI2 and WERL codes. • In modeling fission gas release, the Weisman, Booth, Mason and T.S. models were used. • By this method, it is possible to increase cycle length and enhance heat transfer. - Abstract: As gaseous fission products, e.g. xenon and krypton have adverse effects such as reducing the rate of heat transfer in fuel rods and adding negative reactivity to the reactor core, the present manuscript was dedicated to development of a novel method for improving these defects. In the proposed method, chabazite absorber tubes were used for online removal of the released gaseous fission products from gaseous gap spaces. Moreover, in this research, feasibility of using chabazite absorber tubes instead of burnable absorber rods was examined. To perform the required modeling and calculations to successfully meet the mentioned objectives, a thermo-neutronic computational cycle was designed using the coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic calculations. In addition, in modeling the release process of gaseous fission products, the Weisman, Booth, Mason, and T.S. models were examined. It is worth mentioning that in this research, calculations and modeling procedures were based on the first cycle of Bushehr’s VVER-1000 reactor to study the feasibility of the proposed solution. The obtained results revealed that with application of the proposed method in this research, it is possible to increase cycle length, improve safety thresholds, and enhance heat transfer in the core of nuclear reactors.

  8. Techniques for processing remote field eddy current signals from bend regions of steam generator tubes of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thirunavukkarasu, S. [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India); Rao, B.P.C., E-mail: bpcrao@igcar.gov.in [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India); Jayakumar, T.; Raj, Baldev [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)

    2011-04-15

    Steam generator (SG) is one of the most critical components of sodium cooled fast breeder reactor. Remote field eddy current (RFEC) technique has been chosen for in-service inspection (ISI) of these ferromagnetic SG tubes made of modified 9Cr-1Mo steel (Grade 91). Expansion bends are provided in the SGs to accommodate differential thermal expansion. During ISI using RFEC technique, in expansion bend regions, exciter-receiver coil misalignment, bending stresses, probe wobble and magnetic permeability variations produce disturbing noise hindering detection of defects. Fourier filtering, cross-correlation and wavelet transform techniques have been studied for noise reduction as well as enhancement of RFEC signals of defects in bend regions, having machined grooves and localized defects. Performance of these three techniques has been compared using signal-to-noise ratio (SNR). Fourier filtering technique has shown better performance for noise reduction while cross-correlation technique has resulted in significant enhancement of signals. Wavelet transform technique has shown the combined capability of noise reduction and signal enhancement and resulted in unambiguous detection of 10% of wall loss grooves and localized defects in the bend regions with SNR better than 7 dB.

  9. An Investigation on Cocombustion Behaviors of Hydrothermally Treated Municipal Solid Waste with Coal Using a Drop-Tube Reactor

    Directory of Open Access Journals (Sweden)

    Liang Lu

    2012-01-01

    Full Text Available This work aims at demonstrating the feasibility of replacing Indonesian coal (INC with hydrothermally treated municipal solid waste (MSWH in cocombustion with high ash Indian coal (IC. The combustion efficiencies and emissions (CO, NO of MSWH, INC and their blends with IC for a series of tests performed under a range of temperatures and air conditions were tested in a drop-tube reactor (DTR. The results showed the following. The combustion efficiency of IC was increased by blending both MSWH and INC and CO emission was reduced with increasing temperature. For NO emission, the blending of MSWH led to the increase of NO concentration whereas the effects of INC depended on the temperature. The combustion behaviors of IC-MSWH blend were comparable to those of the IC-INC blend indicating it is possible for MSWH to become a good substitute for INC supporting IC combustion. Moreover, the CO emission fell while the NO emission rose with increasing excess air for IC-MSWH blend at 900°C and the highest combustion efficiency was obtained at the excess air of 1.9. The existence of moisture in the cocombustion system of IC-MSWH blend could slightly improve the combustion efficiency, reduce CO, and increase NO.

  10. Incorporation of a Helical Tube Heat Transfer Model in the MARS Thermal Hydraulic Systems Analysis Code for the T/H Analyses of the SMART Reactor

    International Nuclear Information System (INIS)

    Young Jin Lee; Bub Dong Chung; Jong Chull Jo; Hho Jung Kim; Un Chul Lee

    2004-01-01

    SMART is a medium sized integral type advanced pressurized water reactor currently under development at KAERI. The steam generators of SMART are designed with helically coiled tubes and these are designed to produce superheated steam. The helical shape of the tubes can induce strong centrifugal effect on the secondary coolant as it flows inside the tubes. The presence of centrifugal effect is expected to enhance the formation of cross-sectional circulation flows within the tubes that will increase the overall heat transfer. Furthermore, the centrifugal effect is expected to enhance the moisture separation and thus make it easier to produce superheated steam. MARS is a best-estimate thermal-hydraulic systems analysis code with multi-phase, multi-dimensional analysis capability. The MARS code was produced by restructuring and merging the RELAP5 and the COBRA-TF codes. However, MARS as well as most other best-estimate systems analysis codes in current use lack the detailed models needed to describe the thermal hydraulics of helically coiled tubes. In this study, the heat transfer characteristics and relevant correlations for both the tube and shell sides of helical tubes have been investigated, and the appropriate models have been incorporated into the MARS code. The newly incorporated helical tube heat transfer package is available to the MARS users via selection of the appropriate option in the input. A performance analysis on the steam generator of SMART under full power operation was carried out using the modified MARS code. The results of the analysis indicate that there is a significant improvement in the code predictability. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  12. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1984-10-01

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

  13. Discussion on amount of water ingress mass in steam generator heat-exchange tube rupture accident of high- temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Yan; Zheng Yanhua; Shi Lei; Li Fu; Sun Ximing

    2009-01-01

    The steam generator heat-exchange tube rupture (SGTR) accident which will result in the water ingress to the primary circuit of reactor is an important and particular accident for high-temperature gas-cooled reactor (HTGR). The analysis of the water ingress accident is significant for verifying the inherent safety characteristics of HTGR. The amount of water ingress mass is one of the decisive factors for the seriousness of the accident consequence. The 250 MW Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) designed by Institute of Nuclear and New Energy Technology of Tsinghua University was selected as an example of analysis. The analysis results show that the amount of water ingress mass is not only affected directly with the broken position and the broken area of the tubes, but also related with the diameter of draining piping and restrictor, draining control valve, action setting of emptier system. With reasonable parameters chosen, the water in steam generator could be drained effectively, so it will prevent the primary circuit of reactor from water ingress in large quantity and reduce the radioactive isotopes ingress to the secondary circuit. (authors)

  14. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1983-08-01

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

  15. Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

    International Nuclear Information System (INIS)

    Takizuka, T.; Tokunaga, S.; Hoshino, K.; Shimizu, K.; Asakura, N.

    2015-01-01

    Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor

  16. Status of October 1, 2004 of the repartition of nuclear reactors connected to the grid and under construction in the world

    International Nuclear Information System (INIS)

    2004-01-01

    This document is a table giving the number of nuclear reactors connected to the grid and under construction in the European countries and in the rest of the world. The percentage of electricity produced by these reactors in 2003 is indicated. (J.S.)

  17. Effect of kinetic parameters on simultaneous ramp reactivity insertion plus beam tube flooding accident in a typical low enriched U{sub 3}Si{sub 2}-Al fuel-based material testing reactor-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nasir, Rubina; Mirza, Nasir M. [Dept. of, Physics, Air University, Islamabad (Pakistan); Mirza, Sikander M. [Dept. of, Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences, Post Office Nilore, Islamabad (Pakistan)

    2017-06-15

    This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density (U{sub 3}Si{sub 2}-Al) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

  18. Motivations to Seek Science Videos on YouTube: Free-Choice Learning in a Connected Society

    Science.gov (United States)

    Rosenthal, Sonny

    2018-01-01

    Do individuals use video sharing sites in their free time to learn about science, and if so, why? This study takes a preliminary look at individual differences that motivate online science video seeking. Among 273 Singapore Internet users who participated in an online survey, most reported using YouTube during the previous week, and one-third…

  19. Ultrasonic testing of canning tubes in stainless steel of the EL 4 reactor; Controle par ultrasons des tubes de gaine en acier inoxydable du reacteur EL 4

    Energy Technology Data Exchange (ETDEWEB)

    Prot, A; Monnier, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    From all the methods possible for controlling thin cans the one chosen, for numerous reasons, vas that making use of ultrasonic techniques. A method has been developed which should make it possible to carry out a rapid and efficient industrial control of canning tubes, The reasons for the choice of the ultrasonic method are given in detail, together with the principles of the method and the actual control parameters. In the present state of our research, it should be possible to control at least 50 000 tubes a year. Improvements brought about in the details of the control technique itself should make it possible to increase this rate considerably. (authors) [French] Parmi toutes les methodes possibles de controle des gaines minces, le procede retenu pour de multiples raisons a ete celui faisant appel a la technique des ultrasons. Une methode a ete mise au point qui doit permettre un controle industriel rapide et efficace des tubes de gaine. Sont exposes en detail, les raisons du choix de la methode par ultrasons, les principes de cette methode et les parametres du controle proprement dit. Dans l'etat actuel de nos etudes la cadence devrait permettre le controle de 50000 tubes par an au minimum. Des ameliorations de detail portant sur la technique de controle elle-meme, doivent permettre d'accelerer tres notablement cette cadence. (auteurs)

  20. Reactors

    International Nuclear Information System (INIS)

    Onuki, Koji; Sasanuma, Katsumi.

    1980-01-01

    Purpose: To make it possible to correctly measure the flow rate and temperatures of the coolants flowing through fuel assemblies. Constitution: One or more holes are formed at the side surface of the guide tube of a control rod driving mechanism thereby to reduce the flow path resistance within the guide tube of the control rod driving mechanism and to prevent the outlet coolant of the control rod guide tube from flowing into the guide tube of the mechanism as it is and also from flowing into ambient rectifying lattice guide tubes, so that the quantities and temperatures of the coolants flowing through respective fuel assemblies can be measured correctly. (Kamimura, M.)

  1. Connecting world youth with tobacco brands: YouTube and the internet policy vacuum on Web 2.0.

    Science.gov (United States)

    Elkin, Lucy; Thomson, George; Wilson, Nick

    2010-10-01

    The internet is an ideal forum for tobacco marketing, as it is largely unregulated and there is no global governing body for controlling content. Nevertheless, tobacco companies deny advertising on the internet. To assess the extent and nature of English language videos available on the Web 2.0 domain 'YouTube' that contain tobacco brand images or words. The authors conducted a YouTube search using five leading non-Chinese cigarette brands worldwide. The themes and content of up to 40 of the most viewed videos returned for each search were analysed: a total of 163 videos. A majority of the 163 tobacco brand-related videos analysed (71.2%, 95% CI 63.9 to 77.7) had pro-tobacco content, versus a small minority (3.7%) having anti-tobacco content (95% CI 1.4 to 7.8). Most of these videos contained tobacco brand content (70.6%), the brand name in the title (71.2%) or smoking imagery content (50.9%). One pro-smoking music video had been viewed over 2 million times. The four most prominent themes of the videos were celebrity/movies, sports, music and 'archive', the first three of which represent themes of interest to a youth audience. Pro-tobacco videos have a significant presence on YouTube, consistent with indirect marketing activity by tobacco companies or their proxies. Since content may be removed from YouTube if it is found to breach copyright or if it contains offensive material, there is scope for the public and health organisations to request the removal of pro-tobacco content containing copyright or offensive material. Governments should also consider implementing Framework Convention on Tobacco Control requirements on the internet, to further reduce such pro-tobacco content.

  2. Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-08-15

    The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  3. Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Sang June Ahn

    2016-08-01

    Full Text Available The prototype generation IV sodium-cooled fast reactor (PGSFR has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS and the safety of the primary heat-transfer system (PHTS. In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  4. Two design aspects connected with the safety of the PIK reactor presently under construction

    International Nuclear Information System (INIS)

    Gostev, V.V.; Zakharov, A.S.; Konoplev, K.A.; Levandovskii, N.V.; Ploshchanskii, L.M.; Smolsky, S.L.

    1993-01-01

    The PIK reactor is designed for physical research with neutron beams and sample irradiation. In the central trap the thermal neutrons flux is 4x10 15 n/cm 2 s. The reactor power is 100 MW, the thermal neutron flux in the reflector at the maximum of distribution is 1x10 15 n/cm 2 s. The core with a high uranium concentration of 600 g/l is light water-cooled, heavy water being used in the reflector. The Chernobyl disaster happened at the time of equipment installation at the PIK. The code revision, a change of the authors ideas about the safety, and a change of public attitude towards nuclear installations resulted in a stopping of construction and project revision. Reconstruction project has led to a change of all safety systems and involved in various degrees all essential reactor systems. The construction is presently resumed in spite of economic difficulties in Russia. The reactor was inspected by experts from a number of European countries, USA, and European Commission delegated by their governments to prepare a report on whether supporting the construction to its completion would be reasonable. In the course of inspection the experts from USA and EU expressed doubts concerning two systems, namely, the containment and scram. These two points are discussed in the present paper. Three type of containments are proposed and an analysis of their efficiency is presented. The PIK reactor is controlled by eight rods in the heavy-water reflector -and an absorbing cylinder at the boundary between the core and the central light-water neutron trap. The rods are used for emergency protection and reactor start-up. The central control cylinder called here the shutter serves several functions, namely, as scram, automatic control, and burnup compensation. The delay time before the onset of negative reactivity is 1.05 sec for rods and 0.25 sec for the shutter

  5. Locking device of a guiding ring on a plate including an aperture; application to guide tube of nuclear reactor

    International Nuclear Information System (INIS)

    Cauquelin, C.; Poitrenaud, P.

    1987-01-01

    To make easy to take to pieces a guide tube, by a simple tool, this device includes a guide ring. This guide ring aligned with an aperture in a plate has a tubular support fixed to the plate and coaxial with the aperture and lock the guide tube by rotation [fr

  6. Tests for development of estimation technology of reactor core deformation. Report No.1: fundamental mechanical properties of wrapper tube (test report)

    International Nuclear Information System (INIS)

    Nishiura, Takeo; Shimazaki, Yuji; Horikiri, Morito

    1998-10-01

    Mechanical properties such as local contact compression stiffness, bending stiffness, deformation properties, material properties, and friction properties of a wrapper tube structure were clarified experimentally, which can be used as the basic data for development of estimation technology of reactor core deformation. Contents of the Tests data as follows: (1) Effects of load supporting boundary conditions, whether or not a contact-proof pad is attached, and length of duct, on cross section deformation of wrapper tube were made clear as the local contact compression stiffness characteristics. (2) Bending stiffness does not depend on the difference of load supporting boundary conditions. The property of cross section deformation under bending load was obtained. (3) The deformation modes and the strain distributions were obtained by the deformation tests of wrapper tube. (4) The stress-strain diagrams including plastic range under various strain variation rates were obtained by the material tests at room temperature. (5) The static and the dynamic friction coefficients by various contact angles and the contact loads between contact-proof pads of two wrapper tubes were obtained by friction property tests. (author)

  7. A review of current knowledge on the effects of hydrogen on the pressure tubes of Ontario Hydro operating reactors

    International Nuclear Information System (INIS)

    Leger, M.

    1982-01-01

    Since the occurrence of cracking in Zr-2.5 wt% Nb pressure tubes in Pickering 'A' units 3 and 4 in 1974/75 a great deal of information on the behaviour of hydrogen in pressure tube materials has been generated through research effort by both AECL and Ontario Hydro. In order to use this information effectively and to provide direction and co-ordination for ongoing research, a review of available information and current concerns on hydrogen in pressure tubes was undertaken. The review was divided into two main areas of interest: hydrogen ingress and hydride effects. The uncertainties in the rates of hydrogen ingress into the pressure tubes have been found to be very large. On the basis of current knowledge, predictions of the future behaviour of pressure tubes due to hydride effects are extremely difficult

  8. Influence of structure improvement of guide tubes and bundles in pressurized water reactor (PWR) on drop of control rods

    International Nuclear Information System (INIS)

    Shen Xiuzhong; Yu Pingan; Yang Guanyue

    1996-01-01

    In order to alleviate the cross hydraulic load on control rod guide tubes and bundles, some protective sleeves are added to those near the upper plenum outlet nozzles (4 symmetric bundles: 02-26, 03-25, 11-29, 12-28). In a 1/4 scale transparent model of the PWR upper plenum of Qinshan Nuclear Power Station, water was chosen as the fluid and hydraulic experiments with improved control rod guide tubes and bundles were carried out. The results were carefully compared with those of the experiments with unimproved control rod guide tubes and bundles. It is concluded that adding protective sleeves to the control rod guide tubes and bundles near the outlet nozzles will help to lighten the hydraulic load on them and make certain of the free movement and rapid dropping of control rods in the tubes and bundles in emergency by order

  9. Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: eleodor.nichita@uoit.ca; Kovaltchouk, V., E-mail: vitali.kovaltchouk@uoit.ca

    2015-12-15

    Highlights: • Typical PT-SCWR fuel uses single-region pins consisting of a homogeneous mixture of ThO{sub 2} and PuO{sub 2}. • Using two regions (central for the ThO{sub 2} and peripheral for the PuO{sub 2}) reduces the fuel temperature. • Single-region-pin melting-to-average power ratio is 2.5 at 0.0 MW d/kg and 2.3 at 40 MW d/kg. • Two-region-pin melting-to-average power ratio is 36 at 0.0 MW d/kg and 10.5 at 40 MW d/kg. • Two-region-pin performance drops with burnup due to fissile-element buildup in the ThO{sub 2} region. - Abstract: The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625 K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO{sub 2}) and fissile (PuO{sub 2}) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650 K for fresh fuel compared to the case of a homogeneous (Th–Pu)O{sub 2} mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is

  10. A vapor generator equipped with an advanced drain device for the secondary side of the tubes plate

    International Nuclear Information System (INIS)

    Valadon, C.

    1995-01-01

    A draining design is proposed for the tube plate secondary side in a PWR type reactor, that does not interfere with the water flush 'street' thus allowing for an easy inspection and maintenance in the lower part of the tube bundle. The draining system is composed of a main groove on the upper side of the tube plate, which is connected to draining means situated outside the vapor generator. 6 fig

  11. Specific induced activity profile at the rotary specimen rack of IPR-R1 TRIGA reactor after the introduction of a new pneumatic transfer tube

    International Nuclear Information System (INIS)

    Souza, Luiz Claudio Andrade; Zangirolami, Dante Marco; Maretti Junior, Fausto; Ferreira, Andrea Vidal

    2011-01-01

    The IPR-R1 TRIGA nuclear reactor is located in Belo Horizonte, Brazil, at the Nuclear Technology Development Center (Centro de Desenvolvimento da Tecnologia Nuclear, CDTN) of the National Committee on Nuclear Energy (Comissao Nacional de Energia Nuclear, CNEN). One of its irradiation devices is the rotary specimen rack (RSR), outside the reactor core, with forty irradiation positions arranged in a cylindrical geometry. In a previous work, the neutron fluence rate distribution at the RSR and its variation under different irradiation conditions were evaluated by means of specific induced activity measurements in samples of Al-0.1%Au reference material. Since then the core's configuration has been altered with the (re)introduction of another irradiation device, the pneumatic transfer tube 1 (PT-1). This paper aims at identifying and quantifying any changes in neutron fluence that such modification may have caused. (author)

  12. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung-Kyu, E-mail: power@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Kim, Kwangmin; Park, Minwon [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of)

    2015-11-15

    Highlights: • A 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC transmission system. • The 400 mH class HTS DC reactor was connected to real power network via the HVDC system. • The DC current flowed in HTS DC reactor has several harmonic components and it was analyzed using FFT. - Abstract: High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  13. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    International Nuclear Information System (INIS)

    Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

    2015-01-01

    Highlights: • A 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC transmission system. • The 400 mH class HTS DC reactor was connected to real power network via the HVDC system. • The DC current flowed in HTS DC reactor has several harmonic components and it was analyzed using FFT. - Abstract: High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  14. A nuclear reactor core fuel reload optimization using Artificial-Ant-Colony Connective Networks

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Schirru, Roberto

    2005-01-01

    A Pressurized Water Reactor core must be reloaded every time the fuel burnup reaches a level when it is not possible to sustain nominal power operation. The nuclear core fuel reload optimization consists in finding a burned-up and fresh-fuel-assembly pattern that maximizes the number of full operational days. This problem is NP-hard, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Besides that, the problem is non-linear and its search space is highly discontinual and multimodal. In this work a parallel computational system based on Ant Colony System (ACS) called Artificial-Ant-Colony Networks is introduced to solve the nuclear reactor core fuel reload optimization problem. ACS is a system based on artificial agents that uses the reinforcement learning technique and was originally developed to solve the Traveling Salesman Problem, which is conceptually similar to the nuclear fuel reload problem. (author)

  15. Helically coiled tube heat exchanger

    International Nuclear Information System (INIS)

    Harris, A.M.

    1981-01-01

    In a heat exchanger such as a steam generator for a nuclear reactor, two or more bundles of helically coiled tubes are arranged in series with the tubes in each bundle integrally continuing through the tube bundles arranged in series therewith. Pitch values for the tubing in any pair of tube bundles, taken transverse to the path of the reactor coolant flow about the tubes, are selected as a ratio of two unequal integers to permit efficient operation of each tube bundle while maintaining the various tube bundles of the heat exchanger within a compact envelope. Preferably, the helix angle and tube pitch parallel to the path of coolant flow are constant for all tubes in a single bundle so that the tubes are of approximately the same length within each bundle

  16. Realistic thermal transient margin analysis of 'MONJU' based on plant performance measurements. Reactor vessel outlet nozzle and evaporator feed water inlet tube sheet of the manual reactor plant trip

    International Nuclear Information System (INIS)

    Yamada, Fumiaki; Mori, Takero

    2005-01-01

    In order to develop technologies and achieve safe and stable operation of Monju' as well as realize optimized design and construction of safe and economically competitive fast breeder reactors, the authors are evaluating design approach applied to 'Monju' based on actually measured behavioral data during plant operations. This report uses actual measured characteristic data of 'Monju' during a plant trip test obtained at a commissioning stage with up to 40% power output and introduces plant thermal hydraulic behavior analysis in a representative thermal transient event, i.e. a manual plant trip. Thermal transient driven loads incurred by the reactor vessel outlet nozzle and by the evaporator feed water inlet tube sheet were further derived by structural analyses and were compared with the previously derived values in the design stage and with the limit values. Though the reactor vessel outlet nozzle was exposed to larger temperature change in the trip test than the analytical prediction, the newly calculated mechanical load was about 50% of the previous evaluation in the design stage. Also, the newly analyzed mechanical load incurred by the evaporator feed water inlet tube sheet in this event had a large margin against the limit value of cumulative damage cycle fraction, although the observed temperature disturbance in a steam blow test was wilder than the analytical prediction. Thus we concluded that the Monju' plant has an assured safety margin against thermal transient in plant trip events. (author)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  18. Parallel connecting poloidal coil system for a doublet tokamak fusion reactor

    International Nuclear Information System (INIS)

    Toffolo, W.E.; Chen, W.Y.; Purcell, J.R.; Wesley, J.C.

    1977-09-01

    A method has been developed for parallel connection of the ohmic heating (OH) coil. The method involves subdividing the OH-coil into a number of parallel connected subcoils, with each subcoil having about 20 turns. Each of the field shaping coils (F-coils) also contains 20 turns, so that when connected to a common power supply, the OH and F-coils are decoupled. The advantages resulting from the scheme are numerous: (1) each F-coil contains a much smaller number of turns compared with the previous design concept, thus the construction and maintenance will be easier; (2) the parallel connected OH-coils form a constant flux envelope, resulting in an inherently lower error field at the plasma and the TF coil region, and this low error field is not sensitive to the variation in location of the OH-coils; (3) the voltage and current ratings of the individual OH coil conductors are reduced; and (4) the low impedance of the OH-coil system greatly improves the possibility of using a homopolar motor generator as a means of achieving flux reversal during startup and plasma current control during the burn cycle

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

    International Nuclear Information System (INIS)

    2006-05-01

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

  20. Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

    International Nuclear Information System (INIS)

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-01-01

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology

  1. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Schirru, Roberto; Carvalho da Silva, Fernando; Medeiros, Jose Antonio Carlos Canedo

    2008-01-01

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem

  2. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: alanmmlima@yahoo.com.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br; Carvalho da Silva, Fernando [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: fernando@con.ufrj.br; Medeiros, Jose Antonio Carlos Canedo [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br

    2008-09-15

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem.

  3. Benchmark simulation of turbulent flow through a staggered tube bundle to support CFD as a reactor design tool. Part 1. SRANS CFD simulation

    International Nuclear Information System (INIS)

    Ridluan, Artit; Tokuhiro, Akira

    2008-01-01

    Time-invariant and time-variant numerical simulations of flow through a staggered tube bundle array, idealizing the lower plenum (LP) subsystem configuration of a very high temperature reactor (VHTR), were performed. In Part 1, the CFD prediction of fully periodic isothermal tube-bundle flow using steady Reynolds-averaged Navier-Stokes (SRANS) equations with common turbulence models was investigated at a Reynolds number (Re) of 1.8x10 4 , based on the tube diameter and inlet velocity. Three first-order turbulence models, standard k-ε turbulence, renormalized group (RNG) k-ε, and shear stress transport (SST) k-ω models, and a second-order turbulence model, Reynolds stress model (RSM), were considered. A comparison of CFD simulations and experiment results was made at five locations along (x,y) coordinates. The SRANS simulation showed that no universal model predicted the turbulent Reynolds stresses, and generally, the results were marginal to poor. This is because these models cannot accurately model the periodic, spatiotemporal nature of the complex wake flow structure. (author)

  4. Investigation of the deposit formation in pipelines connecting liquefaction reactors; 1t/d PSU ni okeru ekika hanno tokan fuchakubutsu no seisei yoin ni kansuru ichikosatsu

    Energy Technology Data Exchange (ETDEWEB)

    Okada, Y.; Nogami, Y.; Inokuchi, K. [Mitsui SRC Development Co. Ltd., Tokyo (Japan); Mochizuki, M.; Imada, K. [Nippon Steel Corp., Tokyo (Japan)

    1996-10-28

    The liquefaction reaction system of an NEDOL process coal liquefaction 1t/d PSU was opened and checked to investigate the cause of the rise of differential pressure between liquefaction reactors of the PSU. The liquefaction test at a coal concentration of 50 wt% using Tanito Harum coal was conducted, and it was found that the differential pressure between reactors was on the increase. By the two-phase flow pressure loss method, deposition thickness of deposit in pipelines was estimated at 4.4mm at the time of end operation, which agreed with a measuring value obtained from a {gamma} ray. The rise of differential pressure was caused by deposit formation in pipelines connecting reactors. The main component of the deposit is calcite (CaCO3 60-70%) and is the same as the usual one. It is also the same type as the deposit on the reactor wall. Ca in coal ash is concerned with this. To withdraw solid matters deposited in the reactor, there are installed pipelines for the withdrawal at the reactor bottom. The solid matters are regularly purged by reverse gas for prevention of clogging. As the frequency of purge increases, the deposit at the reactor bottom decreases, but the deposit attaches strongly to pipelines connecting reactors. It is presumed that this deposit is what Ca to be discharged out of the system as a form of deposition solid matter naturally in the Ca balance precipitated as calcite in the pipeline connecting the reactor. 3 refs., 5 figs., 4 tabs.

  5. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  6. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    Science.gov (United States)

    Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

    2015-11-01

    High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  7. Biodegradation of 2,4,6-trichlorophenol in a packed-bed biofilm reactor equipped with an internal net draft tube riser for aeration and liquid circulation

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-De Jesus, A.; Romano-Baez, F.J.; Leyva-Amezcua, L.; Juarez-Ramirez, C.; Ruiz-Ordaz, N. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico); Galindez-Mayer, J. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico)], E-mail: cmayer@encb.ipn.mx

    2009-01-30

    For the aerobic biodegradation of the fungicide and defoliant 2,4,6-trichlorophenol (2,4,6-TCP), a bench-scale packed-bed bioreactor equipped with a net draft tube riser for liquid circulation and oxygenation (PB-ALR) was constructed. To obtain a high packed-bed volume relative to the whole bioreactor volume, a high A{sub D}/A{sub R} ratio was used. Reactor's downcomer was packed with a porous support of volcanic stone fragments. PB-ALR hydrodynamics and oxygen mass transfer behavior was evaluated and compared to the observed behavior of the unpacked reactor operating as an internal airlift reactor (ALR). Overall gas holdup values {epsilon}{sub G}, and zonal oxygen mass transfer coefficients determined at various airflow rates in the PB-ALR, were higher than those obtained with the ALR. When comparing mixing time values obtained in both cases, a slight increment in mixing time was observed when reactor was operated as a PB-ALR. By using a mixed microbial community, the biofilm reactor was used to evaluate the aerobic biodegradation of 2,4,6-TCP. Three bacterial strains identified as Burkholderia sp., Burkholderia kururiensis and Stenotrophomonas sp. constituted the microbial consortium able to cometabolically degrade the 2,4,6-TCP, using phenol as primary substrate. This consortium removed 100% of phenol and near 99% of 2,4,6-TCP. Mineralization and dehalogenation of 2,4,6-TCP was evidenced by high COD removal efficiencies ({approx}95%), and by the stoichiometric release of chloride ions from the halogenated compound ({approx}80%). Finally, it was observed that the microbial consortium was also capable to metabolize 2,4,6-TCP without phenol as primary substrate, with high removal efficiencies (near 100% for 2,4,6-TCP, 92% for COD and 88% for chloride ions)

  8. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  9. Examination of parameters affecting overload fracture behavior of flaw-tip hydrides in Zr-2.5Nb pressure tubes in Candu reactors

    International Nuclear Information System (INIS)

    Cui, J.; Shek, G.K.; Wang, Z.R.

    2007-01-01

    Service-induced flaws in Zr-2.5Nb alloy pressure tubes in Candu (Canada Deuterium Uranium Reactors) nuclear reactors are susceptible to a crack initiation and growth mechanism known as Delayed Hydride Cracking (DHC), which is a repetitive process that involves hydrogen diffusion, hydride precipitation, growth and fracture of a hydride region at the flaw-tip under a constant load. Crack initiation may also occur under another loading condition when the hydride region is subjected to an overload. An overload occurs when the hydride region at the flaw tip is loaded to a stress higher than that at which this region is formed such as when the reactor experiences a transient pressure higher than the normal operating pressure where the hydride region is formed. Flaw disposition requires justification that the hydride region overload will not fracture the hydride region, and initiate DHC. In this work, monotonically increasing load experiments were performed on unirradiated Zr-2.5Nb pressure tube specimens containing simulated debris frets (V-notch) and bearing pad frets (BPF, U-shape notch) to examine overload fracture behavior of flaw-tip hydrides formed under hydride ratcheting conditions. Hydride cracking in the overload tests was detected by the acoustic emission technique and confirmed by post-test metallurgical examination. Test results indicate that the resistance to overload fracture is affected by a number of parameters including hydride formation stress, flaw shape (V-notch vs. BPF) and flaw radius (0.015 mm vs. 0.1 mm). The notch-tip hydride morphologies were examined by optical microscopy and scanning electron microscopy (SEM) which show that they are affected by the hydride formation conditions, resulting in different overload fracture resistance. Finite element stress analyses were also performed to obtain flaw-tip stress distributions for interpretation of the test results. (authors)

  10. Electron tube

    Science.gov (United States)

    Suyama, Motohiro [Hamamatsu, JP; Fukasawa, Atsuhito [Hamamatsu, JP; Arisaka, Katsushi [Los Angeles, CA; Wang, Hanguo [North Hills, CA

    2011-12-20

    An electron tube of the present invention includes: a vacuum vessel including a face plate portion made of synthetic silica and having a surface on which a photoelectric surface is provided, a stem portion arranged facing the photoelectric surface and made of synthetic silica, and a side tube portion having one end connected to the face plate portion and the other end connected to the stem portion and made of synthetic silica; a projection portion arranged in the vacuum vessel, extending from the stem portion toward the photoelectric surface, and made of synthetic silica; and an electron detector arranged on the projection portion, for detecting electrons from the photoelectric surface, and made of silicon.

  11. Remotely controlled device for tightening, the nuts on locating pins for guide tubes in pressurized water reactors

    International Nuclear Information System (INIS)

    Styskal, P.

    1991-01-01

    The device has a support having a horizontal guide radial to the guide tube with a trolley moving on the guide and mounted on it a tool carrier. The tightening tool it self consists of a motor and an assembly of reducing gears mounted on the tool carrier. The final gear wheel in the assembly turns about a vertical axis and has a ferrule on its face for tightening the nut of the guide tube locating pin. The force of reaction on the tool carrier may be measured thus allowing the torque applied by the tool to be regulated [fr

  12. Simulation of steam condensation in the presence of noncondensable gases in horizontal condenser tubes using RELAP5 for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Macedo, Luiz Alberto; Torres, Walmir Maximo

    2009-01-01

    Horizontal heat exchangers are used in advanced light water nuclear reactors in their passive cooling systems, such as residual heat removal (RHRS) and passive containment cooling system (PCCS). Condensation studies of steam and noncondensable gases mixtures in these heat exchangers are very important due to the phenomena multidimensional nature and the condensate stratification effects. This work presents a comparison between simulation results and experimental data in steady state conditions for some inlet pressure, steam and noncondensable gases (air) inlet mass fractions. The test section is three meters long and consists of two concentric tubes containing pressure, temperature and flow rate sensors. The internal tube, called condenser, contains steam-air mixture flow and external tube is a counter current cooler with water flow rate at low temperature. This test section was modeled and simulations were performed with RELAP5 code. Experimental tests were carried out for 200 to 400 kPa inlet pressure and 5, 10, 15 and 20% of inlet air mass fractions. Comparisons between experimental data and simulation results are presented for 200 and 400 kPa pressure conditions and showed good agreement. However, for 400 kPa inlet steam pressure and inlet air mass fractions above 5%, the simulated temperatures are lower than the experimental data at the final third from the inlet condenser tube, indicating a code overestimation of heat transfer coefficient. New correlations for heat transfer coefficient in these steam-air conditions must be theoretical and experimentally studied and implemented in RELAP5 code for better representing the condensation phenomena. (author)

  13. Steam generator tube failures

    International Nuclear Information System (INIS)

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

    1996-04-01

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

  14. Coaxial tube array space transmission line characterization

    International Nuclear Information System (INIS)

    Switzer, C.A.; Bents, D.J.

    1987-01-01

    The coaxial tube array tether/transmission line used to connect an SP-100 nuclear power system to the space station was characterized over the range of reactor-to-platform separation distances of 1 to 10 km. Characterization was done with respect to array performance, physical dimensions and masses. Using a fixed design procedure, a family of designs was generated for the same power level (300 kWe), power loss (1.5 percent), and meteoroid survival probability (99.5 percent over 10 yr). To differentiate between vacuum insulated and gas insulated lines, two different maximum values of the E field were considered: 20 kV/cm (appropriate to vacuum insulation) and 50 kV/cm (compressed SF6). Core conductor, tube, bumper, standoff, spacer and bumper support dimensions, and masses were also calculated. The results of the characterization show mainly how transmission line size and mass scale with reactor-to-platform separation distance

  15. Electrical measuring device for a high temperature reactor

    International Nuclear Information System (INIS)

    Elter, C.; Handel, H.; Schoening, J.; Schmitt, H.

    1982-01-01

    The device for measuring the low or high neutron flux during start-up or at load is accommodated in an armoured guide tube projecting into the floor. A gas-tight capsule is formed as the measuring column with outer dome with a lid solidly connected by a flange to the armoured tube situated on the side wall of the concrete reactor vessel, together with the armoured guide tube. Two shielding shutters prevent the passage of radiation through the armoured tube. (DG) [de

  16. The Texts of the Instruments connected with the Agency's Assistance to the Democratic Republic of the Congo in Continuing a Research Reactor Project Project. Extension Agreement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-11-18

    The text of the Project Extension Agreement between the Agency and the Government of the Democratic Republic of the Congo in connection with the Agency's additional assistance to that Government in continuing a research reactor project is reproduced in this document for the information of all Members. This Agreement entered into force on 27 September 1966.

  17. The Texts of the Instruments connected with the Agency's Assistance to Finland in relation to a Research Reactor Project. A Fourth Supply Agreement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1970-02-05

    As a sequel to the assistance which the Agency provided to the Government of Finland in connection with a research reactor project, a fourth Supply Agreement has been concluded between the Agency and the Governments of Finland and the United States of America. This Agreement entered into force on 27 November 1969, and the text is reproduced herein for the information of all Members.

  18. The Texts of the Instruments connected with the Agency's Assistance to Finland in relation to a Research Reactor Project. A Third Supply Agreement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-01-12

    As a sequel to the assistance which the Agency provided to the Government of Finland in connection with a research reactor project), a third Supply Agreement has been concluded between the Agency and the Governments of Finland and the United States of America. This Agreement entered into force on 5 November 1967, and the text) is reproduced herein for the information of all Members.

  19. The Texts of the Instruments connected with the Agency's Assistance to Yugoslavia in Establishing a Research Reactor Project. Third Supply Agreement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1971-03-11

    As a sequel to the assistance which the Agency provided to the Government of Yugoslavia in connection with a research reactor project, a third Supply Agreement has been concluded between the Agency and the Governments of the United States of America and Yugoslavia. This Agreement entered into force on 30 December 1970, and the text is reproduced herein for the information of all Members.

  20. The Texts of the Instruments connected with the Agency's Assistance to Finland in relation to a Research Reactor Project. A Third Supply Agreement

    International Nuclear Information System (INIS)

    1968-01-01

    As a sequel to the assistance which the Agency provided to the Government of Finland in connection with a research reactor project), a third Supply Agreement has been concluded between the Agency and the Governments of Finland and the United States of America. This Agreement entered into force on 5 November 1967, and the text) is reproduced herein for the information of all Members

  1. The Texts of the Instruments connected with the Agency's Assistance to Yugoslavia in Establishing a Research Reactor Project. Third Supply Agreement

    International Nuclear Information System (INIS)

    1971-01-01

    As a sequel to the assistance which the Agency provided to the Government of Yugoslavia in connection with a research reactor project, a third Supply Agreement has been concluded between the Agency and the Governments of the United States of America and Yugoslavia. This Agreement entered into force on 30 December 1970, and the text is reproduced herein for the information of all Members

  2. The Texts of the Instruments connected with the Agency's Assistance to Finland in relation to a Research Reactor Project. A Fourth Supply Agreement

    International Nuclear Information System (INIS)

    1970-01-01

    As a sequel to the assistance which the Agency provided to the Government of Finland in connection with a research reactor project, a fourth Supply Agreement has been concluded between the Agency and the Governments of Finland and the United States of America. This Agreement entered into force on 27 November 1969, and the text is reproduced herein for the information of all Members

  3. Thermo-hydraulic characteristics of serpentine tubing in the boilers of gas cooled reactors under condition of rapid and slow depressurization

    International Nuclear Information System (INIS)

    Abouhadra, D.S.; Byrne, J.E.

    2003-01-01

    In nuclear reactors of the magnox or advanced gas cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accidents using two phase flow codes requires knowledge of the heat transfer behaviour of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear electric . The tests were carried out on the thermal hydraulics experimental research assembly (THERA) loop at manchester university. Depressurization from an initial pressure of 60 bar, with fluid subcooling of 5 k, 50 k, and 100 k was controlled by discharging the test section contents through suitably chosen orifices to produce blowdown to 10% of the initial pressure over a time scale of 30 s to 3600 s. pressures and temperatures in the serpentine were measured at average time intervals of approximately 1 s

  4. USE OF A GRIFFITH TUBE TO EVALUATE THE ANAEROBIC SLUDGE SEDIMENTATION IN A UASB REACTOR TREATING AN EFFLUENT WITH LONG-CHAIN FATTY ACIDS

    Directory of Open Access Journals (Sweden)

    L. A. S. Miranda

    Full Text Available Abstract This paper proposes to study the sedimentation characteristics of anaerobic sludge, by determining the settling velocity of sludge granules with the Griffith Tube. This is a simple, low-cost method, suitable for use in full-scale treatment plants. The settling characteristics of sludge from two laboratory-scale UASB reactors fed with saccharose and different concentrations of sodium oleate and sodium stereate were evaluated. Addition of fatty acids caused a gradual destabilization of the system, affecting overall performance. The sedimentation profile changed after addition of fatty acids to the synthetic substrate, decreased sedimentation velocity and increased granule diameter. This behaviour was attributed to the adsorption of fatty acids onto the granules, modifying the diameter, shape and density of these bioparticles.

  5. Analysis of multiple-tube ruptures in both steam generators for the Three Mile Island-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1985-01-01

    The operator guidelines were followed for both transients described. Both transients resulted in SG overfill and the tube-rupture flow did not terminate in either transient. The following statements can be deducted from the results of the calculations: the tube-rupture flow could not be stopped for either case during 2600 s (43 min) of transient time; each accident scenario resulted in SG overfill; both SGs overfilled by 1600 s (27 min) and 1800 s (30 min) for Cases 1 and 2, respectively; conditions for isolation of the SGs were not reached; and core subcooling was not lost in either case but the upper head was voided in Case 2. Comparison of the cooldown rates in the two cases after 1200 s (20 min) shows that these rates are equal (i.e., restart of the RCPs did not change the primary-system cooldown rate). However, in Case 2, a steam bubble was formed in the upper head, which did not disappear during the simulated time. One of the immediate actions in the guidelines was to fill both SGs to 95% level. This step was almost unnecessary because the tube-rupture flow was large enough that the 15.4 - mK/s (100 - 0 F h) cooldown-rate limit was exceeded and AFW could not be injected. Also, the guidelines did not address the SG overfill issue

  6. Plating end fittings to reduce hydrogen ingress at rolled joints in CANDU reactors

    International Nuclear Information System (INIS)

    White, A.J.; Urbanic, V.F.; Bahurmuz, A.A.; Clendening, W.R.; Joynes, R.; McDougall, G.M.; Skinner, B.C.; Venkatapathi, S.

    1993-10-01

    Zr-2.5Nb pressure tubes in CANDU nuclear reactors absorb hydrogen at a low rate from the primary heat transport water circulated through the tubes. Extra hydrogen is picked up at the rolled joints that connect the pressure tubes to out-of-core steel piping. This enhanced ingress may contribute to pressure-tube cracking at incorrectly assembled joints. The risk of pressure-tube failure has been decreased by ensuring correct joint assembly, and could be further decreased by reducing hydrogen ingress at rolled joints. This paper reviews progress toward using plated end fittings to reduce rolled-joint hydrogen ingress

  7. The simulation of the process of sodium freezing in the tubes for the optimization of fast breeder reactor units maintenance

    International Nuclear Information System (INIS)

    Tashlykov, O.L.; Shcheklein, S.E.; Annikov, S.V.

    2013-01-01

    The peculiarities of the repair works of the fast breeder reactor sodium systems are considered. The requirements for the sodium melting exclusion inside the equipment and piping during their opening and repair are given. The results of the sodium cooling process simulation with SolidWorks software are also described [ru

  8. A method for emergency flooding of the gland in the main circulating pump of pressurized water reactors and the connection therefor

    International Nuclear Information System (INIS)

    Skalicky, A.

    1978-01-01

    A method is described for the emergency flooding of the main circulating pumps of a pressurized water reactor such that in pressure drop in the flooded gland owing to pump suction, the pump head is connected by the pressure difference action to the flooding gland pipe, this via the heat sink and the filter of the emergency flooding circuit connected to the pump head. The emergency flooding circuit consisting of a pressure reducing valve, a check valve and a stop valve is connected to the pump head, behind the heat sink and the filter. The pressure reducing valve separates two pressure spaces. The former is connected to the pump head via the check valve and to the flooding pipe via the stop valve and the check valve. The latter is connected to the suction pump. (B.S.)

  9. Cooling device for reactor container

    International Nuclear Information System (INIS)

    Akiba, Miyuki.

    1996-01-01

    In a cooling device for a reactor container, a low pressure vessel is connected to an incondensible gas vent tube by way of an opening/closing valve. Upon occurrence of a loss of coolant accident, among steams and incondensible gases contained in the reactor container, steams are cooled and condensed in a heat exchanger. The incondensible gases are at first discharged from the heat exchanger to a suppression pool by way of the incondensible gas vent tube, but subsequently, they are stagnated in the incondensible gas vent tube to hinder heat exchanging and steam cooling and condensing effects in the heat exchanger thereby raising temperature and pressure in the reactor. However, if the opening/closing valve is opened when the incondensible gases are stagnated in the incondensible gas vent tube, since the incondensible gases stagnated in the heat exchanger are sucked and discharged to the low pressure vessel, the performance of the heat exchanger is maintained satisfactorily thereby enabling to suppress elevation of temperature and pressure in the reactor container. (N.H.)

  10. Reactor water sampling device

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo.

    1992-01-01

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  11. The laminar flow tube reactor as a quantitative tool for nucleation studies: Experimental results and theoretical analysis of homogeneous nucleation of dibutylphthalate

    International Nuclear Information System (INIS)

    Mikheev, Vladimir B.; Laulainen, Nels S.; Barlow, Stephan E.; Knott, Michael; Ford, Ian J.

    2000-01-01

    A laminar flow tube reactor was designed and constructed to provide an accurate, quantitative measurement of a nucleation rate as a function of supersaturation and temperature. Measurements of nucleation of a supersaturated vapor of dibutylphthalate have been made for the temperature range from -30.3 to +19.1 degree sign C. A thorough analysis of the possible sources of experimental uncertainties (such as defining the correct value of the initial vapor concentration, temperature boundary conditions on the reactor walls, accuracy of the calculations of the thermodynamic parameters of the nucleation zone, and particle concentration measurement) is given. Both isothermal and the isobaric nucleation rates were measured. The experimental data obtained were compared with the measurements of other experimental groups and with theoretical predictions made on the basis of the self-consistency correction nucleation theory. Theoretical analysis, based on the first and the second nucleation theorems, is also presented. The critical cluster size and the excess of internal energy of the critical cluster are obtained. (c) 2000 American Institute of Physics

  12. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Y. J.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2008-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes.

  13. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Yoon, C.; Rhee, B. W.; Chung, B. D.; Cho, Y. J.; Kim, M. W.

    2008-01-01

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes

  14. Device for the selective positioning of a component on a tube plate

    International Nuclear Information System (INIS)

    1974-01-01

    The invention relates to a device for the selective positioning of a component on a tube plate. It particularly applies to the positioning of a guide tube head successively opposite all the tubes of the tube bundle of a nuclear reactor steam generator. The large number of tubes in the tube bundle of the steam generator in a pressure water nuclear power station must be checked periodically for any likely corrosion. This check is effected with a Foucault current probe which is inserted in each tube in turn and is connected to a probe signal processing unit. The probe is placed in a flexible guide tube brought in turn in front of each tube of the bundle to be checked. The invention concerns a device to move the opening of a tube guide for a Foucault current detector over the entire surface of the tube plate, thereby providing access to all the tubes whilst limiting the interventions to a single positioning and a single withdrawal of the apparatus for testing all the bundle. Between the two interventions at the beginning and end of the operation, all displacements are remote controlled from outside the dangerous radioacive area [fr

  15. Manipulator for inspection or repair of heat exchanger tubes, in particular in steam generators for nuclear reactors

    International Nuclear Information System (INIS)

    Gugel, G.

    1979-01-01

    The manipulator used to inspect or repair pipes in the steam generator chamber of a PWR can be introduced and removed through a penetration nozzle which can be sealed tightly by means of a blind flange. The front end of the manipulator carries a swivel arm which can be operated remotely to be moved in a plane parallel to the tube plate. The end of the swivel arm carries a holder for a mouthpiece which can be extended and retracted. This carrier can also be operated remotely so as to be aligned to the pipe orifices in a direction normal to the swivel plane of the swivel arm. The manipulator is supported in antifriction bearings in the penetration nozzle so as to be movable longitudinally. (DG) [de

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  17. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  18. Double wall steam generator tubing

    International Nuclear Information System (INIS)

    Padden, T.R.; Uber, C.F.

    1983-01-01

    Double-walled steam generator tubing for the steam generators of a liquid metal cooled fast breeder reactor prevents sliding between the surfaces due to a mechanical interlock. Forces resulting from differential thermal expansion between the outer tube and the inner tube are insufficient in magnitude to cause shearing of base metal. The interlock is formed by jointly drawing the tubing, with the inside wall of the outer tube being already formed with grooves. The drawing causes the outer wall of the inner tube to form corrugations locking with the grooves. (author)

  19. The Texts of the Instruments Concerning the Agency's Assistance to Pakistan in connection with the Establishment of a Nuclear Power Reactor Project

    International Nuclear Information System (INIS)

    1968-01-01

    The terms of the Supply Agreement between the Agency and the Governments of Pakistan and the United States of America, and of the Project Agreement between the Agency and the Government of Pakistan concerning the Agency's assistance to that Government in connection with the establishment of a nuclear power reactor project, are reproduced herein for the information of all Members. Both Agreements entered into force on 17 June 1968

  20. The Texts of the Instruments concerning the Agency's Assistance to Pakistan in connection with the Establishment of a Nuclear Power Reactor Project. A Second Supply Agreement

    International Nuclear Information System (INIS)

    1971-01-01

    As a sequel to the assistance which the Agency provided to the Government of Pakistan in connection with a nuclear power reactor project, a Second Supply Agreement has been concluded between the Agency and the Governments of Pakistan and the United States of America. This Agreement entered into force on 22 June 1971, and the text is reproduced herein for the information of all Members

  1. The Texts of the Instruments Concerning the Agency's Assistance to Pakistan in connection with the Establishment of a Nuclear Power Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-09-06

    The terms of the Supply Agreement between the Agency and the Governments of Pakistan and the United States of America, and of the Project Agreement between the Agency and the Government of Pakistan concerning the Agency's assistance to that Government in connection with the establishment of a nuclear power reactor project, are reproduced herein for the information of all Members. Both Agreements entered into force on 17 June 1968.

  2. The Texts of the Instruments connected with the Agency's Assistance to Argentina in Establishing a Research and Isotope Production Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-11-04

    The texts of the Title Transfer Agreement between the Agency and the Governments of Argentina and the United States of America, and of the Project Agreement between the Agency and the Government of Argentina, in connection with the Agency's assistance to that Government in establishing a research and isotope production reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 2 December 1964.

  3. The Texts of the Instruments connected with the Agency's Assistance to the Democratic Republic of the Congo in Continuing a Research Reactor Project. A Second Supply Agreement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1971-10-04

    As a sequel to the assistance which the Agency provided to the Government of the Democratic Republic of the Congo in connection with a research reactor project, a Second Supply Agreement has been concluded between the Agency and the Governments of the Congo and the United States of America. This Agreement entered into force on 15 April 1971, and the text is reproduced herein for the information of all Members.

  4. The Texts of the Instruments concerning the Agency's Assistance to Pakistan in connection with the Establishment of a Nuclear Power Reactor Project. A Second Supply Agreement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1971-10-08

    As a sequel to the assistance which the Agency provided to the Government of Pakistan in connection with a nuclear power reactor project, a Second Supply Agreement has been concluded between the Agency and the Governments of Pakistan and the United States of America. This Agreement entered into force on 22 June 1971, and the text is reproduced herein for the information of all Members.

  5. The Texts of the Instruments connected with the Agency's Assistance to the Democratic Republic of the Congo in Continuing a Research Reactor Project. A Second Supply Agreement

    International Nuclear Information System (INIS)

    1971-01-01

    As a sequel to the assistance which the Agency provided to the Government of the Democratic Republic of the Congo in connection with a research reactor project, a Second Supply Agreement has been concluded between the Agency and the Governments of the Congo and the United States of America. This Agreement entered into force on 15 April 1971, and the text is reproduced herein for the information of all Members

  6. The Texts of the Instruments connected with the Agency's Assistance to Mexico in Establishing a Research Reactor Project. A Second Supply Agreement

    International Nuclear Information System (INIS)

    1972-01-01

    As a sequel to the assistance which the Agency has provided to the Government of Mexico in connection with a research reactor project, a Second Supply Agreement has been concluded between the Agency and the Governments of Mexico and the United States of America. This Agreement entered into force on 4 October 1972, and the text is reproduced herein for the information of all Members

  7. The Texts of the Instruments connected with the Agency's Assistance to Mexico in Establishing a Research Reactor Project. A Second Supply Agreement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-11-17

    As a sequel to the assistance which the Agency has provided to the Government of Mexico in connection with a research reactor project, a Second Supply Agreement has been concluded between the Agency and the Governments of Mexico and the United States of America. This Agreement entered into force on 4 October 1972, and the text is reproduced herein for the information of all Members.

  8. The Texts of the Instruments connected with the Agency's Assistance to Argentina in Establishing a Research and Isotope Production Reactor Project

    International Nuclear Information System (INIS)

    1965-01-01

    The texts of the Title Transfer Agreement between the Agency and the Governments of Argentina and the United States of America, and of the Project Agreement between the Agency and the Government of Argentina, in connection with the Agency's assistance to that Government in establishing a research and isotope production reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 2 December 1964

  9. The Texts of the Instruments connected with the Agency's Assistance to Pakistan in Establishing a Research Reactor Project. A Third Supply Agreement

    International Nuclear Information System (INIS)

    1974-01-01

    As a sequel to the assistance which the Agency has provided to the Government of Pakistan in connection with a research reactor project, a Third Supply Agreement has been concluded between the Agency and the Governments of Pakistan and the United States of America. The Agreement entered into force on 14 June 1974 pursuant to Article V, and the text is reproduced herein for the information of all Members

  10. NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  11. Device for providing a leak-tight penetration for electric cables through a reactor vault roof

    International Nuclear Information System (INIS)

    Eyral, M.; Mahe, A.

    1979-01-01

    The device for providing a cable penetration through the vault roof of a liquid sodium cooled fast reactor comprises a vertical tube closed at the top end by a flange-plate. Electric cables connected to measuring and detecting instruments are passed through the flange-plate which is joined to the reactor vault roof in leak-tight manner and enclosed within a removable hood. At least one horizontal plate is mounted within the vertical tube and provided with orifices for the leak-tight passage of the cables. Cable storage reels are placed within the tube and can be locked in position or released by controlled mechanical means

  12. Effects of heat exchanger tubes on hydrodynamics and CO 2 capture of a sorbent-based fluidized bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lai, Canhai; Xu, Zhijie; Li, Tingwen; Lee, Andrew; Dietiker, Jean-François; Lane, William; Sun, Xin

    2017-12-01

    In virtual design and scale up of pilot-scale carbon capture systems, the coupled reactive multiphase flow problem must be solved to predict the adsorber’s performance and capture efficiency under various operation conditions. This paper focuses on the detailed computational fluid dynamics (CFD) modeling of a pilot-scale fluidized bed adsorber equipped with vertical cooling tubes. Multiphase Flow with Interphase eXchanges (MFiX), an open-source multiphase flow CFD solver, is used for the simulations with custom code to simulate the chemical reactions and filtered models to capture the effect of the unresolved details in the coarser mesh for simulations with reasonable simulations and manageable computational effort. Previously developed two filtered models for horizontal cylinder drag, heat transfer, and reaction kinetics have been modified to derive the 2D filtered models representing vertical cylinders in the coarse-grid CFD simulations. The effects of the heat exchanger configurations (i.e., horizontal or vertical) on the adsorber’s hydrodynamics and CO2 capture performance are then examined. The simulation result subsequently is compared and contrasted with another predicted by a one-dimensional three-region process model.

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  14. Heat exchanger tube tool

    International Nuclear Information System (INIS)

    Gugel, G.

    1976-01-01

    Certain types of heat-exchangers have tubes opening through a tube sheet to a manifold having an access opening offset from alignment with the tube ends. A tool for inserting a device, such as for inspection or repair, is provided for use in such instances. The tool is formed by a flexible guide tube insertable through the access opening and having an inner end provided with a connector for connection with the opening of the tube in which the device is to be inserted, and an outer end which remains outside of the chamber, the guide tube having adequate length for this arrangement. A flexible transport hose for internally transporting the device slides inside of the guide tube. This hose is long enough to slide through the guide tube, into the heat-exchanger tube, and through the latter to the extent required for the use of the device. The guide tube must be bent to reach the end of the heat-exchanger tube and the latter may be constructed with a bend, the hose carrying anit-friction elements at interspaced locations along its length to make it possible for the hose to negotiate such bends while sliding to the location where the use of the device is required

  15. Fuel element clusters for nuclear reactors

    International Nuclear Information System (INIS)

    Anthony, A.J.; Hutchinson, J.J.

    1975-01-01

    In the fuel element assembly for nuclear reactors the influence of temperature cycles upon the stability of the joints between the individual components, especially between the control rod guide tubes and the connecting rods and end plates, respectively, is reduced. For this purpose, the connection is designed as a bolted connection connecting, on the one hand, the guide tubes and guide bolts and, on the other hand, these two components and the end plates. Moreover, the materials of the guide tubes, bolts and end plates are selected so that their respective thermal expansion coefficients differ. The material which can be used for the end plates and the guide bolts is stainless steel and stainless steel plus inconel (nickel-chrome-iron alloy), respectively; for the guide tubes it is a zirconium alloy (zircaloy). In addition to some technical designs of the bolted connections the materials and lengths of the components are selected in such a way that the expansion path of the components held by a bolted connection is equal to that of the stressing part. (DG/RF) [de

  16. The text of the Agreement of 25 September 1980 between Cuba and the Agency for the application of safeguards in connection with the supply of a nuclear research reactor from the Union of Soviet Socialist Republics

    International Nuclear Information System (INIS)

    1982-08-01

    The full text of the agreement between Cuba and the International Atomic Energy Agency for the application of safeguards in connection with the supply of a nuclear research reactor from the USSR is presented

  17. The text of the Agreement of 12 June 1981 between Viet Nam and the Agency relating to the application of safeguards in connection with the supply of nuclear fuel for the Da Lat research reactor

    International Nuclear Information System (INIS)

    1982-02-01

    The full text of the agreement between Viet Nam and the International Atomic Energy Agency relating to the application of safeguards in connection with the supply of nuclear fuel for the DA LAT Research Reactor is presented

  18. Expansion lyre-shaped tube

    International Nuclear Information System (INIS)

    Andro, Jean.

    1973-01-01

    The invention relates the expansion lyre-shaped tube portions formed in dudgeoned tubular bundles between two bottom plates. An expansion lyre comprises at least two sets of tubes of unequal lengths coplanar and symmetrical with respect to the main tube axis, with connecting portions between the tubes forming said sets. The invention applies to apparatus such as heat exchangers, heaters, superheaters or breeders [fr

  19. Fabrication of seamless calandria tubes

    International Nuclear Information System (INIS)

    Saibaba, N.; Phanibabu, C.; Bhaskara Rao, C.V.; Kalidas, R.; Ganguly, C.

    2002-01-01

    Full text: Calandria tube is a large diameter, thin walled zircaloy-4 tube and is an important structural component of PHWR type of reactors. These tubes are lifetime components and remain during the full life of the reactor. Calandria tubes are classified as extremely thin walled tubes with a diameter to wall thickness ratio of around 96. Such thin walled tubes are conventionally produced by seam welded route comprising of extrusion of slabs followed by a series of hot and rolling passes, shaping into O-shape and eventual welding. An alternative and superior method of fabricating the calandria tubes, the seamless route, has been developed, which involves hot extrusion of mother blanks followed by three successive cold pilger reductions. Eccentricity correction of the extruded blanks is carried out on a special purpose grinding equipment to bring the wall thickness variation within permissible limits. Predominant wall thickness reductions are given during cold pilgering to ensure high Q-factor values. The texture in the finished tubes could be closely, controlled with an average f r value of 0.65. Pilgering parameters and tube guiding system have been specially designed to facilities rolling of thin walled tubes. Seamless calandria tubes have distinct advantages over welded tubes. In addition to the absence of weld, they are dimensionally more stable, lighter in weight and possess uniform grains with superior grain size. The cycle time from billet to finished product is substantially reduced and the product is amenable to high level of quality assurance. The most significant feature of the seamless route is its material recovery over welded route. Residual stresses measured in the tubes indicate that these are negligible and uniform along the length of the tube. In view of their superior quality, the first charge of seamless calandria tubes will be rolled into the first 500 MWe Pressurised Heavy Water Reactor at Tarapur

  20. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor; Etude des consequences de la rupture d'un tube de force dans la cuve d'un reacteur modere a l'eau lourde et refroidi au gaz

    Energy Technology Data Exchange (ETDEWEB)

    Hareux, F; Roche, R; Vrillon, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [French] L'eclatement d'un tube de force dans la cuve d'un reacteur de puissance modere a l'eau lourde et refroidi par un gaz sous pression est un accident qui a ete etudie experimentalement a propos d'EL-4. Un premier essai a l'echelle 1 ayant montre que l'eclatement d'un tube ne provoque pas celui des tubes voisins, des essais relatifs a la tenue de la cuve ont ete effectues sur maquettes a echelle tres reduite (l/lO). Il a ete trouve que la cuve peut supporter plusieurs eclatements de tubes sans deformations notables. La pression transitoire dans la cuve a une allure oscillatoire amortie dont le maximum (pression de pic) a fait l'objet d'une etude experimentale detaillee. Il apparait que les parametres essentiels influant sur cette pression sont: la pression du gaz contenu dans le tube de force, le volume du gaz qui participe a l'eclatement, la flexibilite de la cuve, la masse d'air empoisonnee dans la cuve, la nature du gaz explosant. Une methode generale d'estimation des pics de pression dans

  1. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

    International Nuclear Information System (INIS)

    Scheuer, A.; Gutsmiedl, E.

    1999-01-01

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256 deg. C and 250 deg. C. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was take into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256 deg. C and 150 deg. C to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to take into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼ 1x10 22 n/cm 2 was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak before rupture

  2. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

    International Nuclear Information System (INIS)

    Gutsmiedl, Erwin

    2001-01-01

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256degC and 250degC. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was taken into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256degC and 150degC to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to taken into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼1·10 22 n/cm 2 was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak before rupture criteria of

  3. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  4. Prospects for stronger calandria tubes

    International Nuclear Information System (INIS)

    Ells, C.E.; Coleman, C.E.; Hosbons, R.R.; Ibrahim, E.F.; Doubt, G.L.

    1990-12-01

    The CANDU calandria tubes, made of seam welded and annealed Zircaloy-2, have given exemplary service in-reactor. Although not designed as a system pressure containment, calandria tubes may remain intact even in the face of pressure tube rupture. One such incident at Pickering Unit 2 demonstrated the economic advantage of such an outcome, and a case can be made for increasing the probability that other calandria tubes would perform in a similar fashion. Various methods of obtaining stronger calandria tubes are available, and reviewed here. When the tubes are internally pressurized, the weld is the weak section of the tube. Increasing the oxygen concentration in the starting sheet, and thickening the weld, are promising routes to a stronger tube

  5. The text of the instruments connected with the Agency's assistance to Argentina in establishing a research and isotope production reactor project

    International Nuclear Information System (INIS)

    1995-01-01

    The Agreement between the Republic of Argentina, the Federative Republic of Brazil, the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials and the International Atomic Energy Agency for the Application of Safeguards came into force on 4 March 1994. As a result of the coming into force of the aforesaid Agreement for Argentina, the application of safeguards under the Project Agreement of 2 December 1964 between Argentina and the IAEA in connection with the Agency's assistance to Argentina in establishing a research and isotope production reactor project has been suspended

  6. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  7. Tube plug

    International Nuclear Information System (INIS)

    Zafred, P. R.

    1985-01-01

    The tube plug comprises a one piece mechanical plug having one open end and one closed end which is capable of being inserted in a heat exchange tube and internally expanded into contact with the inside surface of the heat exchange tube for preventing flow of a coolant through the heat exchange tube. The tube plug also comprises a groove extending around the outside circumference thereof which has an elastomeric material disposed in the groove for enhancing the seal between the tube plug and the tube

  8. Analysis of steam-generator tube-rupture events combined with auxiliary-feedwater control-system failure for Three Mile Island-Unit 1 and Zion-Unit 1 pressurized water reactors

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1986-01-01

    A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx.63 K (approx.113 0 F) for TMI-1 and approx.44 K (approx.80 0 F) for Zion-1

  9. 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)

  10. Optimizing The Efficiency of a Dielectric Barrier Discharge Reactor for Removal of Nitric Oxides in Gas Phase

    International Nuclear Information System (INIS)

    Siti Aiasah Hashim; Wong, C.S.; Abas, M.R.

    2016-01-01

    A dielectric barrier discharge (DBD) reactor was built and used to remove nitric oxides in gas phase. In the preliminary work, it was found that the DBD reactor can used for direct processing of contaminated air stream. It was observed that if the applied energy is sufficiently high, reduction can overcome the oxidation process. The other characteristics that can affect the efficiency of the reactor are the processing flow rate, number of DBD tubes used and how the tubes are connected. The composition of the feed gas also plays important role. To improve the efficiency, more tubes were added and configured in combination of serial and parallel connections to achieve the best result. The reactor was found to be most efficient when using 6 tubes configured to have 2 sets of 3 tubes in series connected in parallel. The maximum flow rate that can be treated is 5 scfh. When operated with the optimum input voltage of 32 kV, the reactor can remove up to 80 % nitric oxide in the reduction mode. This means that the energy is sufficiently high to sustain the reduction mode and prevent further oxidation. (author)

  11. Gastrostomy Tube (G-Tube)

    Science.gov (United States)

    ... any of these problems: a dislodged tube a blocked or clogged tube any signs of infection (including redness, swelling, or warmth at the tube site; discharge that's yellow, green, or foul-smelling; fever) excessive bleeding or drainage from the tube site severe abdominal pain lasting ...

  12. REACTOR SHIELD

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  13. PROBLEMS IN THE TUBING/PACKER SYSTEM

    OpenAIRE

    Davorin Matanović; Mario Livaja

    1993-01-01

    When gas and oil wells are completed and produced or treated through the tubing connected to packer, there is a great number of problems to be solved. Changes in temperatures and pressures that occure during various operations ussually result in changes in tubing lengths or tubing to packer forces, depending on tubing to packer connections. This paper summarises some earlier papers and explains partly elaborated details. It also gives a complete approach to solve problems in uniform strings r...

  14. Tubing misconnections: normalization of deviance.

    Science.gov (United States)

    Simmons, Debora; Symes, Lene; Guenter, Peggi; Graves, Krisanne

    2011-06-01

    Accidental connection of an enteral system to an intravenous (IV) system frequently results in the death of the patient. Misconnections are commonly attributed to the presence of universal connectors found in the majority of patient care tubing systems. Universal connectors allow for tubing misconnections between physiologically incompatible systems. The purpose of this review of case studies of tubing misconnections and of current expert recommendations for safe tubing connections was to answer the following questions: In tubing connections that have the potential for misconnections between enteral and IV tubing, what are the threats to safety? What are patient outcomes following misconnections between enteral and IV tubing? What are the current recommendations for preventing misconnections between enteral and IV tubing? Following an extensive literature search and guided by 2 models of threats and errors, the authors analyzed case studies and expert opinions to identify technical, organizational, and human errors; patient-related threats; patient outcomes; and recommendations. A total of 116 case studies were found in 34 publications. Each involved misconnections of tubes carrying feedings, intended for enteral routes, to IV lines. Overwhelmingly, the recommendations were for redesign to eliminate universal connectors and prevent misconnections. Other recommendations were made, but the analysis indicates they would not prevent all misconnections. This review of the published case studies and current expert recommendations supports a redesign of connectors to ensure incompatibility between enteral and IV systems. Despite the cumulative evidence, little progress has been made to safeguard patients from tubing misconnections.

  15. Tube-in-shell heat exchangers

    International Nuclear Information System (INIS)

    Richardson, J.

    1976-01-01

    Tube-in-shell heat exchangers normally comprise a bundle of parallel tubes within a shell container, with a fluid arranged to flow through the tubes in heat exchange with a second fluid flowing through the shell. The tubes are usually end supported by the tube plates that separate the two fluids, and in use the tube attachments to the tube plates and the tube plates can be subject to severe stress by thermal shock and frequent inspection and servicing are required. Where the heat exchangers are immersed in a coolant such as liquid Na such inspection is difficult. In the arrangement described a longitudinally extending central tube is provided incorporating axially spaced cylindrical tube plates to which the opposite ends of the tubes are attached. Within this tube there is a tubular baffle that slidably seals against the wall of the tube between the cylindrical tube plates to define two co-axial flow ducts. These ducts are interconnected at the closed end of the tube by the heat exchange tubes and the baffle comprises inner and outer spaced walls with the interspace containing Ar. The baffle is easily removable and can be withdrawn to enable insertion of equipment for inspecting the wall of the tube and tube attachments and to facilitate plugging of defective tubes. Cylindrical tube plates are believed to be superior for carrying pressure loads and resisting the effects of thermal shock. Some protection against thermal shock can be effected by arranging that the secondary heat exchange fluid is on the tube side, and by providing a thermal baffle to prevent direct impingement of hot primary fluid on to the cylindrical tube plates. The inner wall of the tubular baffle may have flexible expansible region. Some nuclear reactor constructions incorporating such an arrangement are described, including liquid metal reactors. (U.K.)

  16. Switching overvoltage when disconnecting a combined 400 kVcable/overhead line with permanently connected shunt reactor

    DEFF Research Database (Denmark)

    Bak, Claus Leth; Søgaard, Kim

    2008-01-01

    consisting of overhead lines, crossbonded cable sections and shunt reactor has been created in PSCAD/EMTDC and verified against measurements with good results. Main focus has been put on the generation of switching overvoltages. It is shown that the generation of switching overvoltages is caused by slightly...... different resonant frequencies of the three phases which are reflected between the phases by mutual couplings in such a way that a low-frequent modulation appears in the phase voltages....

  17. Connected analysis nuclear-thermo-hydraulic of parallel channels of a BWR reactor using distributed computation; Analisis acoplado nuclear-termohidraulico de canales paralelos de un reactor BWR empleando computacion distribuida

    Energy Technology Data Exchange (ETDEWEB)

    Campos Gonzalez, Rina Margarita

    2007-07-15

    This work consists of the integration of three models previously developed which are described widely in Literature: model of the thermo-hydraulic channel, model of the modal neutronic and the model of the recirculation bows. The tool used for this connection of models is the PVM system, Parallel Virtual Machine that allowed paralleling the model by means of the concept of distributed computation. The purpose of making this connection of models is the one of obtaining a more complete tool than better represents the real configuration and the phenomenology of the nucleus of a BWR reactor, thus obtaining better results. In addition to maintaining the flexibility to improve the resulting model at any time, since the very complex or sophisticated models are difficult to improve being impossible to modify the equations they use and can include variables that are not of primary importance in the tackled problem or that mask relations among variables due to the excess of results. Also maintaining the flexibility for adding component of models or systems of the BWR reactor, all of this following the modeling needs. The Swedish Ringhals power plant was chosen to characterize the resulting connected model for counting on a Stability Benchmark that offers the opportunity to count on real plant data. Besides that in case 9 of cycle 14 of this Benchamark oscillations outside phase appeared, which are from great interest because the detection systems that register the average of the power of the nucleus do not detect them. Additionally in this work the model of the recirculation bows as an independent module is obtained in an individual way, since this model belongs to another work and works connected to the reactor vessel. The model of the recirculation bows is able to model several transients of interest, as it is shown in the Appendix A of this work, among which are found the tripping of recirculation pumps or the transference at low or high velocity of them. The scope of the

  18. Guide tube sleeve

    International Nuclear Information System (INIS)

    Attix, D.J.

    1983-01-01

    The invention increases the operating capacity of a nuclear reactor by causing a modification in the flow pattern of the coolant which enhances the coolant's effectiveness. The apparatus provides a thin-walled tubular sleeve closely surrounding but not attached to the exterior surface of a guide tube in a fuel assembly. The wall of the sleeve has tabs projecting outwardly into adjacent flow channels. The sleeve is attached to the wall of a cellular void through which passes the guide tube associated with said sleeve. The tabs increase the flow of water in the channel and thus increase the heat transfer

  19. Nuclear reactor container

    International Nuclear Information System (INIS)

    Shioiri, Akio.

    1992-01-01

    In a nuclear reactor container, a vent tube communication port is disposed to a pressure suppression pool at a position higher than the pool water therein for communication with an upper dry well, and the upper end opening of a dry well communication pipe is disposed at a position higher than the communication port. When condensate return pipeline is ruptured in the upper dry well, water in a water source pool is injected to the pressure vessel and partially discharged out of the ruptured port and a depressurization valve connected to the pressure vessel to the inside of the upper dry well. The discharged water stays in the upper dry well and, when the water level reaches the height of the vent tube communication port, it flows into the pressure suppression pool. Even in a state that the entire amount of water in the water source pool is supplied, since water does not reach the upper opening port of the dry well communication pipe, water does not flow into a lower dry well. Accordingly, the motor of a control rod drives disposed in the lower dry well can be prevented from submerging. The reactor core can be cooled more reliably, to improve the reliability of the pressure suppression function. (N.H.)

  20. Operating performance of CANDU pressure tubes

    International Nuclear Information System (INIS)

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

    1989-04-01

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

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

    International Nuclear Information System (INIS)

    Quinet, J.L.; Lannou, L.

    1978-01-01

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

  2. The Texts of the Instruments connected with the Agency's Assistance to Yugoslavia in Establishing a Research Reactor Project. The Third Supply Agreement. Amendments

    International Nuclear Information System (INIS)

    1975-01-01

    On 29 December 1972 the Agency and the Governments of the United States of America and Yugoslavia amended by letter of agreement Sections l(i), 4 and 5 of the Third Supply Agreement, concluded in connection with the Agency's assistance to Yugoslavia for the continuation of a research reactor project and reproduced in document INFCIRC/32/Add. 3, to provide for the transfer to Yugoslavia of up to a total net amount of 4800 grams of uranium-23 5 contained in uranium enriched up to approximately 70 per cent in the isotope uranium-235, and for payment by Yugoslavia within 20 days from the date of the invoice received from the Agency and by the Agency within 30 days from the date of the invoice received from the United States Atomic Energy Commission

  3. Numerical analysis of performance of steam reformer of methane reforming hydrogen production system connected with high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yin Huaqiang; Jiang Shengyao; Zhang Youjie

    2007-01-01

    Methane conversion rate and hydrogen output are important performance indexes of the steam reformer. The paper presents numerical analysis of performance of the reformer connected with high-temperature gas-cooled reactor HTR-10. Setting helium inlet flow rate fixed, performance of the reformer was examined with different helium inlet temperature, pressure, different process gas temperature, pressure, flow rate, and different steam to carbon ratio. As the range concerned, helium inlet temperature has remarkable influence on the performance, and helium inlet temperature, process gas temperature and pressure have little influence on the performance, and improving process gas flow rate, methane conversion rate decreases and hydrogen output increases, however improving steam to carbon ratio has reverse influence on the performance. (authors)

  4. A study concerning tritium concentration evolution in the moderator of a CANDU reactor connected on-line to a detritiation facility

    International Nuclear Information System (INIS)

    Bidica, Nicolae; Bornea, Anisia

    2005-01-01

    The present work is a theoretical study on the tritium concentration evolution in the CANDU reactor moderator connected on-line with a detritiation facility. This study is based on a calculation model which takes into account the evolution curve of the tritium concentration in the absence of detritiation process in both the moderator and SPTC of the Unit 1 CANDU reactor at Cernavoda NPP. This study leads to determination of the tritium concentration evolution in the moderator in the presence of the detritiation process for both a range of intake flows and initial concentration. Also, the intake flow change will be analyzed for a detritiation facility as a function of tritium initial concentration existing in the moderator in the case of a survey of the detritiation over a given period of time. The conclusions of this study were the following: - an optimum of the detritiation factor can be determined; - detritiation starts at a lower value for the tritium concentration in moderator which reduces the strain upon the detritiation facility and therefore the costs of its building, maintenance and operation. (authors)

  5. Design and analysis on super-critical water cooled power reactors

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki

    2005-01-01

    The Super-Critical Water Cooled Power Reactors (SCPR) is cooled by 25 MPa supercritical water of 280degC at reactor inlet and greater than 500degC at reactor outlet and directly connected with turbine/generators with high energy conversion efficiency. This corresponds to the deletion of recirculation system and steam-water separation system of BWR type reactors or of pressurizer and steam generator of PWR type reactors. In addition to the design study of the university of Tokyo, technology development of the SCPR for practical use has started under the collaboration of industry and academia since 2000. Mockup single tube and bundle tests for heat transfer/fluid flow characteristics of the design have been conducted with 3D heat transfer analysis. Materials compatible with coolant conditions for fuel cans and reactor internals are also assessed. Overall evaluation of the reactor concept is under way. (T. Tanaka)

  6. The Economical Application of Non-Destructive Testing to Reactor Components, Especially Jacket Tubing; Avantages Economiques du Controle Non Destructif des Pieces de Reacteurs, Notamment des Tubes de Gainage; Ehkonomicheskoe primenenie nedestruktivnykh ispytanij dlya reaktornykh komponentov, v chastnosti obolochechnykh trub; Aplicacion en Condiciones Economicas de Ensayos No Destructivos a las Piezas de los Reactores, en Especial a los Tubos de Revestimiento

    Energy Technology Data Exchange (ETDEWEB)

    Renken, C. J. [Metallurgy Division Argonne National Laboratory Argonne, IL (United States)

    1965-10-15

    The ideal reactor design, in addition to its other desirable characteristics, would require no non-destructive testing. This ideal, like others, will probably never be attained. In any reactor design where cost is an important factor, the question of whether components can be economically tested should be proposed at the same time that questions of fabricability are being considered. Some development of these points as well as a discussion of the importance of non- destructive tests in specification writing is included in this section. Responsibility also rests on the fabricator to use the help provided by non-destructive testing in maintaining quality in the product through various stages in the fabrication process, and to use the test results to indicate those steps in the process most likely to introduce defects in the component. Often it develops that non-destructive testing in earlier stages of component fabrication cannot be replaced economically, if at all, by inspection of the component in finished or semi-finished form. Examples are cited to illustrate this point, particularly with regard to tubing for fuel jacket and heat-exchanger applications. The application of various non-destructive tests during a tube-fabrication development programme is described in some detail. The fabrication and inspection costs for some tubing used for jacket applications by Argonne National Laboratory are compared. Although component inspection in finished form can be minimized by these procedures, it cannot in all cases be eliminated entirely. The economical testing of plates and tubes, especially the latter, is discussed in detail. The discussion is centred around components of stainless steel, Zircaloy, and certain refractory metal alloys. It is shown through various examples that although the use of radiography and penetrants may be useful or even essential steps in the testing, critical inspection of thin-wall tubing must usually be made by either an ultrasonic or an

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  8. Theroretical modelling of the plate-tubes coupling in the hydroelasticity of the perforated plates

    International Nuclear Information System (INIS)

    Dzhupanov, V.A.; Manoach, E.S.

    1983-01-01

    In the previous investigations on the perforated plate hydroelasticity the problem of the plates-tubes-liquid interaction in the process of the general structural vibration is stated. But the interaction of the vibrating plates with the tubes, passing through them, is taken into account considering the tubes only as absolutely rigid supports. This is one of the possible technical realizations. In the present article the case when the tubes are taking part in the plate motion (vibration) is studied. Two circular perforated plates are supported by the absolutely rigid wall of the modelled roundcircular reactor barrel. The distance between the plates is given. They are connected by tubes, passing through, and clamped into the perforation holes. The plates and the tubes are made by any elastic HOOKIAN material. The volume between the two plates and outwardly to the tubes, but intrinsically of the barrel is filled by ideal, compressible and heavy liquid. Evidently the liquid volume is multiconnected one. The free vibration of the whole system is considered with the purposes: i) to give a theoretical model of the plates-tubes-liquid interaction including governing equations and boundary conditions; ii) to trace the solution of the eigen-value problem for the modelled structure; iii) to underline the engineering sides of the modelling process. (orig./GL)

  9. SG tube identification

    International Nuclear Information System (INIS)

    Hoogstraten, P. van

    1994-01-01

    A ''Tracker'' system is described which is designed to identify any tube in a reactor steam generator quickly and safely. Occupational radiation doses to maintenance workers are reduced by using a Tracker and emergency down times are shortened. The system employs a television camera and light source in a stainless steel box with a large window. Both the camera and spotlight can be panned and tilted to reach any point on the tubesheet and are remotely controlled. An operator at a safe working distance can identify any tube visible on a real time video by comparison with the tubesheet pattern stored earlier in the computer memory. The identified tube can then be spotlighted and dealt with quickly by a maintenance worker inside the channel head. (UK)

  10. Caring for Your Percutaneous Nephrostomy Tube

    Science.gov (United States)

    ... to the nephrostomy tube for 15 seconds. 5. Disconnect the drainage bag from the tube. 6. Put the used bag aside. 7. With a new alcohol pad, swab the open end of the nephrostomy tube for 15 seconds. 8. Connect a new bag. 9. Secure the drainage bag ...

  11. Pump element for a tube pump

    DEFF Research Database (Denmark)

    2011-01-01

    The invention relates to a tube pump comprising a tube and a pump element inserted in the tube, where the pump element comprises a rod element and a first and a second non-return valve member positioned a distance apart on the rod element. The valve members are oriented in the same direction...... relative to the rod element so as to allow for a fluid flow in the tube through the first valve member, along the rod element, and through the second valve member. The tube comprises an at least partly flexible tube portion between the valve members such that a repeated deformation of the flexible tube...... portion acts to alternately close and open the valve members thereby generating a fluid flow through the tube. The invention further relates to a pump element comprising at least two non-return valve members connected by a rod element, and for insertion in an at least partly flexible tube in such tube...

  12. Managing a chest tube and drainage system.

    Science.gov (United States)

    Durai, Rajaraman; Hoque, Happy; Davies, Tony W

    2010-02-01

    Intercostal drainage tubes (ie, chest tubes) are inserted to drain the pleural cavity of air, blood, pus, or lymph. The water-seal container connected to the chest tube allows one-way movement of air and liquid from the pleural cavity. The container should not be changed unless it is full, and the chest tube should not be clamped unnecessarily. After a chest tube is inserted, a nurse trained in chest-tube management is responsible for managing the chest tube and drainage system. This entails monitoring the chest-tube position, controlling fluid evacuation, identifying when to change or empty the containers, and caring for the tube and drainage system during patient transport. This article provides an overview of indications, insertion techniques, and management of chest tubes. Copyright 2010 AORN, Inc. Published by Elsevier Inc. All rights reserved.

  13. Ceramic oxygen transport membrane array reactor and reforming method

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.; Christie, Gervase Maxwell; Robinson, Charles; Wilson, Jamie R.; Gonzalez, Javier E.; Doraswami, Uttam R.

    2016-11-08

    The invention relates to a commercially viable modular ceramic oxygen transport membrane reforming reactor configured using repeating assemblies of oxygen transport membrane tubes and catalytic reforming reactors.

  14. The text of the Agreement of 7 October 1983 between Cuba and the Agency for the application of safeguards in connection with the supply of a zero-power nuclear reactor from the Hungarian People's Republic

    International Nuclear Information System (INIS)

    1984-01-01

    The full text of the agreement of 7 October 1983 between Cuba and the Agency for the application of safeguards in connection with the supply of a zero-power nuclear reactor from the Hungarian People's Republic and to the nuclear material to be used therein to be supplied by the Union of Soviet Socialist Republics is presented

  15. In situ sampling for pressure tube deuterium concentration

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  16. Resonant behaviour of MHD waves on magnetic flux tubes. I - Connection formulae at the resonant surfaces. II - Absorption of sound waves by sunspots

    Science.gov (United States)

    Sakurai, Takashi; Goossens, Marcel; Hollweg, Joseph V.

    1991-01-01

    The present method of addressing the resonance problems that emerge in such MHD phenomena as the resonant absorption of waves at the Alfven resonance point avoids solving the fourth-order differential equation of dissipative MHD by recourse to connection formulae across the dissipation layer. In the second part of this investigation, the absorption of solar 5-min oscillations by sunspots is interpreted as the resonant absorption of sounds by a magnetic cylinder. The absorption coefficient is interpreted (1) analytically, under certain simplifying assumptions, and numerically, under more general conditions. The observed absorption coefficient magnitude is explained over suitable parameter ranges.

  17. Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

    International Nuclear Information System (INIS)

    Bucholz, J.A.

    2000-01-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source

  18. Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    2000-07-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

  19. Sealed ion accelerator tubes (survey)

    International Nuclear Information System (INIS)

    Voitsik, L.R.

    1985-01-01

    The first publications on developing commercial models of small-scale sealed accelerator tubes in which neutrons are generated appeared in the foreign press in 1954 to 1957; they were very brief and were advertising-oriented. The tubes were designed for neutron logging of oil wells instead of ampule neutron sources (Po + Be, Ra + Be). Later, instruments of this type began to be called neutron tubes from the resulting neutron radiation that they gave off. In Soviet Union a neutron tube was developed in 1958 in connection with the development of the pulsed neutron-neutron method of studying the geological profile of oil wells. At that time the tube developed was intended, in the view of its inventors, to replace standard isotope sources with constant neutron yield. A fairly detailed survey of neutron tubes was made in the studies. 8 refs., 8 figs

  20. Heat exchanger with layers of helical tubes provided with improved tube supports

    International Nuclear Information System (INIS)

    Carnoy, M.; Mathieu, B.; Renaux, C.

    1986-01-01

    The present heat exchanger comprises coaxial layers of helically wound tubes; these tubes are supported by support plates, each comprising a row of perforations through which the tubes of a same layer pass. Truncated sleeves are in compression around the tubes within the perforations and mounted on the support plates. Pins fix the plates of different layers together against transverse movement but allowing radial movement. The present invention finds an application with nuclear reactor steam generators [fr

  1. Mechanical and Radiological Characterization of Different parts of an Irradiation Coolant Channel Tube from Atucha I Nuclear Plant; Caracterizacion Mecanica y Radiologica de Partes de Canales Refrigerantes Irradiados Extraidos del Reactor de la Central Nuclear Atucha I

    Energy Technology Data Exchange (ETDEWEB)

    Piquin, Ruben [Instituto Balseiro, Universidad Nacional de Cuyo, Centro Atomico Bariloche, Universidad Nacional de Mar del Plata (Argentina)

    2001-07-01

    The widespread replacement of reactor internals has generated a substantial volume of active material. It is essential to work with these components at least in a partial way before the next planned stop, which will take place during the second semester of the year 2002. Due to the fact that the reactor internals pool and the storage pool for irradiated nuclear fuel have limited capacities, it has been proposed to compact an experimental shift of 50 irradiated coolant channels, that are currently placed in storage pools. Basically the processed waste will be put in baskets at the bottom pools.The alternative choice proposes to divide an irradiation coolant channel tube into different parts: stainless steel section, zircaloy-4 section and stainless steel section with hardened zones with cobalt alloys named Estelite-6. The person in charge has already planned the constructive and operative solutions but the mechanical characterization of the different parts of the channel tube is necessary in order to dimension the compaction tool needed for the semi-industrial installation.In the present special report, two well-differentiated actions will be described. The necessary compacted strength of the irradiation coolant channel tube will be estimated for the stainless steel section and the zircaloy-4 section starting from experiment with unirradiated material and considering effects of radiation damage and hydrides on the ductility.These results will be used to design the necessary compacted tools for the semi-industrial installation. The necessary equipment for the radiological characterization of the different material sections already specified will be described and the most important emitting particles of radiation that could be detected will be mentioned. Also the decontamination process to use including the radiological characterization of every stage of the process will be described in order to establish the decontamination factor. Finally the most important

  2. Pressurized water reactor flow arrangement

    International Nuclear Information System (INIS)

    Gibbons, J.F.; Knapp, R.W.

    1980-01-01

    A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

  3. Numerical analysis on the calandria tubes in the moderator of a heavy water reactor using OpenFOAM and other codes

    International Nuclear Information System (INIS)

    Chang, S.M.; Kim, H.T.

    2013-01-01

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors. (authors)

  4. Self operation type reactor scram device

    International Nuclear Information System (INIS)

    Saito, Makoto; Gunji, Minoru.

    1992-01-01

    A control rod having neutron absorbers therein is held by a curie point electromagnet by way of a control rod extension shaft. The electromagnet is suspended from a vertically movable driving shaft in an upper guide tube. Then, a heater is disposed at the lower portion in the inner side of the upper guide tube. Upon a function confirmation test, the electromagnet is at first pulled up to the inside of the upper guide tube. Subsequently, the electromagnet is heated by the heater by a temperature higher than the curie point of the temperature sensing magnetic material. If the function is normal, armature connected to the control rod extension tube is separated. With such a constitution, the electromagnetic portion is isolated from a coolant main stream, thereby enabling to avoid the cooling effect by the stream of coolants. Accordingly, the operation test for confirming the integrity of the function of the curie point electromagnet can be conducted while placing the electromagnet in the reactor core as it is during actual reactor operation. (I.N.)

  5. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  6. Tube closure device, especially for sample irradiation

    International Nuclear Information System (INIS)

    Klahn, F.C.; Nolan, J.H.; Wills, C.

    1979-01-01

    Device for closing the outlet of a bore and temporarily locking components in this bore. Specifically, it concerns a device for closing a tube containing a set of samples for monitoring irradiation in a nuclear reactor [fr

  7. Transient Effects in Fischer-Tropsch Reactor with a Fixed Bed of Catalyst Particles

    Directory of Open Access Journals (Sweden)

    I. V. Derevich

    2015-01-01

    Full Text Available Based on analysis of small temperature disturbances in the Fischer-Tropsch reactor with a fixed bed of catalyst particles various scenarios of thermal instability were investigated. There are two possible scenarios of thermal instability of the reactor. First, thermal explosion may occur due to growth of temperature disturbances inside a catalytic granule. Second scenario connected with loss of thermal stability as a result of an initial increase in temperature in the reactor volume. The boundaries of thermal stability of the reactor were estimated by solving the eigenvalue problems for spherical catalyst particles and cylindrical reactor. Processes of diffusional resistance inside the catalytic granule and heat transfer from wall of the reactor tube are taken into account. Estimation of thermal stability area is compared with the results of numerical simulation of behavior of temperature and concentration of synthesis gas.

  8. Analysis of the effect of tube arrangement and inclination on pressure drop in an intermediate heat exchanger of liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    ChoiI, Seok Ki; Choi, Il Kon; Nam, Ho Yun; Choi, Jong Hyeun [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    An experimental study on the effect of tube arrangement and inclination on the pressure drop in the intermediate heat exchanger is performed. Measurements are made for pressure drop in the triangular and rotated triangular tue arrays whose inclined angles are 30, 45, 60, 75 and 90 degrees. The pitch to tube diameter ratio is 1.6 and the range of Reynolds number based on the free stream velocity and tube diameter is 870-64,000. The experimental results show that the magnitude of dimensionless pressure drop increases with the inclined angle and decreases significantly when the inclined angle is less than 45 degree. The previous correlations are evaluated using the experimental data. The ESDU correlation agrees well with the present data for the triangular arrays. But some discrepancies are observed for the rotated triangular arrays when the inclined angles are 45 and 30 degrees. The Idel'chik correlation generally agrees well with the measured data for the rotated triangular arrays except for inclined angle of 30 degree. The Idel'chik correlation needs modification for the triangular arrays. The modified Idel'chik correlation agrees well with the measure data within 10%. 32 refs., 59 figs., 11 tabs. (Author)

  9. Behavior of concentrically loaded CFT braces connections

    Directory of Open Access Journals (Sweden)

    Maha M. Hassan

    2014-03-01

    Full Text Available Concrete filled tubes (CFTs composite columns have many economical and esthetic advantages, but the behavior of their connections is complicated. Through this study, it is aimed to investigate the performance and behavior of different connection configurations between concrete filled steel tube columns and bracing diagonals through an experimental program. The study included 12 connection subassemblies consisting of a fixed length steel tube and gusset plate connected to the tube end with different details tested under half cyclic loading. A notable effect was observed on the behavior of the connections due to its detailing changes with respect to capacity, failure mode, ductility, and stress distribution.

  10. Connecting ring and process to fix heaters in a pressure vessel by means of these rings

    International Nuclear Information System (INIS)

    Bailleul, G.; Caloine, P.; Coville, P.

    1984-01-01

    The invention can applies to the installation of heaters for nuclear reactor pressurizer or to the installation of any kind of reheaters by means of electric resistances when these reheaters have to work under important pressures. The connecting ring is made of a single metallic piece, two coaxial tubes joined each other by a skirt nearly radial; the skirt joins an end of the outer cylindrical tube and an intermediate zone of the inner cylindrical tube. The invention concerns also a heater provided with such a connecting ring, substituted for a part of its metallic envelope, and a process of fastening of these heaters on a pressure vessel. The description given in the frame of a pressurizer applies to the case of a gas reheater or to a reheater for liquid under pressure such as liquid sodium in a tank [fr

  11. Feeding Tubes

    Science.gov (United States)

    ... feeding therapies have been exhausted. Please review product brand and method of placement carefully with your physician ... Total Parenteral Nutrition. Resources: Oley Foundation Feeding Tube Awareness Foundation Children’s Medical Nutrition Alliance APFED’s Educational Webinar ...

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  13. Mechanical characterization and modeling of SiCF/SiC composite tubes

    International Nuclear Information System (INIS)

    Rohmer, E.

    2013-01-01

    This work is part of the development of the 4. generation of nuclear reactors. It relates more precisely to the composite portion of the sandwich type tubular cladding considered by the CEA for RNR-NA/Gaz type reactors. The texture is formed by a braiding technique and the study focuses on interlocks braided composite. These relatively new structures require extensive mechanical characterization. Two experimental protocols were developed to conduct tensile and internal pressure tests on tubes. Three different textures have been characterized. In addition, a multi-scale model was developed to connect the microstructure of the tube to its mechanical properties. This model is validated for the elastic behavior of a characterized texture. A first approach to the damage in the structure is proposed and a possible improved protocol is discussed. (author) [fr

  14. Analysis of autofrettaged metal tubes

    International Nuclear Information System (INIS)

    Malik, M. Afzaal; Khan, Muddasar; Rashid, Badar; Khushnood, Shahab

    2007-01-01

    Thick-walled cylinders are widely used as compressor cylinders, pump cylinders, high pressure tubing, process reactors and vessels, nuclear reactors, isostatic vessels and gun barrels. In practice, cylinders are generally subjected to sudden and frequently drastic pressure fluctuations, such as the pressure generated in a gun barrel upon the firing of the weapon, pressure reversals in pump cylinders or in process reactors employing high-pressure piping, necessitating enhanced strength of such cylinders. A process for enhancing the strength of thick-walled cylinders has been in service, and is referred to as 'autofrettage'. It extends the service life of the cylinder. The autofrettage is achieved by increasing elastic strength of a cylinder with various methods such as hydraulic pressurization, mechanical swaging, or by utilizing the pressure of a powder gas. This research work deals with the hydraulic and mechanical autofrettage of metal tubes with the objective to attain enhanced strength. Five metal tubes are taken randomly for analysis purpose. The experimental data for five metal tubes is obtained to analyze the behavior of different parameters used during, before, and after autofrettage process. For this research, two-stage autofrettage is taken into consideration. The modeling of the metal tube is carried out in WildFire-ProEngineering, and for analysis purpose, finite element software ANSYS7 and COSMOS are used. The graphical analysis of swage autofrettage is carried out using MATLAB7. The results are validated using available experimental and numerical data. (author)

  15. Tube in shell heat exchangers

    International Nuclear Information System (INIS)

    Hayden, O.; Willby, C.R.; Sheward, G.E.; Ormrod, D.T.; Firth, G.F.

    1980-01-01

    An improved tube-in-shell heat exchanger to be used between liquid metal and water is described for use in the liquid metal coolant system of fast breeder reactors. It is stated that this design is less prone to failures which could result in sodium water reactions than previous exchangers. (UK)

  16. Experience in utilizing research reactors in Yugoslavia

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J.; Raisic, N. [Boris Kidric Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia); Copic, M.; Gabrovsek, Z. [Jozef Stefan Institute Ljubljana (Yugoslavia)

    1972-07-01

    The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

  17. Experience in utilizing research reactors in Yugoslavia

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.; Raisic, N.; Copic, M.; Gabrovsek, Z.

    1972-01-01

    The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

  18. Importance of crevices formed between tubes and tube plate for the operational behaviour of heat exchangers

    International Nuclear Information System (INIS)

    Achten, N.; Herbsleb, G.; Wieling, N.

    1986-01-01

    It must be guaranteed by construction and manufacture of heat exchangers that primary and secondary medium are completely separated from each other. When this requirement is fullfilled, the operational use of heat exchangers can be impaired by corrosion reactions within the crevice formed between tube and tube plate which may result in corrosion damage. The various techniques which are in use to connect tubes and tube plate and which are described in the present report, must be valued with respect to the tightness of the connection as well as to the formation of crevices between tubes and tube plate. Corrosion resistant copperbase alloys and stainless steels are the most important materials which are in use for the construction of heat exchangers. The mechanisms of crevice corrosion with unalloyed and low alloy carbon steels, stainless steels, and mixed connections between tube and tube plate with these materials are described in detail. Crevice corrosion may be caused also by the formation of galvanic cells between materials of differing electrochemical response. Furthermore, the concentration of aggressive media in crevices between tubes and tube plate can lead to corrosion damage of heat exchanger tubes. For the service operation of heat exchangers without any hazard of corrosion damage in crevices between tubes and tube plate, such crevices must be avoided by proper construction and manufacture. As a model for suitable measures to avoid crevices, the manufacture of steam generators for PWR's is described. (orig.) [de

  19. The Mashups of YouTube

    DEFF Research Database (Denmark)

    Simonsen, Thomas Mosebo

    2013-01-01

    This article focuses on YouTube mashups and how we can understand them as a specific subgenre on YouTube. The Mashups are analysed as audiovisual recontextualizations that are given new meaning, e.g., via collaborative social communities or for individual promotional purposes. This is elaborated......, but rather in its social and communicative abilities within the YouTube community. This leads to the article’s overall argument that the main characteristic of the YouTube Mashup can be explained in terms of connectivity. It is argued that Mashups reveal a double articulation of connectivity; one...... that involves the social mechanisms of the Mashups, and another mode, which concerns the explicit embedding of structural connectivity that accentuates the medium-specific infrastructure of YouTube. This double articulation of connectivity is furthermore elaborated on by including Grusin and Bolter’s concept...

  20. Improvements in gas supply systems for heavy-water moderated reactors

    International Nuclear Information System (INIS)

    Aubert, G.; Hassig, J.M.; Laurent, N.; Thomas, B.

    1964-01-01

    In a heavy-water moderated reactor cooled by pressurized gas, an important problem from the point of view, of the reactor block and its economics is the choice of the gas supply system. In the pressure tube solution, the whole of the reactor block structure is at a relatively low temperature, whereas the gas supply equipment is at that of the gas, which is much higher. These parts, through which passes the heat carrying fluid have to present as low a resistance as possible to it so as to avoid costly extra blowing power. Finally, they may only be placed in the reactor block after it has been built; the time required for putting them in position should therefore not be too long. The work reported here concerns the various problems arising in the case of each channel being supplied individually by a tube at the entry and the exit which is connected to a main circuit made up of large size collectors. This individual tubing is sufficiently flexible to absorb the differential expansion and the movement of its ends without stresses or prohibitive reactions being produced; the tubing is also of relatively short length so as to reduce the pressure head of the pressurized gas outside the channels; the small amount of space taken up by the tubing makes it possible to assemble it in a manner which is satisfactory from the point of view both of the time required and of the technical quality. (authors) [fr

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  2. Advanced pressure tube sampling tools

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. Fuel handling system of nuclear reactor plants

    International Nuclear Information System (INIS)

    Faulstich, D.L.

    1991-01-01

    This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

  4. Benchmark simulation of turbulent flow through a staggered tube bundle to support CFD as a reactor design tool. Part 2. URANS CFD simulation

    International Nuclear Information System (INIS)

    Ridluan, Artit; Tokuhiro, Akira

    2008-01-01

    In Part II, we described the unsteady flow simulation and proposed a modification of a traditional turbulence flow model. Computational fluid dynamics (CFD) simulations of an isothermal, fully periodic flow across a tube bundle using unsteady Reynolds averaged Navier-Stokes (URANS) equations, with turbulence models such as the Reynolds stress model (RSM) were investigated at a Reynolds number of 1.8x10 4 , based on the tube diameter and inlet velocity. As noted in Part I, CFD simulation and experimental results were compared at five positions along (x,y) coordinates. The steady RANS simulation showed that four diverse turbulence models were efficient for predicting the Reynolds stresses, and generally, SRANS results were marginal to poor, using a consistent evaluation terminology. In the URANS simulation, we modeled the turbulent flow field in a manner similar to the approach used for large eddy simulation (LES). The time-dependent URANS results showed that the simulation reproduces the dynamic stability as characterized by transverse oscillatory flow structures in the near-wake region. In particular, the inclusion of terms accounting for the time scales associated with the production range and dissipation rate of turbulence generates unsteady statistics of the mean and fluctuation flow. In spite of this, the model implemented produces better agreement with a benchmark data set and is thus recommended. (author)

  5. Steam generator tube extraction

    International Nuclear Information System (INIS)

    Delorme, H.

    1985-05-01

    To enable tube examination on steam generators in service, Framatome has now developed a process for removing sections of steam generator tubes. Tube sections can be removed without being damaged for treating the tube section expanded in the tube sheet

  6. Nuclear reactor spacer assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.; Groves, M.D.

    1979-01-01

    A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

  7. Fuel cladding tube leak detection device

    International Nuclear Information System (INIS)

    Naito, Makoto.

    1992-01-01

    The device of the present invention can detect even a minute leakage or a continuous leakage during reactor operation. That is, the device of the present invention comprises a detector for analyzing nuclides of gases incorporated in a gas waste processing system, and a calculation device connected to the detector and detecting leakage from a fuel cladding tube by calculation for variation coefficient of long-life nuclides. By using theses devices, radioactivity contained in gases incorporated in the gas waste processing system is analyzed for the nuclides. Among the analized nuclides, if the amount of the long-life nuclides exceeds a predetermined value, it is judged as leakage of the fuel cladding tube. For example, the long-life nuclides include Xe-133. The device of the present invention can certainly detect occurrence of leakage even when it is minute or continues leakage. Accordingly, countermeasures can be taken in an early stage, thereby enabling to contribute improvement for the safety of a nuclear power plant. (I.S.)

  8. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

    1982-01-01

    A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

  9. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-09-01

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

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

    International Nuclear Information System (INIS)

    1998-08-01

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

  12. Fuel element for nuclear reactors

    International Nuclear Information System (INIS)

    Cadwell, D.J.

    1982-01-01

    The invention concerns a fuel element for nuclear reactors with fuel rods and control rod guide tubes, where the control rod guide tubes are provided with flat projections projecting inwards, in the form of local deformations of the guide tube wall, in order to reduce the radial play between the control rod concerned and the guide tube, and to improve control rod movement. This should ensure that wear on the guide tubes is largely prevented which would be caused by lateral vibration of the control rods in the guide tubes, induced by the flow of coolant. (orig.) [de

  13. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  14. Reactor and method for production of nanostructures

    Science.gov (United States)

    Sunkara, Mahendra Kumar; Kim, Jeong H.; Kumar, Vivekanand

    2017-04-25

    A reactor and method for production of nanostructures, including metal oxide nanowires or nanoparticles, are provided. The reactor includes a regulated metal powder delivery system in communication with a dielectric tube; a plasma-forming gas inlet, whereby a plasma-forming gas is delivered substantially longitudinally into the dielectric tube; a sheath gas inlet, whereby a sheath gas is delivered into the dielectric tube; and a microwave energy generator coupled to the dielectric tube, whereby microwave energy is delivered into a plasma-forming gas. The method for producing nanostructures includes providing a reactor to form nanostructures and collecting the formed nanostructures, optionally from a filter located downstream of the dielectric tube.

  15. Support tube of in-core instruments

    International Nuclear Information System (INIS)

    Suzumura, Takeshi; Saito, Shozo; Yasuda, Tetsuo; Shirosaki, Kiyotaka.

    1975-01-01

    Object: To permit satisfactory output measurement by preventing the bending of a in-core instrument tube within a reactor due to vibrations by means of a spring and thereby preventing mechanical damage of an adjacent fuel channel box. Structure: At a corner of a channel box of a fuel assembly, a in-core instrument tube is arranged along a channel box and has its surface provided with a plurality of removable leaf springs arranged in the direction of axis of the in-core instrument tube and each having an arcular tip. Thus, when the in-core instrument tube is inserted into the reactor, the arcular tip portions of the leaf springs are brought into plane contact with the corner of the channel box so that the in-core instrument tube is elastically supported on the channel box. Thus, there is no possibility of causing damage to the adjacent fuel channel box. (Kamimura, M.)

  16. Ear Tubes

    Science.gov (United States)

    ... of the ear drum or eustachian tube, Down Syndrome, cleft palate, and barotrauma (injury to the middle ear caused by a reduction of air pressure, ... specialist) may be warranted if you or your child has experienced repeated ... fluid in the middle ear, barotrauma, or have an anatomic abnormality that ...

  17. Reactor fuel exchanging facility

    International Nuclear Information System (INIS)

    Kubota, Shin-ichi.

    1981-01-01

    Purpose: To enable operation of an emergency manual operating mechanism for a fuel exchanger with all operatorless trucks and remote operation of a manipulator even if the exchanger fails during the fuel exchanging operation. Constitution: When a fuel exchanging system fails while connected to a pressure tube of a nuclear reactor during a fuel exchanging operation, a stand-by self-travelling truck automatically runs along a guide line to the position corresponding to the stopping position at that time of the fuel exchanger based on a command from a central control chamber. At this time the truck is switched to manual operation, and approaches the exchanger while being monitored through a television camera and then stops. Then, a manipurator is connected to the emergency manual operating mechanism of the exchanger, and is operated through necessary emergency steps by driving the snout, the magazine, the grab or the like in the exchanger in response to the problem, and necessary operations for the emergency treatment are thus performed. (Sekiya, K.)

  18. Ultrasonic measurement of gap between calandria tube and liquid injection shutdown system tube in PHWR

    International Nuclear Information System (INIS)

    Kim, Tae Ryong; Sohn, Seok Man; Lee, Jun Shin; Lee, Sun Ki; Lee, Jong Po

    2001-01-01

    Sag of CT or liquid injection shutdown system tubes in pressurized heavy water reactor is known to occur due to irradiation creep and growth during plant operation. When the sag of CT is big enough, the CT tube possibly comes in contact with liquid injection shutdown system tube (LIN) crossing beneath the CT, which subsequently may prevent the safe operation. It is therefore necessary to check the gap between the two tubes in order to confirm no contacts when using a proper measure periodically during the plant life. An ultrasonic gap measuring probe assembly which can be fed through viewing port installed on the calandria was developed and utilized to measure the sags of both tubes in a pressurized heavy water reactor in Korea. It was found that the centerlines of CT and LIN can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. But the measured gap data observed at the viewing port were actually not the data at the crossing point of CT and LIN. To get the actual gap between two tubes, mathematical modeling for the deflection curves of two tubes was used. The sags of CT and LIN tubes were also obtained by comparison of the present centerlines with the initial elevations at the beginning of plant operation. The gaps between two tubes in the unmeasurable regions were calculated based on the measurement data and the channel power distribution

  19. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  20. Refurbishment programme of the reactor and progress of work

    International Nuclear Information System (INIS)

    Astruc, J.M.

    1992-01-01

    During 20 years of operation, since its start-up the ILL there have been some problems, like ruptured heavy water collector, in the upper part of the reflector tank, replacement of all the beam tubes due to the evolution of the mechanical characteristics of the aluminium alloy under irradiation. Some days after regular shutdown for maintenance, an inspection of the internal elements of the reactor discovered cracks on the grids which ensure the regular flow of cooling water. The investigations showed that the cracks are due to a design fault, aggravated by the effects of mechanical fatigue on highly irradiated material. It was not possible to repair the cracked grid, and it had to be replaced. This involved the dismantling of the internals parts of the reactor tank. The reactor refurbishment programme was set up. It provides for the replacement of the reactor block, the coupling sleeves, the anti turbulence grids and the diffuser, and of the ancillary elements. The main items to be replaced are: the reactor block consisting of the reactor vessel and its cover, known as the 'upper structure'; the heavy water collectors; connecting sleeves between the reactor block and the flanges of the various beam tubes. These three items constitute the primary circuit in the swimming pool. It is also planned to replace some internal parts of the reactor tank, such as the beam-tubes, the grid and diffuser and the chimney. Some parts of the present reactor, which are not at the end of their life, would be reused, for instance the two cold sources, the safety rods, and some other pieces. The parts replaced would be cut up and packaged in accordance with current standards and disposed of. All items are in principle to be replaced by identical equipment. This concerns in particular performance, mechanical characteristics and the choice of materials. The replacement of the reactor block necessitates a complete dismantling of the equipment in the reactor block, and of the structures in

  1. Transient CFD studies on multiple jets issuing from injection tube

    International Nuclear Information System (INIS)

    Kumawat, Ganesh Lal; Kansal, Anuj Kumar; Maheshwari, Naresh Kumar; Rama Rao, A.

    2016-01-01

    Shut down system 2 of Advanced Heavy Water reactor incorporates the injection of liquid poison into moderator through injection tubes. The injection tubes consist of several holes distributed axially and circumferentially. Investigation of the poison jet progression and spreading from the holes of injection tube is important aspect of determining negative reactivity injection rate. This paper presents the CFD simulation to investigate poison jet progression and its spreading from the holes of injection tube. (author)

  2. Chest tube insertion

    Science.gov (United States)

    Chest drainage tube insertion; Insertion of tube into chest; Tube thoracostomy; Pericardial drain ... Be careful there are no kinks in your tube. The drainage system should always sit upright and be placed ...

  3. Depressurisation study of a tank-tubing assemble

    International Nuclear Information System (INIS)

    Freitas, R.L.

    1975-08-01

    The depressurisation of a nuclear reactor following the rupture of the primary coolant circuit is studied, using the simple analogy of the rupture of the tubing connected to a pressurised tank. The method of characteristics has been used in this theoretical analysis. The partial differential equations of conservation of mass, momentum and energy forming a hyperbolic system and defining real characteristic directions, allow the integration of these equations to be carried out along these directions. The method allows calculations to be made of the pressure, temperature, density and fluid velocity in the reactor circuit at any time after the beginning of depressurisation. A computer code MECA I has been written to calculate all the parameters after the rupture for any point in the coolant tubing. The computers used for these calculations were the IBM 360/40 and 370/145 and the Burroughs 6700. In this preliminary study, the simplest case of a system using a perfect gas coolant has been used [pt

  4. Mass production of CNTs using CVD multi-quartz tubes

    Energy Technology Data Exchange (ETDEWEB)

    Yousef, Samy; Mohamed, Alaa [Dept. of Production Engineering and Printing Technology, Akhbar Elyom Academy, Giza (Egypt)

    2016-11-15

    Carbon nanotubes (CNTs) have become the backbone of modern industries, including lightweight and heavy-duty industrial applications. Chemical vapor deposition (CVD) is considered as the most common method used to synthesize high yield CNTs. This work aims to develop the traditional CVD for the mass production of more economical CNTs, meeting the growing CNT demands among consumers by increasing the number of three particular reactors. All reactors housing is connected by small channels to provide the heat exchange possibility between the chambers, thereby decreasing synthesis time and reducing heat losses inside the ceramic body of the furnace. The novel design is simple and cheap with a lower reacting time and heat loss compared with the traditional CVD design. Methane, hydrogen, argon, and catalyzed iron nanoparticles were used as a carbon source and catalyst during the synthesis process. In addition, CNTs were produced using only a single quartz tube for comparison. The produced samples were examined using XRD, TEM, SEM, FTIR, and TGA. The results showed that the yield of CNTs increases by 287 % compared with those synthesized with a single quartz tube. Moreover, the total synthesis time of CNTs decreases by 37 % because of decreased heat leakage.

  5. Irradiation of reactor materials within projects VISA-2 and 3, 3. Procedure for construction and testing the capsules and test-tubes - Phase I (Parts I and II) Part II; Ozracivanje reaktorskih materijala po projektima VISA-2 i 3, 3. Osvajanje postupka izrade i ispitivanja kapsula i kenera VISA - I faza (I i II deo), II deo

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1964-02-15

    Experiments concerned with Projects VISA-2 and 3 demand construction of hermetization test-tubes, irradiation capsules, experimental devices and reactor channels as well as welding of fuel element claddings. For this purpose special materials as stainless steels, aluminium alloys, pure aluminium, magnesium, zirconium were chosen. these materials demand special procedure for welding. This report includes design and construction data with drawings of the special device for semiautomated circular welding.

  6. The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Kweon, Young Chul

    2002-01-01

    A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR 1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR 1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet

  7. Extending the maximum operation time of the MNSR reactor.

    Science.gov (United States)

    Dawahra, S; Khattab, K; Saba, G

    2016-09-01

    An effective modification to extend the maximum operation time of the Miniature Neutron Source Reactor (MNSR) to enhance the utilization of the reactor has been tested using the MCNP4C code. This modification consisted of inserting manually in each of the reactor inner irradiation tube a chain of three polyethylene-connected containers filled of water. The total height of the chain was 11.5cm. The replacement of the actual cadmium absorber with B(10) absorber was needed as well. The rest of the core structure materials and dimensions remained unchanged. A 3-D neutronic model with the new modifications was developed to compare the neutronic parameters of the old and modified cores. The results of the old and modified core excess reactivities (ρex) were: 3.954, 6.241 mk respectively. The maximum reactor operation times were: 428, 1025min and the safety reactivity factors were: 1.654 and 1.595 respectively. Therefore, a 139% increase in the maximum reactor operation time was noticed for the modified core. This increase enhanced the utilization of the MNSR reactor to conduct a long time irradiation of the unknown samples using the NAA technique and increase the amount of radioisotope production in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

  8. Viewing device of a steam generator tube-plate

    International Nuclear Information System (INIS)

    Denis, J.; Poirier, D.

    1984-01-01

    The invention proposes a device to observe the tubular plate of a steam generator including rows of parallel tubes situated in a shell provided with at least one entrance situated face to the interval between two adjacent rows. The device comprises a boom of which transversal dimension is less important than the interval; the boom can be inserted by the entrance; it contains a rigid endoscope terminated in an eyepiece and an optical fibre lighguide in the same vertical plane for illumination of the far end. The respective rotary angled mirrors are driven simultaneously by drums connected to a rack-and-pinion mechanism which is operated by a plunger held by a spring against a rocking lever driven by a motor and cam. As the mirrors rotate, the illuminated zone overlaps the field of view of the endoscope. The tube plate area in the shadow of the endoscope mirror (20) is illuminated separately by an ailiary fibre with a fixed terminal mirror. The invention enables the observation of the tube plate on both sides of the boom. It can be used in the case of the inspection of the steam generator of a pressurized water reactor [fr

  9. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  10. Russian RERTR program as a part of Joint US DOE-RF MINATOM collaboration on elimination of the threat connected to the use of HEU in research reactors

    International Nuclear Information System (INIS)

    Arkhangelsky, N.

    2002-01-01

    The Russian RERTR Program started at the end of 70's, the final goal of the program is to eliminate supplies of HEU in fuel elements and assemblies for foreign research reactors that were designed according to Russian projects. Basic directions of the work include: completion of the development of the fuel elements and assemblies on a basis of uranium dioxide; development of the fuel on a basis of U-Mo alloy; and development of pin type fuel elements. Fuel assemblies of WWR-M2 type with LEU were developed and qualified for using in foreign research reactors that use such type of fuel assemblies. These assemblies are ready for the supplying several operating foreign research reactors. There are more than 20 sites in Eastern European countries, former Soviet republics and another countries that have big amount of Russian origin HEU in fresh and spent fuel. The problem of the shipment of SNF from sites of research reactors is also very important for domestic Russian research reactors. More than ten years from its beginning the Russian RERTR program developed practically independently from the international RERTR program and only at the begin of 90's the Russian specialists started to contact with foreign scientists and the exchange of the scientific information has become more intensive. In September 1994, representatives of Minatom and DOE signed a protocol of intent to reduce an enrichment of uranium in research reactors. The main aspects of collaboration involve: Several domestic Russian research reactors such as WWR-M, IR-8 and others were investigated from the point of view of possibility of reducing of enrichment; financial support of the program from US DOE which is insufficient. The important part of international collaboration is the import of Russian origin spent and fresh fuel of research reactors to Russia. In August 2002 an impressive result of the Russian-American collaboration with support of IAEA and with the help and assistance of Yugoslavian side was

  11. Heavy water moderated tubular type nuclear reactor

    International Nuclear Information System (INIS)

    Oohashi, Masahisa.

    1986-01-01

    Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

  12. Methods and codes for neutronic calculations of the MARIA research reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.; Bretscher, M.M.; Hanan, N.A.; Matos, J.E.

    1998-01-01

    The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6x8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminium. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization. (author)

  13. photomultiplier tubes

    CERN Multimedia

    photomultiplier tubes. A device to convert light into an electric signal (the name is often abbreviated to PM). Photomultipliers are used in all detectors based on scintillating material (i.e. based on large numbers of fibres which produce scintillation light at the passage of a charged particle). A photomultiplier consists of 3 main parts: firstly, a photocathode where photons are converted into electrons by the photoelectric effect; secondly, a multiplier chain consisting of a serie of dynodes which multiply the number of electron; finally, an anode, which collects the resulting current.

  14. photomultiplier tube

    CERN Multimedia

    photomultiplier tubes. A device to convert light into an electric signal (the name is often abbreviated to PM). Photomultipliers are used in all detectors based on scintillating material (i.e. based on large numbers of fibres which produce scintillation light at the passage of a charged particle). A photomultiplier consists of 3 main parts: firstly, a photocathode where photons are converted into electrons by the photoelectric effect; secondly, a multiplier chain consisting of a serie of dynodes which multiply the number of electron; finally, an anode, which collects the resulting current.

  15. Modifying the Heysham 2 and Torness guide tubes

    International Nuclear Information System (INIS)

    Salter, I.D.

    1988-01-01

    In 1986 the National Nuclear Corporation carried out the unfuelled engineering run on Torness Reactor One. Subsequent inspection revealed wear on the reactor control rods, following severe spinning caused by gas cross flow swirls at the guide tube castellated ring. The adopted solution was to machine 32 radial holes through the guide tube wall and blank off the existing castellated slots, however, man access to the guide tubes is extremely difficult. This paper describes how Taylor Hitec produced, in only 12 weeks, three remote drilling machines, together with associated debris collection systems, cleaning equipment and remote video/CCTV inspection systems, and then carried out the modifications to the reactors. (author)

  16. Core barrel inner tube lifter

    Energy Technology Data Exchange (ETDEWEB)

    Jeffers, J P

    1968-07-16

    A core drill with means for selectively lifting a core barrel inner tube consists of a lifting means connected to the core barrel inner tube assembly. It has a closable passage to permit drilling fluid normally to pass through it. The lifting means has a normally downward facing surface and a means to direct drilling fluid pressure against that surface so that on closure of the passage to fluid flow, the pressure of the drilling fluid is caused to act selectively on it. This causes the lifting means to rise and lift the core barrel. (7 claims)

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

    International Nuclear Information System (INIS)

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

    2013-12-01

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

  18. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  19. Single-phase AutoReClosure ARC failure on 400 kV combinedcable/overhead line with permanently connected shunt reactor

    DEFF Research Database (Denmark)

    Bak, Claus Leth; Søgaard, Kim

    2008-01-01

    consisting of overhead lines, crossbonded cable sections and shunt reactor has been created in PSCAD/EMTDC and verified against measurements with good results. Main focus has been put on the likelihood of having a successful single-phase autoreclosure ARC in such a combined cable/OHL line....

  20. The Texts of the Instruments connected with the Agency's Assistance to Pakistan in Establishing a Research Reactor Project. A Second Supply Agreement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-01-12

    As a sequel to the assistance which the Agency provided to the Government of Pakistan in establishing a research reactor project, a second Supply Agreement has been concluded between the Agency and the Governments of Pakistan and the United States of America. This Agreement entered into force on 19 October 1967, and the text is reproduced herein for the information of all Members.