A study on integrity of LMFBR secondary cooling system to hypothetical tube failure propagation in the steam generator

International Nuclear Information System (INIS)

Yoshihisa Shindo; Kazuo Haga

2005-01-01

Full text of publication follows: A fundamental safety issue of liquid-metal-cooled fast breeder reactor (LMFBR) is to maintain the integrity of the secondary cooling system components against violent chemical sodium-water reaction caused by the water leak from the heat transfer tube of steam generators (SG). The produced sodium-water reaction jet would attack more severely surrounding tubes and would cause other tube failures (tube failure propagation), if it was assumed that the water leak was not detected by function-less detectors and proper operating actions to mitigate the tube failure propagation, such as isolations of the SG from the secondary cooling system and turbine water/steam system, and blowing water and steam inside tubes in the SG, were not taken. This study has been made focusing on the affection of large-scale water leak enlarged due to SG tube failure propagation to the structural integrity of the secondary cooling system because the generated pressure pulse caused by a large-scale sodium-water reaction might break heat transfer tubes of the intermediate heat exchanger (IHX). The present work has been made as one part of the study of probabilistic safety assessment (PSA) of LMFBR, because if the heat-transfer tubes of IHX were failed, the reactor core may be affected by the pressure pulse and/or by the sodium-water reaction products transported through the primary cooling system. As tools for PSA of the water leak incident of SG, we have developed QUARK-LP Version 4 code that mainly analyzes the high temperature rupture phenomena and estimates the number of failed tubes during the middle-scale water leak. The pressure pulse behavior generated by sodium-water reaction in the failure SG and the pressure propagation in the secondary cooling system are calculated by using the SWAAM-2 code developed by ANL. Furthermore, the quasi-steady state high pressure and temperature of the secondary cooling system in a long term is estimated by using the SWAAM

Ballooning of CANDU pressure tube in local thermal transients

International Nuclear Information System (INIS)

Mihalache, Maria; Ionescu, Viorel

2008-01-01

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

Consequences of pressure tube rupture on in-core components

International Nuclear Information System (INIS)

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

1982-12-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

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

Science.gov (United States)

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

2017-11-01

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

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)

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.)

Experimental Study on Solar Cooling Tube Using Thermal/Vacuum Emptying Method

Directory of Open Access Journals (Sweden)

Huizhong Zhao

2012-01-01

Full Text Available A solar cooling tube using thermal/vacuum emptying method was experimentally studied in this paper. The coefficient of performance (COP of the solar cooling tube was mostly affected by the vacuum degree of the system. In past research, the thermal vacuum method, using an electric oven and iodine-tungsten lamp to heat up the adsorbent bed and H2O vapor to expel the air from the solar cooling tube, was used to manufacture solar cooling tubes. This paper presents a novel thermal vacuum combined with vacuum pump method allowing an increased vacuum state for producing solar cooling tubes. The following conclusions are reached: the adsorbent bed temperature of solar cooling tube could reaches up to 233°C, and this temperature is sufficient to meet desorption demand; the refrigerator power of a single solar cooling tube varies from 1 W to 12 W; the total supply refrigerating capacity is about 287 kJ; and the COP of this solar cooling tube is about 0.215.

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)

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)

Thermal Sizing of Heat Exchanger Tubes for Air Natural Convective Cooling System of Emergency Cooling Tank

Energy Technology Data Exchange (ETDEWEB)

Kim, Myoung Jun; Lee, Hee Joon [Kookmin Univ., Seoul (Korea, Republic of); Moon, Joo Hyung; Bae, Youngmin; Kim, Youngin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

For the long operation of secondary passive cooling system, however, water level goes down by evaporation in succession at emergency cooling tank. At the end there would be no place to dissipate heat from condensation heat exchanger. Therefore, steam cooling heat exchanger is put on the top of emergency cooling tank to maintain appropriate water level by collecting evaporating steam. Steam cooling heat exchanger is installed inside an air chimney and evaporated steam is cooled down by air natural convection. In this study, thermal sizing of steam cooling heat exchanger under air natural convection was conducted by TSCON program for the design of experimental setup as shown in Fig. 2. Thermal sizing of steam cooling heat exchanger tube under air natural convection was conducted by TSCON program for the design of experimental setup. 25 - 1' tubes which has a length 1687 mm was determined as steam cooling heat exchanger at 2 kW heat load and 100 liter water pool in emergency cooling tank (experimental limit condition). The corresponding width of two tubes is 50 mm and has 5 by 5 tube array for heat exchanger.

Thermal Sizing of Heat Exchanger Tubes for Air Natural Convective Cooling System of Emergency Cooling Tank

International Nuclear Information System (INIS)

Kim, Myoung Jun; Lee, Hee Joon; Moon, Joo Hyung; Bae, Youngmin; Kim, Youngin

2014-01-01

For the long operation of secondary passive cooling system, however, water level goes down by evaporation in succession at emergency cooling tank. At the end there would be no place to dissipate heat from condensation heat exchanger. Therefore, steam cooling heat exchanger is put on the top of emergency cooling tank to maintain appropriate water level by collecting evaporating steam. Steam cooling heat exchanger is installed inside an air chimney and evaporated steam is cooled down by air natural convection. In this study, thermal sizing of steam cooling heat exchanger under air natural convection was conducted by TSCON program for the design of experimental setup as shown in Fig. 2. Thermal sizing of steam cooling heat exchanger tube under air natural convection was conducted by TSCON program for the design of experimental setup. 25 - 1' tubes which has a length 1687 mm was determined as steam cooling heat exchanger at 2 kW heat load and 100 liter water pool in emergency cooling tank (experimental limit condition). The corresponding width of two tubes is 50 mm and has 5 by 5 tube array for heat exchanger

Constructal design of finned tubes used in air-cooled heat exchangers

Energy Technology Data Exchange (ETDEWEB)

Shokouhmand, Hossein; Mahjoub, Shoeib [University of Tehran, Tehran (Iran, Islamic Republic of); Salimpour, Mohammad Reza [Isfahan University of Technology, Isfahan (Iran, Islamic Republic of)

2014-06-15

The present study documents the constructal design and optimization of finned tubes used in air-cooled heat exchangers. The considered tubes are equipped with annular fins. The aim is to minimize the overall thermal resistance by morphing the geometry. The geometrical and thermo-physical parameters considered are the number of fins, ratio of fin height to tube diameter, Stanton number, ratio of fin conductivity to air conductivity, ratio of in-tube fluid conductivity to air conductivity and dimensionless pressure drop. Two constraints are applied in the optimization process: fixed overall volume of heat exchanger and fixed volume fraction of fin material. It is found that there exist optimal values for the number and the height of fins. Moreover, the optimal heat transfer has an extremum in a special volume fraction of fin material.

Constructal design of finned tubes used in air-cooled heat exchangers

International Nuclear Information System (INIS)

Shokouhmand, Hossein; Mahjoub, Shoeib; Salimpour, Mohammad Reza

2014-01-01

The present study documents the constructal design and optimization of finned tubes used in air-cooled heat exchangers. The considered tubes are equipped with annular fins. The aim is to minimize the overall thermal resistance by morphing the geometry. The geometrical and thermo-physical parameters considered are the number of fins, ratio of fin height to tube diameter, Stanton number, ratio of fin conductivity to air conductivity, ratio of in-tube fluid conductivity to air conductivity and dimensionless pressure drop. Two constraints are applied in the optimization process: fixed overall volume of heat exchanger and fixed volume fraction of fin material. It is found that there exist optimal values for the number and the height of fins. Moreover, the optimal heat transfer has an extremum in a special volume fraction of fin material.

Optimization of the breeder zone cooling tubes of the DEMO Water-Cooled Lithium Lead breeding blanket

Energy Technology Data Exchange (ETDEWEB)

Di Maio, P.A.; Arena, P.; Bongiovì, G. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Chiovaro, P., E-mail: pierluigi.chiovaro@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Del Nevo, A. [ENEA Brasimone, Camugnano, BO (Italy); Forte, R. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy)

2016-11-01

Highlights: • Determination of an optimal configuration for the breeder zone cooling tubes. • Attention has been focused on the toroidal–radial breeder zone cooling tubes lay out. • A theoretical-computational approach based on the Finite Element Method (FEM) has been followed, adopting a qualified commercial FEM code. • Five different configurations have been investigated to optimize the breeder zone cooling tubes arrangement fulfilling all the rules prescribed by safety codes. - Abstract: The determination of an optimal configuration for the breeder zone (BZ) cooling tubes is one of the most important issues in the DEMO Water-Cooled Lithium Lead (WCLL) breeding blanket R&D activities, since BZ cooling tubes spatial distribution should ensure an efficient heat power removal from the breeder, avoiding hotspots occurrence in the thermal field. Within the framework of R&D activities supported by the HORIZON 2020 EUROfusion Consortium action on the DEMO WCLL breeding blanket design, a campaign of parametric analyses has been launched at the Department of Energy, Information Engineering and Mathematical Models of the University of Palermo (DEIM), in close cooperation with ENEA-Brasimone, in order to assess the potential influence of BZ cooling tubes number on the thermal performances of the DEMO WCLL outboard breeding blanket equatorial module under the nominal steady state operative conditions envisaged for it, optimizing their geometric configuration and taking also into account that a large number of cooling pipes can deteriorate the tritium breeding performances of the module. In particular, attention has been focused on the toroidal-radial option for the BZ tube bundles lay-out and a parametric study has been carried out taking into account different tube bundles arrangement within the module. The study has been carried out following a numerical approach, based on the finite element method (FEM), and adopting a qualified commercial FEM code. Results

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)

Numerical study of the thermo-flow performances of novel finned tubes for air-cooled condensers in power plant

Science.gov (United States)

Guo, Yonghong; Du, Xiaoze; Yang, Lijun

2018-02-01

Air-cooled condenser is the main equipment of the direct dry cooling system in a power plant, which rejects heat of the exhaust steam with the finned tube bundles. Therefore, the thermo-flow performances of the finned tubes have an important effect on the optimal operation of the direct dry cooling system. In this paper, the flow and heat transfer characteristics of the single row finned tubes with the conventional flat fins and novel jagged fins are investigated by numerical method. The flow and temperature fields of cooling air for the finned tubes are obtained. Moreover, the variations of the flow resistance and average convection heat transfer coefficient under different frontal velocity of air and jag number are presented. Finally, the correlating equations of the friction factor and Nusselt number versus the Reynolds number are fitted. The results show that with increasing the frontal velocity of air, the heat transfer performances of the finned tubes are enhanced but the pressure drop will increase accordingly, resulting in the average convection heat transfer coefficient and friction factor increasing. Meanwhile, with increasing the number of fin jag, the heat transfer performance is intensified. The present studies provide a reference in optimal designing for the air-cooled condenser of direct air cooling system.

Water-cooled U-tube grids for continuously operated neutral-beam injectors

International Nuclear Information System (INIS)

Hoffman, M.A.; Duffy, T.J.

1979-01-01

A design for water-cooled extractor grids for long-pulse and continuously operated ion sources for neutral-beam injectors is described. The most serious design problem encountered is that of minimizing the thermal deformation (bowing) of these slender grid rails, which have typical overall spans of 150 mm and diameters on the order of 1 mm. A unique U-tube design is proposed that offers the possibility of keeping the thermal bowing down to about 0.05 mm (about 2.0 mils). However, the design requires high-velocity cooling water at a Reynolds number of about 3 x 10 4 and an inlet pressure on the order of 4.67 x 10 6 Pa (677 psia) in order to keep the axial and circumferential temperature differences small enough to achieve the desired small thermal bowing. It appears possible to fabricate and assemble these U-tube grids out of molybdenum with high precision and with a reasonably small number of brazes

Inert medium (helium) irradiation testing of pressure tube samples

International Nuclear Information System (INIS)

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

2001-01-01

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

Flow instability research on steam generator with straight double-walled heat transfer tube for FBR. Pressure drop under high pressure condition

International Nuclear Information System (INIS)

Liu, Wei; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

2008-01-01

For the Steam Generator (SG) with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments were performed under high pressure condition in JAEA with using a straight tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper evaluates the pressure drop data with TRAC-BF1 code. The Pffan's correlation for single phase flow and the Martinelli-Nelson's two-phase flow multiplier are found can be well predicted the present pressure drop data under high pressure condition. (author)

High temperature deformation behavior of gradually pressurized zircaloy-4 tubes

International Nuclear Information System (INIS)

Suzuki, Motoye

1982-03-01

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

Preventing freezing of condensate inside tubes of air cooled condenser

International Nuclear Information System (INIS)

Joo, Jeong A; Hwang, In Hwan; Lee, Dong Hwan; Cho, Young Il

2012-01-01

An air cooled condenser is a device that is used for converting steam into condensate by using ambient air. The air cooled condenser is prone to suffer from a serious explosion when the condensate inside the tubes of a heat exchanger is frozen; in particular, tubes can break during winter. This is primarily due to the structural problem of the tube outlet of an existing conventional air cooled condenser system, which causes the backflow of residual steam and noncondensable gases. To solve the backflow problem in such condensers, such a system was simulated and a new system was designed and evaluated in this study. The experimental results using the simulated condenser showed the occurrence of freezing because of the backflow inside the tube. On the other hand, no backflow and freezing occurred in the advanced new condenser, and efficient heat exchange occurred

Experimental Study on Cooling Heat Transfer of Supercritical Carbon Dioxide Inside Horizontal Micro-Fin Tubes

Science.gov (United States)

Kuwahara, Ken; Higashiiu, Shinya; Ito, Daisuke; Koyama, Shigeru

This paper deals with the experimental study on cooling heat transfer of supercritical carbon dioxide inside micro-fin tubes. The geometrical parameters in micro-fin tubes used in the present study are 6.02 mm in outer diameter, 4.76 mm to 5.11 mm in average inner diameter, 0.15 mm to 0.24 mm in fin height, 5 to 25 in helix angle, 46 to 52 in number of fins and 1.4 to 2.3 in area expansion ratio. Heat transfer coefficients were measured at 8-10 MPa in pressure, 360-690 kg/(m2•s) in mass velocity and 20-75 °C in CO2 temperature. The measured heat transfer coefficients of micro-fin tubes were 1.4 to 2 times higher than those of the smooth tube having 4.42 in inner diameter. The predicted heat transfer coefficients using the correlation equation, which was developed for single-phase turbulent fluid flow inside micro-fin-tubes, showed large deviations to the measured values. The new correlation to predict cooling heat transfer coefficient of supercritical carbon dioxide inside micro-fin tubes was developed taking into account the shape of fins based on experimental data empirically. This correlation equation agreed within ±20% of almost all of the experimental data.

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

Science.gov (United States)

Wang, Xinyi; Shen, Jialong; Liu, Xinbo

2018-01-01

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

Development of NDT techniques for the inspection of WSGHWR pressure tubes

International Nuclear Information System (INIS)

Gray, B.S.; Highmore, P.J.; Rudlin, J.R.; Cooper, A.G.

1979-01-01

The fuel for the Steam Generating Heavy Water Reactor at Winfrith Heath is contained in vertical Zircaloy pressure tubes and is cooled by boiling light water. This paper describes the development of NDT techniques for the inservice examination of the pressure tubes to provide continuing assurance of the absence of axial crack-like defects. The resultant equipment has to operate in water-filled tubes in the presence of the radiation field due to the irradiated fuel elements in adjacent tubes. Also, a layer of surface oxide on the inside of the tubes has been found to significantly affect the behaviour of a prototype inspection device. To provide adequate sensitivity in these conditions, without the occurrence of unnecessary spurious indications, a combination of techniques has been developed. This involves the use of ultrasonics in both pulse-echo and 'pitch and catch' mode together with a single frequency eddy current technique. Laboratory work using artificial defects is described and also how the development programme was modified to accommodate the results of in-reactor tests using a prototype device. Reference is also made to the development of CCTV equipment to provide a supplementary visual examination. (author)

Evaluation of water cooled supersonic temperature and pressure probes for application to 1366 K flows

Science.gov (United States)

Lagen, Nicholas; Seiner, John M.

1990-01-01

Water cooled supersonic probes are developed to investigate total pressure, static pressure, and total temperature in high-temperature jet plumes and thereby determine the mean flow properties. Two probe concepts, designed for operation at up to 1366 K in a Mach 2 flow, are tested on a water cooled nozzle. The two probe designs - the unsymmetric four-tube cooling configuration and the symmetric annular cooling design - take measurements at 755, 1089, and 1366 K of the three parameters. The cooled total and static pressure readings are found to agree with previous test results with uncooled configurations. The total-temperature probe, however, is affected by the introduction of water coolant, and effect which is explained by the increased heat transfer across the thermocouple-bead surface. Further investigation of the effect of coolant on the temperature probe is proposed to mitigate the effect and calculate more accurate temperatures in jet plumes.

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

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

Achieving reduced fouling of cooling water exchangers with stainless steel tubes

International Nuclear Information System (INIS)

Iftikhar, A.; Mir, N.

2010-01-01

Good performance of cooling water heat exchangers plays a vital role in the over all energy efficiency of a chemical plant. Heavy fouling on carbon steel tubes of the cooling water exchangers was causing poor performance and frequent cleaning requirement. The carbon steel tubes were replaced with stainless steel tubes. Improved performance was achieved and cleaning frequency reduced. The paper covers the details of study and methodology applied for the above changes along with summary of results. (author)

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

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

Design and performance prediction of an adsorption heat pump with multi-cooling tubes

Energy Technology Data Exchange (ETDEWEB)

Wang, D.C.; Zhang, J.P. [College of Electromechanical Engineering, Qingdao University, Qingdao 266071 (China)

2009-05-15

Widespread application of adsorption heat pumps has been delayed not only by poor heat and mass transfer performance but also by low operating reliability because high vacuum must be maintained in the adsorption cooling system, especially in a water system. An adsorption cooling tube is a tube in which an adsorber, a condenser and an evaporator are all completely housed to construct a small scale adsorption cooling unit. In this work, an adsorption cooling tube and an adsorption heat pump with multi-cooling tubes are designed. A theoretical model is built to simulate the performance of the designed chiller. According to the results, the coefficient of performance and specific cooling power reach about 0.5 and 85 W/kg adsorbent, respectively, at the hot water temperature of 85 C. These results indicate that the designed heat pump in this work would provide a better choice if the operating reliability became crucial for an adsorption heat pump. (author)

Design and performance prediction of an adsorption heat pump with multi-cooling tubes

International Nuclear Information System (INIS)

Wang, D.C.; Zhang, J.P.

2009-01-01

Widespread application of adsorption heat pumps has been delayed not only by poor heat and mass transfer performance but also by low operating reliability because high vacuum must be maintained in the adsorption cooling system, especially in a water system. An adsorption cooling tube is a tube in which an adsorber, a condenser and an evaporator are all completely housed to construct a small scale adsorption cooling unit. In this work, an adsorption cooling tube and an adsorption heat pump with multi-cooling tubes are designed. A theoretical model is built to simulate the performance of the designed chiller. According to the results, the coefficient of performance and specific cooling power reach about 0.5 and 85 W/kg adsorbent, respectively, at the hot water temperature of 85 deg. C. These results indicate that the designed heat pump in this work would provide a better choice if the operating reliability became crucial for an adsorption heat pump.

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)

Dynamics of explosively imploded pressurized tubes

Science.gov (United States)

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

2011-04-01

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

Development of Zirconium alloys (for pressure tubes)

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

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

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro

1994-11-01

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

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.)

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

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.

Selected reading on introduction to pressure tube technology

International Nuclear Information System (INIS)

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

1981-10-01

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

Flat tile armour cooled by hypervapotron tube: a possible technology for ITER

International Nuclear Information System (INIS)

Schlosser, J.; Escourbiac, F.; Bayetti, P.; Missirlian, M.; Mitteau, R.; Schedler, B.; Bobin-Vastra, I.

2003-01-01

Carbon fibre composite (CFC) flat tile armours for actively cooled plasma facing components (PFC's) are an important challenge for controlled fusion machine. Flat tile concepts, water cooled by tubes, were studied, developed, tested and finally experienced with success in Tore Supra. The components were designed for 10 MW/m 2 and mock-ups were successfully fatigue tested at 15 MW/m 2 , 1000 cycles. For ITER, a tube-in-tile concept was developed and mock-ups sustained up to 25 MW/m 2 for 1000 cycles without failure. Recently flat tile armored mock-ups cooled by Hypervapotron tube successfully sustained a cascade failure test under a mean heat flux of 10 MW/m 2 but with a doubling of the heat flux on some tiles to simulate missing tiles (500 cycles). This encouraging results lead to reconsider the limits for flat tile concept when cooled by Hypervapotron tube. New tests are now scheduled to investigate these limits notably in regards to the ITER requirements. The concept could also be experimented in Tore Supra by installing a new limiter into the machine. (authors)

Hot hardness studies on zircaloy 2 pressure tube along three orientations

International Nuclear Information System (INIS)

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

2002-01-01

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

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

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.

Two phase cooling for superconducting magnets

International Nuclear Information System (INIS)

Eberhard, P.H.; Gibson, G.A.; Green, M.A.; Ross, R.R.; Smits, R.G.

1986-01-01

Comments on the use of two phase helium in a closed circuit tubular cooling system and some results obtained with the TPC superconducting magnet are given. Theoretical arguments and experimental evidence are given against a previously suggested method to determine helium two phase flow regimes. Two methods to reduce pressure in the magnet cooling tubes during quenches are discussed; 1) lowering the density of helium in the magnet cooling tubes and 2) proper location of pressure relief valves. Some techniques used to protect the refrigerator from too much cold return gas are also mentioned

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

International Nuclear Information System (INIS)

Larson, G.C.

1980-01-01

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

Flat tile armour cooled by hypervapotron tube: a possible technology for ITER

Energy Technology Data Exchange (ETDEWEB)

Schlosser, J.; Escourbiac, F.; Bayetti, P.; Missirlian, M.; Mitteau, R. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Merola, M. [European Fusion Development Agreement - Close Support Unit (EFDA), Garching (Germany); Schedler, B. [Plansee Aktiengesellschaft, Technology Center, Reutte/Tirol (Austria); Bobin-Vastra, I. [FRAMATOME-ANP, Centre Technique, 71 - Le Creusot (France)

2003-07-01

Carbon fibre composite (CFC) flat tile armours for actively cooled plasma facing components (PFC's) are an important challenge for controlled fusion machine. Flat tile concepts, water cooled by tubes, were studied, developed, tested and finally experienced with success in Tore Supra. The components were designed for 10 MW/m{sup 2} and mock-ups were successfully fatigue tested at 15 MW/m{sup 2}, 1000 cycles. For ITER, a tube-in-tile concept was developed and mock-ups sustained up to 25 MW/m{sup 2} for 1000 cycles without failure. Recently flat tile armored mock-ups cooled by Hypervapotron tube successfully sustained a cascade failure test under a mean heat flux of 10 MW/m{sup 2} but with a doubling of the heat flux on some tiles to simulate missing tiles (500 cycles). This encouraging results lead to reconsider the limits for flat tile concept when cooled by Hypervapotron tube. New tests are now scheduled to investigate these limits notably in regards to the ITER requirements. The concept could also be experimented in Tore Supra by installing a new limiter into the machine. (authors)

Flat Tile Armour Cooled by Hypervapotron Tube: a Possible Technology for ITER

Science.gov (United States)

Schlosser, J.; Escourbiac, F.; Merola, M.; Schedler, B.; Bayetti, P.; Missirlian, M.; Mitteau, R.; Robin-Vastra, I.

Carbon fibre composite (CFC) flat tile armours for actively cooled plasma facing components (PFC’s) are an important challenge for controlled fusion machines. Flat tile concepts, water cooled by tubes, were studied, developed, tested and finally operated with success in Tore Supra. The components were designed for 10 MW/m2 and mock-ups were successfully fatigue tested at 15 MW/m2, 1000 cycles. For ITER, a tube-in-tile concept was developed and mock-ups sustained up to 25 MW/m2 for 1000 cycles without failure. Recently flat tile armoured mock-ups cooled by a hypervapotron tube successfully sustained a cascade failure test under a mean heat flux of 10 MW/m2 but with a doubling of the heat flux on some tiles to simulate missing tiles (500 cycles). This encouraging results lead to reconsider the limits for flat tile concept when cooled by hypervapotron (HV) tube. New tests are now scheduled to investigate these limits in regard to the ITER requirements. Experimental evidence of the concept could be gained in Tore Supra by installing a new limiter into the machine.

Flat Tile Armour Cooled by Hypervapotron Tube: a Possible Technology for ITER

Energy Technology Data Exchange (ETDEWEB)

Schlosser, J.; Escourbiac, F.; Bayetti, P.; Missirlian, M.; Mitteau, R. [Association Euratom CEA, DSM/DRFC/SIPP, St Paul lez Durance (France); Merola, M. [EFDA Close Support Unit, Garching (Germany); Schedler, B. [Plansee Aktiengesellschaft, Reutte (Austria). Technology Center; Bobin-Vastra, I. [Framatome-ANP, Le Creusot (France). Centre Technique

2004-08-01

Carbon fibre composite (CFC) flat tile armours for actively cooled plasma facing components (PFC's) are an important challenge for controlled fusion machines. Flat tile concepts, water cooled by tubes, were studied, developed, tested and finally operated with success in Tore Supra. The components were designed for 10MW/m{sup 2} and mock-ups were successfully fatigue tested at 15MW/m{sup 2}; 1000 cycles. For ITER, a tube-in-tile concept was developed and mock-ups sustained up to 25MW/m{sup 2} for 1000 cycles without failure. Recently flat tile armoured mock-ups cooled by a hypervapotron tube successfully sustained a cascade failure test under a mean heat flux of 10MW/m{sup 2} but with a doubling of the heat flux on some tiles to simulate missing tiles (500 cycles). This encouraging results lead to reconsider the limits for flat tile concept when cooled by hypervapotron (HV) tube. New tests are now scheduled to investigate these limits in regard to the ITER requirements. Experimental evidence of the concept could be gained in Tore Supra by installing a new limiter into the machine.

Pressure tube rupture in a closed tank

International Nuclear Information System (INIS)

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

1985-06-01

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

21 CFR 868.5860 - Pressure tubing and accessories.

Science.gov (United States)

2010-04-01

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

Evaluation of a leaking crack in an irradiated CANDU pressure tube

International Nuclear Information System (INIS)

Coleman, C.E.; Simpson, L.A.

1988-06-01

Leak-before-break is used in CANDU reactors as part of the defence against rupture of the pressure tubes. Two important features of this technique are the action time available for detection of a leaking crack and the size of the leak allowing crack location. Support for continued reliance on leak-before-break is being obtained from experiments, on irradiated Zr-2.5 Nb pressure tubes attached to their end fittings, that simulate the behaviour of a leaking crack in a reactor. At reactor operating temperatures leaking cracks grow more slowly than dry cracks in the laboratory because they are cooled when pressurised water flashes to steam on their surface. These cracks remain stable till they are at least 70 mm long. From the results of these experiments the action time is at least 100 h. The leak rate increases rapidly when a through-wall crack extends a small amount, thus greatly assisting with crack location

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

Two phase cooling for superconducting magnets

International Nuclear Information System (INIS)

Eberhard, P.H.; Gibson, G.A.; Green, M.A.; Ross, R.R.; Smits, R.G.; Taylor, J.D.; Watt, R.D.

1986-01-01

Comments on the use of two phase helium in a closed circuit tubular cooling system and some results obtained with the TPC superconducting magnet are given. Theoretical arguments and experimental evidence are given against a previously suggested method to determine helium two phase flow regimes. Two methods to reduce pressure in the magnet cooling tubes during quenches are discussed; (1) lowering the density of helium in the magnet cooling tubes and (2) proper location of pressure relief valves. Some techniques used to protect the refrigerator from too much cold return gas are also mentioned. 10 refs., 1 fig., 5 tabs

Tube failures due to cooling process problem and foreign materials in power plants

Energy Technology Data Exchange (ETDEWEB)

Ahmad, J. [Kapar Energy Ventures Sdn Bhd, Jalan Tok Muda, Kapar 42200 (Malaysia); Purbolaksono, J., E-mail: judha@uniten.edu.m [Department of Mechanical Engineering, Universiti Tenaga Nasional, Km 7 Jalan Kajang-Puchong, Kajang 43009, Selangor (Malaysia); Beng, L.C. [Kapar Energy Ventures Sdn Bhd, Jalan Tok Muda, Kapar 42200 (Malaysia)

2010-07-15

Cooling process which uses water for heat transfer is an essential factor in coal-fired and nuclear plants. Loss of cooling upset can force the plants to shut down. In particular, this paper reports visual inspections and metallurgical examinations on the failed SA210-A1 right-hand side (RHS) water wall tube of a coal-fired plant. The water wall tube showed the abnormal outer surface colour and has failed with wide-open ductile rupture and thin edges indicating typical signs of short-term overheating. Metallurgical examinations confirmed the failed tube experiencing higher temperature operation. Water flow starvation due to restriction inside the upstream tube is identified as the main root cause of failure. The findings are important to take failure mitigation actions in the future operation. Discussion on the typical problems related to the cooling process in nuclear power plants is also presented.

Tube failures due to cooling process problem and foreign materials in power plants

International Nuclear Information System (INIS)

Ahmad, J.; Purbolaksono, J.; Beng, L.C.

2010-01-01

Cooling process which uses water for heat transfer is an essential factor in coal-fired and nuclear plants. Loss of cooling upset can force the plants to shut down. In particular, this paper reports visual inspections and metallurgical examinations on the failed SA210-A1 right-hand side (RHS) water wall tube of a coal-fired plant. The water wall tube showed the abnormal outer surface colour and has failed with wide-open ductile rupture and thin edges indicating typical signs of short-term overheating. Metallurgical examinations confirmed the failed tube experiencing higher temperature operation. Water flow starvation due to restriction inside the upstream tube is identified as the main root cause of failure. The findings are important to take failure mitigation actions in the future operation. Discussion on the typical problems related to the cooling process in nuclear power plants is also presented.

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

International Nuclear Information System (INIS)

Shewfelt, R.S.W.

1986-01-01

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

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

Numerical analysis on the condensation heat transfer and pressure drop characteristics of the horizontal tubes of modular shell and tube-bundle heat exchanger

International Nuclear Information System (INIS)

Ko, Seung Hwan; Park, Hyung Gyu; Kim, Charn Jung; Park, Byung Kyu

2001-01-01

A numerical analysis of the heat and mass transfer and pressure drop characteristics in modular shell and tube bundle heat exchanger was carried out. Finite concept method based on FVM and κ-ε turbulent model were used for this analysis. Condensation heat transfer enhanced total heat transfer rate 4∼8% higher than that of dry heat exchanger. With increasing humid air inlet velocity, temperature and relative humidity, and with decreasing heat exchanger aspect ratio and cooling water velocity, total heat and mass transfer rate could be increased. Cooling water inlet velocity had little effect on total heat transfer

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

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.

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

Low pressure cooling seal system for a gas turbine engine

Science.gov (United States)

Marra, John J

2014-04-01

A low pressure cooling system for a turbine engine for directing cooling fluids at low pressure, such as at ambient pressure, through at least one cooling fluid supply channel and into a cooling fluid mixing chamber positioned immediately downstream from a row of turbine blades extending radially outward from a rotor assembly to prevent ingestion of hot gases into internal aspects of the rotor assembly. The low pressure cooling system may also include at least one bleed channel that may extend through the rotor assembly and exhaust cooling fluids into the cooling fluid mixing chamber to seal a gap between rotational turbine blades and a downstream, stationary turbine component. Use of ambient pressure cooling fluids by the low pressure cooling system results in tremendous efficiencies by eliminating the need for pressurized cooling fluids for sealing this gap.

Dry cooling tower operating experience in the LOFT reactor

International Nuclear Information System (INIS)

Hunter, J.A.

1980-01-01

A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

Sample summary report for ARG 1 pressure tube sample

International Nuclear Information System (INIS)

Belinco, C.

2006-01-01

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

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

Directory of Open Access Journals (Sweden)

Zeki ARGUNHAN

2006-02-01

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

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)

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

Evaluation of water cooled supersonic temperature and pressure probes for application to 2000 F flows

Science.gov (United States)

Lagen, Nicholas T.; Seiner, John M.

1990-01-01

The development of water cooled supersonic probes used to study high temperature jet plumes is addressed. These probes are: total pressure, static pressure, and total temperature. The motivation for these experiments is the determination of high temperature supersonic jet mean flow properties. A 3.54 inch exit diameter water cooled nozzle was used in the tests. It is designed for exit Mach 2 at 2000 F exit total temperature. Tests were conducted using water cooled probes capable of operating in Mach 2 flow, up to 2000 F total temperature. Of the two designs tested, an annular cooling method was chosen as superior. Data at the jet exit planes, and along the jet centerline, were obtained for total temperatures of 900 F, 1500 F, and 2000 F, for each of the probes. The data obtained from the total and static pressure probes are consistent with prior low temperature results. However, the data obtained from the total temperature probe was affected by the water coolant. The total temperature probe was tested up to 2000 F with, and without, the cooling system turned on to better understand the heat transfer process at the thermocouple bead. The rate of heat transfer across the thermocouple bead was greater when the coolant was turned on than when the coolant was turned off. This accounted for the lower temperature measurement by the cooled probe. The velocity and Mach number at the exit plane and centerline locations were determined from the Rayleigh-Pitot tube formula.

Suppression of acoustic streaming in tapered pulse tubes

International Nuclear Information System (INIS)

Olson, J.R.; Swift, G.W.

1998-01-01

In a pulse tube cryocooler, the gas in the pulse tube can be thought of as an insulating piston, transmitting pressure and velocity from the cold heat exchanger to the hot end of the pulse tube. Unfortunately, convective heat transfer can carry heat from the hot end to the cold end and reduce the net cooling power. Here, the authors discuss one driver of such convection: steady acoustic streaming as generated by interactions between the boundary and the oscillating pressure, velocity, and temperature. Using a perturbation method, they have derived an analytical expression for the streaming in a tapered pulse tube with axially varying mean temperature in the acoustic boundary layer limit. The calculations showed that the streaming depends strongly on the taper angle, the ratio of velocity and pressure amplitudes, and the phase between the velocity and pressure, but it depends only weakly on the mean temperature profile and is independent of the overall oscillatory amplitude. With the appropriate tapering of the tube, streaming can be eliminated for a particular operating condition. Experimentally, the authors have demonstrated that an orifice pulse tube cryocooler with the calculated zero-streaming taper has more cooling power than one with either a cylindrical tube or a tapered pulse tube with twice the optimum taper angle

Heat transfer study of water-cooled swirl tubes for neutral beam targets

International Nuclear Information System (INIS)

Kim, J.; Davis, R.C.; Gambill, W.R.; Haselton, H.H.

1977-01-01

Heat transfer considerations of water-cooled swirl-tubes including heat transfer correlations, burnout data, and 2-D considerations are presented in connection with high power neutral beam target applications. We also discuss performance results of several swirl tube targets in use at neutral beam development facilities

Low-pressure water-cooled inductively coupled plasma torch

Science.gov (United States)

Seliskar, Carl J.; Warner, David K.

1988-12-27

An inductively coupled plasma torch is provided which comprises an inner tube, including a sample injection port to which the sample to be tested is supplied and comprising an enlarged central portion in which the plasma flame is confined; an outer tube surrounding the inner tube and containing water therein for cooling the inner tube, the outer tube including a water inlet port to which water is supplied and a water outlet port spaced from the water inlet port and from which water is removed after flowing through the outer tube; and an r.f. induction coil for inducing the plasma in the gas passing into the tube through the sample injection port. The sample injection port comprises a capillary tube including a reduced diameter orifice, projecting into the lower end of the inner tube. The water inlet is located at the lower end of the outer tube and the r.f. heating coil is disposed around the outer tube above and adjacent to the water inlet.

Rejection index for pressure tubes

International Nuclear Information System (INIS)

Mitchell, A.B.; Meneley, D.

1989-10-01

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

Bulging of pressure tubes at hot spots under LOCA conditions

International Nuclear Information System (INIS)

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

1996-01-01

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

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

Directory of Open Access Journals (Sweden)

Mukul Kumar Jain

2011-01-01

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

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

International Nuclear Information System (INIS)

Tingkuan, C.; Xuanzheng, C.

1987-01-01

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

Remotised sliver sample removal from irradiated pressure tube

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Remotised sliver sample removal from irradiated pressure tube

International Nuclear Information System (INIS)

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

1994-01-01

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

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.

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

Development of technology on the material surveillance of CANDU pressure tubes

Energy Technology Data Exchange (ETDEWEB)

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

1995-05-01

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

Development of technology on the material surveillance of CANDU pressure tubes

International Nuclear Information System (INIS)

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

1995-05-01

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

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)

Development of delayed hydride cracking resistant-pressure tube

Energy Technology Data Exchange (ETDEWEB)

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

2000-10-01

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

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

The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

2001-03-01

In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary.

The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

International Nuclear Information System (INIS)

Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

2001-03-01

In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Innovation in pressure tube life assessment

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Delayed hydrogen cracking of zirconium alloy pressure tubes

International Nuclear Information System (INIS)

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

1976-10-01

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

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

International Nuclear Information System (INIS)

Hu, M.H.

1998-01-01

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

Mechanical and experimental study on freeze proof solar powered adsorption cooling tube using active carbon/methanol working pair

International Nuclear Information System (INIS)

Zhao Huizhong; Zhang Min; Liu Zhenyan; Liu Yanling; Ma Xiaodong

2008-01-01

The freeze proof solar cooling tube, which can produce cooling capacity with the refrigerant temperature below 0 deg. C using solar light as energy and active carbon-methanol as working pair, was firstly designed and made in this research. This paper focused on mechanical and experimental study on a freeze proof solar powered adsorption cooling tube. The following experimental results could be concluded: at the solar radiation value between 15.3 and 17.1 MJ m -2 , the highest adsorbent bed temperature is below 110 deg. C. The freeze proof solar cooling tube's cooling capacity was about 87-99 kJ, and the coefficient of performance (COP) was more than 0.11 when the evaporation temperature was about -4 deg. C

Signal characteristics of guided wave for condenser tube of NPP

International Nuclear Information System (INIS)

Min, Lee Dong; Hoon, Choi Sang; Yeong, Yang Tae

2012-01-01

A Condenser is a large heat exchanger of the shell and tube type. Cooling water enters through the water box, through the tubesheet and into the tubes(about 80,000 tubes/unit). The shell side of the condenser receives steam from the low pressure turbines exhaust. The steam is cooled to a liquid by passing over the tubes where the cooling water is circulated. Because seawater is used as a coolant, condenser tubes are easily damaged. For such a reason, nondestructive testing conducted periodically. But nondestructive testing takes a lot of manpower and time. Guided wave technique can overcome these shortcomings. In this study, we made an effort evaluating a guided wave defect signal

Database for Pressure Tube Diameter and Operation Data of Wolsong NPP

International Nuclear Information System (INIS)

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

2010-12-01

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

Linac drift tube tank upgrade engineering - cooling solution

International Nuclear Information System (INIS)

Li, G.; Heilbrunn, W.; Potter, J.

1999-01-01

Components from the injector of the canceled SSC project are being modified by JPAW to make a commercial radioisotope production linac for I 3 in Denton, TX. The biggest challenge of the upgraded design is the increased average power of the DTL, 40 times the original. With the thermo-mechanical analysis backed by a thorough understanding of the thermal physics, 156 drift tubes have been redesigned according to the RF power deposition. Increasing flow rate in the original cooling channels and adding four flow paths reduces the average tank temperature to an acceptable level. The Δf tolerance budget is controlled without the use of additional temperature control units. The unfinished SSC endwall parts have been modified for additional cooling of the nose and the wall. The different LINAC cooling subsystems are connected to a manifold in parallel through independent flow control valves to balance the required flow rate for each branch

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

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

Measurement of digital blood pressure after local cooling

DEFF Research Database (Denmark)

Nielsen, S L; Lassen, N A

1977-01-01

A double-inlet plastic cuff was designed for local cooling and systolic blood pressure measurement on the middle phalanx of the fingers. With a tourniquet on the proximal phalanx of one finger, cooling for 5 min made the digital artery temperature equal the skin temperature. The difference between...... the systolic pressure in a control finger and in the cooled finger give the reopening pressure in the digital arteries. At 30, 25, 20, 15, and 10 degrees C, respectively the percent decrease of the finger pressure was 0.2 (0.2), 1.5 (2.5), 8.5 (3.7), 11.4 (3.4), and 15.3 (3.1) in normal young women...

Compact, single-tube scanning tunneling microscope with thermoelectric cooling.

Science.gov (United States)

Jobbins, Matthew M; Agostino, Christopher J; Michel, Jolai D; Gans, Ashley R; Kandel, S Alex

2013-10-01

We have designed and built a scanning tunneling microscope with a compact inertial-approach mechanism that fits inside the piezoelectric scanner tube. Rigid construction allows the microscope to be operated without the use of external vibration isolators or acoustic enclosures. Thermoelectric cooling and a water-ice bath are used to increase temperature stability when scanning under ambient conditions.

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

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

International Nuclear Information System (INIS)

Azer, N.; Wong, H.

1994-01-01

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

TEM examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

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

2005-09-01

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

Comparison of evaluation method for planar flaw in pressure tube

International Nuclear Information System (INIS)

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

2009-01-01

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

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

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.

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)

Quantitative experiments on thermal hydraulic characteristics of an annular tube with twisted fins

International Nuclear Information System (INIS)

Ezato, Koichiro; Dairaku, Masayuki; Taniguchi, Masaki; Sato, Kazuyoshi; Suzuki, Satoshi; Akiba, Masato

2003-11-01

Thermal hydraulic experiments measuring critical heat flux (CHF) and pressure drop of an annular tube with twisted fins, ''annular swirl tube'', has been performed to examine its applicability to the ITER divertor cooling structure. The annular swirl tube consists of two concentric circular tubes, the outer and inner tubes. The outer tube with outer and inner diameters (OD and ID) of 21 mm and 15 mm is made of Cu-alloy that is CuCrZr and oe of candidate materials of the ITER divertor cooling tube. The inner tube with OD of 11 mm and ID of 9 mm is made of stainless steal. It has an external swirl fin with twist ratio (y) of three to enhance its heat transfer performance. In this tube, cooling water flows inside of the inner tube first, and then returns into an annulus between the outer and inner tubes with a swirl flow at an end-return of the cooling tube. The CHF experiments show that no degradation of CHF of the annular swirl tube in comparison with the conventional swirl tube whose dimensions are similar to those of the outer tube of the annular swirl tube. A minimum axial velocity of 7.1 m/s is required to remove the incident heat flux of 28MW/m 2 , the ITER design value. Applicability of the JAERI's correlation for the heat transfer to the annular swirl tube is also demonstrated by comparing the experimental results with those of the numerical analysis. The friction factor correlation for the annular flow with the twisted fins is also proposed for the hydrodynamic design of the ITER vertical target. The least pressure drop at the end-return is obtained by using the hemispherical end-plug. Its radius is the same as that of ID of the outer cooling tube. These results show that thermal-hydraulic performance of the annular swirl tube is promising in application to the cooling structure for the ITER vertical target. (author)

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Delayed hydrogen cracking test design for pressure tubes

International Nuclear Information System (INIS)

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

1999-01-01

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

Experimental study on a simple Ranque-Hilsch vortex tube

NARCIS (Netherlands)

Gao, C.; Bosschaart, K.J.; Zeegers, J.C.H.; Waele, de A.T.A.M.

2005-01-01

The Ranque-Hilsch vortex tube is a device by which cold gas can be generated using compressed gas. To understand the cooling mechanism of this device, it is necessary to know the pressure, temperature, and velocity distributions inside the tube. In order to investigate this, a simple vortex tube is

Ultra-low-vibration pulse-tube cryocooler system - cooling capacity and vibration

Science.gov (United States)

Ikushima, Yuki; Li, Rui; Tomaru, Takayuki; Sato, Nobuaki; Suzuki, Toshikazu; Haruyama, Tomiyoshi; Shintomi, Takakazu; Yamamoto, Akira

2008-09-01

This report describes the development of low-vibration cooling systems with pulse-tube (PT) cryocoolers. Generally, PT cryocoolers have the advantage of lower vibrations in comparison to those of GM cryocoolers. However, cooling systems for the cryogenic laser interferometer observatory (CLIO), which is a gravitational wave detector, require an operational vibration that is sufficiently lower than that of a commercial PT cryocooler. The required specification for the vibration amplitude in cold stages is less than ±1 μm. Therefore, during the development of low-vibration cooling systems for the CLIO, we introduced advanced countermeasures for commercial PT cryocoolers. The cooling performance and the vibration amplitude were evaluated. The results revealed that 4 K and 80 K PT cooling systems with a vibration amplitude of less than ±1 μm and cooling performance of 4.5 K and 70 K at heat loads of 0.5 W and 50 W, respectively, were developed successfully.

Heat exchanger for cooling a liquid metal with air, including panels of identical tubes

International Nuclear Information System (INIS)

Malaval, C.

1985-01-01

The heat exchanger includes panels of identical tubes, each one comprising two horizontal collectors situated at the vertical of each other and a group of vertical tubes for cooling arranged in a horizontal parallelepiped casing opened on two of its opposite sides. The air flows from the inlet to the outlet face of the casing. The panels of tubes are arranged side by side so that their outlet faces form a prismatic surface of which the height is vertical and the inner space communicates with a vertical axis chimney. Each one of the panels is hanging from a fixed structure by means of articulated fasteners, by means of its upper collector only. The invention applies, more particularly, for cooling the primary sodium of fast neutron reactors after they are stopped [fr

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)

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

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)

Prevention of biliary complication in radiofrequency ablation for hepatocellular carcinoma-Cooling effect by endoscopic nasobiliary drainage tube

Energy Technology Data Exchange (ETDEWEB)

Ogawa, Tsuneyoshi [Department of Gastroenterology and Hepatology, Okayama University Graduate School of Medicine, Dentistry, and Pharmaceutical Sciences, 2-5-1 Shikata-cho, Okayama 700-8558 (Japan); Kawamoto, Hirofumi [Department of Gastroenterology and Hepatology, Okayama University Graduate School of Medicine, Dentistry, and Pharmaceutical Sciences, 2-5-1 Shikata-cho, Okayama 700-8558 (Japan)], E-mail: h-kawamo@md.okayama-u.ac.jp; Kobayashi, Yoshiyuki; Nakamura, Shinichiro; Miyatake, Hirokazu; Harada, Ryo; Tsutsumi, Koichiro; Fujii, Masakuni; Kurihara, Naoko; Kato, Hironari; Hirao, Ken; Mizuno, Osamu; Ishida, Etsuji; Okada, Hiroyuki; Yamamoto, Kazuhide [Department of Gastroenterology and Hepatology, Okayama University Graduate School of Medicine, Dentistry, and Pharmaceutical Sciences, 2-5-1 Shikata-cho, Okayama 700-8558 (Japan)

2010-02-15

Background and study aims: Biliary stricture after radiofrequency ablation (RFA) for nodules of hepatocellular carcinoma (HCC) close to major bile ducts sometimes causes septic complications and liver failure. Therefore, it may require interventional drainage for decompression during the follow-up period. The purpose of this study is to clarify the feasibility and safety of bile duct cooling using an endoscopic nasobiliary drainage (ENBD) tube in RFA for HCC close to major bile ducts. Patients and methods: Between August 2003 and July 2007, 14 consecutive patients (14 nodules) undergoing RFA with cooling by an ENBD tube for HCCs close to major bile ducts were enrolled in this study. We infused chilled saline solution via the ENBD tube at 1 ml/s to prevent heat damage during RFA. As controls, 11 patients (13 nodules) undergoing RFA without cooling close to major bile ducts between April 2001 and August 2003 were reviewed. The major outcomes for evaluation were biliary complications and the secondary outcome was local tumor recurrence. Results: There were no significant differences in tumor recurrence between the two groups. However, the rate of biliary complications was significantly lower in the cooling group than in the non-cooling group (0% vs. 39%, P = 0.02). Conclusions: Cooling of bile ducts via an ENBD tube can prevent biliary complications induced by RFA of HCC close to major bile ducts without increasing local recurrence. This technique increases indication of RFA in difficult cases.

Flow-Induced Vibration Measurement of an Inner Cladding Tube in a Simulated Dual-Cooled Fuel Rod

Energy Technology Data Exchange (ETDEWEB)

Lee, Kang Hee; Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho; Kim, Jae Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2009-05-15

To create an internal coolant flow passage in a dual cooled fuel rod, an inner cladding tube cannot have intermediate supports enough to relieve its vibration. Thus it can be suffered from a flow-induced vibration (FIV) more severely than an outer cladding tube which will be supported by series of spacer grids. It may cause a fatigue failure at welding joints on the cladding's end plug or fluid elastic instability of long, slender inner cladding due to decrease of a critical flow velocity. This is one of the challenging technical issues when a dual cooled fuel assembly is to be realized into a conventional reactor core To study an actual vibration phenomenon of a dual cooled fuel rod, FIV tests using a small-scale test bundle are being carried out. Measurement results of inner cladding tube of two typically simulated rods are presented. Causes of the differences in the vibration amplitude and response spectrum of the inner cladding tube in terms of intermediate support condition and pellet stacking are discussed.

Subchannel flow analysis in Candu and ACR pressure tubes with radial and axial diameter variation

Energy Technology Data Exchange (ETDEWEB)

Catana, A.; Prodea, L. [RAAN, Institute for Nuclear Research, Arges (Romania); Danila, N.; Prisecaru, I.; Dupleac, D. [Bucharest Univ. Politehnica(Romania)

2007-07-01

The Candu (Canada Deuterium Uranium) and ACR (Advanced Candu Reactor) are pressure tubes (PT) heavy water moderated reactors. Candu are heavy water and ACR are light water cooled reactors. The pressure tube is filled with 12 bundles, each consisting of 37 respectively 43 fuel rods. One Candu reactor is in operation at Cernavoda, Romania since 1996. ACR is a proposed advanced Candu. PT diameter variation has a significant impact on the thermal-hydraulic parameters. Almost all thermal-hydraulic parameters change, but some of them have a greater significance. In this work we have considered a set of radial and axial PT diameter variations both for Candu-600 and ACR-700 reactors using various types of fuel bundles. We can conclude the following: 1) some thermal-hydraulic parameters are significantly influenced: critical heat flux (CHF), pressure drop, or void fraction; 2) the most significant parameter CHF is worsening which reduces the safety margin; 3) some fuel types present a better thermal-hydraulic behavior; and 4) fuel bundles with fresh fuel or low burnup have a worse thermal-hydraulic behaviour than those at average burn-up.

Subchannel flow analysis in Candu and ACR pressure tubes with radial and axial diameter variation

International Nuclear Information System (INIS)

Catana, A.; Prodea, L.; Danila, N.; Prisecaru, I.; Dupleac, D.

2007-01-01

The Candu (Canada Deuterium Uranium) and ACR (Advanced Candu Reactor) are pressure tubes (PT) heavy water moderated reactors. Candu are heavy water and ACR are light water cooled reactors. The pressure tube is filled with 12 bundles, each consisting of 37 respectively 43 fuel rods. One Candu reactor is in operation at Cernavoda, Romania since 1996. ACR is a proposed advanced Candu. PT diameter variation has a significant impact on the thermal-hydraulic parameters. Almost all thermal-hydraulic parameters change, but some of them have a greater significance. In this work we have considered a set of radial and axial PT diameter variations both for Candu-600 and ACR-700 reactors using various types of fuel bundles. We can conclude the following: 1) some thermal-hydraulic parameters are significantly influenced: critical heat flux (CHF), pressure drop, or void fraction; 2) the most significant parameter CHF is worsening which reduces the safety margin; 3) some fuel types present a better thermal-hydraulic behavior; and 4) fuel bundles with fresh fuel or low burnup have a worse thermal-hydraulic behaviour than those at average burn-up

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.)

Film cooling adiabatic effectiveness measurements of pressure side trailing edge cooling configurations

Directory of Open Access Journals (Sweden)

R. Becchi

2015-12-01

Full Text Available Nowadays total inlet temperature of gas turbine is far above the permissible metal temperature; as a consequence, advanced cooling techniques must be applied to protect from thermal stresses, oxidation and corrosion the components located in the high pressure stages, such as the blade trailing edge. A suitable design of the cooling system for the trailing edge has to cope with geometric constraints and aerodynamic demands; state-of-the-art of cooling concepts often use film cooling on blade pressure side: the air taken from last compressor stages is ejected through discrete holes or slots to provide a cold layer between hot mainstream and the blade surface. With the goal of ensuring a satisfactory lifetime of blades, the design of efficient trailing edge film cooling schemes and, moreover, the possibility to check carefully their behavior, are hence necessary to guarantee an appropriate metal temperature distribution. For this purpose an experimental survey was carried out to investigate the film covering performance of different pressure side trailing edge cooling systems for turbine blades. The experimental test section consists of a scaled-up trailing edge model installed in an open loop suction type test rig. Measurements of adiabatic effectiveness distributions were carried out on three trailing edge cooling system configurations. The baseline geometry is composed by inclined slots separated by elongated pedestals; the second geometry shares the same cutback configuration, with an additional row of circular film cooling holes located upstream; the third model is equipped with three rows of in-line film cooling holes. Experiments have been performed at nearly ambient conditions imposing several blowing ratio values and using carbon dioxide as coolant in order to reproduce a density ratio close to the engine conditions (DR=1.52. To extend the validity of the survey a comparison between adiabatic effectiveness measurements and a prediction by

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

International Nuclear Information System (INIS)

Sim, Woo Gun

2013-01-01

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

Numerical simulation of tubes-in-tube heat exchanger in a mixed refrigerant Joule-Thomson cryocooler

Science.gov (United States)

Damle, R. M.; Ardhapurkar, P. M.; Atrey, M. D.

2017-02-01

Mixed refrigerant Joule-Thomson (MRJT) cryocoolers can produce cryogenic temperatures with high efficiency and low operating pressures. As compared to the high system pressures of around 150-200 bar with nitrogen, the operational pressures with non-azeotropic mixtures (e.g., nitrogen-hydrocarbons) come down to 10-25 bar. With mixtures, the heat transfer in the recuperative heat exchanger takes place in the two-phase region. The simultaneous boiling and condensation of the cold and hot gas streams lead to higher heat transfer coefficients as compared to single phase heat exchange. The two-phase heat transfer in the recuperative heat exchanger drastically affects the performance of a MRJT cryocooler. In this work, a previously reported numerical model for a simple tube-in-tube heat exchanger is extended to a multi tubes-in-tube heat exchanger with a transient formulation. Additionally, the J-T expansion process is also considered to simulate the cooling process of the heat exchanger from ambient temperature conditions. A tubes-in-tube heat exchanger offers more heat transfer area per unit volume resulting in a compact design. Also, the division of flow in multiple tubes reduces the pressure drop in the heat exchanger. Simulations with different mixtures of nitrogen-hydrocarbons are carried out and the numerical results are compared with the experimental data.

A statistical approach to the prediction of pressure tube fracture toughness

International Nuclear Information System (INIS)

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

2008-01-01

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

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)

Experimental and numerical studies of choked flow through adiabatic and diabatic capillary tubes

International Nuclear Information System (INIS)

Deodhar, Subodh D.; Kothadia, Hardik B.; Iyer, K.N.; Prabhu, S.V.

2015-01-01

Capillary tubes are extensively used in several cooling applications like refrigeration, electronic cooling etc. Local pressure variation in adiabatic straight capillary tube (mini channel) is studied experimentally and numerically with R134a as the working fluid. Experiments are performed on two straight capillary tubes. It is found that the diameter is the most sensitive design parameter of the capillary tube. Experiments are performed on five helically coiled capillary tubes to quantify the effect of pitch and curvature of helically coiled capillary tube on the pressure drop. Non dimensionalized factor to account coiling of capillary tube is derived to calculate mass flow rate in helically coiled capillary tubes. Flow visualization in adiabatic capillary tube confirms the bubbly nature of two phase flow. Numerical and experimental investigations in diabatic capillary tube suggest that the use of positive displacement pump and choking at the exit of the channel ensures flow stability. - Highlights: • Model is developed to design capillary tube in adiabatic and diabatic condition. • Effect of coil curvature on pressure drop is studied experimentally. • Correlation is developed to predict mass flow rate in helical capillary tubes. • Flow visualization is carried out to check the type of two phase flow. • Effect of choked flow on diabatic capillary tubes is studied experimentally.

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

A cryogenic tensile testing apparatus for micro-samples cooled by miniature pulse tube cryocooler

International Nuclear Information System (INIS)

Chen, L B; Liu, S X; Gu, K X; Zhou, Y; Wang, J J

2015-01-01

This paper introduces a cryogenic tensile testing apparatus for micro-samples cooled by a miniature pulse tube cryocooler. At present, tensile tests are widely applied to measure the mechanical properties of materials; most of the cryogenic tensile testing apparatus are designed for samples with standard sizes, while for non-standard size samples, especially for microsamples, the tensile testing cannot be conducted. The general approach to cool down the specimens for tensile testing is by using of liquid nitrogen or liquid helium, which is not convenient: it is difficult to keep the temperature of the specimens at an arbitrary set point precisely, besides, in some occasions, liquid nitrogen, especially liquid helium, is not easily available. To overcome these limitations, a cryogenic tensile testing apparatus cooled by a high frequency pulse tube cryocooler has been designed, built and tested. The operating temperatures of the developed tensile testing apparatus cover from 20 K to room temperature with a controlling precision of ±10 mK. The apparatus configurations, the methods of operation and some cooling performance will be described in this paper. (paper)

Structural integrity evaluations of CANDU pressure tubes

International Nuclear Information System (INIS)

Radu, Vasile

2003-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Highlights of the metallurgical behaviour of CANDU pressure tubes

International Nuclear Information System (INIS)

Price, E.G.

1984-10-01

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

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

An experimental investigation on air-side performances of finned tube heat exchangers for indirect air-cooling tower

Directory of Open Access Journals (Sweden)

Du Xueping

2014-01-01

Full Text Available A tremendous quantity of water can be saved if the air cooling system is used, comparing with the ordinary water-cooling technology. In this study, two kinds of finned tube heat exchangers in an indirect air-cooling tower are experimentally studied, which are a plain finned oval-tube heat exchanger and a wavy-finned flat-tube heat exchanger in a cross flow of air. Four different air inlet angles (90°, 60 °, 45°, and 30° are tested separately to obtain the heat transfer and resistance performance. Then the air-side experimental correlations of the Nusselt number and friction factor are acquired. The comprehensive heat transfer performances for two finned tube heat exchangers under four air inlet angles are compared. For the plain finned oval-tube heat exchanger, the vertical angle (90° has the worst performance while 45° and 30° has the best performance at small ReDc and at large ReDc, respectively. For the wavy-finned flat-tube heat exchanger, the worst performance occurred at 60°, while the best performance occurred at 45° and 90° at small ReDc and at large ReDc, respectively. From the comparative results, it can be found that the air inlet angle has completely different effects on the comprehensive heat transfer performance for the heat exchangers with different structures.

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-09-15

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

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

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

Study on corrosion of thermal power plant condenser tubes

Energy Technology Data Exchange (ETDEWEB)

Mohammadi, Abdolreza Rashidi; Zhaam, Ali Akbar [Niroo Research Institute, end of Poonak Bakhtari blvd., Shahrak Ghods, Tehran (Iran)

2004-07-01

The aim of this investigation is to study kinds of corrosion mechanisms in thermal power plant condenser tubes. Condenser is a shell and tube heat exchanger in which cooling water flows through its tubes. While the steam from low pressure turbine passes within condenser tubes, it is condensed by cooling water. The exhausted steam from low pressure turbine is condensed on external surface of condenser tubes and heat is transferred to cooling water which flow into tubes. Tubes composition is usually copper-based alloys, stainless steel or titanium. Annual damages due to corrosion cause much cost for replacement and repairing metallic equipment and installations in electric power industry. Because of existence of different contaminants in water and steam cycle, condenser tubes surfaces are exposed to corrosion. Contaminants like oxygen, carbon dioxide, chloride ion and ammonia in water and steam cycle originate several damages such as pitting and crevice corrosion, erosion, galvanic attack, SCC, condensed corrosion, de-alloying in thermal power plant condenser. The paper first states how corrosion damage takes place in condensers and then introduces types of usual alloys used in condensers and also their corrosion behavior. In continuation, a brief explanation is presented about kinds of condenser failures due to corrosion. Then, causes and locations of different mechanisms of corrosion events on condenser tubes and effects of different parameters such as composition, temperature, chloride and sulfide ion concentration, pH, water velocity and biological precipitation are examined and finally protection methods are indicated. Also some photos of tubes specimens related to power plants are studied and described in each case of mentioned mechanisms. (authors)

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Development and performance of inspection equipment for pressure tubes in Fugen

International Nuclear Information System (INIS)

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

1984-01-01

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

YouTube Video Project: A "Cool" Way to Learn Communication Ethics

Science.gov (United States)

Lehman, Carol M.; DuFrene, Debbie D.; Lehman, Mark W.

2010-01-01

The millennial generation embraces new technologies as a natural way of accessing and exchanging information, staying connected, and having fun. YouTube, a video-sharing site that allows users to upload, view, and share video clips, is among the latest "cool" technologies for enjoying quick laughs, employing a wide variety of corporate activities,…

Evaluation of Fatigue Crack Initiation for Volumetric Flaw in Pressure Tube

International Nuclear Information System (INIS)

Choi, Sung Nam; Yoo, Hyun Joo

2005-01-01

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

GOTHIC code evaluation of alternative passive containment cooling features

International Nuclear Information System (INIS)

Gavrilas, M.; Todreas, E.N.; Driscoll, M.J.

1996-01-01

Reliance on passive cooling has become an important objective in containment design. Several reactor concepts have been set forth, which are equipped with entirely passively cooled containments. However, the problems that have to be overcome in rejecting the entire heat generated by a severe accident in a high-rating reactor (i.e. one with a rating greater than 1200 MW e ) have been found to be substantial and without obvious solutions. The GOTHIC code was verified and modified for containment cooling applications; optimal mesh sizes, computational time steps and applicable heat transfer correlations were examined. The effect of the break location on circulation patterns that develop inside the containment was also evaluated. The GOTHIC code was then employed to assess the effectiveness of several original heat rejection features that make it possible to cool high-rating containments. Two containment concepts were evaluated: one for a 1200 MW e new pressure tube light-water reactor, and one for a 1300 MW e pressurized-water reactor. The effectiveness of various containment configurations that include specific pressure-limiting features has been predicted. The best-performance configurations-worst-case-accident scenarios that were examined yielded peak pressures of less than 0.30 MPa for the 1200 MW e pressure tube light-water reactor, and less than 0.45 MPa for the 1300 MW e pressurized-water reactor. (orig.)

Average Natural Convective Heat Transfer of Air-cooled Condensing Heat Exchanger of Emergency Cooldown Tank - Effect of Tube Banks

International Nuclear Information System (INIS)

Huh, Seon Jeong; Lee, Hee Joon; Kim, Myoung Jun; Moon, Joo Hyung; Bae, Youngmin; Kim, Young-In

2016-01-01

Recently emergency cooldown tank(ECT) is a great concern of passive cooling system for the safety of nuclear reactor. After the operation of a conventional passive cooling system for an extended period, however, the water level falls as a result of the evaporation from the ECT, as steam is emitted from the open top of the tank. In this study, the effect of heat transfer area at the air cooled condensing heat exchanger was investigated by changing 5×5 tube banks into 4×4 and 3×3. Moreover, each of air-side natural convective heat transfer coefficient of tube banks was compared to existing correlations. This study presents the effect of heat transfer area at air-cooled condensing heat exchanger. As heat transfer area decreased, the temperature of outlet increased. In other words, the cooling performance got lower with the decrease of heat transfer area. In addition, the average natural convective heat transfer coefficient was 15.3 W/m"2/K from the 4×4 tube banks, and 4.92 W/m"2/K from the 3×3 tube banks, which had quite a large error more than 46% especially with the value of 4×4 tube banks compared to the value from correlation equation. Therefore, according to this result, it is needed to measure the local heat transfer coefficient of vertical cylinder more elaborately in further study

Average Natural Convective Heat Transfer of Air-cooled Condensing Heat Exchanger of Emergency Cooldown Tank - Effect of Tube Banks

Energy Technology Data Exchange (ETDEWEB)

Huh, Seon Jeong; Lee, Hee Joon [Kookmin University, Seoul (Korea, Republic of); Kim, Myoung Jun; Moon, Joo Hyung; Bae, Youngmin; Kim, Young-In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Recently emergency cooldown tank(ECT) is a great concern of passive cooling system for the safety of nuclear reactor. After the operation of a conventional passive cooling system for an extended period, however, the water level falls as a result of the evaporation from the ECT, as steam is emitted from the open top of the tank. In this study, the effect of heat transfer area at the air cooled condensing heat exchanger was investigated by changing 5×5 tube banks into 4×4 and 3×3. Moreover, each of air-side natural convective heat transfer coefficient of tube banks was compared to existing correlations. This study presents the effect of heat transfer area at air-cooled condensing heat exchanger. As heat transfer area decreased, the temperature of outlet increased. In other words, the cooling performance got lower with the decrease of heat transfer area. In addition, the average natural convective heat transfer coefficient was 15.3 W/m{sup 2}/K from the 4×4 tube banks, and 4.92 W/m{sup 2}/K from the 3×3 tube banks, which had quite a large error more than 46% especially with the value of 4×4 tube banks compared to the value from correlation equation. Therefore, according to this result, it is needed to measure the local heat transfer coefficient of vertical cylinder more elaborately in further study.

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

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.)

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

Experimental study of air-cooled water condensation in slightly inclined circular tube using infrared temperature measurement technique

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyungdae [Nuclear Engineering Department, Kyung Hee University, Yongin (Korea, Republic of); Kwon, Tae-Soon [Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Dong Eok, E-mail: dekim@knu.ac.kr [Department of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

2016-11-15

Highlights: • Air-cooled condensation experiments in an inclined Pyrex glass tube were performed. • High-resolution wall temperature data and flow regime formations could be obtained. • The local heat flux was strongly dependent on the air-side heat transfer. • A CFD analysis was conducted for calculating the local heat flux distribution. - Abstract: This study presents the results of an investigation of the air-cooled water condensation heat transfer characteristics inside a slightly inclined circular tube made of transparent Pyrex glass. The high-resolution wall temperature data and stratified film formations could be obtained with the assistance of an infrared (IR) thermometry technique and side-view visualization using a CCD camera. In all experimental cases, the condensation flow patterns were in the fully-stratified flow region. In addition, the experimentally measured void fraction corresponded well with the logarithmic mean void fraction model. The local temperature differences in the cooling air flow across the condenser tube and high-resolution temperature profiles on the tube’s outer wall were obtained in the experimental measurements. Under the experimental conditions of this study, the local heat flux distributions in the longitudinal direction of the test tube were strongly dependent on the cooling air velocity. And, with the help of IR thermometry, the tube outer wall temperature data at 45 local points could be measured. From the data, the asymmetry distribution of the local wall temperatures and the accurate location of the transition from two-phase mixture to single phase liquid inside the tube could be obtained. Also, the analysis of the thermal resistances by condensation, wall conduction and air convection showed that the air convective heat transfer behavior can play a dominant role to the local heat transfer characteristics. Finally, in order to obtain the local heat flux distribution along the tube’s outer wall, a two

Pressure heat pumping in the orifice pulse-tube refrigerator

International Nuclear Information System (INIS)

Boer, P.C.T. de

1996-01-01

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

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

DEFF Research Database (Denmark)

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

2010-01-01

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

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

DEFF Research Database (Denmark)

Gaihede, Michael; Jacobsen, Henrik; Tveterås, Kjell

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

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)

On random pressure pulses in the turbine draft tube

Science.gov (United States)

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

2017-04-01

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

Microstructural examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

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

2005-01-01

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

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)

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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)

Performance optimization of evacuated tube collector for solar cooling of a house in hot climate

Science.gov (United States)

Ghoneim, Adel A.

2018-02-01

Evacuating the space connecting cover and absorber significantly improves evacuated tube collector (ETC) performance. So, ETCs are progressively utilised all over the world. The main goal of current study is to explore ETC thermal efficiency in hot and severe climate like Kuwait weather conditions. A collector test facility was installed to record ETC thermal performance for one-year period. An extensively developed model for ETCs is presented, employing complete optical and thermal assessment. This study analyses separately optics and heat transfer in the evacuated tubes, allowing the analysis to be extended to different configurations. The predictions obtained are in agreement with experimental. The optimum collector parameters (collector tube length and diameter, mass flow rate and collector tilt angle) are determined. The present results indicate that the optimum tube length is 1.5 m, as at this length a significant improvement is achieved in efficiency for different tube diameters studied. Finally, the heat generated from ETCs is used for solar cooling of a house. Results of the simulation of cooling system indicate that an ETC of area 54 m2, tilt angle of 25° and storage tank volume of 2.1 m3 provides 80% of air-conditioning demand in a house located in Kuwait.

Method of detecting leakage from sealing attached to pressure tube

International Nuclear Information System (INIS)

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

1990-01-01

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

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

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

International Nuclear Information System (INIS)

2013-01-01

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

Measurement and analysis of the re-wetting front velocity during quench cooling of hot horizontal tubes

Energy Technology Data Exchange (ETDEWEB)

Takrouri, Kifah, E-mail: takroukj@mcmaster.ca [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada); Luxat, John, E-mail: luxatj@mcmaster.ca [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada); Hamed, Mohamed [Thermal Processing Laboratory (TPL), Department of Mechanical Engineering, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada)

2017-01-15

Highlights: • Two phase flow & re-wetting front velocity were studied for quench of hot tubes. • The velocity decreased as temperature difference between tube and coolant decreased. • Increasing surface curvature was found to decrease the re-wetting front velocity. • Increasing tube thermal conductivity decreased the velocity. • Correlations were developed to predict the front velocity. - Abstract: When a liquid is put into contact with a hot dry surface, there exists a maximum temperature called the re-wetting temperature below which the liquid is in actual contact with the surface. Re-wetting occurs after destabilization of a vapor film that exists between the hot surface and the liquid. If re-wetting is established at a location on the hot surface, a wet patch appears at that location and starts to spread to cover and cool the entire surface. The outer edge of the wet patch is called the re-wetting front and can proceed only if the surface ahead of it cools down to the re-wetting temperature. Study of re-wetting heat transfer is very important in nuclear reactor safety for limiting the extent of core damage during the early stages of severe accidents after loss of coolant accidents LOCA and is essential for predicting the rate at which the coolant cools an overheated core. One of the important parameters in re-wetting cooling is the velocity at which the re-wetting front moves on the surface. In this study, experimental tests were carried out to investigate the re-wetting front velocity on hot horizontal cylindrical tubes being cooled by a vertical rectangular water multi-jet system. Effects of initial surface temperature in the range 400–740 °C, water subcooling in the range 15–80 °C and jet velocity in the range 0.17–1.43 m/s on the re-wetting front velocity were investigated. The two-phase flow behavior was observed by using a high-speed camera. The re-wetting front velocity was found to increase by increasing water subcooling, decreasing

GOTHIC code evaluation of alternative passive containment cooling features

International Nuclear Information System (INIS)

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

1994-01-01

The GOTHIC code was employed to assess the effectiveness of several original heat rejection features that make it possible to cool large rating containments. The code was first verified and modified for specific containment cooling applications; optimal mesh sizes, computational time steps, and applicable heat transfer correlations were examined. The effect of the break location on circulation patterns that develop inside the containment was also evaluated. GOTHIC was then used to obtain performance predictions for two containment concepts: a 1200 MW e new pressure tube light water reactor, and a 1300 MW e pressurized water reactor. The effectiveness of various containment configurations that include specific pressure-limiting features have been predicted. For the 1200 MW e pressure tube light water reactor, the evaluated pressure-limiting features are: a large water pool connected to the calandria, large containment free volume and an air-convection annulus. For the 1300 MW e pressurized water reactor, an external moat, an internal water pool, and an air-convection annulus were evaluated. The performance of the proposed containment configurations is dependent on the extent of thermal stratification inside the containment. The best-performance configurations/worst-case-accident scenarios that were examined yielded peak pressures of less than 0.30 MPa for the 1200 MW e pressure tube light water reactor, and less than 0.45 MPa for the 1300 MW e pressurized water reactor. The low peak pressure predicted for the 1200 MW e pressure tube light water reactor can be in part attributed to its relatively large free volume, while the relatively high peak pressure predicted for the 1300 MW e pressurized water reactor can be attributed to its relatively small free volume (i.e., the size used was that of a pressurized water reactor containment designed with active heat removal features). (author)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Status and Plans for work on pressure tube creep at AECL

International Nuclear Information System (INIS)

Bickel, Grant A.

2013-01-01

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

Gamma sensitivity of pressurized drift tubes

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-06-01

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

A numerical study on the heat transfer in a swirl-tube heated/cooled on the half periphery of the tube wall

International Nuclear Information System (INIS)

Aoyama, Yoshiyuki; Kunugi, Tomoaki

2002-01-01

Convection heat transfer in a swirl tube was numerically analyzed so as to investigate a characteristic of heat removal when the cooling fluid flows within the swirl tube mounted in a solid structure represented as like a slab. Since the condition of heat inflow was treated as being transmitted only on the one-side surface of the structure, heat conduction through the structure was analyzed in linkage with the convection. Some results for the change in the coefficient of heat transmission along the tube axis are shown. The performance of heat removal was found to be strengthened due to the continuous renovation of thermal boundary layer close to the inside tube surface because the fluid flows in helical motion to shift the range alternate higher and lower temperature. (author)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Pressure relief experiments on a cyclindrical carbon brick tube

International Nuclear Information System (INIS)

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

1978-08-01

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

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

Science.gov (United States)

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

2013-09-01

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

Delayed hydride cracking in Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

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

2007-01-01

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

Pressure tube replacement in Pickering NGS A units 1 and 2

International Nuclear Information System (INIS)

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

1986-10-01

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

Effects of Tube Diameter and Tubeside Fin Geometry on the Heat Transfer Performance of Air-Cooled Condensers

Science.gov (United States)

Wang, H. S.; Honda, Hiroshi

A theoretical study has been made on the effects of tube diameter and tubeside fin geometry on the heat transfer performance of air-cooled condensers. Extensive numerical calculations of overall heat transfer from refrigerant R410A flowing inside a horizontal microfin tube to ambient air were conducted for a typical operating condition of the air-cooled condenser. The tubeside heat transfer coefficient was calculated by applying a modified stratified flow model developed by Wang et al.8). The numerical results show that the effects of tube diameter, fin height, fin number and helix angle of groove are significant, whereas those of the width of flat portion at the fin tip, the radius of round corner at the fin tip and the fin half tip angle are small.

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)

Flooding of a large, passive, pressure-tube LWR

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Microstructural characterization and mechanical properties of Excel alloy pressure tube material

Science.gov (United States)

Sattari, Mohammad

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

Experimental study and modelisation of a pulse tube refrigerator

International Nuclear Information System (INIS)

Ravex, A.; Rolland, P.; Liang, J.

1992-01-01

A test bench for pulse tube refrigerator characterization has been built. In various configurations (basic pulse tube, orifice pulse tube and double inlet pulse tube), the ultimate temperature and the cooling power have been measured as a function of pressure wave amplitude and frequency for various geometries. A lowest temperature of 28 K has been achieved in a single staged double inlet configuration. A modelisation taking into account wall heat pumping, enthalpy flow and regenerator inefficiency is under development. Preliminary calculation results are compared with experimental data

Flow Analysis of Isobutane (R-600A) Inside AN Adiabatic Capillary Tube

Science.gov (United States)

Alok, Praveen; Sahu, Debjyoti

2018-02-01

Capillary tubes are simple narrow tubes but the phase change which occurs inside the capillary tubes is complex to analyze. In the present investigation, an attempt is made to analyze the flow of Isobutane (R-600a) inside the coiled capillary tubes for different load conditions by Homogeneous Equilibrium Model. The Length and diameter of the capillary tube not only depend on the pressure and temperature of the condenser and evaporator but also on the cooling load. The present paper investigates the change in dimensions of the coil capillary tube with respect to the change in cooling load on the system for the constant condenser and evaporator conditions. ANSYS CFX (Central Florida Expressway) software is used to study the flow characteristics of the refrigerant. Appropriate helical coil is selected for this analysis.

Spallation Neutron Source Drift Tube Linac Resonance Control Cooling System Modeling

CERN Document Server

Tang, Johnny Y; Champion, Marianne M; Feschenko, Alexander; Gibson, Paul; Kiselev, Yuri; Kovalishin, A S; Kravchuk, Leonid V; Kvasha, Adolf; Schubert, James P

2005-01-01

The Resonance Control Cooling System (RCCS) for the warm linac of the Spallation Neutron Source was designed by Los Alamos National Laboratory. The primary design focus was on water cooling of individual component contributions. The sizing the RCCS water skid was accomplished by means of a specially created SINDA/FLUINT model tailored to these system requirements. A new model was developed in Matlab Simulink and incorporates actual operational values and control valve interactions. Included is the dependence of RF input power on system operation, cavity detuning values during transients, time delays that result from water flows through the heat exchanger, the dynamic process of water warm-up in the cooling system due to dissipated RF power on the cavity surface, differing contributions on the cavity detuning due to drift tube and wall heating, and a dynamic model of the heat exchanger with characteristics in close agreement to the real unit. Because of the Matlab Simulink model, investigation of a wide range ...

Moderator mixing after a pressure tube failure

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Directory of Open Access Journals (Sweden)

Thomas Berlet

2016-01-01

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

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

International Nuclear Information System (INIS)

Takikawa, N.

1982-01-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Tube Plugging Criterion for the TPCCW Heat Exchanger of Yonggwang NPP 1 and 2

International Nuclear Information System (INIS)

Kim, Hyung Nam; Yoo, Hyun Ju; Choi, Sung Nam; Song, Seok Yoon

2009-01-01

The turbine plant component cooling water(TPCCW) system circulates the cooling water to cool the components in the turbine building and discharges the heat from the components through the TPCCW heat exchanger. Recently, Yonggwang NPP 1 and 2 replaced the TPCCW heat exchanger because of tube degradation. The tubing material of new TPCCW heat exchanger of Yonggwang NPP 1 and 2 is titanium. If the tube wall cannot withstand the pressure, the cooling water with the chemicals flows into the tube side and it is discharged to the open water. The chemicals can pollute the open water. Therefore, the tubes of the TPCCW heat exchanger should be inspected and degraded tubes should be plugged. It is inevitable for the materials of the components to be degraded as the power plants become older. The degradation accompanies increasing maintenance cost as well as creating safety issues. The materials and wall thickness of heat exchanger tubes in nuclear power plants are selected to withstand system temperature, pressure, and corrosion. However, tubes have experienced leaks and failures and plugged based upon eddy current testing (ET) results. There are some problems for plugging the heat exchanger tubes since the criterion and its basis are not clearly described. For this reason, the criteria for the tube wall thickness are addressed in order to operate the heat exchangers in nuclear power plant without trouble during the cycle. There are many codes and standards to be referred for calculating the minimum thickness of the heat exchanger tube in the designing stage. However, the codes and standards related to show the tube plugging criteria may not exist currently. In this paper, a method to establish the tube plugging criteria of BOP heat exchangers, which is based on the USNRC Regulatory Guide 1.121, is introduced and the tube plugging criteria for the TPCCW heat exchanger of Yonggwang NPP No. 1 and 2. This method relies on the similar plugging criteria used in the steam generator

Tube Plugging Criterion for the TPCCW Heat Exchanger of Yonggwang NPP 1 and 2

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyung Nam; Yoo, Hyun Ju; Choi, Sung Nam; Song, Seok Yoon [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2009-05-15

The turbine plant component cooling water(TPCCW) system circulates the cooling water to cool the components in the turbine building and discharges the heat from the components through the TPCCW heat exchanger. Recently, Yonggwang NPP 1 and 2 replaced the TPCCW heat exchanger because of tube degradation. The tubing material of new TPCCW heat exchanger of Yonggwang NPP 1 and 2 is titanium. If the tube wall cannot withstand the pressure, the cooling water with the chemicals flows into the tube side and it is discharged to the open water. The chemicals can pollute the open water. Therefore, the tubes of the TPCCW heat exchanger should be inspected and degraded tubes should be plugged. It is inevitable for the materials of the components to be degraded as the power plants become older. The degradation accompanies increasing maintenance cost as well as creating safety issues. The materials and wall thickness of heat exchanger tubes in nuclear power plants are selected to withstand system temperature, pressure, and corrosion. However, tubes have experienced leaks and failures and plugged based upon eddy current testing (ET) results. There are some problems for plugging the heat exchanger tubes since the criterion and its basis are not clearly described. For this reason, the criteria for the tube wall thickness are addressed in order to operate the heat exchangers in nuclear power plant without trouble during the cycle. There are many codes and standards to be referred for calculating the minimum thickness of the heat exchanger tube in the designing stage. However, the codes and standards related to show the tube plugging criteria may not exist currently. In this paper, a method to establish the tube plugging criteria of BOP heat exchangers, which is based on the USNRC Regulatory Guide 1.121, is introduced and the tube plugging criteria for the TPCCW heat exchanger of Yonggwang NPP No. 1 and 2. This method relies on the similar plugging criteria used in the steam generator

Performance Optimization of the Water Cooling System for Resonance Frequency Control of the PEFP DTL

International Nuclear Information System (INIS)

Kim, K. Y.; Kim, H. K.; Kim, H. S.; Yoon, J. C.; Sohn, Y. K.; Kweon, S. J.; Park, J.; Kim, K. S.

2010-03-01

The objective of in this research project is prototype cooling water skid of separated closed loop in order to supply and withdraw low conductivity deionized water in drift tube of drift tube linac as core components of proton accelerates. This report is dealt with design specification of J-PARC 400 MeV Linac cooling water system, PEFP DTL cooling system, specification of RCCS21-24, RCCS101 with pump, loss coefficient for DTL2 modeling, pressure drop with flow rate of heat exchanger.

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.)

Fundamental research on the cooling characteristic of passive containment cooling system

International Nuclear Information System (INIS)

Kawakubo, M.; Kikura, H.; Aritomi, M.; Inaba, N.; Yamauchi, T.

2004-01-01

The objective of this experimental study is to clarify the heat transfer characteristics of the Passive Containment Cooling System (PCCS) with vertical heat transfer tubes for investigating the influence of non-condensable gas on condensation. Furthermore, hence we obtained new experimental correlation formula to calculate the transients in system temperature and pressure using the simulation program of the PCCS. The research was carried out using a forced circulation experimental loop, which simulates atmosphere inside PCCS with vertical heat transfer tubes if a loss of coolant accident (LOCA) occurs. The experimental facility consists of cooling water supply systems, an orifice flowmeter, and a tank equipped with the heat transfer pipe inside. Cooling water at a constant temperature is injected to the test part of heat transfer pipe vertically installed in the tank by forced circulation. At that time, the temperature of the cooling water between inlet and outlet of the pipe was measured to calculate the overall heat transfer coefficient between the cooling water and atmosphere in the tank. Thus, the heat transfer coefficient between heat transfer surface and the atmosphere in the tank considering the influence of the non-condensable gas was clarified. An important finding of this study is that the amount of condensation in the steamy atmosphere including non-condensable gas depends on the cooling water Reynolds number, especially the concentration of non-condensable gas that has great influence on the amount of condensation. (authors)

Innovative technologies for Faraday shield cooling

International Nuclear Information System (INIS)

Rosenfeld, J.H.; Lindemuth, J.E.; North, M.T.; Goulding, R.H.

1995-01-01

Alternative advanced technologies are being evaluated for use in cooling the Faraday shields used for protection of ion cyclotron range of frequencies (ICR) antennae in Tokamaks. Two approaches currently under evaluation include heat pipe cooling and gas cooling. A Monel/water heat pipe cooled Faraday shield has been successfully demonstrated. Heat pipe cooling offers the advantage of reducing the amount of water discharged into the Tokamak in the event of a tube weld failure. The device was recently tested on an antenna at Oak Ridge National Laboratory. The heat pipe design uses inclined water heat pipes with warm water condensers located outside of the plasma chamber. This approach can passively remove absorbed heat fluxes in excess of 200 W/cm 2 ;. Helium-cooled Faraday shields are also being evaluated. This approach offers the advantage of no liquid discharge into the Tokamak in the event of a tube failure. Innovative internal cooling structures based on porous metal cooling are being used to develop a helium-cooled Faraday shield structure. This approach can dissipate the high heat fluxes typical of Faraday shield applications while minimizing the required helium blower power. Preliminary analysis shows that nominal helium flow and pressure drop can sufficiently cool a Faraday shield in typical applications. Plans are in progress to fabricate and test prototype hardware based on this approach

Evaluation of techniques for inspection and diagnostics of HWR pressure tubes

International Nuclear Information System (INIS)

Choi, Jong-Ho

2008-01-01

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

Deadly pressure pneumothorax after withdrawal of misplaced feeding tube

DEFF Research Database (Denmark)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Comparison of Cooling Different Parts in a High Pressure Ratio Centrifugal Compressor

Directory of Open Access Journals (Sweden)

S. Mostafa Moosania

2016-12-01

Full Text Available Cooling in a centrifugal compressor can improve the performance and reduce the impeller temperature. In a centrifugal compressor, external walls can be cool down, which is known as the shell cooling. This method avoids undesirable effects induced by other cooling methods. Cooling can be applied on different external walls, such as the shroud, diffuser or the back plate. This paper focuses on seeking the most effective cooling place to increase the performance and reduce the impeller temperature. It is found that shroud cooling improves the compressor performance the most. Shroud cooling with 2400 W of cooling power increases the pressure ratio by 4.6% and efficiency by 1.49%. Each 500 W increase in the shroud cooling power, increases the efficiency by 0.3%. Diffuser cooling and back plate cooling have an identical effect on the polytropic efficiency. However, back plate cooling increases the pressure ratio more than diffuser cooling. Furthermore, only back plate cooling reduces the impeller temperature, and with 2400 W of cooling power, the impeller temperature reduces by 45 K.

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

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)

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.

Boussignac continuous positive airway pressure for weaning with tracheostomy tubes

NARCIS (Netherlands)

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

2008-01-01

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

Solar-Enhanced Air-Cooled Heat Exchangers for Geothermal Power Plants

Directory of Open Access Journals (Sweden)

Kamel Hooman

2017-10-01

Full Text Available This paper focuses on the optimization of a Solar-Enhanced Natural-Draft Dry-Cooling Tower (SENDDCT, originally designed by the Queensland Geothermal Energy Centre of Excellence (QGECE, as the air-cooled condenser of a geothermal power plant. The conventional method of heat transfer augmentation through fin-assisted area extension is compared with a metal foam-wrapped tube bundle. Both lead to heat-transfer enhancement, albeit at the expense of a higher pressure drop when compared to the bare tube bundle as our reference case. An optimal design is obtained through the use of a simplified analytical model and existing correlations by maximizing the heat transfer rate with a minimum pressure drop goal as the constraint. Sensitivity analysis was conducted to investigate the effect of sunroof diameter, as well as tube bundle layouts and tube spacing, on the overall performance of the system. Aiming to minimize the flow and thermal resistances for a SENDDCT, an optimum design is presented for an existing tower to be equipped with solar panels to afterheat the air leaving the heat exchanger bundles, which are arranged vertically around the tower skirt. Finally, correlations are proposed to predict the total pressure drop and heat transfer of the extended surfaces considered here.

Heat transfer test in a vertical tube using CO2 at supercritical pressures

International Nuclear Information System (INIS)

Kim, Hwan Yeol; Kim, Hyungrae; Song, Jin Ho; Cho, Bong Hyun; Bae, Yoon Yeong

2007-01-01

Heat transfer test facility, SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), was constructed at KAERI (Korea Atomic Energy Research Institute) for an investigation of the thermal-hydraulic behaviors of supercritical CO 2 at the various geometries of the test section. The test data will be used for the reactor core design of the SCWR (SuperCritical Water-cooled Reactor). As a working fluid, CO 2 was selected to make use of the low critical pressure and temperature of CO 2 compared with water. An experimental study was carried out in the SPHINX to investigate the characteristics of heat transfer and pressure drop at a vertical single tube with an inside diameter of 4.4 mm in case of an upward flow of supercritical CO 2 . The heat and mass fluxes were varied at a given pressure. The mass flux was in the range of 400-1,200 kg/m 2 s and the heat flux was chosen up to 150 kW/m 2 . The selected pressures were 7.75, 8.12, and 8.85 MPa. A heat transfer deterioration occurred at the lower mass fluxes. The experimental heat transfer coefficients were compared with the ones predicted by several existing correlations. The standard deviation was about 20% for each correlation and an apparent discrepancy was not found among the correlations. The major components of the pressure drop were a gravitational pressure drop and a frictional pressure drop. The frictional pressure drop increases as the mass flux and heat flux increase. (author)

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

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)

Emergency core cooling system

International Nuclear Information System (INIS)

Abe, Nobuaki.

1993-01-01

A reactor comprises a static emergency reactor core cooling system having an automatic depressurization system and a gravitationally dropping type water injection system and a container cooling system by an isolation condenser. A depressurization pipeline of the automatic depressurization system connected to a reactor pressure vessel branches in the midway. The branched depressurizing pipelines are extended into an upper dry well and a lower dry well, in which depressurization valves are disposed at the top end portions of the pipelines respectively. If loss-of-coolant accidents should occur, the depressurization valve of the automatic depressurization system is actuated by lowering of water level in the pressure vessel. This causes nitrogen gases in the upper and the lower dry wells to transfer together with discharged steams effectively to a suppression pool passing through a bent tube. Accordingly, the gravitationally dropping type water injection system can be actuated faster. Further, subsequent cooling for the reactor vessel can be ensured sufficiently by the isolation condenser. (I.N.)

Propagation of atmospheric-pressure ionization waves along the tapered tube

Science.gov (United States)

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

2018-02-01

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

A software tool for evaluation of hydrogen ingress in CANDU pressure tubes

International Nuclear Information System (INIS)

Mihalache, Maria; Vasile, Radu; Deaconu, Mariea

2009-01-01

The prediction of hydrogen isotopes concentration into the body and in the rolled joints of operating pressure tubes as a function of reactor hot hours is very important in many fitness-for-service assessments and end of life estimates. The rolled joints are high stress zones with potential for delayed hydride cracking. Predictive models for assessing the long-term deuterium ingress in both body and rolled joint of the pressure tubes have been implemented in a software tool, ROHID, developed in INR-Pitesti. ROHID is a PC-based Windows application with a user-friendly interface that predicts the equivalent hydrogen ingress for Zr-2.5Nb pressure tubes. It uses colour-coded reactor core maps to display the predicted deuterium concentration as a function of time for selected axial locations. Plots of deuterium versus axial location and time for individual pressure tubes are also available. Also, the software tool can predict the exceeding of hydrogen terminal solid solubility (HTSS) from hydrides during precipitation and dissolving processes as a function of time and axial location. (authors)

Reflood Heat Transfer in SiC and Graphene Oxide Coated Tube

International Nuclear Information System (INIS)

Kim, Kyung Mo; Lee, Seung Won; Bang, In Cheol

2013-01-01

The reflood tests have been performed flowing water into bare tube and nanoparticles coated tube at constant flow rate (3 cm/s). The quenching curves have been obtained at atmospheric pressure. Finally, Scanning Electron Microscopy (SEM) images are acquired and contact angles are measured in order to observe the surface structures and wettability effect on cooling performance. The quenching time decreases and quenching velocity increases as the coating time of nanoparticles on the tube increases, because the nanoparticles deposited on the tube destabilize and rupture the vapor film early in the effect of increased Leidenfrost point temperature. The SiC nanoparticles coated tubes have better quenching performance than GO nanoparticles coated tubes. The SEM images and contact angle observations proved the enhanced wettability and rough surface due to deposition of SiC nanoparticles. And the wettability of GO nanoparticles coated tubes shows the increase at 600 s coating. But, the wettability decreases on GO nanoparticles tube coated for 900 s despite the enhanced quenching performance. Thus, the porous structure affects to the better cooling performance in case of GO nanoparticles coated tubes

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

Science.gov (United States)

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

2015-03-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Stress relaxation analysis and irradiation creep and swelling in pressure tubes

International Nuclear Information System (INIS)

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

1979-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Shock tubes: compressions in the low pressure chamber

International Nuclear Information System (INIS)

Schins, H.; Giuliani, S.

1986-01-01

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

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

International Nuclear Information System (INIS)

Gulshani, P.; So, C.B.

1986-10-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Cooling methods of station blackout scenario for LWR plants

International Nuclear Information System (INIS)

2012-01-01

The objective of this study is to analyze the cooling method of station blackout scenario for both the BWR and PWR plants by RELAP5 code and to check the validity of the cooling method proposed by the utilities. In the BWR plant cooling scenario, the Reactor Core Isolation Cooling System (RCIC), which is operated with high pressure steam from the reactor, injects cooling water into the reactor to keep the core water level. The steam generated in the core is released into the suppression pool at containment vessel to condense. To restrict the containment vessel pressure rising, the ventilation from the wet-well is operated. The scenario is analyzed by RELAP5 and CONTEMPT-LT code. In the PWR plant scenario, the primary pressure is decreased by the turbine-driven auxiliary feed water system operated with secondary side steam of the steam generators (SGs). And the core cooling is kept by the natural circulation flow at the primary loop. The analytical method of un-uniform flow behavior among the SG U-tubes, which affects the natural circulation flow rate, is developed. (author)

Heat transfer and pressure drop of condensation of hydrocarbons in tubes

Science.gov (United States)

Fries, Simon; Skusa, Severin; Luke, Andrea

2018-03-01

The heat transfer coefficient and pressure drop are investigated for propane. Two different mild steel plain tubes and saturation pressures are considered for varying mass flux and vapour quality. The pressure drop is compared to the Friedel-Correlation with two different approaches to determine the friction factor. The first is calculation as proposed by Friedel and the second is through single phase pressure drop investigations. For lower vapour qualities the experimental results are in better agreement with the approach of the calculated friction factor. For higher vapour qualities the experimental friction factor is more precise. The pressure drop increases for a decreasing tube diameter and saturation pressure. The circumferential temperature profile and heat transfer coefficients are shown for a constant vapour quality at varying mass fluxes. The subcooling is highest for the bottom of the tube and lowest for the top. The average subcooling as well as the circumferential deviation decreases for rising mass fluxes. The averaged heat transfer coefficients are compared to the model proposed by Thome and Cavallini. The experimental results are in good agreement with both correlations, however the trend is better described with the correlation from Thome. The experimental heat transfer coefficients are under predicted by Thome and over predicted by Cavallini.

Complex cooling water systems optimization with pressure drop consideration

CSIR Research Space (South Africa)

Gololo, KV

2012-12-01

Full Text Available Pressure drop consideration has shown to be an essential requirement for the synthesis of a cooling water network where reuse/recycle philosophy is employed. This is due to an increased network pressure drop associated with additional reuse...

Experiments on vibration of heat exchanger tube arrays in cross flow

International Nuclear Information System (INIS)

Blevins, R.D.; Gibert, R.J.; Villard, B.

1981-08-01

A series of tests have been made at the Commissariat a l'Energie Atomique, in cooperation with General Atomic Company, SAN DIEGO (U.S.A.) on the flow-induced vibration of heat exchanger tube bundles in cross flow. These tests were made in air on tube bundles which simulated heat exchangers in the high temperature gas cooled reactors. The tests were of two types. In the first type, an instrumented tube was inserted at various locations into a tube bundle. Measurements were made of pressure at a number of points along the tube and about the circumference of the tube. These measurements were processed to obtain the spectra of turbulent pressure fluctuations on the tube, the spanwise correlation and the lift force. The second set of tests was made on tube bundles with flexible tubes. As the flow velocity was increased, these tests clearly show an instability. Nine tube configurations were tested with both plastic and metallic tubes and the effect of tube-to-tube difference in natural frequency was investigated

Detection of pressure tube leaks relying on moisture beetles only

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Study on the manufacturing process, causes of the pressure tube failure and methods for improving its performance

Energy Technology Data Exchange (ETDEWEB)

You, Ho Sik; Jeong, Jin Kon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-07-01

Manufacturing processes of Zr-2.5Nb pressure tube used in CANDU reactor, effects of impurities on the properties of the pressure tube, experiences and causes of the pressure tube cracking accident and the development programs on the fuel channel at AECL have been described. Fabrication processes on the pressure tube have been explained in detail from the sponge production step to the final product. Test methods that are performed to verify the integrity of the final product have also been described. Most of the pressure tube rupture accidents were caused by DHC (Delayed Hydride Cracking). In cases of Pickering units 3 and 4 and Bruce unit 2, excessive residual stresses induced by improper rolled joint process had played a role to cause DHC. In Pickering unit 2, cracks formed by contact between pressure and calandria tubes due to the movement of garter spring were direct cause of failure. After the accidents, a lot of R and D programs on each component of the fuel channel have been carried out. The study on the improvement of manufacturing processes such as increasing cold working rate, performing the intermediate and final annealing and adding the third element like Fe, V, Cr for enhancing the pressure tube performance are on progress. To suppress hydrogen uptake into the pressure tube, the methods such as zirconia coating on the pressure tube, Cr-plating on the end fitting and placing the yttrium getter on the pressure tube are considered. Experiments on each test specimen are currently under way. Owing to such an effort, more advanced fuel channel can be installed in the next CANDU reactor. 6 tabs., 20 figs., 20 refs. (Author).

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

International Nuclear Information System (INIS)

Kaempf, F.; Pietsch, H.B.

1978-01-01

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

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.)

The evolution of US helium-cooled blankets

International Nuclear Information System (INIS)

Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.

1991-01-01

This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors. (orig.)

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

Science.gov (United States)

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

2011-09-01

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

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

International Nuclear Information System (INIS)

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

2000-12-01

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

Failure Pressure Estimates of Steam Generator Tubes Containing Wear-type Defects

International Nuclear Information System (INIS)

Yoon-Suk Chang; Jong-Min Kim; Nam-Su Huh; Young-Jin Kim; Seong Sik Hwang; Joung-Soo Kim

2006-01-01

It is commonly requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. The critical defect dimensions have been determined based on the concept of plastic instability. This criterion, however, is known to be too conservative for some locations and types of defects. In this context, the accurate failure estimation for steam generator tubes with a defect draws increasing attention. Although several guidelines have been developed and are used for assessing the integrity of defected tubes, most of these guidelines are related to stress corrosion cracking or wall-thinning phenomena. As some of steam generator tubes are also failed due to fretting and so on, alternative failure estimation schemes for relevant defects are required. In this paper, three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of steam generator tubes with different defect configurations; elliptical wastage type, wear scar type and rectangular wastage type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of the steam generator tube. After investigating the effect of key parameters such as wastage depth, wastage length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wastage region. Comparison of failure pressures predicted according to the proposed estimation scheme with some corresponding burst test data shows good agreement, which provides a confidence in the use of the proposed equations to assess the integrity of steam generator tubes with wear-type defects. (authors)

Hierarchically structured nanoporous carbon tubes for high pressure carbon dioxide adsorption

Directory of Open Access Journals (Sweden)

Julia Patzsch

2017-05-01

Full Text Available Mesoscopic, nanoporous carbon tubes were synthesized by a combination of the Stoeber process and the use of electrospun macrosized polystyrene fibres as structure directing templates. The obtained carbon tubes have a macroporous nature characterized by a thick wall structure and a high specific surface area of approximately 500 m²/g resulting from their micro- and mesopores. The micropore regime of the carbon tubes is composed of turbostratic graphitic areas observed in the microstructure. The employed templating process was also used for the synthesis of silicon carbide tubes. The characterization of all porous materials was performed by nitrogen adsorption at 77 K, Raman spectroscopy, infrared spectroscopy, thermal gravimetric analysis (TGA, scanning electron microscopy (SEM as well as transmission electron microscopy (TEM. The adsorption of carbon dioxide on the carbon tubes at 25 °C at pressures of up to 30 bar was studied using a volumetric method. At 26 bar, an adsorption capacity of 4.9 mmol/g was observed. This is comparable to the adsorption capacity of molecular sieves and vertically aligned carbon nanotubes. The high pressure adsorption process of CO2 was found to irreversibly change the microporous structure of the carbon tubes.

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.

Deformation of in-service pressure tubes

International Nuclear Information System (INIS)

Sarce, A.L.

1993-01-01

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

A statistical method for draft tube pressure pulsation analysis

International Nuclear Information System (INIS)

Doerfler, P K; Ruchonnet, N

2012-01-01

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

A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

Fuel-element vibration and bearing pad to pressure tube fretting

International Nuclear Information System (INIS)

Fisher, N.J.; Taylor, C.E.; Pettigrew, M.J.

1990-08-01

Fuel channel operation under boiling condition results in increased flow velocities, which may lead to unacceptable fuel-element vibration and bearing pad to pressure tube fretting. The existing endurance test database does not fully cover the range of future channel operating conditions. In particular, after refuelling, some channels for future designs may operate with two-phase flow conditions outside the range of endurance test conditions. Full-scale endurance testing at realistic steam-water conditions involves substantial energy costs. Therefore, fundamental laboratory investigations were conducted to define and endurance test matrix which adequately envelops the future range of operating conditions while minimizing both the number of tests and the energy requirement of individual tests. The main focus of the laboratory investigations was to establish the relationships between: fuel channel flow conditions and fuel-element vibration; and fuel-element vibration and bearing pad to pressure tube fretting. The vibration response of a single fuel element was measured over a wide range of operating conditions covering realistic fuel channel conditions and simulated endurance testing conditions. For higher void fractions, the vibration amplitudes measured in air/water were much higher than in steam/water, while for low void fractions, the amplitudes were similar. The measured amplitudes in steam/water varied very little over the range of temperature and pressure investigated. The effects of temperature, pressure tube oxide thickness, vibration amplitude and bearing pad manufacturer on pressure tube fretting were investigated. The fretting rate is extremely temperature dependent. For vibration amplitudes about three or four times greater than expected in-reactor conditions, peak fretting rates were observed in the 225 to 286 degrees C temperature range. Fretting rates were seven times less at the higher temperatures of 300 and 315 degrees C, and the lower temperatures

Falling film flow, heat transfer and breakdown on horizontal tubes

International Nuclear Information System (INIS)

Rogers, J.T.

1980-11-01

Knowledge of falling film flow and heat transfer characteristics on horizontal tubes is required in the assessment of certain CANDU reactor accident sequences for those CANDU reactors which use moderator dump as one of the shut-down mechanisms. In these reactors, subsequent cooling of the calandria tubes is provided by falling films produced by sprays. This report describes studies of falling film flow and heat transfer characteristics on horizontal tubes. Analyses using integral methods are given for laminar and turbulent flow, ignoring and accounting for momentum effects in the film. Preliminary experiments on film flow stability on horizontal tubes are described and various mechanisms of film breakdown are examined. The work described in this report shows that in LOCA with indefinitely delayed ECI in the NPD or Douglas Point (at 70 percent power) reactors, the falling films on the calandria tubes will not be disrupted by any of the mechanisms considered, provided that the pressure tubes do not sag onto the calandria tubes. However, should the pressure tubes sag onto the calandria tubes, film disruption will probably occur

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

International Nuclear Information System (INIS)

Agostini, F.

2008-07-01

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

Pressure Separators for District Cooling; Tryckvaexlare foer fjaerrkyla - Teknik och funktion

Energy Technology Data Exchange (ETDEWEB)

Gustafson, Bror-Arne [Fludex AB, Goeteborg (SE)] [and others

2003-07-01

About 10 pressure separators run today in commercial operation in district heating networks. Undoubtedly, the pressure separator has become a new tool for more efficient operation of district heating systems. The pressure separator makes it possible to keep different parts of the pipe network at different pressure levels without the unavoidable temperature losses of heat exchangers. The objectives of this project are to find the answers of two questions. The first question is: Will the pressure separator function in district cooling systems if it is designed in the same way as for district heating? The only difference should then be the temperatures of operation. The second question is: Is there a modified design that will perform better in district cooling systems? To find the answers of the two questions, a test rig was built in Rosenlund Power Station in Goeteborg. Also computer simulations were carried out to clarify actual phenomena. The answer to the first question is: Yes/no. Measurements show that the original design of the pressure separator can be made working at temperatures of operation typical for district cooling. It will, however, be very sensitive and is not recommended for practical applications. The answer to the second question is: Yes, there is. This makes the details of the original design less interesting. The modified design is characterized by 'Differential pressure control' instead of temperature layer control'. This means a working principle that is completely independent of operational temperatures. The TVX-tank is replaced by a 'short cut' with a spring loaded checkvalve. One of the control valves creates a differential pressure that is too low to open the check valve but high enough to keep the next checkvalve in the next short cut closed. Measurements and computer simulations show that the pressure separator in the modified design works very well in district cooling applications. The temperature losses of heat

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.

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

International Nuclear Information System (INIS)

Yetisir, M.; Fisher, N.J.

1997-01-01

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

Proliferation attractiveness of nuclear material in a small modular pressure tube SCWR

Energy Technology Data Exchange (ETDEWEB)

McDonald, M.; Pencer, J., E-mail: mcdonamh@aecl.ca, E-mail: pencerj@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

The SuperSafe© Reactor (SSR), has been recently proposed as a small modular version of the Canadian supercritical water cooled reactor (SCWR). This reactor is a heavy water moderated, pressure tube reactor using supercritical light water as coolant. The current SSR design is to generate 300 MWe taking advantage of the expected high thermal efficiency (assumed 45%). As one of the reactor types being considered by the Generation-IV International Forum, it is expected that this SCWR design will feature enhanced proliferation resistance over current generation technologies. Proliferation resistance assessments are wide-ranging, multidisciplinary efforts that are typically performed at a number of levels, from a state level down to a specific facility level. One small, but particularly important, sub-assessment is that of nuclear material attractiveness, that is, assessing the quality of nuclear materials throughout the fuel cycle for use in making a nuclear explosive device. The attractiveness of materials for three different SSR fuel options is examined in this work. (author)

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)

Airfoil, platform, and cooling passage measurements on a rotating transonic high-pressure turbine

Science.gov (United States)

Nickol, Jeremy B.

An experiment was performed at The Ohio State University Gas Turbine Laboratory for a film-cooled high-pressure turbine stage operating at design-corrected conditions, with variable rotor and aft purge cooling flow rates. Several distinct experimental programs are combined into one experiment and their results are presented. Pressure and temperature measurements in the internal cooling passages that feed the airfoil film cooling are used as boundary conditions in a model that calculates cooling flow rates and blowing ratio out of each individual film cooling hole. The cooling holes on the suction side choke at even the lowest levels of film cooling, ejecting more than twice the coolant as the holes on the pressure side. However, the blowing ratios are very close due to the freestream massflux on the suction side also being almost twice as great. The highest local blowing ratios actually occur close to the airfoil stagnation point as a result of the low freestream massflux conditions. The choking of suction side cooling holes also results in the majority of any additional coolant added to the blade flowing out through the leading edge and pressure side rows. A second focus of this dissertation is the heat transfer on the rotor airfoil, which features uncooled blades and blades with three different shapes of film cooling hole: cylindrical, diffusing fan shape, and a new advanced shape. Shaped cooling holes have previously shown immense promise on simpler geometries, but experimental results for a rotating turbine have not previously been published in the open literature. Significant improvement from the uncooled case is observed for all shapes of cooling holes, but the improvement from the round to more advanced shapes is seen to be relatively minor. The reduction in relative effectiveness is likely due to the engine-representative secondary flow field interfering with the cooling flow mechanics in the freestream, and may also be caused by shocks and other

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-15

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

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

International Nuclear Information System (INIS)

Oh, Young-Jin; Chang, Yoon-Suk

2014-01-01

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

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

Science.gov (United States)

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

2015-01-01

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

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

Experimental studies on the crystallographic and plastic anisotropies of zircaloy-4 tubes

International Nuclear Information System (INIS)

Costa Viana, C.S. da

1982-01-01

The crystallographic and plastic anisotropies of a zircaloy-4 tubing using direct pole figures and experimental yield loci are analyzed. Tensile and plane-strain compression tests were used to assess the mecahnical behaviour. The results are discussed with respect to the dimensional stability and mechanical behaviour expected for the tube in its use in the core of pressurized water cooled reactors. (Author) [pt

Design of a high-pressure single pulse shock tube for chemical kinetic investigations

International Nuclear Information System (INIS)

Tranter, R. S.; Brezinsky, K.; Fulle, D.

2001-01-01

A single pulse shock tube has been designed and constructed in order to achieve extremely high pressures and temperatures to facilitate gas-phase chemical kinetic experiments. Postshock pressures of greater than 1000 atmospheres have been obtained. Temperatures greater than 1400 K have been achieved and, in principle, temperatures greater than 2000 K are easily attainable. These high temperatures and pressures permit the investigation of hydrocarbon species pyrolysis and oxidation reactions. Since these reactions occur on the time scale of 0.5--2 ms the shock tube has been constructed with an adjustable length driven section that permits variation of reaction viewing times. For any given reaction viewing time, samples can be withdrawn through a specially constructed automated sampling apparatus for subsequent species analysis with gas chromatography and mass spectrometry. The details of the design and construction that have permitted the successful generation of very high-pressure shocks in this unique apparatus are described. Additional information is provided concerning the diaphragms used in the high-pressure shock tube

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)

Tracheal tube and laryngeal mask cuff pressure during anaesthesia - mandatory monitoring is in need

DEFF Research Database (Denmark)

Rokamp, K.Z.; Secher, N.H.; Møller, Ann

2010-01-01

ABSTRACT: BACKGROUND: To prevent endothelium and nerve lesions, tracheal tube and laryngeal mask cuff pressure is to be maintained at a low level and yet be high enough to secure air sealing. METHOD: In a prospective quality-control study, 201 patients undergoing surgery during anaesthesia (without...... the use of nitrous oxide) were included for determination of the cuff pressure of the tracheal tubes and laryngeal masks. RESULTS: In the 119 patients provided with a tracheal tube, the median cuff pressure was 30 (range 8 - 100) cm H2O and the pressure exceeded 30 cm H2O (upper recommended level) for 54...... patients. In the 82 patients provided with a laryngeal mask, the cuff pressure was 95 (10 - 121) cm H2O and above 60 cm H2O (upper recommended level) for 56 patients and in 34 of these patients, the pressure exceeded the upper cuff gauge limit (120 cm H2O). There was no association between cuff pressure...

Fuel bundle to pressure tube fretting in Bruce and Darlington

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Fuel bundle to pressure tube fretting in Bruce and Darlington

International Nuclear Information System (INIS)

Norsworthy, A.G.; Ditschun, A.

1995-01-01

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

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

International Nuclear Information System (INIS)

Majumdar, S.; Kasza, K.

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Experimental and visual study on flow patterns and pressure drops in U-tubes

International Nuclear Information System (INIS)

Da Silva Lima, J. R.

2011-01-01

In single- and two-phase flow heat exchangers (in particular 'coils'), besides the straight tubes there are also many singularities, in particular the 180° return bends (also called return bends or U-bends). However, contrary to the literature concerning pressure drops and heat transfer in straight tubes, where many experimental data and predicting methods are available, only a limited number of studies concerning U-bends can be found. Neither reliable experimental data nor proven prediction methods are available. Indeed, flow structure, pressure drop and heat transfer in U-bends are an old unresolved design problem in the heat transfer industry. Thus, the present study aims at providing further insight on two-phase pressure drops and flows patterns in U-bends. Based on a new type of U-bend test section, an extensive experimental study was conducted. The experimental campaign covered five test sections with three internal diameters (7.8, 10.8 and 13.4 mm), five bend diameters (24.8, 31.7, 38.1, 54.8 and 66.1 mm), tested for three orientations (horizontal, vertical upflow and vertical downflow), two fluids (R134a and R410A), two saturation temperatures (5 and 10 °C) and mass velocities ranging from 150 to 1000 kg s -1 m -2 . The flow pattern observations identified were stratified-wavy, slug-stratified-wavy, intermittent, annular, dryout and mist flows. The effects of the U-bend on the flow patterns were also observed. A total of 5655 pressure drop data were measured at seven different locations in the test section ( straight tubes and U-bend) providing a total of almost 40,000 data points. The straight tube data were first used to improve the actual two-phase straight tube model of Moreno-Quibén and Thome. This updated model was then used to developed a two-phase U-bend pressure drop model. Based on a comparison between experimental and predicted values, it is concluded that the new two-phase frictional pressure drop model for U-bends successfully

Effect of pressure on the vacuum cooling of iceberg lettuce

Energy Technology Data Exchange (ETDEWEB)

Ozturk, Hande Mutlu [Pamukkale University, Food Engineering Department, Faculty of Engineering, Denizli (Turkey); Ozturk, Harun Kemal [Pamukkale University, Mechanical Engineering Department, Faculty of Engineering, 20070 Kinikli, Denizli (Turkey)

2009-05-15

Vacuum cooling is known as a rapid evaporative cooling technique for any porous product which has free water. The aim of this paper is to apply vacuum cooling technique to the cooling of the iceberg lettuce and show the pressure effect on the cooling time and temperature decrease. The results of vacuum cooling are also compared with conventional cooling (cooling in refrigerator) for different temperatures. Vacuum cooling of iceberg lettuce at 0.7 kPa is about 13 times faster than conventional cooling of iceberg lettuce at 6 C. It has been also found that it is not possible to decrease the iceberg lettuce temperature below 10 C if vacuum cooling method is used and vacuum pressure is set to 1.5 kPa. (author) [French] Le refroidissement sous vide est connu comme une technique evaporative rapide refroidissant pour n'importe quel produit poreux qui a de l'eau libre. Le but de ce papier est d'appliquer le refroidissement sous vide pour le refroidissement de la laitue et examiner l'effet de la pression sur le temps de refroidissement et la diminution de temperature. Les resultats de refroidissement sous vide sont aussi compares avec le refroidissement conventionnel (refroidissement dans le refrigerateur) pour les differentes temperatures. Le refroidissement a vide de laitue a 0.7 kPa est environ 13 fois plus vite que le refroidissement conventionnel de laitue croquante a 6 C. Il a ete aussi constate qu'il n'est pas possible de diminuer la temperature de laitue ci-dessous 10 C si le refroidissement sous vide est utilise comme methode et la pression a vide est montree a 1.5 kPa. (orig.)

Comparison of enterotomy leak pressure among fresh, cooled, and frozen-thawed porcine jejunal segments.

Science.gov (United States)

Aeschlimann, Kimberly A; Mann, F A; Middleton, John R; Belter, Rebecca C

2018-05-01

OBJECTIVE To determine whether stored (cooled or frozen-thawed) jejunal segments can be used to obtain dependable leak pressure data after enterotomy closure. SAMPLE 36 jejunal segments from 3 juvenile pigs. PROCEDURES Jejunal segments were harvested from euthanized pigs and assigned to 1 of 3 treatment groups (n = 12 segments/group) as follows: fresh (used within 4 hours after collection), cooled (stored overnight at 5°C before use), and frozen-thawed (frozen at -12°C for 8 days and thawed at room temperature [23°C] for 1 hour before use). Jejunal segments were suspended and 2-cm enterotomy incisions were made on the antimesenteric border. Enterotomies were closed with a simple continuous suture pattern. Lactated Ringer solution was infused into each segment until failure at the suture line was detected. Leak pressure was measured by use of a digital transducer. RESULTS Mean ± SD leak pressure for fresh, cooled, and frozen-thawed segments was 68.3 ± 23.7 mm Hg, 55.3 ± 28.1 mm Hg, and 14.4 ± 14.8 mm Hg, respectively. Overall, there were no significant differences in mean leak pressure among pigs, but a significant difference in mean leak pressure was detected among treatment groups. Mean leak pressure was significantly lower for frozen-thawed segments than for fresh or cooled segments, but mean leak pressure did not differ significantly between fresh and cooled segments. CONCLUSIONS AND CLINICAL RELEVANCE Fresh porcine jejunal segments or segments cooled overnight may be used for determining intestinal leak pressure, but frozen-thawed segments should not be used.

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

International Nuclear Information System (INIS)

Royen Van, E.

2011-11-01

The complexity of two-phase flow boiling on a tube bundle presents many challenges to the understanding of the physical phenomena taking place. It is important to quantify these numerous heat flow mechanisms in order to better describe the performance of tube bundles as a function of the operational conditions. In the present study, the bundle boiling facility at the Laboratory of Heat and Mass Transfer (LTCM) was modified to obtain high-speed videos to characterise the two-phase regimes and some bubble dynamics of the boiling process. It was then used to measure heat transfer on single tubes and in bundle boiling conditions. Pressure drop measurements were also made during adiabatic and diabatic bundle conditions. New enhanced boiling tubes from Wolverine Tube Inc. (Turbo-B5) and the Wieland-Werke AG (Gewa-B5) were investigated using R134a and R236fa as test fluids. The tests were carried out at saturation temperatures T sat of 5 °C and 15 °C, mass flow rates from 4 to 35 kg/m 2 s and heat fluxes from 15 to 70 kW/m 2 , typical of actual operating conditions. The flow pattern investigation was conducted using visual observations from a borescope inserted in the middle of the bundle. Measurements of the light attenuation of a laser beam through the intertube two-phase flow and local pressure fluctuations with piezo-electric pressure transducers were also taken to further help in characterising the complex flow. Pressure drop measurements and data reduction procedures were revised and used to develop new, improved frictional pressure drop prediction methods for adiabatic and diabatic two-phase conditions. The physical phenomena governing the enhanced tube evaporation process and their effects on the performance of tube bundles were investigated and insight gained. A new method based on a theoretical analysis of thin film evaporation was used to propose a new correlating parameter. A large new database of local heat transfer coefficients were obtained and then

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)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

Directory of Open Access Journals (Sweden)

Rafael de Macedo Coelho

2016-04-01

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

A passive emergency heat sink for water cooled reactors with particular application to CANDU reg-sign reactors

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU reg-sign moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners

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.)

A Method to Establishing Tube Plugging Criterion for Heat Exchangers with Straight Tubes

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyungnam [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The difference of thermal expansion coefficients between the shell and tube materials causes the stress in axial direction of tube. Because of the axial stress due to thermal load, the straight tubes are used for heat exchangers operated in low temperature such as CCW (Component Cooling Water) heat exchangers and condensers. It is inevitable for the materials of the components to be degraded as the power plants become older. The degradation accompanies increasing maintenance cost as well as creating safety issues. The materials and wall thickness of heat exchanger tubes in nuclear power plants are selected to withstand system temperature, pressure, and corrosion. There are many codes and standards to be referred for calculating the minimum thickness of the heat exchanger tube in the designing stage. However, the codes and standards related to show the tube plugging criteria may not exist currently. In this paper, a method to establish the tube plugging criteria of BOP heat exchangers, which is based on the USNRC Regulatory Guide 1.121, is introduced and the tube plugging criteria for the TPCCW heat exchanger of Yonggwang NPP No. 1 and 2. A method to establish the tube plugging criteria of heat exchangers with straight tubes are introduced based on the USNRC Regulatory Guide 1.121. As an example, the tube plugging criterion for the CCW heat exchanger of a nuclear power plant is provided.

Manufacture of thin-walled clad tubes by pressure welding of roll bonded sheets

Science.gov (United States)

Schmidt, Hans Christian; Grydin, Olexandr; Stolbchenko, Mykhailo; Homberg, Werner; Schaper, Mirko

2017-10-01

Clad tubes are commonly manufactured by fusion welding of roll bonded metal sheets or, mechanically, by hydroforming. In this work, a new approach towards the manufacture of thin-walled tubes with an outer diameter to wall thickness ratio of about 12 is investigated, involving the pressure welding of hot roll bonded aluminium-steel strips. By preparing non-welded edges during the roll bonding process, the strips can be zip-folded and (cold) pressure welded together. This process routine could be used to manufacture clad tubes in a continuous process. In order to investigate the process, sample tube sections with a wall thickness of 2.1 mm were manufactured by U-and O-bending from hot roll bonded aluminium-stainless steel strips. The forming and welding were carried out in a temperature range between RT and 400°C. It was found that, with the given geometry, a pressure weld is established at temperatures starting above 100°C. The tensile tests yield a maximum bond strength at 340°C. Micrograph images show a consistent weld of the aluminium layer over the whole tube section.

Effect of Ovality on Maximum External Pressure of Helically Coiled Steam Generator Tubes with a Rectangular Wear

Energy Technology Data Exchange (ETDEWEB)

Shin, Dong In; Lim, Eun Mo; Huh, Nam Su [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of); Choi, Shin Beom; Yu, Je Yong; Kim, Ji Ho; Choi, Suhn [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

A structural integrity of steam generator tubes of nuclear power plants is one of crucial parameters for safe operation of nuclear power plants. Thus, many studies have been made to provide engineering methods to assess integrity of defective tubes of commercial nuclear power plants considering its operating environments and defect characteristics. As described above, the geometric and operating conditions of steam generator tubes in integral reactor are significantly different from those of commercial reactor. Therefore, the structural integrity assessment of defective tubes of integral reactor taking into account its own operating conditions and geometric characteristics, i. e., external pressure and helically coiled shape, should be made to demonstrate compliance with the current design criteria. Also, ovality is very specific characteristics of the helically coiled tube because it is occurred during the coiling processes. The wear, occurring from FIV (Flow Induced Vibration) and so on, is main degradation of steam generator tube. In the present study, maximum external pressure of helically coiled steam generator tube with wear is predicted based on the detailed 3-dimensional finite element analysis. As for shape of wear defect, the rectangular shape is considered. In particular, the effect of ovality on the maximum external pressure of helically coiled tubes with rectangular shaped wear is investigated. In the present work, the maximum external pressure of helically coiled steam generator tube with rectangular shaped wear is investigated via detailed 3-D FE analyses. In order to cover a practical range of geometries for defective tube, the variables affecting the maximum external pressure were systematically varied. In particular, the effect of tube ovality on the maximum external pressure is evaluated. It is expected that the present results can be used as a technical backgrounds for establishing a practical structural integrity assessment guideline of

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-10-15

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

An assessment of the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes of PHWR

International Nuclear Information System (INIS)

Sah, D.N.

1992-01-01

In view of the deleterious effect of hydriding on the operating life of zircaloy-2 pressure tubes in PHWRs there is an urgent need for the assessment of the status of the pressure tubes with respect to corrosion and hydrogen pick-up in the operating PHWRs. A model has been developed for analysing the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes under reactor operating conditions. This model predicts the axial profiles of oxide layer thickness and hydrogen pick-up in the pressure tubes as a function of the operating time of the reactor. The prediction of hydrogen pick-up by the model in the F-10 pressure tube of RAPS-I have been found to be in good agreement with the measured value of hydrogen content. This report gives a brief description of the model and its predictions on the present status of hydrogen pick-up in the pressure tubes of lead reactor RAPS-II. (author). 6 refs., 5 figs., 2 tabs

Calculation of cooling tower plumes for high pressure wintry situations

International Nuclear Information System (INIS)

Gassmann, F.; Tinguely, M.; Haschke, D.

1982-12-01

The diffusion of the plumes of the projected nuclear power plants at Kaiseraugst and Schwoerstadt, during high pressure wintry conditions, has been examined using a mathematical model to simulate the plumes. For these calculations, microaerological measurements were made in the proximity of Kaiseraugst and Schwoerstadt. These give a typical image of the weather during high pressure wintry conditions, which is normally associated with an inversion, sometimes strong, at a low height. Dry cooling towers with natural draught, which offer an alternative solution to the wet cooling towers proposed for Kasieraugst, are examined equally. (Auth./G.T.H.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Development of Evaluation Technology of the Integrity of HWR Pressure Tubes

Energy Technology Data Exchange (ETDEWEB)

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

2007-06-15

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

Study on frictional pressure drop of steam-water two phase flow in optimized four-head internal-ribbed tube

International Nuclear Information System (INIS)

Wang Weishu; Zhu Xiaojing; Bi Qincheng; Wu Gang; Yu Shuiqing

2012-01-01

The optimized internal-ribbed tube is different from the normal internal-ribbed tube on the frictional pressure drop characteristics. The frictional pressure drop characteristics of steam-water two phase flow in horizontal four-head optimized internal-ribbed were studied under adiabatic condition. According to the experimental and calculation results, the two-phase multiplier is greatly affected by the steam quality and pressure. The two-phase multiplier increases with increasing quality, and decreases with increasing pressure. In the near-critical pressure region, the two-phase multiplier is close to 1. The frictional pressure drop of two phase flow in optimized tube is less than that in the normal tube under the same work condition. The good hydrodynamic condition could be achieved when the optimized internal-ribbed tube is used in the heat transfer equipment because the self-compensating characteristics exist due to the reduction of frictional pressure drop. (authors)

Final report on development evaluation of Task Group 3 pressure tubes

International Nuclear Information System (INIS)

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

1983-11-01

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

A pulse tube cryocooler with a cold reservoir

Science.gov (United States)

Zhang, X. B.; Zhang, K. H.; Qiu, L. M.; Gan, Z. H.; Shen, X.; Xiang, S. J.

2013-02-01

Phase difference between pressure wave and mass flow is decisive to the cooling capacity of regenerative cryocoolers. Unlike the direct phase shifting using a piston or displacer in conventional Stirling or GM cryocoolers, the pulse tube cyocooler (PTC) indirectly adjusts the cold phase due to the absence of moving parts at the cold end. The present paper proposed and validated theoretically and experimentally a novel configuration of PTC, termed cold reservoir PTC, in which a reservoir together with an adjustable orifice is connected to the cold end of the pulse tube. The impedance from the additional orifice to the cold end helps to increase the mass flow in phase with the pressure wave at the cold end. Theoretical analyses with the linear model for the orifice and double-inlet PTCs indicate that the cooling performance can be improved by introducing the cold reservoir. The preliminary experiments with a home-made single-stage GM PTC further validated the results on the premise of minor opening of the cold-end orifice.

Study of heat transfer and pressure drop characteristics of air heat exchanger using PCM for free cooling applications

Directory of Open Access Journals (Sweden)

Kalaiselvam Sivakumar

2016-01-01

Full Text Available Free cooling is the process of storing the cool energy available in the night ambient air and using it during the day. The heat exchanger used in this work is a modular type which is similar to the shell and tube heat exchanger. The shell side is filled with Phase Change Materials (PCM and air flow is through the tubes in the module. The modules of the heat exchanger are arranged one over other with air spacers in between each module. The air space provided in between the module in-creases the retention time of the air for better heat transfer. Transient Computational Fluid Dynamics modeling is carried out for single air passage in a modular heat exchanger. It shows that the PCM phase transition time in the module in which different shape of fins is adopted. The module with rectangular fins has 17.2 % reduction in solidification compared with the plain module. Then steady state numerical analysis is accomplished to the whole module having the fin of high heat transfer, so that pressure drop, flow and thermal characteristics across the module and the air spacers are deter-mined for various air inlet velocities of 0.4 to 1.6 m/s. To validate the computational results, experiments are carried out and the agreement was found to be good.

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

International Nuclear Information System (INIS)

Gyongyosi, T.

2005-01-01

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

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

International Nuclear Information System (INIS)

Gyongyosi, T.

2005-01-01

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

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)

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

Displacer Diameter Effect in Displacer Pulse Tube Refrigerator

Science.gov (United States)

Zhu, Shaowei

2017-12-01

Gas driving displacer pulse tube refrigerators are one of the work recovery type of pulse tube refrigerators whose theoretical efficiency is the same as Stirling refrigerators'. Its cooling power is from the displacement of the displacer. Displace diameter, rod diameter and pressure drop of the regenerator influence the displacement, which are investigated by numerical simulation. It is shown that the displacement ratio of the displacer over the piston is almost not affected by the displacer diameter at the same rod diameter ratio, or displacer with different diameters almost has the same performance.

Fracture Toughness Round Robin Test International in pressure tube materials

International Nuclear Information System (INIS)

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

1993-01-01

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

Measurements of convective heat transfer to vertical upward flows of CO{sub 2} in circular tubes at near-critical and supercritical pressures

Energy Technology Data Exchange (ETDEWEB)

Zahlan, H., E-mail: hussamzahlan@gmail.com [Canadian Nuclear Laboratories, Chalk River, K0J 1J0 (Canada); Department of Mechanical Engineering, University of Ottawa, Ottawa, ON K1N 6N5 (Canada); Groeneveld, D. [Canadian Nuclear Laboratories, Chalk River, K0J 1J0 (Canada); Department of Mechanical Engineering, University of Ottawa, Ottawa, ON K1N 6N5 (Canada); Tavoularis, S. [Department of Mechanical Engineering, University of Ottawa, Ottawa, ON K1N 6N5 (Canada)

2015-08-15

Highlights: • We present and discuss results of thermal–hydraulic measurements in CO{sub 2} for the near critical and supercritical pressure region. • We report the full heat transfer and pressure drop database. - Abstract: An extensive experimental program of heat transfer measurements has been completed recently at the University of Ottawa's supercritical pressure test facility (SCUOL). Thermal–hydraulics tests were performed for vertical upflow of carbon dioxide in directly heated tubes with inner diameters of 8 and 22 mm, at high subcritical, near-critical and supercritical pressures. The test conditions, when converted to water-equivalent values, correspond to conditions of interest to current Super-Critical Water-Cooled Reactor designs, and include many measurements under conditions for which few data are available in the literature. These data significantly complement the existing experimental database and are being used for the derivation and validation of a new heat transfer prediction method in progress at the University of Ottawa. The same data are also suitable for the assessment of the accuracy of other heat transfer prediction methods and fluid-to-fluid scaling laws for near-critical and supercritical pressures. In addition, they permit further examination of previously suggested relationships describing the critical heat flux and post-dryout heat transfer coefficient at high subcritical pressures and the boundaries of the deteriorated/enhanced heat transfer regions for near-critical and supercritical pressures. The measurements reported in this paper cover several subcritical heat transfer modes, including single phase liquid heat transfer, nucleate boiling, critical heat flux, post-dryout heat transfer and superheated vapor heat transfer; they also cover several supercritical heat transfer modes, including heat transfer to liquid-like supercritical fluid and heat transfer to vapor-like supercritical fluid, which occurred in the

Low cryogen inventory, forced flow Ne cooling system with room temperature compression stage and heat recuperation

CERN Document Server

Shornikov, A; Wolf, A

2014-01-01

We present design and commissioning results of a forced flow cooling system utilizing neon at 30 K. The cryogen is pumped through the system by a room-temperature compression stage. To decouple the cold zone from the compression stage a recuperating counterflow tube-in-tube heat exchanger is used. Commissioning demonstrated successful condensation of neon and transfer of up to 30 W cooling power to the load at 30 K using only 30 g of the cryogen circulating in the system at pressures below 170 kPa.

Correlations of CO2 at supercritical pressures in a vertical circular tube

International Nuclear Information System (INIS)

Li Zhihui; Jiang Peixue

2010-01-01

The experiment results of convection heat transfer of CO 2 at supercritical pressures in a 2 mm diameter vertical circular tube for upward flow and downward flow were analyzed for pressures ranging from 78 to 95 bar, inlet temperatures from to 25 to 40 degree C, and inlet Re numbers from 3000 to 20000. The results were compared with some well known empirical correlations for the heat transfer without buoyancy effects and the heat transfer with strong buoyancy effects. It is found that there is a big deviation between the experiment results and empirical correlations. Based on the experiment data, correlations are developed for the local Nusselt correlations of CO 2 at supercritical pressures in vertical circular tubes.(authors)

Creep collapse of thick-walled heat transfer tube subjected to external pressure at high temperature

International Nuclear Information System (INIS)

Ioka, Ikuo; Kaji, Yoshiyuki; Terunuma, Isao; Nekoya, Shin-ichi; Miyamoto, Yoshiaki

1994-09-01

A series of creep collapse tests of thick-walled heat transfer tube were examined experimentally and analytically to confirm an analytical method for creep deformation behavior of a heat transfer tube of an intermediate heat exchanger (IHX) at a depressurization accident of secondary cooling system of HTTR (High Temperature Engineering Test Reactor). The tests were carried out using thick-walled heat transfer tubes made of Hastelloy XR at 950degC in helium gas environment. The predictions of creep collapse time obtained by a general purpose FEM-code ABAQUS were in good agreement with the experimental results. A lot of cracks were observed on the outer surface of the test tubes after the creep collapse. However, the cracks did not pass through the tube wall and, therefore, the leak tightness was maintained regardless of a collapse deformation for all tubes tested. (author)

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

Directory of Open Access Journals (Sweden)

A. M. Mohsen

2012-06-01

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

The Sensitivity Analysis of Axial Pressure Tube Creep Profile for Dryout Power in PHWR

Energy Technology Data Exchange (ETDEWEB)

Ryu, Euiseung; Kim, Youngae [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Stern Laboratory performed the CHF tests with only one axial pressure tube creep profile per 3.3%, 5.1% peak crept channel and made CHF correlation including creep factor from the CHF test results. Wolsong nuclear power plants also have utilized the same CHF correlation derived by CNL. Pressure tube diameter creep rate is function of fast neutron, coolant temperature, and coolant pressure in a channel. It means that various axial pressure tube creep profiles exist in PHWR due to the history of operating conditions. Usually, CHF correlation is used during ROP(Regional Overpower Protection) Trip Setpoint Analysis or Safety Analysis in PHWR. The sensitivity analysis for CHF effects using various creep profiles is needed. This paper summarizes the comparison results of dryout power between CHF test creep profile and estimated creep profiles of Wolsong units. The effect of axial pressure tube creep profile for dryout power in fuel channel is evaluated by using Stern Lab. CHF test creep profile and 380 channel creep profiles of Wolsong. The dryout powers at 3.3% and 5.1% test conditions are slightly smaller when using 380 Wolsong channels creep profiles. These also show that the simulated dryout powers maintain consistency regardless of flow conditions.

Burst pressure and leak rate from fretted SG tubes

International Nuclear Information System (INIS)

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

2005-01-01

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

Brittle-fracture potential of irradiated Zircaloy-2 pressure tubes

Science.gov (United States)

Huang, F. H.

1993-12-01

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

Plane strain analytical solutions for a functionally graded elastic-plastic pressurized tube

International Nuclear Information System (INIS)

Eraslan, Ahmet N.; Akis, Tolga

2006-01-01

Plane strain analytical solutions to functionally graded elastic and elastic-plastic pressurized tube problems are obtained in the framework of small deformation theory. The modulus of elasticity and the uniaxial yield limit of the tube material are assumed to vary radially according to two parametric parabolic forms. The analytical plastic model is based on Tresca's yield criterion, its associated flow rule and ideally plastic material behaviour. Elastic, partially plastic and fully plastic stress states are investigated. It is shown that the elastoplastic response of the functionally graded pressurized tube is affected significantly by the material nonhomogeneity. Different modes of plasticization may take place unlike the homogeneous case. It is also shown mathematically that the nonhomogeneous elastoplastic solution presented here reduces to that of a homogeneous one by appropriate choice of the material parameters

Heat transfer test in a tube using CO2 at supercritical pressures

International Nuclear Information System (INIS)

Kim, Hwan Yeol; Kim, Hyungrae; Song, Jin Ho; Cho, Bong Hyun; Bae, Yoon Yeong

2005-01-01

Heat transfer test facility, which is named as SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), has been constructed in KAERI for the study of heat transfer and pressure drop characteristics in a single tube, single rod and rod bundle at supercritical CO 2 conditions. The tests with supercritical water are difficult it terms of cost and effort, since the critical pressure and temperature of water are as high as 22.12 MPa and 374.14degC. As a substitute for water, CO 2 is selected for the test since the critical pressure and temperature of CO 2 are 7.38 MPa and 31.05degC that are much lower than those of water. This paper describes the design characteristics of the SPHINX and the experimental investigations on the heat transfer and pressure drop of a vertical single tube with an inside diameter of 4.4 mm with upward flow of supercritical CO 2 . The geometry of the single tube is the same as that of Kyushu University test performed with Freon (R22) for the direct comparison of a medium effect. The tests were performed with various heat and mass fluxes at a given pressure. The range of mass flux is 400∼1200 kg/m 2 s and the heat flux is chosen up to 150 kW/m 2 . The selected pressure are 7.75, 8.12, and 8.85 MPa. The test results are investigated and compared with the previous tests. (author)

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

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)

Critical heat flux data in a vertical tube at low and medium pressures

Energy Technology Data Exchange (ETDEWEB)

Teyssedou, A [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Olekhnowitch, A [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Tapucu, A [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Champagne, P [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Groeneveld, D [Chalk River Laboratories, AECL Research, Chalk River (Canada)

1994-09-01

AECL Research and Ecole Polytechnique have been cooperating on the validation of the critical heat flux (CHF) look-up table (D.C. Groeneveld et al., Heat Transfer Eng. 7(1-2) (1986) 46-62). For low and medium pressures the values in the table have been obtained by extrapolation and curve fitting; therefore, errors could be expected. To reduce these possible extrapolation errors, CHF experiments are being carried out in water cooled 8mm internal diameter (ID) tubes, at conditions where the data are scarce. This paper presents some of the experimental CHF data obtained for vertical up flow in an 8mm ID test section, for a wide range of exit qualities (5-70%) and the exit pressure ranging from 5 to 30bar. The experiments were carried out for heated lengths of 0.75, 1, 1.4 and 1.8m. In general, the collected data show parametric trends similar to those described in the open literature. However, it was observed that for low pressure conditions CHF depends on the heated length; this dependence begins to disappear for exit pressure of about 30bar. The CHF data have also been compared with predictions of well-known correlations (L. Biasi et al., Energia Nucl. 14(9) (1967) 530-536; R. Bowring, Br. Report AEEW-R789, Winfrith, UK, 1972; Y. Khatto and H. Ohno, Int. J. Heat Mass Transfer 27 (1984) 1641-1648) and those of the look-up table given by Groeneveld et al. For low pressures and low mass fluxes the look-up table seems to yield better predictions of the CHF than the correlations. However, for medium pressures and mass fluxes the correlations perform better than the look-up table; among those tested, Katto and Ohno's correlation gives the best results. ((orig.))

Fabrication of gas turbine water-cooled composite nozzle and bucket hardware employing plasma spray process

Science.gov (United States)

Schilke, Peter W.; Muth, Myron C.; Schilling, William F.; Rairden, III, John R.

1983-01-01

In the method for fabrication of water-cooled composite nozzle and bucket hardware for high temperature gas turbines, a high thermal conductivity copper alloy is applied, employing a high velocity/low pressure (HV/LP) plasma arc spraying process, to an assembly comprising a structural framework of copper alloy or a nickel-based super alloy, or combination of the two, and overlying cooling tubes. The copper alloy is plamsa sprayed to a coating thickness sufficient to completely cover the cooling tubes, and to allow for machining back of the copper alloy to create a smooth surface having a thickness of from 0.010 inch (0.254 mm) to 0.150 inch (3.18 mm) or more. The layer of copper applied by the plasma spraying has no continuous porosity, and advantageously may readily be employed to sustain a pressure differential during hot isostatic pressing (HIP) bonding of the overall structure to enhance bonding by solid state diffusion between the component parts of the structure.

Manufacturing of a HCLL cooling plate mock up

International Nuclear Information System (INIS)

Rigal, E.; Dinechin, G. de; Rampal, G.; Laffont, G.; Cachon, L.

2007-01-01

The European DEMO blankets and associated Test Blanket Modules (TBM) are made of a set of components cooled by flowing helium at 80bar pressure. Hot Isostatic Pressing (HIP) is one of the very few processes that allow manufacturing such components exhibiting complex cooling channels. In HIP technology, the parts used to manufacture components with embedded channels are usually machined plates, blocks and tubes. Achievable geometries are limited in shape because it is not always possible to figure the channels by bent tubes. This occurs for example when channels present sharp turns, when the cross section of the channels is rectangular or when the rib between channels is so small that very thin tubes would be required. In these cases, bending is unpractical. The breeder unit cooling plates of the Helium Cooled Lithium Lead (HCLL) blanket have eight 4 x 4.5 mm parallel channels that run following a double U scheme. Turns are sharp and the wall thickness is small (1mm), so the manufacturing process described above cannot be used. An alternative process has been developed which has many advantages. It consists in machining grooves in a base plate, then closing the top of the grooves using thin welded strips, and finally adding a plate by HIP. There is then no need for the use of tubes with associated bending and deformation issues. The final component contains welds, but it must be stressed out that these potentially brittle zones do not connect the channels to the external surface because they are covered by the HIPed plate. Furthermore, the welds are homogenised during the HIP operation and further heat treatments. This paper describes the design of a simplified cooling plate mock up and its fabrication using this so-called weld+HIP process. The thermal fatigue testing of this mock up is presented somewhere else in this conference. (orig.)

Condensation of pure and near-azeotropic refrigerants in microfin tubes: A new computational procedure

Energy Technology Data Exchange (ETDEWEB)

Cavallini, A; Del Col, D; Mancin, S; Rossetto, L [Dipartimento di Fisica Tecnica, University of Padova, Via Venezia 1, Padova 35131 (Italy)

2009-01-15

Microfin tubes are widely used in air cooled and water cooled heat exchangers for heat pump and refrigeration applications during condensation or evaporation of refrigerants. In order to design heat exchangers and to optimize heat transfer surfaces, accurate procedures for computing pressure drops and heat transfer coefficients are necessary. This paper presents a new simple model for the prediction of the heat transfer coefficient to be applied to condensation in horizontal microfin tubes of halogenated and natural refrigerants, pure fluids or nearly azeotropic mixtures. The updated model accounts for refrigerant physical properties, two-phase flow patterns in microfin tubes and geometrical characteristics of the tubes. It is validated against a data bank of 3115 experimental heat transfer coefficients measured in different independent laboratories all over the world including diverse inside tube geometries and different condensing refrigerants among which R22, R134a, R123, R410A and CO{sub 2}. (author)

Benchmark of WIMS-IST against MCNP for CANDU pressure tube fast fluxes

International Nuclear Information System (INIS)

Donders, R.E.; Douglas, S.R.

2002-01-01

Pressure tube fast-flux data in CANDU are currently calculated using the multi-group neutron transport code WIMS-IST. In this study, the WIMS-IST fast flux calculations are benchmarked against MCNP calculations (a Monte Carlo particle transport code), over the range of fuel burnup and coolant density in CANDU. The comparison shows good agreement between WIMS and MCNP, with WIMS fast fluxes being 1.5% to 4% lower than the MCNP values. The difference is smallest for fresh fuel, and increases with burnup. The fast flux gradient across the pressure tube (factor of 1.23 from inner edge to outer edge) is accurately calculated by WIMS. When reporting fast fluxes in pressure tubes, these are generally given as >1.000 MeV fluxes. For WIMS, this requires an extra conversion step, since the WIMS ENDF/B libraries do not have a group boundary at 1 MeV. The conversion step is based on a fictitious isotope ONEMEV in the WIMS nuclear data library. The conversion factor in WIMS was found to be about one percent too high. When providing >1 MeV fluxes from WIMS, this partially compensates for the slight under prediction of the fast flux. Pressure tube >1 MeV fluxes from WIMS are therefore 0.5% to 3% lower than MCNP values. To obtain accurate fast flux data, neutron transport calculations must be performed on a critical cell. For this study, all calculations were performed with radial albedo boundary conditions giving a critical cell. This required the use of an albedo version of MCNP, developed at AECL. (author)

Rotating anode X-ray tubes

International Nuclear Information System (INIS)

Webley, R.S.

1981-01-01

In a rotating anode x-ray tube it is proposed to mount the rotating anode, or means such as a shaft affixed to it, to rotate on bearings in a race the seating for which is cooled by a suitable coolant flow. A suitable bellows arrangement allows the coolant pressure to determine the contact pressure of the seating on the bearings. This allows the thermal impedance to be varied and the bearing wear to be optimised therewith as well as allowing adjustment for wear. The use of two bellows allows the seating section therebetween to move towards the other section as the rollers wear. (author)

The development of an auto-sealing system using an electrically shrinkable tube under a low-pressure condition

Energy Technology Data Exchange (ETDEWEB)

Okano, Yoshihiro; Kitagawa, Takao [NKK Corp, Tsu, Mie (Japan); Shoji, Norio [NKK Corp., Yokohama (Japan); Namioka, Toshiyuki [Nippon Kokan Koji Corp., Yokohama (Japan). Research and Development Dept.; Komura, Minoru [Nitto Denko Corp., Fukaya, Saitama (Japan)

1997-04-01

This article describes the development of a system to create high quality, automatic sealing of field joints of polyethylene coated pipelines. The system uses a combination of an electrically heated shrinkable tube and a low-pressure chamber. The self-heating shrinkable tube includes electric heater wires that heat when connected to electricity. A method was developed to eliminate air trapped between the tube and the steel pipe by shrinking the tube under a low-pressure condition. The low-pressure condition was automatic and easily attained by using a vacuum chamber. It was verified that the system produced high quality sealing of the field joints.

An analysis of system pressure and temperature distribution in self-pressurizer of SMART considering thermal stratification at intermediate cavity

Energy Technology Data Exchange (ETDEWEB)

Kang, Yeon Moon; Lee, Doo Jeong; Yoon, Ju Hyun; Kim, Hwan Yeol [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-03-01

Because the pressurizer is in reactor vessel, the heat transfer from primary water would increase the temperatures of fluids in pressurizer to same temperature of hotleg, if no cooling equipment were supplied. Thus, heat exchanger and thermal insulator are needed to minimize heat transferred from primary water and to remove heat in pressurizer. The temperatures in cavities of pressurizer for normal operation are 70 deg C and 74 deg C for intermediate and end cavity, respectively, which considers the solubility of nitrogen gas in water. Natural convection is the mechanism of heat balance in pressurizer of SMART. In SMART, the heat exchanger in pressurizer is placed in lower part of intermediate cavity, so the heat in upper part of intermediate cavity can't be removed adequately and it can cause thermal stratification. If thermal stratification occurred, it increases heat transfers to nitrogen gas and system pressure increases as the result. Thus, proper evaluation of those effects on system pressure and ways to mitigate thermal stratification should be established. This report estimates the system pressure and temperatures in cavities of pressurizer with considering thermal stratification in intermediate cavity. The system pressure and temperatures for each cavities considered size of wet thermal insulator, temperature of upper plate of reactor vessel, parameters of heat exchanger in intermediate cavity such as flow rate and temperature of cooling water, heat transfer area, effective tube height, and location of cooling tube. In addition to the consideration of thermal stratification thermal mixing of all water in intermediate cavity also considered and compared in this report. (author). 6 refs., 60 figs., 2 tabs.

The thermal interaction of a buoyant plume from a calandria tube with an oblique jet

Energy Technology Data Exchange (ETDEWEB)

Rossouw, D.J.; Atkins, M.D.; Beharie, K. [Nuclear Science Division, School of Mechanical & Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Kim, T., E-mail: tong.kim@wits.ac.za [Nuclear Science Division, School of Mechanical & Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Rhee, B.W.; Kim, H.T. [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejun (Korea, Republic of)

2016-12-15

Highlights: • A crucial role of relative orientation between mixed convection modes is observed. • The extent of thermal interaction strongly depends on the relative orientation. • Coolant flow is substantially diffused by a buoyant plume if counter-acting. • Slightly oblique coolant flow to the gravitational axis provides the best cooling. - Abstract: Severe reactor core damage may occur from fuel channel failure as a consequence of excessive heat emitted from calandria tubes (CTs) in a pressurised heavy water (D{sub 2}O) reactor (CANDU). The heating of the CTs is caused by creep deformation of the pressure tubes (PTs), which may be ballooning or sagging depending on the internal pressure of the PTs. The deformation of the pressure tube is due to overheating as a result of a loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) failure. To prevent the exacerbation of the LOCA, circulating D{sub 2}O in the moderator tank may be utilized by forming a secondary jet that externally cools the individual CTs. The buoyant plume develops around the CTs and interacts with the secondary jet at a certain oblique angle with respect to the gravitational axis, depending on the spatial location of the hot calandria tubes (or the hot reactor core region). This study reports on how the local and overall heat transfer characteristics on a calandria tube where the buoyant plume develops, are altered by the obliqueness of the external secondary jet (from a co-current jet to a counter-current jet) in a simplified configuration at the jet Reynolds number of Re{sub j} = 1500 for the Archimedes number of Ar{sub D} = 0.11 and Rayleigh number of Ra{sub D} = 1.6 × 10{sup 6} (modified Rayleigh number of 3.0 × 10{sup 7}).

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

International Nuclear Information System (INIS)

Marinescu, R.; Mincu, M.

2015-01-01

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

Device for automatically operating cooling mode of water in a pressure suppression chamber

International Nuclear Information System (INIS)

Sato, Hideyuki.

1975-01-01

Object: To provide a system for removing residual heat in a reactor safety system, which can automatically cool water in a pressure suppression chamber when a load on a generator is cut off, so as not to scram the reactor. Structure: When a load cut-off signal is generated by means of rapid closure of a turbine regulating valve or due to the load unbalance relay of generator output, or the like, a sea water pump is started to fully open an outlet valve for the sea water pump, a heat exchanging inlet valve and a minimum crow valve and to fully close a heat exchanging bypass valve. In this manner, cooling water for the heat exchanger is secured to start the pump in the system for removing residual heat, and when the pump discharge pressure is in normal condition, the inlet valve in pressure suppression chamber and the spray valve in the pressure suppression chamber are fully opened to automatically cool water in the pressure suppression chamber. (Hanada, M.)

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

International Nuclear Information System (INIS)

Saibaba, N.

2010-01-01

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

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

International Nuclear Information System (INIS)

Aly, Wael I.A.

2014-01-01

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

Study on transport safety of fresh MOX fuel. Performance of the cladding tube of fresh MOX fuel against external water pressure

International Nuclear Information System (INIS)

Ito, Chihiro

1999-01-01

It is important to know the ability of the cladding tube for fresh MOX fuel against external water pressure when they were hypothetically sunk into the sea for unknown reasons. In order to evaluate the ability of cladding tubes for MOX fresh fuel against external water pressure, external water pressure tests were carried out. Resistible limit of cladding tubes against external water pressure is defined when cladding tubes are deformed largely due to buckling etc. The test results show cladding tube of BWR type can resist an external water pressure of 69 MPa (a depth of water of 7,000 m) and that of PWR type fuel can resist an external water pressure of 54 MPa (a depth of water of 5,500 m). Moreover, leak tightness is maintained at an external water pressure of 73 MPa (a depth of water of 7,400 m) for BWR type cladding tubes and at an external water pressure of 98 MPa (a depth of water of 10,000 m) for PWR type cladding tubes. (author)

Endotracheal tube cuff pressures during general anaesthesia while using air versus a 50% mixture of nitrous oxide and oxygen as inflating agents

Directory of Open Access Journals (Sweden)

Jesni Joseph Manissery

2007-01-01

Full Text Available The present study was aimed at assessing the efficacy of filling a 50% mixture of nitrous oxide : oxygen (50%N 2 O:O 2 in the endotracheal tube cuff to provide stable cuff pressures during general anaesthesia with 67%N 2 O. The endotracheal tube cuff pressures with air (control as the inflating agent in the tubes were found to have a total mean pressure of 62.60±12.33 at the end of one hour of general anaesthesia. When comparing the endotracheal tube cuff pressures in the Mallinckrodt tubes with that of the Portex tubes, with air as the inflating agent, the Portex tubes showed a significantly lower cuff pressures at the end of one hour. The endotracheal tube cuff pressures with 50%N 2 O:O 2 as the inflating agent showed a total mean pressure of 27.63 ± 3.221 at the end of one hour of general anaesthesia. This indicates that inflation of the cuff of the endotracheal tubes with a 50%N 2 O:O 2 rather than air maintains a stable intra cuff pressure. Therefore, the method of using a 50%N 2 O:O 2 for filling endotracheal tube cuff can be adopted for endotracheal tubes with high-volume, low-pressure cuffs to prevent both excessive cuff pressure and disruption of cuff seal, during general anaesthesia lasting up to one hour.

Supercritical-pressure light water cooled reactors

CERN Document Server

Oka, Yoshiaki

2014-01-01

This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

Superconducting cable cooling system by helium gas at two pressures

International Nuclear Information System (INIS)

Dean, J.W.

1977-01-01

Thermally contacting, oppositely streaming, cryogenic fluid streams in the same enclosure in a closed cycle changes the fluid from a cool high pressure helium gas to a cooler reduced pressure helium gas in an expander so as to be at different temperature ranges and pressures respectively in go and return legs that are in thermal contact with each other and in thermal contact with a longitudinally extending superconducting transmission line enclosed in the same cable enclosure that insulates the line from the ambient at a temperature T 1 . By first circulating the fluid from a refrigerator at one end of the line as a cool gas at a temperature range T 2 to T 3 in the go leg, then circulating the gas through an expander at the other end of the line where the gas becomes a cooler gas at a reduced pressure and at a reduced temperature T 4 and finally by circulating the cooler gas back again to the refrigerator in a return leg at a temperature range T 4 to T 5 , while in thermal contact with the gas in the go leg, and in the same enclosure therewith for compression into a higher pressure gas at T 2 in a closed cycle, where T 2 greater than T 3 and T 5 greater than T 4 , the fluid leaves the enclosure in the go leg as a gas at its coldest point in the go leg, and the temperature distribution is such that the line temperature decreases along its length from the refrigerator due to the cooling from the gas in the return leg

Emergency core cooling systems

International Nuclear Information System (INIS)

Kubokoya, Takashi; Okataku, Yasukuni.

1984-01-01

Purpose: To maintain the fuel soundness upon loss of primary coolant accidents in a pressure tube type nuclear reactor by injecting cooling heavy water at an early stage, to suppress the temperature of fuel cans at a lower level. Constitution: When a thermometer detects the temperature rise and a pressure gauge detects that the pressure for the primary coolants is reduced slightly from that in the normal operation upon loss of coolant accidents in the vicinity of the primary coolant circuit, heavy water is caused to flow in the heavy water feed pipeway by a controller. This enables to inject the heavy water into the reactor core in a short time upon loss of the primary coolant accidents to suppress the temperature rise in the fuel can thereby maintain the fuel soundness. (Moriyama, K.)

Design of distributed JT (Joule-Thomson) effect heat exchanger for superfluid 2 K cooling device

Science.gov (United States)

Jeong, S.; Park, C.; Kim, K.

2018-03-01

Superfluid at 2 K or below is readily obtained from liquid helium at 4.2 K by reducing its vapour pressure. For better cooling performance, however, the cold energy of vaporized helium at 2 K chamber can be effectively utilized in a recuperator which is specially designed in this paper for accomplishing so-called the distributed Joule-Thomson (JT) expansion effect. This paper describes the design methodology of distributed JT effect heat exchanger for 2 K JT cooling device. The newly developed heat exchanger allows continuous significant pressure drop at high-pressure part of the recuperative heat exchanger by using a capillary tube. Being different from conventional recuperative heat exchangers, the efficient JT effect HX must consider the pressure drop effect as well as the heat transfer characteristic. The heat exchanger for the distributed JT effect actively utilizes continuous pressure loss at the hot stream of the heat exchanger by using an OD of 0.64 mm and an ID of 0.4 mm capillary tube. The analysis is performed by dividing the heat exchanger into the multiple sub-units of the heat exchange part and JT valve. For more accurate estimation of the pressure drop of spirally wound capillary tube, preliminary experiments are carried out to investigate the friction factor at high Reynolds number. By using the developed pressure drop correlation and the heat transfer correlation, the specification of the heat exchanger with distributed JT effect for 2 K JT refrigerator is determined.

SCW Pressure-Channel Nuclear Reactor Some Design Features

Science.gov (United States)

Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

Experimental facility design for a gap heat transfer in a double wall tube

International Nuclear Information System (INIS)

Nam, Ho Yun; Hong, Jong Gan; Kim, Jong Man; Kim, Jong Bum; Jeong, Ji Young

2012-01-01

A reliable steam generator design is one of the most critical issues in developing a sodium cooled fast reactor (SFR), and various efforts to avoid potential sodium water reaction (SWR) have been made. For this reason, SFR steam generators have been developed to improve its reliability using a double wall tube (DWT), which has two barriers between the sodium and water. Most steam generators for SFRs are the shell and tube type. Steam at high pressure and low temperature flows inside the inner tubes, which are heated by the shell side sodium at low pressure and high temperature. Since the inner and outer tubes of conventional DWTs are made of identical materials, the degree of thermal expansion is somewhat different between the two concentric tubes owing to their temperature difference. Therefore, a greater temperature difference results in less contact pressures between the inner and outer tubes. This feature results in a deterioration of the heat transfer capability of DWTs. Current developments are focused on an improvement of heat transfer capability by investigating the gap conductance between the two concentric tubes. To improve the heat transfer capability of DWTs, it is preferable to use different tube materials (Fig. 1). It is recommended to choose the inner tube material whose thermal expansion coefficient is greater than that of the outer tube by 10 to 15%

Gas pressure in bubble attached to tube circular outlet

OpenAIRE

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

2016-01-01

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

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

Heat Transfer Characteristics for an Upward Flowing Supercritical Pressure CO{sub 2} in a Vertical Circular Tube

Energy Technology Data Exchange (ETDEWEB)

Kang, Deog Ji

2008-02-15

The SCWR(Super Critical Water-cooled Reactor) is one of the feasible options for the 4th generation nuclear power plant, which is being pursued by an international collaborative organization, the Gen IV International Forum(GIF). The major advantages of the SCWR include a high thermal efficiency and a maximum use of the existing technologies. In the SCWR, the coolant(water) of a supercritical pressure passes the pseudo-critical temperature as it flows upward through the sub-channels of the fuel assemblies. At certain conditions a heat transfer deterioration occurs near the pseudo-critical temperature and it may cause an excessive rise of the fuel surface temperature. Therefore, an accurate estimation of the heat transfer coefficient is necessary for the thermal-hydraulic design of the reactor core. A test facility, SPHINX(Supercritical Pressure Heat Transfer Investigation for the Next Generation), dedicated to produce heat transfer data and study flow characteristics, uses supercritical pressure CO{sub 2} as a surrogate medium to take advantage of the relatively low critical temperature and pressure: and similar physical properties with water. The produced data includes the temperature of the heating surface and the heat transfer coefficient at varying mass fluxes, heat fluxes, and operating pressures. The test section is a circular tube of ID 6.32 mm: it is almost the same as the hydraulic diameter of the sub-channel in the conceptional design presented by KAERI. The test range of the mass flux is 285 to 1200 kg/m{sup 2}s and the maximum heat flux is 170 kW/m{sup 2}. The tests were mainly performed for an inlet pressure of 8.12 MPa which is 1.1 times of critical pressure. With the test results of the wall temperature and the heat transfer coefficient, effects of mass flux, heat flux, inlet pressure, and the tube diameter on the heat transfer were studied. And the test results were compared with the existing correlations of the Nusselt number. In addition, New

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Science.gov (United States)

2013-10-25

... comments were received. A companion guide, DG-1277, ``Initial Test Program of Emergency Core Cooling... NUCLEAR REGULATORY COMMISSION [NRC-2011-0129] Preoperational Testing of Emergency Core Cooling... (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors...

Pressure drop and friction factor correlations of supercritical flow

International Nuclear Information System (INIS)

Fang Xiande; Xu Yu; Su Xianghui; Shi Rongrong

2012-01-01

Highlights: ► Survey and evaluation of friction factor models for supercritical flow. ► Survey of experimental study of supercritical flow. ► New correlation of friction factor for supercritical flow. - Abstract: The determination of the in-tube friction pressure drop under supercritical conditions is important to the design, analysis and simulation of transcritical cycles of air conditioning and heat pump systems, nuclear reactor cooling systems and some other systems. A number of correlations for supercritical friction factors have been proposed. Their accuracy and applicability should be examined. This paper provides a comprehensive survey of experimental investigations into the pressure drop of supercritical flow in the past decade and a comparative study of supercritical friction factor correlations. Our analysis shows that none of the existing correlations is completely satisfactory, that there are contradictions between the existing experimental results and thus more elaborate experiments are needed, and that the tube roughness should be considered. A new friction factor correlation for supercritical tube flow is proposed based on 390 experimental data from the available literature, including 263 data of supercritical R410A cooling, 45 data of supercritical R404A cooling, 64 data of supercritical carbon dioxide (CO 2 ) cooling and 18 data of supercritical R22 heating. Compared with the best existing model, the new correlation increases the accuracy by more than 10%.

Low-cost evacuated-tube solar collector appendices. Final report

Energy Technology Data Exchange (ETDEWEB)

Beecher, D.T.

1980-05-31

A low cost solar heat energy collector module and array has been designed using the evacuated tube, selective absorber, air cooled concept. Glass tubing as used in fluorescent lamps with automatic sealing methods is a key feature of the evacuated tube design. A molded fiber glass concentrating reflector panel and sheet metal header assembly are proposed. Major design problems involved included the cost of materials and labor, thermal expansion and distortion problems, high stagnation and operating temperatures, isolation, thermal efficiency, sealing, joining, air pressure drop, and weight of the preassembled module. A cost of less than $5 per active square foot of collecting surface has been estimated for materials and labor of the module and its mounting frame.

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

International Nuclear Information System (INIS)

Griffiths, M.; Winegar, J.E.

1994-12-01

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

The application of ductile-fracture analysis to predictions of pressure-tube failure

International Nuclear Information System (INIS)

Simpson, L.A.

1981-08-01

Progress during the past six years towards establishing a method for predicting critical crack length in a reactor pressure tube, based on data from tests on small fracture-mechanics specimens, is reviewed. The disadvantages of relying on data from burst tests alone are described along with the benefits of a small-specimen method. It is clear from the work reviewed that only an approach that can account for the ability of the presssure tube material to increase its crack-growth resistance during stable crack extension is suitable for the prediction of critical crack length. A method that utilizes crack-growth resistance curves based on crack-opening displacement, or the J integral, is described, along with a large body of experimental data. It is concluded that the resistance curve approach provides a viable method for the analysis of fracture in pressure tubes that can greatly improve our understanding of the material's behaviour

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-15

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