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Sample records for candu pressure tubes

  1. Performance of pressure tubes in CANDU reactors

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    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

  2. In-reactor performance of pressure tubes in CANDU reactors

    Science.gov (United States)

    Rodgers, D. K.; Coleman, C. E.; Griffiths, M.; Bickel, G. A.; Theaker, J. R.; Muir, I.; Bahurmuz, A. A.; Lawrence, S. St.; Resta Levi, M.

    2008-12-01

    The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS, and fracture toughness), and delayed-hydride-cracking properties (crack growth rate Vc, and threshold stress intensity factor, KIH) change with irradiation; the former reach a limiting value at a fluence of Pressure tubes exhibit elongation and diametral expansion. The deformation behaviour is a function of operating conditions and material properties that vary from tube-to-tube and as a function of axial location. Semi-empirical predictive models have been developed to describe the deformation response of average tubes as a function of operating conditions. For corrosion and, more importantly deuterium pickup, semi-empirical predictive models have also been developed to represent the behaviour of an average tube. The effect of material variability on corrosion behaviour is less well defined compared with other properties. Improvements in manufacturing have increased fracture resistance by minimising trace elements, especially H and Cl, and reduced variability by tightening controls on forming parameters, especially hot-working temperatures.

  3. Development of an Integrity Assessment Procedure for CANDU Pressure Tubes

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    Chung, Han Sub [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The pressure tubes used in a CANDU reactor are made from Zr-2.5Nb. During service the pressure tubes operate at temperatures between about 150 and 310 .deg. C, and with variable coolant pressures up to 11MPa corresponding to hoop stress of up to 130MPa. The maximum flux of fast neutrons (E>1MeV) from the fuel is about 4X10{sup 17}nm{sup -2}{sub s}{sup -1}. The pressure tubes are exposed to very severe degradation environment. The aging degradation of the pressure tubes are summarized as below. - Geometric deformation; axial elongation, diametric creep, and wall thinning. - Deuterium uptake; some fraction of the deuterium generated by the corrosion of pressure tubes is absorbed into the pressure tubes. Total equivalent hydrogen content in the pressure tube is the sum of the initial hydrogen content before operation and the deuterium uptake during operation. High concentration of hydrogen inside the pressure tubes makes the metal susceptible to Delayed Hydride Cracking. The DHC is a degradation mechanism of prime importance for CANDU pressure tubes. Mechanical properties, in particular fracture toughness, are deteriorated by high concentration of dissolved hydrogen. - Flaws; volumetric flaws are generated during operation. Wear scars by debris fretting, and bearing pad fretting are common. These volumetric flaws can be a site of crack initiation by fatigue or DHC. Cracks can propagate by DHC or fatigue crack propagation if conditions are met. - Material properties degradation; mechanical properties are affected by neutron irradiation. Yield strength and tensile strength are increased, and fracture toughness is deteriorated. The susceptibility to DHC is also affected. The integrity assessment of the pressure tube is a procedure to determine if the risk of pressure tube failure is controlled to maintain acceptably low. CSA N285.4 and 285.8 are two important guidelines regarding the integrity of pressure tubes. N285.4 is to guide in-service inspection, and N285

  4. Development of CANDU pressure tube integrity evaluation system

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    Kwac, S. L.; Kim, Y. J. [Sungkyunkwan Univ., Seoul (Korea, Republic of); Lee, J. S. [Kyonggi Univ., Suwon (Korea, Republic of); Park, Y. W. [KINS, Taejon (Korea, Republic of)

    1999-05-01

    The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw or contact with their calandria tubes is found during the periodic inspection, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to perform the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the integrity evaluation process. For this reason, an integrity evaluation system was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL. The evaluation procedure includes the crack growth calculation both by DHC and by fatigue. It also provides the prediction of fracture initiation, plastic collapse and leak-before-break(LBB), blister formation and blister growth. This system provides various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

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

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    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.

  6. Failure probability estimation of flaw in CANDU pressure tube considering the dimensional change

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    Kwak, Sang Log; Kim, Young Jin [Sungkyunkwan Univ., Suwon (Korea, Republic of); Lee, Joon Seong [Kyonggi Univ., Suwon (Korea, Republic of); Park, Youn Won [KINS, Taejon (Korea, Republic of)

    2002-11-01

    The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and heavy water coolant. Pressure tubes are installed horizontally inside the reactor and only selected samples are periodically examined during in-service inspection. In this respect, a probabilistic safety assessment method is more appropriate for the assessment of overall pressure tube safety. The failure behavior of CANDU pressure tubes, however, is governed by delayed hydride cracking which is the major difference from pipings and reactor pressure vessels. Since the delayed hydride cracking has more widely distributed governing parameters, it is impossible to apply a general PFM methodology directly. In this paper, a PFM methodology for the safety assessment of CANDU pressure tubes is introduced by applying Monte Carlo simulation in determining failure probability. Initial hydrogen concentration, flaw shape and depth, axial and radial crack growth rate and fracture toughness were considered as probabilistic variables. Parametric study has been done under the base of pressure tube dimension and hydride precipitation temperature in calculating failure probability. Unstable fracture and plastic collapse are used for the failure assessment. The estimated failure probability showed about three-order difference with changing dimensions of pressure tube.

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

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

    1997-04-01

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

  8. Pressure tube creep impact on the physics parameters for CANDU-6 reactors

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    Kim, W. Y.; Min, B. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kam, S. C.; Kim, M. E. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The lattice cell calculations are performed to assess the sensitivity of the reactor physics parameters to pressure tube creep resulting from radiation aging. The physics parameters of the lattice cell are calculated by using WIMSD-5B code, WIMS- AECL code, and MCNP code. The reference model(normal state) and two perturbed models accounting for the pressure tube creep are developed on the basis of CANDU-6 lattice cell. The 2.5% and 5% values of pressure tube diameter creep are considered. Also, The effects of the analyzed lattice parameters which are the coolant void reactivity, the fuel fission density and the atom density of Pu isotopes on the lattice.

  9. Probabilistic integrity assessment of CANDU pressure tube for the consideration of flaw generation time

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    Kwak, Sang Log; Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Lee, Joon Seong [Kyonggi Univ., Seoul (Korea, Republic of); Park, Youn Won [KINS, Taejon (Korea, Republic of)

    2001-07-01

    This paper describes a Probabilistic Fracture Mechanics (PFM) analysis based on Monte Carlo (MC) simulation. In the analysis of CANDU pressure tube, it is necessary to perform the PFM analyses based on statistical consideration of flaw generation time. A depth and an aspect ratio of initial semi-elliptical surface crack, a fracture toughness value, Delayed Hydride Cracking (DHC) velocity, and flaw generation time are assumed to be probabilistic variables. In all the analyses, degradation of fracture toughness due to neutron irradiation is considered. Also, the failure criteria considered are plastic collapse, unstable fracture and crack penetration. For the crack growth by DHC, the failure probability was evaluated in due consideration of flaw generation time.

  10. Diagnostic technology for degradation of feeder pipes and fuel channels in CANDU reactor; development of aging assessment technology for CANDU pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Kim, Yun Jae; Huh, Nam Su; Kwak, Sang Log; Lee, Kyu Ho [Sungkyunkwan University, Seoul (Korea)

    2002-04-01

    This research project attempts to resolve two issues related to integrity assessment of CANDU pressure tubes; (1) FE analysis of blister formation and growth, and (2) engineering estimation scheme to predict creep deflection of pressure tubes. Results for blister formation and growth can be summarised as follows. Comparing the results from the FE analysis, developed within this project, with experimental data shows some differences ranging from 10-57%. Such difference results from two possible sources. One source is neglecting two phase diffusion. The present FE analysis considers only single phase diffusion, and thus blister growth can not be accurately modeled. The other source would be inherent errors associated with experimental measurement. Thus it has been concluded that further efforts should be made on two phase diffusion modeling. For developing mechanistic model of creep deflection, the proposed reference stress based model is simple to use. Extensive validation against creep FE results shows that the proposed model is also quite accurate. More important aspect of the proposed method is that it can be easily generalized to more complex problems. Thus it is believed that the present results provide a sound basis for sagging assessment of CANDU pressure tubes. 16 refs., 12 figs., 6 tabs. (Author)

  11. Some applications related to the structural integrity analyses of CANDU 6 pressure tubes

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    Radu, V.S. [Inst. for Nuclear Research Pitesti, Pitesti (Romania)]. E-mail: vradu@nuclear.ro

    2006-07-01

    The flaws found during in-service inspection of Zr-2.5%Nb CANDU pressure tubes include fuel bundle bearing pad fretting flaws and debrise fretting flaws. In-service flaws are evaluated using fitness-for-service procedures to justify continued operation of pressure tube containing the flaw. The flaw evaluation procedures address crack initiation due to Delayed Hydride Cracking (DHC) under constant loading and also address fracture initiation and plastic collapse. The paper presents some applications related to the influence of the residual hoop stresses at roll-expanded joint region into stainless steel fittings at both ends on the structural integrity evaluation in the presence of blunt flaws. Two cases of blunt flaws were considered for evaluation: fuel bundle bearing pad fretting flaws and debrise fretting flaws. The blunt flaw geometry modeling and stress-strain analyses were performed by means finite element method (FEM) with FEA-Flaw computer code. The allowable peak flaw-tip stress and the Failure Assessment Diagram (FAD) for DHC initiation criterion were used for integrity assessment for the mentioned blunt flaws. Applications are performed as part of the research program address to evaluation of the in-service inspection results of the fuel channels from Cernavoda NPP. (author)

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

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

    2012-12-15

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

  13. PARTICULARITIES REGARDING THE OPERATING PROCESS OF THE CUTTING AND EXTRACTION DEVICE IN THE CANDU HORIZONTAL FUEL CHANNELS PRESSURE TUBE DECOMMISSIONING PART II: CUTTING AND EXTRACTING PRESSURE TUBE PROCESS

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2016-05-01

    Full Text Available This paper presents some details of operation process for a Cutting and Extraction Device (CED in order to achieve the decommissioning of the horizontal fuel channels pressure tube in the CANDU 6 nuclear reactor. The most important characteristic of the Cutting and Extraction Device (CED is his capability of totally operator’s protection against the nuclear radiation during pressure tube decommissioning. The cutting and extracting pressure tube processes present few particularities due to special adopted technical solutions: a special module with three cutting rollers (system driven by an actuator, a guiding-extracting and connecting module (three fixing claws which are piloted by an actuator and block the device in the connecting position with extracting plugs. The Cutting and Extraction Device (CED is a train of modules equipped with special systems to be fully automated, connected with a Programmable Logic Controller (PLC and controlled by an operator panel type Human Machine Interface (HMI. All processes are monitored by video cameras. In case of error, the process is automatically stopped, the operator receiving an error message and the last sequence could be reinitialized or aborted due to safety reasons.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Yong [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of); Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2011-08-15

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

  16. Characteristics of U-tube assembly design for CANDU 6 type steam generators

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    Park, Jun Su; Jeong, Seung Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    Since the first operation of nuclear steam generator early 1960s, its performance requirements have been met but the steam generator problems have been met but the steam generator problems have been major cause of reducing the operational reliability, plant safety and availability. U-tube assembly of steam generator forms the primary system pressure boundary of the plant and have experienced several types of tube degradation problems. Tube failure and leakage resulting from the degradation will cause radioactive contamination of secondary system by the primary coolant, and this may lead to unplanned plant outages and costly repair operations such as tube plugging or steam generator replacement. For the case of steam generators for heavy water reactors, e.g. Wolsong 2, 3, and 4 NPP, a high cost of heavy water will be imposed additionally. During the plant operation, steam generator tubes can potentially be subject to adverse environmental conditions which will cause damages to U-tube assembly. Types of the damage depend upon the combined effects of design factors, materials and chemical environment of steam generator, and they are the pure water stress corrosion cracking, intergranular attack, pitting, wastage, denting, fretting and fatigue, etc. In this report, a comprehensive review of major design factors of recirculating steam generators has been performed against the potential tube damages. Then the design characteristics of CANDU-type Wolsong steam generator were investigated in detail, including tube material, thermalhydraulic aspects, tube-to-tubesheet joint, tube supports, water chemistry and sludge management. 9 tabs., 18 figs., 38 refs. (Author) .new.

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

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    Choi, Sung Nam; Kim, Hyung Nam; Yoo, Hyun Joo [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Hwang, Won Gul [Chonnam National University, Gwangju (Korea, Republic of)

    2009-07-01

    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.

  18. Diametral creep prediction of pressure tube using statistical regression methods

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    Kim, D. [Korea Advanced Inst. of Science and Technology, Daejeon (Korea, Republic of); Lee, J.Y. [Korea Electric Power Research Inst., Daejeon (Korea, Republic of); Na, M.G. [Chosun Univ., Gwangju (Korea, Republic of); Jang, C. [Korea Advanced Inst. of Science and Technology, Daejeon (Korea, Republic of)

    2010-07-01

    Diametral creep prediction of pressure tube in CANDU reactor is an important factor for ROPT calculation. In this study, pressure tube diametral creep prediction models were developed using statistical regression method such as linear mixed model for longitudinal data analysis. Inspection and operating condition data of Wolsong unit 1 and 2 reactors were used. Serial correlation model and random coefficient model were developed for pressure tube diameter prediction. Random coefficient model provided more accurate results than serial correlation model. (author)

  19. Impact of aging and material structure on CANDU plant performance

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    Nadeau, E.; Ballyk, J.; Ghalavand, N. [Candu Energy Inc., Mississauga, Ontario (Canada)

    2011-07-01

    In-service behaviour of pressure tubes is a key factor in the assessment of safety margins during plant operation. Pressure tube deformation (diametral expansion) affects fuel bundle dry out characteristics resulting in reduced margin to trip for some events. Pressure tube aging mechanisms also erode design margins on fuel channels or interfacing reactor components. The degradation mechanisms of interest are primarily deformation, loss of fracture resistance and hydrogen ingress. CANDU (CANada Deuterium Uranium, a registered trademark of the Atomic Energy of Canada Limited used under exclusive licence by Candu Energy Inc.) owners and operators need to maximize plant capacity factor and meet or exceed the reactor design life targets while maintaining safety margins. The degradation of pressure tube material and geometry are characterized through a program of inspection, material surveillance and assessment and need to be managed to optimize plant performance. Candu is improving pressure tubes installed in new build and life extension projects. Improvements include changes designed to reduce or mitigate the impact of pressure tube elongation and diametral expansion rates, improvement of pressure tube fracture properties, and reduction of the implications of hydrogen ingress. In addition, Candu provides an extensive array of engineering services designed to assess the condition of pressure tubes and address the impact of pressure tube degradation on safety margins and plant performance. These services include periodic and in-service inspection and material surveillance of pressure tubes and deterministic and probabilistic assessment of pressure tube fitness for service to applicable standards. Activities designed to mitigate the impact of pressure tube deformation on safety margins include steam generator cleaning, which improves trip margins, and trip design assessment to optimize reactor trip set points restoring safety and operating margins. This paper provides an

  20. Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

    Directory of Open Access Journals (Sweden)

    Park Joo Hwan

    2016-01-01

    Full Text Available A CANDU-6 reactor, which has 380 fuel channels of a pressure tube type, is suffering from aging or creep of the pressure tubes. Most of the aging effects for the CANDU primary heat transport system were originated from the horizontal crept pressure tubes. As the operating years of a CANDU reactor proceed, a pressure tube experiences high neutron irradiation damage under high temperature and pressure. The crept pressure tube can deteriorate the Critical Heat Flux (CHF of a fuel channel and finally worsen the reactor operating performance and thermal margin. Recently, the modification of the central subchannel area with increasing inner pitch length of a standard 37-element fuel bundle was proposed and studied in terms of the dryout power enhancement for the uncrept pressure tube since a standard 37-element fuel bundle has a relatively small flow area and high flow resistance at the central region. This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. Also, it was discussed how much the crept pressure tubes affected the thermalhydraulic characteristics of the fuel channel as well as the dryout power for the modification of a standard 37-element fuel bundle.

  1. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II. [USA

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site.

  2. Localization of CANDU technology

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    Alizadeh, Ala

    2010-09-15

    The CANDU pressurized heavy water reactor's principal design features suit it particularly well for technology transfer and localization. When the first commercial CANDU reactors of 540 MWe entered service in 1971, Canada's population of less than 24 million supported a 'medium' level of industrial development, lacking the heavy industrial capabilities of larger countries like the USA, Japan and Europe. A key motivation for Canada in developing the CANDU design was to ensure that Canada would have the autonomous capacity to build and operate nuclear power reactors without depending on foreign sources for key components or enriched fuel.

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

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    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.

  4. Development of delayed hydride cracking resistant-pressure tube

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    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

  5. Tensile properties and fracture toughness of Zr–2.5Nb alloy pressure tubes of IPHWR220

    Energy Technology Data Exchange (ETDEWEB)

    Khandelwal, H.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Singh, R.N., E-mail: rnsingh@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Bind, A.K.; Sunil, S.; Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Ghosh, A.; Dhandharia, P.; Bhachawat, D. [Engineering Directorate, Nuclear Power Corporation of India Ltd., NUB, Anushaktinagar, Mumbai 400094 (India); Shekhar, R.; Kumar, Sunil Jai [National Centre for Compositional Characterisation of Materials, Bhabha Atomic Research Centre, ECIL (PO), Hyderabad 500 062 (India)

    2015-11-15

    Highlights: • Evaluated tensile properties & fracture toughness of Zr–2.5Nb pressure tube alloy. • Studied the effect of test temperature, sample location and ingot melting. • Quadruple melting improves fracture toughness despite variation in fabrication route. • Fracture toughness of IPHWR220 pressure tubes compared with CANDU material. - Abstract: The pressure tubes of Indian Pressurized Heavy Water Reactor (IPHWR) of 220 MWe are made of Zr–2.5Nb alloy manufactured either from Double Melted (DM) or from Quadruple Melted (QM) ingots. These pressure tubes are manufactured by hot extrusion, two stages of cold pilgering with intermediate annealing and autoclaving. To achieve good in-reactor performance, it is required to have minimum variability in the mechanical properties of the pressure tube across its length and between tube to tube. In this work, tensile properties and fracture toughness parameters (J{sub max}, dJ/da and CCL determined as per ASTM E1820-11 standard) of unirradiated Zr–2.5Nb alloy pressure tubes manufactured from DM and QM ingots using samples obtained from front and back end of the tubes is presented. The mechanical properties were evaluated in temperature range of 25–450 °C and compared with the corresponding data reported in literature for CANDU pressure tubes.

  6. An emergency water injection system (EWIS) for future CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre L.F. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: momarques@uol.com.br; Todreas, Neil E.; Driscoll, Michael J. [Massachusetts Inst.of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.

    2000-07-01

    This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m{sup 2}: the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

  7. Initiation of delayed hydride cracking in zirconium-niobium micro pressure tubes

    Science.gov (United States)

    Sundaramoorthy, Ravi Kumar

    Pressure tubes pick up hydrogen while they are in service within CANDU reactors. Sufficiently high hydrogen concentration can lead to hydride precipitation during reactor shutdown/repair at flaws, resulting in the potential for eventual rupture of the pressure tubes by a process called Delayed Hydride Cracking (DHC). The threshold stress intensity factor (KIH) below which the cracks will not grow by delayed hydride cracking of Zr-2.5Nb micro pressure tubes (MPTs) has been determined using a load increasing mode (LIM) method at different temperatures. MPTs have been used to allow easy study of the impact of properties like texture and grain size on DHC. Previous studies on MPTs have focused on creep and effects of stress on hydride orientation; here the use of MPTs for DHC studies is confirmed for the first time. Micro pressure tube samples were hydrided to a target hydrogen content of 100 ppm using an electrolytic method. For DHC testing, 3 mm thick half ring samples were cut out from the tubes using Electrical Discharge Machining (EDM) with a notch at the center. A sharp notch with a root radius of 15 microm was introduced by broaching to facilitate crack initiation. The direct current potential drop method was used to monitor crack growth during the DHC tests. For the temperature range tested the threshold stress intensity factors for the micro pressure tube used were found to be 6.5--10.5 MPa.m 1/2 with the value increasing with increasing temperature. The average DHC velocities obtained for the three different test temperatures 180, 230 and 250°C were 2.64, 10.87 and 8.45 x 10-8 m/s, respectively. The DHC data obtained from the MPTs are comparable to the data published in the literature for full sized CANDU pressure tubes.

  8. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  9. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, R.H.; Macici, N [Hydro-Quebec, Montreal, Quebec (Canada); Gibb, R. [New Brunswick Power, Lepreau, NB (Canada); Purdy, P.L.; Manzer, A.M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Hydro, Toronto, Ontario (Canada)

    1997-07-01

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO{sub 2} fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly

  10. Gettering of hydrogen from Zr-2. 5Nb pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Cann, C.D.; Sexton, E.E.; Bahurmuz, A.A.; White, A.J.; Balness, H.R.; Ledoux, G.A. (AECL Research, Whiteshell Labs., Pinawa, Manitoba (Canada))

    1991-09-10

    Yttrium is being investigated as a hydrogen getter to prevent delayed hydride cracking in Zr-2.5Nb pressure tubes in CANDU nuclear reactors. Yttrium strips have been encapsulated in zirconium alloy and attached to the ends of hydrided pressure tube sections to determine the effect of the degree of contact between the yttrium and the encapsulation on the gettering rate. Rates for strips hot isostatically pressed into the encapsulation were in good agreement with diffusion model predictions assuming complete contact. Rates for strips brought into contact by cold rolling were slightly lower than those for the hot-pressed strips, while little gettering was observed for loose strips sealed in the encapsulation by tungsten-inert gas welding. The effect of hydrogen flux rate to the yttrium on gettering was determined at 313degC for hydrogen fluxes from three to nine times those predicted in reactor. It was found that these fluxes did not affect the gettering rate for hydrogen concentrations up to 58 at.% in the hot isostatically pressed yttrium inserts. Inserts that were thermally cycled and inserts that had not been hot pressed achieved similar gettering rates. (orig.).

  11. Leak detection capability in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Azer, N.; Barber, D.H.; Boucher, P.J. [and others

    1997-04-01

    This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis.

  12. An improved statistical model for predicting the deuterium ingress in zirconium alloy pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Pandey, M.D., E-mail: mdpandey@uwaterloo.ca [Department of Civil and Environmental Engineering University of Waterloo, Waterloo, Ontario, N2L 3G1 (Canada); Xin, L. [Department of Civil and Environmental Engineering University of Waterloo, Waterloo, Ontario, N2L 3G1 (Canada)

    2012-09-15

    In the CANDU pressurized heavy water reactor (PHWR), the nuclear fuel is contained in hundreds of Zr-2.5 Nb alloy pressure tubes. The corrosion of zirconium alloy produces deuterium that is absorbed by the body of the pressure tube. The presence of this deuterium causes hydrogen embrittlement of zirconium alloy with an adverse effect on the integrity of the pressure tube. An accurate prediction of deuterium accumulation over time is an important step for ensuring the fitness-for-service of pressure tubes. Deuterium ingress data collected from in-service inspection of pressure tubes exhibit heteroscedasticity, i.e., the variance of deuterium concentration is dependent on operating time (or exposure) and temperature. The currently used model by the nuclear industry involves a logarithmic regression of deuterium content over time and temperature. Since this approach does not deal with heteroscedasticity precisely, it results in a conservative prediction of the deuterium ingress. The paper presents a new approach for predicting deuterium ingress based on a weighted least-squares (WLS) regression that overcomes the limitations of the existing model, and it provides realistic prediction bounds of deuterium ingress.

  13. 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

  14. Eddy current monitoring of fatigue crack growth in Zr-2.5% Nb pressure tube

    Science.gov (United States)

    Krause, T. W.; Martin, A. E.; Sheppard, R. R.; Schankula, J. J.

    2000-05-01

    Zr-2.5% wt. Nb pressure tubes (PTs) form the core of the heat transport system in CANDU nuclear reactors. These 6 m long, 100 mm diameter tubes are operated at elevated temperatures (nominally 300 °C) and at pressures that produce hoop stresses that are 25% of the ultimate tensile strength of the PT (120 Mpa). Therefore, detection and characterization of flaws in these components becomes crucial for their continued pressure retaining integrity. If a flaw is detected, however, the cost of PT replacement is expensive. Periodic in-service inspection of a flaw that demonstrates no change in flaw characteristics can be used to allow a pressure tube to remain in-service. This requires confidence in the accuracy and reliability of methods used to inter flaw characteristics. Such confidence can only be developed by comparing nondestructive predictions with results from destructive examinations. In this work, eddy current testing was used to monitor the progressive stages of a fatigue crack, grown through pressure cycling from a notch on the inner surface of a PT. Results from a differential lift-off compensated eddy current probe were used to produce sizing estimates of the crack grown between 35% (base of notch) and 74% of the PT wall. A comparison with a destructive examination of the crack demonstrated sensitivity too changes in crack depth accurate to 5% of the tube wall thickness. Such results, combined with similar information obtained from ultrasonics will increase confidence in interpretation of PT inspection data.

  15. Hybrid Monte Carlo deterministic and probabilistic core assessment for flaws and leak-before break for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shi, P.; Mok, D.H.B. [AMEC NSS, Toronto, Ontario (Canada)

    2011-07-01

    Even though pressure tubes are major components of a CANDU reactor, only small proportions of pressure tubes are sampled for in-service inspections due to execution cost, outage duration, and site cumulative radiation exposure limits. In general, a realistic core assessment was not carried out based on all known information related to in-service degradation mechanisms. Recently, a hybrid deterministic and probabilistic core assessment (HDPCA) has been introduced to address the uncertainties associated with uninspected pressure tubes and diverse degradation mechanisms. In the present paper, the HDPCA was carried out for a CANDU unit based on cumulative operating experience and history in order to satisfy the requirements of Clause 7 of CSA Standard N285.8 by considering the uncertainties associated with the estimated distribution parameters, the limited inspected data, and pressure tube properties. The HDPCA is composed of two parts: a simulation part and a deterministic evaluation part. The outcome of the core assessment is the expected pressure tube failure frequency due to pressure tube flaws. In the simulations, pressure tube material properties were sampled from distributions derived from material surveillance and testing programs. The flaw dimensions and intensities were sampled from distributions fitted to in-service inspection data. The pressure tubes were then populated with flaws. Each simulated flaw was evaluated for DHC initiation under constant loading conditions. When Delayed Hydride Cracking initiation from a flaw was predicted, the pressure tube was evaluated for rupture in the Leak-Before-Break evaluation. Based on all the predicted pressure tube ruptures from simulations, the failure frequency was calculated on an annual basis. The largest expected mean and the 95% upper bound of the mean failure frequencies for any evaluation subinterval to the end of pressure tube design life of 210,000 EFPH are significantly below the allowable failure frequency

  16. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building.

  17. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-09-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles.

  18. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Z., E-mail: chengz@aecl.ca; Rao, Y.F., E-mail: raoy@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-10-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles.

  19. CFX Analysis of the CANDU Moderator Thermal-Hydraulics in the Stern Lab. Test Facility

    Science.gov (United States)

    Kim, Hyoung Tae

    2014-06-01

    A numerical calculation with the commercial CFD code CFX is conducted for a test facility simulating the CANDU moderator thermal-hydraulics. Two kinds of moderator thermal-hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the CANDU moderator circulating vessel, which is called a Calandria, housing a matrix of horizontal rod bundles simulating the Calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the Calandria. In the present study the full geometric details of the Calandria are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data.

  20. CANDU, building the future

    Energy Technology Data Exchange (ETDEWEB)

    Stern, F. [Stern Laboratories (Canada)

    1997-07-01

    The CEO of Stern Laboratories delivered a speech on the problems and challenges facing the nuclear industry. The CANDU system is looked at as the practical choice for the future of our energy source. The people of the industry must be utilized and respected to deliver to the best of their ability.

  1. CANDU in-reactor quantitative visual-based inspection techniques

    Science.gov (United States)

    Rochefort, P. A.

    2009-02-01

    This paper describes two separate visual-based inspection procedures used at CANDU nuclear power generating stations. The techniques are quantitative in nature and are delivered and operated in highly radioactive environments with access that is restrictive, and in one case is submerged. Visual-based inspections at stations are typically qualitative in nature. For example a video system will be used to search for a missing component, inspect for a broken fixture, or locate areas of excessive corrosion in a pipe. In contrast, the methods described here are used to measure characteristic component dimensions that in one case ensure ongoing safe operation of the reactor and in the other support reactor refurbishment. CANDU reactors are Pressurized Heavy Water Reactors (PHWR). The reactor vessel is a horizontal cylindrical low-pressure calandria tank approximately 6 m in diameter and length, containing heavy water as a neutron moderator. Inside the calandria, 380 horizontal fuel channels (FC) are supported at each end by integral end-shields. Each FC holds 12 fuel bundles. The heavy water primary heat transport water flows through the FC pressure tube, removing the heat from the fuel bundles and delivering it to the steam generator. The general design of the reactor governs both the type of measurements that are required and the methods to perform the measurements. The first inspection procedure is a method to remotely measure the gap between FC and other in-core horizontal components. The technique involves delivering vertically a module with a high-radiation-resistant camera and lighting into the core of a shutdown but fuelled reactor. The measurement is done using a line-of-sight technique between the components. Compensation for image perspective and viewing elevation to the measurement is required. The second inspection procedure measures flaws within the reactor's end shield FC calandria tube rolled joint area. The FC calandria tube (the outer shell of the FC) is

  2. Research on method of pressure grouting piling of driven tube

    Institute of Scientific and Technical Information of China (English)

    Dianqi PAN; Zupei ZHANG; Diancai PAN; Yong CHEN; Maosen TAN

    2006-01-01

    The pressure grouting pile of driven tube can improve the load bearing capacity of the single pile from the mechanism of pressure grouting pile of driven tube. On the basis of analyzing the mechanism, the authors designed the machines and tools of pressure grouting, determined the operating manufacture and technology parameter on the pressure grouting secondly. The result shows that the pressure grouting pile of driven tube not only changes the pile type but also reduce the length of the pile and its engineering cost, it enhances the load bearing capacity of single pile an the same time.

  3. Remote Pressure Control - Considering Pneumatic Tubes in Controller Design

    OpenAIRE

    Rager, David; Neumann, Rüdiger; Murrenhoff, Hubertus

    2016-01-01

    In pneumatic pressure control applications the influence of tubes that connect the valve with the control volume ist mainly neglected. This can lead to stability and robustness issues and limit either control performance or tube length. Modeling and considering tube behavior in controller design procedure allows longer tubes while maintaining the required performance and robustness properties without need for manual tuning. The author\\'s previously published Simplified Fluid Transmission Line...

  4. CANDU steam generator life management: laboratory data and plant experience

    Energy Technology Data Exchange (ETDEWEB)

    Tapping, R.L. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Nickerson, J.H.; Subash, N. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Wright, M.D

    2001-10-01

    As CANDU reactors enter middle age, and the potential value of the plants in a deregulated market is realized, life management and life extension issues become increasingly important. An accurate assessment of critical components, such as the CANDU 6 steam generators (SGs), is crucial for successful life extension, and in this context, material issues are a key factor. For example, service experience with Alloy 900 tubing indicates very low levels of degradation within CANDU SGs; the same is also noted worldwide. With little field data for extrapolation, life management and life extension decisions for the tube bundles rely heavily on laboratory data. Similarly, other components of the SGs, in particular the secondary side internals, have only limited inspection data upon which to base a condition assessment. However, in this case there are also relatively little laboratory data. Decisions on life management and life extension are further complicated--not only is inspection access often restricted, but repair or replacement options for internal components are, by definition, also limited. The application of CANDU SG life management and life extension requires a judicious blend of in-service data, laboratory research and development (R and D) and materials and engineering judgment. For instance, the available laboratory corrosion and fretting wear data for Alloy 800 SG tubing have been compared with plant experience (with all types of tubing), and with crevice chemistry simulations, in order to provide an appropriate inspection guide for a 50-year SG life. A similar approach has been taken with other SG components, where the emphasis has been on known degradation mechanisms worldwide. This paper provides an outline of the CANDU SG life management program, including the results to date, a summary of the supporting R and D program showing the integration with condition assessment and life management activities, and the approach taken to life extension for a typical

  5. Pressure effect on the sensitivity of quartz Bourdon tube gauges.

    Science.gov (United States)

    Szaniszlo, A. J.

    1972-01-01

    The sensitivity change for a commercial fused quartz Bourdon tube precision pressure gauge, due to a change in absolute pressure level, has been analytically computed and experimentally confirmed. The computed differential pressure error is 2.5% of full scale at a 100 atm absolute pressure level. The experimental method compared the fused quartz Bourdon tube gauge digital output to the results obtained from a nitrogen gas pressure system which had a high pressure, well-type mercury manometer as the differential pressure reference.

  6. EC6{sup TM} - Enhanced Candu 6{sup TM} reactor safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.; Snell, V.; Cormier, M.; Hopwood, J. [Atomic Energy of Canada Ltd., 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2010-07-01

    The EC6 is a 740 MWe-class natural-uranium-fuelled, heavy-water-cooled and -moderated pressure-tube reactor, which has evolved from the eleven (11) CANDU{sup R} 6 plants operating in five countries (on four continents). CANDU 6 has over 150 reactor-years of safe operation. The most recent CANDU 6 - at Qinshan, in China - is the Reference Design for EC6. The EC6 shares many inherent, passive and engineered safety characteristics with the Reference Design. However EC6 has been designed to meet modern regulatory requirements and safety expectations. The resulting design changes have improved these safety characteristics, and this paper provides a convenient summary. The paper addresses the safety functions of reactivity control, heat removal, and containment of radioactive material. For each safety function, the EC6 characteristics are categorized as inherent, passive, or engineered. The paper focuses mostly on the first two. The Enhanced CANDU 6 uses an appropriate mix of passive, inherent, and engineered safety functions. Reactivity transients are generally slow, mild and inherently limited due to the natural uranium core and use of on-power refuelling. Only the coolant void coefficient can cause a large reactivity insertion, particularly in a large LOCA. This is mitigated by the long prompt neutron lifetime and the large delayed neutron fraction, and terminated by either of the two shutdown systems. For EC6, the large LOCA power transient has been reduced significantly by speeding up the slower of the two shutdown systems. Redundant shutdown and the LOCA power pulse improvements mitigate the limiting large positive reactivity insertion. Decay heat removal shows a very high component of passive safety, from thermo-siphoning in the Reactor Coolant System to passive heat removal in severe accidents via the moderator or reactor vault. The latter two can maintain the fuel in a more predictable and favourable geometry than 'core on the floor'. The containment

  7. Effects of extrusion-billet preheating on the microstructure and properties of Zr-2.5Nb pressure tube materials

    Energy Technology Data Exchange (ETDEWEB)

    Choubey, R.; Cann, C.D. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada). Whiteshell Labs.; Aldridge, S.A. [Nu-Tech Precision Metals, Inc., Arnprior, Ontario (Canada); Theaker, J.R.; Coleman, C.E. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada). Chalk River Labs.

    1996-12-31

    The effects of extrusion temperature and pre-heat soak time for billets on the mechanical properties of Zr-2.5Nb pressure tubes for CANDU reactors have been examined. The {beta}-quenched billets from a quadruple-melted ingot containing approximately 1,200 ppm of oxygen were extruded at 780, 815, and 850 C with pre-heat soak times of 15 to 300 min. The extruded hollows were finished by cold drawing (with a 28% reduction in area) and then stress relieving at 400 C. The {alpha}-phase grain structure, tensile strength, and fracture toughness properties were found to vary with the pre-heat temperature and soak time. All the materials were tough because embrittling impurities were absent. The tubes with 780 C preheat had a very fine and uniform {alpha}-grain structure, giving high strength and toughness at all soak times. The opposite was true for the 850 C soaks; the grain structure was coarse and inhomogeneous and the materials tended to be less strong and less touch. The tubes with the 815 C soaks showed intermediate values of strength and toughness. These variations in mechanical properties are discussed in terms of {alpha}-grain refinement and oxygen enrichment.

  8. 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......: An 84-year-old Caucasian woman with dysphagia and at risk of aspiration underwent routine insertion of a nasogastric feeding tube; however, shortly after insertion she developed respiratory distress. A chest X-ray showed the tube had been misplaced into our patient's right lung. The tube was removed......, 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...

  9. 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.

  10. Estimation of Aging Effects on LOHS for CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Yong Ki; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    To evaluate the Wolsong Unit 1's capacity to respond to large-scale natural disaster exceeding design, the loss of heat sink(LOHS) accident accompanied by loss of all electric power is simulated as a beyond design basis accident. This analysis is considered the aging effects of plant as the consequences of LOHS accident. Various components of primary heat transport system(PHTS) get aged and some of the important aging effects of CANDU reactor are pressure tube(PT) diametral creep, steam generator(SG) U-tube fouling, increased feeder roughness, and feeder orifice degradation. These effects result in higher inlet header temperatures, reduced flows in some fuel channels, and higher void fraction in fuel channel outlets. Fresh and aged models are established for the analysis where fresh model is the circuit model simulating the conditions at retubing and aged model corresponds to the model reflecting the aged condition at 11 EFPY after retubing. CATHENA computer code[1] is used for the analysis of the system behavior under LOHS condition. The LOHS accident is analyzed for fresh and aged models using CATHENA thermal hydraulic computer code. The decay heat removal is one of the most important factors for mitigation of this accident. The major aging effect on decay heat removal is the reduction of heat transfer efficiency by steam generator. Thus, the channel failure time cannot be conservatively estimated if aged model is applied for the analysis of this accident.

  11. Modulated pressure waves in large elastic tubes.

    Science.gov (United States)

    Mefire Yone, G R; Tabi, C B; Mohamadou, A; Ekobena Fouda, H P; Kofané, T C

    2013-09-01

    Modulational instability is the direct way for the emergence of wave patterns and localized structures in nonlinear systems. We show in this work that it can be explored in the framework of blood flow models. The whole modified Navier-Stokes equations are reduced to a difference-differential amplitude equation. The modulational instability criterion is therefore derived from the latter, and unstable patterns occurrence is discussed on the basis of the nonlinear parameter model of the vessel. It is found that the critical amplitude is an increasing function of α, whereas the region of instability expands. The subsequent modulated pressure waves are obtained through numerical simulations, in agreement with our analytical expectations. Different classes of modulated pressure waves are obtained, and their close relationship with Mayer waves is discussed.

  12. Pressure Loss across Tube Bundles in Two-phase Flow

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Woo Gun; Banzragch, Dagdan [Hannam Univ., Daejon (Korea, Republic of)

    2016-03-15

    An analytical model was developed by Sim to estimate the two-phase damping ratio for upward two-phase flow perpendicular to horizontal tube bundles. The parameters of two-phase flow, such as void fraction and pressure loss evaluated in the model, were calculated based on existing experimental formulations. However, it is necessary to implement a few improvements in the formulations for the case of tube bundles. For the purpose of the improved formulation, we need more information about the two-phase parameters, which can be found through experimental test. An experiment is performed with a typical normal square array of cylinders subjected to the two-phase flow of air-water in the tube bundles, to calculate the two-phase Euler number and the two-phase friction multiplier. The pitch-to-diameter ratio is 1.35 and the diameter of cylinder is 18mm. Pressure loss along the flow direction in the tube bundles is measured with a pressure transducer and data acquisition system to calculate the two-phase Euler number and the two-phase friction multiplier. The void fraction model by Feenstra et al. is used to estimate the void fraction of the two-phase flow in tube bundles. The experimental results of the two phase friction multiplier and two-phase Euler number for homogeneous and non-homogeneous two-phase flows are compared and evaluated against the analytical results given by Sim's model.

  13. Radiological Characteristics of decommissioning waste from a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahmed, Rizwan; Heo, Gyun Young [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

    2011-11-15

    The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be 1.04 x 10{sup 16} Bq, 2.09 x 10{sup 3} W, 5.31 x 10{sup 14} m{sup 3}-water, 4.69 x 10{sup 5} kg, and 7.38 x 10{sup 1} m{sup 3}, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

  14. Advanced CFD simulations of turbulent flows around appendages in CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Hadaller, G.I.; Fortman, R.A., E-mail: fabbasian@sternlab.com [Stern Laboratories Inc., Hamilton, Ontario (Canada)

    2013-07-01

    Computational Fluid Dynamics (CFD) was used to simulate the coolant flow in a modified 37-element CANDU fuel bundle, in order to investigate the effects of the appendages on the flow field. First, a subchannel model was created to qualitatively analyze the capabilities of different turbulence models such as k.ε, Reynolds Normalization Group (RNG), Shear Stress Transport (SST) and Large Eddy Simulation (LES). Then, the turbulence model with the acceptable quality was used to investigate the effects of positioning appendages, normally used in CANDU 37-element Critical Heat Flux (CHF) experiments, on the flow field. It was concluded that the RNG and SST models both show improvements over the k.ε method by predicting cross flow rates closer to those predicted by the LES model. Also the turbulence effects in the k.ε model dissipate quickly downstream of the appendages, while in the RNG and SST models appear at longer distances similar to the LES model. The RNG method simulation time was relatively feasible and as a result was chosen for the bundle model simulations. In the bundle model simulations it was shown that the tunnel spacers and leaf springs, used to position the bundles inside the pressure tubes in the experiments, have no measureable dominant effects on the flow field. The flow disturbances are localized and disappear at relatively short streamwise distances. (author)

  15. Development of CANDU ECCS performance evaluation methodology and guides

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Park, Kyung Soo; Chu, Won Ho [Korea Maritime Univ., Jinhae (Korea, Republic of)

    2003-03-15

    The objectives of the present work are to carry out technical evaluation and review of CANDU safety analysis methods in order to assist development of performance evaluation methods and review guides for CANDU ECCS. The applicability of PWR ECCS analysis models are examined and it suggests that unique data or models for CANDU are required for the following phenomena: break characteristics and flow, frictional pressure drop, post-CHF heat transfer correlations, core flow distribution during blowdown, containment pressure, and reflux rate. For safety analysis of CANDU, conservative analysis or best estimate analysis can be used. The main advantage of BE analysis is a more realistic prediction of margins to acceptance criteria. The expectation is that margins demonstrated with BE methods would be larger that when a conservative approach is applied. Some outstanding safety analysis issues can be resolved by demonstration that accident consequences are more benign than previously predicted. Success criteria for analysis and review of Large LOCA can be developed by top-down approach. The highest-level success criteria can be extracted from C-6 and from them, the lower level criteria can be developed step-by-step, in a logical fashion. The overall objectives for analysis and review are to verify radiological consequences and frequency are met.

  16. Decay of weak pressure waves in a low-pressure tube

    Energy Technology Data Exchange (ETDEWEB)

    Takiya, Toshio; Terada, Yukihiro; Komura, Akio [Hitachi Zosen Corp., Osaka (Japan); Higashino, Fumio; Abe, Hideaki; Ando, Masami

    1997-05-01

    In this study, the characteristics of pressure wave propagation in a vacuum tube have been investigated experimentally from the viewpoint of vacuum protection in the beamlines of a synchrotron radiation facility. Baffle plates having a single orifice of 5, 10 or 15 mm in diameter were installed in shock tubes 5 m in length and 36.6 or 68.8 mm in diameter, in order to slow the pressure wave or shock wave propagation, as a model for the beamline. To evaluate the decay of pressure waves, pressure changes with time at several locations along the side wall as well as at the end wall of the tube were measured. The results showed that the effect of the orifices on pressure wave propagation and its decay was significant. The present investigation may contribute to the design and construction of high-energy synchrotron radiation facilities with long beamlines. (author)

  17. Nondestructive examination of PHWR pressure tube using eddy current technique

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee Jong; Choi, Sung Nam; Cho, Chan Hee; Yoo, Hyun Joo; Moon, Gyoon Young [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2014-06-15

    A pressurized heavy water reactor (PHWR) core has 380 fuel channels contained and supported by a horizontal cylindrical vessel known as the calandria, whereas a pressurized water reactor (PWR) has only a single reactor vessel. The pressure tube, which is a pressure-retaining component, has a 103.4 mm inside diameter x 4.19 mm wall thickness, and is 6.36 m long, made of a zirconium alloy (Zr-2.5 wt% Nb). This provides support for the fuel while transporting the D2O heat-transfer fluid. The simple tubular geometry invites highly automated inspection, and good approach for all inspection. Similar to all nuclear heat-transfer pressure boundaries, the PHWR pressure tube requires a rigorous, periodic inspection to assess the reactor integrity in accordance with the Korea Nuclear Safety Committee law. Volumetric-based nondestructive evaluation (NDE) techniques utilizing ultrasonic and eddy current testing have been adopted for use in the periodic inspection of the fuel channel. The eddy current testing, as a supplemental NDE method to ultrasonic testing, is used to confirm the flaws primarily detected through ultrasonic testing, however, eddy current testing offers a significant advantage in that its ability to detect surface flaws is superior to that of ultrasonic testing. In this paper, effectiveness of flaw detection and the depth sizing capability by eddy current testing for the inside surface of a pressure tube, will be introduced. As a result of this examination, the ET technique is found to be useful only as a detection technique for defects because it can detect fine defects on the surface with high resolution. However, the ET technique is not recommended for use as a depth sizing method because it has a large degree of error for depth sizing.

  18. Effect of DUPIC cycle on CANDU reactor safety parameters

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M. A. [Atomic Energy Authority, ETRR-2, Cairo (Egypt); Badawi, Alya [Dept. of Nuclear and Radiation Engineering, Alexandria University, Alexandria (Egypt)

    2016-10-15

    Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO{sub 2} enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

  19. Microhole High-Pressure Jet Drill for Coiled Tubing

    Energy Technology Data Exchange (ETDEWEB)

    Ken Theimer; Jack Kolle

    2007-06-30

    Tempress Small Mechanically-Assisted High-Pressure Waterjet Drilling Tool project centered on the development of a downhole intensifier (DHI) to boost the hydraulic pressure available from conventional coiled tubing to the level required for high-pressure jet erosion of rock. We reviewed two techniques for implementing this technology (1) pure high-pressure jet drilling and (2) mechanically-assisted jet drilling. Due to the difficulties associated with modifying a downhole motor for mechanically-assisted jet drilling, it was determined that the pure high-pressure jet drilling tool was the best candidate for development and commercialization. It was also determined that this tool needs to run on commingled nitrogen and water to provide adequate downhole differential pressure and to facilitate controlled pressure drilling and descaling applications in low pressure wells. The resulting Microhole jet drilling bottomhole assembly (BHA) drills a 3.625-inch diameter hole with 2-inch coil tubing. The BHA consists of a self-rotating multi-nozzle drilling head, a high-pressure rotary seal/bearing section, an intensifier and a gas separator. Commingled nitrogen and water are separated into two streams in the gas separator. The water stream is pressurized to 3 times the inlet pressure by the downhole intensifier and discharged through nozzles in the drilling head. The energy in the gas-rich stream is used to power the intensifier. Gas-rich exhaust from the intensifier is conducted to the nozzle head where it is used to shroud the jets, increasing their effective range. The prototype BHA was tested at operational pressures and flows in a test chamber and on the end of conventional coiled tubing in a test well. During instrumented runs at downhole conditions, the BHA developed downhole differential pressures of 74 MPa (11,000 psi, median) and 90 MPa (13,000 psi, peaks). The median output differential pressure was nearly 3 times the input differential pressure available from the

  20. Neutronics-thermalhydraulics coupling in a CANDU SCWR

    Science.gov (United States)

    Adouki, Pierre

    In order to implement new nuclear technologies as a solution to the growing demand for energy, 10 countries agreed on a framework for international cooperation in 2002, to form the Generation IV International Forum (GIF). The goal of the GIF is to design the next generation of nuclear reactors that would be cost effective and would enhance safety. This forum has proposed several types of Generation IV reactors including the Supercritical Water-Cooled Reactor (SCWR). The SCWR comes in two main configurations: pressure vessel SCWR and pressure tube SCWR (PT-SCWR). In this study, the CANDU SCWR (a PT-SCWR) is considered. This reactor is oriented vertically and contains 336 channels with a length of 5 m. The target coolant inlet and outlet temperatures are 350 Celsius and 625 Celsius, respectively. The coolant flows downwards, and the reactor power is 2540 MWth. Various fuel designs have been considered in order not to exceed the linear element rating. However, the dependency between the core power and thermalhydraulics parameters results in the necessity to use a neutronics/thermalhydaulics coupling scheme to determine the core power and the thermalhydraulics parameters. The core power obtained has a power peaking factor of 1.4. The bundle power distribution for all channels has a peak at the third bundle from the inlet, but the value of this peak increases with the channel power. The heat-transfer coefficient and the specific-heat capacity have a peak at the same location in a channel, and this location shifts toward the inlet as the channel power increases. The exit coolant temperature increases with the channel power, while the exit coolant density and pressure decrease with the channel power. Also, higher channel powers lead to higher fuel and cladding temperatures. Moreover, as the coupling method is applied, the effective multiplication factor and the values of thermalhydaulics parameters oscillate as they converge.

  1. The small (or large) modular CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Meneley, D.; Harvel, G. [Univ. of Ontario Inst. of Tech., Oshawa, Ontario (Canada)

    2013-07-01

    This presentation outlines the design for small (or large) modular CANDU. The origins of this work go back many years to a comment by John Foster, then President of AECL CANDU. Foster noted that the CANDU reactor, with its many small fuel channels, was like a wood campfire. To make a bigger fire, just throw on some more logs (channels). If you want a smaller fire, just use fewer logs. The design process is greatly simplified.

  2. Collapse of composite tubes under uniform external hydrostatic pressure

    Science.gov (United States)

    Smith, P. T.; Ross, C. T. F.; Little, A. P. F.

    2009-08-01

    This paper describes an experimental and a theoretical investigation into the collapse of 22 circular cylindrical composite tubes under external hydrostatic pressure. The investigations were on the collapse of fibre reinforced plastic tube specimens made from a mixture of three carbon and two E-glass fibre layers. The theoretical investigations were carried out using an in-house finite element computer program called BCLAM, together with the commercial computer package, namely ANSYS. It must be emphasised here that BS 5500 does not appear to exclusively cater for the buckling of composite shells under external hydrostatic pressure, so the work presented here is novel and should be useful to industry. The experimental investigations showed that the composite specimens behaved similarly to isotropic materials previously tested, in that the short vessels collapsed through axisymmetric deformation while the longer tubes collapsed through non-symmetric bifurcation buckling. Furthermore it was discovered that the models failed at changes of the composite lay-up due to the manufacturing process of these models. These changes seemed to be the weak points of the specimens.

  3. Towards a shock tube method for the dynamic calibration of pressure sensors.

    Science.gov (United States)

    Downes, Stephen; Knott, Andy; Robinson, Ian

    2014-08-28

    In theory, shock tubes provide a pressure change with a very fast rise time and calculable amplitude. This pressure step could provide the basis for the calibration of pressure transducers used in highly dynamic applications. However, conventional metal shock tubes can be expensive, unwieldy and difficult to modify. We describe the development of a 1.4 MPa (maximum pressure) shock tube made from unplasticized polyvinyl chloride pressure tubing which provides a low-cost, light and easily modifiable basis for establishing a method for determining the dynamic characteristics of pressure sensors.

  4. Magnetic pressure in electromagnetic tube forming with echelon coil

    Institute of Scientific and Technical Information of China (English)

    ZHAO Zhi-heng; YU Hai-ping; LI Chun-feng; LI Zhong

    2008-01-01

    The effects of geometrical characteristics of echelon coil on the magnetic pressure distribution and their contribution to the final shape of parts were focused and investigated through experiments and numerical simulation using FEM software ANSYS.The results show that the geometrical characteristics of echelon coil play a key role in controlling the magnetic pressure acting on the tube.They show a hump·like distribution near the interface between bigger diameter region and transition region of echelon coil,and affect the final shape of tubular parts then.With the reduction of relative diameter,the magnetic pressure in smaller diameter region decreases and its distribution gradient in transition region increases.With the augment of relative length,the magnetic pressure increases in bigger diameter region,while it almost remains constant in smaller diameter region,and the gradient in transition region enhances sharply.The distribution of magnetic pressure in the axial direction of tube agrees well with the profile of specimen.

  5. Development of a System Dynamics Model for Evaluating the Economics of an Advanced CANDU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jong Yeob; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    Since the early 1990's, the Korea Atomic Energy Research Institute (KAERI) and the Atomic Energy of Canada Limited (AECL) have cooperated to develop, verify, and demonstrate the advanced CANDU fuel, so called CANFLEX-NU (Natural Uranium). The CANFLEX-NU fuel bundle consists of 43 fuel elements and has the buttons on the outer surface of the fuel elements for improving the CHF (Critical-Heat-Flux) characteristics. Because of this features of CANFLEXNU fuel, it offers higher operating and safety margins than current 37-element fuel. Recently, the interest for a CANFLEX-NU has been increased because of the power de-rating due to aging of CANDU reactors. Wolsong Unit 1 CANDU reactor has been operated over 25 years and the operating power at the present time is less than 90% of a full power because of a reduction of the margin of ROP trip set point. The most appropriate way to overcome such a power de-rating due to a crept pressure tube is the introduction of a CANFLEX-NU fuel into a CANDU reactor. Now, a CANFLEX-NU fuel is ready to be commercialized in a CANDU-6 reactor because the design and demonstration irradiation have been completed in both Korea and Canada. Economic evaluation for commercializing a CANFLEX-NU fuel in Wolsong Units was carried out by calculating the unit prime cost of electricity production. Throughout the economic evaluation, it was found that the introduction of CANFLEX-NU fuel into Wolsong Units would have much economic benefits due to a better operating performance. However, the amount of economic profit due to introducing CANFLEX-NU fuel depends on several parameters such as the required time to get license from regulatory institute before commercializing, licensing cost, failure probability of commercializing etc. Therefore, it is necessary to determine the optimum condition to get the highest economic profit. In this paper, an economic evaluation was carried out based on the starting year of the licensing study with considering the

  6. Steam generator tube failures

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  7. Effects of Fin Shape on Condensation Heat Transfer and Pressure Drop inside Herringbone Micro Fin Tubes

    Science.gov (United States)

    Miyara, Akio; Otsubo, Yusuke; Ohtsuka, Satoshi

    Experiments of in-tube condensation of R410A have been carried out for as mooth tube, a h elical micro fin tube and five types of herringbone micro fin tubes. In the herringbone micro fin tube, the micro fins work to remove liquid at fin-diverging parts and collect liquid at fin-converging parts. In the high mass velocity region, heat transfer coefficient of all the herringbone tubes is about 2-4 times higher than that of the helical micro fin tube. In the low mass velocity region, however, the heat transfer coefficients of the herringbone micro fin tubes are equal to or smaller than those of the helical micro fin tube. Up to the fin height of 0.18 mm, the heat transfer coefficient is higher for higher fin, whereas that of ah igher fin tube is saturated. The pressure drop increases with increasing fin height. The helix angle strongly affects the heat transfer and pressure drop. Higher helix angle causes higher heat transfer coefficient and higher pressure drop. In the case of the herringbone tube which has shorter fin and/or smaller helix angle, pressure drops are equal to or lower than that of the helical micro fin tube, whereas those of other tubes are higher.

  8. Leningrad nuclear power plant pressure tube failure investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bruchertseifer, H.; Bart, G.; Restani, R. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Aden, V.G.; Abramov, V.Y.; Kalachikov, V.E.; Kozlov, A.V. [Research and Development Inst. of Power Engineering (RDIPE), Moscow and Sverdlovsk (Russian Federation); Subbotin, A.V.; Smirnov, E.A. [Moscow Engineering Physics Inst., Moscow (Russian Federation)

    1996-09-01

    During March 1992 a fuel pressure tube of a reactor channel of the Leningrad Nuclear Power Plant underwent a temperature excursion after a coolant flow blockage and was destroyed. In the following, within the Swiss Eastern European aid program a collaboration was set up for a project between the Moscow Research and Development Institute of Power Engineering, the designer of the RBMK-reactors, and the Paul Scherrer Institute. An intensive failure analysis program was started, based on modern equipment available at PSI for analysis of highly radioactive material and on the experience of both institutes in investing failures of reactor structure materials, with the goal of establishing the accident temperature evolution in time. This report presents the results of studies undertaken in order to determine the parameters which govern the events during the accident obtained from an analysis of the tube failure material together with evaluations of the apparent phase and structure changes. Our analysis of experimental data for oxygen distribution and the diffusion coefficient calculations showed that the temperatures exceeded 1300{sup o}C, which is much higher than results from previous studies performed in standard failure post-irradiation examination. The results obtained are important in that they have allowed to revise the previous assessments of the initial thermal conditions of the accident progression. In particular, they already served as a basis for determining the efficiency of the RBMK safety improvement measures carried out in response to the accident. (author) 8 figs., 5 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. S.; Jeong, Y. M.; Ahn, S. B. (and others)

    2005-03-15

    Major degradation of the feeder pipe is the thinning due to the flow accelerated corrosion and the cracking in the bent region due to the stress corrosion cracking. The feeder pipe in a PHWR is a pipe to supply the coolant to the pressure tube and the heated coolant to the steam generator for power generation. Approximately 380 pipes are installed on the inlet side and outlet side each with two bent regions in the 600 MW-class PHWR. After a leakage in the bent region of the feeder pipe, it is required to examine all the pipes in order to ensure the integrity of the pressure boundaries. It is not easy, however, to examine all the pipes with the conventional ultrasonic method, because of a high dose of radiation exposure and a limited accessibility to the pipe. In order to get rid of the limited accessibility, the ultrasonic guided wave method are developed for detection and evaluation of the cracks in the feeder pipe. The dispersion mode analysis was performed for the development of long-range guided wave inspection for the feeder pipe. An analytical approach for the straight pipe as well as numerical approach for the bent pipe with 2-D FFT were accomplished. A computer program for the calculation of the dispersion curves and wave structures was developed. Based on the dispersion curves and wave structure of the feeder pipe, candidates for the optimal parameters on the frequencies and vibration modes were selected. A time-frequency analysis methodology was developed for the mode identification of received ultrasonic signal. A high power tone-burst ultrasonic system has been setup for the generation of guided waves. Various artificial notches were fabricated on the bent feeder pipes for the experiment on the flaw detection. Considering the results of dispersion analysis and field condition, the torsional vibration mode, T(0,1) is selected for the first choice. An array of electromagnetic acoustic transducers (EMAT) was designed and fabricated for the generation of T

  10. Convection heat transfer and pressure drop in cross flow over finned tubes

    Energy Technology Data Exchange (ETDEWEB)

    Baran, M.; Pronobis, M.

    1984-05-15

    This paper reports the results of an experimental study on the heat transfer and pressure drop in finned tube banks. The measurements were carried out for the tubes with fins arranged parallel and a certain angle to the flow direction. The performance of such a heat exchanger with that of the conventional one i.e. plain tube heat exchanger, is compared.

  11. A pressurized drop-tube furnace for coal reactivity studies

    Science.gov (United States)

    Ouyang, Shan; Yeasmin, Hasina; Mathews, Joseph

    1998-08-01

    The design and characterization of a pressurized drop-tube furnace for investigation of coal devolatilization, gasification, and combustion are presented. The furnace is designed for high-temperature, isothermal operation in a developing laminar flow regime. It can be operated at pressures up to 1600 kPa, and temperatures up to 1673 K, with variable reaction time, particle feeding rate, and with inert and various oxidizing atmospheres. Particle residence times can be varied between ˜0.02 and ˜10 s depending upon operating conditions and positions of injection and sampling probes. Observations ports are available for sample collections and for optical investigation of the reactions or temperature measurements. Characterization of gas temperature in the furnace shows that, although the gas temperature profile in the furnace is affected by the water-cooled injection probe, the furnace is able to achieve isothermal operation in a developing laminar flow regime. Results from a series of brown coal devolatilization tests demonstrated the suitability of the furnace for experiments in coal research.

  12. LBB in Candu plants

    Energy Technology Data Exchange (ETDEWEB)

    Kozluk, M.J.; Vijay, D.K. [Ontario Hydro Nuclear, Toronto, Ontario (Canada)

    1997-04-01

    Postulated catastrophic rupture of high-energy piping systems is the fundamental criterion used for the safety design basis of both light and heavy water nuclear generating stations. Historically, the criterion has been applied by assuming a nonmechanistic instantaneous double-ended guillotine rupture of the largest diameter pipes inside of containment. Nonmechanistic, meaning that the assumption of an instantaneous guillotine rupture has not been based on stresses in the pipe, failure mechanisms, toughness of the piping material, nor the dynamics of the ruptured pipe ends as they separate. This postulated instantaneous double-ended guillotine rupture of a pipe was a convenient simplifying assumption that resulted in a conservative accident scenario. This conservative accident scenario has now become entrenched as the design basis accident for: containment design, shutdown system design, emergency fuel cooling systems design, and to establish environmental qualification temperature and pressure conditions. The requirement to address dynamic effects associated with the postulated pipe rupture subsequently evolved. The dynamic effects include: potential missiles, pipe whipping, blowdown jets, and thermal-hydraulic transients. Recent advances in fracture mechanics research have demonstrated that certain pipes under specific conditions cannot crack in ways that result in an instantaneous guillotine rupture. Canadian utilities are now using mechanistic fracture mechanics and leak-before-break assessments on a case-by-case basis, in limited applications, to support licensing cases which seek exemption from the need to consider the various dynamic effects associated with postulated instantaneous catastrophic rupture of high-energy piping systems inside and outside of containment.

  13. Condensation inside tubes: Computer program for pressure drop in straight tubes (horizontal and vertical with downflow)

    Science.gov (United States)

    1993-12-01

    ESDU 93014 introduces a Fortran program that implements the calculation procedures of ESDU 90024 and 91023 respectively for vertical and horizontal cases. Those documents should be consulted for details of the empirical correlation used. Since vapor density is an important variable in the calculation and is usually available as a function of saturation temperature, the relationship between pressure and saturation temperature is required at points along the tube, although a constant value of vapor density may be used if the user wishes. The program provides options to use an Antoine or Wagner equation, or to provide a set of values of saturation pressure and temperature; for the vapor density the options are to use the ideal gas law, to provide a set of values of saturation temperature and density or to use a specific correlation equation (log density as a fraction of critical as a five term polynomial function of reciprocal reduced temperature minus one). For a wide range of pure compounds the ESDU Physical Data, Chemical Engineering Sub-series provides values of the constants in the correlation equations for saturation temperature and vapor density. The program (ESDUpac A9314) is provided on disc (uncompiled) in the software volume, and also compiled within ESDUview, a user-friendly shell running under MS DOS that prompts on screen for the input data. A worked example illustrates the use of the program and the formats of the input data and the output.

  14. Deadly pressure pneumothorax after withdrawal of misplaced feeding tube

    DEFF Research Database (Denmark)

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

    2016-01-01

    : An 84-year-old Caucasian woman with dysphagia and at risk of aspiration underwent routine insertion of a nasogastric feeding tube; however, shortly after insertion she developed respiratory distress. A chest X-ray showed the tube had been misplaced into our patient's right lung. The tube was removed...

  15. Verification tests for CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1997-07-01

    For the development of a CANDU advanced fuel, the CANFLEX-NU fuel bundles were tested under reactor operating conditions at the CANDU-Hot test loop. This report describes test results and test methods in the performance verification tests for the CANFLEX-NU bundle design. The main items described in the report are as follows. - Fuel bundle cross-flow test - Endurance fretting/vibration test - Freon CHF test - Production of technical document. (author). 25 refs., 45 tabs., 46 figs.

  16. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, M. K.; Lee, W. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented. 12 refs., 26 figs., 3 tabs. (Author)

  17. Computer simulation of the behaviour and performance of a CANDU fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Marino, A.C. [Comison Nacional de Energia Atomica (Argentina)

    1997-07-01

    At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

  18. Endotracheal tube cuff pressure increases significantly during anterior cervical fusion with the Caspar instrumentation system.

    Science.gov (United States)

    Sperry, R J; Johnson, J O; Apfelbaum, R I

    1993-06-01

    To determine whether endotracheal tube cuff pressure increases significantly with surgical retraction and cervical spine distraction during anterior cervical spine surgery with Caspar instrumentation, we prospectively studied 10 patients undergoing this procedure. The tracheas of all patients were intubated with a Mallinckrodt Hi-Lo endotracheal tube. Tracheal tube cuff pressures measured with a transducer system were 42.4 mm Hg +/- 7.0 mm Hg (SEM) after intubation and cuff inflation. Air was removed from the endotracheal tube cuff until the trachea was just barely sealed at a cuff pressure of 15.2 mm Hg +/- 1.6 mm Hg. The endotracheal tube cuff pressure was readjusted to "just-seal" pressure before the surgeons introduced the Caspar instrumentation. The cuff pressure with traction and distraction was 43.2 mm Hg +/- 5.0 mm Hg. This pressure was significantly increased from the "just-seal" pressure, and from the cuff pressure after instrumentation was discontinued (9.8 mm Hg +/- 2.3 mm Hg). We conclude that anterior cervical spine surgery with Caspar instrumentation is associated with a significant increase in endotracheal tube cuff pressure.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ardhapurkar, P. M. [Mechanical Engineering Department, Indian Institute of Technology Bombay, Mumbai, MS 400 076 India and S. S. G. M. College of Engineering Shegaon, MS 444 203 (India); Sridharan, Arunkumar; Atrey, M. D. [Mechanical Engineering Department, Indian Institute of Technology Bombay, Mumbai, MS 400 076 (India)

    2014-01-29

    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.

  20. Zirconium pressure tube testing: Test procedures, Production Assurance Program (Project H-700)

    Energy Technology Data Exchange (ETDEWEB)

    Zaloudek, F.R.; Lewis, M. [Pacific Northwest Lab., Richland, WA (United States)

    1986-06-01

    UNC Nuclear Industries (UNC) has initiated a plan for the fabrication of zirconium alloy pressure tubes required for the future operation of N-Reactor. As part of this plan, UNC is establishing a program to qualify and develop a process capable of fabricating these pressure tubes to the requirements of UNC specification HWS 6502, REV. 4, Amendment 1. The objective of the Pressure Tube Testing Task is to support the UNC program-by performing physical, mechanical and chemical testing on prototype tube sections produced during FY-1986, 1987 and 1988 and to test samples from production runs after 1988 as may be required. The types of tests included in the Zirconium Pressure Tube Testing Program will be as follows: tensile tests; burst tests; fracture toughness tests; corrosion tests; chemical composition analyses; grain structure evaluations. The purpose of this document is to define the procedures that will be used in each type of test included in this task.

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

    OpenAIRE

    Thomas Berlet; Mathias Marchon

    2016-01-01

    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, air...

  2. Development of High-Performance Pressure Tube Material for the Canadian SCWR Concept

    Science.gov (United States)

    Walters, L.; Donohue, S.

    2016-02-01

    The Canadian super-critical water-cooled reactor (SCWR) concept is moderated by using heavy water, while the coolant is light water at 25 MPa with an inlet temperature of 625 K and an outlet temperature of 900 K. The fuel assemblies reside in vertical pressure tubes that are the pressure boundary. The pressure tubes are insulated from the fuel assemblies and operate at temperatures near the moderator temperature, at 390 K. The zirconium alloy Excel has been selected as a candidate material for the pressure tube based on favorable properties such as high strength, resistance to radiation-induced diametral strain, and high terminal solid solubility. However, significant future effort will be required to obtain material properties and crack initiation mechanisms at super-critical water (SCW) conditions to verify that annealed Excel is a viable option as a pressure tube material in the Canadian SCWR.

  3. DETERMINISTIC EVALUATION OF DELAYED HYDRIDE CRACKING BEHAVIORS IN PHWR PRESSURE TUBES

    Directory of Open Access Journals (Sweden)

    YOUNG-JIN OH

    2013-04-01

    Full Text Available Pressure tubes made of Zr-2.5 wt% Nb alloy are important components consisting reactor coolant pressure boundary of a pressurized heavy water reactor, in which unanticipated through-wall cracks and rupture may occur due to a delayed hydride cracking (DHC. The Canadian Standards Association has provided deterministic and probabilistic structural integrity evaluation procedures to protect pressure tubes against DHC. However, intuitive understanding and subsequent assessment of flaw behaviors are still insufficient due to complex degradation mechanisms and diverse influential parameters of DHC compared with those of stress corrosion cracking and fatigue crack growth phenomena. In the present study, a deterministic flaw assessment program was developed and applied for systematic integrity assessment of the pressure tubes. Based on the examination results dealing with effects of flaw shapes, pressure tube dimensional changes, hydrogen concentrations of pressure tubes and plant operation scenarios, a simple and rough method for effective cooldown operation was proposed to minimize DHC risks. The developed deterministic assessment program for pressure tubes can be used to derive further technical bases for probabilistic damage frequency assessment.

  4. 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.

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

    Science.gov (United States)

    Berlet, Thomas; Marchon, Mathias

    2016-01-01

    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.

  6. Decay of weak pressure waves in a low pressure tube; Teiatsu kannai ni okeru bisho atsuryokuha no gensui

    Energy Technology Data Exchange (ETDEWEB)

    Takiya, T.; Terada, Y.; Komura, A. [Hitachi Zosen Corp., Osaka (Japan); Higashino, F.; Abe, H. [Tokyo Univ. of Agriculture and Technology, Tokyo (Japan). Faculty of Technology; Abe, M. [National Lab. for High Energy Physics, Tsukuba (Japan)

    1996-04-25

    The characteristics of pressure wave propagation in a vacuum tube have been investigated experimentally from the viewpoint of vacuum protection in the beam lines of a synchrotrons radiation facility. Baffle plates having a single orifice of 5, 10 or 15 mm in diameter were installed in shock tubes 5 m in length, and 36.6 or 68.8 mm in diameter, in order to show the pressure wave or shock wave propagation as a model for the beam line. To evaluate the decay of pressure waves pressure changes with time at several locations along the side wall as well as at the end wall of the tube were measured. The results show that the effect of the orifices on pressure wave propagation and its decay is significant. The present investigation may contribute to the design and construction of high energy synchrotrons radiation facilities with long beam lines. 11 refs., 9 figs., 2 tabs.

  7. 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.

  8. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    Science.gov (United States)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  9. Sensitivity Analysis of Dousing Spray Trip on Radioactive Release in Pressure Tube Rupture Accident with Both End Fitting Failures

    Energy Technology Data Exchange (ETDEWEB)

    Jang, M. S.; Kang, H. S; Kim, S. R. [NESS, Daejeon (Korea, Republic of)

    2015-10-15

    We analyzed the sensitivity analysis of dousing spray trip conditions on radioactive release. In terms of conservativeness, the set 1 trip would be more appropriate in RR analysis than set 2 trip, which is the general condition of RR analysis. Radioactive releases from the containment building is related to containment air pressure, which increases by the coolant discharge from loss of coolant accident and the actuation conditions of dousing spray and so on. In LOCA analysis, the dousing spray trip conditions are set for the analysis objectives; for peak pressure (PP), for pressure signal (PS), for radioactive release (RR) and etc. In RR analysis, we would determine the dousing spray trip condition to increase radioactive release to the public for conservatism. Therefore, we carried out the sensitivity analysis of dousing spray trip condition on radioactive release from containment building using GOTHIC and SMART program for CANDU.

  10. Overview of methods to increase dryout power in CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com [Chalk River Laboratories, AECL, Chalk River (Canada); University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada); Leung, L.K.H. [Chalk River Laboratories, AECL, Chalk River (Canada); Park, J.H. [Korean Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-06-15

    Highlights: • Small changes in bundle geometry can have noticeable effects on the bundle CHF. • Rod spacing devices can results in increases of over 200% in CHF. • CHF enhancement decays exponentially downstream from spacers. • CHF-enhancing bundle appendages also increase the post-CHF heat transfer. - Abstract: In CANDU reactors some degradation in the CCP (critical channel power, or power corresponding to the first occurrence of CHF in any fuel channel) will occur with time because of ageing effects such as pressure-tube diametral creep, increase in reactor inlet-header temperature, increased hydraulic resistance of feeders. To compensate for the ageing effects, various options for recovering the loss in CCP are described in this paper. They include: (i) increasing the bundle heated perimeter, (ii) optimizing the bundle configuration, (iii) optimizing core flow and flux distribution, (iv) reducing the bundle hydraulic resistance, (v) use of CHF-enhancing bundle appendages, (vi) more precise experimentation, and (vii) redefining CHF. The increase in CHF power has been quantified based on experiments on full-scale bundles and subchannel code predictions. The application of several of these CHF enhancement principles has been used in the development of the 43-rod CANFLEX bundle.

  11. Simulation of In-Core Dose Rates for an Offline CANDU Reactor

    Science.gov (United States)

    Gilbert, Jordan

    This thesis describes the development of a Monte Carlo simulation to predict the dose rates that will be encountered by a novel robotic inspection system for the pressure tubes of an offline CANDU reactor. Simulations were performed using the Monte Carlo N-Particle (MCNP) radiation transport code, version 6.1. The radiation fields within the reactor, even when shut down, are very high, and can cause significant damage to certain structural components and the electronics of the inspection system. Given that the robotic system will rely heavily on electronics, it is important to know the dose rates that will be encountered, in order to estimate the component lifetimes. The MCNP simulation was developed and benchmarked against information obtained from Ontario Power Generation and the Canadian Nuclear Laboratories. The benchmarking showed a good match between the simulated values and the expected values. This simulation, coupled with the accompanying user interface, represent a tool in dose field prediction that is currently unavailable. Predicted dose rates for a postulated inspection at 7 days after shutdown, with 2:5 cm of tungsten shielding around the key components, would survive for approximately 7 hours in core. This is anticipated to be enough time to perform an inspection and shows that the use of this tool can aid in designing the new inspection system.

  12. Surface modification of tube inner wall by transferred atmospheric pressure plasma

    Science.gov (United States)

    Chen, Faze; Liu, Shuo; Liu, Jiyu; Huang, Shuai; Xia, Guangqing; Song, Jinlong; Xu, Wenji; Sun, Jing; Liu, Xin

    2016-12-01

    Tubes are indispensable in our daily life, mechanical engineering and biomedical fields. However, the practical applications of tubes are sometimes limited by their poor wettability. Reported herein is hydrophilization of the tube inner wall by transferred atmospheric pressure plasma (TAPP). An Ar atmospheric pressure plasma jet (APPJ) is used to induce He TAPP inside polytetrafluoroethylene (PTFE) tube to perform inner wall surface modification. Optical emission spectrum (OES) is used to investigate the distribution of active species, which are known as enablers for surface modification, along the TAPP. Tubes' surface properties demonstrate that after TAPP treatment, the wettability of the tube inner wall is well improved due to the decrease of surface roughness, the removal of surface fluorine and introduction of oxygen. Notably, a deep surface modification can significantly retard the aging of the obtained hydrophilicity. The results presented here clearly demonstrate the great potential of TAPP for surface modification of the inner wall of tube or other hollow bodies, and thus a uniform, effective and long-lasting surface modification of tube with any length can be easily realized by moving the tube along its axis.

  13. Pressure tube replication techniques using the advanced NDE system

    Energy Technology Data Exchange (ETDEWEB)

    Isherwood, A.; Jarron, D.; Travers, J.; Hanley, K. [Ontario Power Generation, Pickering, Ontario (Canada)]. E-mail: andrew.isherwood@opg.com

    2006-07-01

    and rotary position of the flaw, as determined by ultrasonics, is entered and the tool is positioned in the channel and locked in place. A twenty minute auto-sequence is then initiated which injects material into the flaw and allows it to cure. Once complete, the delivery machine comes off channel and the tool is pushed out of the machine onto a loading trough. The operators remove the replica and recharge the tool for the next set of indications. Once the removed replica has been examined by trained technicians, it is moved to an on-site laser profilometry device to convert the positive of the flaw to a 3D computer image. This image is then curve fitted to determine the smallest radius of the flaw. The combination of in-vault recharging and a two plate tool has reduced critical path times for this phase of the outages. The ANDE Replication System has achieved rates of 10 replicas in just 42 hours. The ANDE Replication System has been successfully used at four different CANDU stations on over 30 flaws. It has been used with both the UDM and the ADM (Advanced Delivery Machine). This paper will present details of the new tool, control system, and field execution of the ANDE Replication System at Ontario Power Generation reactors. (author)

  14. Automated Control of Endotracheal Tube Cuff Pressure during Simulated Flight

    Science.gov (United States)

    2016-06-21

    controllers—Intellicuff, Hamilton Medical; Pyton, ARM Medical; Cuff Sentry, Outcome Solutions—were used to manage cuff pressures . The fourth group had...Medical; Pyton, ARM Medical; Cuff Sentry, Outcome Solutions—were used to manage cuff pressures . The fourth group had cuff pressure set at sea level...ventilator, while the other two devices are stand-alone products. The fourth group of ETTs had the cuff pressure measured by the respiratory

  15. Results of experimental tests simulating supply pressure decrease in a K process tube

    Energy Technology Data Exchange (ETDEWEB)

    Toyoda, K.G.; Calkin, J.F.

    1957-11-13

    Simultaneous reduction of coolant to several or all reactor tubes raises concern not only for the adequacy of protection in the individual process tube but also the reactor as a whole. In event of such flow reduction, the heat generation does not decrease until at least 1.4 seconds have elapsed following the accident. Thus, the water temperature from each tube will rise, and result in an increase in the bulk water temperature. If the increase in bulk water temperature is such that saturation temperature at the top of downcomer is reached, pressurization may occur at that point and exceed the maximum recommended working pressure limit (approximately 1 to 2 psig). The purpose of this report is to present experimental data on a series of tests which were made to simulate flow reductions to a K type process tube by simulated front header pressure decreases.

  16. Isothermal flow measurement using planar PIV in the 1/4 scaled model of CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Im, Sunghyuk; Sung, Hyung Jin [KAIST, Daejeon (Korea, Republic of); Seo, Han; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the flow field and local temperature distribution in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV) and Laser Induced Fluorescence (LIF) techniques. Small-scale 1/40 and 1/8 small-scale model tests were performed prior to installation of the main MCT facility to identify the potential problems of the flow visualization and measurement expected in the 1/4 scale MCT facility. In the 1/40 scale test, a flow field was measured with a PIV measurement technique under an iso-thermal state, and the temperature field was visualized using a LIF technique. In this experiment, the key point was to illuminate the region of interest as uniformly as possible since the velocity and temperature fields in the shadow regions were distorted and unphysical. In the 1/8 scale test, the flow patterns from the inlet nozzles to the top region of the tank were investigated using PIV measurement at two different positions of the inlet nozzle. For each position of laser beam exposure the measurement sections were divided to 7 groups to overcome the limitation of the laser power to cover the relatively large test section. The MCT facility is the large-scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. It is reduced in a 1/4 scale and a moderator test vessel is built to the specifications of the CANDU6 reactor design, where a working fluid is sub-cooled water with atmospheric pressure. Previous studies were

  17. Dimensional Measurements of Fresh CANDU Fuel Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Jo, Chang Keun; Jung, Jong Yeob; Koo, Dae Seo; Cho, Moon Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    This paper intends to provide the dimensional measurements of fresh CANDU fuel (37-element) bundle for the estimation of deformation of post-irradiated (PI) bundle. It is expensive and difficult to measure the fretting wear of bearing pad, the element bowing and the waviness of endplate at the two-phase high flow condition (above 24 kg/s) of out-of-reactor test. So, it is recommended to compare the geometry of fresh bundle with that of PI bundle to estimate the integrity of fuel bundle in the CANDU-6 fuel channel with two-phase flow condition. The measurement system has been developed to provide the visual inspection and the dimensional measurements within the accuracy of 10 {mu}m. It is applicable in-air and underwater to the CANDU bundle as well as the CANFLEX bundle. The in-air measurements of the 36 fresh CANDU bundles (S/N: B400892 {approx} B400927) are done by this system from February 2004 to March 2004 in the PHWR fresh fuel storage building of KNFC. These bundles are produced by KNFC manufacturing procedure and are waiting for the delivery to the Wolsong-3 plant, and are planned to load into the proposed test channels. The detail measurements contain the outer rod profile (including the bearing pad), the diameter of bundle, the bowing of bundle, the rod length and the surface profile of end plate (waviness)

  18. Effects of diameter, length, and circuit pressure on sound conductance through endotracheal tubes.

    Science.gov (United States)

    Räsänen, Jukka O; Rosenhouse, Giora; Gavriely, Noam

    2006-07-01

    We evaluated the acoustic frequency response of endotracheal tubes (ETs) to assess their effect on respiratory system sound transmission studies. White noise 150-3300 Hz was introduced into 4.0-, 6.0-, and 8.0-mm ETs and recorded at their proximal and distal ends. Four tubes of each size were studied at their original and normalized lengths, in straight and bent configurations, and at circuit pressures from 0 to 20 cmH2O. The characteristics of the sound transmission were compared using an analysis of variance for repeated measures. The average transmission amplitude varied directly with tube diameter. The position of peaks and troughs on the amplitude frequency distribution depended on tube length but not on tube diameter. The angle of the phase-frequency plot correlated well with the length of the tube and was independent of its diameter. A 90 degrees bend in the tube had no effect on its sound transmission. Increasing the circuit pressure above ambient modified the frequency response only if volume changes occurred in the test lung. When used to conduct sound into the respiratory system an ET affects the incident signal predictably depending on its length and diameter but not on its curvature or circuit pressure.

  19. 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%.

  20. A feasibility study on the use of the MOOSE computational framework to simulate three-dimensional deformation of CANDU reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle A., E-mail: Kyle.Gamble@inl.gov [Royal Military College of Canada, Chemistry and Chemical Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada); Williams, Anthony F., E-mail: Tony.Williams@cnl.ca [Canadian Nuclear Laboratories, Fuel and Fuel Channel Safety, 1 Plant Road, Chalk River, Ontario, Canada K0J 1J0 (Canada); Chan, Paul K., E-mail: Paul.Chan@rmc.ca [Royal Military College of Canada, Chemistry and Chemical Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada); Wowk, Diane, E-mail: Diane.Wowk@rmc.ca [Royal Military College of Canada, Mechanical and Aerospace Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada)

    2015-11-15

    Highlights: • This is the first demonstration of using the MOOSE framework for modeling CANDU fuel. • Glued and frictionless contact algorithms behave as expected for 2D and 3D cases. • MOOSE accepts and correctly interprets functions of arbitrary form. • 3D deformation calculations accurately compare against analytical solutions. • MOOSE is a viable simulation tool for modeling accident reactor conditions. - Abstract: Horizontally oriented fuel bundles, such as those in CANada Deuterium Uranium (CANDU) reactors present unique modeling challenges. After long irradiation times or during severe transients the fuel elements can laterally deform out of plane due to processes known as bow and sag. Bowing is a thermally driven process that causes the fuel elements to laterally deform when a temperature gradient develops across the diameter of the element. Sagging is a coupled mechanical and thermal process caused by deformation of the fuel pin due to creep mechanisms of the sheathing after long irradiation times and or high temperatures. These out-of-plane deformations can lead to reduced coolant flow and a reduction in coolability of the fuel bundle. In extreme cases element-to-element or element-to-pressure tube contact could occur leading to reduced coolant flow in the subchannels or pressure tube rupture leading to a loss of coolant accident. This paper evaluates the capability of the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed at the Idaho National Laboratory to model these deformation mechanisms. The material model capabilities of MOOSE and its ability to simulate contact are also investigated.

  1. 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%.

  2. CFD simulations of the single-phase and two-phase coolant flow of water inside the original and modified CANDU 37-element bundles

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Hadaller, G.I.; Fortman, R.A. [Stern Laboratories, Hamilton, Ontario (Canada)

    2010-07-01

    Single-phase (inlet temperature of 180° C, outlet pressure of 9 MPa, total power of 2 MW and flow rate of 13.5 Kg/s), and two-phase (inlet temperature of 265° C, outlet pressure of 10 MPa, total power of 7.126 MW and flow rate of 19 Kg/s) water flows inside a CANDU thirty seven element fuel string are simulated using a Computational Fluid Dynamics (CFD) code with parallel processing and results are presented in this paper. The analyses have been performed for the original and modified (11.5 mm center element diameter) designs with skewed cosine axial heat flux distribution and 5.1% diametral creep of the pressure tube. The CFD results are in good agreement with the expected temperature and velocity cross-sectional distributions. The effect of the pressure tube creep on the flow bypass is detected, and the CHF improvement in the core region of the modified design is confirmed. The two-phase flow model reasonably predicted the void distribution and the role of interfacial drag on increasing the pressure drop. In all CFD models, the appendages were shown to enhance the production of cross flows and their corresponding flow mixing and asymmetry. (author)

  3. Studies on an improved indigenous pressure wave generator and its testing with a pulse tube cooler

    Science.gov (United States)

    Jacob, S.; Karunanithi, R.; Narsimham, G. S. V. L.; Kranthi, J. Kumar; Damu, C.; Praveen, T.; Samir, M.; Mallappa, A.

    2014-01-01

    Earlier version of an indigenously developed Pressure Wave Generator (PWG) could not develop the necessary pressure ratio to satisfactorily operate a pulse tube cooler, largely due to high blow by losses in the piston cylinder seal gap and due to a few design deficiencies. Effect of different parameters like seal gap, piston diameter, piston stroke, moving mass and the piston back volume on the performance is studied analytically. Modifications were done to the PWG based on analysis and the performance is experimentally measured. A significant improvement in PWG performance is seen as a result of the modifications. The improved PWG is tested with the same pulse tube cooler but with different inertance tube configurations. A no load temperature of 130 K is achieved with an inertance tube configuration designed using Sage software. The delivered PV power is estimated to be 28.4 W which can produce a refrigeration of about 1 W at 80 K.

  4. Experimental Investigation of Heat Transfer and Pressure Drop Characteristics of H-type Finned Tube Banks

    OpenAIRE

    2014-01-01

    H-type finned tube heat exchanger elements maintain a high capacity for heat transfer, possess superior self-cleaning properties and retain the ability to effect flue gas waste heat recovery in boiler renovations. In this paper, the heat transfer and pressure drop characteristics of H-type finned tube banks are studied via an experimental open high-temperature wind tunnel system. The effects of fin width, fin height, fin pitch and air velocity on fin efficiency, convective heat transfer coe...

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

    Directory of Open Access Journals (Sweden)

    Møller Ann M

    2010-12-01

    Full Text Available 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 and age, body mass index, type of surgery, or time from induction of anaesthesia to the time the cuff pressure was measured. Conclusion For maintenance of epithelia flow and nerve function and at the same time secure air sealing, this evaluation indicates that the cuff pressure needs to be checked as part of the procedures involved in induction of anaesthesia and eventually checked during surgery.

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

    Science.gov (United States)

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

    2015-04-01

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

  7. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 9 - CUTTING AND EXTRACTING DEVICE FUNCTIONING

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2015-05-01

    Full Text Available This paper presents a constructive solution proposed by the authors in order to achieve of a cutting and extracting device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components. It's a flexible and modular device, which is designed to work inside the fuel channel and has the following functions: moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. The Cutting and Extraction Device (CED consists of following modules: guiding-fixing module, traction modules, cutting module, guiding-extracting module and flexible elements for modules connecting. The guiding-fixing module is equipped with elastic guiding rollers and fixing claws in working position, the traction modules are provided with variable pitch rollers for allowing variable travel speed through the fuel channel. The cutting module is positioned in the middle of the device and it is equipped with three knife rolls for pressure tube cutting, using a system for cutting place video surveillance and pyrometers for monitoring cutting place temperature. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  8. Validation of WIMS-CANDU using Pin-Cell Lattices

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Young; Min, Byung Joo; Park, Joo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The WIMS-CANDU is a lattice code which has a depletion capability for the analysis of reactor physics problems related to a design and safety. The WIMS-CANDU code has been developed from the WIMSD5B, a version of the WIMS code released from the OECD/NEA data bank in 1998. The lattice code POWDERPUFS-V (PPV) has been used for the physics design and analysis of a natural uranium fuel for the CANDU reactor. However since the application of PPV is limited to a fresh fuel due to its empirical correlations, the WIMS-AECL code has been developed by AECL to substitute the PPV. Also, the WIMS-CANDU code is being developed to perform the physics analysis of the present operating CANDU reactors as a replacement of PPV. As one of the developing work of WIMS-CANDU, the U{sup 238} absorption cross-section in the nuclear data library of WIMS-CANDU was updated and WIMS-CANDU was validated using the benchmark problems for pin-cell lattices such as TRX-1, TRX-2, Bapl-1, Bapl-2 and Bapl-3. The results by the WIMS-CANDU and the WIMS-AECL were compared with the experimental data.

  9. Comparative evaluation of intraocular pressure changes subsequent to insertion of laryngeal mask airway and endotracheal tube.

    Directory of Open Access Journals (Sweden)

    Ghai B

    2001-07-01

    Full Text Available AIMS: To evaluate the intraocular pressure and haemodynamic changes subsequent to insertion of laryngeal mask airway and endotracheal tube. SUBJECTS AND METHODS: The study was conducted in 50 adult patients. A standard general anaesthesia was administered to all the patients. After 3 minutes of induction of anaesthesia baseline measurements of heart rate, non-invasive blood pressure and intraocular pressure were taken following which patients were divided into two groups: laryngeal mask airway was inserted in group 1 and tracheal tube in group 2. These measurements were repeated at 15-30 second, every minute thereafter up to 5 minutes after airway instrumentation. RESULTS: A statistically significant rise in heart rate, systolic blood pressure, diastolic blood pressure and intraocular pressure was seen in both the groups subsequent to insertion of laryngeal mask airway or endotracheal tube. Mean maximum increase was statistically more after endotracheal intubation than after laryngeal mask airway insertion. The duration of statistically significant pressure responses was also longer after endotracheal intubation. CONCLUSION: Laryngeal mask airway is an acceptable alternative technique for ocular surgeries, offering advantages in terms of intraocular pressure and cardiovascular stability compared to tracheal intubation.

  10. 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

    HYPOTHESIS:: Middle ear pressure (MEP) is actively regulated by both the Eustachian tube and the mastoid air cell system. BACKGROUND:: MEP is a highly significant factor involved in many clinical conditions related to otitis media. Basic knowledge on its overall regulation remains insufficient...... of these distinct mechanisms were found. CONCLUSION:: The human mastoid as well as the Eustachian tube was capable of active counter-regulation of the MEP in short-term experimental pressure changes in healthy ears. Thus, these 2 systems seemed to function in a complementary way, where the mastoid was related...

  11. 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.

  12. 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...

  13. Low-frequency pressure wave propagation in liquid-filled, flexible tubes. (A)

    DEFF Research Database (Denmark)

    Bjørnø, Leif; Bjelland, C.

    1992-01-01

    A model has been developed for propagation of low-frequency pressure waves in viscoelastic tubes with distensibility of greater importance than compressibility of the liquid. The dispersion and attenuation are shown to be strongly dependent on the viscoelastic properties of the tube wall....... The complex, frequency-dependent moduli of relevant tube materials have been measured in a series of experiments using three different experimental procedures, and the data obtained are compared. The three procedures were: (1) ultrasonic wave propagation, (2) transversal resonance in bar samples, and (3......) moduli determined by stress wave transfer function measurements in simple extension experiments. The moduli are used in the model to produce realistic dispersion relations and frequency dependent attenuation. Signal transfer functions between positions in the liquid-filled tube can be synthesized from...

  14. Dynamic neck development in a polymer tube under internal pressure loading

    DEFF Research Database (Denmark)

    Lindgreen, Britta; Tvergaard, Viggo; Needleman, Alan

    2008-01-01

    and a short wave length imperfection. After some thinning down at the necks, the mode of deformation switches to neck propagation along the circumference of the tube. A case is shown in which the necks have propagated along the entire tube wall, so that network locking in the polymer results in high stiffness......The initiation and growth of necks in polymer tubes subjected to rapidly increasing internal pressure is analyzed numerically. Plane strain conditions are assumed to prevail in the axial direction. The polymer is characterized by a finite strain elastic-viscoplastic constitutive relation...... and the calculations are carried out using a dynamic finite element program. Numerical results for neck development are illustrated and discussed for tubes of various thicknesses. The sensitivity to the wave number of the thickness imperfections is studied with a focus on comparing a long wave length imperfection...

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

    Energy Technology Data Exchange (ETDEWEB)

    Da Silva Lima, J. R.

    2011-07-01

    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{sup -1} m{sup -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

  16. Experimental studies on pressure drop characteristics of cryogenic cross-counter flow coiled finned tube heat exchangers

    Science.gov (United States)

    Gupta, Prabhat Kumar; Kush, P. K.; Tiwari, Ashesh

    2010-04-01

    Cross-counter flow coiled finned tube heat exchangers used in medium capacity helium liquefiers/refrigerators were developed in our lab. These heat exchangers were developed using integrated low finned tubes. Experimental studies have been performed to know the pressure drop characteristics of tube side and shell side flow of these heat exchangers. All experiments were performed at room temperature in the Reynolds number range of 3000-30,000 for tube side and 25-155 for shell side. The results of present experiments indicate that available correlations for tube side can not be used for prediction of tube side pressure drop data due to complex surface formation at inner side of tube during formation of fins over the outer surface. Results also indicate that surface roughness effect becomes more pronounced as the value of di/ D m increases. New correlations based on present experimental data are proposed for predicting the friction factors for tube side and shell side.

  17. A numerical assessment of the load bearing capacity of externally pressurized moderately thick tubes

    Energy Technology Data Exchange (ETDEWEB)

    Corradi, Leone [Politecnico di Milano - Department of Energy, Enrico Fermi Center for Nuclear Studies (CeSNEF), via Ponzio 34/3 - 20133 Milano (Italy)], E-mail: leone.corradi@polimi.it; Di Marcello, Valentino; Luzzi, Lelio; Trudi, Fulvio [Politecnico di Milano - Department of Energy, Enrico Fermi Center for Nuclear Studies (CeSNEF), via Ponzio 34/3 - 20133 Milano (Italy)

    2009-08-15

    The collapse behavior of cylindrical shells pressurized from outside is examined. Attention is focused on tubes of moderate thickness, as required by very deep water pipelines or some innovative nuclear power plant proposals. Their collapse is expected to be dominated by yielding but, because of the decreasing nature of the post-collapse evolution, interaction with instability is likely to be significant enough to demand consideration. At present, no quantitative assessment of such effect is available, because little study has been devoted to tubes in this thickness range. Plasticity-instability interaction is activated by imperfections and to assess their influence on a systematic numerical study is undertaken. Computations produce a meaningful measure of the collapse pressure and it is proposed that the allowable pressure be determined on its basis, by introducing a suitable safety factor. This is chosen so that results reproduce those provided by presently accepted procedures in the well explored and reliable range of medium-thin tubes. When the same factor is applied to thicker tubes, the resulting allowable pressure is significantly higher than the values suggested by codes, which apparently react to the present lack of knowledge by assuming an extremely conservative attitude.

  18. Viscous Inner and Outer Pressure Forming Method of Thin-walled Tube and Its Application

    Institute of Scientific and Technical Information of China (English)

    GAO Tiejun; LIU Yang; WANG Zhongjin

    2015-01-01

    Aiming at overcoming the difficulties in integral forming of thin-walled tubes with complex shapes, a novel forming method by inner and outer pressure through viscous was proposed. In this method, by dividing large deformation of the part into inner and outer pressure forming deformations, the limit deformation of tube part can be increased by several times. Meanwhile, the principle of viscous inner and outer pressure forming was provided, and key problems during the forming process such as reduction of the wall-thickness and instability wrinkling were analyzed. Thereby, the complex curved surface super-alloy GH3044 thin-walled tube with varying diameter ratio of 1.35 (the ratio between the maximum and minimum diameters of the part) can be integrally formed by this method. The experimental surface of the formed part is superior in quality and the wall-thickness distribution is uniform. The results show that the viscous inner and outer pressure forming can provide a new approach for integral forming of thin-walled tubes with complex shapes.

  19. Measurements of endotracheal tube cuff contact pressure using fibre Bragg gratings

    Science.gov (United States)

    Hernandez, F. U.; Correia, R.; Korposh, S.; Morgan, S. P.; Hayes-Gill, B. R.; James, S. W.; Evans, D.; Norris, A.

    2015-09-01

    An optical fibre Bragg grating (FBG) was used to measure local strain (due to contact pressure) at the interface of a cuffed endotracheal tube (ETT) tested in a tracheal model. The tracheal model consisted of a corrugated tube. Two FBG sensors written in a single optical fibre were attached to the outside wall of the cuff of the ETT. Intracuff endotracheal pressure was measured using a digital manometer, while the contact pressure between the model trachea and the ETT was measured using Flexiforce sensors. Changes in the Bragg wavelengths in response to the inflation of the cuff of the ETT, and concomitant pressure increase, were observed to be dependent on the location of the FBGs at the corrugations, i.e., the annular peaks and troughs of the corrugated tube. The performance of both contact pressure sensors FBG and Flexiforce suggests that FBG technology is better suited to this application as it allows the measurement of contact pressures on non-uniform surfaces such as in the tracheal model.

  20. Two-Phase Critical Discharge of Initially Saturated or Subcooled Water Flowing in Sharp-Edgred Tubes at High Pressure

    Institute of Scientific and Technical Information of China (English)

    1995-01-01

    The transient critical flow experiment with sharp-deged tubes as the break geometries is conducted in high pressure convective circulation test loop of Xi'an Jiantong University.The initial Steady operation pressure is up to 22.0MPa.An empirical correlation was made to obtain the critical mass flow rates,the critical pressure ratio and the thermal nonequilibrium number were correlated as the functions of the tube length to tube diameter ratio L/D.The predicted critical mass flow rate gets a higher accureacy for short tubes with L/D 12.

  1. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 10 - PRESENTATION OF THE DECOMMISSIONING DEVICE OPERATING

    Directory of Open Access Journals (Sweden)

    Constantin D. STANESCU,

    2015-05-01

    Full Text Available This paper presents a solution proposed by the authors in order to achieve of a cutting and extracting device operating panel for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components, moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. All operations can be monitored and controlled from a operating panel. The PLC fully command the device in automatic or manually mode, to control the internal sensors, transducers, electrical motors, video surveillance and pyrometers for monitoring cutting place temperature. The device controller has direct access to the measured values with these sensors, interprets and processes them, preparing the next actionafter confirming the action in progress. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  2. Experience on management of CANDU spent fuel in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H.-Y.; Choi, B.-I.; Yoon, J.-H.; Seo, U.-S. [Korea Hydro and Nuclear Power Co. Ltd., Nuclear Environment Technology Inst. (KHNP/NETEC), Yusung-Gu, Daejeon (Korea, Republic of)

    2002-07-01

    In Korea, national policy on the management of spent fuel from both PWR and CANDU reactors demands that all the spent fuel be kept within reactor site in until 2016 the time spent fuel interim storage facility might open. Based on the end of 2001, KHNP has 4 CANDU reactors in operation generating approximately 5,000 bundles of spent fuels per each unit annually. The generation, accumulation, and management of CANDU spent fuel by KHNP in Korea are reviewed. CANDU spent fuel storage technology including pool storage in fuel building, concrete silo storage, and on going project for consolidating storage adapting modular vault type MACSTOR concept are outlined. Especially current joint development of storage of CANDU spent fuel for improving land usage is addressed. The explanation of the new consolidated dry storage system includes description of the storage facility, its safety evaluations, and final implementation. Finally future movement on management of spent fuel in Korea is also briefly introduced. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

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

  4. Prediction of Burst Pressure in Multistage Tube Hydroforming of Aerospace Alloys.

    Science.gov (United States)

    Saboori, M; Gholipour, J; Champliaud, H; Wanjara, P; Gakwaya, A; Savoie, J

    2016-08-01

    Bursting, an irreversible failure in tube hydroforming (THF), results mainly from the local plastic instabilities that occur when the biaxial stresses imparted during the process exceed the forming limit strains of the material. To predict the burst pressure, Oyan's and Brozzo's decoupled ductile fracture criteria (DFC) were implemented as user material models in a dynamic nonlinear commercial 3D finite-element (FE) software, ls-dyna. THF of a round to V-shape was selected as a generic representative of an aerospace component for the FE simulations and experimental trials. To validate the simulation results, THF experiments up to bursting were carried out using Inconel 718 (IN 718) tubes with a thickness of 0.9 mm to measure the internal pressures during the process. When comparing the experimental and simulation results, the burst pressure predicated based on Oyane's decoupled damage criterion was found to agree better with the measured data for IN 718 than Brozzo's fracture criterion.

  5. Wrinkling behavior in tube hydroforming coupled with internal and external pressure

    Directory of Open Access Journals (Sweden)

    Cui X.L.

    2015-01-01

    Full Text Available Control and use of wrinkles is a challenge in tube hydroforming because wrinkle was always considered as one of the defects of tubes from the traditional view. In this paper, a dedicated experimental setup was designed and manufactured, using which the investigation of wrinkling behavior coupled with internal and external pressure can be realized. The effect of internal or external pressure on 5A02 aluminum alloy tubes and 0Cr18Ni9 stainless steel tubes were all investigated using this setup. It was found that the number and shape of wrinkles are strongly dependent on the internal or external pressure. More important is that the geometrical configuration of wrinkles can be perfectly characterized using the Gauss function rather than the sine function adopted in the published literature. In addition, the fitted Gauss functions for every wrinkle were integrated in order to compare their area with the corresponding area of die cavity, so as to obtain the appropriate process parameters for the useful wrinkle, which can be formed in advance and then flatted in the calibration stage.

  6. Spatial resolution of thin-walled high-pressure drift tubes

    CERN Document Server

    Davkov, V I; Tikhomirov, V O; Smirnov, S Y; Gregor, I; Senger, P; Naumann, L; Myalkovskiy, V V; Mouraviev, S V; Peshekhonov, V D; Russakovich, N A; Rufanov, I A; Rembser, C

    2011-01-01

    A small prototype detector based on high pressure thin-walled tubes (straws) has been developed and its parameters have been studied on a bench at JINR, Dubna, and SPS at CERN. The inner diameter of the straws is 9.53 mm. The pressure of the active gas mixture Ar/CO(2) (80/20) was varied from 1 to 5 bar. The best spatial resolution achieved in this pressure range is similar to 40 mu m. Both the high efficiency and high rate capability are retained. (C) 2011 Published by Elsevier B.V.

  7. 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...... and age, body mass index, type of surgery, or time from induction of anaesthesia to the time the cuff pressure was measured. CONCLUSION: For maintenance of epithelia flow and nerve function and at the same time secure air sealing, this evaluation indicates that the cuff pressure needs to be checked...

  8. Variation of pressure limits of flame propagation with tube diameter for various isooctane-oxygen-nitrogen mixtures

    Science.gov (United States)

    Spakowski, Adolph, A; Belles, Frank E

    1952-01-01

    An investigation was made of the change in the pressure limits of flame propagation with tube diameter for various isooctane-oxygen-nitrogen mixtures. Pressure limits were measured in cylindrical glass tubes of four different inside diameters at six different oxygen-nitrogen ratios. Under the experimental conditions, flame propagation was found to be impossible in isooctane-oxygen mixtures with oxygen concentrations less than 11 to 12 percent. Critical tube diameters for flame propagation were calculated and the effect of pressure was determined and compared with the effect of pressure on quenching distance. Critical diameters were related to flame speeds for various isooctane-oxygen-nitrogen mixtures.

  9. Temperature, Pressure and Velocity measurements on the Ranque-Hilsch Vortex Tube

    Science.gov (United States)

    Liew, R.; Zeegers, J. C. H.; Kuerten, J. G. M.; Michałek, W. R.

    2012-11-01

    Temperatures, pressures and velocities were measured in a Ranque-Hilsch vortex tube. Results show that the cooling power is larger than the heating power due to a heat loss to the surroundings. This heat loss became the more dominant thermodynamic process at large cold fractions (the ratio of cold mass flow over total mass flow). The velocities were obtained by means of Laser Doppler Anemometry. By this method, the three dimensional velocities of the gas and their standard deviations in the vortex tube are revealed by an non-intrusive measurement method. The turbulent fluctuations, characterized by the standard deviations, show that the turbulence is isotropic in the core region of the vortex tube.

  10. 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

  11. Pressure wave propagation in fluid-filled co-axial elastic tubes. Part 1: Basic theory.

    Science.gov (United States)

    Berkouk, K; Carpenter, P W; Lucey, A D

    2003-12-01

    Our work is motivated by ideas about the pathogenesis of syringomyelia. This is a serious disease characterized by the appearance of longitudinal cavities within the spinal cord. Its causes are unknown, but pressure propagation is probably implicated. We have developed an inviscid theory for the propagation of pressure waves in co-axial, fluid-filled, elastic tubes. This is intended as a simple model of the intraspinal cerebrospinal-fluid system. Our approach is based on the classic theory for the propagation of longitudinal waves in single, fluid-filled, elastic tubes. We show that for small-amplitude waves the governing equations reduce to the classic wave equation. The wave speed is found to be a strong function of the ratio of the tubes' cross-sectional areas. It is found that the leading edge of a transmural pressure pulse tends to generate compressive waves with converging wave fronts. Consequently, the leading edge of the pressure pulse steepens to form a shock-like elastic jump. A weakly nonlinear theory is developed for such an elastic jump.

  12. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  13. Development of the advanced CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Na, Y. H.; Lee, S. Y.; Choi, J. H.; Lee, B. C.; Kim, S. N.; Jo, C. H.; Paik, J. S.; On, M. R.; Park, H. S.; Kim, S. R. [Korea Electric Power Co., Taejon (Korea, Republic of)

    1997-07-01

    The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs.

  14. Shock tube study of n-decane ignition at low pressures

    Institute of Scientific and Technical Information of China (English)

    Xiao-Fei Nie; Ping Li; Chang-Hua Zhang; Wei Xie; Cong-Shan Li; Xiang-Yuan Li

    2012-01-01

    Ignition delay times for n-decane/O2/Ar mixtures were measured behind reflected shock waves using endwall pressure and CH* emission measurements in a heated shock tube.The initial postshock conditions cover pressures of 0.09-0.26 MPa,temperatures of 1 227-1 536 K,and oxygen mole fractions of 3.9%-20.7% with an equivalence ratio of 1.0.The correlation formula of ignition delay dependence on pressure,temperature,and oxygen mole fraction was obtained.The current data are in good agreement with available low-pressure experimental data,and they are then compared with the prediction of a kinetic mechanism.The current measurements extend the kinetic modeling targets for the n-decane combustion at low pressures.

  15. Reconstruction of an acoustic pressure field in a resonance tube by particle image velocimetry.

    Science.gov (United States)

    Kuzuu, K; Hasegawa, S

    2015-11-01

    A technique for estimating an acoustic field in a resonance tube is suggested. The estimation of an acoustic field in a resonance tube is important for the development of the thermoacoustic engine, and can be conducted employing two sensors to measure pressure. While this measurement technique is known as the two-sensor method, care needs to be taken with the location of pressure sensors when conducting pressure measurements. In the present study, particle image velocimetry (PIV) is employed instead of a pressure measurement by a sensor, and two-dimensional velocity vector images are extracted as sequential data from only a one- time recording made by a video camera of PIV. The spatial velocity amplitude is obtained from those images, and a pressure distribution is calculated from velocity amplitudes at two points by extending the equations derived for the two-sensor method. By means of this method, problems relating to the locations and calibrations of multiple pressure sensors are avoided. Furthermore, to verify the accuracy of the present method, the experiments are conducted employing the conventional two-sensor method and laser Doppler velocimetry (LDV). Then, results by the proposed method are compared with those obtained with the two-sensor method and LDV.

  16. Implementation of an on-line monitoring system for transmitters in a CANDU nuclear power plant

    Science.gov (United States)

    Labbe, A.; Abdul-Nour, G.; Vaillancourt, R.; Komljenovic, D.

    2012-05-01

    Many transmitters (pressure, level and flow) are used in a nuclear power plant. It is necessary to calibrate them periodically to ensure that their measurements are accurate. These calibration tasks are time consuming and often contribute to worker radiation exposure. Human errors can also sometimes degrade their performance since the calibration involves intrusive techniques. More importantly, experience has shown that the majority of current calibration efforts are not necessary. These facts motivated the nuclear industry to develop new technologies for identifying drifting instruments. These technologies, well known as on-line monitoring (OLM) techniques, are non-intrusive and allow focusing the maintenance efforts on the instruments that really need a calibration. Although few OLM systems have been implemented in some PWR and BWR plants, these technologies are not commonly used and have not been permanently implemented in a CANDU plant. This paper presents the results of a research project that has been performed in a CANDU plant in order to validate the implementation of an OLM system. An application project, based on the ICMP algorithm developed by EPRI, has been carried out in order to evaluate the performance of an OLM system. The results demonstrated that the OLM system was able to detect the drift of an instrument in the majority of the studied cases. A feasibility study has also been completed and has demonstrated that the implementation of an OLM system at a CANDU nuclear power plant could be advantageous under certain conditions.

  17. Backup and Ultimate Heat Sinks in CANDU Reactors For Prolonged SBO Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Brown, M. J. [Atomic Energy of Canada Limited, Ontario (Canada)

    2013-10-15

    In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ∼2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

  18. Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU`s. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work.

  19. Experimental Study on Heat Transfer and Pressure Drop of Micro-Sized Tube Heat Exchanger

    Institute of Scientific and Technical Information of China (English)

    王秋香; 戴传山

    2014-01-01

    A micro-sized tube heat exchanger (MTHE) was fabricated, and its performance in heat transfer and pres-sure drop was experimentally studied. The single-phase forced convection heat transfer correlation on the sides of the MTHE tubes was proposed and compared with previous experimental data in the Reynolds number range of 500-1 800. The average deviation of the correlation in calculating the Nusselt number was about 6.59%. The entrance effect in the thermal entrance region was discussed. In the same range of Reynolds number, the pressure drop and friction coefficient were found to be considerably higher than those predicted by the conventional correlations. The product of friction factor and Reynolds number was also a constant, but much higher than the conventional.

  20. Stress and integrity analysis of steam superheater tubes of a high pressure boiler

    Directory of Open Access Journals (Sweden)

    Neves Daniel Leite Cypriano

    2004-01-01

    Full Text Available Sources that can lead to deterioration of steam superheater tubes of a high pressure boiler were studied by a stress analysis, focused on internal pressure and temperature experienced by the material at real operating conditions. Loss of flame control, internal deposits and unexpected peak charge are factors that generate loads above the design limit of tube materials, which can be subjected to strain, buckling, cracks and finally rupture in service. To evaluate integrity and dependability of these components, the microstructure of selected samples along the superheater was studied by optical microscopy. Associated with this analysis, dimensional inspection, nondestructive testing, hardness measurement and deposit examination were made to determine the resultant material condition after twenty three years of operation.

  1. Remote field eddy current technique for gap measurement of horizontal flux detector guide tube in pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Hoon; Jung, Hyun Kyu; Yang, Dong Ju; Cheong, Yong Moo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2004-11-15

    The fuel channels including the pressure tube(PT) and the calandria tube(CT) are important components of the pressurized heavy water reactor(PHWR). A sagging of fuel channel increases by heat and radiation exposure with the increasing operation time. The contact of fuel channel to the Horizontal flux Detector(HFD) guide tube is needed for the power plant safety. In order to solve this safety issue, the electromagnetic technique was applied to measure the status of the guide tube. The Horizontal flux Detector(HFD) guide tube and the Calandria tube(CT) in the Pressurized Heavy Water Reactor(PHWR) are cross-aligned horizontally. The remote field eddy current(RFEC) technology is applied for gap measurement between the HFD guide tube and the CT HFD guide tube can be detected by inserting the RFEC probe into pressure tube(PT) at the crossing point directly. The RFEC signals using the volume integral method(VIM) were simulated for obtaining the optimal inspection parameters. This paper shows that the simulated eddy current signals and the experimental results in variance with the CT/HFD gap.

  2. The travesty of discarding used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ottensmeyer, P. [Univ. of Toronto, Toronto, Ontario (Canada)

    2016-09-15

    The current plan worldwide for virtually all used nuclear fuels is costly deep burial to attempt to isolate their long-term radiotoxicity permanently. Alternatively Canada's 50,000 tons spent CANDU fuel, of which only 0.74% of the heavy atoms have been fissioned to extract their energy, could supply 130 times more non-carbon energy using proven economical recycling and fast-neutron technologies. The result in this country alone would currently be the creation of $74 trillion of reliable electricity on demand without greenhouse gas emissions. It would avoid adding 475 billion tons CO{sub 2} to the atmosphere compared to the use of coal, to mitigate climate change. Worldwide recycling of stored spent nuclear fuel and replenishing with depleted uranium in fast-neutron reactors could avoid emitting over 20 trillion tons CO{sub 2}, or over six times the current total atmospheric CO{sub 2} content. As added bonus the long-term radiotoxicity of the used CANDU fuel is effectively eliminated, making a long-term deep geological repository unnecessary. Even the shorter-lived radioisotope fission products become valuable stable atoms and minerals that would fetch $3 million per ton. Such an alternative is certainly worth pursuing. (author)

  3. Pressure losses during steam flow and condensation in tubes and channels

    Science.gov (United States)

    Leontiev, A. I.; Milman, O. O.

    2014-12-01

    Theoretical and experimental investigations have revealed the dependence of parameters of the process of steam condensation in tubes and channels on the scheme of heat-exchange fluid flow, including counter, forward, and cross flow systems. The total pressure losses in the case of counter flow are greater than those in the case of forward and cross flow. This dependence is valid for the flow of gases and plasma in channels with significant density variation (e.g., due to heating and cooling). Pressure losses have been evaluated using various computational models, and the results are compared to experimental data.

  4. Minimum Safety Factor for Evaluation of Critical Buckling Pressure of Zirconium Alloy Tube

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-15

    We consider the uncertainty in the elastic buckling formula for a thin tube. We take into account the measurement uncertainty of Young's modulus and Poisson's ratio and the tolerance of the tube thickness and diameter. Elastic buckling must be prohibited for a thin tube such as a nuclear fuel rod that must satisfy a self-stand criterion. Since the predicted critical buckling pressure overestimated that found in the experiment, the determination of the minimum safety factor is crucial. The uncertainty in each parameter (i.e., Young's modulus, Poisson's ratio, thickness, and diameter) is mutually independent, so the safety factor is evaluated as the sum of the inverse of each uncertainty. We found that the thickness variation greatly affects the uncertainty. The minimum safety factor of a thin tube of Zirconium alloy is evaluated as 1.547 for a thickness of 0.87 mm and 3.487 for a thickness of 0.254 mm.

  5. Experiments of draining and filling processes in a collapsible tube at high external pressure

    Science.gov (United States)

    Flaud, P.; Guesdon, P.; Fullana, J.-M.

    2012-02-01

    The venous circulation in the lower limb is mainly controlled by the muscular action of the calf. To study the mechanisms governing the venous draining and filling process in such a situation, an experimental setup, composed by a collapsible tube under external pressure, has been built. A valve preventing back flows is inserted at the bottom of the tube and allows to model two different configurations: physiological when the fluid flow is uni-directional and pathological when the fluid flows in both directions. Pressure and flow rate measurements are carried out at the inlet and outlet of the tube and an original optical device with three cameras is proposed to measure the instantaneous cross-sectional area. The experimental results (draining and filling with physiological or pathological valves) are confronted to a simple one-dimensional numerical model which completes the physical interpretation. One major observation is that the muscular contraction induces a fast emptying phase followed by a slow one controlled by viscous effects, and that a defect of the valve decreases, as expected, the ejected volume.

  6. Advanced CANDU reactor pre-licensing progress

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca

    2005-07-01

    The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6 reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The Canadian Nuclear Safety Commission (CNSC) staff are currently reviewing the ACR design to determine whether, in their opinion, there are any fundamental barriers that would prevent the licensing of the design in Canada. This CNSC licensability review will not constitute a licence, but is expected to reduce regulatory risk. The CNSC pre-licensing review started in September 2003, and was focused on identifying topics and issues for ACR-700 that will require a more detailed review. CNSC staff reviewed about 120 reports, and issued to AECL 65 packages of questions and comments. Currently CNSC staff is reviewing AECL responses to all packages of comments. AECL has recently refocused the design efforts to the ACR-1000, which is a larger version of the ACR design. During the remainder of the pre-licensing review, the CNSC review will be focused on the ACR-1000. AECL Technologies Inc. (AECLT), a wholly-owned US subsidiary of AECL, is engaged in a pre-application process for the ACR-700 with the US Nuclear Regulatory Commission (USNRC) to identify and resolve major issues prior to entering a formal process to obtain standard design certification. To date, the USNRC has produced a Pre-Application Safety Assessment Report (PASAR), which contains their reviews of key focus topics. During the remainder of the pre-application phase, AECLT will address the issues identified in the PASAR. Pursuant to the bilateral agreement between AECL and the Chinese nuclear regulator, the National Nuclear Safety Administration (NNSA) and its Nuclear Safety Center (NSC), NNSA/NSC are reviewing the ACR in seven focus areas. The review started in September 2004, and will take three years. The main objective of the review is to determine how the ACR complies

  7. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  8. Structural safety of coolant channel components under excessively high pressure tube diametral expansion rate at garter spring location

    Energy Technology Data Exchange (ETDEWEB)

    Aravind, M. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Sinha, S.K., E-mail: sunilks@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2013-08-15

    Structural safety of coolant channel assembly in the event of high diametral expansion of pressure tube in a 220 MWe pressurised heavy water reactor was investigated using axisymmetric and 3-D finite element models. The axisymmetric analyses were performed and stresses were evaluated for pressure tube, girdle wire and calandria tube at different point of time for diametral expansion rates of 0.2%, 0.25% and 0.3% per year of the pressure tube inside diameter. The results of this study indicated that for the case of 0.3% per year of diametral expansion rate (worst case scenario), occurrence of complete circumferential interference of garter spring with calandria tube at the location of maximum expansion would take place much earlier at around 14 years or 4.2% of the total expansion of pressure tube as opposed to its anticipated design life (30 years). This fact was further corroborated by 3-D finite element analysis performed for the actual assembly configuration under actual loadings. The latter analysis revealed that net section yielding of calandria tube occurs in just 1 year after the occurrence of total circumferential interference between calandria tube and garter spring spacer. It has also been observed that the maximum stress intensity in girdle wire does not increase beyond the ultimate tensile strength even when maximum stress intensity in calandria tube reaches its yield strength. These analyses also revealed that the structural as well as functional integrity of pressure tube and the garter spring is not affected as result of this interference.

  9. Highly sensitive contact pressure measurements using FBG patch in endotracheal tube cuff

    Science.gov (United States)

    Correia, R.; Blackman, O. R.; Hernandez, F. U.; Korposh, S.; Morgan, S. P.; Hayes-Gill, B. R.; James, S. W.; Evans, D.; Norris, A.

    2016-05-01

    A method for measuring the contact pressure between an endotracheal tube cuff and the trachea was designed and developed by using a fibre Bragg grating (FBG) based optical fibre sensor. The FBG sensor is encased in an epoxy based UV-cured cuboid patch and transduces the transversely loaded pressure into an axial strain that induces wavelength shift of the Bragg reflection. The polymer patch was created by using a PTFE based mould and increases tensile strength and sensitivity of the bare fibre FBG to pressure to 2.10×10-2 nm/kPa. The characteristics of the FBG patch allow for continuous measurement of contact pressure. The measurement of contact pressure was demonstrated by the use of a 3D printed model of a human trachea. The influence of temperature on the measurements is reduced significantly by the use of a second FBG sensor patch that is not in contact with the trachea. Intracuff pressure measurements performed using a commercial manometer agreed well with the FBG contact pressure measurements.

  10. Finite element analysis of free expansion of aluminum alloy tube under magnetic pressure

    Institute of Scientific and Technical Information of China (English)

    YU Hai-ping; LI Chun-feng

    2005-01-01

    A link between the electromagnetic code, ANSYS/Emag and the structural code, Ls-dyna was developed, and the numerical modeling of electromagnetic forming for aluminum alloy tube expansion was performed by means of them (discharge energy 0.75 kJ). A realistic distribution of magnetic pressure was calculated. The calculated values of displacement along the tube axis and versus time are in very good agreement with the measured ones.The maximum strain rate is 1 122 s-1, which is not large enough to change the constitutive equations of aluminum alloy. With the augment of discharge energy (0. 5 - 1.0 kJ), the relative errors of the maximum deformation increase from 2.93% to 11.4%. Therefore, coupled numerical modeling of the electromagnetic field and the structural field should be performed to investigate the electromagnetic forming with larger deformation.

  11. Monte Carlo Study on Gas Pressure Response of He-3 Tube in Neutron Porosity Logging

    Directory of Open Access Journals (Sweden)

    TIAN Li-li;ZHANG Feng;WANG Xin-guang;LIU Jun-tao

    2016-10-01

    Full Text Available Thermal neutrons are detected by (n,p reaction of Helium-3 tube in the compensated neutron logging. The helium gas pressure in the counting area influences neutron detection efficiency greatly, and then it is an important parameter for neutron porosity measurement accuracy. The variation law of counting rates of a near detector and a far one with helium gas pressure under different formation condition was simulated by Monte Carlo method. The results showed that with the increasing of helium pressure the counting rate of these detectors increased firstly and then leveled off. In addition, the neutron counting rate ratio and porosity sensitivity increased slightly, the porosity measurement error decreased exponentially, which improved the measurement accuracy. These research results can provide technical support for selecting the type of Helium-3 detector in developing neutron porosity logging.

  12. Pulsed electron beam propagation in gases under pressure of 6.6 kPa in drift tube

    Science.gov (United States)

    Kholodnaya, G. E.; Sazonov, R. V.; Ponomarev, D. V.; Remnev, G. E.; Poloskov, A. V.

    2017-02-01

    This paper presents the results of an investigation of pulsed electron beam transport propagated in a drift tube filled with different gases (He, H2, N2, Ar, SF6, and CO2). The total pressure in the drift tube was 6.6 kPa. The experiments were carried out using a TEA-500 pulsed electron accelerator. The electron beam was propagated in the drift tube composed of two sections equipped with reverse current shunts. Under a pressure of 6.6 kPa, the maximum value of the electron beam charge closed on the walls of the drift tube was recorded when the beam was propagated in hydrogen and carbon dioxide. The minimum value of the electron beam charge closed on the walls of the drift tube was recorded for sulfur hexafluoride. The visualization of the pulsed electron beam energy losses onto the walls of the drift chamber was carried out using radiation-sensitive film.

  13. 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.

  14. Degradation and buckling of metal tubes under cyclic bending and external pressure

    Science.gov (United States)

    Corona, Edmundo

    The response and stability of long tubular components under bending and external pressure were investigated. The behavior of the structure under monotonic as well as cyclic bending was examined through combined experimental and analytical efforts. The experiments involved metal seamless tubes with diameter-to-thickness ratios in the range of 17 to 35. Long specimens were tested under combined bending and pressure in a specially developed test facility. Bending-pressure interaction collapse envelopes were first generated for monotonically increasing loading histories. The two loads were found to interact strongly through the ovalization of the cross section and the collapse envelopes to depend on the loading history followed. Cyclic bending under various curvature controlled and moment controlled histories was considered. The factors influencing the rate of accumulation of ovalization and the resulting instabilities were studied parametrically. Buckling under cyclic loads occurred when the ovalization of the tubes reached a critical value approximately equal to the critical value developed under the corresponding monotonically applied loads. The problem was analyzed numerically using kinematics which capture the ovalization of the cross section. The predicted response was found to be very sensitive to the elastic-plastic constitutive models used. This sensitivity was carefully analyzed using state-of-the-art models. In the case of cyclic loading histories, the hardening rules used in such models were found to play a pivotal role in the accuracy of the predictions. The reasons for this sensitivity were studied through a parallel investigation of the behavior of the material under cyclic loads.

  15. Novel method for estimating the dynamic characteristics of pressure sensor in shock tube calibration test.

    Science.gov (United States)

    Li, Qiang; Wang, Zhongyu; Wang, Zhuoran; Yan, Hu

    2015-06-01

    A shock tube is usually used to excite the dynamic characteristics of the pressure sensor used in an aircraft. This paper proposes a novel estimation method for determining the dynamic characteristic parameters of the pressure sensor. A preprocessing operation based on Grey Model [GM(1,1)] and bootstrap method (BM) is employed to analyze the output of a calibrated pressure sensor under step excitation. Three sequences, which include the estimated value sequence, upper boundary, and lower boundary, are obtained. The processing methods on filtering and modeling are used to explore the three sequences independently. The optimal estimated, upper boundary, and lower boundary models are then established. The three models are solved, and a group of dynamic characteristic parameters corresponding to the estimated intervals are obtained. A shock tube calibration test consisting of two experiments is performed to validate the performance of the proposed method. The results show that the relative errors of the dynamic characteristic parameters of time and frequency domains do not exceed 9% and 10%, respectively. Moreover, the nominal and estimated values of the parameters fall into the estimated intervals limited by the upper and lower values.

  16. Numerical simulation of the processes in the normal incidence tube for high acoustic pressure levels

    Science.gov (United States)

    Fedotov, E. S.; Khramtsov, I. V.; Kustov, O. Yu.

    2016-10-01

    Numerical simulation of the acoustic processes in an impedance tube at high levels of acoustic pressure is a way to solve a problem of noise suppressing by liners. These studies used liner specimen that is one cylindrical Helmholtz resonator. The evaluation of the real and imaginary parts of the liner acoustic impedance and sound absorption coefficient was performed for sound pressure levels of 130, 140 and 150 dB. The numerical simulation used experimental data having been obtained on the impedance tube with normal incidence waves. At the first stage of the numerical simulation it was used the linearized Navier-Stokes equations, which describe well the imaginary part of the liner impedance whatever the sound pressure level. These equations were solved by finite element method in COMSOL Multiphysics program in axisymmetric formulation. At the second stage, the complete Navier-Stokes equations were solved by direct numerical simulation in ANSYS CFX in axisymmetric formulation. As the result, the acceptable agreement between numerical simulation and experiment was obtained.

  17. Experimental Investigation of Heat Transfer and Pressure Drop Characteristics of H-type Finned Tube Banks

    Directory of Open Access Journals (Sweden)

    Heng Chen

    2014-11-01

    Full Text Available H-type finned tube heat exchanger elements maintain a high capacity for heat transfer, possess superior self-cleaning properties and retain the ability to effect flue gas waste heat recovery in boiler renovations. In this paper, the heat transfer and pressure drop characteristics of H-type finned tube banks are studied via an experimental open high-temperature wind tunnel system. The effects of fin width, fin height, fin pitch and air velocity on fin efficiency, convective heat transfer coefficient, integrated heat transfer capacity and pressure drop are examined. The results indicate that as air velocity, fin height and fin width increase, fin efficiency decreases. Convective heat transfer coefficient is proportional to fin pitch, but inversely proportional to fin height and fin width. Integrated heat transfer capacity is related to fin efficiency, convective heat transfer coefficient and finned ratio. Pressure drop increases with the increase of fin height and fin width. Finally, predictive correlations of fin efficiency, Nusselt number and Euler Number are developed based on the experimental data.

  18. Heat transfer and pressure drop performance of smooth and internally finned tubes with oil and refrigerant 22 mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Schlager, L.M. (Indiana-Purdue Univ., Ft. Wayne, IN (US)); Pate, M.B. (Iowa State Univ., Ames, IA (US)); Bergles, A.E. (Rensselaer Polytechnic Inst., Troy, NY (US))

    1989-01-01

    The overall performance of a smooth tube and two augmented tubes is compared by using an enhancement performance ratio, defined as the ratio of heat transfer enhancement to pressure drop increase. The augmented tubes are compared to the smooth tube with pure R-22 and with mixtures of R-22 plus 150-or 300-SUS naphthenic mineral oil. Additionally, the performance of all three tubes with refrigerant-oil mixtures is compared to performance of the same tube with pure refrigerant. Various oil concentrations up to 5% by weight were tested and mass flux was varied from 92,000 to 294,000 lb/h{circle dot}ft{sup 2} (125 to 400 kg/m{sup 2}{circle dot}s). Nominal evaporation conditions were 37{degrees}F(3{degrees}C) with inlet and outlet qualities of 15% and 85%, respectively. The condensation conditions were 105{degrees}F (41{degrees}C) with inlet and outlet qualities of 85% and 15%, respectively. The enhancement performance ratio of the micro-fin tube is consistently higher than that of the low-fin tube with either pure refrigerant or refrigerant-oil mixtures. With pure refrigerant, the enhancement performance ratio of the augmented tubes is generally greater than unity, indicating improved performance relative to a smooth tube.

  19. Exact solution of unsteady flow generated by sinusoidal pressure gradient in a capillary tube

    Directory of Open Access Journals (Sweden)

    M. Abdulhameed

    2015-12-01

    Full Text Available In this paper, the mathematical modeling of unsteady second grade fluid in a capillary tube with sinusoidal pressure gradient is developed with non-homogenous boundary conditions. Exact analytical solutions for the velocity profiles have been obtained in explicit forms. These solutions are written as the sum of the steady and transient solutions for small and large times. For growing times, the starting solution reduces to the well-known periodic solution that coincides with the corresponding solution of a Newtonian fluid. Graphs representing the solutions are discussed.

  20. Experimental study on the pressure and pulse wave propagation in viscoelastic vessel tubes-effects of liquid viscosity and tube stiffness.

    Science.gov (United States)

    Ikenaga, Yuki; Nishi, Shohei; Komagata, Yuka; Saito, Masashi; Lagrée, Pierre-Yves; Asada, Takaaki; Matsukawa, Mami

    2013-11-01

    A pulse wave is the displacement wave which arises because of ejection of blood from the heart and reflection at vascular bed and distal point. The investigation of pressure waves leads to understanding the propagation characteristics of a pulse wave. To investigate the pulse wave behavior, an experimental study was performed using an artificial polymer tube and viscous liquid. A polyurethane tube and glycerin solution were used to simulate a blood vessel and blood, respectively. In the case of the 40 wt% glycerin solution, which corresponds to the viscosity of ordinary blood, the attenuation coefficient of a pressure wave in the tube decreased from 4.3 to 1.6 dB/m because of the tube stiffness (Young's modulus: 60 to 200 kPa). When the viscosity of liquid increased from approximately 4 to 10 mPa·s (the range of human blood viscosity) in the stiff tube, the attenuation coefficient of the pressure wave changed from 1.6 to 3.2 dB/m. The hardening of the blood vessel caused by aging and the increase of blood viscosity caused by illness possibly have opposite effects on the intravascular pressure wave. The effect of the viscosity of a liquid on the amplitude of a pressure wave was then considered using a phantom simulating human blood vessels. As a result, in the typical range of blood viscosity, the amplitude ratio of the waves obtained by the experiments with water and glycerin solution became 1:0.83. In comparison with clinical data, this value is much smaller than that seen from blood vessel hardening. Thus, it can be concluded that the blood viscosity seldom affects the attenuation of a pulse wave.

  1. The effect of flexible tube vibration on pressure drop and heat transfer in heat exchangers considering viscous dissipation effects

    Science.gov (United States)

    Shokouhmand, H.; Sangtarash, F.

    2008-04-01

    The pressure drop and heat transfer coefficient in tube bundle of shell and tube heat exchangers are investigated considering viscous dissipation effects. The governing equations are solved numerically. Because of temperature-dependent viscosity the equations should be solved simultaneously. The flexible tubes vibration is modeled in a quasi-static method by taking the first tube of the row to be in 20 asymmetric positions with respect to the rest of the tubes which are assumed to be fixed and time averaging the steady state solutions corresponding to each one of these positions .The results show that the eccentricity of the first tube increases pressure drop and heat transfer coefficients significantly comparing to the case of rigid tube bundles, symmetrically placed. In addition, these vibrations not only compensate the effect of viscous dissipations on heat transfer coefficient but also increase heat transfer coefficient. The constant viscosity results obtained from our numerical method have a good agreement with the available experimental data of constant viscosity for flexible tube heat exchangers.

  2. Resonant tube for measurement of sound absorption in gases at low frequency/pressure ratios

    Science.gov (United States)

    Zuckerwar, A. J.; Griffin, W. A.

    1980-01-01

    The paper describes a resonant tube for measuring sound absorption in gases, with specific emphasis on the vibrational relaxation peak of N2, over a range of frequency/pressure ratios from 0.1 to 2500 Hz/atm. The experimental background losses measured in argon agree with the theoretical wall losses except at few isolated frequencies. Rigid cavity terminations, external excitation, and a differential technique of background evaluation were used to minimize spurious contributions to the background losses. Room temperature measurements of sound absorption in binary mixtures of N2-CO2 in which both components are excitable resulted in the maximum frequency/pressure ratio in Hz/atm of 0.063 + 123m for the N2 vibrational relaxation peak, where m is mole percent of added CO2; the maximum ratio for the CO2 peak was 34,500 268m where m is mole percent of added N2.

  3. The pressure field in the liquid column in the tube-arrest method

    Institute of Scientific and Technical Information of China (English)

    Ying Chong-Fu; Li Chao; Xu De-Long; Deng Jing-Jun

    2008-01-01

    We have been using the method of tube-arrest as a means of producing transient single cavitation bubble. In the present paper we seek to comprehend the mechanism of production and inquire into the structure of the ab initio pressure field in the arrested liquid column. The generated pressure wave is shown by combining the theoretical analysis with the experimental observation to be a slightly varied version of water hammer. With relatively clean liquid, the magnitude of the tension peak generating the TSB is likely to reach of several millions Pa. It is also shown that the so generated cavitation bubble originating from the gas-containing bulk liquid is in 'violent' motion.

  4. Unusual cause of a facial pressure ulcer: the helmet securing the Sengstaken-Blakemore tube.

    Science.gov (United States)

    Kim, S M; Ju, R K; Lee, J H; Jun, Y J; Kim, Y J

    2015-06-01

    Many medical devices, such as pulse oximetry, ventilation masks and other splints are put on critically ill patients. Although these devices are designed to deliver relatively low physical pressure to the skin of the patient, they can still cause pressure ulcers (PUs) in critically ill patients. There are reports of medical device-related PUs on the face. Here we describe forehead skin necrosis caused by the securing helmet for the Sengstaken-Blakemore tube. It is difficult to detect this kind of PU early, because most of the patients have decreased mental status or delirium due to varix bleeding. For this reason, medical staff should be aware of the risk of developing a PU by the device and take preventive measures accordingly.

  5. Endotracheal Tube Cuff Pressures in Patients Intubated Prior to Helicopter EMS Transport

    Directory of Open Access Journals (Sweden)

    Joseph Tennyson

    2016-11-01

    Full Text Available Introduction: Endotracheal intubation is a common intervention in critical care patients undergoing helicopter emergency medical services (HEMS transportation. Measurement of endotracheal tube (ETT cuff pressures is not common practice in patients referred to our service. Animal studies have demonstrated an association between the pressure of the ETT cuff on the tracheal mucosa and decreased blood flow leading to mucosal ischemia and scarring. Cuff pressures greater than 30 cmH2O impede mucosal capillary blood flow. Multiple prior studies have recommended 30 cmH2O as the maximum safe cuff inflation pressure. This study sought to evaluate the inflation pressures in ETT cuffs of patients presenting to HEMS. Methods: We enrolled a convenience sample of patients presenting to UMass Memorial LifeFlight who were intubated by the sending facility or emergency medical services (EMS agency. Flight crews measured the ETT cuff pressures using a commercially available device. Those patients intubated by the flight crew were excluded from this analysis as the cuff was inflated with the manometer to a standardized pressure. Crews logged the results on a research form, and we analyzed the data using Microsoft Excel and an online statistical analysis tool. Results: We analyzed data for 55 patients. There was a mean age of 57 years (range 18-90. The mean ETT cuff pressure was 70 (95% CI= [61-80] cmH2O. The mean lies 40 cmH2O above the maximum accepted value of 30 cmH2O (p120 cmH2O, the maximum pressure on the analog gauge. Conclusion: Patients presenting to HEMS after intubation by the referral agency (EMS or hospital have ETT cuffs inflated to pressures that are, on average, more than double the recommended maximum. These patients are at risk for tracheal mucosal injury and scarring from decreased mucosal capillary blood flow. Hospital and EMS providers should use ETT cuff manometry to ensure that they inflate ETT cuffs to safe pressures.

  6. Pressure drop and stability of flow in Archimedean spiral tube with transverse corrugations

    Directory of Open Access Journals (Sweden)

    Đorđević Milan

    2016-01-01

    Full Text Available Isothermal pressure drop experiments were carried out for the steady Newtonian fluid flow in Archimedean spiral tube with transverse corrugations. Pressure drop correlations and stability criteria for distinguishing the flow regimes have been obtained in a continuous Reynolds number range from 150 to 15 000. The characterizing geometrical groups which take into account all the geometrical parameters of Archimedean spiral and corrugated pipe has been acquired. Before performing experiments over the Archimedean spiral, the corrugated straight pipe having high relative roughness e/d = 0.129 of approximately sinusoidal type was tested in order to obtain correlations for the Darcy friction factor. Insight into the magnitude of pressure loss in the proposed geometry of spiral solar receiver for different flow rates is important because of its effect upon the efficiency of the receiver. Although flow in spiral and corrugated geometries has the advantages of compactness and high heat transfer rates, the disadvantage of greater pressure drops makes hydrodynamic studies relevant. [Projekat Ministarstva nauke Republike Srbije, br. III 42006 i br. TR 33015

  7. Experimental study on the minimum drag coefficient of supercritical pressure water in horizontal tubes

    Energy Technology Data Exchange (ETDEWEB)

    Lei, Xianliang, E-mail: xianlianglei@mail.xjtu.edu.cn; Li, Huixiong; Guo, YuMeng; Zhang, Qing; Zhang, Weiqiang; Zhang, Qian

    2016-05-15

    Highlights: • The minimum drag coefficient phenomenon (MDC) has been observed and further investigated. • Effects of heat flux, mass flux and pressure to MDC have been discussed. • A series of comparisons between existing correlations and data have been conducted. • Two correlations of drag coefficient are proposed for isothermal and nonisothermal flow. - Abstract: Hydraulic resistance and its components are of great importance for understanding the turbulence nature of supercritical fluid and establishing prediction methods. Under supercritical pressures, the hydraulic resistance of the fluid exhibits a “pit” in the regions near its pseudo-critical point, which is hereafter called the minimum drag coefficient phenomenon. However, this special phenomenon was paid a little attention before. Hence systematical experiments have been carried out to investigate the hydraulic resistance of supercritical pressure water in both adiabatic and heated horizontal tubes. Parametric effects of heat flux, pressure and mass fluxes to drag coefficient are further compared. It is found that almost all of the existing correlations don’t agree well with the experimental data due to the insufficient consideration of thermal-properties near the pseudocritical point. Two correlations of the drag coefficients are finally proposed by introducing the new variable of the derivative of density with respect to temperature or Prandtl number, which can better predict the drag coefficient of isothermal and nonisothermal flow respectively.

  8. Severe accident analysis of a station blackout accident using MAAP-CANDU for the Point Lepreau station refurbishment project level 2 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Petoukhov, S.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station, using the MAAP-CANDU code to simulate the progression of severe core damage accidents and fission product releases. Five representative severe accidents were selected: Station Blackout, Small Loss-of-Coolant, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State. Analysis results for the reference station blackout accident are discussed in this paper. (author)

  9. 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.

  10. Evaluation of Pressure Stable Chip-to-Tube Fittings Enabling High-Speed Chip-HPLC with Mass Spectrometric Detection.

    Science.gov (United States)

    Lotter, Carsten; Heiland, Josef J; Stein, Volkmar; Klimkait, Michael; Queisser, Marco; Belder, Detlev

    2016-08-01

    Appropriate chip-to-tube interfacing is an enabling technology for high-pressure and high-speed liquid chromatography on chip. For this purpose, various approaches, to connect pressure resistant glass chips with HPLC pumps working at pressures of up to 500 bar, were examined. Three side-port and one top-port connection approach were evaluated with regard to pressure stability and extra column band broadening. A clamp-based top-port approach enabled chip-HPLC-MS analysis of herbicides at the highest pressure and speed.

  11. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  12. Performance assessment of an inline horizontal swirl tube cyclone for gas-liquid separation at high pressure

    Institute of Scientific and Technical Information of China (English)

    Nurhayati Mellon; Azmi M. Shariff

    2011-01-01

    The application of swirl tube cyclone for gas-liquid separation is attractive due to its small size and weight.However,very scarce information on the performance of the swirl tube cyclone especially at high operating pressure emulating actual field condition was published in journals.Performance assessment was usually done at a low operating pressure using either air-water,air-fine particle mixtures or dense gas such as SF6.This paper fills the existing gaps and reports the initial findings on the performance assessment of a horizontal swirl tube cyclone for gas-liquid separation operating at a flow rate of 5 MMSCFD at 40-60 bar operating pressure.

  13. Automatic ultrasonic inspection system for wear determination in calandria tubes of Embalse Nuclear Power Plant

    Science.gov (United States)

    Katchadjian, Pablo; Desimone, Carlos; Garcia, Alejandro; Antonaccio, Carlos; Schroeter, Fernando; Molina, Héctor

    2015-03-01

    Embalse Nuclear Power Plant (CNE) (CANDU design) is reaching its end of life and due to elapsed operating time the problem of deformation by accelerated creep occurs in the pressure tubes (PT), leading to a possible contact between calandria tubes (CT), concentric to the PT, and some Liquid Injection Shutdown System (LISS) nozzles that pass underneath them. With determination of CT wear, after the predicted contact occurs, the wear rate of the TC could be determined and thus take less conservative measures over the remaining life of the component. This paper presents the development of an ultrasonic technique for measuring wear in CT, with nominal thickness of 1.34 mm. Because the only access is through the interior of PT, to perform this measurement it is necessary to pass through three different interfaces.

  14. FURTHER MICROSTRUCTURAL EXAMINATIONS OF V-4Cr-4Ti PRESSURIZED CREEP TUBES

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, David S.; Kurtz, Richard J.

    2004-06-30

    Pressurized thermal creep tubes of V-4Cr-4Ti have been examined following testing in the range 650 to 800°C for tests lasting ~104 h. Creep deformation was found to be controlled by climb-controlled dislocation glide at all temperatures below 800°C whereas at 800°C, sub-grain boundary structure predominated and represented the main obstacle for dislocation motion. At 650 and 700°C after ~104 h an increased density of (Ti,V) oxy-carbo-nitride precipitates near the outer surface extending inwards a distance of 30 and 70 µm, respectively, was found. At 800°C, enhanced (Ti,V) oxy-carbo-nitride precipitation was observed across the entire tube wall thickness and may have affected creep response. Also, evidence for internal precipitation associated with the dislocation structure could be identified. The discussion section addresses differences in the controlling creep mechanisms between grain boundary sliding, sub-grain boundary controlled dislocation climb and individual dislocation climb processes.

  15. Shock tube investigation of dynamic response of pressure transducers for validation of rotor performance measurements

    Science.gov (United States)

    Bershader, Daniel

    1988-01-01

    For some time now, NASA has had a program under way to aid in the validation of rotor performance and acoustics codes associated with the UH-60 rotary-wing aircraft; and to correlate results of such studies with those obtained from investigations of other selected aircraft rotor performance. A central feature of these studies concerns the dynamic measurement of surface pressure at various locations up to frequencies of 25 KHz. For this purpose, fast-response gauges of the Kulite type are employed. The latter need to be buried in the rotor; they record surface pressures which are transmitted by a pipette connected to the gauge. The other end of the pipette is cut flush with the surface. In certain locations, the pipette configuration includes a rather sharp right-angle bend. The natural question has arisen in this connection: In what way are the pipettes modifying the signals received at the rotor surface and subsequently transmitted to the sensitive Kulite transducer element. The basic details and results of the program performed and recently completed in the High Pressure Shock Tube Laboratory of the Department of Aeronautics and Astronautics at Stanford University are given.

  16. A study for good regulatin of the CANDU's in Korea. Development of safety regulatory requirement for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Ki; Shin, Y. K.; Kim, J. S.; Yu, Y. J.; Lee, Y. J. [Ajou Univ., Suwon (Korea, Republic of)

    2001-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the first phase of this 2 years study, following research activities were done. On-site survey and analysis on CANDU plants regulation. Review on CANDU plants regulating experiences and current constraints. Review and analysis on the new Canadian regulatory approach.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  18. A unifying model for elongational flow of polymer melts and solutions based on the interchain tube pressure concept

    Science.gov (United States)

    Wagner, Manfred Hermann; Rolón-Garrido, Víctor Hugo

    2015-04-01

    An extended interchain tube pressure model for polymer melts and concentrated solutions is presented, based on the idea that the pressures exerted by a polymer chain on the walls of an anisotropic confinement are anisotropic (M. Doi and S. F. Edwards, The Theory of Polymer Dynamics, Oxford University Press, New York, 1986). In a tube model with variable tube diameter, chain stretch and tube diameter reduction are related, and at deformation rates larger than the inverse Rouse time τR, the chain is stretched and its confining tube becomes increasingly anisotropic. Tube diameter reduction leads to an interchain pressure in the lateral direction of the tube, which is proportional to the 3rd power of stretch (G. Marrucci and G. Ianniruberto. Macromolecules 37, 3934-3942, 2004). In the extended interchain tube pressure (EIP) model, it is assumed that chain stretch is balanced by interchain tube pressure in the lateral direction, and by a spring force in the longitudinal direction of the tube, which is linear in stretch. The scaling relations established for the relaxation modulus of concentrated solutions of polystyrene in oligomeric styrene (M. H. Wagner, Rheol. Acta 53, 765-777, 2014, M. H. Wagner, J. Non-Newtonian Fluid Mech. http://dx.doi.org/10.1016/j.jnnfm.2014.09.017, 2014) are applied to the solutions of polystyrene (PS) in diethyl phthalate (DEP) investigated by Bhattacharjee et al. (P. K. Bhattacharjee et al., Macromolecules 35, 10131-10148, 2002) and Acharya et al. (M. V. Acharya et al. AIP Conference Proceedings 1027, 391-393, 2008). The scaling relies on the difference ΔTg between the glass-transition temperatures of the melt and the glass-transition temperatures of the solutions. ΔTg can be inferred from the reported zero-shear viscosities, and the BSW spectra of the solutions are obtained from the BSW spectrum of the reference melt with good accuracy. Predictions of the EIP model are compared to the steady-state elongational viscosity data of PS

  19. Fuel management simulation for CANFLEX-RU in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fuel management simulations have been performed for CANFLEX-09% RU fuel in the CANDU 6 reactor. In this study, the bi-directional 4-bundle shift fuelling scheme was assumed. The lattice cell and time-average calculation were carried out. The refuelling simulation calculations were performed for 600 full power days. Time-averaged results show good axial power profile with the CANFLEX-RU fuel. During the simulation period, the maximum channel and bundle power were maintained below the licensing limit of CANDU 6 reactor. 7 refs., 4 figs. (Author)

  20. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.; Snell, V.; Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); West, J. [Candesco Co., Toronto, Ontario (Canada)

    2006-09-15

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross electrical output of 1165 MWe. The ACR-1000 design has evolved from AECL's in-depth knowledge of CANDU systems, components, and materials, as well as the experience and feedback received from owners and operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. It also features major improvements in economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The CANDU system is ideally suited to this evolutionary approach since the modular fuel channel reactor design can be modified, through a series of incremental changes in the reactor core design, to increase the power output and improve the overall safety, economics, and performance. The safety enhancements made in ACR-1000 encompass improved safety margins, performance and reliability of safety related systems. In particular, the use of the CANFLEX-ACR fuel bundle, with lower linear rating and higher critical heat flux, provides increased operating and safety margins. Safety features draw from those of the existing CANDU plants (e.g., the two

  1. Experimental investigation of heat transfer and pressure drop in fin-tube waste heat recovery heat exchangers

    OpenAIRE

    2014-01-01

    The aim of this master thesis was to investigate heat transfer and pressure drop of fin-tube heat exchangers. Experimental investigations of heat transfer and pressure drop in fin-tube bundles has been performed. The main focus was to investigate the influence of the fin height and the fin tip clearance. The effect of the uneven heat transfer distribution on the heat transfer coefficient has been analyzed.A literature survey has been dedicated to investigate the influence of the fin height an...

  2. Thermodynamic and fluid mechanic analysis of rapid pressurization in a dead-end tube

    Science.gov (United States)

    Leslie, Ian H.

    1989-01-01

    Three models have been applied to very rapid compression of oxygen in a dead-ended tube. Pressures as high as 41 MPa (6000 psi) leading to peak temperatures of 1400 K are predicted. These temperatures are well in excess of the autoignition temperature (750 K) of teflon, a frequently used material for lining hoses employed in oxygen service. These findings are in accord with experiments that have resulted in ignition and combustion of the teflon, leading to the combustion of the stainless steel braiding and catastrophic failure. The system analyzed was representative of a capped off-high-pressure oxygen line, which could be part of a larger system. Pressurization of the larger system would lead to compression in the dead-end line, and possible ignition of the teflon liner. The model consists of a large plenum containing oxygen at the desired pressure (500 to 6000 psi). The plenum is connected via a fast acting valve to a stainless steel tube 2 cm inside diameter. Opening times are on the order of 15 ms. Downstream of the valve is an orifice sized to increase filling times to around 100 ms. The total length from the valve to the dead-end is 150 cm. The distance from the valve to the orifice is 95 cm. The models describe the fluid mechanics and thermodynamics of the flow, and do not include any combustion phenomena. A purely thermodynamic model assumes filling to be complete upstream of the orifice before any gas passes through the orifice. This simplification is reasonable based on experiment and computer modeling. Results show that peak temperatures as high as 4800 K can result from recompression of the gas after expanding through the orifice. An approximate transient model without an orifice was developed assuming an isentropic compression process. An analytical solution was obtained. Results indicated that fill times can be considerably shorter than valve opening times. The third model was a finite difference, 1-D transient compressible flow model. Results from

  3. Instrumented thick-walled tube method for measuring thermal pressure in fluids and isotropic stresses in thermosetting resins

    Science.gov (United States)

    Merzlyakov, Mikhail; Simon, Sindee L.; McKenna, Gregory B.

    2005-06-01

    We have developed a method for measuring the thermal pressure coefficient and cure-induced and thermally induced stresses based on an instrumented thick-walled tube vessel. The device has been demonstrated at pressures up to 330 MPa and temperatures to 300 °C. The method uses a sealed stainless steel thick-walled tube to impose three-dimensional isotropic constraints. The tube is instrumented with strain gauges in hoop and in axial directions and can be used in open or closed configurations. By making measurements of the isotropic stresses as a function of temperature, the method allows determination of the thermal pressure coefficient in both the glassy and rubbery (or liquid) states. The method also can be used to measure isotropic stress development in thermosetting resins during cure and subsequent thermal cycling. Experimental results are presented for sucrose benzoate, di-2-ethylhexylsebacate, and an epoxy resin. The current report shows that the method provides reliable estimates for the thermal pressure coefficient. The thermal pressure coefficient is determined with resolution on the order of 10kPa/K. Among advantages of the method is that the tubes are reusable, even when measurements are made for cure response of thermosetting resins.

  4. Dynamic Runner Forces and Pressure Fluctuations on the Draft Tube Wall of a Model Pump-Turbine

    Science.gov (United States)

    Kirschner, O.; Ruprecht, A.; Göde, E.; Riedelbauch, S.

    2016-11-01

    When Francis-turbines and pump-turbines operate at off-design conditions, typically a vortex rope develops. The vortex rope causes pressure oscillations leading to fluctuations of the forces affecting the runner. The presence of dynamic runner forces over a long period of time might damage the bearings and possibly the runner. In this experimental investigation, the fluctuating part of the runner forces and the pressure oscillations on the draft tube wall were measured on a model pump-turbine with a simplified straight cone draft tube in different operating conditions. The investigation focuses on the correlation of the pressure fluctuations frequency measured at the draft tube wall with the frequency of the fluctuating forces on the runner. The comparison between pressure fluctuations and dynamic forces shows a significant correlation in all operating points. For the comparison of different components in the spatial directions of the forces, the pressure fluctuations were separated in a synchronous part and a rotating part for operating points with higher amplitudes. The rotating pressure fluctuations correlate with the radial forces especially in the operating points with a rotating vortex rope. At frequencies with higher amplitudes in the pressure fluctuations caused by the vortex rope movement, there are also higher amplitudes in the radial forces at the same frequencies.

  5. A review of CANDU feeder wall thinning

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Han Sub [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Flow Accelerated Corrosion is an active degradation mechanism of CANDU feeder. The tight bend downstream to Grayloc weld connection, close to reactor face, suffers significant wall thinning by FAC. Extensive in-service inspection of feeder wall thinning is very difficult because of the intense radiation field, complex geometry, and space restrictions. Development of a knowledge-based inspection program is important in order to guarantee that adequate wall thickness is maintained throughout the whole life of feeder. Research results and plant experiences are reviewed, and the plant inspection databases from Wolsong Units One to Four are analyzed in order to support developing such a knowledge-based inspection program. The initial thickness before wall thinning is highly non-uniform because of bending during manufacturing stage, and the thinning rate is non-uniform because of the mass transfer coefficient distributed non-uniformly depending on local hydraulics. It is obvious that the knowledge-based feeder inspection program should focus on both fastest thinning locations and thinnest locations. The feeder wall thinning rate is found to be correlated proportionately with QV of each channel. A statistical model is proposed to assess the remaining life of each feeder using the QV correlation and the measured thicknesses. W-1 feeder suffered significant thinning so that the shortest remaining life barely exceeded one year at the end of operation before replacement. W-2 feeder showed far slower thinning than W-1 feeder despite the faster coolant flow. It is believed that slower thinning in W-2 is because of higher chromium content in the carbon steel feeder material. The average Cr content of W-2 feeder is 0.051%, while that value is 0.02% for W-1 feeder. It is to be noted that FAC is reduced substantially even though the Cr content of W-2 feeder is still very low

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

    Science.gov (United States)

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

    2008-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Vaibhaw, Kumar [Nuclear Fuel Complex, ECIL Post, Hyderabad 500 062 (India)], E-mail: krvaibhaw@yahoo.co.in; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N. [Nuclear Fuel Complex, ECIL Post, Hyderabad 500 062 (India)

    2008-12-15

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

  8. Flow condensation pressure drop characteristics of R410A-oil mixture inside small diameter horizontal microfin tubes

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Xiangchao; Ding, Guoliang; Hu, Haitao; Zhu, Yu. [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Gao, Yifeng [International Copper Association Shanghai Office, Shanghai 200020 (China); Deng, Bin [Institute of Heat Transfer Technology, Golden Dragon Precise Copper Tube Group Inc., Shanghai 200135 (China)

    2010-11-15

    Flow condensation pressure drop characteristics of R410A-oil mixture inside small diameter (5.0 mm and 4.0 mm O.D.) horizontal microfin tubes were investigated experimentally covering nominal oil concentrations from 0% to 5%. The research results indicate that, comparing with the frictional pressure drop of pure R410A, the frictional pressure drop of R410A-oil mixture may decrease by maximum of 18% when the vapor quality is lower than 0.6, and increase by maximum of 13% when the vapor quality is higher than 0.6. A new frictional pressure drop correlation for R410A-oil mixture flow condensation inside microfin tubes is developed based on the refrigerant-oil mixture properties, and can agree with 94% of the experimental data within a deviation of -30% to +30%. (author)

  9. Overview of activities on CANDU fuel in Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez, L.; Valesi, J., E-mail: lalvarez@cnea.gov.ar [National Commission on Atomic Energy, Fuel Engineering Department (Argentina)

    2013-07-01

    This paper gives an outline of activities on CANDU fuel in Argentina. It discusses the nuclear activities and electricity production in Argentina, evolution of the activities in fuel engineering, fuel fabrication, fuel performance at Embalse nuclear power plant and spent fuel storage options.

  10. Proceedings of the fourth international conference on CANDU maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance.

  11. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  12. Estimation of CANDU spent fuel disposal canister lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Kook, Dong Hak; Lee, Min Soo; Hwang, Yong Soo; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    Active nuclear energy utilization causes significant spent fuel accumulation problem. The cumulative amount of spent fuel is about 10,083 ton as of Dec. 2008, and is expected to increase up to 19,000 ton by 2020. Of those, CANDU spent fuels account for more than 60% of the total amounts. CANDU spent fuels had been stored in dry concrete silos since 1991 and during the past 15 years, 300 silos were constructed and {approx}3,200 ton of spent fuels are stored now. Another dry storage facility MACSTOR /KN-400 will store new-coming CANDU spent fuels from 2009. But, after intermediate storage ends, all CANDU spent fuels have to be disposed within multi-layer metallic canister which is composed of cast iron inside and copper outside. Canister lifetime estimation, therefore, is very important for the final disposal safety analysis. The most significant factor of lifetime is copper corrosion, and Y. S. Hwang developed a corrosion model in order to predict the general corrosion effect on copper canister lifetime during the final disposal period. This research applied his model to KURT1 where many disposal researches are being performed actively and the results shows safe margin of the copper canister for the very long-term disposal.

  13. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  14. Ludwig: A Training Simulator of the Safety Operation of a CANDU Reactor

    Directory of Open Access Journals (Sweden)

    Gustavo Boroni

    2011-01-01

    Full Text Available This paper presents the application Ludwig designed to train operators of a CANDU Nuclear Power Plant (NPP by means of a computer control panel that simulates the response of the evolution of the physical variables of the plant under normal transients. The model includes a close set of equations representing the principal components of a CANDU NPP plant, a nodalized primary circuit, core, pressurizer, and steam generators. The design of the application was performed using the object-oriented programming paradigm, incorporating an event-driven process to reflect the action of the human operators and the automatic control system. A comprehensive set of online graphical displays are provided giving an in-depth understanding of transient neutronic and thermal hydraulic response of the power plant. The model was validated against data from a real transient occurring in the Argentine NPP Embalse Río Tercero, showing good agreement. However, it should be stressed that the aim of the simulator is in the training of operators and engineering students.

  15. Anomalous memory effect in the breakdown of low-pressure argon in a long discharge tube

    Energy Technology Data Exchange (ETDEWEB)

    Meshchanov, A. V.; Korshunov, A. N.; Ionikh, Yu. Z., E-mail: y.ionikh@spbu.ru [St. Petersburg State University (Russian Federation); Dyatko, N. A. [Troitsk Institute for Innovation and Fusion Research (Russian Federation)

    2015-08-15

    The characteristics of breakdown of argon in a long tube (with a gap length of 75 cm and diameter of 2.8 cm) at pressures of 1 and 5 Torr and stationary discharge currents of 5–40 mA were studied experimentally. The breakdown was initiated by paired positive voltage pulses with a rise rate of ∼10{sup 8}–10{sup 9} V/s and duration of ∼1–10 ms. The time interval between pairs was varied in the range of Τ ∼ 0.1–1 s, and that between pulses in a pair was varied from τ = 0.4 ms to ≈Τ/2. The aim of this work was to detect and study the so-called “anomalous memory effect” earlier observed in breakdown in nitrogen. The effect consists in the dynamic breakdown voltage in the second pulse in a pair being higher than in the first pulse (in contrast to the “normal” memory effect, in which the relation between the breakdown voltages is opposite). It is found that this effect is observed when the time interval between pairs of pulses is such that the first pulse in a pair is in the range of the normal memory effect of the preceding pair (under the given conditions, Τ ≈ 0.1–0.4 s). In this case, at τ ∼ 10 ms, the breakdown voltage of the second pulse is higher than the reduced breakdown voltage of the first pulse. Optical observations of the ionization wave preceding breakdown in a long tube show that, in the range of the anomalous memory effect and at smaller values of τ, no ionization wave is detected before breakdown in the second pulse. A qualitative interpretation of the experimental results is given.

  16. Heat transfer enhancement accompanying pressure-loss reduction with winglet-type vortex generators for fin-tube heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Torii, K.; Kwak, K.M.; Nishino, K. [Yokohama National Univ. (Japan). Dept. of Mechanical Engineering

    2002-08-01

    This paper proposes a novel technique that can augment heat transfer but nevertheless can reduce pressure-loss in a fin-tube heat exchanger with circular tubes in a relatively low Reynolds number flow, by deploying delta winglet-type vortex generators. The winglets are placed with a heretofore-unused orientation for the purpose of augmentation of heat transfer. This orientation is known as ''common flow up'' configuration. The proposed configuration causes significant separation delay, reduces form drag, and removes the zone of poor heat transfer from the near-wake of the tubes. This enhancement strategy has been successfully verified by experiments in the proposed configuration. In case of staggered tube banks, the heat transfer was augmented by 30% to 10%, and yet the pressure loss was reduced by 55% to 34% for the Reynolds number (based on two times channel height) ranging from 350 to 2100, when the present winglets were added. In case of in-line tube banks, these were found to be 20% to 10% augmentation, and 15% to 8% reduction, respectively. (author)

  17. EXPERIMENTAL INVESTIGATION ON HEAT TRANSFER AND PRESSURE DROP CHARACTERISTICS OF AIR FLOW OVER A STAGGERED FLAT TUBE BANK IN CROSSFLOW

    Directory of Open Access Journals (Sweden)

    M. Ishak

    2013-06-01

    Full Text Available This paper presents an experimental investigation into the heat transfer and pressure drop characteristics of air flow in a staggered flat tube bank in crossflow with laminar-forced convection. Measurements were conducted for sixteen tubes in the direction of flow and four tubes in rows. The air velocity varies between 0.6–1.0 m/s and the Reynolds number varied from 373 to 623. The total heat flux supplied in all tubes are changed from 967.92 to 3629.70 W/m2. The results indicate that the average Nusselt number for all the flat tubes increased by 11.46–46.42%, with the Reynolds numbers varying from 373 to 623 at the fixed heat flux. The average Nusselt number increased by 21.39–84%, and the total heat flux varyied between 967.92–3629.70 W/m2 with a constant Reynolds number Re = 498. In addition, the pressure drop decreased with an increase in the Reynolds number. A new mean Nusselt number-Reynolds number correlation was found, and the correlation yielded good predictions for the measured data with a confidence interval of 98.9%.

  18. Computational analysis of heat transfer and pressure drop performance for internally finned tubes with three different longitudinal wavy fins

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Qiu-Wang; Lin, Mei; Zeng, Min; Tian, Lin [Xi' an Jiaotong University, State Key Laboratory of Multiphase Flow in Power Engineering, Xi' an, Shaanxi (China)

    2008-12-15

    Turbulent pressure drop and heat transfer characteristics in tubes with three different kinds of internally longitudinal fin patterns (interrupted wavy, sinusoidal wavy and plain) are numerically investigated for Re=904-4,520. The channel velocity, temperature, and turbulence fields are obtained to discern the mechanisms of heat transfer enhancement. Numerical results indicate that the steady and spatially periodic growth and disruption of cross-sectional vortices occur near the tube/fin walls along the streamwise locations. The thermal boundary layers near the tube/fin surfaces are thereby periodically interrupted, with heat transfer near the recirculation zones being enhanced. The overall heat transfer coefficients in wavy channels are higher than those in a plain fin channel, while with larger pressure drop penalties. At the same waviness, the interrupted wavy fin tube could enhance heat transfer by 72-90%, with more than 2-4 times of pressure drop penalty. Among the fins studied, the sinusoidal wavy fin has the best comprehensive performance. (orig.)

  19. Characteristics of two-phase flow pattern transitions and pressure drop of five refrigerants in horizontal circular small tubes

    Energy Technology Data Exchange (ETDEWEB)

    Pamitran, A.S. [Department of Mechanical Engineering, University of Indonesia, Kampus Baru UI, Depok 16424 (Indonesia); Choi, Kwang-Il [Graduate School, Chonnam National University, San 96-1, Dunduk-Dong, Yeosu, Chonnam 550-749 (Korea); Oh, Jong-Taek [Department of Refrigeration and Air Conditioning Engineering, Chonnam National University, San 96-1, Dunduk-Dong, Yeosu, Chonnam 550-749 (Korea); Hrnjak, Pega [Department of Mechanical Science and Engineering, ACRC, University of Illinois at Urbana-Champaign, 1206 West Green Street, Urbana, IL 61801 (United States)

    2010-05-15

    An experimental investigation on the characteristics of two-phase flow pattern transitions and pressure drop of R-22, R-134a, R-410A, R-290 and R-744 in horizontal small stainless steel tubes of 0.5, 1.5 and 3.0 mm inner diameters is presented. Experimental data were obtained over a heat flux range of 5-40 kW/m{sup 2}, mass flux range of 50-600 kg/(m{sup 2} s), saturation temperature range of 0-15 C, and quality up to 1.0. Experimental data were evaluated with Wang et al. and Wojtan et al. [Wang, C.C., Chiang, C.S., Lu, D.C., 1997. Visual observation of two-phase flow pattern of R-22, R-134a, and R-407C in a 6.5-mm smooth tube. Exp. Therm. Fluid Sci. 15, 395-405; Wojtan, L., Ursenbacher, T., Thome, J.R., 2005. Investigation of flow boiling in horizontal tubes: part I - a new diabatic two-phase flow pattern map. Int. J. Heat Mass Transfer 48, 2955-2969.] flow pattern maps. The effects of mass flux, heat flux, saturation temperature and inner tube diameter on the pressure drop of the working refrigerants are reported. The experimental pressure drop was compared with the predictions from some existing correlations. A new two-phase pressure drop model that is based on a superposition model for two-phase flow boiling of refrigerants in small tubes is presented. (author)

  20. Heat transfer and pressure drop characteristics of plain finned heat exchangers having 5.0 mm tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Nae Hyun; Ham, Jung Ho; Oh, Wang Ku [Incheon Univ., Incheon (Korea, Republic of); Choi, Yong Hwa; Gaku, Hayase [Samsung Electric Company, Suwon (Korea, Republic of)

    2007-07-01

    In this study, pressure drop and heat transfer characteristics of plain finned heat exchangers having 5.0 diameter (fin collar 5.3 mm) tubes were investigated. Six samples having different fin pitches and tube rows were tested. The fin pitch had a negligible effect on j and f factors. Both j and f factors decreased as the number of tube row increased, although the difference was not significant for the f factor. When compared with the previous 7.3 mm diameter data, both the present j and f factors yielded lower values. However, the j/f ratio was larger at low Reynolds numbers. Possible reasoning is provided from the flow pattern consideration. Comparison with existing correlations were made.

  1. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2003-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  2. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Jae Young; Bang, Kwang Hyun [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2001-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SOS-1, SOS-2, ECCS and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  3. Simultaneous and long-lasting hydrophilization of inner and outer wall surfaces of polytetrafluoroethylene tubes by transferring atmospheric pressure plasmas

    Science.gov (United States)

    Chen, Faze; Song, Jinlong; Huang, Shuai; Xu, Sihao; Xia, Guangqing; Yang, Dezheng; Xu, Wenji; Sun, Jing; Liu, Xin

    2016-09-01

    Plasma hydrophilization is a general method to increase the surface free energy of materials. However, only a few works about plasma modification focus on the hydrophilization of tube inner and outer walls. In this paper, we realize simultaneous and long-lasting plasma hydrophilization on the inner and outer walls of polytetrafluoroethylene (PTFE) tubes by atmospheric pressure plasmas (APPs). Specifically, an Ar atmospheric pressure plasma jet (APPJ) is used to modify the PTFE tube’s outer wall and meanwhile to induce transferred He APP inside the PTFE tube to modify its inner wall surface. The optical emission spectrum (OES) shows that the plasmas contain many chemically active species, which are known as enablers for various applications. Water contact angle (WCA) measurements, x-ray photoelectron spectroscopy (XPS) and atomic force microscopy (AFM) are used to characterize the plasma hydrophilization. Results demonstrate that the wettability of the tube walls are well improved due to the replacement of the surface fluorine by oxygen and the change of surface roughness. The obtained hydrophilicity decreases slowly during more than 180 d aging, indicating a long-lasting hydrophilization. The results presented here clearly demonstrate the great potential of transferring APPs for surface modification of the tube’s inner and outer walls simultaneously.

  4. Changes in endotracheal tube cuff pressure during laparoscopic surgery in head-up or head-down position

    Science.gov (United States)

    2014-01-01

    Background The abdominal insufflation and surgical positioning in the laparoscopic surgery have been reported to result in an increase of airway pressure. However, associated effects on changes of endotracheal tube cuff pressure are not well established. Methods 70 patients undergoing elective laparoscopic colorectal tumor resection (head-down position, n = 38) and laparoscopic cholecystecomy (head-up position, n = 32) were enrolled and were compared to 15 patients undergoing elective open abdominal surgery. Changes of cuff and airway pressures before and after abdominal insufflation in supine position and after head-down or head-up positioning were analysed and compared. Results There was no significant cuff and airway pressure changes during the first fifteen minutes in open abdominal surgery. After insufflation, the cuff pressure increased from 26 ± 3 to 32 ± 6 and 27 ± 3 to 33 ± 5 cmH2O in patients receiving laparoscopic cholecystecomy and laparoscopic colorectal tumor resection respectively (both p < 0.001). The head-down tilt further increased cuff pressure from 33 ± 5 to 35 ± 5 cmH2O (p < 0.001). There six patients undergoing colorectal tumor resection (18.8%) and eight patients undergoing cholecystecomy (21.1%) had a total increase of cuff pressure more than 10 cm H2O (18.8%). There was no significant correlation between increase of cuff pressure and either the patient's body mass index or the common range of intra-abdominal pressure (10-15 mmHg) used in laparoscopic surgery. Conclusions An increase of endotracheal tube cuff pressure may occur during laparoscopic surgery especially in the head-down position. PMID:25210501

  5. Time-resolved detection of temperature, concentration, and pressure in a shock tube by intracavity absorption spectroscopy

    Science.gov (United States)

    Fjodorow, Peter; Fikri, Mustapha; Schulz, Christof; Hellmig, Ortwin; Baev, Valery M.

    2016-06-01

    In this paper, we demonstrate the first application of intracavity absorption spectroscopy (ICAS) for monitoring species concentration, total pressure, and temperature in shock-tube experiments. ICAS with a broadband Er3+-doped fiber laser is applied to time-resolved measurements of absorption spectra of shock-heated C2H2. The measurements are performed in a spectral range between 6512 and 6542 cm-1, including many absorption lines of C2H2, with a time resolution of 100 µs and an effective absorption path length of 15 m. Up to 18-times increase of the total pressure and a temperature rise of up to 1200 K have been monitored. Due to the ability of simultaneously recording many absorption lines in a broad spectral range, the presented technique can also be applied to multi-component analysis of transient single-shot processes in reactive gas mixtures in shock tubes, pulse detonation engines, or explosions.

  6. A step towards closing the CANDU fuel cycle: an innovative scheme for reprocessing used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Collins, F.; Lister, D. [Univ. of New Brunswick, UNB Nuclear, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada)

    2011-07-01

    Disposal versus reprocessing costs for used CANDU fuel was recently discussed by Rozon and Lister in a report produced for the Nuclear Waste Management Organization (NWMO). Their study discussed the economic incentives for reprocessing, not for the recovery of fissile uranium but for the recovery of plutonium ash. A $370/kg break-even price of uranium was calculated, and their model was found to be very sensitive to the reprocessing costs of the chosen technology. Findings were consistent with earlier studies done by Harvard University. Various reprocessing technologies (most based on solvent extraction) have been in use for many decades, but there appears to be no conceptual engineering study available in the open literature for a spent fuel reprocessing facility - one that includes process flows, operating costs and economic analysis. A deeper engineering study of the design and economics of re-processing technologies has since been undertaken by the nuclear group at the University of New Brunswick. An improved fluorination process was developed and modeled using ASPEN process simulation software. This study examines the impact of chosen technology on the spent fuel re-processing costs. (author)

  7. Analysis of stress-induced Burgers vector anisotropy in pressurized tube specimens of irradiated ferritic-martensitic steel: JLF-1

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States); Shibayama, T. [Univ. of Hokkaido, Oarai, Ibaraki (Japan). Inst. for Materials Research

    1998-09-01

    A procedure for determining the Burgers vector anisotropy in irradiated ferritic steels allowing identification of all a<100> and all a/2<111> dislocations in a region of interest is applied to a pressurized tube specimen of JLF-1 irradiated at 430 C to 14.3 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) or 61 dpa. Analysis of micrographs indicates large anisotropy in Burgers vector populations develop during irradiation creep.

  8. Post-irradiation examinations of a Zr2.5Nb pressure tube of the Karachi nuclear power plant (KANUPP)

    Science.gov (United States)

    Zaheer, Mohammed Sajjad; Akhtar, Javed Iqbal; Ahmad, Ejaz; Saleem, Muhammad; Hussain, Syed Mukarrum; Qureshi, Masroor Ahmad; Khan, Azmatullah; Ali, Rafaqat; Zafarullah, Muhammad

    1996-09-01

    The results of post-irradiation examinations of a pressure tube of fuel channel No. G-12 of KANUPP have been described. A detailed study was made in Canada by AECL. A parallel investigation on its seven rings of about 50 mm length each was also carried out at PINSTECH. Visual inspection showed normal oxidation effects. Gamma spectrometry showed the presence of 95Zr and 95Nb. Microstructural study revealed the characteristic alpha plus a transformed beta phase structure.

  9. Comparison Study on Thermal-Hydraulic Analysis Depending on Liquid Relief Valve Response for an Station Blackout in CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. M.; Kho, D. W. [KHNP-CRI, Daejeon (Korea, Republic of); Choi, S. H.; Moon, B. J.; Kim, S. R. [Nuclear Engineering Service and Solution Co., Daejeon (Korea, Republic of)

    2014-10-15

    The purpose of this analysis is to compare the results of thermal-hydraulic analysis depending on liquid relief valve response during a station black out (SBO) events in CANDU-6. The primary heat transport system (PHTS) behavior following the postulated SBO is analyzed using CATHENA code. In the paper, analysis was performed to evaluate the effect on coolant system where LRVs are assumed to be opened or opened according to normal open characteristics in the condition of SBO. The result showed that the primary pressure boundary is extended from LRV to DCT and the effects on primary system behavior were neglectable.

  10. Scenarios for the transmutation of actinides in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, Bronwyn, E-mail: hylandb@aecl.ca [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Gihm, Brian, E-mail: gihmb@aecl.ca [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2011-12-15

    With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100-1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

  11. Qualification of inspection systems in the CANDU nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

    Baron, J.A. [CANDU Owners Group, CANDU Inspection Qualification Bureau, Toronto, Ontario (Canada)

    2014-01-15

    Most jurisdictions that generate electricity through nuclear-electric plants have imposed requirements on inspection systems beyond the typical Level 1, 2 and 3 found in personnel qualification/certification schemes. The paper discusses the rationale for this obligation and describes how the requirement for inspection qualification has been implemented for CANDU plants. The paper discusses the qualification structure and process, including a brief overview of experience to-date in qualifying Inspection Procedures. (author)

  12. Convective Heat and Mass Transfer in Water at Super—Critical Pressures under Heating or Cooling Conditions in Vertical Tubes

    Institute of Scientific and Technical Information of China (English)

    Pei-XueJiang; Ze-PeiRen; 等

    1995-01-01

    Forced and mixed convection heat and mass transfer are studied numerically for water containing metallic corrosion products in a heated or cooled vertical tube with variable thermophysical properties at super-citical pressures.the fouling mechanisms and fouling models are presented.The influence of variable properties at super-critical pressures on forced or mixed convection has been analyzed.The differences between heat and mass transfer under heating and cooling conditions are discussed.It is found that variable properties,especially buoyancy,greatly influence the fluid flow and heat mass fransfer.

  13. A novel thermobaric analyser: in situ measurement of gas pressure during synthesis in sealed quartz tube at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, A.G.; Orlando, M.T.D. [Departamento de Fisica, Universidade Federal do Espirito Santo, 29060-900 Vitoria-ES (Brazil); Centro Brasileiro de Pesquisas Fisicas, Rua Dr Xavier Sigaud 150-Urca, 22290-180 Rio de Janeiro (Brazil); Sin, A.; Granados, X.; Calleja, A.; Pinol, S.; Obradors, X. [Institut de Ciencia de Materials de Barcelona (CSIC), Campus de la UAB, Bellaterra E-08193, Barcelona (Spain); Emmerich, F.G. [Departamento de Fisica, Universidade Federal do Espirito Santo, 29060-900 Vitoria-ES (Brazil); Baggio-Saitovitch, E. [Centro Brasileiro de Pesquisas Fisicas, Rua Dr Xavier Sigaud 150-Urca, 22290-180 Rio de Janeiro (Brazil)

    2000-11-01

    We have developed a novel technique (thermobaric analysis or TBA) to measure, in situ up to 900 deg. C, the pressure of gases such as Hg and O{sub 2} in sealed quartz tubes. The pressure determination in closed systems enables us to obtain information on the synthesis of compounds which involve solid-gas reactions. The concept of the TBA set-up is described, including the calibration method and the verification with HgO decomposition. The technique is applied to the optimized synthesis of the ceramic Hg, Re-1223 superconductor. (author)

  14. Assessment of DUPIC fuel compatibility with CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Choi, H. B.; Roh, G. H.; Jeong, C. J.; Rhee, B. W.; Choi, J. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    The compatibility of DUPIC fuel with the existing CANDU reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will require a quantitative analysis later. (author).

  15. Advancement of safeguards inspection technology for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Sung; Park, W. S.; Cha, H. R.; Ham, Y. S.; Lee, Y. G.; Kim, K. P.; Hong, Y. D

    1999-04-01

    The objectives of this project are to develop both inspection technology and safeguards instruments, related to CANDU safeguards inspection, through international cooperation, so that those outcomes are to be applied in field inspections of national safeguards. Furthermore, those could contribute to the improvement of verification correctness of IAEA inspections. Considering the level of national inspection technology, it looked not possible to perform national inspections without the joint use of containment and surveillance equipment conjunction with the IAEA. In this connection, basic studies for the successful implementation of national inspections was performed, optimal structure of safeguards inspection was attained, and advancement of safeguards inspection technology was forwarded. The successful implementation of this project contributed to both the improvement of inspection technology on CANDU reactors and the implementation of national inspection to be performed according to the legal framework. In addition, it would be an opportunity to improve the ability of negotiating in equal shares in relation to the IAEA on the occasion of discussing or negotiating the safeguards issues concerned. Now that the national safeguards technology for CANDU reactors was developed, the safeguards criteria, procedure and instruments as to the other item facilities and fabrication facilities should be developed for the perfection of national inspections. It would be desirable that the recommendations proposed and concreted in this study, so as to both cope with the strengthened international safeguards and detect the undeclared nuclear activities, could be applied to national safeguards scheme. (author)

  16. Two-phase heat transfer and pressure drop of LNG during saturated flow boiling in a horizontal tube

    Science.gov (United States)

    Chen, Dongsheng; Shi, Yumei

    2013-12-01

    Two-phase heat transfer and pressure drop of LNG (liquefied natural gas) have been measured in a horizontal smooth tube with an inner diameter of 8 mm. The experiments were conducted at inlet pressures from 0.3 to 0.7 MPa with a heat flux of 8-36 kW m-2, and mass flux of 49.2-201.8 kg m-2 s-1. The effect of vapor quality, inlet pressure, heat flux and mass flux on the heat transfer characteristic are discussed. The comparisons of the experimental data with the predicted value by existing correlations are analyzed. Zou et al. (2010) correlation shows the best accuracy with 24.1% RMS deviation among them. Moreover four frictional pressure drop methods are also chosen to compare with the experimental database.

  17. Modeling and experiments with low-frequency pressure wave propagation in liquid-filled, flexible tubes

    DEFF Research Database (Denmark)

    Bjelland, C; Bjarnø, Leif

    1992-01-01

    A model for wave propagation in a liquid-filled viscoelastic tube with arrays of receivers inside, is being used to analyze the influence of noise generated by in-line vibrational noise sources. In this model, distensibility is of greater importance than compressibility of the liquid....... The dispersion and attenuation is shown to be strongly dependent on the viscoelastic properties of the tube wall. The complex, frequency-dependent moduli of relevant tube materials have been measured in stress wave transfer function experiments. The moduli are used in the model to produce realistic dispersion...... relations and frequency-dependent attenuation. A 12-m-long, liquid-filled tube with interior stress members and connectors in each end is hanging vertically from an upper fixture. The lower end connector is excited by a power vibrator to generate the relevant wave modes. Measurements with reference...

  18. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  19. Experimental study of heat transfer of ultra-supercritical pressure water in vertical upward internally ribbed tube

    Institute of Scientific and Technical Information of China (English)

    Wang Weishu; Chen Tingkuan; Luo Yushan; Gu Hongfang; Yin Fei

    2007-01-01

    Under ultra-supercritical pressure, the heat transfer characteristics of water in vertical upward 4-head internally ribbed tubes with a diameter of 28.65mm and thickness of 8mm were experimentally studied.The experiments were performed at P=25~34MPa,G=450~1800kg/(m2·s)and q=200~600kW/m2. The results show that the pressure has only a moderate effect on the heat transfer of ultra-supercritical water when the water temperature is below the pseudocritical point. Sharp rise of the wall temperature near the pesudocritical region occurs earlier at a higher pressure. Increasing the mass velocity improves the heat transfer with a much stronger effect below the pesudocritical point than that above the pesudocritical point. For given pressure and mass velocity, the inner wall heat flux also shows a significant effect on the inner wall temperature, with a higher inner wall heat flux leading to a higher inner wall temperature. Increasing of inner wall heat flux leads to an early occurrence of sharp rise of the wall temperature. Correlations of heat transfer coefficients are also presented for vertical upward internally ribbed tubes.

  20. Study of the velocity distribution influence upon the pressure pulsations in draft tube model of hydro-turbine

    Science.gov (United States)

    Sonin, V.; Ustimenko, A.; Kuibin, P.; Litvinov, I.; Shtork, S.

    2016-11-01

    One of the mechanisms of generation of powerful pressure pulsations in the circuit of the turbine is a precessing vortex core, formed behind the runner at the operation points with partial or forced loads, when the flow has significant residual swirl. To study periodic pressure pulsations behind the runner the authors of this paper use approaches of experimental modeling and methods of computational fluid dynamics. The influence of velocity distributions at the output of the hydro turbine runner on pressure pulsations was studied based on analysis of the existing and possible velocity distributions in hydraulic turbines and selection of the distribution in the extended range. Preliminary numerical calculations have showed that the velocity distribution can be modeled without reproduction of the entire geometry of the circuit, using a combination of two blade cascades of the rotor and stator. Experimental verification of numerical results was carried out in an air bench, using the method of 3D-printing for fabrication of the blade cascades and the geometry of the draft tube of hydraulic turbine. Measurements of the velocity field at the input to a draft tube cone and registration of pressure pulsations due to precessing vortex core have allowed building correlations between the velocity distribution character and the amplitude-frequency characteristics of the pulsations.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

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

  2. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-03-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.

  3. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)

  4. Analysis of the impact of coolant density variations in the high efficiency channel of a pressure tube super critical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scriven, M.G.; Hummel, D.W.; Novog, D.R.; Luxat, J.C. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    The Pressure Tube (PT) Supercritical Water Reactor (SCWR) is based on a light water coolant operating at pressures above the thermodynamic critical pressure; a separate low temperature and low pressure moderator. The coolant density changes by an order of magnitude depending on its local enthalpy in the porous ceramic insulator tube. This causes significant changes in the neutron transport characteristics, axially and radially, in the fuel channel. This work performs lattice physics calculations for a 78-element Pu-Th fuel at zero burnup and examines the effect of assumptions related to coolant density in the radial direction of a HEC, using the neutron transport code WIMS-AECL. (author)

  5. Severe accident analysis of a small LOCA accident using MAAP-CANDU support level 2 PSA for the Point Lepreau station refurbishment project

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J.; Mathew, P.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP4-CANDU code was used to calculate the progression of postulated severe core damage accidents and fission product releases. Five representative severe core damage accidents were selected: Station Blackout, Small Loss-of-Coolant Accident, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State Accident. Analysis results for only the reference Small LOCA Accident scenario (which is a very low probability event) are discussed in this paper. (author)

  6. Comparison of the cuff pressure of a TaperGuard endotracheal tube and a cylindrical endotracheal tube after lateral rotation of head during middle ear surgery

    Science.gov (United States)

    Choi, Eunkyung; Park, Yongmin; Jeon, Younghoon

    2017-01-01

    Abstract Background: Positional change affects the cuff pressure of an endotracheal tube (ETT) in treacheally intubated patients. We compared the cuff pressure of a TaperGuard ETT and a cylindrical ETT after lateral rotation of head during middle ear surgery. Methods: Fifty-two patients aged 18–70 years underwent a tympanomastoidectomy under general anesthesia were randomly allocated to receive endotracheal intubation with cylindrical (group C, n = 26) or TaperGuard ETTs (group T, n = 26). After endotracheal intubation, the ETT cuff pressure was set at 22 cmH2O in the neutral position of head. After lateral rotation of head, the cuff pressure was measured again and readjusted to 22 cmH2O. In addition, the change of distance from the carina to the tip of the ETT was measured before and after the positional change. The incidence of cough, sore throat, and hoarseness was assessed at 30 minutes, 6 hours, and 24 hours after surgery. Results: There was no difference in demographic data between groups. After lateral rotation of head, the cuff pressure significantly increased in group T (11.9 ± 2.3 cmH2O) compared with group C (6.0 ± 1.9 cmH2O) (P 30 cmH2O was higher in group T (96.2%) than in group C (30.8%) (P < 0.001). In addition, the degree of displacement of an ETT was greater in group T (11.0 ± 1.7 mm) than in group C (7.2 ± 2.6 mm) (P < 0.001). The overall incidences of postoperative sore throat, hoarseness, and cough at 30 minutes, 6 hours, and 24 hours after surgery were comparable between two groups. Conclusion: The cuff pressure was higher in the TaperGuard ETT than in the cylindrical ETT after positional change of head from neutral to lateral rotation. In addition, after a positional change, the extent of displacement of ETT was greater in the TaperGuard ETT than in the cylindrical ETT. PMID:28272230

  7. Experimental Research on Heat Transfer and Pressure Drop of Two Configurations of Pin Finned—Tubes in an In—line Array

    Institute of Scientific and Technical Information of China (English)

    ShouGuangYao; DeShuZhu

    1994-01-01

    In this paper,a local simulation method is employed to investigate the heat transfer and pressure drop characteristics of two configurations of pin finned tubes deployed in an in-line array,In this research,heat pipes are adopted as heating elements.Therefore,the experimental equipment becomes simple and has an advantage of sufficient reducibility.The air-side heat transfer and pressure drop correlations for each type of pin fin surface including the effect of the tube-row number are obtained in the Reynolds number range commonly encountered in engineering.These correlations may be used in the design of pin finned tube heat exchangers.

  8. Experimental and numerical investigations of three-dimensional turbulent flow of water surrounding a CANDU simulation fuel bundle structure inside a channel

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Yu, S.D. [Department of Mechanical and Industrial Engineering, Ryerson University, 350 Victoria Street, Toronto, Ontario, M5B 2K3 (Canada); Cao, J. [Department of Mechanical and Industrial Engineering, Ryerson University, 350 Victoria Street, Toronto, Ontario, M5B 2K3 (Canada)], E-mail: jcao@ryerson.ca

    2009-11-15

    Computational fluid dynamics (CFD) is used to simulate highly turbulent coolant flows surrounding a simulation CANDU fuel bundle structure inside a flow channel. Three CFD methods are used: large eddy simulation (LES), detached eddy simulation (DES), and Reynolds stress model (RSM). The outcome of the simulations is compared with the experimental pressure data measured using an in-water microphone and a miniature pressure transducer placed at various locations in the vicinity of the bundle structure. Among all the three methods employed in developing computational models, LES provides the most accurate results for turbulent pressures.

  9. Linear and nonlinear viscoelastic properties of bidisperse linear polymers: Mixing law and tube pressure effect

    DEFF Research Database (Denmark)

    van Ruymbeke, E.; Nielsen, J.; Hassager, Ole

    2010-01-01

    In this manuscript, we extend the tube-based model that we developed for predicting the linear viscoelasticity of entangled polymers [van Ruymbeke et al., J. Non-Newtonian Fluid Mech. 128, 7-22 (2005)] to the prediction of the extensional rheology of monodisperse and bidisperse linear polymers...

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. S.; Jeong, Y. M.; Ahn, S. B. (and others)

    2005-03-15

    Major degradation of the feeder pipe is the thinning due to the flow accelerated corrosion and the cracking in the bent region due to the stress corrosion cracking. The feeder pipe in a PHWR is a pipe to supply the coolant to the pressure tube and the heated coolant to the steam generator for power generation. Approximately 380 pipes are installed on the inlet side and outlet side each with two bent regions in the 600 MW-class PHWR. After a leakage in the bent region of the feeder pipe, it is required to examine all the pipes in order to ensure the integrity of the pressure boundaries. It is not easy, however, to examine all the pipes with the conventional ultrasonic method, because of a high dose of radiation exposure and a limited accessibility to the pipe. In order to get rid of the limited accessibility, the ultrasonic guided wave method are developed for detection and evaluation of the cracks in the feeder pipe. The dispersion mode analysis was performed for the development of long-range guided wave inspection for the feeder pipe. An analytical approach for the straight pipe as well as numerical approach for the bent pipe with 2-D FFT were accomplished. A computer program for the calculation of the dispersion curves and wave structures was developed. Based on the dispersion curves and wave structure of the feeder pipe, candidates for the optimal parameters on the frequencies and vibration modes were selected. A time-frequency analysis methodology was developed for the mode identification of received ultrasonic signal. A high power tone-burst ultrasonic system has been setup for the generation of guided waves. Various artificial notches were fabricated on the bent feeder pipes for the experiment on the flaw detection. Considering the results of dispersion analysis and field condition, the torsional vibration mode, T(0,1) is selected for the first choice. An array of electromagnetic acoustic transducers (EMAT) was designed and fabricated for the generation of T

  11. Measurement and correlation of frictional pressure drop of refrigerant-based nanofluid flow boiling inside a horizontal smooth tube

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Hao; Ding, Guoliang; Jiang, Weiting; Hu, Haitao [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, 800 Dongchuan Road, Shanghai 200240 (China); Gao, Yifeng [International Copper Association Shanghai Office, 381 Huaihaizhong Road, Shanghai 200020 (China)

    2009-11-15

    The objective of this paper is to investigate the effect of nanoparticle on the frictional pressure drop characteristics of refrigerant-based nanofluid flow boiling inside a horizontal smooth tube, and to present a correlation for predicting the frictional pressure drop of refrigerant-based nanofluid. R113 refrigerant and CuO nanoparticle were used for preparing refrigerant-based nanofluid. Experimental conditions include mass fluxes from 100 to 200 kg m{sup -2} s{sup -1}, heat fluxes from 3.08 to 6.16 kW m{sup -2}, inlet vapor qualities from 0.2 to 0.7, and mass fractions of nanoparticles from 0 to 0.5 wt%. The experimental results show that the frictional pressured drop of refrigerant-based nanofluid increases with the increase of the mass fraction of nanoparticles, and the maximum enhancement of frictional pressure drop is 20.8% under above conditions. A frictional pressure drop correlation for refrigerant-based nanofluid is proposed, and the predictions agree with 92% of the experimental data within the deviation of {+-}15%. (author)

  12. Amorphous carbon film deposition on inner surface of tubes using atmospheric pressure pulsed filamentary plasma source

    CERN Document Server

    Pothiraja, Ramasamy; Awakowicz, Peter

    2011-01-01

    Uniform amorphous carbon film is deposited on the inner surface of quartz tube having the inner diameter of 6 mm and the outer diameter of 8 mm. A pulsed filamentary plasma source is used for the deposition. Long plasma filaments (~ 140 mm) as a positive discharge are generated inside the tube in argon with methane admixture. FTIR-ATR, XRD, SEM, LSM and XPS analyses give the conclusion that deposited film is amorphous composed of non-hydrogenated sp2 carbon and hydrogenated sp3 carbon. Plasma is characterized using optical emission spectroscopy, voltage-current measurement, microphotography and numerical simulation. On the basis of observed plasma parameters, the kinetics of the film deposition process is discussed.

  13. The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

    Energy Technology Data Exchange (ETDEWEB)

    Agafonova, N. [St. Petersburg State Technical Univ. (Russian Federation); Banati, J. [Lappeenranta Univ. of Technology (Finland)

    1997-12-31

    RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

  14. Experimental study of vapor local characteristics in upward low pressure boiling tube

    Institute of Scientific and Technical Information of China (English)

    SUN Qi; ZHAO Hua; XI Zhao; YANG Rui-Chang

    2003-01-01

    Radial distribution of vapor local parameters, including local void fraction, interfacial velocity, bubblesize, bubble frequency and interfacial area concentration, are investigated through the measurement in an upwardboiling tube using dual-sensor optical probe. In addition, a new local parameter -"local bubble number concentra-tion" is developed on the basis of bubble frequency. The analysis shows that this parameter can reflect bubble numberdensity in space, and has clear physical meaning.

  15. Pressure drop measurements in the transition region for a circular tube with a square-edged entrance

    Science.gov (United States)

    Ghajar, Afshin J.; Augustine, Jody R.

    1990-06-01

    Pressure drop measurements were made in a horizontal circular straight tube with a square-edged entrance under isothermal flow conditions. The experiments covered a Reynolds number range from 512 to 14,970. A total of thirty-three sets of experimental data for the twenty pressure tap locations along the 20 ft length of the test section were gathered. For the square-edged entrance the range of Reynolds number for which transition flow exists was determined to be between 2070 to 2840. A correlation for prediction of fully developed skin friction coefficient in this region is recommended. In the entrance region the length required for the friction factor to become fully developed in both the laminar and turbulent regions was found to be inversely proportional to the Reynolds number, with the turbulent data showing a stronger dependency. A correlation for prediction of entrance length in the turbulent region is offered.

  16. Fabrication of Zr-2.5Nb pressure tubes to minimize the harmful effects of trace elements

    Energy Technology Data Exchange (ETDEWEB)

    Theaker, J.R.; Coleman, C.E. [AECL Research, Chalk River, Ontario (Canada). Chalk River Labs.; Choubey, R. [AECL Research, Pinawa, Manitoba (Canada). Whiteshell Labs.; Moan, G.D. [AECL CANDU, Mississauga, Ontario (Canada); Aldridge, S.A. [Nu-Tech Precision Metals Inc., Arnprior, Ontario (Canada); Davis, L.; Graham, R.A. [Teledyne Wah Chang Albany, OR (United States)

    1994-12-31

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

  17. Design and analysis of 19 pin annular fuel rod cluster for pressure tube type boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Deokule, A.P., E-mail: abhijit.deokule1986@gmail.com [Homi Bhabha National Institute, Trombay 400 085, Mumbai (India); Vishnoi, A.K.; Dasgupta, A.; Umasankari, K.; Chandraker, D.K.; Vijayan, P.K. [Bhabha Atomic Research Centre, Trombay 400 085, Mumbai (India)

    2014-09-15

    Highlights: • Development of 19 pin annular fuel rod cluster. • Reactor physics study of designed annular fuel rod cluster. • Thermal hydraulic study of annular fuel rod cluster. - Abstract: An assessment of 33 pin annular fuel rod cluster has been carried out previously for possible use in a pressure tube type boiling water reactor. Despite the benefits such as negative coolant void reactivity and larger heat transfer area, the 33 pin annular fuel rod cluster is having lower discharge burn up as compared to solid fuel rod cluster when all other parameters are kept the same. The power rating of this design cannot be increased beyond 20% of the corresponding solid fuel rod cluster. The limitation on the power is not due to physics parameters rather it comes from the thermal hydraulics side. In order to increase power rating of the annular fuel cluster, keeping same pressure tube diameter, the pin diameter was increased, achieving larger inside flow area. However, this reduces the number of annular fuel rods. In spite of this, the power of the annular fuel cluster can be increased by 30% compared to the solid fuel rod cluster. This makes the nineteen pin annular fuel rod cluster a suitable option to extract more power without any major changes in the existing design of the fuel. In the present study reactor physics and thermal hydraulic analysis carried out with different annular fuel rod cluster geometry is reported in detail.

  18. Future CANDU nuclear power plant design requirements document executive summary

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Usmani, S.A. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    The future CANDU Requirements Document (FCRED) describes a clear and complete statement of utility requirements for the next generation of CANDU nuclear power plants including those in Korea. The requirements are based on proven technology of PHWR experience and are intended to be consistent with those specified in the current international requirement documents. Furthermore, these integrated set of design requirements, incorporate utility input to the extent currently available and assure a simple, robust and more forgiving design that enhances the performance and safety. The FCRED addresses the entire plant, including the nuclear steam supply system and the balance of the plant, up to the interface with the utility grid at the distribution side of the circuit breakers which connect the switchyard to the transmission lines. Requirements for processing of low level radioactive waste at the plant site and spent fuel storage requirements are included in the FCRED. Off-site waste disposal is beyond the scope of the FCRED. 2 tabs., 1 fig. (Author) .new.

  19. Shock Tube Investigation of Pressure and Ion Sensors Used in Pulse Detonation Engine Research

    Science.gov (United States)

    2004-06-01

    is a gas which follows the equation RTP ρ= and is generally applied to gases at low temperatures and pressures ( Cengel and Boles, 2002:88). A non...ideal or real gas does not follow this equation at sufficiently high temperature or pressure ( Cengel and Boles, 2002:622). This is the case when the

  20. Reliable experimental setup to test the pressure modulation of Baerveldt Implant tubes for reducing post-operative hypotony

    Science.gov (United States)

    Ramani, Ajay

    Glaucoma encompasses a group of conditions that result in damage to the optic nerve and can cause loss of vision and blindness. The nerve is damaged due to an increase in the eye's internal (intraocular) pressure (IOP) above the nominal range of 15 -- 20 mm Hg. There are many treatments available for this group of diseases depending on the complexity and stage of nerve degradation. In extreme cases where drugs or laser surgery do not create better conditions for the patient, ophthalmologists use glaucoma drainage devices to help alleviate the IOP. Many drainage implants have been developed over the years and are in use; but two popular implants are the Baerveldt Glaucoma Implant and the Ahmed Glaucoma Valve Implant. Baerveldt Implants are non-valved and provide low initial resistance to outflow of fluid, resulting in post-operative complications such as hypotony, where the IOP drops below 5 mm of Hg. Ahmed Glaucoma Valve Implants are valved implants which initially restrict the amount of fluid flowing out of the eye. The long term success rates of Baerveldt Implants surpass those of Ahmed Valve Implants because of post-surgical issues; but Baerveldt Implants' initial effectiveness is poor without proper flow restriction. This drives the need to develop new ways to improve the initial effectiveness of Baerveldt Implants. A possible solution proposed by our research team is to place an insert in the Baerveldt Implant tube of inner diameter 305 microns. The insert must be designed to provide flow resistance for the early time frame [e.g., first 30 -- 60 post-operative days] until sufficient scar tissue has formed on the implant. After that initial stage with the insert, the scar tissue will provide the necessary flow resistance to maintain the IOP above 5 mm Hg. The main objective of this project was to develop and validate an experimental apparatus to measure pressure drop across a Baerveldt Implant tube, with and without inserts. This setup will be used in the

  1. Study of a high-temperature and high-pressure FBG sensor with Al2O3 thin-wall tube substrate

    Institute of Scientific and Technical Information of China (English)

    ZHOU Hong; QIAO Xue-guang; WANG Hong-liang; FENG De-quan; WANG Wei

    2008-01-01

    A fiber Bragg grating (FBG) high-temperature and high pressure sensor has been designed and fabricated by using the Al2O3 thin-wall tube as a substrate. The test results show that the sensor can withstand a pressure range of 0-45 MPa and a temperature range of-10-300℃, and has a pressure sensitivity of 0.0426 nm/MPa and a temperature sensitivity of 0.0112nm/℃

  2. Pressure wave propagation in fluid-filled co-axial elastic tubes. Part 2: Mechanisms for the pathogenesis of syringomyelia.

    Science.gov (United States)

    Carpenter, P W; Berkouk, K; Lucey, A D

    2003-12-01

    Our aim in this paper is to use a simple theoretical model of the intraspinal cerebrospinal-fluid system to investigate mechanisms proposed for the pathogenesis of syringomyelia. The model is based on an inviscid theory for the propagation of pressure waves in co-axial, fluid-filled, elastic tubes. According to this model, the leading edge of a pressure pulse tends to steepen and form an elastic jump, as it propagates up the intraspinal cerebrospinal-fluid system. We show that when an elastic jump is incident on a stenosis of the spinal subarachnoid space, it reflects to form a transient, localized region of high pressure within the spinal cord that for a cough-induced pulse is estimated to be 50 to 70 mm Hg or more above the normal level in the spinal subarachnoid space. We propose this as a new mechanism whereby pressure pulses created by coughing or sneezing can generate syrinxes. We also use the same analysis to investigate Williams' suck mechanism. Our results do not support his concept, nor, in cases where the stenosis is severe, the differential-pressure-propagation mechanism recently proposed by Greitz et al. Our analysis does provide some support for the piston mechanism recently proposed by Oldfield et al. and Heiss et al. For instance, it shows clearly how the spinal cord is compressed by the formation of elastic jumps over part of the cardiac cycle. What appears to be absent for this piston mechanism is any means whereby the elastic jumps can be focused (e.g., by reflecting from a stenosis) to form a transient, localized region of high pressure within the spinal cord. Thus it would seem to offer a mechanism for syrinx progression, but not for its formation.

  3. Shock-tube calibration of a fast-response pressure transducer

    Science.gov (United States)

    Chung, Kung-Ming; Lu, Frank K.

    1990-01-01

    The sensitivity of a miniature fast-response piezoresistive pressure transducer determined dynamically was found to be slightly higher than that determined statically. Thus, mean pressures in a turbulent or unsteady flowfield that are measured using statically-calibrated pressure transducers would be slightly above true values. Unsteady pressure measurements to obtain space-time correlations and spectra can, however, be properly performed if the slight error is acceptable. These measurements are, obviously, subjected to limitations imposed by the bandwidth and the spatial resolution of the transducer. The noise spectrum revealed that the noise is predominantly above the transducer's resonant frequency. Filtering to improve the signal-to-noise ratio is particularly necessary when using the transducers at their low range. Transducer drift increases the signal-to-noise ratio and can adversely affect mean measurements.

  4. 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

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi [Japan Atomic Power Company, Tokyo (Japan)

    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)

  5. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy Of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2005-03-15

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. (author)

  6. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)

  7. CURRENT STATUS OF INTEGRITY ASSESSMENT BY SIPPING SYSTEM OF SPENT FUEL BUNDLES IRRADIATED IN CANDU REACTOR

    Directory of Open Access Journals (Sweden)

    JONG-YOUL PARK

    2014-12-01

    Full Text Available In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

  8. Numerical Study on the Heat Transfer of Carbon Dioxide in Horizontal Straight Tubes under Supercritical Pressure.

    Science.gov (United States)

    Yang, Mei

    2016-01-01

    Cooling heat transfer of supercritical CO2 in horizontal straight tubes with wall is numerically investigated by using FLUENT. The results show that almost all models are able to present the trend of heat transfer qualitatively, and the stand k-ε with enhanced wall treatment model shows the best agreement with the experimental data, followed by LB low Re turbulence model. Then further studies are discussed on velocity, temperature and turbulence distributions. The parameters which are defined as the criterion of buoyancy effect on convection heat transfer are introduced to judge the condition of the fluid. The relationships among the inlet temperature, outlet temperature, the mass flow rate, the heat flux and the diameter are discussed and the difference between the cooling and heating of CO2 are compared.

  9. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.K.; Snell, V. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca; West, J. [Candesco, Toronto, Ontario (Canada); Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. AECL initiated pre-licensing reviews of the ACR reactor design in Canada, US and China, with an objective to take into account regulatory feedback early in the design process. The Canadian Nuclear Safety Commission (CNSC) is performing a pre-project pre-licensing assessment of the ACR design. The objective of the assessment is to issue a formal statement as to whether there are any fundamental barriers that would prevent the licensing of the new CANDU reactor design in Canada under the Nuclear Safety and Control Act. The CNSC review is being conducted in four phases. In Phase 1 (September 2003 to September 2004) CNSC performed a pre-licensing review of the ACR-700, and focused on the design process, methodology, design concepts and R and D. CNSC staff reviewed about 100 reports, and submitted to AECL questions and comments. In Phase 2 (September 2004 to August 2005) AECL provided responses and additional information to CNSC on their comments and questions in Phase 1. Phase 3 is the Transition Phase (September 2005 to May 2006), bridging the transition from the ACR-700 to the ACR-1000 design. Phase 3 focused on review of generic aspects of the ACR design, on the Safety

  10. Operating Experience of MACSTOR Modules at CANDU 6 Stations

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, Robert R. [Atomic Energy Canada Ltd., Chalk River (Canada)

    2005-11-15

    Over the last three decades, Atomic Energy of Canada Limited (AECL) has contributed to the technology development and implementation of dry spent fuel management facilities in Canada, Korea and Romania During that period, AECL has developed a number of concrete canister models and the MACSTOR200 module, a medium size air-cooled vault with a 228 MgU (Mega grams of Uranium) capacity. AECL's dry storage technologies were used for the construction of eight large-scale above ground dry storage facilities for CANDU spent fuel. As of 2005, those facilities have an installed capacity in excess of 5,000 MgU. Since 1995, the two newest dry storage installations built for CANDU 6 reactors at Gentilly 2 (Canada) and Cernavoda (Romania) used the MACSTOR 200 module. Seven such modules have been built at Gentilly 2 during the 1995 to 2004 period and one at Cernavoda in 2003. The construction and operating experience of those modules is reviewed in this paper. The MACSTOR 200 modules were initially designed for a 50-year service life, with recent units at Gentilly 2 licensed for a 100-year service life in a rural (non-maritime) climate. During the 1995-2005 period, six of the eight modules were loaded with fuel. Their operation has brought a significant amount of experience on loading operations, performance of fuel handling equipment, radiation shielding, heat transfer, monitoring of the two confinement boundaries and radiation dose to personnel. Heat dissipation performance of the MACSTOR 200 was initially licensed using values derived from full scale tests made at AECL's Whiteshell Research Laboratories, that were backed-up by temperature measurements made on the first two modules. Results and computer models developed for the MACSTOR 200 module are described. Korea Hydro and Nuclear Power (KHNP) and its subsidiary Nuclear Environment Technology Institute (NETEC), in collaboration with Hyundai Engineering Company Ltd. (HEC) and AECL, are developing a new dry storage

  11. Proceedings of the fifth international conference on CANDU fuel. V.1,2

    Energy Technology Data Exchange (ETDEWEB)

    Lau, J.H. [ed.

    1997-07-01

    The First International Conference on CANDU Fuel was held in Chalk River in 1986. The CANDU Fuel community has gathered every three years since. The papers presented include topics on international experience, CANFLEX fuel bundles, Fuel design, Fuel modelling, Manufacturing and Quality assurance, Fuel performance and Safety, Fuel cycles and Spent Fuel management. Volume One was published in advance of the conference and Volume Two was printed after the conference.

  12. An empirical investigation on thermal characteristics and pressure drop of Ag-oil nanofluid in concentric annular tube

    Science.gov (United States)

    Abbasian Arani, A. A.; Aberoumand, H.; Aberoumand, S.; Jafari Moghaddam, A.; Dastanian, M.

    2016-08-01

    In this work an experimental study on Silver-oil nanofluid was carried out in order to present the laminar convective heat transfer coefficient and friction factor in a concentric annulus with constant heat flux boundary condition. Silver-oil nanofluid prepared by Electrical Explosion of Wire technique with no nanoparticles agglomeration during nanofluid preparation process and experiments. The average sizes of particles were 20 nm. Nanofluids with various particle Volume fractions of 0.011, 0.044 and 0.171 vol% were employed. The nanofluid flowing between the tubes is heated by an electrical heating coil wrapped around it. The effects of different parameters such as flow Reynolds number, tube diameter ratio and nanofluid particle concentration on heat transfer coefficient are studied. Results show that, heat transfer coefficient increased by using nanofluid instead of pure oil. Maximum enhancement of heat transfer coefficient occurs in 0.171 vol%. In addition the results showed that, there are slight increases in pressure drop of nanofluid by increasing the nanoparticle concentration of nanofluid in compared to pure oil.

  13. Experimental investigation of heat transfer and pressure drop of turbulent flow inside tube with inserted helical coils

    Science.gov (United States)

    Sharafeldeen, M. A.; Berbish, N. S.; Moawed, M. A.; Ali, R. K.

    2016-08-01

    The heat transfer and pressure drop were experimentally investigated in a coiled wire inserted tube in turbulent flow regime in the range of Reynolds number of 14,400 ≤ Re ≤ 42,900. The present work aims to extend the experimental data available on wire coil inserts to cover wire diameter ratio of 0.044 ≤ e/d ≤ 0.133 and coil pitch ratio of 1 ≤ p/d ≤ 5. Uniform heat flux was applied to the external surface of the tube and air was selected as fluid. The effects of Reynolds number and wire diameter and coil pitch ratios on the Nusselt number and friction factor were studied. The enhancement efficiency and performance criteria ranges are of (46.9-82.6 %) and (100.1-128 %) within the investigated range of the different parameters, respectively. Correlations are obtained for the average Nusselt number and friction factor utilizing the present measurements within the investigated range of geometrical parameters and Re. The maximum deviation between correlated and experimental values for Nusselt number and friction factor are ±5 and ±6 %, respectively.

  14. Atmospheric pressure argon surface discharges propagated in long tubes: physical characterization and application to bio-decontamination

    Science.gov (United States)

    Kovalova, Zuzana; Leroy, Magali; Jacobs, Carolyn; Kirkpatrick, Michael J.; Machala, Zdenko; Lopes, Filipa; Laux, Christophe O.; DuBow, Michael S.; Odic, Emmanuel

    2015-11-01

    Pulsed corona discharges propagated in argon (or in argon with added water vapor) at atmospheric pressure on the interior surface of a 49 cm long quartz tube were investigated for the application of surface bio-decontamination. H2O molecule dissociation in the argon plasma generated reactive species (i.e. OH in ground and excited states) and UV emission, which both directly affected bacterial cells. In order to facilitate the evaluation of the contribution of UV radiation, a DNA damage repair defective bacterial strain, Escherichia coli DH-1, was used. Discharge characteristics, including propagation velocity and plasma temperature, were measured. Up to ~5.5 and ~5 log10 reductions were observed for E. coli DH-1 bacteria (from 106 initial load) exposed 2 cm and 44 cm away from the charged electrode, respectively, for a 20 min plasma treatment. The factors contributing to the observed bactericidal effect include desiccation, reactive oxygen species (OH) plus H2O2 accumulation in the liquid phase, and UV-B (and possibly VUV) emission in dry argon. The steady state temperature measured on the quartz tube wall did not exceeded 29 °C the contribution of heating, along with that of H2O2 accumulation, was estimated to be low. The effect of UV-B emission alone or in combination with the other stress factors of the plasma process was examined for different operating conditions.

  15. HeatTransfer Coefficients and Pressure Drops of The Finned Tube Heat Exchangers with Small Diameter Pipes

    Science.gov (United States)

    Tanaka, Hiroyoshi; Aoyama, Shigeo; Koma, Hachirou; Adachi, Masaaki

    In order to enhance the heat transfer coefficient of the fin used in the finned tube heat exchanger, newly designed fin surfaces, especially, with small diameter (≅4mm) pipes are developed. The experiments are made by the transient testing technique, and used the plastic fins scaling up 4 times of the actual metal fin size. The data of the heat transfer coefficient and the pressure drop are transformed to the actual metal fin data. The fin with the anomalous staggered pipe arrangement and the bridge-like cutting-out with inclined leg portion from stream line is found to have very high overall heat transfer coefficient which is about 1.8-fold increase in comparison with the conventional Louvered fin. In this paper the reason why such enhancement is caused is clarified by mean of the calculation based on the rectangular duct flow. The calculated values are coincident with the data of the experiment well.

  16. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 6 - PRESENTATION OF THE DECOMMISSIONING DEVICE

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

    dismantling and extraction of the channel closure plug and shield plug, extraction of the end fitting, cutting and extraction of the pressure tube. The fuel channel decommissioning device consists of following major components: coupling and locking fuel channel module, assembly valve for access to the fuel channel, storage tubes assembly for extracted components, handling elements assembly, cutting and extraction device and housing device. The design of the device and platform support is achieved according to the particular features of the fuel channel components to be dismantled in the program of nuclear reactor decommissioning according to all the safety aspects and environmental protection during the activities, resulting from the decommissioning plan developed.

  17. Exergoeconomic optimization of coaxial tube evaporators for cooling of high pressure gaseous hydrogen during vehicle fuelling

    DEFF Research Database (Denmark)

    Jensen, Jonas Kjær; Rothuizen, Erasmus Damgaard; Markussen, Wiebke Brix

    2014-01-01

    -stage evaporator. The main contribution to the total cost was the cost associated with exergy destruction, the capital investment cost contributed with 5-14 %. The main contribution to the exergy destruction was found to be thermally driven. The pressure driven exergy destruction accounted for 3-9 %....

  18. An experimental investigation of pressure drop of aqueous foam in laminar tube flow

    Science.gov (United States)

    Blackwell, B. F.; Sobolik, K. B.

    1987-04-01

    This report is the first of two detailing pressure-drop and heat-transfer measurements made at the Foam Flow Heat Transfer Loop. The work was motivated by a desire to extend the application of aqueous foam from petroleum drilling to geothermal drilling. Pressure-drop measurements are detailed in this report; a forthcoming report (SAND85-1922) will describe the heat-transfer measurements. The pressure change across a 2.4-m (8-ft) length of the 2.588-cm (1.019-in.) ID test section was measured for liquid volume fractions between 0.05 and 0.35 and average velocities between 0.12 and 0.80 m/s (0.4 and 2.6 ft/s). The resulting pressure-drop/flow-rate data were correlated to a theoretical model for a Bingham plastic. Simple expressions for the dynamic viscosity and the yield stress as a function of liquid volume fraction were estimated.

  19. Measurements of Speed of Sound in Lean and Rich Natural Gas Mixtures at Pressures up to 37 MPa Using a Specialized Rupture Tube

    Science.gov (United States)

    Botros, K. K.

    2010-12-01

    Measurements of the speed of sound in 42 different compositions of lean, medium, and rich natural-gas mixtures using a specialized high-pressure rupture tube have been conducted. The rupture tube is made of stainless steel (internal diameter = 38.1 mm and length = 42 m), and is instrumented with 13 high-frequency-response dynamic pressure transducers (Endevco) mounted very close to the rupture end and along the length of the tube to capture the pressure-time traces of the decompression wave. Tests were conducted for initial pressures ranging from 10 MPa to 37 MPa and a temperature range from -25°C to+68°C. Gas mixture compositions were controlled by mixing conventional natural-gas mixtures from an adjacent gas pipeline with richer components of alkanes. Temperature control is achieved by a heat tracer along the tube with a set point at the desired gas temperature of the particular test. Uncertainty analysis indicated that the uncertainty in the experimentally determined speed of sound in the undisturbed gas mixture at the initial pressure and temperature is on the order of 0.306 %. The measured speeds of sound were compared to predictions by five equations of state, namely; the Benedict-Webb-Rubin-Starling (BWRS), AGA-8, Peng-Robinson (PR), Redlich-Kwong-Soave (RK-Soave), and Groupe Européen de Recherches Gaziéres (GERG-2004) equations.

  20. Pressure dependence and branching ratios in the decomposition of 1-pentyl radicals: shock tube experiments and master equation modeling.

    Science.gov (United States)

    Awan, Iftikhar A; Burgess, Donald R; Manion, Jeffrey A

    2012-03-22

    The decomposition and intramolecular H-transfer isomerization reactions of the 1-pentyl radical have been studied at temperatures of 880 to 1055 K and pressures of 80 to 680 kPa using the single pulse shock tube technique and additionally investigated with quantum chemical methods. The 1-pentyl radical was generated by shock heating dilute mixtures of 1-iodopentane and the stable products of its decomposition have been observed by postshock gas chromatographic analysis. Ethene and propene are the main olefin products and account for >97% of the carbon balance from 1-pentyl. Also produced are very small amounts of (E)-2-pentene, (Z)-2-pentene, and 1-butene. The ethene/propene product ratio is pressure dependent and varies from about 3 to 5 over the range of temperatures and pressures studied. Formation of ethene and propene can be related to the concentrations of 1-pentyl and 2-pentyl radicals in the system and the relative rates of five-center intramolecular H-transfer reactions and β C-C bond scissions. The 3-pentyl radical, formed via a four-center intramolecular H transfer, leads to 1-butene and plays only a very minor role in the system. The observed (E/Z)-2-pentenes can arise from a small amount of beta C-H bond scission in the 2-pentyl radical. The current experimental and computational results are considered in conjunction with relevant literature data from lower temperatures to develop a consistent kinetics model that reproduces the observed branching ratios and pressure effects. The present experimental results provide the first available data on the pressure dependence of the olefin product branching ratio for alkyl radical decomposition at high temperatures and require a value of = (675 ± 100) cm(-1) for the average energy transferred in deactivating collisions in an argon bath gas when an exponential-down model is employed. High pressure rate expressions for the relevant H-transfer reactions and β bond scissions are derived and a Rice Ramsberger

  1. A Shock-Tube Study of the CO + OH Reaction Near the Low-Pressure Limit.

    Science.gov (United States)

    Nasir, Ehson F; Farooq, Aamir

    2016-06-09

    Rate coefficients for the reaction between carbon monoxide and hydroxyl radical were measured behind reflected shock waves over 700-1230 K and 1.2-9.8 bar. The temperature/pressure conditions correspond to the predicted low-pressure limit of this reaction, where the channel leading to carbon dioxide formation is dominant. The reaction rate coefficients were inferred by measuring the formation of carbon dioxide using quantum cascade laser absorption near 4.2 μm. Experiments were performed under pseudo-first-order conditions with tert-butyl hydroperoxide (TBHP) as the OH precursor. Using ultraviolet laser absorption by OH radicals, the TBHP decomposition rate was measured to quantify potential facility effects under extremely dilute conditions used here. The measured CO + OH rate coefficients are provided in Arrhenius form for three different pressure ranges: kCO+OH(1.2-1.6 bar) = (9.14 ± 2.17) × 10(-13) exp(-(1265 ± 190)/T) cm(3) molecule(-1) s(-1); kCO+OH(4.3-5.1 bar) = (8.70 ± 0.84) × 10(-13) exp(-(1156 ± 83)/T) cm(3) molecule(-1) s(-1); and kCO+OH(9.6-9.8 bar) = (7.48 ± 1.92) × 10(-13) exp(-(929 ± 192)/T) cm(3) molecule(-1) s(-1). The measured rate coefficients are found to be lower than the master equation modeling results by Weston et al. [J. Phys. Chem. A, 2013, 117, 821] at 819 K and in closer agreement with the expression provided by Joshi and Wang [Int. J. Chem. Kinet., 2006, 38, 57].

  2. A Shock Tube Study of the CO + OH Reaction Near the Low-Pressure Limit

    KAUST Repository

    Nasir, Ehson Fawad

    2016-05-16

    Rate coefficients for the reaction between carbon monoxide and hydroxyl radical were measured behind reflected shock waves over 700 – 1230 K and 1.2 – 9.8 bar. The temperature/pressure conditions correspond to the predicted low-pressure limit of this reaction, where the channel leading to carbon dioxide formation is dominant. The reaction rate coefficients were inferred by measuring the formation of carbon dioxide using quantum cascade laser absorption near 4.2 µm. Experiments were performed under pseudo-first order conditions with tert-butyl hydroperoxide (TBHP) as the OH precursor. Using ultraviolet laser absorption by OH radicals, the TBHP decomposition rate was measured to quantify potential facility effects under extremely dilute conditions used here. The measured CO + OH rate coefficients are provided in Arrhenius form for three different pressure ranges: kCO+OH (1.2 – 1.6 bar) = 9.14 x 10-13 exp(-1265/T) cm3 molecule-1 s-1 kCO+OH (4.3 – 5.1 bar) = 8.70 x 10-13 exp(-1156/T) cm3 molecule-1 s-1 kCO+OH (9.6 – 9.8 bar) = 7.48 x 10-13 exp(-929/T) cm3 molecule-1 s-1 The measured rate coefficients are found to be lower than the master equation modeling results by Weston et al. [J. Phys. Chem. A, 117 (2013) 821] at 819 K and in closer agreement with the expression provided by Joshi and Wang [Int. J. Chem. Kinet., 38 (2006) 57].

  3. Numerical Simulations of Pressure Spikes within a Cylindrical Launch Tube due to a Bursting Helium Flask

    Science.gov (United States)

    2011-11-09

    Above 50 atm pressure, the system deviates from ideal gas law behavior as shown by the red curve. This curve was obtained using the Cheetah 6.0...equation of state for Helium6 and calculating 1 1 2 2PV PV directly from the 8 Harold D. Ladouceur and Benjamin Gould Cheetah output. Note...that the Cheetah code utilizes the ideal gas law to calculate A as indicated by the black dots in Figure 6. A key point of this figure is the

  4. Heat transfer enhancement and pressure drop analysis in a helically coiled tube using Al{sub 2}O{sub 3} / water nanofluid

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, P. C. Mukesh; Tamilarasan, R.; Nathan, S. Sendhil [University College of Engineering Pattukkottai, Rajamadam (India); Kumar, J. [Sasurie College of Engineering, Tiruppur (India); Suresh, S. [National Institute of Technology, Tiruchirappalli (India)

    2014-05-15

    In this experimental investigation, the heat transfer and pressure drop analysis of a shell and helically coiled tube heat exchanger by using Al{sub 2}O{sub 3} / water nanofluids have been carried out under turbulent flow condition. The Al{sub 2}O{sub 3} / water nanofluids of 0.1%, 0.4%, and 0.8% particle volume concentration have been prepared by using two step method. The tube side experimental Nusselt number of 0.1%, 0.4% and 0.8% nanofluids were found to be 28%, 36% and 56%, respectively higher than water. These enhancements are due to higher thermal conductivity of nanofluid, better fluid mixing and strong secondary flow formation in coiled tube. The pressure drop of 0.1%, 0.4% and 0.8% were found to be 4%, 6%, and 9%, respectively higher than water. The increase in pressure drop is due to increase in nanofluid viscosity while adding nanoparticles. The measurement of nanofluid thermal performance factor is found to be greater than unity. It is concluded that the Al{sub 2}O{sub 3} nanofluid can be applied as a coolant in helically coiled tube heat exchanger to enhance heat transfer with negligible pressure drop.

  5. Bubble-assisted film evaporation correlation for saline water at sub-atmospheric pressures in horizontal-tube evaporator

    KAUST Repository

    Shahzad, Muhammad Wakil

    2013-01-01

    In falling film evaporators, the overall heat transfer coefficient is controlled by film thickness, velocity, liquid properties and the temperature differential across the film layer. This article presents the heat transfer behavior for evaporative film boiling on horizontal tubes, but working at low pressures of 0.93-3.60 kPa (corresponding solution saturation temperatures of 279-300 K) as well as seawater salinity of 15,000 to 90,000 mg/l or ppm. Owing to a dearth of literature on film-boiling at these conditions, the article is motivated by the importance of evaporative film boiling in the desalination processes such as the multi-effect distillation (MED) or multi-stage flashing (MSF): It is observed that in addition to the above-mentioned parameters, evaporative heat transfer of seawater is affected by the emergence of micro-bubbles within the thin film layer, particularly when the liquid saturation temperatures drop below 298 K (3.1 kPa). Such micro bubbles are generated near to the tube wall surfaces and they enhanced the heat transfer by two or more folds when compared with the predictions of conventional evaporative film boiling. The appearance of micro-bubbles is attributed to the rapid increase in the specific volume of vapor, i.e., dv/dT, at low saturation temperature conditions. A new correlation is thus proposed in this article and it shows good agreement to the measured data with an experimental uncertainty of 8% and regression RMSE of 3.5%. © 2012 Elsevier Ltd. All rights reserved.

  6. FLUID-STRUCTURE INTERACTION IN A U-TUBE WITH SURFACE ROUGHNESS AND PRESSURE DROP

    Directory of Open Access Journals (Sweden)

    GYUN-HO GIM

    2014-10-01

    Full Text Available In this research, the surface roughness affecting the pressure drop in a pipe used as the steam generator of a PWR was studied. Based on the CFD (Computational Fluid Dynamics technique using a commercial code named ANSYS-FLUENT, a straight pipe was modeled to obtain the Darcy frictional coefficient, changed with a range of various surface roughness ratios as well as Reynolds numbers. The result is validated by the comparison with a Moody chart to set the appropriate size of grids at the wall for the correct consideration of surface roughness. The pressure drop in a full-scale U-shaped pipe is measured with the same code, correlated with the surface roughness ratio. In the next stage, we studied a reduced scale model of a U-shaped heat pipe with experiment and analysis of the investigation into fluid-structure interaction (FSI. The material of the pipe was cut from the real heat pipe of a material named Inconel 690 alloy, now used in steam generators. The accelerations at the fixed stations on the outer surface of the pipe model are measured in the series of time history, and Fourier transformed to the frequency domain. The natural frequency of three leading modes were traced from the FFT data, and compared with the result of a numerical analysis for unsteady, incompressible flow. The corresponding mode shapes and maximum displacement are obtained numerically from the FSI simulation with the coupling of the commercial codes, ANSYS-FLUENT and TRANSIENT_STRUCTURAL. The primary frequencies for the model system consist of three parts: structural vibration, BPF(blade pass frequency of pump, and fluid-structure interaction.

  7. Integrated evolution of the medium power CANDU{sup MD} reactors; Evolution integree des reacteurs CANDU{sup MD} de moyenne puissance

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F. [AECL Accelerators, Kanata, ON (Canada)

    2002-07-01

    The aim of this document is the main improvements of the CANDU reactors in the economic, safety and performance domains. The presentation proposes also other applications as the hydrogen production, the freshening of water sea and the bituminous sands exploitation. (A.L.B.)

  8. Remote field Eddy Current Technique Development for Gap Measurement of Neighboring Tubes of Nuclear Fuel Channel in Pressurized Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, H. K.; Lee, D. H.; Lee, Y. S.; Huh, H.; Cheong, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2004-04-15

    Liquid Injection Nozzle(LIN) tube and Calandria tube(CT) in pressurized Heavy Water Reactor (PHWR) are cross-aligned horizontally. These neighboring tubes can contact each other due to the sag of the calandria tube resulting from the irradiation creep and thermal creep, and fuel load, etc. In order to judge the contact which might be the safety concern, the remote field eddy current (RFEC) technology is applied for the gap measurement in this paper. LIN can be detected by inserting the RFEC probe into pressure tube (PT) at the crossing point directly. To obtain the optimal conditions of the RFEC inspection, the sensitivity, penetration and noise signals are considered simultaneously. The optimal frequency and coil spacing are 1kHz and 200mm respectively. Possible noises during LIN signal acquisition are caused by lift-off, PT thickness variation, and gap variation between PT and CT. The simulated noise signals were investigated by the Volume Integral Method(VIM). Signal analysis on the voltage plane describes the amplitude and shape of LIN and possible defects at several frequencies. All the RFEC measurements in the laboratory were done in variance with the CT/LIN gap and showed the relationship between the LIN gap and the signal parameters by analyzing the voltage plane signals

  9. Proof of Concept of Crack Localization Using Negative Pressure Waves in Closed Tubes for Later Application in Effective SHM System for Additive Manufactured Components

    Directory of Open Access Journals (Sweden)

    Michaël F. Hinderdael

    2016-01-01

    Full Text Available Additive manufactured components have a different metallurgic structure and are more prone to fatigue cracks than conventionally produced metals. In earlier papers, an effective Structural Health Monitoring solution was presented to detect fatigue cracks in additive manufactured components. Small subsurface capillaries are embedded in the structure and pressurized (vacuum or overpressure. A crack that initiated at the component’s surface will propagate towards the capillary and finally breach it. One capillary suffices to inspect a large area of the component, which makes it interesting to locate the crack on the basis of the pressure measurements. Negative pressure waves (NPW arise from the abrupt encounter of high pressure fluid with low pressure fluid and can serve as a basis to locate the crack. A test set-up with a controllable leak valve was built to investigate the feasibility of using NPW to localize a leak in closed tubes with small lengths. Reflections are expected to occur at the ends of the tube, possibly limiting the localization accuracy. In this paper, the results of the tests on the test set-up are reported. It will be shown that the crack could be localized with high accuracy (millimeter accuracy which proves the concept of crack localization on basis of NPW in a closed tube of small length.

  10. Ear Tubes

    Science.gov (United States)

    ... of the ear drum or eustachian tube, Down Syndrome, cleft palate, and barotrauma (injury to the middle ear caused by a reduction of air pressure, ... specialist) may be warranted if you or your child has experienced repeated ... fluid in the middle ear, barotrauma, or have an anatomic abnormality that ...

  11. Heat transfer and pressure drop of surfactant solutions at crossflown finned helical tubes. Waermeuebergang und Druckverlust waessriger Tensidloesungen an einer querangestroemten berippten Rohrwendel

    Energy Technology Data Exchange (ETDEWEB)

    Weber, M. (Huels AG, Marl (Germany)); Kleuker, H.H.; Steiff, A.; Weinspach, P.M. (Dortmund Univ. (Germany). Lehrstuhl fuer Thermische Verfahrenstechnik)

    1992-09-01

    The addition of suitable drag reducers to water in district heating networks either reduces the pressure drop significantly or the electrical power consumption of the conveying pump can be reduced at the same flow rate. New surfactant additive systems accomplish the requirements on the thermal and mechanical capacitance of district heating systems. One of the important aspects for the application of surfactant solutions is the influence on the heat transfer in the installed heat exchangers in district heating networks. In earlier publications heat transfer and pressure drop of surfactant solutions in straight pipes and in helical tubes have been discussed. Developing from the scientific findings heat transfer and pressure drop at crossflown finned helical tubes were investigated in this work. The main purpose is the presentation of the occurred effects of drag reducing solutions. Due to the complex flow conditions a prediction could not be developed as yet. (orig.).

  12. 针翅管传热与压降特性研究%Research on Heat Transfer and Pressure Drop Characteristics of Integral Pin-Fin Tube

    Institute of Scientific and Technical Information of China (English)

    石帅; 阎昌琪; 牛广林; 陈哲雨

    2012-01-01

    以润滑油为换热介质,对整体针翅管传热与阻力特性进行了理论分析与试验研究,研究结果可为针翅管的优化设计提供参考.在换热介质纵向冲刷换热管的条件下,对不同针翅长度的3种整体针翅管与光管进行了传热与阻力试验.结果表明:整体针翅管对润滑油换热具有很好的强化能力,在本试验范围内,整体针翅管对油流体扰动强烈,换热强度是同条件下光管的2~6倍;针翅长度是影响针翅管压降的主要因素,在雷诺数达300时,压降曲线出现转折.%Taking lubricating oil as the heat transfer medium, heat transfer and pressure drop characteristics of integral pin-fin tubes were researched both in terms of theoretical and experiments. The results can provide a reference for pin-fin tube optimization. Under the condition of heat transfer medium longitudinally flushing the heat exchanger tube surface, the heat transfer and resistance experiments of plain tube and integral pin-fin tubes (in three different length) were carried out in the present work. The results show that the integral pin-fin tubes can improve the heat transfer ability of lubricating oil. The oil flow fluctuation caused by integral pin-fin tubes is intense and the heat transfer intension of integral pin-fin tubes is 2-6 times of plain tube at the same experiment condition. The length of fin makes main influence on pressure drop, and the pressure drop curve turns around when the values of Reynolds number reaches to 300.

  13. Effect of tube size on electromagnetic tube bulging

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    The commercial finite code ANSYS was employed for the simulation of the electromagnetic tube bulging process. The finite element model and boundary conditions were thoroughly discussed. ANSYS/EMAG was used to model the time varying electromagnetic field in order to obtain the radial and axial magnetic pressure acting on the tube. The magnetic pressure was then used as boundary conditions to model the high velocity deformation of various length tube with ANSYS/LSDYNA. The time space distribution of magnetic pressure on various length tubes was presented. Effect of tube size on the distribution of radial magnetic pressure and axial magnetic pressure and high velocity deformation were discussed. According to the radial magnetic pressure ratio of tube end to tube center and corresponding dimensionless length ratio of tube to coil, the free electromagnetic tube bulging was studied in classification. The calculated results show good agreements with practice.

  14. An Effective Approach for Coupling Direct Analysis in Real Time with Atmospheric Pressure Drift Tube Ion Mobility Spectrometry

    Science.gov (United States)

    Keelor, Joel D.; Dwivedi, Prabha; Fernández, Facundo M.

    2014-09-01

    Drift tube ion mobility spectrometry (DTIMS) has evolved as a robust analytical platform routinely used for screening small molecules across a broad suite of chemistries ranging from food and pharmaceuticals to explosives and environmental toxins. Most modern atmospheric pressure IM detectors employ corona discharge, photoionization, radioactive, or electrospray ion sources for efficient ion production. Coupling standalone DTIMS with ambient plasma-based techniques, however, has proven to be an exceptional challenge. Device sensitivity with near-ground ambient plasma sources is hindered by poor ion transmission at the source-instrument interface, where ion repulsion is caused by the strong electric field barrier of the high potential ion mobility spectrometry (IMS) inlet. To overcome this shortfall, we introduce a new ion source design incorporating a repeller point electrode used to shape the electric field profile and enable ion transmission from a direct analysis in real time (DART) plasma ion source. Parameter space characterization studies of the DART DTIMS setup were performed to ascertain the optimal configuration for the source assembly favoring ion transport. Preliminary system capabilities for the direct screening of solid pharmaceuticals are briefly demonstrated.

  15. Influence of Fe content on corrosion and hydrogen pick up behavior of Zr–2.5Nb pressure tube material

    Energy Technology Data Exchange (ETDEWEB)

    Choudhuri, Gargi, E-mail: gargi@barc.gov.in [Quality Assurance Division, BARC, Mumbai 400 085 (India); Jagannath [Theoretical Physics Division, BARC, Mumbai 400 085 (India); Kiran Kumar, M.; Kain, V.; Srivastava, D. [Material Science Division, BARC, Mumbai 400 085 (India); Basu, S. [Solid State Physics Division, BARC, Mumbai 400 085 (India); Shah, B.K. [Quality Assurance Division, BARC, Mumbai 400 085 (India); Saibaba, N. [Nuclear Fuel Complex, Hyderabad 500 062 (India); Dey, G.K. [Material Science Division, BARC, Mumbai 400 085 (India)

    2013-10-15

    The effects of Fe addition in the range of 300–1250 ppm in cold worked stress-relieved Zr–2.5Nb pressure tube on oxidation and hydrogen pick up behavior have been studied after 415 °C steam autoclaving. Microstructure and micro-chemistry of second phase and precipitates were characterized using electron microscope. Addition of 800 ppm Fe in Zr–2.5Nb alloy led to better oxidation resistance. With further addition of Fe no significant improvement of oxidation resistance was observed but hydrogen-pickup was found to increase. Zr–Nb–Fe bearing precipitates were observed in Zr–2.5Nb alloy containing 800 ppm Fe. Further addition of Fe led to formation of Zr–Fe intermetallic. The chemical state of oxide has been determined by X-ray photo electron spectroscopy. Grazing Incidence X-ray Diffraction revealed that oxide in alloys with higher Fe, contained a higher fraction of tetragonal-Zirconia which is indicative of a protective oxide film and hence better oxidation resistance of the alloy.

  16. Crack initiation at long radial hydrides in Zr-2. 5Nb pressure tube material at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Choubey, R.; Puls, M.P. (AECL Research, Pinawa, Manitoba (Canada). Whiteshell Labs.)

    1994-05-01

    Crack initiation at hydrides in smooth tensile specimens of Zr-2.5Nb pressure tube material was investigated at elevated temperatures up to 300 C using an acoustic emission (AE) technique. The test specimens contained long, radial hydride platelets. These hydrides have their plate normals oriented in the applied stress direction. Below [approximately]100 C, widespread hydride cracking was initiated at stresses close to the yield stress. An estimate of the hydride's fracture strength from this data yielded a value of [approximately]520 MPa at 100 C. Metallography showed that up to this temperature, cracking occurred along the length of the hydrides. However, at higher temperatures, there was no clear evidence of lengthwise cracking of hydrides, and fewer of the total hydride population fractured during deformation, as indicated by the AE record and the metallography. Moreover, the hydrides showed significant plasticity by-being able to flow along with the matrix material and align themselves parallel to the applied stress direction without fracturing. Near the fracture surface of the specimen, transverse cracking of the flow-reoriented hydrides had occurred at various points along the lengths of the hydrides. These fractures appear to be the result of stresses produced by large plastic strains imposed by the surrounding matrix on the less ductile hydrides.

  17. [Variations in the internal pressure of the pneumatic cuffs of endotracheal tubes according to their contents and the anesthetic mixtures used. Experimental study].

    Science.gov (United States)

    de Santos, P; Castillo, J; Bogdanovich, A; Nalda, M A

    1989-01-01

    With the purpose of measuring pressure changes in the pneumatic cuffs of endotracheal tubes when the composition of the mixture of gases used for ventilation had to change for the same content, we designed a model of artificial respiration that consisted of a tube with a low pressure pneumatic cuff measuring 8.5 mm in inner diameter introduced in a replica of a human trachea, adjusted to two anesthetic bags. The cuff valve was connected to a pressure transducer by a three-ended stopcock and, after aspiration of its content, it was inflated with air, saline or nitrous oxide and oxygen at 60% up to a basal pressure of 20 mmHg. The tube was connected to a respirator adjusted to inflate 10 l/min at a rate of 15 insufflations/min of: oxygen 100% for 5 minutes, then nitrous oxide and oxygen at 60% for 30 minutes and oxygen 100% again for 15 minutes. When inflating the pneumatic cuff with air and ventilating with nitrous oxide and oxygen at 60%, its pressure reached a maximum mean value of 58 mmHg (190% with respect to base values). When insufflating with saline and ventilating in the same conditions, pressure reached a maximum mean value of 33 mmHg (65% with respect to base values). When the pneumatic cuff was inflated with nitrous oxide and oxygen at 60%, important changes in pressure were observed when the characteristics of the inspired gases were modified. We conclude that some method for monitoring pneumatic cuff pressure should be systematized.

  18. Analyse du transfert de chaleur et de la perte de pression pour des ecoulements supercritiques dans le reacteur CANDU-SCWR

    Science.gov (United States)

    Zoghlami, Sarra

    The supercritical water reactor is one of the six concepts of generation IV nuclear reactors that has been selected by the International Generation IV Forum (GIF). Canada has chosen to conduct advanced research on this type of reactor. For the design and safety analysis of the reactor concept, the development of numerical simulation codes is needed. The ARTHUR code is a thermal-hydraulic computer code developed by Fassi-Fehri (2008), at the Ecole Polytechnique de Montreal, to analyse the CANDU-6 reactor. The purpose of this project is to modify this numerical code so that it can be used to treat the CANDU-SCWR. To calculate the coolant thermal-hydraulics properties in the fuel channel of a CANDU-SCWR, it was assumed that the water flows under supercritical conditions is a one-phase flow. Thus within this code, we developed the conservation equations for one-phase flow. Hydraulic resistance and heat transfer at supercritical pressure are two important aspects to be considered in the modeling of a fuel channel in a nuclear reactor. To choose the accurate correlation to predict the pressure friction factor, we compared numerical calculations, using different correlations found in literature, to experimental data. We concluded that the Garimella (2008) correlation is the most consistent, to be incorporated in the ARTHUR &barbelow;SCWR code. We proved that the choice of the friction factor correlation affects slightly the distribution of thermal-hydraulic properties in the fuel channel. Under supercritical conditions, water thermal-physical properties are characterized by significant variations in the pseudo-critical region. This behavior influences the forced convection heat transfer phenomena. To choose the adequate correlation to calculate the forced convection heat transfer coefficient, we compared numerical results to experimental data, and we found that the standard deviation given by Mokry et al. (2010) correlation is the lowest. In order to model the fuel

  19. Computer code applicability assessment for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wren, D.J.; Langman, V.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    AECL Technologies, the 100%-owned US subsidiary of Atomic Energy of Canada Ltd. (AECL), is currently the proponents of a pre-licensing review of the Advanced Candu Reactor (ACR) with the United States Nuclear Regulatory Commission (NRC). A key focus topic for this pre-application review is the NRC acceptance of the computer codes used in the safety analysis of the ACR. These codes have been developed and their predictions compared against experimental results over extended periods of time in Canada. These codes have also undergone formal validation in the 1990's. In support of this formal validation effort AECL has developed, implemented and currently maintains a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper discusses the SQA program used to develop, qualify and maintain the computer codes used in ACR safety analysis, including the current program underway to confirm the applicability of these computer codes for use in ACR safety analyses. (authors)

  20. Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

  1. Overall heat transfer coefficient and pressure drop in a typical tubular exchanger employing alumina nano-fluid as the tube side hot fluid

    Science.gov (United States)

    Kabeel, A. E.; Abdelgaied, Mohamed

    2016-08-01

    Nano-fluids are used to improve the heat transfer rates in heat exchangers, especially; the shell-and-tube heat exchanger that is considered one of the most important types of heat exchangers. In the present study, an experimental loop is constructed to study the thermal characteristics of the shell-and-tube heat exchanger; at different concentrations of Al2O3 nonmetallic particles (0.0, 2, 4, and 6 %). This material concentrations is by volume concentrations in pure water as a base fluid. The effects of nano-fluid concentrations on the performance of shell and tube heat exchanger have been conducted based on the overall heat transfer coefficient, the friction factor, the pressure drop in tube side, and the entropy generation rate. The experimental results show that; the highest heat transfer coefficient is obtained at a nano-fluid concentration of 4 % of the shell side. In shell side the maximum percentage increase in the overall heat transfer coefficient has reached 29.8 % for a nano-fluid concentration of 4 %, relative to the case of the base fluid (water) at the same tube side Reynolds number. However; in the tube side the maximum relative increase in pressure drop has recorded the values of 12, 28 and 48 % for a nano-material concentration of 2, 4 and 6 %, respectively, relative to the case without nano-fluid, at an approximate value of 56,000 for Reynolds number. The entropy generation reduces with increasing the nonmetallic particle volume fraction of the same flow rates. For increase the nonmetallic particle volume fraction from 0.0 to 6 % the rate of entropy generation decrease by 10 %.

  2. Free Piston Double Diaphragm Shock Tube

    OpenAIRE

    OGURA, Eiji; FUNABIKI, Katsushi; SATO, Shunichi; Abe, Takashi; 小倉, 栄二; 船曳, 勝之; 佐藤, 俊逸; 安部, 隆士

    1997-01-01

    A free piston double diaphragm shock tube was newly developed for generation of high Mach number shock wave. Its characteristics was investigated for various operation parameters; such as a strength of the diaphragm at the end of the comparession tube, an initial pressure of low pressure tube, an initial pressure of medium pressure tube and the volume of compression tube. Under the restriction of fixed pressures for the driver high pressure tube (32×10^5Pa) and the low pressure tube (40Pa) in...

  3. Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Wren, D.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd, (Canada)

    2004-07-01

    Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

  4. CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J. [Korea Atomic Energy Research Inst., Yusong, Taejon (Korea, Republic of)

    1997-07-01

    KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

  5. Safety and reliability of the sealing cuff pressure of the Microcuff pediatric tracheal tube for prevention of post-extubation morbidity in children: A comparative study

    Directory of Open Access Journals (Sweden)

    Roshdi Roshdi Al-Metwalli

    2014-01-01

    Full Text Available Objectives: The objective of this study is to evaluate the efficacy and safety of sealing pressure as an inflation technique of the Microcuff pediatric tracheal cuffed tube. Materials and Methods: A total of 60 children were enrolled in this study. After induction of anesthesia and intubation with Microcuff pediatric tracheal tube, patients were randomly assigned, to one of the three groups. Control group (n = 20 the cuff was inflated to a cuff pressure of 20 cm H 2 O; sealing group (n = 20 the cuff was inflated to prevent the air leak at peak airway pressure of 20 cm H 2 O and the finger group (n = 20 the cuff was inflated to a suitable pressure using the finger estimation. Tracheal leak, incidence and severity of post-extubation cough, stridor, sore throat and hoarseness were recorded. Results: The cuff pressure as well as the volume of air to fill the cuff was significantly low in the sealing group when compared with the control group (P < 0.001; however, their values were significantly high in the finger group compared with both the control and the sealing group (P < 0.001. The incidence and severity of sore throat were significantly high in the finger group compared with both the control and the sealing group (P = 0.0009 and P = 0.0026. Three patients in the control group developed air leak around the endotracheal tube cuff. The incidence and severity of other complications were similar in the three groups. Conclusion: In pediatric N 2 O, free general anesthesia using Microcuff pediatric tracheal tub, sealing cuff pressure is safer than finger palpation technique regarding post-extubation morbidities and more reliable than recommended safe pressure in prevention of the air leak.

  6. Irradiation creep and density changes observed in MA957 pressurized tubes irradiated to doses of 40-110 dpa at 400-750 °C in FFTF

    Science.gov (United States)

    Toloczko, M. B.; Garner, F. A.; Maloy, S. A.

    2012-09-01

    An irradiation creep and swelling study was performed on tubing constructed from the yttrium/titanium oxide dispersion strengthened (ODS) ferritic steel MA957. As a result of the reduction operations during manufacture, the grains in the tubing were highly elongated in the direction of the tubing longitudinal axis. Pressurized creep tubes were irradiated in the Fast Flux Test Facility (FFTF) to doses ranging from 40 dpa to 110 dpa at target temperatures ranging from 400 to 750 °C. The diametral strains produced during irradiation exhibit primary (transient) creep strains that are dependent on stress and increase with irradiation temperature and are followed by a temperature-independent steady-state creep rate of ˜0.75 × 10-6 (MPa dpa)-1, a value similar to that of traditional tempered ferritic/martensitic steels. Contributions to primary creep strains may arise not only from classical thermal creep or irradiation creep considerations, but also may result from an irradiation-stimulated growth process whereby the highly elongated grain structure shrinks somewhat in the elongated direction, reducing the tubing aspect ratio to produce slightly fatter grains and thereby increasing the tube diameter. One manifestation of this process is a change in tube diameter that is not accompanied by a density change characteristic of either void swelling or precipitation-induced changes in lattice parameter. These results provide the first demonstration that resistance to irradiation creep can be extended to higher temperatures by dispersoid addition, and most importantly, this resistance is maintained to high radiation damage levels at least for temperatures of 600 °C or less.

  7. Development of modern CANDU PHWR cross-section libraries for SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Shoman, Nathan T., E-mail: nshoman@vols.utk.edu; Skutnik, Steven E., E-mail: sskutnik@utk.edu

    2016-06-15

    Highlights: • New ORIGEN libraries for CANDU 28 and 37-element fuel assemblies have been created. • These new reactor data libraries are based on modern ENDF/B-VII.0 cross-section data. • The updated CANDU data libraries show good agreement with radiochemical assay data. • Eu-154 overestimated when using ENDF-VII.0 due to a lower thermal capture cross-section. - Abstract: A new set of SCALE fuel lattice models have been developed for the 28-element and 37-element CANDU fuel assembly designs using modern cross-section data from ENDF-B/VII.0 in order to produce new reactor data libraries for SCALE/ORIGEN depletion analyses. These new libraries are intended to provide users with a convenient means of evaluating depletion of CANDU fuel assemblies using ORIGEN through pre-generated cross sections based on SCALE lattice physics calculations. The performance of the new CANDU ORIGEN libraries in depletion analysis benchmarks to radiochemical assay data were compared to the previous version of the CANDU libraries provided with SCALE (based on WIMS-AECL models). Benchmark comparisons with available radiochemical assay data indicate that the new cross-section libraries perform well at matching major actinide species (U/Pu), which are generally within 1–4% of experimental values. The library also showed similar or better results over the WIMS-AECL library regarding fission product species and minor actinoids (Np, Am, and Cm). However, a notable exception was in calculated inventories of {sup 154}Eu and {sup 155}Eu, where the new library employing modern nuclear data (ENDF/B-VII.0) performed substantially poorer than the previous WIMS-AECL library (which used ENDF-B/VI.8 cross-sections for these species). The cause for this discrepancy appears to be due to differences in the {sup 154}Eu thermal capture cross-section between ENDF/B-VI.8 and ENDF/B-VII.0, an effect which is exacerbated by the highly thermalized flux of a CANDU heavy water reactor compared to that of a

  8. Application of principal component analysis for the diagnosis of neutron overpower system oscillations in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara; Gabbar, Hossam A., E-mail: Hossam.gabbar@uoit.ca

    2014-04-01

    Highlights: • Diagnosis of neutron overpower protection (NOP) in CANDU reactors. • Accurate reactor detector modeling. • NOP detectors response analysis. • Statistical methods for quantitative analysis of NOP detector behavior. - Abstract: An accurate fault modeling and troubleshooting methodology is required to aid in making risk-informed decisions related to design and operational activities of current and future generation of CANDU{sup ®} designs. This paper attempts to develop an explanation for the unanticipated detector response and overall behavior phenomena using statistical methods to compliment traditional engineering analysis techniques. Principal component analysis (PCA) methodology is used for pattern recognition using a case study of Bruce B zone-control level oscillations.

  9. A study for good regulation of the CANDU's in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Ki; Shin, Y. K.; Joe, S. K.; Kim, J. S.; Yu, Y. J.; Lee, Y. J. [Ajou Univ., Suwon (Korea, Republic of)

    2002-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the second phase of this 2 years study, following research activities were done. Review the technical basis and framework of the new Canadian Regulation System and IAEA. Analysis on the interview of Wolsung operation staffs to identify important safety issues and regulation problems experienced at operation. Providing a plan of CANDU regulation system enhancement program.

  10. Experimental investigation of syngas flame stability using a multi-tube fuel injector in a high pressure combustor

    Science.gov (United States)

    Maldonado, Sergio Elzar

    Over 92% of the coal consumed by power plants is used to generate electricity in the United States (U.S.). The U.S. has the world's largest recoverable reserves of coal, it is estimated that reserves of coal will last more than 200 years based in current production and demand levels. Integrated Gasification Combined Cycle (IGCC) power plants aim to reduce the amount of pollutants by gasifying coal and producing synthesis gas. Synthesis gas, also known as syngas, is a product of coal gasification and can be used in gas turbines for energy production. Syngas is primarily a mixture of hydrogen and carbon monoxide and is produced by gasifying a solid fuel feedstock such as coal or biomass. The objective of the thesis is to create a flame stability map by performing various experiments using high-content hydrogen fuels with varying compositions of hydrogen representing different coal feedstocks. The experiments shown in this thesis were performed using the High-Pressure Combustion facility in the Center for Space Exploration Technology Research (CSETR) at the University of Texas at El Paso (UTEP). The combustor was fitted with a novel Multi-Tube fuel Injector (MTI) designed to improve flame stability. This thesis presents the results of testing of syngas fuels with compositions of 20, 30, and 40% hydrogen concentrations in mixtures with carbon monoxide. Tests were completed for lean conditions ranging from equivalence ratios between 0.6 and 0.9. The experimental results showed that at an equivalence ratio of 0.6, a stable flame was not achieved for any of the fuel mixtures tested. It was also observed that the stability region of the syngas flame increased as equivalence ratio and the hydrogen concentration in syngas fuel increases with the 40% hydrogen-carbon monoxide mixture demonstrating the greatest stability region. Design improvements to the MTI are also discussed as part of the future work on this topic.

  11. 管内高压智能封堵机器人%The In-tube Pressurized Intelligent Plugging Robot

    Institute of Scientific and Technical Information of China (English)

    刘华洁; 张策; 张仕民; 朱吉祥

    2013-01-01

    为满足国内管道快速维修的需要,开展了管道智能封堵技术研究.在介绍管内智能封堵机器人的封堵作业流程后,描述了封堵机器人的结构组成,包括双向清管式封堵单元、远程控制系统和地面控制中心,给出了主要技术参数.随后简要介绍和分析了封堵机器人的性能试验情况,包括通过性能试验、双向通信和压力试验以及解封试验.试验证明封堵机器人可在一段管道内实现多次封堵和解堵作业,大大缩短管道停输时间,且操作简单,封堵性能良好,能够封堵20MPa的高压,无渗漏.该智能封堵机器人的研制成功为国内管道维抢修技术提供了补充.%To meet the domestic needs of fast maintenance ot pipeline,research on the plugging technology ofpipeline was conducted.The paper first introduces the plugging process of in-tube intelligent plugging robot,describes the structural composition of the robot,including two-way pigging plugging unit,remote control system and ground control center,and offers the main technological parameters.Then,it briefly introduces and analyzes the performance test of the robot,including passage capacity test,two-way communication,pressure test and plug re-moval test.The tests have proved that the robot can achieve multiple plugging and plug removal operations in a sec-tion of pipeline.This remarkably shortens the pipeline shutdown time.The operation is simple and the plugging per-formance is desirable.The robot can plug as high as 20 MPa pressure with no leakage.The successful developmentof the robot serves as a supplementation for domestic pipeline maintenance technology.

  12. High Pressure Pneumatic Forming of Ti-3Al-2.5V Titanium Tubes in a Square Cross-Sectional Die

    Directory of Open Access Journals (Sweden)

    Gang Liu

    2014-08-01

    Full Text Available A new high strain rate forming process for titanium alloys is presented and named High Pressure Pneumatic Forming (HPPF, which might be applicable to form certain tubular components with irregular cross sections with high efficiency, both with respect to energy cost and time consumption. HPPF experiments were performed on Ti-3Al-2.5V titanium alloy tubes using a square cross-sectional die with a small corner radius. The effects of forming of pressure and temperature on the corner filling were investigated and the thickness distributions after the HPPF processes at various pressure levels are discussed. At the same time, the stress state, strain and strain rate distribution during the HPPF process were numerically analyzed by the finite element method. Microstructure evolution of the formed tubes was also analyzed by using electron back scattering diffraction (EBSD. Because of different stress states, the strain and strain rate are very different at different areas of the tube during the corner filling process, and consequently the microstructure of the formed component is affected to some degree. The results verified that HPPF is a potential technology to form titanium tubular components with complicated geometrical features with high efficiency.

  13. ENHANCING ADVANCED CANDU PROLIFERATION RESISTANCE FUEL WITH MINOR ACTINIDES

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Chang

    2010-05-01

    The advanced nuclear system will significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. Minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

  14. Safety design of next generation SUI of CANDU stations

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe St. N., Oshawa, L1H 7K4 ON (Canada); Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe St. N., Oshawa, L1H 7K4 ON (Canada)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Review of current SUI technologies and challenges. Black-Right-Pointing-Pointer Propose a new type of SUI detectors. Black-Right-Pointing-Pointer Propose a new SUI system architecture and layout. Black-Right-Pointing-Pointer Propose implementation procedure for SUI with reduced risks. - Abstract: Due to the age and operating experience of Nuclear Power Plants, equipment ageing and obsolescence has become one of the main challenges that need to be resolved for all systems, structures and components in order to ensure a safe and reliable production of energy. This paper summarizes the research into a methodology for modernization of Start-Up Instrumentation (SUI), both in-core and Control Room equipment, using a new generation of detectors and cables in order to manage obsolescence. The main objective of this research is to develop a new systematic approach to SUI installation/replacement procedure development and optimization. Although some additional features, such as real-time data monitoring and storage/archiving solutions for SUI systems are also examined to take full advantage of today's digital technology, the objectives of this study do not include detailed parametrical studies of detector or system performance. Instead, a number of technological, operational and maintenance issues associated with Start-Up Instrumentation systems at Nuclear Power Plants (NPPs) will be identified and a structured approach for developing a replacement/installation procedure that can be standardized and used across all of the domestic CANDU (Canadian Deuterium Uranium) stations is proposed.

  15. Assessment of System Behavior and Actions Under Loss of Electric Power For CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kang, San Ha; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    For the analysis, the CANDU-6 plant in Korea is considered and only the passive components are operable. The other systems are assumed to be at failed condition due to the loss of electric power. At this accident, only the inventories remained in the primary heat transport system (PHTS) and steam generator can be used for the decay heat removal. Due to the transfer of decay heat, the inventory of steam generator secondary side is discharged to the air through passive operation of main steam safety valves (MSSVs). After the steam generators are dried, the PHTS is over-pressurized and the coolant is discharged to fuelling machine vault through passive operation of degasser condenser tank relief valves (DCRVs). Under this situation, the maintenance of the integrity of PHTS is important for the protection of radionuclides release to the environment. Thus, deterministic analysis using CATHENA code is carried out for the simulation of the accident and the appropriate operator action is considered. The loss of electric power results in the depletion of steam generator inventory which is necessary for the decay heat removal. If only the passive system is credited, the PT can be failed after the steam generator is depleted. For the prevention of the PT failure, the feedwater should be supplied to the steam generator before 4,800s after the accident. The feedwater can be supplied using water in dousing tank if the steam generators are depressurized. The decay heat from the core is removed through natural circulation if the feedwater can be supplied continuously.

  16. Signal processing system design for improved shutdown system of CANDU{sup ®} nuclear reactors in large break LOCA events

    Energy Technology Data Exchange (ETDEWEB)

    Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Xia, Lingzhi; Isham, Manir U. [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Ponomarev, Vladimir [Megawatt Solutions, 1235 Radom St., unit 68, Pickering, ON, Canada L1W 1J3 (Canada)

    2016-03-15

    Highlights: • Neutronic signal processing system design to improve CANDU SDS1 performance. • Reactor modeling for CANDU LLOCA transient. • MATLAB/Simulink system implementation for the SDS1 trip logic. • Increasing the SDS1 trip response. - Abstract: For CANDU reactors, several options to improve CANDU nuclear power plant operation safety margin have been investigated in this paper. A particular attention is paid to the response time of CANDU shutdown system number 1 (SDS1) in case of large break loss of coolant accident (LLOCA). Based on point kinetic method, a systematic fundamental analysis is performed to CANDU LLOCA event, and the power transient signal is generated. In order to improve the SDS1 response time during LLOCA events, an innovative power measurement and signal processing system is particularly designed. The new signal processing system is implemented with the input of the LLOCA power transient, and the simulation results of the reactor trip time and signal are compared to those of the existing system in CANDU power plants. It is demonstrated that the new signal processing system can not only achieve a shorter reactor trip time than the existing system, but also accommodate the spurious trip immunity. This will significantly enhance the safety margin for the power plant operation, or bring extra economical benefits to the power plant units.

  17. Prompt improvement of a pressure ulcer by the administration of high viscosity semi-solid nutrition via a nasogastric tube in a man with tuberculosis: a case report

    Directory of Open Access Journals (Sweden)

    Hatsuda Kazuyoshi

    2010-01-01

    Full Text Available Abstract Introduction Semi-solid nutrition with high viscosity has the advantage of reducing gastroesophageal reflux and diarrhea and shortens the duration of administration compared with liquid nutrition. This is the first report describing the administration of semi-solid nutrition with high viscosity via a nasogastric tube, which achieved a remarkable improvement in the patient's nutritional state. Case presentation A 67-year-old man (mongoloid race, Japanese with tuberculosis, a pressure ulcer and malnutrition was admitted to our hospital. He also had right hemiplegia, dysphagia and aphasia as sequelae of a cerebral hemorrhage. Before his admission, he had been treated at another hospital with 600 kcal/day of liquid nutrition via a nasogastric tube, which was insufficient and induced severe malnutrition. After he was admitted to our hospital, we increased the quantity of his liquid nutrition without success because of complications, specifically diarrhea and gastroesophageal reflux. As it was difficult to confirm whether or not he would accept gastrostomy feeding, we administered semi-solid nutrition with high viscosity (20,000 mPa x s via a large-bore nasogastric tube (18 French. Soon after he was started on semi-solid nutrition, his pressure ulcer and malnutrition improved without diarrhea or complications accompanying the large-bore nasogastric tube. Conclusion When patients have problems with liquid nutrition, such as diarrhea or gastroesophageal reflux, semi-solid nutrition via a nasogastric tube is a useful method of achieving improvements in nutritional state in a short period of time.

  18. Evaluation of CANDU6 PCR (power coefficient of reactivity) with a 3-D whole-core Monte Carlo Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Motalab, Mohammad Abdul; Kim, Woosong; Kim, Yonghee, E-mail: yongheekim@kaist.ac.kr

    2015-12-15

    Highlights: • The PCR of the CANDU6 reactor is slightly negative at low power, e.g. <80% P. • Doppler broadening of scattering resonances improves noticeably the FTC and make the PCR more negative or less positive in CANDU6. • The elevated inlet coolant condition can worsen significantly the PCR of CANDU6. • Improved design tools are needed for the safety evaluation of CANDU6 reactor. - Abstract: The power coefficient of reactivity (PCR) is a very important parameter for inherent safety and stability of nuclear reactors. The combined effect of a relatively less negative fuel temperature coefficient and a positive coolant temperature coefficient make the CANDU6 (CANada Deuterium Uranium) PCR very close to zero. In the original CANDU6 design, the PCR was calculated to be clearly negative. However, the latest physics design tools predict that the PCR is slightly positive for a wide operational range of reactor power. It is upon this contradictory observation that the CANDU6 PCR is re-evaluated in this work. In our previous study, the CANDU6 PCR was evaluated through a standard lattice analysis at mid-burnup and was found to be negative at low power. In this paper, the study was extended to a detailed 3-D CANDU6 whole-core model using the Monte Carlo code Serpent2. The Doppler broadening rejection correction (DBRC) method was implemented in the Serpent2 code in order to take into account thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. Time-average equilibrium core was considered for the evaluation of the representative PCR of CANDU6. Two thermal hydraulic models were considered in this work: one at design condition and the other at operating condition. Bundle-wise distributions of the coolant properties are modeled and the bundle-wise fuel temperature is also considered in this study. The evaluated nuclear data library ENDF/B-VII.0 was used throughout this Serpent2 evaluation. In these Monte Carlo calculations, a large number

  19. Development of the advanced PHWR technology -Design and analysis of CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Hoh Chun; Shim, Kee Sub; Byun, Taek Sang; Park, Kwang Suk; Kang, Heui Yung; Kim, Bong Kee; Jung, Chang Joon; Lee, Yung Wook; Bae, Chang Joon; Kwon, Oh Sun; Oh, Duk Joo; Im, Hong Sik; Ohn, Myung Ryong; Lee, Kang Moon; Park, Joo Hwan; Lee, Eui Joon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel design and analysis project, and describes CANFLEX fuel design and mechanical integrity analysis, reactor physics analysis and safety analysis of the CANDU-6 with the CANFLEX-NU. The following is the R and D scope of this fiscal year : (1) Detail design of CANFLEX-NU and detail analysis on the fuel integrity, reactor physics and safety. (a) Detail design and mechanical integrity analysis of the bundle (b) CANDU-6 refueling simulation, and analysis on the Xe transients and adjuster system capability (c) Licensing strategy establishment and safety analysis for the CANFLEX-NU demonstration demonstration irradiation in a commercial CANDU-6. (2) Production and revision of CANFLEX-NU fuel design documents (a) Production and approval of CANFLEX-NU reference drawing, and revisions of fuel design manual and technical specifications (b) Production of draft physics design manual. (3) Basic research on CANFLEX-SEU fuel. 55 figs, 21 tabs, 45 refs. (Author).

  20. EVALUATION OF THE APPLICABLE REACTIVITY RANGE OF A REACTIVITY COMPUTER FOR A CANDU-6 REACTOR

    Directory of Open Access Journals (Sweden)

    EUN KI LEE

    2014-04-01

    Full Text Available Recently, a CANDU digital reactivity computer system (CDRCS to measure the worth of the liquid zone controller in a CANDU-6 was developed and successfully applied to a physics test of refurbished Wolsong Unit 1. In advance of using the CDRCS, its measureable reactivity range should be investigated and confirmed. There are two reasons for this investigation. First, the CANDU-6 has a larger reactor and smaller excore detectors than a general PWR and consequently the measured reactivity is likely to reflect the peripheral power variation only, not the whole core. The second reason is photo neutrons generated from the interaction of the moderator and gamma-rays, which are never considered in a PWR. To evaluate the limitations of the CDRCS, several tens of three-dimensional steady and transient simulations were performed. The simulated detector signals were used to obtain the dynamic reactivity. The difference between the dynamic reactivity and the static worth increases in line with the water level changes. The maximum allowable reactivity was determined to be 1.4 mk in the case of CANDU-6 by confining the difference to less than 1%.

  1. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  2. ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca; Nava-Dominguez, A., E-mail: navadoma@aecl.ca

    2014-08-15

    Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

  3. Ultrasonic water level determination of the high-pressure boilers tubes; Determinacao do nivel d'agua em tubos verticais de caldeiras aquatubulares por ultra-som

    Energy Technology Data Exchange (ETDEWEB)

    Goettems, Felipe Samuel; Reolon, Amon Marques; Avancini, Flavio; Braga, Rubem Manoel de; Reguly, Afonso [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Lab. de Metalurgia Fisica], e-mail: fgoettems@demet.ufrgs.br

    2006-07-01

    Electric power is very important to our society and thermoelectric power plant. They are especially important mainly in the summer when there is a scarcity in water supply to hydroelectric power plants. Southern Brazilian thermoelectric power plants employ high-pressure boilers in order to generate water vapor which, in turn, moves turbines to produce electricity. These high-pressure boilers must work in a continuous way to avoid damages caused by emergency halts. To accomplish this, some actions must be taken. The water height inside of the tubes must be kept in a strict level to avoid thermal gradient in both water walls and super-heater header. In this water walls the water become in vapor. The best way to regulate the valves that command the water level is through the control of the water height and this is the main purpose of this work. The ultrasound is a nondestructive test which is able in doing this control without damaging the tube. This method allows determining the water level, improving the system performance and reducing the maintenance costs caused by tube collapse. (author)

  4. Automated refueling simulations of a CANDU for the exploitation of thorium fuels

    Science.gov (United States)

    Holmes, Bradford

    CANDU nuclear reactors are in a unique circumstance where they are able to utilize and exploit a number of different fuel options to provide power as a utility. Thorium, a fertile isotope found naturally, is one option that should be explored. Thorium is more abundant than uranium, which is the typical fuel in the reactor and the availability of thorium makes nuclear energy desirable to more countries. This document contains the culmination of a project that explores, tests, and analyzes the feasibility of using thorium in a CANDU reactor. The project first develops a set of twodimensional lattice and three dimensional control rod simulations using the DRAGON Version 4 nuclear physics codes. This step is repeated for many concentrations of thorium. The data generated in these steps is then used to determine a functional enrichment of thorium. This is done via a procedural elimination and optimization of certain key parameters including but not limited to average exit burnup and reactivity evolution. For the purposes of this project, an enrichment of 1 % thorium was found viable. Full core calculations were done using the DONJON 4 code. CANFUEL, a program which simulates the refueling operations of a CANDU reactor for this fuel type was developed and ran for a simulation period of one hundred days. The program and the fuel selection met all selected requirements for the entirety of the simulation period. CANFUEL requires optimization for fuel selection before it can be used extensively. The fuel selection was further scrutinized when a reactivity insertion event was simulated. The adjuster rod 11 withdrawal from the core was analyzed and compared to classical CANDU results in order to ensure no significant deviations or unwanted evolutions were encountered. For this case, the simulation results were deemed acceptable with no significant deviations from the classical CANDU case.

  5. THE INSIDE PRESSURE OF STENT TUBE ON CHOLEDOCO-JEJUNOSTOMY SCAR: A STUDY ON SCAR TISSUE COLLAGEN

    Institute of Scientific and Technical Information of China (English)

    郭善禹; 周林斌; 姚德成; 孙建民

    2002-01-01

    Objective As the beneficial effect to the skin scar under external bandage compression, intra-choledocal stent must have the same effect on splanchnic scar formation. The experiment consists to work out the time optimum to yield a minimum scar formation. Methods By means of transmitting electronic microscope (TEM), computer assisted three-dimensional morphometry (CAM), and biochemical analysis to determine the extracellular collagen volume density (ECVD) and biochemical collagen content (BCC), to analyze the ultrastructure and components within scar tissues removed from the specimens in 3 groups of experimental animals were detailed. Results In the animals of simple choledoco-jejunostomy (CJ) group, active scar proliferation was seen in all specimens excised within one year after operation. In the stent group, decreasing collagen fibers arranged in orientation began to appear in the 6-month specimens and scar maturation existed in the 9- and 12-month specimens. In periodic tube withdrawal group, 3 months following tube ablation, scar proliferation recurred in the 6th month tube retaining animals, whereas scar maturation without recurrence happened in animals following 9 to 12 months tube retaining. Conclusion 9~12 months of tube stent is necessary for stable scar maturation.

  6. SAFIRE - a robotic inspection system for CANDU feeders

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, R. [OC Robotics, Bristol (United Kingdom)

    2011-07-01

    The condition of primary circuit feeder pipes in CANDU reactors is relevant to the commercial viability and plant life. One known wear mechanism is external fretting between feeder pipes and adjacent services or support structures, particularly within the Upper Feeder Cabinet (UFC). Fretting leads to wall thinning which must not exceed certain agreed limits. Chafe shields have been added to protect the feeder pipes. Regular inspections are required of the chafe shields, feeder pipes and other structures that may cause feeder damage. Historically, the dose received by inspectors conducting this work has been significant. For this reason Ontario Power Generation has invested in a remotely operated robot system to conduct visual inspections within the UFC. This system, called SAFIRE for 'Snake-Arm Feeder Inspection Robot Equipment' has been deployed at Pickering during 2010 and 2011 and has been used to inspect areas that are extremely difficult to inspect with existing manual techniques. The 2011 scope of work included inspection of a total of 660 feeder pipes in three UFC quadrants, in two reactors. The full scope was completed over a one-month period in Autumn 2011 in which SAFIRE was used during 23, twelve hour shifts. This included two periods each of 72 hours of continuous operation using multiple teams of operators. SAFIRE is remote controlled delivery system for multiple cameras to record still images and video. The main system elements include a snake-arm robot mounted on a mobile vehicle. It can be controlled from up to 500m away using a fibre/copper connection. The snake-arm is 2.2m long, 25mm wide and has 18 degrees of freedom. It is designed to snake between the rows of feeder pipes to inspect feeder/hanger interfaces, both above and below the feeder cabinet catwalks. Future upgrades offer the potential to add additional tools to increase functionality. This paper describes the SAFIRE development process from inception to operational experience

  7. Investigation of the effects of baffle orientation, baffle cut and fluid viscosity on shell side pressure drop and heat transfer coefficient in an e-type shell and tube heat exchanger

    OpenAIRE

    Mohammadi, Koorosh

    2011-01-01

    The commercial CFD code FLUENT is used to determine the effect of baffle orientation and baffle cut as well as viscosity of the working fluid on the shell-side heat transfer and pressure drop of a shell and tube heat exchanger. The shell and tube heat exchangers considered follow the TEMA standards. The investigation has been completed in three stages: 1. The shell and tube heat exchanger consists of 660 plain tubes with fixed outside diameter which are arranged in a triangular layout. Hor...

  8. Experimental investigation and correlation of two-phase frictional pressure drop of R410A-oil mixture flow boiling in a 5 mm microfin tube

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Guoliang; Hu, Haitao; Huang, Xiangchao [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Deng, Bin [Institute of Heat Transfer Technology, Golden Dragon Precise Copper Tube Group Inc., Shanghai 200135 (China); Gao, Yifeng [International Copper Association, Shanghai Office, Shanghai 200020 (China)

    2009-01-15

    This study presents experimental two-phase frictional data for R410A-oil mixture flow boiling in an internal spiral grooved microfin tube with outside diameter of 5 mm. Experimental parameters include the evaporation temperature of 5 C, the mass flux from 200 to 400 kg m{sup -2} s{sup -1}, the heat flux from 7.46 to 14.92 kW m{sup -2}, the inlet vapor quality from 0.1 to 0.8, and nominal oil concentration from 0 to 5%. The test results show that the frictional pressure drop of R410A initially increases with vapor quality and then decreases, presenting a local maximum in the vapor quality range between 0.7 and 0.8; the frictional pressure drop of R410A-oil mixture increases with the mass flux, the presence of oil enhances two-phase frictional pressure drop, and the effect of oil on frictional pressure drop is more evident at higher vapor qualities where the local oil concentrations are higher. The enhanced factor is always larger than unity and increases with nominal oil concentration at a given vapor quality. The range of the enhanced factor is about 1.0-2.2 at present test conditions. A new correlation to predict the local frictional pressure drop of R410A-oil mixture flow boiling inside the internal spiral grooved microfin tube is developed based on local properties of refrigerant-oil mixture, and the measured local frictional pressure drop is well correlated with the empirical equation proposed by the authors. (author)

  9. Compact exhaust gas boilers. Investigation of heat transfer and pressure drop for serrated finned tubes. (Abbreviated edition); Kompakte avgasskjeler. Undersoekelse av varmeovergang og trykktap for serraterte finnede roer. (Forkortet utgave.)

    Energy Technology Data Exchange (ETDEWEB)

    Midtbust, H.O.; Naess, E.

    1995-07-03

    This report discusses investigations of pressure drops and heat transfer in cross-current flow of gases (air) on bundles of serrated finned tubes. For the various geometries the tube spacing and tube diameter varied while the geometry of the fins remained unchanged. Pressure drop and heat transfer were measured at six different air flows for each geometry, and the results compared with available correlations from the literature. The measurements are at variance with the correlations and indicate that the pressure loss coefficient for all the tested geometries are less influenced by the flow conditions (air speed) than predicted by the correlations. Compared with the correlation recommended by the tube supplier (Weierman`s correlation) the measured results are mostly somewhat higher than predicted for the larger air flows. The maximum observed deviation is 70%. The deviation between the published pressure loss correlations is also considerable. The heat transfer measurements agree qualitatively with the published correlations with respect to the flow conditions. Comparison with the heat transfer correlation recommended by the tube supplier indicates that the correlation over-predicts the heat transfer quite considerably. The deviation increases systematically with reduced tube diameter and with increased angle of the tube arrangement. 16 figs., 7 tabs.

  10. Heat transfer and pressure drop comparison of louver- and plain-finned heat exchangers where one fluid passes through flattened tubes

    Directory of Open Access Journals (Sweden)

    J.M. Gorman

    2015-03-01

    Full Text Available Louvered fins constitute a major methodology for heat transfer enhancement. Of critical significance in evaluating the worthiness of such fins is the comparison between the heat transfer and pressure drop for a thus-finned heat exchanger with the baseline case of a counterpart plain-finned heat exchanger. Up to the present, it appears that such comparisons are confined to heat exchangers in which one of the participating fluids passes through circular tubes. In another basic geometry in which louvered fins have been employed, the aforementioned participating fluid passes through flattened tubes which are virtually rectangular in cross section. The focus of the present paper is to obtain results for the latter basic geometry for both louver-fin-based heat exchangers and counterpart plain-fin-based heat exchangers. The results were obtained by means of numerical simulation over a range of Reynolds numbers spanning approximately a factor of five. Over this range, enhancements of the heat transfer rate ranged from factors of approximately 2.2–2.8. Over this same Reynolds number range, the pressure drop increased by factors of 2.3–3.6. This outcome is attributable to the fact that the rate of heat transfer is less sensitive to the velocity than is the pressure drop.

  11. FAST: A Fuel And Sheath Modeling Tool for CANDU Reactor Fuel

    Science.gov (United States)

    Prudil, Andrew Albert

    Understanding the behaviour of nuclear fuel during irradiation is a complicated multiphysics problem involving neutronics, chemistry, radiation physics, material-science, solid mechanics, heat transfer and thermal-hydraulics. Due to the complexity and interdependence of the physics and models involved, fuel modeling is typically clone with numerical models. Advancements in both computer hardware and software have made possible new more complex and sophisticated fuel modeling codes. The Fuel And Sheath modelling Tool (FAST) is a fuel performance code that has been developed for modeling nuclear fuel behaviour under normal and transient conditions. The FAST code includes models for heat generation and transport, thermal expansion, elastic strain, densification, fission product swelling, pellet relocation, contact, grain growth, fission gas release, gas and coolant pressure and sheath creep. These models are coupled and solved numerically using the Comsol Multiphysics finite-element platform. The model utilizes a radialaxial geometry of a fuel pellet (including dishing and chamfering) and accompanying fuel sheath allowing the model to predict circumferential ridging. This model has evolved from previous treatments developed at the Royal Military College. The model has now been significantly advanced to include: a more detailed pellet geometry, localized pellet-to-sheath gap size and contact pressure, ability to model cracked pellets, localized fuel burnup for material property models, improved U02 densification behaviour, fully 2-dimensional model for the sheath, additional creep models, additional material models, an FEM Booth-diffusion model for fission gas release (including ability to model temperature and power changes), a capability for end-of-life predictions, the ability to utilize text files as model inputs, and provides a first time integration of normal operating conditions (NOC) and transient fuel models into a single code (which has never been achieved

  12. ACCEPT: a three-dimensional finite element program for large deformation elastic-plastic-creep analysis of pressurized tubes (LWBR/AWBA Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Hutula, D.N.; Wiancko, B.E.

    1980-03-01

    ACCEPT is a three-dimensional finite element computer program for analysis of large-deformation elastic-plastic-creep response of Zircaloy tubes subjected to temperature, surface pressures, and axial force. A twenty-mode, tri-quadratic, isoparametric element is used along with a Zircaloy materials model. A linear time-incremental procedure with residual force correction is used to solve for the time-dependent response. The program features an algorithm which automatically chooses the time step sizes to control the accuracy and numerical stability of the solution. A contact-separation capability allows modeling of interaction of reactor fuel rod cladding with fuel pellets or external supports.

  13. Experimental Study on Heat Transfer and Pressure Drop Characteristics of Four Types of Plate Fin-and-Tube Heat Exchanger Surfaces

    Institute of Scientific and Technical Information of China (English)

    1994-01-01

    In this paper,air side heat transfer and pressure drop characteristics of twelve three-row plate fin-and-tube heat exchanger cores of four types of fin configurations have been experimentally investigated .The heat transfer and friction factor correlations for the twelve cores are provided in a wide range of Reynolds number.It is found that in the range of Reynolds number tested.the Nusselt number of the slotted fin surface is the largest and that of the plain plate fin is the lowest while the Nusselt numbers of two types of wavy fins are somewhere in between.

  14. Achieving a Safe Endotracheal Tube Cuff Pressure in the Prehospital Setting: Is It Time to Revise the Standard Cuff Inflation Practice?

    Science.gov (United States)

    Carhart, Elliot; Stuck, Logan H; Salzman, Joshua G

    2016-01-01

    Numerous studies have reported unsafe endotracheal tube (ETT) cuff pressures (CP) in the prehospital environment. The purpose of this study was to identify an optimal cuff inflation volume (CIV) to achieve a safe CP (20-30 cmH2O). This observational study utilized 30 recently harvested ovine tracheae, which were warmed from refrigeration in a water bath at 85°F prior to testing. Each trachea was intubated with five different ETT sizes (6.0-8.0 mm), and each size tube was tested with six cuff inflation volumes (5-10 cc). The order of ETT size for each trachea and CIV for each size ETT was randomly pre-assigned. Data were descriptively summarized and categorized before mixed-effects logistic regression was used to determine optimal CIV. Only 113 CP measurements (12.6%, N = 900) were within the optimal range (M = 54.75 cmH2O, SD = 38.52), all of which resulted from a CIV 6 or 7 cc (61% and 39%, respectively). CIVs of 5 cc (n = 150) resulted in underinflation (30 cmH2O) in all instances, regardless of ETT size. The odds of achieving a safe CP were greater with CIV of 6 cc for tube sizes 6.0 (OR = 15.9, 95% CI = 3.85-65.58, p safe CP between CIV of 6 and 7 cc for tube sizes 7.0, 7.5, or 8.0 mm. Neither trachea circumference (M = 7.11 cm, SD = 0.40), nor tissue temperature (M = 81.32°F, SD = 0.93) were found to be significant predictors of CP (p = 0.20 and 0.81, respectively). Our study showed a high frequency of CP measurements outside of the desired norms. The CIV range of 6-7 cc resulted in the highest likelihood of achieving the desired cuff pressure range, while cuffs inflated with 8-10 cc resulted in dangerously high CPs in all instances. In the absence of a more ideal solution, the results of this study suggest that narrowing the recommended CIV from 5-10 cc to 6-7 cc might be a reasonable target for any tube size.

  15. Does objective measurement of tracheal tube cuff pressures minimise adverse effects and maintain accurate cuff pressures? A systematic review and meta-analysis.

    Science.gov (United States)

    Hockey, C A; van Zundert, A A J; Paratz, J D

    2016-09-01

    Correct inflation pressures of the tracheal cuff are recommended to ensure adequate ventilation and prevent aspiration and adverse events. However there are conflicting views on which measurement to employ. The aim of this review was to examine whether adjustment of cuff pressure guided by objective measurement, compared with subjective measurement or observation of the pressure value alone, was able to prevent patient-related adverse effects and maintain accurate cuff pressures. A search of PubMed, Web of Science, Embase, CINAHL and ScienceDirect was conducted using keywords 'cuff pressure' and 'measure*' and related synonyms. Included studies were randomised or pseudo-randomised controlled trials investigating mechanically ventilated patients both in the intensive care unit and during surgery. Outcomes included adverse effects and the comparison of pressure measurements. Pooled analyses were performed to calculate risk ratios, effect sizes and 95% confidence intervals. Meta-analysis found preliminary evidence that adjustment of cuff pressure guided by objective measurement as compared with subjective measurement or observation of the pressure value alone, has benefit in preventing adverse effects. These included cough at two hours (odds ratio [OR] 0.42, confidence interval [CI] 0.23 to 0.79, P=0.007), hoarseness at 24 hours (OR 0.49, CI 0.31 to 0.76, P measurement to guide adjustment or observation of the pressure value alone may lead to patient-related adverse effects and inaccuracies. It is recommended that an objective form of measurement be used.

  16. Method of Measuring the Vapor Pressure and Concentration of Fluids using VLE and Vibrating Tube Densitometer Apparatuses

    OpenAIRE

    Abdalla, Momin Elhadi; Pannir, Siddharth

    2016-01-01

    This work presents the vapor pressure and concentration measurement of newly discovered environmentally friendly refrigerants 1, 1-difluoroethane (R152a) and 1,1,1,3,3-Pentafluorbutane (R365mfc), besides their mixture. The experimental procedure used in this work was a VLE recirculation type apparatus in which the liquid phase is circulating around the equilibrium cell. Special attention was given to enable a highly accurate vapor pressure measurement up to maximum pressure of 25 bar. The li...

  17. 测压管路动态特性实测技术研究%Study on measurement technology of dynamics characteristics of typical tubes for pressure measurements

    Institute of Scientific and Technical Information of China (English)

    余世策; 韩新刚; 冀晓华; 屠荣伟; 蒋建群

    2012-01-01

    利用声音振动发生原理研制了多功能声音振动发生装置,开发了测压管路动态特性的实测技术,并对风洞试验中典型测压管路的频响特性进行了实测.采用正弦压力波对不同的测压管路进行激励,采用多点联合扫描技术提高采样频率,得到完整的正弦波动曲线.实验研究结果表明,该实验技术可以得到高频的动态压力信号和准确的频响特性曲线,为误差修正提供了依据.%By using independently developed sound vibration generating device, the measurement technology of dynamics characteristics of typical tubes for pressure measurements was developed, and the frequency response characteristics of typical tubes for fluctuating wind pressure measurements were measured. By using principle of sound vibrations, a multi-function sound vibration generating device was developed. Different pipes were excited by sine pressure waves, multi-point scanning technology was used to improve the sampling frequency for getting full curves of sine waves. Experimental results show that the experimental technique developed can be ' used to obtain high-frequency dynamic pressure signals and accurate frequency response curve for providing a basis for the error correction.

  18. Study on the use of slightly enriched uranium fuel cycle in an existing CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae [Institute for Advanced Engineering, Seoul (Korea, Republic of)

    1997-12-31

    To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers. 8 refs., 2 figs., 1 tab. (Author)

  19. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-04-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.

  20. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    1998-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. Finally improvement areas of model development for auditing tool were established based on the identified phenomena. 8 refs., 21 figs., 19 tabs. (Author)

  1. CANDU fuel attribution through the analysis of delayed neutron temporal behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Sellers, M.T.; Corcoran, E.C.; Kelly, D.G., E-mail: David.Kelly@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2012-07-01

    Delayed Neutron Counting (DNC) is an established technique in the Canadian nuclear industry as it is used for the detection of defective fuel in several CANDU reactors and the assay of uranium in geological samples. This paper describes the possible expansion of DNC to the discipline of nuclear forensics analysis. The temporal behaviour of experimentally measured delayed neutron spectra were used to determine the relative contributions of {sup 233}U and {sup 235}U to the overall fissile content present in mixtures with average absolute errors of ±4 %. The characterization of fissile content in current and proposed CANDU fuels (natural UO{sub 2}, thoria and mixed oxide (MOX) based) by DNC analysis is evaluated through Monte Carlo simulations. (author)

  2. A study on the interlink of CANDU safety analysis codes with development of GUI system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. J.; Jeo, Y. J.; Park, Q. C. [Seoul National Univ., Seoul (Korea, Republic of); Kim, H. T.; Min, B. J. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    In order to improve the CANDU safety analysis code system, the interlink of containment analysis code, PRESCON2 to the system thermal hydraulics analysis code, CATHENA, has been implemented with development of the GUI system. Before the GUI development, we partly corrected two codes to optimize on the PC environment. The interlink of two codes could be executed by introducing three interlinking variables, mass flux, mixture enthalpy, and mixture specific volume. To guarantee the robustness of the codes, two codes are extremely linked by using the GUI system. The GUI system provides much of user-friendly functions and will be improved step by step. This study is expected to improve the safety assessment system and technology for CANDU NPPs.

  3. Influences of guide-tube and bluff-body on advanced atmospheric pressure plasma source for single-crystalline polymer nanoparticle synthesis at low temperature

    Science.gov (United States)

    Kim, Dong Ha; Park, Choon-Sang; Kim, Won Hyun; Shin, Bhum Jae; Hong, Jung Goo; Park, Tae Seon; Seo, Jeong Hyun; Tae, Heung-Sik

    2017-02-01

    The use of a guide-tube and bluff-body with an advanced atmospheric pressure plasma source is investigated for the low-temperature synthesis of single-crystalline high-density plasma polymerized pyrrole (pPPy) nano-materials on glass and flexible substrates. Three process parameters, including the position of the bluff-body, Ar gas flow rate, and remoteness of the substrate from the intense and broadened plasma, are varied and examined in detail. Plus, for an in-depth understanding of the flow structure development with the guide-tube and bluff-body, various numerical simulations are also conducted using the same geometric conditions as the experiments. As a result, depending on both the position of the bluff-body and the Ar gas flow rate, an intense and broadened plasma as a glow-like discharge was produced in a large area. The production of the glow-like discharge played a significant role in increasing the plasma energy required for full cracking of the monomers in the nucleation region. Furthermore, a remote growth condition was another critical process parameter for minimizing the etching and thermal damage during the plasma polymerization, resulting in single- and poly-crystalline pPPy nanoparticles at a low temperature with the proposed atmospheric pressure plasma jet device.

  4. Exergy analysis and performance of a counter flow Ranque-Hilsch vortex tube having various nozzle numbers at different inlet pressures of oxygen and air

    Energy Technology Data Exchange (ETDEWEB)

    Kirmaci, Volkan [Bartin University, Faculty of Engineering, Mechanical Engineering Department, 74100 Bartin (Turkey)

    2009-11-15

    An experimental investigation is made to determine the effects of the orifice nozzle number and the inlet pressure on the heating and cooling performance of the counter flow Ranque-Hilsch vortex tube when air and oxygen used as a fluid. The orifices used at these experiments are made of the polyamide plastic material. The thermal conductivity of polyamide plastic material is 0.25 W/m C. Five orifices with nozzle numbers of 2, 3, 4, 5 and 6 have been manufactured and used during the experiments. For each one of the orifices (nozzle numbers) when used with two different fluids, inlet pressures were adjusted from 150 kPa to 700 kPa with 50 kPa increments, and the exergy efficiency was determined. During the experiments, a vortex tube is used with an L/D ratio of 15, and cold mass fraction is held constant at 0.5. As a result of the experimental study, it is determined that the temperature gradient between the hot and cold fluid is decreased with increasing of the orifice nozzle number. (author)

  5. A controllability study of TRUMOX fuel for load following operations in a CANDU-900 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Trudell, D.A., E-mail: trudelda@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    Using a core model of a generic CANDU-900 reactor in RFSP-IST, load following simulations have been performed to assess the controllability of the reactor due to Xenon transients. Week long load following simulations have been performed with daily power cycles 12 hours in duration. Simulations have shown that Natural Uranium fuel can be safely cycled between 100 and 90% Full Power without adjuster rod movement while TRUMOX fuel can be safely cycled between 100 and 85% Full Power. (author)

  6. Extending the world's uranium resources through advanced CANDU fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    De Vuono, Tony; Yee, Frank; Aleyaseen, Val; Kuran, Sermet; Cottrell, Catherine

    2010-09-15

    The growing demand for nuclear power will encourage many countries to undertake initiatives to ensure a self-reliant fuel source supply. Uranium is currently the only fuel utilized in nuclear reactors. There are increasing concerns that primary uranium sources will not be enough to meet future needs. AECL has developed a fuel cycle vision that incorporates other sources of advanced fuels to be adaptable to its CANDU technology.

  7. Validation of WIMS-AECL reactivity device calculations for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Donnelly, J. V. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1997-06-01

    An important component of the overall program to validate WIMS-AECL for use with RFSP in the analysis of CANDU-6 reactors for design and safety analysis calculations is the validation of calculations of incremental cross sections used to represent reactivity devices. A method has been developed for the calculation of the three-dimensional neutron flux distribution in and around CANDU reactor fuel channels and reactivity control devices. The methods is based on one- and two dimensional transport calculations with the WIMS-AECL lattice cell code, SPH homogenization, and three-dimensional flux calculations with finite-difference diffusion theory using the MULTICELL code. Simulations of Wolsung 1 Phase-B commissioning measurements and Point Lepreau restart tests have been performed, as a part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor simulations in RFSP. The incremental cross section properties of the Wolsung 1 and Point Lepreau adjusters, Mechanical Control Absorbers(MCA) and liquid Zone Control Unit (ZCU) is based on the WIMS-AECL/MULTICELL modelling methods and the results are compared with those of WIMS-AECL/DRAGON-2 modelling methods. (author). 13 tabs., 4 figs., 11 refs.

  8. Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  9. Mechanical engineering utilizing advanced engineering tools for the CANDU 9 project

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F.; Yu, S.K.W.; Hedges, K.R. [Atomic Energy of Canada Limited (AECL), Ontario (Canada)

    1998-05-01

    To meet the increasing challenging project requirements such as cost and schedule reduction, AECL has incorporated a comprehensive suite of integrated, advanced engineering tools for CANDU project engineering and delivery. This paper provides a description of the advanced engineering tools developed and used by AECL in the pre-project engineering of the CANDU 9 product and the construction projects such as the construction of two CANDU 6 units in Qinshan, China. The advanced mechanical engineering tools described include: the Process and Instrument Diagram ; the mechanical/piping 3D models; the CADDS/piping analysis interface (PAI) tool; the pipe support design system (SDS) tool; and the powerful equipment database tool - TeddyBase. A description of the enhanced work process will also be provided. The work process improvement is a direct result of the implementation of advanced information technology and the integration of AECL tools with commercial engineering and project tools available in the market. The use of these advanced tools results in better design quality; enhanced presentation of the engineering deliverables to construction and commissioning staff; and potential support to future plant operations (Ref.1). (author). 2 refs., 3 figs.

  10. R and D activities on CANDU-type fuel in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Suripto, A.; Badruzzaman, M.; Latief, A. [Nuclear Fuel Element Centre, National Atomic Energy Agency of Indonesia (BATAN), Puspiptek, Serpong (Indonesia)

    1997-07-01

    The status of R and D activities in Indonesia with respect of CANDU-type fuel development is presented. The activities have been started since the first feasibility study to introduce nuclear power plants was carried out in 1970s. The early research comprised the in-situ pilot production of yellow-cake in Kalimantan (Borneo) experimental mining site, uranium purification and pellet preparation. This program continued to gain a full support from the Government which culminated in the realisation of the construction by BATAN of a large fuel development laboratory in Serpong, starting from 1984 in co-operation with NIRA Ansaldo of Italy. The laboratory, which is called the Power Reactor Experimental Fuel Element Installation (EFEI) was originally designed as an experimental facility to integrate the acquired domestic R and D results gained so far on the CANDU-type fuel technology and the additional know-how received from NIRA Ansaldo which at that time was engaged, in developing a CANDU-type fuel, called the CIRENE fuel design. In the present days the facility houses the power reactor fuel development activities carried out to build up the national capability on power reactor fuel fabrication technology in anticipation to embark upon the nuclear energy era in the near future. (author)

  11. Power distribution control of CANDU reactors based on modal representation of reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Xia, Lingzhi, E-mail: lxia4@uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Jiang, Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Luxat, John C., E-mail: luxatj@mcmaster.ca [Department of Engineering Physics, McMaster University, Hamilton, Ontario L8S 4L7 (Canada)

    2014-10-15

    Highlights: • Linearization of the modal synthesis model of neutronic kinetic equations for CANDU reactors. • Validation of the linearized dynamic model through closed-loop simulations by using the reactor regulating system. • Design of a LQR state feedback controller for CANDU core power distribution control. • Comparison of the results of this new controller against those of the conventional reactor regulation system. - Abstract: Modal synthesis representation of a neutronic kinetic model for a CANDU reactor core has been utilized in the analysis and synthesis for reactor control systems. Among all the mode shapes, the fundamental mode of the power distribution, which also coincides with the desired reactor power distribution during operation, is used in the control system design. The nonlinear modal models are linearized around desired operating points. Based on the linearized model, linear quadratic regulator (LQR) control approach is used to synthesize a state feedback controller. The performance of this controller has been evaluated by using the original nonlinear models under load-following conditions. It has been demonstrated that the proposed reactor control system can produce more uniform power distribution than the traditional reactor regulation systems (RRS); in particular, it is more effective in compensating the Xenon induced transients.

  12. Tracheal tube cuff inflation guided by pressure volume loop closure associated with lower postoperative cuff-related complications: Prospective, randomized clinical trial

    Directory of Open Access Journals (Sweden)

    Waleed A Almarakbi

    2014-01-01

    Full Text Available Background: The main function of an endotracheal tube (ETT cuff is to prevent aspiration. High cuff pressure is usually associated with postoperative complications. We tried to compare cuff inflation guided by pressure volume loop closure (PV-L with those by just to seal technique (JS and assess the postoperative incidence of sore throat, cough and hoarseness. Materials and Methods: In a prospective, randomized clinical trial, 100 patients′ tracheas were intubated. In the first group (n = 50, ETT cuff inflation was guided by PV-L, while in the second group (n. = 50 the ETT cuff was inflated using the JS technique. Intracuff pressures and volumes were measured. The incidence of postoperative cuff-related complications was reported. Results: Demographic data and durations of intubation were comparable between the groups. The use of PV-L was associated with a lesser amount of intracuff air [4.05 (3.7-4.5 vs 5 (4.8-5.5, P < 0.001] and lower cuff pressure than those in the JS group [18.25 (18-19 vs 33 (32-35, P ≤ 0.001]. The incidence of postextubation cuff-related complications was significantly less frequent among the PV-L group patients as compared with the JS group patients (P ≤ 0.009, except for hoarseness of voice, which was less frequent among the PV-L group, but not statistically significant (P ≤ 0.065. Multiple regression models for prediction of intra-cuff pressure after intubation and before extubation revealed a statistically significant association with the technique used for cuff inflation (P < 0.0001. Conclusions : The study confirms that PV-L-guided ETT cuff inflation is an effective way to seal the airway and associates with a lower ETT cuff pressure and lower incidence of cuff-related complications.

  13. Speciation of iodine (I-127) in the natural environment around Canadian CANDU sites

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, S.J.; Kotzer, T.G.; Chant, L.A

    2001-06-01

    In Canada, very little data is available regarding the concentrations and chemical speciation of iodine in the environment proximal and distal to CANDU Nuclear Power Generating Stations (NPGS). In the immediate vicinity of CANDU reactors, the short-lived iodine isotope {sup 131}I (t{sub 1/2} = 8.04 d), which is produced from fission reactions, is generally below detection and yields little information about the environmental cycling of iodine. Conversely, the fission product {sup 129}I has a long half-life (t{sub 1/2} = 1.57x10{sup 7} y) and has had other anthropogenic inputs (weapons testing, nuclear fuel reprocessing) other than CANDU over the past 50 years. As a result, the concentrations of stable iodine ({sup 127}I) have been used as a proxy. In this study, a sampling system was developed and tested at AECL's Chalk River Laboratories (CRL) to collect and measure the particulate and gaseous inorganic and organic fractions of stable iodine ({sup 127}I) in air and associated organic and inorganic reservoirs. Air, vegetation and soil samples were collected at CRL, and at Canadian CANDU Nuclear Power Generating Stations (NPGS) at OPG's (Ontario Power Generation) Pickering (PNGS) and Darlington NPGS (DNGS) in Ontario, as well as at NB Power's Pt. Lepreau NPGS in New Brunswick. The concentrations of particulate and inorganic iodine in air at CRL were extremely low, and were often found to be below detection. The concentrations are believed to be at this level because the sediments in the CRL area are glacial fluvial and devoid of marine ionic species, and the local atmospheric conditions at the sampling site are very humid. Concentrations of a gaseous organic species were comparable to worldwide levels. The concentrations of particulate and inorganic iodine in air were also found to be low at PNGS and DNGS, which may be attributed to reservoir effects of the large freshwater lakes in southern Ontario, which might serve to dilute the atmospheric iodine

  14. Exploring the polymerization of bioactive nano-cones on the inner surface of an organic tube by an atmospheric pressure pulsed micro-plasma jet

    Science.gov (United States)

    Xu, H. M.; Yu, J. S.; Chen, G. L.; Qiu, X. P.; Hu, W.; Chen, W. X.; Bai, H. Y.

    2015-12-01

    In this paper, the successful deposition of acrylic acid polymer (PAA) nano-cones on the inner surface of a polyvinyl chloride (PVC) tube using an atmospheric pressure pulsed plasma jet (APPJ) with acrylic acid (AA) monomer is presented. Optical emission spectroscopy (OES) measurements indicated that various reactive radicals, such as rad OH and rad O, existed in the plasma jet. Moreover, the pulsed current proportionally increased with the increase in the applied voltage. The strengthened stretching vibration of the carbonyl group (Cdbnd O) at 1700 cm-1, shown in the ATR-FTIR spectra, clearly indicated that the PAA was deposited on the PVC surface. The maximum height of the PAA nano-cones deposited by this method ranged from 150 to 200 nm. FTIR and XPS results confirmed the enhanced exposure of the carboxyl groups on the modified PVC surface, which was considered highly beneficial for successfully immobilizing a high density of biomolecules. The XPS data showed that the carbon ratios of the Csbnd OH/R and COOH/R groups increased from 7.03% and 2.6% to 18.69% and 6.81%, respectively (more than doubled) when an Ar/O2 plasma with AA monomer was applied to treat the inner surface of the PVC tube. Moreover, the enhanced attachment density of MC3T3-E1 bone cells was observed on the PVC inner surface coated with PAA nano-cones.

  15. A comparison of intraocular pressure and hemodynamic responses to insertion of laryngeal mask airway or endotracheal tube using anesthesia with propofol and remifentanil in cataract surgery

    Directory of Open Access Journals (Sweden)

    Mohsen Ziyaeifard

    2012-01-01

    Full Text Available Background: The aim of this study was to evaluate intraocular pressure (IOP and hemodynamic responses following insertion of laryngeal mask airway (LMA or endotracheal tube (ETT after anesthesia induction with propofol and remifentanil in cataract surgery. Materials and Methods: In a randomized controlled study, 50 adults scheduled for elective cataract extraction procedure under general anesthesia were allocated to LMA insertion (n = 25 or ETT (n = 25 groups. IOP, systolic blood pressure (SBP, diastolic blood pressure (DBP, and heart rate (HR were measured after insertion of the airway device every minute up to 5 min. Results: There were no significant differences between LMA and ETT groups in SBP, DBP, HR, and IOP immediately after airway instrumentation up to 5 min, except in 4th min in DBP, 2nd min in HR, and 5th min in IOP (7.9 ± 2.3 mmHg in LMA and 9.4 ± 2.5 mmHg in ETT group; P = 0.030. There was good surgeon satisfaction for providing acceptable surgical field in both groups (88% in LMA and 80% in ETT group; P = 0.702. Conclusion: Propofol combined with remifentanil provides good and excellent conditions for insertion of LMA or ETT with minimal hemodynamic disturbances in cataract surgery. Considering LMA insertion is less traumatic than ETT, using LMA may be better than ETT for airway securing in these patients.

  16. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, D.W.; Novog, D.R. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO{sub 2} in ThO{sub 2}) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO{sub 2} (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  17. Manufacturers of Copper Tube for Central Air Conditioner Use Face Mounting Pressure in the Final Quarter of the Year

    Institute of Scientific and Technical Information of China (English)

    2014-01-01

    <正>This year,the investment growth rate of real estate industry continued to slow down,sold area of commercial housing also dropped significantly,which brought huge pressure to the domestic air conditioning manufactures.In the first half of the year,by relying on high growth in national financial expenditure,along with investment in public infrastructure

  18. 变截面管管内传热与阻力性能研究%The Experimental Performance of Heat Transfer and Pressure Drop in Variable Cross-section Tube

    Institute of Scientific and Technical Information of China (English)

    甄亮; 江楠; 徐百平

    2000-01-01

    On the basis of the orthogonal experiment, nine groups of structure parameters of variable cross-section tubes are issued. After the regression analysis of the experimental data, the correlation of heat transfer and pressure drop of variable cross-section tube in the tube side are obtained in this article. The optimal structure parameter of variable cross-section tube is recommended in the last part of the paper.%根据正交实验设计提出了9种变截面管强化传热管的结构尺寸,并由实验结果回归得到了变截面管的传热与阻力关联式.经过管型的分析比较,筛选出性能比较好的变截面管尺寸参数.

  19. Measurement of gap and grain-boundary inventories of {sup 129}I in used CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Moir, D.L.; Kolar, M.; Porth, R.J.; McConnell, J.L.; Kerr, A.H. [AECL Research, Pinawa, Manitoba (Canada). Whiteshell Labs.

    1995-12-31

    Combined gap and grain-boundary inventories of {sup 129}I in 14 used CANDU fuel elements were measured by crushing and simultaneously leaching fuel segments for 4 h in a solution containing KI carrier. From analogy with previous work a near one-to-one correlation was anticipated between the amount of stable Xe and the amount of {sup 128}I in the combined gap and grain-boundary regions of the fuel. However, the results showed that such a correlation was only apparent for low linear power rating (LLPR) fuels with an average linear power rating of < 42 kW/m. For high linear power rating (HLPR) fuels (> 44 kW/m), the {sup 129}I values were considerably smaller than expected. The combined gap and grain-boundary inventories of {sup 129}I in the 14 fuels tested varied from 1.8 to 11.0%, with an average value of 3.6 {+-} 2.4% which suggests that the average value of 8.1 {+-} 1% used in safety assessment calculations overestimates the instant release fraction for {sup 129}I. Segments of used CANDU fuels were leached for 92 d (samples taken at 5, 28 and 92 d) to determine the kinetics of {sup 129}I release. Results could be fitted tentatively to half-order reaction kinetics, implying that {sup 129}I release is a diffusion-controlled process for LLPR fuels, and also for HLPR fuels, once the gap inventory has been leached. However, more data are needed over longer leaching periods to gain more understanding of the processes that control grain-boundary release of {sup 129}I from used CANDU fuel.

  20. Assessment of CANDU reactor physics effects using a simplified whole-core MCNP model

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

    A whole-core Monte Carlo n-particle (MCNP) model of a simplified CANDU reactor was developed and used to study core configurations and reactor physics phenomena of interest in CANDU safety analysis. The resulting reactivity data were compared with values derived from corresponding WIMS-AECL/RFSP, two-neutron-energy-group diffusion theory core simulations, thereby extending the range of CANDU-related code-to-code benchmark comparisons to include whole-core representations. These comparisons show a systematic discrepancy of about 6 mk between the respective absolute k{sub eff} values, but very good agreement to within about -0.15 {+-} 0.06 mk for the reactivity perturbation induced by G-core checkerboard coolant voiding. These findings are generally consistent with the results of much simpler uniform-lattice comparisons involving only WIMS-AECL and MCNP. In addition, MCNP fission-energy tallies were used to evaluate other core-wide properties, such as fuel bundle and total-channel power distributions, as well as intra-bundle details, such as outer-fuel-ring relative power densities and outer-ring fuel element azimuthal power variations, which cannot be determined directly from WIMS-AECL/RFSP core calculations. The average MCNP values for the ratio of outer fuel element to average fuel element power density agreed well with corresponding values derived from WIMS-AECL lattice-cell cases, showing a small systematic discrepancy of about 0.5 %, independent of fuel bum-up. For fuel bundles containing the highest-power fuel elements, the maximum peak-to-average outer-element azimuthal power variation was about 2.5% for cases where a statistically significant trend was observed, while much larger peak-to-average outer-element azimuthal power variations of up to around 42% were observed in low-power fuel bundles at the core/radial-neutron-reflector interface. (author)

  1. THE IMPACT OF POWER COEFFICIENT OF REACTIVITY ON CANDU 6 REACTORS

    Directory of Open Access Journals (Sweden)

    D. KASTANYA

    2013-10-01

    Full Text Available The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The CANDU® reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC and Large Break Loss of Coolant Accident (LBLOCA events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

  2. Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Young; Hwang, Gi Suk; Jung, Yun Sik [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon; Moon, Young Min [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2003-03-15

    The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items : the one is the development of experimental facility for the safety evaluation of CANDU, the other is the results of comparison with the existing correlations and data. The literature reviews are performed and the database for previous off-take experiments are built. By a survey of state-of-the-articles, the deficiencies of previous works and limitations of existing models are examined. The hydraulic behavior branching through the feeder pipes from the header pipe is analyzed and the test facility of off-take experiment is designed and manufactured as the prototype CANDU6, by a proper scaling methodologies. The test facility contains various branch pipes not only for three directions (top, side and bottom), but for arbitrary directions. The experiments about the onset of entrainment and branch quality only for three directions (top, side and bottom) are carried out by using air-water as working fluids. On the whole, the existing correlations predict the present experimental results well branch quality, entrainment, which validates the availability of experimental facility and methodology. Especially, for the branch quality with top and bottom branches, the different results are shown because of the unstable flow regimes in the horizontal pipe and the different branch diameters. The deficiencies of previous works and limitations of existing models are considered. The off-take experiment for arbitrary branch angles continues as the next year research.

  3. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  4. 高效传热管内凝结换热性能及阻力性能的实验研究%EXPERIMENTAL~INVESTIGATION~ ON~ CONDENSATION HEAT TRANSFER COEFFICIENT AND PRESSURE DROP IN HIGH PERFORMANCE HEAT TRANSFER TUBE

    Institute of Scientific and Technical Information of China (English)

    解旭斌; 王维城; 王栋

    2000-01-01

    Using HFC134a and HCFC22 as the working fluid, the comparative experimentalstudies on horizontal condensation in smooth tube and two other augumented heattransfer tubes with different groove shapes (DAE-2 tube and DAEC tube)areconducted. As the experimental results indicated, compared with the smoothtube, the average heat transfer coefficient of DAE-2 is improved by140%to 170% and the pressure drop per unit length increased by 50%to100%. In addition, the average heat transfer coefficient of DAEC tube isenhanced by 160%to 200% and the pressure drop per unit length increased by70% to 130%. Further more, the empirical formulas about average heat transfercoefficient and pressure drop of the DAE-2 tube and DAEC tube are given inthis paper,which can be used to design condensers%本文以HFC134a和HCFC22为工质对光管及两种不同槽型的强化传热管(DAE-2管与DAEC管)的水平管内凝结换热进行了对比实验研究。研究发现,DAE-2管平均换热系数比光管提高了140%170%, 而单位长度阻力损失增加了50%100%, DAEC管平均换热系数比光管提高了160%200%, 同时单位长度阻力损失增加了70%130%。此外,本文给出了DAE-2管和DAEC管平均换热系数及阻力损失的计算关联式,可用于冷凝器设计

  5. Incorporating single detector failure into the ROP detector layout optimization for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kastanya, Doddy, E-mail: Doddy.Kastanya@snclavalin.com

    2015-12-15

    Highlights: • ROP TSP value needs to be adjusted when any detector in the system fails. • Single detector failure criterion has been incorporated into the detector layout optimization as a constraint. • Results show that the optimized detector layout is more robust with respect to its vulnerability to a single detector failure. • An early rejection scheme has been introduced to speed-up the optimization process. - Abstract: In CANDU{sup ®} reactors, the regional overpower protection (ROP) systems are designed to protect the reactor against overpower in the fuel which could reduce the safety margin-to-dryout. In the CANDU{sup ®} 600 MW (CANDU 6) design, there are two ROP systems in the core, each of which is connected to a fast-acting shutdown system. Each ROP system consists of a number of fast-responding, self-powered flux detectors suitably distributed throughout the core within vertical and horizontal flux detector assemblies. The placement of these ROP detectors is a challenging discrete optimization problem. In the past few years, two algorithms, DETPLASA and ADORE, have been developed to optimize the detector layout for the ROP systems in CANDU reactors. These algorithms utilize the simulated annealing (SA) technique to optimize the placement of the detectors in the core. The objective of the optimization process is typically either to maximize the TSP value for a given number of detectors in the system or to minimize the number of detectors in the system to obtain a target TSP value. One measure to determine the robustness of the optimized detector layout is to evaluate the maximum decrease (penalty) in TSP value when any single detector in the system fails. The smaller the penalty, the more robust the design is. Therefore, in order to ensure that the optimized detector layout is robust, the single detector failure (SDF) criterion has been incorporated as an additional constraint into the ADORE algorithm. Results from this study indicate that there

  6. Once-through CANDU reactor models for the ORIGEN2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

  7. Preliminary study of CANDU moderator thermal hydraulics using the CUPID code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gi; Jeong Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Lee, Jae Ryong; Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    When the moderator cooling system fails, moderator may act as to remove decay heat which occurs in fuel. During loss of coolant accident (LOCA), the film boiling occurs in the Calandria tube (CT) because the hot pressure tube would deform into contacting with the calandria tube. And lower subcooling would decrease the margin of the CT to dryout. So, it is important to estimate a local subcooling of the moderator inside the Calandria vessel. However, in order to predict the internal temperature the study of empirical experiments and calculations are needed because only the inlet/outlet temperature can be measured in real reactor. In this study, the internal flow of the moderator was predicted by using the CUPID code, which has been developed in KAERI. The CUPID adopts three dimensional, transient, two phase and three field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two fluid model. The CUPID code shows single phase and two phase flow through two phase flow calculations of virtual can be applied.

  8. Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Constantin, A.; Constantin, M. [Institute for Nuclear Research, Pitesti (Romania)

    2013-07-15

    The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated. (orig.)

  9. Optimization of the self-sufficient thorium fuel cycle for CANDU power reactors

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available The results of optimization calculations for CANDU reactors operating in the thorium cycle are presented in this paper. Calculations were performed to validate the feasibility of operating a heavy-water thermal neutron power reactor in a self-sufficient thorium cycle. Two modes of operation were considered in the paper: the mode of preliminary accumulation of 233U in the reactor itself and the mode of operation in a self-sufficient cycle. For the mode of accumulation of 233U, it was assumed that enriched uranium or plutonium was used as additional fissile material to provide neutrons for 233U production. In the self-sufficient mode of operation, the mass and isotopic composition of heavy nuclei unloaded from the reactor should provide (after the removal of fission products the value of the multiplication factor of the cell in the following cycle K>1. Additionally, the task was to determine the geometry and composition of the cell for an acceptable burn up of 233U. The results obtained demonstrate that the realization of a self-sufficient thorium mode for a CANDU reactor is possible without using new technologies. The main features of the reactor ensuring a self-sufficient mode of operation are a good neutron balance and moving of fuel through the active core.

  10. Development of scaling laws on thermal-hydraulic effect test facility for CANDU-6 moderator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Jung, Y. S.; Kim, N. S. [Handong University, Pohang (Korea, Republic of); Kim, M. W.; Kim, H. J. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The scaling laws on thermal-hydraulic effect test facility for CANDU-6 moderator (HGU-KINS) have been investigated and manufactured. The basic laws are the satisfaction of energy conservation and dimensionless number, Ar and Re, for the similarities of thermal-hydraulic properties. And then the thermal-hydraulic scaling analyses of test facilities, SPEL(1/10 scale) and STERN(1/4 scale), have been identified by the present method. As a result, in the case of SPEL, the energy conservation is confirmed, but the similarities of Ar and the heat density are not considered. In the case of STERN, the energy conservation and the characteristics of Ar were well defined. But the similarity of the heat density is unsatisfied, either. Therefore the present method was applied with 1/8 length scale. For the performance test, CFD analysis has been accomplished by CFX5. The results of flow pattern certifications and variation of axial properties with CANDU show that the present scaling method is acceptable.

  11. The application of Plant Reliability Data Information System (PRINS) to CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, S. W.; Lim, Y. H.; Park, H. C. [Korea Hydro and Nuclear Power Co., Ltd., Naah-ri 260, Yangnam-myun, Gyeongju-si, Gyeong Buk (Korea, Republic of)

    2012-07-01

    As risk-informed applications (RIAs) are actively implanted in the nuclear industry, an issue associated with technical adequacy of Probabilistic Safety Assessment (PSA) arises in its modeling and data sourcing. In Korea, PSA for all Korean NPPs has been completed and KHNP(Korea Hydro and Nuclear Power Plant Company) developed the database called the Plant Reliability Data Information System (PRinS). It has several characteristics that distinguish it from other database system such as NPRDs (INPO,1994), PRIS (IAEA), and SRDF (EdF). This database has the function of systematic data management such as automatic data-gathering, periodic data deposition and updating, statistical analysis including Bayesian method, and trend analysis of failure rate or unavailability. In recent PSA for CANDU reactor, the component failure data of EPRI ALWR URD and Component Reliability Database were preferentially used as generic data set. The error factor for most component failure data was estimated by using the information NUREG/CR-4550 and NUREG/CR-4639. Also, annual trend analysis was performed for the functional losses of components using the statistical analysis and chart module of PRinS. Furthermore, the database has been updated regularly and maintained as a living program to reflect the current status. This paper presents the failure data analysis using PRinS which provides Bayesian analysis on main components in the CANDU reactor. (authors)

  12. Design and Development of a Robotic Crawler for CANDU Fuel Channel Inspection

    Science.gov (United States)

    Shukla, Shivam

    For the design of a new robotic crawler drive unit for CANDU fuel channel inspection, a complete design and screening process was done in order to fulfil the objective of this research. A brief explanation of CANDU reactors is provided along with a discussion of the inspection systems that are currently in use. A study of some existing inspection systems is presented which was used for the development of the new robotic crawler design. A number of concepts were generated which underwent a screening process with the help of two design tools. With the help of these tools, a concept was chosen as the final design and details of it are presented. To demonstrate a proof-of-concept, the physical prototype of the robotic crawler was manufactured and assembled. A speed controller was implemented in the final design of the robotic crawler. A set of test procedures were performed on the final design and the results are discussed. Some improvements that can be done on the final design of the robotic crawler are also discussed in the final section of this thesis.

  13. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  14. Two-Phase condensation Heat Transfer Coefficients Heat Transfer Coefficients and Pressure drops of R-404A for different Condensing Temperatures in a smooth and Micro-Fin Tube

    Directory of Open Access Journals (Sweden)

    DR. S.N. Sapali

    2009-11-01

    Full Text Available Two phase heat transfer coefficients and pressure drops of R-404A in a smooth (8.56 mm ID and micro-fin tube (8.96 mm ID are experimentally investigated. Different from previous studies, the present experiments are performed for different condensing temperatures, with superheating and sub cooling and using hermetically sealed compressor. The test runs are done at average saturated condensing temperatures ranging from 35oC to 60oC. The mass fluxes are between 90 and 800 kg m-2s-1 . The experimental results from both smooth and micro-fin tubes show that the average heat transfer coefficient and pressure drop increases with mass flux but decreases with increasing condensing temperature. The average heat transfer coefficient is 30-210% higher for micro-fin tube than that of smooth tube, with moderate increase in pressure drop ranging from 10-55%. New correlations based on the data gathered during the experimentation for predicting condensation heat transfer coefficients are proposed for wide range of practical applications.

  15. Soret Effect Study on High-Pressure CO2-Water Solutions Using UV-Raman Spectroscopy and a Concentric-Tube Optical Cell

    Energy Technology Data Exchange (ETDEWEB)

    Windisch, Charles F.; McGrail, B. Peter; Maupin, Gary D.

    2012-01-01

    Spatially resolved deep-UV Raman spectroscopy was applied to solutions of CO2 and H2O (or D2O), which were subject to a temperature gradient in a thermally regulated high-pressure concentric-tube Raman cell in an attempt to measure a Soret effect in the vicinity of the critical point of CO2. Although Raman spectra of solutions of CO2 dissolved in D2O at 10 MPa and temperatures near the critical point of CO2 had adequate signal-to-noise and spatial resolution to observe a Soret effect with a Soret coefficient with magnitude of |ST| > 0.03, no evidence for an effect of this size was obtained for applied temperature gradients up to 19oC. The presence of 1 M NaCl did not make a difference. In contrast, the concentration of CO2 dissolved in H2O was shown to vary significantly across the temperature gradient when excess CO2 was present, but the results could be explained simply by the variation in CO2 solubility over the temperature range and not to kinetic factors. For mixtures of D2O dissolved in scCO2 at 10 MPa and temperatures close to the critical point of CO2, the Raman peaks for H2O were too weak to measure with confidence even at the limit of D2O solubility.

  16. Numerical simulation of pressure drop characteristics in a circular tube with self-rotating twisted tape inserts%内置自旋扭带圆管内压降特性的数值模拟研究

    Institute of Scientific and Technical Information of China (English)

    冯振飞; 孙瑞娟; 林清宇

    2013-01-01

      为了直观地描述内置自旋扭带圆管内压降的特性,采用RNG k-ε湍流模型对内置静止扭带、自旋扭带圆管及空管的压降特性进行数值模拟研究,并进行了内置3种型号自旋扭带圆管压降特性的数值模拟和试验研究。研究表明:自旋扭带管的压力降约为空管的2倍,而静止扭带管的压力降差不多达到空管的3倍;在含有自旋扭带或静止扭带的管段内,压力沿轴线方向线性变化,与理论分析趋势一致;扭带压降(流体与自旋扭带的摩擦力引起的压力降)的理论计算值、数值结果均与试验结果比较一致;影响扭带压降的因素是流体轴向流速、扭带宽度及扭带节距;流体轴向流速增大,扭带压降也增大;扭带宽度增大,扭带压降也增大;扭带节距增大,扭带压降略有下降。%In order to visually describe the pressure drop characteristics in a circular tube with self-rotating twisted tape inserts , the RNG k-ε turbulent model was used to simulate the pressure drop characteristics in a circular tube with self-rotating twisted tape , static twisted tape or none inserts . The numerical simulation and experimental study of pressure drop characteristics in a circular tube with 3 types of self-rotating twisted tape inserts were presented .The results indicated that the pres-sure drop of the tube with self-rotating twisted tape is about twice of the empty tube and the pressure drop of the tube with static twisted tape is nearly three times of the empty tube .In the section of tube with self-rotating twisted tape or static twisted tape inserts , the pressure shows a linear variation along the axial direction , which is consistent with the theoretical analysis .The theoretical and nu-merical results of pressure drop caused by friction resistance between the fluid and the twisted tape are consistent with the experimental results .The flow axial velocity , the width and the half

  17. Manually operated piston-driven shock tube

    OpenAIRE

    Reddy, KPJ; Sharath, N

    2013-01-01

    A simple hand-operated shock tube capable of producing Mach 2 shock waves is described. Performance of this miniature shock tube using compressed high pressure air created by a manually operated piston in the driver section of the shock tube as driver gas with air at 1 atm pressure as the test gas in the driven tube is presented. The performance of the shock tube is found to match well with the theoretically estimated values using normal shock relations. Applications of this shock tube named ...

  18. Total hemispherical emissivity of pre-oxidized and un-oxidized Zr-2.5Nb pressure-tube materials at 600 {sup o}C to 1000 {sup o}C under vacuum

    Energy Technology Data Exchange (ETDEWEB)

    Fong, R.W.L.; Paine, M.; Nitheanandan, T., E-mail: randy.fong@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The emissivity of pre-oxidized and un-oxidized pressure-tube specimens has been measured at high temperatures under vacuum. The emissivity values of un-oxidized tube specimens decreased only slightly from 0.34 at 600 {sup o}C to 0.30 at 800 {sup o}C and changed gradually to 0.25 at 1000 {sup o}C. In comparison, the emissivity of pre-oxidized pressure-tube specimens decreased drastically from 0.70 at 600 {sup o}C to 0.35 at 800 {sup o}C, and gradually decreased to 0.25 at 1000 {sup o}C. The oxide layer of the pre-oxidized tube specimens dissolved into the metal matrix when heated to 700 {sup o}C and higher. Using these results, 2 linear correlations were obtained for emissivity with the oxide thickness measured by scanning electron microscopy and secondary ion mass spectroscopy analysis. (author)

  19. Computer-Aided Thermohydraulic Design of TEMA Type E Shell and Tube Heat Exchangers for Use in Low Pressure, Liquid-to-Liquid, Single Phase Applications.

    Science.gov (United States)

    1985-04-01

    subroutine contains data on friction factors correlated by Sieder and Tate for fluids which are being heated or cooled in tubes. The subroutine uses a linear...inter- -". polation algorithm to calculate the friction factor depending on the Reynolds Number of the tube-side fluid. The Sieder and Tate correlated

  20. Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files

    Science.gov (United States)

    Kozier, K. S.; Dyck, G. R.

    2006-04-01

    A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in alternate channels. Results were compared using different room-temperature data files for deuterium, various thermal-scattering-law data files for hydrogen bound in light water and deuterium bound in heavy water, and for pre-ENDF/B-VII and ENDF/B-VI.8 data for uranium. The reactivity differences observed were small (typically <1 mk) and increased with axial neutron leakage.

  1. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore; Ionescu, Drags; Pauna, Eduard [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.

    2012-03-15

    When nuclear power reactors are operated in a load following (LF) mode, the nuclear fuel may be subjected to step changes in power on weekly, daily, or even hourly basis, depending on the grid's needs. Two load following tests performed in TRIGA Research Reactor of Institute for Nuclear Research (INR) Pitesti were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets in the corrosive environment. The 3D finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath at ridge region. This paper summarizes the results of the analytical assessment for SCF and their relation to CANDU fuel performance in LF tests conditions. (orig.)

  2. Waste management issues and their potential impact on technical specifications of CANDU fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Tait, J.C.; Johnson, L.H. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

    1997-07-01

    The technical specifications for the composition of nuclear fuels and materials used in Canada's CANDU reactors have been developed by AECL and materials manufacturers, taking into account considerations specific to their manufacture and the effect of minor impurities on fuel behaviour in reactor. Nitrogen and chlorine are examples of UO{sub 2} impurities, however, where there is no technical specification limit. These impurities are present in the source materials or introduced in the fabrication process and are neutron activated to {sup 14}C and {sup 36}C1, which after {sup 129}I , are the two most significant contributors to dose in safety assessments for the disposal of used fuel. For certain impurities, environmental factors, particularly the safety of the disposal of used fuels, should be taken into consideration when deriving 'allowable' impurity limits for nuclear fuel materials. (author)

  3. Design requirements of a consolidating dry storage module for CANDU spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Ho; Yoon, Jeong Hyoun; Yang, Ke Hyung; Choi, Byung Il; Lee, Heung Young [KHNP/NETEC, Taejon (Korea, Republic of); Cho, Gyu Seong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2003-10-01

    This paper presents a technical description of design requirement document covers the requirements of the MACSTOR/KN-400 module, which is under development to densely accommodate CANDU spent fuels with more efficient way. The design requirement is for the module that will be constructed within a dry storage site after successfully licensed by the regulatory body. This temporary outdoor spent fuel dry storage facility provides for safe storage of spent nuclear fuel after it has been removed from the plant's storage pool after being allowed to decay for a period of at least 6 years. The MACSTOR/KN-400 module is being designed to the envelope of site environmental conditions encountered at the Wolsong station. The design requirements of MACSTOR/KN-400 module meets the requirements of the appropriate Codes and Standards for dry storage of spent fuel from nuclear power reactors such as lOCFR72, and Korea Atomic Energy Act and relevant technical standard.

  4. Structural design concept and static analysis of CANDU spent fuel compact dry storage system

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K. S.; Yang, K. H.; Paek, C. R.; Jung, J. S.; Lee, H. Y. [Korea Hydro and Nuclear Power Company, Taejon (Korea, Republic of)

    2003-07-01

    In this study, an structural design concept on CANDU spent fuel compact dry storage system MACSTOR/KN-400 module has been established with a view to optimally design the structural members of the system. Design loads, loading combination and structural safety criteria of the module were reviewed assuming W olsung Site. The static analysis of the module showed that compressive stress concentration due to dead load and live load occurred around the center of roof slab. Maximum stress resulted from dead load is about twice as much as the stress from live load, and structural behavior of module caused by wind load was not significant. The static analysis results will have influence on the reinforcement bar design of structural members with other structural analyses.

  5. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  6. The Application of Advancements in Computer Technology to the Control and Safety System of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chan, P. S. W. [AECL CANDU/Sheridan Park Research Community, Ontario (Canada)

    1992-04-15

    The present spatial control algorithm in CANDU reactors is based on flux synthesis from a set of parti-coloured harmonic flux modes. The design of the Rop system is also based on parti-coloured flux shapes, including both normal and abnormal reactor operating conditions. The dependency of the control and safety systems on parti-coloured data was necessitated by the slow CPU and by the scarcity of Ram which were available to the computer systems in the early seventies. Recent advancements in high speed microprocessors and high capacity Ram chips enable the development of the Pmfp computer code, which calculates reactor power distribution on-line, using diffusion theory and in-core self-powered flux detector readings as internal boundary conditions. The Pmfp based control and safety systems do not depend on parti-coloured flux shapes or preconceived reactor operating conditions.

  7. Radionuclide Release after End Fitting Failure Accident in CANDU-6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hoon; Kim, Yun Ho; Lee, Kwang Ho [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The total amount of mass and energy discharged into containment building from primary heat transport system in the event of end fitting failure accident in CANDU-6 plant is similar to that of small loss of coolant accident. But the ejection of fuel bundles into fuelling machine room is unique phenomenon and causes radio nuclides release from the physically broken fuel rod to outside containment building. The only objective of containment behavior analysis for end fitting failure event is to assess the amount of radio nuclides release to the ambient atmosphere. Radionuclide release rates in case of end fitting failure with all safety system available, that is containment building is intact, as well as with containment system impairment are analyzed with GOTHIC and SMART code

  8. Optimisation de la gestion du combustible dans les reacteurs CANDU refroidis a l'eau legere

    Science.gov (United States)

    Chambon, Richard

    This research has two main goals. First, we wanted to introduce optimization tools in the diffusion code DONJON, mostly for fuel management. The second objective is more practical. The optimization capabilities are applied to the fuel management problem for different CANDU reactors at refueling equilibrium state. Two kinds of approaches are considered and implemented in this study to solve optimization problems in the code DONJON. The first methods are based on gradients and on the quasi-linear mathematical programming. The method initially developed in the code OPTEX is implemented as a reference approach for the gradient based methods. However, this approach has a major drawback. Indeed, the starting point has to be a feasible point. Then, several approaches have been developed to be more general and not limited by the initial point choice. Among the different methods we developed, two were found to be very efficient: the multi-step method and the mixte method. The second kind of approach are the meta-heuristic methods. We implemented the tabu search method. Initially, it was designed to optimize combinatory variable problems. However, we successfully used it for continuous variables. The major advantage of the tabu method over the gradient methods is the capability to exit from local minima. Optimisation of the average exit burnup has been performed for CANDU-6 and ACR-700 reactors. The fresh fuel enrichment has also been optimized for ACR-700. Results match very well what the reactor physics can predict. Moreover, a comparison of the two totally different types of optimization methods validated the results we obtained.

  9. Seismic Soil-Structure Interaction Analysis of a Consolidated Dry Storage Module for CANDU Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Gon; Yoon, Jeong Hyoun; Cho, Chun Hyung; Lee, Heung Young [Korea Hydro and Nuclear Power Co., Ltd., Taejon (Korea, Republic of); Choi, Kyu Sup; Jeong, In Su; Kim, Jong Soo [KONES Co., Seoul (Korea, Republic of)

    2005-07-01

    The MACSTOR/KN-400 module has been developed as an effective alternative to the existing stand alone concrete canister for dry storage of CANDU spent fuel. The structure is a concrete monolith of 21.67 m long and 12.66 m wide and has a height equal to 7.518 m including the bottom slab. Inside of the concrete module are built 40 storage cylinders accommodating ten 60- bundle dry storage baskets, which are suspended from the top slab and eventually constrained at 10 cm above the bottom slab with horizontal seismic restraints. The main cooling process of the MACSTOR/KN-400 module is by air convection through air inlets and outlets. The civil design parameters, with respect to meteorological and seismic loads applied to the module are identical to those specified for the Wolsung CANDU 3 and 4 plants except for local geologic characteristics. As per USNRC SRP Section 3.7.2 and current US practices, Soil-Structure Interaction (SSI) effect shall be considered for all structures not supported by a rock or rock-like soil foundation materials. An SSI is a very complicated phenomenon of the structure coupled with the soil medium that is usually semi-infinite in extent and highly nonlinear in its behavior. And the effect of the SSI is noticeable especially for stiff and massive structures resting on relatively soft ground. Thus the SSI effect has to be considered in the seismic design of MACSTOR/KN-400 module resting on soil medium. The scope of the this paper is to carry out a seismic SSI analysis of the MACSTOR/KN-400 module, in order to show how much the SSI gives an effect on the structural responses by comparing with the fixed-base analysis.

  10. Update on use of AECL's MACSTOR module at CANDU 6 stations

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Moussalam, G.; Kachef, I. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada)

    2006-07-01

    AECL has contributed to the technology development and implementation of dry spent fuel management facilities in Canada and internationally over the last three decades. During that period, AECL has designed a number of concrete canister models and the MACSTOR module; a medium size air-cooled vault. AECL's dry storage technology was used in Canada, Korea and Romania for the construction of eight large-scale above ground dry storage facilities for CANDU spent fuel. These projects add up to a constructed capacity in excess of 5,000 MgU, that represents a significant share of the total worldwide dry storage capacity. This paper describes basic research and technology developments made at AECL's facilities to develop those dry storage technologies for its own reactors and for the operating CANDU 6 reactors. The current operating status of the facilities using concrete canisters is provided. A description of the MACSTOR 200 modules each having a capacity of 228 MgU that is in use at the Gentilly 2 and Cernavoda stations is provided. The Cernavoda spent fuel management facility was commissioned in 2003. The organisational, licensing, equipment supply and construction aspects that were necessary to deliver this turnkey project by AECL and its Romanian partners in 25 months are described. The paper also provides an outline of the joint program between AECL and KHNP-NETEC to develop the new MACSTOR/KN-400 and provides a description of this module having a capacity of 456 MgU (thus twice the MACSTOR 200 capacity) to be deployed by 2007 at the Wolsong site in Korea. (author)

  11. Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

  12. Fitness for service assessment of degraded CANDU feeder piping-Canadian regulatory expectations

    Energy Technology Data Exchange (ETDEWEB)

    Jin, John C., E-mail: john.jin@cnsc-ccsn.gc.c [Operational Engineering Assessment Div., Canadian Nuclear Safety Commission (Canada); Awad, Raoul [Operational Engineering Assessment Div., Canadian Nuclear Safety Commission (Canada)

    2011-03-15

    Allowance for the continued operation of feeder piping at some Canadian CANDU stations, which is experiencing active degradation mechanisms, has been based primarily on augmented inspection practices and conservative fitness for service assessments. The major degradation mechanisms identified to date are: pipe wall thinning due to Flow Accelerated Corrosion (FAC) and service induced cracking due to Intergranular Cracking due to Stress Corrosion Cracking (SCC) and potentially Low Temperature Creep Cracking (LTCC) mechanisms. Given that currently available industry codes and standards do not provide sufficient guidelines/criteria for assessing the degradation of feeder pipes, the Canadian Nuclear Safety Commission (CNSC) has asked the utilities to establish feeder pipe specific procedures to provide reasonable assurance that the risk associated with the feeder degradation is maintained at an acceptably low level. In response to this requirement, the Canadian CANDU industry has developed and continued to update feeder fitness for service guidelines to provide evaluation procedures and industry standard acceptance criteria for assessing the structural integrity of the feeder pipes. The scope and frequency of inspections are determined based on the results of the fitness for service assessments taking into account the relative susceptibility of feeder pipes to each specific degradation mechanism. While industry practices for the management of degraded feeder pipes have, in general, been complied with the regulatory expectations, outstanding issues still remain. Major regulatory concerns include uncertainties associated with limitations in both the inspection techniques and the mechanistic understanding of the degradation processes, which can impede inspection planning and fitness for service assessments. This paper presents the regulator's view of the current situation with respect to degradation of feeder piping, its implications for nuclear safety and the

  13. Magnesium tube hydroforming

    Energy Technology Data Exchange (ETDEWEB)

    Liewald, M.; Pop, R. [Institute for Metal Forming Technology (IFU), Stuttgart (Germany)

    2008-04-15

    Magnesium alloys reveal a good strength-to-weight ratio in the family of lightweight metals and gains potential to provide up to 30% mass savings compared to aluminium and up to 75 % compared to steel. The use of sheet magnesium alloys for auto body applications is however limited due to the relatively low formability at room temperature. Within the scope of this paper, extruded magnesium tubes, which are suitable for hydroforming applications, have been investigated. Results obtained at room temperature using magnesium AZ31 tubes show that circumferential strains are limited to a maximal value of 4%. In order to examine the influence of the forming temperature on tube formability, investigations have been carried out with a new die set for hot internal high pressure (IHP) forming at temperatures up to 400 C. Earlier investigations with magnesium AZ31 tubes have shown that fractures occur along the welding line at tubes extruded over a spider die, whereby a non-uniform expansion at bursting with an elongation value of 24% can be observed. A maximum circumferential strain of approx. 60% could be attained when seamless, mechanically pre-expanded and annealed tubes of the same alloy have been used. The effect of annealing time on materials forming properties shows a fine grained structure for sufficient annealing times as well as deterioration with a large increase at same time. Hence, seamless ZM21 tubes have been used in the current investigations. With these tubes, an increased tensile fracture strain of 116% at 350 C is observed as against 19% at 20 C, obtained by tensile testing of milled specimens from the extruded tubes. This behaviour is also seen under the condition of tool contact during the IHP forming process. To determine the maximum circumferential strain at different forming temperatures and strain rates, the tubes are initially bulged in a die with square cross-section under plane stress conditions. Thereafter, the tubes are calibrated by using an

  14. Reliability of steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

    Kadokami, E. [Mitsubishi Heavy Industries Ltd., Hyogo-ku (Japan)

    1997-02-01

    The author presents results on studies made of the reliability of steam generator (SG) tubing. The basis for this work is that in Japan the issue of defects in SG tubing is addressed by the approach that any detected defect should be repaired, either by plugging the tube or sleeving it. However, this leaves open the issue that there is a detection limit in practice, and what is the effect of nondetectable cracks on the performance of tubing. These studies were commissioned to look at the safety issues involved in degraded SG tubing. The program has looked at a number of different issues. First was an assessment of the penetration and opening behavior of tube flaws due to internal pressure in the tubing. They have studied: penetration behavior of the tube flaws; primary water leakage from through-wall flaws; opening behavior of through-wall flaws. In addition they have looked at the question of the reliability of tubing with flaws during normal plant operation. Also there have been studies done on the consequences of tube rupture accidents on the integrity of neighboring tubes.

  15. Advanced Alarm Systems: Revision of Guidance and Its Technical Basis

    Science.gov (United States)

    2000-11-01

    Pressurized Water Reactor (Westinghouse) APWR Advanced Pressurized Water Reactor (Mitsubishi) ARP alarm response procedure CAMLS CANDU Annunciation Message...List System CANDU Canadian Deuterium Uranium CE Combustion Engineering CPIAS Critical Parameter Indication and Alarm System CRT cathode ray tube ...bulletins and information notices; inspection and investigative reports; licensee event reports; and Commission papers and their attachments. NRC

  16. Measurement of Velocity and Temperature Profiles in the 1/40 Scaled-Down CANDU-6 Moderator Tank

    Directory of Open Access Journals (Sweden)

    Hyoung Tae Kim

    2015-01-01

    Full Text Available In order to simulate the CANDU-6 moderator circulation phenomena during steady state operating and accident conditions, a scaled-down moderator test facility has been constructed at Korea Atomic Energy Institute (KAERI. In the present work an experiment using a 1/40 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a flow field is visualized with a particle image velocimetry (PIV technique under an isothermal state, and the temperature field is measured using a laser induced fluorescence (LIF technique. A preliminary CFD analysis is also performed to find out the flow, thermal, and heating boundary conditions with which the various flow patterns expected in the prototype CANDU-6 moderator tank can be reproduced in the experiment.

  17. A comprehensive model for in-plane and out-of-plane vibration of CANDU fuel endplate rings

    Energy Technology Data Exchange (ETDEWEB)

    Yu, S.D., E-mail: syu@ryerson.ca; Fadaee, M.

    2016-08-01

    Highlights: • Proposed an effective method for modelling bending and torsional vibration of CANDU fuel endplate rings. • Applied successfully the thick plate theory to curved structural members by accounting for the transverse shear effect. • The proposed method is computationally more efficient compared to the 3D finite element. - Abstract: In this paper, a comprehensive vibration model is developed for analysing in-plane and out-of-plane vibration of CANDU fuel endplate rings by taking into consideration the effects of in-plane extension in the circumferential and radial directions, shear, and rotatory inertia. The model is based on Reddy’s thick plate theory and the nine-node isoparametric Lagrangian plate finite elements. Natural frequencies of various modes of vibration of circular rings obtained using the proposed method are compared with 3D finite element results, experimental data and results available in the literature. Excellent agreement was achieved.

  18. Optimizing in-bay fuel inspection capability to meet the needs of today's CANDU fleet

    Energy Technology Data Exchange (ETDEWEB)

    St-Pierre, J., E-mail: joe.st-pierre@amec.com [AMEC NSS, Toronto, Ontario (Canada); Simons, B. [Stern Laboratories Incorporated, Hamilton, Ontario (Canada)

    2013-07-01

    With the recent return to service of many CANDU units, aging of all others, increasingly competitive energy market and aging hot cell infrastructure - there exists now a greater need for timely, cost-effective and reliable collection of irradiated fuel performance information from fuel bay inspections. The recent development of simple in-bay tools, used in combination with standardized technical specifications, inspection databases and assessment techniques, allows utilities to characterize the condition of irradiated fuel and any debris lodged in the bundle in a more timely fashion and more economically than ever. Use of these tools and 'advanced' techniques permits timely engineering review and disposition of emerging issues to support reliable operation of the CANDU fleet. (author)

  19. Large-break loss-of-coolant accident phenomena identification and ranking table (PIRT) for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.; Snell, V.G.; Sills, H.E.; Langman, V.J.; Boyack, B. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    The Advanced Candu Reactor (ACR) is an evolutionary advancement of the current Candu-6 reactor, aimed at producing electrical power for a capital cost and unit-energy cost significantly less than that of current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by heavy water moderator, as with all Candu reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper is focused on the double-ended guillotine critical inlet header break (CRIHB) loss-of-coolant accident (LOCA) in an ACR reactor, which is considered as a large break LOCA. Large Break LOCA in water-cooled reactors has been used historically as a design basis event by regulators, and it has attracted a very large share of safety analysis and regulatory review. The LBLOCA event covers a wide range of system behaviours and fundamental phenomena. The Phenomena Identification and Ranking Table (PIRT) for LBLOCA therefore provides a good understanding of many of the safety characteristics of the ACR design. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the LOCA phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the final PIRT summary table. (authors)

  20. A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

  1. SARAPAN—A Simulated-Annealing-Based Tool to Generate Random Patterned-Channel-Age in CANDU Fuel Management Analyses

    Directory of Open Access Journals (Sweden)

    Doddy Kastanya

    2017-02-01

    Full Text Available In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the *SIMULATE module of the Reactor Fueling Simulation Program (RFSP code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the *INSTANTAN module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the *INSTANTAN module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Obrutsky, L.S.; Cecco, V.S.; Sullivan, S.P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1997-02-01

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

  3. Alternate tube plugging criteria for steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Cueto-Felgueroso, C.; Aparicio, C.B. [Tecnatom, S.A., Madrid (Spain)

    1997-02-01

    The tubing of the Steam Generators constitutes more than half of the reactor coolant pressure boundary. Specific requirements governing the maintenance of steam generator tubes integrity are set in Plant Technical Specifications and in Section XI of the ASME Boiler and Pressure Vessel Code. The operating experience of Steam Generator tubes of PWR plants has shown the existence of some types of degradatory processes. Every one of these has an specific cause and affects one or more zones of the tubes. In the case of Spanish Power Plants, and depending on the particular Plant considered, they should be mentioned the Primary Water Stress Corrosion Cracking (PWSCC) at the roll transition zone (RTZ), the Outside Diameter Stress Corrosion Cracking (ODSCC) at the Tube Support Plate (TSP) intersections and the fretting with the Anti-Vibration Bars (AVBs) or with the Support Plates in the preheater zone. The In-Service Inspections by Eddy Currents constitutes the standard method for assuring the SG tubes integrity and they permit the monitoring of the defects during the service life of the plant. When the degradation reaches a determined limit, called the plugging limit, the SG tube must be either repaired or retired from service by plugging. Customarily, the plugging limit is related to the depth of the defect. Such depth is typically 40% of the wall thickness of the tube and is applicable to any type of defect in the tube. In its origin, that limit was established for tubes thinned by wastage, which was the predominant degradation in the seventies. The application of this criterion for axial crack-like defects, as, for instance, those due to PWSCC in the roll transition zone, has lead to an excessive and unnecessary number of tubes being plugged. This has lead to the development of defect specific plugging criteria. Examples of the application of such criteria are discussed in the article.

  4. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea. The module provides the benefit of occupying significantly less area than the concrete canisters presently used. The modules are designed for a minimum service life of 50 years. During that period, the spent fuel bundles shall be safely stored. This imposes that failure of a fuel bundle element or unacceptable degradation of an existing defect (from reactor operation) does not occur during the dry storage period. The fuel bundles are stored in an air-filled fuel basket that releases 365 Watts on average and a maximum of 390 Watts when rare fuel loading conditions are postulated. In addition, specific accidental air flow cooling conditions are postulated that consist of 100% blockage of all air inlets on one side of the module. These conditions can generate a peak daily fuel temperature of up to 155{sup o}C during a reference hot summer day during the first year of operation. The fuel temperature decreases over the years and also fluctuates due to daily and seasonal temperature variations. At this temperature, fuel elements with intact Zircaloy sheathing will not experience damage. However, for the few fuel bundle elements that are non-leaktight (less than 1 per 37,000), some re-oxidation of UO{sub 2} into higher oxides such as U{sub 3}O{sub 7} / U{sub 4}O{sub 9} and U{sub 3}O{sub 8} will occur. This latter form of Uranium oxide is undesirable due to its lower density that results in a volumetric increase of the pellet that can overstress the fuel element sheathing. The level of fuel pellet

  5. Incident shock wave attenuation in oscillatory tube and influence on performance of pressure wave refrigerator%振荡管内入射激波衰减及其对冷效应的影响

    Institute of Scientific and Technical Information of China (English)

    郑闽锋; 刘曦; 黄成; 林跃东; 雷晓健; 李学来

    2014-01-01

    The pressure wave refrigerator represents a simple arrangement for gas cooling by its decompression and has many applications in chemical processes and energy transformation. The mechanism of the cooling effect of oscillatory tube is the conversion of the pressure energy of gas to heat through the movement of pressure waves, which are moving shock wave and unsteady expansion wave. In the present paper, the regular pattern of incident shock wave attenuation and its influence on the performance of pressure wave refrigerator are investigated by means of a single-tube set up. In the experiments, the expansion ratio is from 2.0 to 6.0, the relative length of the oscillatory tube L/d is from 87 to 737, and the exciting frequency is from 10 Hz to 240 Hz. The experimental results show that the relative strength of incident shock wave is reduced with the increase of relative position in length x/L because the energy of the reflected shock wave is exhausted by the viscosity and friction of the gas inside the tube. The other reason is the result of the gas in the tube pressurized and heated by the shock wave. The shock wave strength is also influenced by transmission and reflection effects resulted from the reflected shock wave. When the tube is relatively short, the relative strength of incident shock wave is less reduced as the tube length decreases, while the strength of the reflected shock wave at the closed end of the tube increases. The maximum refrigeration efficiencyηmax of the refrigerator increases with the tube length, but the value ofηmax is not affected obviously when the tube length increases to some value. The recommended optimal tube length L/d is 300-435 for the tube in this experiment. It helps to improve the performance of the pressure wave refrigerator under variable work condition when the amplitude of the refrigeration efficiency fluctuation is reduced as the length increases. The relative strength of the incident shock wave attenuation is concerned

  6. 重水堆核电站薄壁压力管的水浸超声检测技术%Ultrasonic Immersion Inspection of Thin-Wall Pressurized Tubes in Heavy Water Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    朱思民; 毕炳荣; 邹廉列; 裴卫农

    2005-01-01

    描述了秦山三期CANDU-6重水堆核电站压力管的内壁划伤情况.针对划伤状况,制定超声探伤检验工艺.采用水浸超声探伤方法对内壁的划伤深度进行估判,阐述了水浸探伤的试验方法与探伤检测过程,同时依据检测结果,对划伤的压力管进行了分析与评价.

  7. Vortex tube optimization theory

    Energy Technology Data Exchange (ETDEWEB)

    Lewins, Jeffery [Cambridge Univ., Magdalene Coll., Cambridge (United Kingdom); Bejan, Adrian [Duke Univ., Dept. of Mechanical Engineering and Materials Science, Durham, NC (United States)

    1999-11-01

    The Ranque-Hilsch vortex tube splits a single high pressure stream of gas into cold and warm streams. Simple models for the vortex tube combined with regenerative precooling are given from which an optimisation can be undertaken. Two such optimisations are needed: the first shows that at any given cut or fraction of the cold stream, the best refrigerative load, allowing for the temperature lift, is nearly half the maximum loading that would result in no lift. The second optimisation shows that the optimum cut is an equal division of the vortex streams between hot and cold. Bounds are obtainable within this theory for the performance of the system for a given gas and pressure ratio. (Author)

  8. Measurement of fluid flow by means of pressure differential devices inserted in circular cross-section conduits running full -- Part 4: Venturi tubes

    CERN Document Server

    International Organization for Standardization. Geneva

    2003-01-01

    ISO 5167-4:2003 specifies the geometry and method of use (installation and operating conditions) of Venturi tubes when they are inserted in a conduit running full to determine the flowrate of the fluid flowing in the conduit. ISO 5167-4:2003 also provides background information for calculating the flow-rate and is applicable in conjunction with the requirements given in ISO 5167-1. ISO 5167-4:2003 is applicable only to Venturi tubes in which the flow remains subsonic throughout the measuring section and where the fluid can be considered as single-phase. In addition, each of these devices can only be used within specified limits of pipe size, roughness, diameter ratio and Reynolds number. ISO 5167-4:2003 is not applicable to the measurement of pulsating flow. It does not cover the use of Venturi tubes in pipes sized less than 50 mm or more than 1 200 mm, or for where the pipe Reynolds numbers are below 20 000. ISO 5167-4:2003 deals with the three types of classical Venturi tubes: cast, machined and rough welde...

  9. Experimental Study on Pressure Drop of Falling Film Flow Across Tube Bundles in Rotated Square Arrangement%转角正方形管束有降膜流动时的压降实验

    Institute of Scientific and Technical Information of China (English)

    刘华; 沈胜强; 陈石; 龚路远; 刘瑞; 陈学

    2013-01-01

    To study the effects of flow resistance in large-scale seawater desalination facility on the performance itself,an experimental setup was built to simulate the steam flow process in a horizontal-tube falling film evaporator,so as to analyze the influence of saturated steam temperature and spray density on the flow resistance in the tube bundle.The new parameter (Reynolds number of spray water) was used to fit the experimental results,and subsequently a pressure drop coefficient formula was obtained for the steam flow across the tube bundle in rotated square arrangement.Results show that for a constant steam flow and spray density,the pressure drop reduces with rising saturated steam temperature,and the error of differential pressure is within-± 15% between predicted value and actual measurement.%为了深入研究大型海水淡化装置中流动阻力对装置性能的影响,建立了大型水平管束降膜流动特性实验台,模拟了水平管降膜蒸发器内蒸汽的流动过程,分析了饱和蒸汽温度和喷淋密度对管束流动阻力的影响,引入新的参数(喷淋雷诺数)对实验数据进行了拟合,得出蒸汽横掠有降膜流动的转角正方形管束的压降系数公式.结果表明:在相同的蒸汽质量流量和喷淋密度下,压降随饱和蒸汽温度的升高而降低;压差预测值与实验值的误差小于±15%.

  10. 负压吸引配合咽鼓管吹张治疗分泌性中耳炎临床经验%Clinical Effect of Negative Pressure Suction with Eustachian Tube in Treatment of Secretory Otitis Media

    Institute of Scientific and Technical Information of China (English)

    马进学

    2013-01-01

      目的:探讨负压吸引配合咽鼓管吹张治疗分泌性中耳炎的临床疗效.方法:回顾性分析我院使用负压吸引配合咽鼓管吹张治疗分泌性中耳炎病例193例并观察其疗效.结果:痊愈164例,显效27例,无效2例,总有效率98.96%.结论:负压吸引配合咽鼓管吹张治疗分泌性中耳炎疗效好,复发率低,患者易于接受,是治疗中耳炎较好的方法之一.%Objective:To explore the clinical effect of negative pressure suction with eustachian tube in treatment of secretory otitis media. Methods:Retrospective analysis of our hospital uses neg-ative pressure suction with eustachian tube in treating secretory otitis media in 193 patients, and ob-serve its curative effect. Results:164 cases were cured, 27 cases markedly effective, 2 cases ineffec-tive, the total efficiency of 98.96%. Conclusion:Negative pressure suction with the eustachian tube in treating secretory otitis media has good curative effect, low recurrence rate, and is easily accept-ed by patients, is one of the better therapy of otitis media.

  11. Dual-tube continuous negative pressure drainage in radical mastectomy%双管持续中心负压引流在乳腺癌根治术中的应用观察

    Institute of Scientific and Technical Information of China (English)

    张刚; 伍万权

    2011-01-01

    Objective: To observe the effect of dual-tube continuous negative pressure drainage on patients receiving radical mastectomy. Methods:Forty-three patients with breast cancer underwent radical mastectomy, and the dual-tube continuous negative pressure drainage was applied. Subcutaneous fluid collection and skin flap necrosis were observed after the operation. Results: Primary healing was reached in 35 cases; subcutaneous fluid collection was detected in 5 cases, 3 of which were in the axilla and 2 in the parasternal;skin flap necrosis occurred in 3 cases,which was mainly in the middle edge of the incision;2 cases suffered from both fluid collection and skin flap necrosis. Conclusions:After radical mastectomy, dual-tube continuous negative pressure drainage can reduce the subcutaneous fluid collection and skin flap necrosis.%目的:观察双管持续中心负压引流在乳腺癌根治术中的应用效果.方法:对43例乳腺癌患者行乳腺癌根治术并双管持续中心负压引流,观察术后皮下积液及皮瓣坏死的情况.结果:术后35例切口一期愈合;并发皮下积液5例,其中位于腋窝处3例,胸骨旁2例;皮瓣坏死3例,主要集中在切口中段边缘,其中皮下积液合并皮瓣坏死2例.结论:乳腺癌根治术后应用双管持续中心负压引流,可减少皮下积液和皮瓣坏死的发生.

  12. The mode of operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available This paper presents the results of calculations for CANDU reactor operation in the thorium fuel cycle. The calculations were performed to estimate feasibility of operation of a heavy-water thermal neutron power reactor in the self-sufficient thorium cycle. The parameters of the active core and the scheme of fuel reloading were considered to be the same as for the standard operation in the uranium cycle. Two modes of operation are discussed in the paper: the mode of preliminary accumulation of 233U and the mode of operation in the self-sufficient cycle. For calculations for the mode of accumulation of 233U, it was assumed that plutonium was used as the additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. The maximum content of 233U in the target channels was about 13 kg/t of ThO2. This was achieved by six year irradiation. The start of reactor operation in the self-sufficient mode requires content of 233U not less than 12 kg/t. For the mode of operation in the self-sufficient cycle, it was assumed that all the channels were loaded with the identical fuel assemblies containing ThO2 and a certain amount of 233U. It was shown that the non-uniform distribution of 233U in a fuel assembly is preferable.

  13. Field measurements of beta ray energy spectra in CANDU nuclear generating stations

    Energy Technology Data Exchange (ETDEWEB)

    Horowitz, Y.S. (Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Physics); Hirning, C.R. (Ontario Hydro, Whitby, ON (Canada)); Yuen, P.S.; Aikens, M.S. (AECL Research, Chalk River, ON (Canada). Chalk River Labs.)

    1994-01-01

    Field measurements of beta ray energy spectra have been carried out at various locations in CANDU nuclear generating stations operated by Ontario Hydro. The beta ray energy spectrometer consists of a 5 cm diameter x 2 cm thick BC-404 plastic scintillator situated behind a 100 [mu]m thick, totally depleted, silicon detector. Photon events are rejected by requiring a coincidence between the two detectors. The spectrometer is capable of measuring electron energies from 125 keV to 3.5 MeV. Beta ray energy spectra have been measured for uncontaminated and contaminated fueling machine components, fueling machine swipes and a reactor containment vault. The degree of protection afforded by various articles of protective clothing has also been investigated for the various fueling machine components. Monte Carlo calculations have been used to estimate beta factors for 100 mg.cm[sup -2] and 240 mg.cm[sup -2] LiF-TLD chips, which are used as 'skin-and 'extremity' dosemeters in the Ontario Hydro Radiation Dosimetry Programme. (Author).

  14. Seismic fragility analysis of a CANDU containment structure for near-fault ground motions

    Energy Technology Data Exchange (ETDEWEB)

    Choi, In Kil; Choun, Young Sun; Seo, Jeong Moon; Ahn, Seong Moon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    The R. G. 1.60 spectrum used for the seismic design of Korean nuclear power plants provides a generally conservative design basis due to its broadband nature. A survey on some of the Quaternary fault segments near Korean nuclear power plants is ongoing. It is likely that these faults will be identified as active ones. If the faults are confirmed as active ones, it will be necessary to reevaluate the seismic safety of the nuclear power plants located near these faults. The probability based scenario earthquakes were identified as near-field earthquakes. In general, the near-fault ground motion records exhibit a distinctive long period pulse like time history with very high peak velocities. These features are induced by the slip of the earthquake fault. Near-fault ground motions, which have caused much of the damage in recent major earthquakes, can be characterized by a pulse-like motion that exposes the structure to a high input energy at the beginning of the motion. It is necessary to estimate the near-fault ground motion effects on the nuclear power plant structures and components located near the faults. In this study, the seismic fragility analysis of a CANDU containment structure was performed based on the results of nonlinear dynamic time-history analyses.

  15. Nonlinear seismic behavior of a CANDU containment building subjected to near-field ground motions

    Energy Technology Data Exchange (ETDEWEB)

    Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. A survey on some of the Quaternary fault segments near Korean nuclear power plants is ongoing. It is likely that these faults will be identified as active ones. If the faults are confirmed as active ones, it will be necessary to reevaluate the seismic safety of the nuclear power plants located near the fault. Near-fault ground motions are the ground motions that occur near an earthquake fault. In general, the near-fault ground motion records exhibit a distinctive long period pulse like time history with very high peak velocities. These features are induced by the slip of the earthquake fault. Near-fault ground motions, which have caused much of the damage in recent major earthquakes, can be characterized by a pulse-like motion that exposes the structure to a high input energy at the beginning of the motion. In this study, nonlinear dynamic time-history analyses were performed to investigate the seismic behavior of a CANDU containment structure subjected to various earthquake ground motions including the near-field ground motions.

  16. Current Status and Characterization of CANDU Spent Fuel for Geological Disposal System Design

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun; Lee, Seung Woo; Cha, Jeong Hun; Choi, Jong Won; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yang [SK Engineering and Construction, Seoul (Korea, Republic of)

    2008-06-15

    Inventories to be disposed of, reference turn up, and source terms for CANDU spent fuel were evaluated for geological disposal system design. The historical and projected inventory by 2040 is expected to be 14,600 MtU under the condition of 30-year lifetime for unit 1 and 40-year lifetime for other units in Wolsong site. As a result of statistical analysis for discharge burnup of the spent fuels generated by 2007, average and stand deviation revealed 6,987 MWD/MtU and 1,167, respectively. From this result, the reference burnup was determined as 8,100 MWD/MtU which covers 84% of spent fuels in total. Source terms such as nuclide concentration for a long-term safety analysis, decay heat, thermo-mechanical analysis, and radiation intensity and spectrum was characterized by using ORIGEN-ARP containing conservativeness in the aspect of decay heat up to several thousand years. The results from this study will be useful for the design of storage and disposal facilities.

  17. Simulation of transient heat transfer in MACSTOR/KN-400 module storing irradiated CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea, the MACSTOR/KN-400. The simulation of transient conditions for AECL's spent fuel dry storage systems, presented in this paper, has not been performed before and is considered a major achievement of the present work. In a fist step, CATHENA was compared to MACSTOR-200 temperature measurements and the accuracy of the results were very good. In a second step, CATHENA was applied to the MACSTOR/KN-400. Four cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter and reduced air flow cases in summer and winter. The maximum local concrete temperatures were predicted to be 63{sup o}C for the off-normal case and 65{sup o}C in the reduced air flow case. The maximum temperature gradients in the concrete are predicted to be 28{sup o}C for the off-normal case and 30{sup o}C in the reduced air flow case, incorporating a 3{sup o}C uncertainty. This paper shows that the maximum temperature for the module is expected to meet the temperature limitations of appropriate standards. (author)

  18. Drop Test of the Candu Spent Fuel Storage Basket in MACSTOR/KN-400

    Energy Technology Data Exchange (ETDEWEB)

    Choi, W.S.; Jeon, J.Y.; Seo, K.S. [KAERI, 1045 Daedeokdaero, Yuseong, Daejeon, 305-353 (Korea, Republic of); Park, J.E.; Yoo, G.S.; Park, W.G. [Korea Hydro Nuclear Power - KHNP (Korea, Republic of)

    2009-06-15

    The MACSTOR/KN-400 of Wolsung power plant in Korea is a dry interim storage facilities. There are 400 long slender cylinders in MACSTOR/KN-400. In one cylinder, ten baskets where Candu spent fuels are loaded are stacked and stored. For this MACSTOR/KN-400 facilities, analyses and tests for the hypothetical accident conditions that might happen during moving and storing baskets into a cylinder were performed. The hypothetical accident conditions to be considered are two cases. One is the case of basket dropping onto the bottom plate of a cylinder. The other is the case of basket dropping onto the other basket top plate stored in the cylinder. For the drop analyses, the case of hanging cylinder and the case of cylinder on the unyielding target surface were considered. Based on the dropping analysis, testing condition was determined as the latter case that is for the cylinder on the target surface. In a basket, 60 dummy fuel bundles are loaded which have the same weight of real spent fuel bundles. On the external surface of the basket, 8 strain gauges and 4 accelerometers were attached for the data acquisition. In order to measure the velocity when a basket impacts, three different devices were utilized. And the impact velocity results were compared and cross-checked. After the dropping tests, helium leak tests were conducted to evaluate the leakage rate. (authors)

  19. Heat transfer analysis of the MACSTOR/KN-400 storage module for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Youn, J. H.; Choi, B. I.; Lee, H. Y. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of)

    2003-10-01

    It was verified through heat transfer analysis that a consolidated dry storage system for CANDU spent fuel, MACSTOR/KN-400 was safe in thermal aspect. In order to validate the computer code of CATHENA which was employed to perform the analysis, the comparison between actual measurement data of MACSTOR-200 at Getilly-2 NPP in Canada and computed values from the code has been carried out. The comparison represented that the computed values acceptably agreed to the measurement data and thus the computer code was verified for its application to MACSTOR/KN-400. The identical K-values(parameter to describe head loss inside the module) and convective heat transfer coefficient of the module obtained by the validation was applied to the heat transfer analysis modelling of MACSTOR/KN-400. The result from the analysis showed that under 40 .deg. C of ambient temperature, maximum average and local temperatures of the concrete module were represented by 53 .deg. C and 69 .deg. C, respectively, which fulfilled well the allowable temperature limit of the concrete structure given by ACI349(American Concrete Institute)

  20. Operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Indian Academy of Sciences (India)

    B R Bergelson; A S Gerasimov; G V Tikhomirov

    2007-02-01

    This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of 233U and mode of operation in self-sufficient cycle. For the mode of accumulation of 233U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. Maximum content of 233U in target channels was estimated to be ∼ 13 kg/t of ThO2. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires 233U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO2 and certain amount of 233U. It is shown that nonuniform distribution of 233U in fuel assembly is preferable.

  1. Parallel CFD simulations of turbulent flows inside a CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Yu, S.D.; Cao, J. [Ryerson Univ., Dept. of Mechanical and Industrial Engineering, Toronto, Ontario (Canada)], E-mail: fabbasia@ryerson.ca

    2008-07-01

    Large Eddy Simulation (LES) is used to study the turbulent flow inside a 43-rod bundle. The two LES models developed in this paper are of dynamic Smagorinsky type, featuring a satisfactory prediction of anisotropic turbulence intensity and frequency. The first model, by taking advantage of the geometric periodicity, deals with one seventh of a rod bundle; it is developed for studying the axial, lateral turbulence intensities and frequencies in the centers of subchannels and narrow-gap regions. The second model, dealing with the full rod bundle inside a pressure tube with nominal eccentricity, is developed for studying the turbulent fluid forces acting on the bundle. In order to accelerate the solution process for the two large CFD models, the parallelized CFD technique is utilized in connection with 24 processors. The numerical results, obtained for a test case (an eight-rod bundle), are in good agreement with those experimental data available in the literature. Numerical simulations of turbulent flow phenomena within subchannels are advantageous since true flow features are difficult or costly to reveal by experiments. (author)

  2. Experimental investigation of friction coefficient in tube hydroforming

    Institute of Scientific and Technical Information of China (English)

    Hyae Kyung YI; Hong Sup YIM; Gun Yeop LEE; Sung Mun LEE; Gi Suk CHUNG; Young-Hoon MOON

    2011-01-01

    The friction coefficient between tube and die in guide zone of tube hydroforming was obtained. In hydroforming, the tube is expanded by an internal pressure against the tool wall. By pushing the tube through tool, a friction force at the contact surface between the tube and the tool occurs. In guiding zone, the friction coefficients between tube and die can be estimated from the measured axial feeding forces. In expansion zone, the friction coefficients between tube and die can be evaluated from the measured geometries of expanded tubes and FE analysis.

  3. Hitachi's proposed DCS solution for new build CANDU EC6 using the G-HIACS unified platform

    Energy Technology Data Exchange (ETDEWEB)

    Tan, D.; Ishii, K.; Otsuka, Y.; Uemura, K., E-mail: daisuke.tan.ye@hitachi.com [Hitachi Ltd., Infrastructure Systems Co., Ibaraki (Japan); Marko, P.E. [Hitachi Power Systems Canada Ltd., Power and Industry Div., Ontario (Canada)

    2013-07-01

    Hitachi Ltd. has developed the safe and secure functional safety DCS controller for potential new build NPP projects in the global market. Hitachi has improved the availability, maintainability, and reliability for its latest DCS systems named G-HIACS. In this latest paper on its DCS product development program, Hitachi would like to report a proposed DCS solution for new build CANDU NSP and BOP based on the G-HIACS Unified Architecture (R800FS/HSC800FS vSAFE Functional Safety Controller and R900/HSC900 General Purpose Controller) hybrid control system. (author)

  4. Effects of Two Types of Nasogastric Tube Fixation on Incidence of Nasal Alar Pressure Ulcers%两种鼻胃管固定方法对鼻翼部压疮发生率的影响

    Institute of Scientific and Technical Information of China (English)

    宋瑞梅; 钱火红; 高青; 刘一; 颜哲; 赵彩霞

    2013-01-01

    目的 观察胃切除术后留置鼻胃管患者鼻翼部压疮的发生情况,寻找更好的鼻胃管固定方法.方法 方便性抽样选取第二军医大学长海医院普外科胃切除术后留置胃管行胃肠减压患者1869例,按患者住院先后时间分为:对照组826例,采用易撕敷料(3M transporeTM white)胶带交叉粘贴于胃管再固定于鼻翼部;观察组1043例,采用黏着性棉布伸缩包带(3M multipore-light brown)“Y”型粘贴于鼻翼部再交叉粘贴于胃管上;统计两组患者留置胃管3~6 d和7~20d时段中鼻翼部压疮发生率.结果 对照组与观察组鼻翼部压疮总体发生率差异有统计学意义(P<0.01),但两组内留置胃管3~6 d和7~20 d鼻翼部压疮发生率的差异无统计学意义(P>0.05).结论 采用黏着性棉布伸缩包带“Y”型固定鼻胃管,可降低鼻胃管留置患者鼻翼部压疮的发生率,值得临床推广应用.%Objective To statistically analyze the incidence of nasal ala pressure ulcers in patients with postoperative nasogastric tube fixation after gastric operation so as to find the causes for nasal ala pressure ulcers. Methods Of 1869 patients undergoing postoperative nasogastric tube fixation after gastric surgery treatment were taken as control group tapped with 3M transpore? white adhesive tape to fix the nasogastric tube on the nasal ala. Then 1043 patients were taken as experimental group tapped with "Y" shaped 3M multipore-light brown adhesive tape to fix the nasogastric tube on the nasal ala. The incidence of nasal ala pressure ulcers in the duration of nasogastric tube from 3 to 6 days and 7 to 20 days were calculated. Results Significant statistical difference was found on the total incidence of nasal ala pressure ulcers between the two groups(P0. 05). Conclusion "Y"shaped 3M multipore-light brown adhesive tape can decrease the incidence of necrosis of nasal ala pressure ulcers.

  5. Dynamic Experimental Study of a Multi—bypass Pulse Tube Refrigerator with Two—bypass Tubes

    Institute of Scientific and Technical Information of China (English)

    YonglinJu; ChaoWang; 等

    1998-01-01

    A dynamic experimental apparatus to measure the instantaneous velocity and pressure in the multibypass pulse tube refrigerator(MPTR) was designed and constructed.Some important experimental results of the instantaneous measurements of the velocity and the pressure in the MPTR with twobypass tubes during actual operation are prsented.The effects of the middle-bypass version on the dynamic pressure and mass flow rate at the cold end of the pulse tube are ev aluated from experimental measurements.DC-flow phenomena are observed in this MPTR.The reasons of the multi-bypass version improved the performance of pulse tube refrigertor are given.

  6. Pulse tube cooler having 1/4 wavelength resonator tube instead of reservoir

    Science.gov (United States)

    Gedeon, David R. (Inventor)

    2008-01-01

    An improved pulse tube cooler having a resonator tube connected in place of a compliance volume or reservoir. The resonator tube has a length substantially equal to an integer multiple of 1/4 wavelength of an acoustic wave in the working gas within the resonator tube at its operating frequency, temperature and pressure. Preferably, the resonator tube is formed integrally with the inertance tube as a single, integral tube with a length approximately 1/2 of that wavelength. Also preferably, the integral tube is spaced outwardly from and coiled around the connection of the regenerator to the pulse tube at a cold region of the cooler and the turns of the coil are thermally bonded together to improve heat conduction through the coil.

  7. Leaching of used CANDU fuel: Results from a 19-year leach test under oxidizing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Johnson, L.H.; Tait, J.C.; McConnell, J.L.; Porth, R.J. [AECL, Pinawa, Manitoba (Canada). Whiteshell Labs.

    1997-12-31

    A fuel leaching experiment has been in progress since 1977 to study the dissolution behavior of used CANDU fuel in aerated aqueous solution. The experiment involves exposure of 50-mm clad segments of an outer element of a Pickering fuel bundle (burnup 610 GJ/kg U; linear and peak power ratings 53 and 58 kW/m, respectively), to deionized distilled water (DDH{sub 2}O, {approximately}2 mg/L carbonate) and tapwater ({approximately}50 mg/L carbonate). In 1992, it was observed that the fuel in at least one of the leaching solutions showed some signs of deterioration and, therefore, in 1993, parts of the fuel samples were sacrificed for a detailed analysis of the physical state of the fuel, using SEM and optical microscopy. Leaching results to date show that even after >6900 days only 5 to 7.7% of the total calculated inventory of {sup 137}Cs has leached out preferentially and that leach rates suggest a development towards congruent dissolution. Total amounts of {sup 137}Cs and {sup 90}Sr leached are slightly larger in tapwater than in DDH{sub 2}O. SEM examinations of leached fuel surface fragments indicate that the fuel surface exposed to DDH{sub 2}O is covered in a needle-like precipitate. The fuel surface exposed to tapwater shows evidence of leaching but no precipitate, likely because uranium is kept in solution by carbonate. Detailed optical and SEM microscopy examinations on fuel cross sections suggest that grain-boundary dissolution in DDH{sub 2}O is not prevalent, and in tapwater appears to be limited to the outer {approximately}0.5 mm (pellet/cladding) region of the fuel. Grain boundary attack seems to be limited to microcracks at or near the surface of the fuel. It thus appears that grain-boundary attack occurs only near the fuel pellet surface and is prevalent only in the presence of carbonate in solution.

  8. Split radius-form blocks for tube benders

    Science.gov (United States)

    Lange, D. R.; Seiple, C. W.

    1970-01-01

    Two-piece, radius-form block permits accurate forming and removing of parts with more than a 180 degree bend. Tube bender can shape flexible metal tubing in applications dealing with plumbing, heating, and pressure transmission lines.

  9. Jejunostomy feeding tube

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/patientinstructions/000181.htm Jejunostomy feeding tube To use the sharing features on this ... vomiting Your child's stomach is bloated Alternate Names Feeding - jejunostomy tube; G-J tube; J-tube; Jejunum ...

  10. Development of P22 Tube Blank Steel for High Pressure Boiler Tube%高压锅炉管用P22管坯钢的开发生产

    Institute of Scientific and Technical Information of China (English)

    袁淑君; 李业才

    2016-01-01

    Based on the technology of steel requirements, the chemical composition and internal control requirements of the steel were designed. Using clean steel technology to control S, P and inclusion, selecting high-quality scrap and molten iron to control As, Sn, Pb, Sb, Bi and other harmful elements, and the Ca treatment can reduce the harm of the inclusion. P22 tube blank steel was produced by the process“EAF-LF-VD-CC”in Laiwu Steel, physical quality inspection shows that the steel is pure, the trace harmful elements are low, the temper brittleness sensitivity coefficient J and the CEF value were controlled ideal, the macrostructure and surface quality of the round billet are better.%依据钢的技术要求,设计了钢的化学成分及内控要求,采用纯净钢技术控制S、P及夹杂物,选用优质废钢和铁水控制As、Sn、Pb、Sb、Bi等有害元素,并通过Ca处理降低夹杂物的危害,莱钢采用EAF-LF-VD-CC工艺流程开发了P22管坯钢。实物质量检测表明,钢质纯净,微量有害元素低,回火脆性敏感系数J、CEF控制理想,圆坯低倍组织和表面质量良好。

  11. Release of [sup 14]C from the gap and grain-boundary regions of used CANDU fuels to aqueous solutions

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Tait, J.C.; Porth, R.J.; McConnell, J.L.; Lincoln, W.J. (Whiteshell Lab., Pinawa, Manitoba (Canada). AECL Research)

    1994-01-01

    This study was undertaken as part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), to measure [sup 14]C inventories of used CANDU fuel. Other objectives were to measure the fraction of the total [sup 14]C inventory that would be instantly released to solution from used CANDU fuels upon sheath failure and to determine if the assumptions made in safety assessment calculations of used fuel waste disposal regarding instant release of [sup 14]C were correct. Results showed that the measured [sup 14]C inventories were a factor of 11.5 [+-] 3.9 lower than the estimated [sup 14]C inventory values used in safety assessment calculations. Measured instant release values for [sup 14]C ranged from 0.06 to 5.04% (of total [sup 14]C inventories) with an average of 2.7 [+-] 1.6%, indicating that instant release fractions for [sup 14]C used in safety assessment calculations (1.2--25%) were overestimated.

  12. Cutting Under Pressure Technology for the Tubing String with Dual Packer in Gas Well%带压切割气井双封分压管柱工艺技术∗

    Institute of Scientific and Technical Information of China (English)

    谢涛; 杨红斌; 徐迎新

    2015-01-01

    复杂多个大直径工具管柱、腐蚀管柱和带喷砂器的管柱带压起钻工艺尚无有效方法,为此,研发了带压切割气井双封分压管柱工艺技术。该工艺技术采用专门的带压切割装置,利用切割刀具在装置内逐段切割,带压起出切割掉的油管,下带压密封捞矛打捞切割掉的油管鱼头,逐级起出复杂管柱。现场试验结果表明,气井带压切割工艺可以实现ø73�0 mm N80油管双封分压管柱的带压切割;对于腐蚀穿孔和双封分压以上复杂管柱可以采取装置内分段切割、密封打捞及分级起出的技术方案。%There is still no effective method of tripping out complex multiple large diameter tool string, corro⁃sion string and string with sandblast under pressure, to address the issue, a technology of cutting dual packer iso⁃lation string in gas well has been developed�A special device capable of cutting string under pressure is used to cut the string by cutter�After tripping out the cut tubing string section, run fishing spear for snubbing operation to fish the tubing string in well, thus, tripping out the complex string piece by piece�Field tests show that the cutting un⁃der pressure technology for gas well is fully capable of cutting the ø73�0 mm N80 tubing string with dual packer un⁃der pressure�For corrosion perforation and the complex string upper dual packer, cutting into piece, fishing with sealing and tripping out piece by piece could be a solution.

  13. A review of qualitative inspection aspects of end fittings in an Indian pressurized heavy water reactor

    OpenAIRE

    Urva Pancholi; Dhaval Dave; Ajay Patel

    2016-01-01

    The paper provides a summarized description of the current state of knowledge and practices used in India, in the qualitative inspection of end fittings – a key component of the fuel channel assembly of a pressurized heavy water reactor (PHWR), generally of a Canadian Deuterium Uranium (CANDU) type. Further it discusses various quality inspection techniques; and the high standards and mechanical precision of the job required, to be accepted as viable nuclear reactor component. The techniqu...

  14. Acoustical studies on corrugated tubes

    Science.gov (United States)

    Balaguru, Rajavel

    Corrugated tubes and pipes offer greater global flexibility combined with local rigidity. They are used in numerous engineering applications such as vacuum cleaner hosing, air conditioning systems of aircraft and automobiles, HVAC control systems of heating ducts in buildings, compact heat exchangers, medical equipment and offshore gas and oil transportation flexible riser pipelines. Recently there has been a renewed research interest in analyzing the flow through a corrugated tube to understand the underlying mechanism of so called whistling, although the whistling in such a tube was identified in early twentieth century. The phenomenon of whistling in a corrugated tube is interesting because an airflow through a smooth walled tube of similar dimensions will not generate any whistling tones. Study of whistling in corrugated tubes is important because, it not only causes an undesirable noise problem but also results in flow-acoustic coupling. Such a coupling can cause significant structural vibrations due to flow-acoustic-structure interaction. This interaction would cause flow-induced vibrations that could result in severe damage to mechanical systems having corrugated tubes. In this research work, sound generation (whistling) in corrugated tubes due to airflow is analyzed using experimental as well as Computational Fluid Dynamics-Large Eddy Simulation (CFD-LES) techniques. Sound generation mechanisms resulting in whistling have been investigated. The whistling in terms of frequencies and sound pressure levels for different flow velocities are studied. The analytical and experimental studies are carried out to understand the influence of various parameters of corrugated tubes such as cavity length, cavity width, cavity depth, pitch, Reynolds numbers and number of corrugations. The results indicate that there is a good agreement between theoretically calculated, computationally predicted and experimentally measured whistling frequencies and sound pressure levels

  15. Experimental Investigation on Heat Transfer and Frictional Characteristics of Shell-and-tube Heat exchangers with Different Baffles and Tubes

    Science.gov (United States)

    Wang, C.; Zhu, J. G.; Sang, Z. F.

    2010-03-01

    In this study, the heat transfer and tube frictional characteristics of the helixchangers (shell-and-tube heat exchanger with helical baffles) with spirally corrugated and smooth tubes and the conventional shell-and-tube heat exchanger with smooth tubes were experimentally obtained. The results show that the helixchangers with the spirally corrugated tube and the smooth tubes enhance the total heat transfer coefficient about 26% and 7% on the average than the segmental baffled heat exchanger. In the tube side, the spirally corrugated tube leads to about 28% average increase on convective heat transfer performance and about 24% average increase on pressure drop than the smooth tube, but its conversion efficiency is still higher. The helical baffle could enhance the shell-side condensation coefficient by 13%, and the spirally corrugated tube could help the helixchanger with it enhance remarkably the condensation performance by 53% than the segmental baffled heat exchanger.

  16. photomultiplier tube

    CERN Multimedia

    photomultiplier tubes. A device to convert light into an electric signal (the name is often abbreviated to PM). Photomultipliers are used in all detectors based on scintillating material (i.e. based on large numbers of fibres which produce scintillation light at the passage of a charged particle). A photomultiplier consists of 3 main parts: firstly, a photocathode where photons are converted into electrons by the photoelectric effect; secondly, a multiplier chain consisting of a serie of dynodes which multiply the number of electron; finally, an anode, which collects the resulting current.

  17. photomultiplier tubes

    CERN Multimedia

    photomultiplier tubes. A device to convert light into an electric signal (the name is often abbreviated to PM). Photomultipliers are used in all detectors based on scintillating material (i.e. based on large numbers of fibres which produce scintillation light at the passage of a charged particle). A photomultiplier consists of 3 main parts: firstly, a photocathode where photons are converted into electrons by the photoelectric effect; secondly, a multiplier chain consisting of a serie of dynodes which multiply the number of electron; finally, an anode, which collects the resulting current.

  18. Heat transfer and pressure drop performance of a finned-tube heat exchanger proposed for use in the NASA Lewis Altitude Wind Tunnel

    Science.gov (United States)

    Vanfossen, G. J.

    1985-01-01

    A segment of the heat exchanger proposed for use in the NASA Lewis Altitude Wind Tunnel (AWT) facility has been tested under dry and icing conditions. The heat exchanger has the largest pressure drop of any component in the AWT loop. It is therefore critical that its performance be known at all conditions before the final design of the AWT is complete. The heat exchanger segment is tested in the NASA Lewis Icing Research Tunnel (IRT) in order to provide an icing cloud environment similar to what will be encountered in the AWT. Dry heat transfer and pressure drop data are obtained and compared to correlations available in the literature. The effects of icing sprays on heat transfer and pressure drop are also investigated.

  19. Performance of multi tubes in tube helically coiled as a compact heat exchanger

    Science.gov (United States)

    Nada, S. A.; El Shaer, W. G.; Huzayyin, A. S.

    2014-12-01

    Multi tubes in tube helically coiled heat exchanger is proposed as a compact heat exchanger. Effects of heat exchanger geometric parameters and fluid flow parameters; namely number of inner tubes, annulus hydraulic diameter, Reynolds numbers and input heat flux, on performance of the heat exchanger are experimentally investigated. Different coils with different numbers of inner tubes, namely 1, 3, 4 and 5 tubes, were tested. Results showed that coils with 3 inner tubes have higher values of heat transfer coefficient and compactness parameter (bar{h} Ah ). Pressure drop increases with increasing both of Reynolds number and number of inner tubes. Correlations of average Nusselt number were deduced from experimental data in terms of Reynolds number, Prandtl number, Number of inner coils tubes and coil hydraulic diameter. Correlations prediction was compared with experimental data and the comparison was fair enough.

  20. Thermal Analysis of CANDU6-Moderator System for Loss of Cooling%CANDU6慢化剂系统丧失冷却情况下的温度分析

    Institute of Scientific and Technical Information of China (English)

    徐珍

    2012-01-01

    The coolant system and moderator system of CANDU6 are independent. The prompt neutrons are moderated as thermal neutrons by the moderator and the continuous nuclear fission in the reactor is maintained. At the same time the moderator system supplies the heat sink for the heat produced by the neutrons moderation. During the in-service maintenance of plant, the standby RCW which will only cool down reactor coolant system operates instead of RCW and can not supply heat sink for moderator system heat exchanger. As the result, the moderator system will lose heat sink during the operation of standby RCW. To estimate the moderator temperature, was compared with the experiment data and evaluated in this paper. the thermal analysis of moderator system for loss of cooling the system failure caused by the temperature raising wasevaluated in this paper.%压力管卧式重水反应堆(CANDU6)具有相互独立的冷却剂系统和慢化剂系统。慢化剂系统将堆芯高能裂变中子慢化到能维持持续裂变所需的热中子水平,并将慢化中子过程中产生的热量带出。在反应堆大修期间,需要对再循环冷却水系统(RCW)进行检修,则需要并投入其备用系统,但是RCW备用系统仅对反应堆冷却剂系统进行冷却,不提供慢化剂系统热交换器冷却水。所以在RCW备用系统投入的情况下,慢化剂系统丧失冷却。为判断在此情况下慢化剂的温度变化情况,本文对CANDU6大修期间慢化剂系统丧失冷却情况下的温度变化进行分析并与试验结果进行比较,评估是否会由于温度过高而导致系统失效。

  1. Sound absorption and reflection with coupled tubes

    NARCIS (Netherlands)

    Eerden, van der Frits

    2000-01-01

    This paper describes a special sound absorbing technique with an accompanying efficient numerical design tool. As a basis pressure waves in a single narrow tube or pore are considered. In such a tube the viscosity and the thermal conductivity of the air, or any other fluid, can have a significant ef

  2. A tube-in-tube thermophotovoltaic generator

    Energy Technology Data Exchange (ETDEWEB)

    Ashcroft, J.; Campbell, B.; Depoy, D.

    1996-12-31

    A thermophotovoltaic device includes at least one thermal radiator tube, a cooling tube concentrically disposed within each thermal radiator tube and an array of thermophotovoltaic cells disposed on the exterior surface of the cooling tube. A shell having a first end and a second end surrounds the thermal radiator tube. Inner and outer tubesheets, each having an aperture corresponding to each cooling tube, are located at each end of the shell. The thermal radiator tube extends within the shell between the inner tubesheets. The cooling tube extends within the shell through the corresponding apertures of the two inner tubesheets to the corresponding apertures of the two outer tubesheets. A plurality of the thermal radiator tubes can be arranged in a staggered or an in-line configuration within the shell.

  3. Development boiling to sprinkled tube bundle

    Directory of Open Access Journals (Sweden)

    Kracík Petr

    2016-01-01

    Full Text Available This paper presents results of a studied heat transfer coefficient at the surface of a sprinkled tube bundle where boiling occurs. Research in the area of sprinkled exchangers can be divided into two major parts. The first part is research on heat transfer and determination of the heat transfer coefficient at sprinkled tube bundles for various liquids, whether boiling or not. The second part is testing of sprinkle modes for various tube diameters, tube pitches and tube materials and determination of individual modes’ interface. All results published so far for water as the falling film liquid apply to one to three tubes for which the mentioned relations studied are determined in rigid laboratory conditions defined strictly in advance. The sprinkled tubes were not viewed from the operational perspective where there are more tubes and various modes may occur in different parts with various heat transfer values. The article focuses on these processes. The tube is located in a low-pressure chamber where vacuum is generated using an exhauster via ejector. The tube consists of smooth copper tubes of 12 mm diameter placed horizontally one above another.

  4. Development boiling to sprinkled tube bundle

    Science.gov (United States)

    Kracík, Petr; Pospíšil, Jiří

    2016-03-01

    This paper presents results of a studied heat transfer coefficient at the surface of a sprinkled tube bundle where boiling occurs. Research in the area of sprinkled exchangers can be divided into two major parts. The first part is research on heat transfer and determination of the heat transfer coefficient at sprinkled tube bundles for various liquids, whether boiling or not. The second part is testing of sprinkle modes for various tube diameters, tube pitches and tube materials and determination of individual modes' interface. All results published so far for water as the falling film liquid apply to one to three tubes for which the mentioned relations studied are determined in rigid laboratory conditions defined strictly in advance. The sprinkled tubes were not viewed from the operational perspective where there are more tubes and various modes may occur in different parts with various heat transfer values. The article focuses on these processes. The tube is located in a low-pressure chamber where vacuum is generated using an exhauster via ejector. The tube consists of smooth copper tubes of 12 mm diameter placed horizontally one above another.

  5. 4333M4超高压反应管的断裂韧度%Fracture Toughness of Ultrahigh Pressure Reaction Tube Made by 4333M4 Steel

    Institute of Scientific and Technical Information of China (English)

    李铜; 陈进

    2012-01-01

    The critical stress intensity factor Kit: of high strength low alloy steel 4333M4 was determinated through non-standard arch bone bending experiment, and flexibility method was used. Then the tube was assessed by the leak before break criterion. The results showed that the fracture toughness of the steel was measured up to standard of SFAC 98--01 Rev. 2 and the pipe met requirements of leak before break.%通过非标准拱形三点弯曲试样的试验,采用柔度法标定,测定了进口4333M4低合金超高强度钢管材的临界应力强度因子Kzc,并对该超高压反应管进行了先漏后破分析。结果表明:该钢材有较高的断裂韧度,符合SFAC98-01Rev.2《超高压用无缝合金钢管技术条件》的要求,且管材符合先漏后破标准。

  6. Effect of initial pressure on propagation of detonation wave in round tube%初始压力对爆轰波在管道内传播的影响

    Institute of Scientific and Technical Information of China (English)

    喻健良; 高远; 闫兴清; 高伟

    2014-01-01

    Detonation tube was built to investigate the effect of initial pressure on the propagation of detonation wave in round tube.The premixed gas of CH4+2O2 was selected as experimental gas. Optical fiber probe was used to measure the local velocity of detonation wave.Smoked foils were used to register the cellular structure of detonation wave in tubes.The experimental results show that there are five distinct modes during the propagation of detonation wave in tubes,which are stable mode,rapid fluctuation mode,stuttering mode,galloping mode and failure mode.Under the mode of stable detonation,the fluctuations of the local velocity of detonation wave are generally small and the averaged velocity of detonation wave is close to the theoretical CJ value.The detonation wave has multi-headed cellular structure.With decreasing of the initial pressure,the fluctuations of the local velocity of detonation wave increase,and the averaged velocity of detonation wave decreases.For the galloping detonation,at the decoupled position,cellular structure disappears.Cellular structure forms again when overdriven detonation occurs.If the initial pressure is further decreased till the detonation failure,no cellular structure is observed.%建立爆轰管道研究不同初始压力下爆轰波在管道内传播规律。选用 CH4+2O2气体,采用光纤探针测量爆轰波在管道内的传播速度,采用烟迹法记录爆轰波胞格结构。结果表明:爆轰波在管道内传播时出现5种不同传播模式,分别为稳态式、快速波动式、结巴式、驰振式与失效模式。在稳态传播模式下,爆轰波局部速度波动很小且平均速度接近理论爆轰 CJ 速度,并呈现多头胞格结构。随着初始压力的降低,爆轰波局部速度波动增加且其平均速度产生衰减。在驰振式爆轰解耦处,爆轰波胞格结构消失,过载爆轰时,重新形成胞格结构。进一步降低初始压力至爆轰失效时,则无胞格结构。

  7. Enhancenment of In-tube Condensation of Nonazeotropic Refrigerant Mixtures with a Micro-fin Tube

    Science.gov (United States)

    Koyama, Shigeru; Gao, Lei; Hujii, Tetsu

    Condensation heat transfer and pressure drop characteristics of nonazeotropic refrigerant mixtures (NARMs) of HCFC22 and CFC114 inside horizontal smooth and micro-fin tubes are experimentally investigated. The local Nusselt number for both tubes based on the temperature difference between bulk refrigerant and tube wall is compared. Data for the micro-fin tube are about 50% higher than those for the smooth tube in both cases of pure refrigerants and NARMs. In case of NARMs, however, the decrease of Nesselt number due to vapor mass transfer resistance is observed for the micro-fin tube as well as for the smooth tube. By assuming that heat transfer characteristics of the condensate of NARMs are similar to those for pure refrigerants, the vapor mass transfer coefficient of NARMs is evaluated. Comparison of frictional pressure drop for both tubes is also performed in the Lockharte-Martinelli relation. Data for the micro-fin tube are higer than those for the smooth tube. The maximum increase of the frictional pressure drop is about 30%.

  8. Finned Small Diameter Tube Heat Exchanger

    Science.gov (United States)

    Dang, Chaobin; Daiguji, Hirofumi; Hihara, Eiji; Tokunaga, Masahide

    The performance of fined small tube heat exchangers was investigated both experimentally and theoretically. The Inner diameters of tubes were 1.0mm, 2.1mm and 4.0mm. Exchanged heat and pressure drop obtained from numerical simulation agreed well with the experimental ones. Calculation results show that the volume of a 2.0mm tube heat exchanger can be reduced to 33% of that of a 4mm tube heat exchanger with the same capacity. In addition the distribution of two-phase flow in a branching unit was investigated by measuring downstream temperature distribution. The flow distribution in a branching unit strongly affects the exchanged heat.

  9. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  10. Feasibility study in aspect of thermal integrity on the dry storage expansion options for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y.; Song, M. J. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of); Cho, K. S. [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-10-01

    In order to expand the capability of the CANDU spent fuel dry storage facilities of the at Wolsong, the alternative concepts based on MACSTOR are suggested to replace with existing concrete silo of Wolsong. For this, the feasibility of its design changes from original MACSTOR is examined in term of heat transfer and thermal hydraulic. In this study, the configuration of the module was conceptually changed from its original 2 rows to 3 and 4 rows for review. Under normal operation, the results of heat transfer and thermal hydraulic shows that storage module can feasibly accomodate four rows of storage cylinders within allowable range in terms of maximum allowable temperature of the fuel basket.

  11. Modelling material effects on flow-accelerated corrosion in primary CANDU coolant and secondary reactor feed-water

    Energy Technology Data Exchange (ETDEWEB)

    Phromwong, P.; Lister, D., E-mail: c7r13@unb.ca [Univ. of New Brunswick, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada); Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

    2012-07-01

    The effects of chromium content on flow-accelerated corrosion (FAC) of carbon steel have been predicted very well by including a passivating layer, which is a chromium-dependent diffusion barrier at the metal-oxide interface. By adjusting the properties of the chromium-dependent layer, described with a Passivation Parameter (PP), we can predict the FAC of carbon steel of different chromium contents in typical reactor feed-water environments (140{sup o}C and neutral or ammoniated chemistry). The model and an appropriate PP are also applied to the environment typical of carbon-steel feeders in the primary coolant of a CANDU reactor (310{sup o}C and lithiated chemistry). The model predicts FAC rate very well (with a deviation of 10% or less) in both situations. (author)

  12. Confined Tube Crimp Using Portable Hand Tools

    Energy Technology Data Exchange (ETDEWEB)

    Reynolds, Joseph James [Los Alamos National Laboratory; Pereyra, R. A. [LANL Retired; Archuleta, Jeffrey Christopher [Los Alamos National Laboratory; Martinez, Isaac P. [Los Alamos National Laboratory; Nelson, A. M. [MST-16 Summer Student (2007); Allen, Ronald Scott [Los Alamos National Laboratory; Page, R. L. [LANL Retired; Freer, Jerry Eugene [Los Alamos National Laboratory; Dozhier, Nathan Gus [Los Alamos National Laboratory

    2016-04-04

    The Lawrence Radiation Laboratory developed handheld tools that crimp a 1/16 inch OD tube, forming a leak tight seal1 (see Figure 1). The leak tight seal forms by confining the 1/16 inch OD tubing inside a die while applying crimp pressure. Under confined pressure, the tube walls weld at the crimp. The purpose of this study was to determine conditions for fabricating a leak tight tube weld. The equipment was used on a trial-and-error basis, changing the conditions after each attempt until successful welds were fabricated. To better confine the tube, the die faces were polished. Polishing removed a few thousandths of an inch from the die face, resulting in a tighter grip on the tubing wall. Using detergent in an ultrasonic bath, the tubing was cleaned. Also, the time under crimp pressure was increased to 30 seconds. With these modifications, acceptable cold welds were fabricated. After setting the conditions for an acceptable cold weld, the tube was TIG welded across the crimped face.

  13. 高压高产气井试气求产过程中油管挤毁失效分析%Failure Analysis on the Collapsed Tubing of the High Pressure and High Production Gas Well in the Process of Testing and Production

    Institute of Scientific and Technical Information of China (English)

    薛继军; 赵滨; 赵赫

    2014-01-01

    One of the tubing used in the gas well under the high pressure and high production has occurred in the split,fall,collapse and severe diapirism of tube body in the trial extraction process.This paper have a test of tube so that certain the reason that split,fall,collapsed and se-vere diapirism of tube body by the analysis of the macroscopic,physical and chemical properties, microscopic metallographic.The result shows that gas,liquid pipe ram and solid multiphase flow in tube may be formatted the instantaneous negative pressure and increase the likelihood of the tubing were collapsed in the process of the gas testing and production.Then it has resulted in clogging,severe erosion,thinning wall and piercing in tube.Finally the tube has occurred in frac-ture because of the tensile strength.%某高压高产气井试气用油管在试提中发生管体断裂掉落,并有管体挤毁和严重刺穿现象。通过对油管进行宏观分析、理化性能分析、微观金相分析试验,以确定油管发生断裂掉落以及挤毁和刺穿的原因。结果表明:在放喷求产过程中油管内存在气、液、固等多相流,可能会形成瞬时负压,增加了油管被挤毁变形的可能性;随后造成流道堵塞,油管内冲蚀严重,壁厚变薄发生刺穿,最后因拉伸强度不足而发生断裂。

  14. Probabilistic seismic safety assessment of a CANDU 6 nuclear power plant including ambient vibration tests: Case study

    Energy Technology Data Exchange (ETDEWEB)

    Nour, Ali [Hydro Québec, Montréal, Québec H2L4P5 (Canada); École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada); Cherfaoui, Abdelhalim; Gocevski, Vladimir [Hydro Québec, Montréal, Québec H2L4P5 (Canada); Léger, Pierre [École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada)

    2016-08-01

    Highlights: • In this case study, the seismic PSA methodology adopted for a CANDU 6 is presented. • Ambient vibrations testing to calibrate a 3D FEM and to reduce uncertainties is performed. • Procedure for the development of FRS for the RB considering wave incoherency effect is proposed. • Seismic fragility analysis for the RB is presented. - Abstract: Following the 2011 Fukushima Daiichi nuclear accident in Japan there is a worldwide interest in reducing uncertainties in seismic safety assessment of existing nuclear power plant (NPP). Within the scope of a Canadian refurbishment project of a CANDU 6 (NPP) put in service in 1983, structures and equipment must sustain a new seismic demand characterised by the uniform hazard spectrum (UHS) obtained from a site specific study defined for a return period of 1/10,000 years. This UHS exhibits larger spectral ordinates in the high-frequency range than those used in design. To reduce modeling uncertainties as part of a seismic probabilistic safety assessment (PSA), Hydro-Québec developed a procedure using ambient vibrations testing to calibrate a detailed 3D finite element model (FEM) of the containment and reactor building (RB). This calibrated FE model is then used for generating floor response spectra (FRS) based on ground motion time histories compatible with the UHS. Seismic fragility analyses of the reactor building (RB) and structural components are also performed in the context of a case study. Because the RB is founded on a large circular raft, it is possible to consider the effect of the seismic wave incoherency to filter out the high-frequency content, mainly above 10 Hz, using the incoherency transfer function (ITF) method. This allows reducing significantly the non-necessary conservatism in resulting FRS, an important issue for an existing NPP. The proposed case study, and related methodology using ambient vibration testing, is particularly useful to engineers involved in seismic re-evaluation of

  15. Feeding tube - infants

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/007235.htm Feeding tube - infants To use the sharing features on this page, please enable JavaScript. A feeding tube is a small, soft, plastic tube placed ...

  16. Mechanical Instabilities of Biological Tubes

    Science.gov (United States)

    Hannezo, Edouard; Prost, Jacques; Joanny, Jean-François

    2012-07-01

    We study theoretically the morphologies of biological tubes affected by various pathologies. When epithelial cells grow, the negative tension produced by their division provokes a buckling instability. Several shapes are investigated: varicose, dilated, sinuous, or sausagelike. They are all found in pathologies of tracheal, renal tubes, or arteries. The final shape depends crucially on the mechanical parameters of the tissues: Young’s modulus, wall-to-lumen ratio, homeostatic pressure. We argue that since tissues must be in quasistatic mechanical equilibrium, abnormal shapes convey information as to what causes the pathology. We calculate a phase diagram of tubular instabilities which could be a helpful guide for investigating the underlying genetic regulation.

  17. Orifice plates and venturi tubes

    CERN Document Server

    Reader-Harris, Michael

    2015-01-01

    This book gives the background to differential-pressure flow measurement and goes through the requirements explaining the reason for them. For those who want to use an orifice plate or a Venturi tube the standard ISO 5167 and its associated Technical Reports give the instructions required.  However, they rarely tell the users why they should follow certain instructions.  This book helps users of the ISO standards for orifice plates and Venturi tubes to understand the reasons why the standards are as they are, to apply them effectively, and to understand the consequences of deviations from the standards.

  18. Closed End Launch Tube (CELT)

    Science.gov (United States)

    Lueck, Dale E.; Immer, Christopher D.

    2004-02-01

    A small-scale test apparatus has been built and tested for the CELT pneumatic launch assist concept presented at STAIF 2001. The 7.5 cm (3-inch) diameter × 305 M (1000 feet) long system accelerates and pneumatically brakes a 6.35 cm diameter projectile with variable weight (1.5 - 5 Kg). The acceleration and braking tube has been instrumented with optical sensors and pressure transducers at 14 stations to take data throughout the runs. Velocity and pressure profiles for runs with various accelerator pressures and projectile weights are given. This test apparatus can serve as an important experimental tool for verifying this concept.

  19. Successful tubes treatment of esophageal fistula

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Aim: To discuss the merits of "tubes treatment" for esophageal fistula (EF). Methods: A 66-year-old female who suffered from a bronchoesophageal and esophagothoratic fistula underwent a successful "three tubes treatment" (close chest drainage, negative pressure suction at the leak, and nasojejunal feeding tube), combination of antibiotics, antacid drugs and nutritional support. Another 55-year-old male patient developed an esophagopleural fistula (EPF) after esophageal carcinoma operation. He too was treated conservatively with the three tubes strategy as mentioned above towards a favorable outcome. Results:The two patients recovered with the tubes treatment, felt well and became able to eat and drink, presenting no complaint. Conclusion: Tubes treatment is an effective basic way for EF. It may be an alternative treatment option.

  20. Experimental Investigation on the Heat Transfer and Pressure Drop Characteristics of Refrigerant-oil Mixture Flow Condensation inside Small Diameter Copper Tubes%小管径铜管内含油制冷剂流动冷凝换热与压降特性的实验研究

    Institute of Scientific and Technical Information of China (English)

    胡海涛; 丁国良; 黄翔超; 朱禹; 高屹峰; 郑永新; 宋吉

    2012-01-01

      The heat transfer and pressure drop characteristics of R410A-oil mixture flow condensation inside small diameter copper tubes are performed experimentally. The test tubes include a smooth tube and a microfin tube, the diameters of which are both 5 mm. The test results show that, the presence of oil always deteriorates the heat transfer for smooth and microfin tubes, and the deterioration effects for smooth tube and microfin tube are 24.8% and 25.1%, respectively. The effects of oil on pressure drop are different for smooth tube and microfin tube. For the smooth tube, the present of oil always decrease the pressure drop, and the maximum of the decreasement effect is 19%. While for the microfin tube, when the vapor quality is smaller than 0.6, the presence of oil decreases the pressure drop by a maximum of 18%; when the vapor quality is larger than 0.6, the presence of oil increases the pressure drop by a maximum of 9%. Comparing with 5 mm smooth tube, the heat transfer coefficient and pressure drop of R410A-oil mixture flow condensation in 5 mm microfin tube increases by 60%-130% and 40%-65%, respectively, due to the presence of microfins.%  实验研究了小管径铜管内 R410A-油混合物的流动冷凝换热与压降特性.测试管为外径为5mm的光管和强化管.实验结果表明,润滑油的存在总是恶化5mm 光管和强化管内的换热特性,最大分别恶化24.8%和25.1%.润滑油的存在对光管和强化管内的冷凝压降影响不同.对于光管,润滑油总是降低冷凝压降,最大降低19%.对于强化管,干度小于0.6时,润滑油的存在降低强化管内的压降,最大降低18%;干度大于0.6时,润滑油的存在增大强化管内的压降,最大增强9%.相同工况下,5 mm 强化管与光管相比,换热系数增大60%~130%、压降增大40%~65%

  1. Numerical Simulation of Pressure Drop and Flow Filed in the Heat Exchanger Tube Inserted Self-rotating Straight-tooth Twisted Tape%内插自旋直齿扭带换热管压降及流场数值模拟

    Institute of Scientific and Technical Information of China (English)

    刘宜仔; 林清宇; 冯振飞; 欧向波; 刘晓林

    2014-01-01

    In order to visually describe the characteristics of pressure drop and flow filed self-ro-tating straight-tooth twisted tape inserted in heat exchanger tube,the computer model for the in-ner basin of heat exchanger tube was established,in which self-rotating straight-tooth twisted tape is inserted with water as the medium,and numerical simulation was taken to study the char-acteristics of pressure drop and flow filed in tube by using RNG k~εturbulence model.The re-sults of numerical simulation indicated that the pressure dropΔp of the tube with twisted tape in-to a parabola relationship with the flow velocity u,the faster of the flow velocity,the larger of the pressure drop,and it was greater than that of smooth tube.The fluid was a rule of three-di-mensional spiral flow in the tube with twisted tape,making the fluid form a secondary vortex flow.The twisted tape with straight serration structure enhanced the turbulence of fluid.In the region near tube wall,the flow axial velocity of the tube with twisted tape was averagely in-creased more in the range of 33.05% ~ 35.17% than that of smooth tube.%为了直观地描述内插自旋直齿扭带管内压降及流场的特性,建立了以水为介质的内插自旋直齿扭带换热管内流场的计算模型,采用RNG k~ε湍流模型对管内压降及流场的特性进行数值模拟研究。计算结果表明,扭带管的压降Δp与流体的流速u成抛物线关系,压降随着流速的增大而增大,并且大于空管的压降;扭带管的流体呈有规律的三维螺旋状流动,且存在二次流,扭带的直齿结构增强了流体的湍流程度;在近管壁区域,扭带管的流体轴向速度比空管平均提高幅度在33.05%~35.17%。

  2. Hybrid endotracheal tubes

    Science.gov (United States)

    Sakezles, Christopher Thomas

    Intubation involves the placement of a tube into the tracheal lumen and is prescribed in any setting in which the airway must be stabilized or the patient anesthetized. The purpose of the endotracheal tube in these procedures is to maintain a viable airway, facilitate mechanical ventilation, allow the administration of anesthetics, and prevent the reflux of vomitus into the lungs. In order to satisfy these requirements a nearly airtight seal must be maintained between the tube and the tracheal lining. Most conventional endotracheal tubes provide this seal by employing a cuff that is inflated once the tube is in place. However, the design of this cuff and properties of the material are a source of irritation and injury to the tracheal tissues. In fact, the complication rate for endotracheal intubation is reported to be between 10 and 60%, with manifestations ranging from severe sore throat to erosion through the tracheal wall. These complications are caused by a combination of the materials employed and the forces exerted by the cuff on the tracheal tissues. In particular, the abrasive action of the cuff shears cells from the lining, epithelium adhering to the cuff is removed during extubation, and normal forces exerted on the basement tissues disrupt the blood supply and cause pressure necrosis. The complications associated with tracheal intubation may be reduced or eliminated by employing airway devices constructed from hydrogel materials. Hydrogels are a class of crosslinked polymers which swell in the presence of moisture, and may contain more than 95% water by weight. For the current study, several prototype airway devices were constructed from hydrogel materials including poly(vinyl alcohol), poly(hydroxyethyl methacrylate), and poly(vinyl pyrrolidone). The raw hydrogel materials from this group were subjected to tensile, swelling, and biocompatibility testing, while the finished devices were subjected to extensive mechanical simulation and animal trials

  3. Gauging U.S.-Indian Strategic Cooperation

    Science.gov (United States)

    2007-03-01

    sabotage. CANDU cores typically are subdivided into two thermo-hydraulic loops. Each loop has hundreds of individual pressure tubes . This feature would...allow international inspections at any of its indigenously constructed reactors. The other pressure driving this deal has been the DAE’s failure to...Indian commercial reactor is the pressurized heavy water reactor (PHWR), based on the Canadian Deuterium Uranium ( CANDU ) design. Fourteen of

  4. Influence of furnace tube shapeon thermal strain of fire-tube boilers

    Directory of Open Access Journals (Sweden)

    Gaćeša Branka

    2014-01-01

    Full Text Available The aim of this paper is to use numerical analysis and fine element method-FEM to investigate the influence of furnace tube shape on the thermal strain of fire-tube boilers. Thermal stresses in corrugated furnace tubes of different shape, i.e. with different corrugation pitch and depth, were analysed first. It was demonstrated that the thermal stresses in corrugated furnace tube are significantly reduced with the increase of corrugation depth. Than deformations and stresses in the structure of a fire-tube boiler were analysed in a real operating condition, for the cases of installed plain furnace tube and corrugated furnace tubes with different shapes. It was concluded that in this fire-tube boiler, which is of larger steam capacity, the corrugated furnace tube must be installed, as well as that the maximal stress in the construction is reduced by the installation of the furnace tube with greater corrugation depth. The analysis of stresses due to pressure and thermal loads pointed out that thermal stresses are not lower-order stresses in comparison to stresses due to pressure loads, so they must be taken into consideration for boiler strength analysis. [Projekat Ministarstva nauke Republike Srbije, br. TR 35040 i br. TR 35011

  5. Left-right differences in Eustachian tube function in children with ventilation tubes.

    NARCIS (Netherlands)

    Heerbeek, N. van; Akkerman, A.E.; Ingels, K.J.A.O.; Engel, J.A.M.; Zielhuis, G.A.

    2003-01-01

    OBJECTIVE: To study the intraindividual variation in Eustachian tube (ET) function in children with ventilation tubes. METHODS: The forced response test, the pressure equilibration test and the sniff test were performed on both ears of 148 children. The results of both ears were compared. RESULTS: N

  6. Circumferential buckling instability of a growing cylindrical tube

    KAUST Repository

    Moulton, D.E.

    2011-03-01

    A cylindrical elastic tube under uniform radial external pressure will buckle circumferentially to a non-circular cross-section at a critical pressure. The buckling represents an instability of the inner or outer edge of the tube. This is a common phenomenon in biological tissues, where it is referred to as mucosal folding. Here, we investigate this buckling instability in a growing elastic tube. A change in thickness due to growth can have a dramatic impact on circumferential buckling, both in the critical pressure and the buckling pattern. We consider both single- and bi-layer tubes and multiple boundary conditions. We highlight the competition between geometric effects, i.e. the change in tube dimensions, and mechanical effects, i.e. the effect of residual stress, due to differential growth. This competition can lead to non-intuitive results, such as a tube growing to be thinner and yet buckle at a higher pressure. © 2011 Elsevier Ltd. All rights reserved.

  7. Shock Tube and Modeling Study of the H + O2 = OH + O Reaction over a Wide Range of Composition, Pressure, and Temperature

    Science.gov (United States)

    Ryu, Si-Ok; Hwang, Soon Muk; Rabinowitz, Martin Jay

    1995-01-01

    The rate coefficient of the reaction H + 02 = OH + 0 was determined using OH laser absorption spectroscopy behind reflected shock waves over the temperature range 1050-2500 K and the pressure range 0.7-4.0 atm. Eight mixtures and three stoichiometries were used. Two distinct and independent criteria were employed in the evaluation of k(sub 1). Our recommended expression for k(sub 1) is k(sub 1) = 7.13 x 10(exp 13)exp(-6957 K/T) cm(exp 3)mol(exp -1)s(exp -1) with a statistical uncertainty of 6%. A critical review of recent evaluations of k(sub 1) yields a consensus expression given by k(sub 1) = 7.82 x 10(exp 13)exp(-7105 K/7) cm(exp 3)mol(exp -1)s(exp -1) over the temperature range 960-5300 K. We do not support a non-Arrhenius rate coefficient expression, nor do we find evidence of composition dependence upon the determination of k(sub 1).

  8. Capillary imbibition in parallel tubes

    Science.gov (United States)

    McRae, Oliver; Ramakrishnan, T. S.; Bird, James

    2016-11-01

    In modeling porous media two distinct approaches can be employed; the sample can be examined holistically, using global variables such as porosity, or it can be treated as a network of capillaries connected in series to various intermediate reservoirs. In forced imbibition this series-based description is sufficient to characterize the flow, due to the presence of an externally maintained pressure difference. However, in spontaneous imbibition, flow is driven by an internal capillary pressure, making it unclear whether a series-based model is appropriate. In this talk, we show using numerical simulations the dynamics of spontaneous imbibition in concentrically arranged capillary tubes. This geometry allows both tubes access to a semi-infinite reservoir but with inlets in close enough proximity to allow for interference. We compare and contrast the results of our simulations with theory and previous experiments. Schlumberger-Doll Research.

  9. Cleaning device for inside of tube

    Energy Technology Data Exchange (ETDEWEB)

    Usuniwa, Yukio; Arai, Sanae; Koyama, Mayumi; Kuramata, Rumi

    1996-07-23

    A large number of small diameter tubes are disposed passing through a lower end plate of a main body of a reactor pressure vessel. The device of the present invention can wash the inside of the small diameter tube efficiently and forecast the completion time of the operation. Namely, purified water at a predetermined temperature is stored in a water reservoir of a purified water recycling unit. The purified water is introduced from a header to the inside of the tube by way of a short tube connected to the lower end of the tube. The purified water introduced into the tube is charged from an opening to an overflow tube and returned to the water reservoir passing through the over flowing tube. The inside of the tube is washed by such circulation of purified water. In the cleaning operation by using the device, (1) the completion time of the operation can be forecast and operationability is improved, and (2) since the operation can be performed rapidly, the period of time for performing the step can be shortened. (I.S.)

  10. Lattice BGK simulations of unsteady flow in a 2D elastic tube

    NARCIS (Netherlands)

    Hoekstra, A.G.; van 't Hoff, J.; Artoli, A.M.M.; Sloot, P.M.A.

    2003-01-01

    We report results of unsteady, harmonic flow simulations with the lattice BGK method in two-dimensional elastic tubes. The tubes are assumed to obey a simple constitutive equation, linearly relating the diameter of the tube to the pressure difference inside and outside the tube. First, as a benchmar

  11. Seismic Structure-Soil-Structure Interaction Analysis of a Consolidated Dry Storage Module for CANDU Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Gon; Yoon, Jeong Hyoun; Kim, Sung Hwan; Yang, Ke Hyung; Lee, Heung Young; Cho, Chun Hyung [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Choi, Kyu Sup; Jeong, In Su [KONES Corporation, Seoul (Korea, Republic of)

    2005-07-01

    The MACSTOR/KN-400 module has been developed as an effective alternative to the existing stand alone concrete canister for dry storage of CANDU spent fuel. The structure is a concrete monolith of 21.67 m long and 12.66 m wide and has a height equal to 7.518 m including the bottom slab. Inside the concrete module consists of 40 storage cylinders accommodating ten 60-bundle dry storage baskets, which are suspended from the top slab and eventually restrained at 10 cm above the bottom slab with horizontal seismic restraints. The main cooling process of the MACSTOR/KN-400 module shall be by air convection through air inlets and outlets. The civil design parameters, with respect to meteorological and seismic loads applied to the module are identical to those specified for the Wolsung CANDU 3 and 4 plants, except for local site characteristics required for soilstructure interaction (SSI) analysis. It is required for the structural integrity to fulfill the licensing requirements. As per USNRC SRP Section 3.7.2, it shall be reviewed how to consider the phenomenon of coupling of the dynamic response of adjacent structures through the soil, which is referred to as structure-soil-structure interaction (SSSI). The presence of closely spaced multiple structural foundations creates coupling between the foundations of individual structures . Some observations of the actual seismic response of structures have indicated that SSSI effects do exist, but they are generally secondary for the overall structural response motions. SSSI effects, however, may be important for a relatively small structure which is to be close to a relatively large structure, while they may be generally neglected for overall structural response of a large massive structure, such as nuclear power plant. As such the scope of the present paper is to carry out a seismic SSSI analysis in case of the MACSTOR/KN- 400 module, in order to investigate whether or not SSSI effect shall be included in the overall seismic

  12. Analysis of clinical value of hydraulic pressure method in diagnosis of fallobian tube patency%输卵管通液测压诊断输卵管通畅性的临床价值分析

    Institute of Scientific and Technical Information of China (English)

    王本立; 郝天然; 张学鸿; 徐自全

    2011-01-01

    Objective To evaluate the clinical value of hydraulic pressure method in diagnosis of fallobian tube patency. Methods Summarize and analyze the hydrotubation diagnosis and hysterosal-pingography (HSG) data of 158 patients with infertility diseases. Use SJ - 1 fallobian tube hydraulic pressure diagnostic and therapeutic instrument to cany out hydrotubation and HSG diagnosis and compare the results of each diagnosis. The results of hydrotubation include patency, incomplete patency and tubal nowhere. The results of HSG diagnosis include normal, incomplete jam and jam. Then analyze the accuracy of the two methods in diagnosis of tubal patency and tubal diseases and conduct x2 examination. Results Using the hydrotubation method, among the 158 cases, we have found 100 cases of patency, 36 cases of incomplete patency and 22 cases of tubal nowhere. Using the HSG method, we have found 66 cases of normal, 7 cases of incomplete jam and 8 cases of jam in regard of light tubal diseases. In regard of serious tubal diseases and using the HSG method, we have found 39 cases of normal, 12 cases of incomplete jam and 18 cases of jam.x2 examination showed that the two methods differ significantly in diagnosis. Conclustions The hydraulic pressure method may cause many errors in diagnosis, for it is unable to distinguish the part, nature and degree of diseases. Thus it doesnt have much clinical vaulue and is not suitable in diagnosis.%目的 评价输卵管通液测压诊断输卵管通畅性的临床价值.方法 总结分析158例不孕症患者通液诊断和子宫输卵管造影( hysterosalpingography,HSG)资料,应用SJ -1宫腔输卵管注液测压诊疗仪分别进行通液诊断、HSG诊断,对比分析每例通液诊断结果与相应HSG诊断结果,通液诊断结果分通畅、不全通畅、不通,HSG相应诊断正常、不全阻塞、阻塞,分别评价两种检查方法诊断输卵管通畅性及输卵管病变的准确性,进行X2检验.结果 158例通液诊断通畅100

  13. Chemical milling of Zircaloy tubing to produce integral OD spiral finned tubes (AWBA development program)

    Energy Technology Data Exchange (ETDEWEB)

    Horwood, W.A.

    1982-02-01

    A detailed process description is provided for producing integral spiral fins on the outside surface of Zircaloy nuclear fuel cladding tubes by masking with pressure sensitive tape strips and then chemical milling (pickling) the tube wall between the tape strips to leave the fins in relief. Fins up to 0.020 inch high by 0.05 to 0.12 inch wide were consistently produced on tubes having wall thickness of 0.008 inch or greater after fin pickling. Wall thickness uniformity was excellent. Information is provided on tube surface preparation to maximize tape mask adhesion time during pickling, acid chemistry control to prevent local tube wall thinning near the fin, and pickling techniques to promote uniform material removal. Simple fixture designs are described for quickly and conveniently applying the tape strips to the tube wall in an accurate spiral. 13 figures, 4 tables.

  14. Chaotic Recurrence Characteristics Analysis of Differential Pressure Fluctuating Signal Across Tube Bundles%管束间压差波动信号的递归特性分析

    Institute of Scientific and Technical Information of China (English)

    洪文鹏; 周云龙; 刘燕

    2011-01-01

    Based on the differential pressure fluctuating signals of the gas-liquid two phase flow measured in two pitch tube bundles, the recurrence quantification analysis method was used to study the dynamic characteristics of differential pressure fluctuating signals of bubbly flow, intermittent flow, mist flow and transition flow patterns. The research indicated isolated points texture appeared on recurrence plot for the bubbly flow. Both dispersed points texture and massive texture appeared on recurrence plot for the intermittent flow with strong chaos characteristic. The recurrence plot of mist flow whose idicated periodic is outstanding has a good diagonal texture. The recursion texture of differential pressure fluctuating signals could well reflect the evolution characteristics of flow pattern, and the recurrence characteristics quantities were obvious to the variations of superficial gas velocity and provided a new ideas for investigation of gas-liquid two-phase flow mechanism.%基于2种节距比管束间的不同流型的压差波动信号,采用递归定量分析法研究了泡状流、间歇流、雾状流3种典型流型及过渡流型的动力学特性。研究结果表明:泡状流在递归图上表现为孤立点状结构,间歇流递归图兼顾点状和块状结构,具有较强的混沌特性,雾状流沿对角线纹理发育好,周期性突出。压差信号的递归纹理结构清晰地演化了其动力学特性,且递归特征量随气相折算流速变化明显,为两相流流型机理的研究提供新思路。

  15. 开放式双套管持续冲洗低负压引流应用于高位肠瘘的护理体会%Using open bicameral perfusion tubes for continuous douche and negative pressure drainage in care of superior position intestinal fistula patients

    Institute of Scientific and Technical Information of China (English)

    张月琴

    2011-01-01

    Objective To explore first - hand experience using open bicameral perfusion tubes for continuous douche and negative pressure drainage in care of patients with superior position intestinal fistula. Methods For 14 diagnosed cases of superior position intestinal fistula patients, we planted open bicameral perfusion tubes for continuous douche and negative pressure drainage. Results In 10 out of the 14 cases, the fistula sealed after continuous douche and negative pressure drainage. The other 4 cases still suffered from high fever, acute diffuse peritonitis, and symptoms of whole body poisoning 3-5 days post therapy, and were subjected to surgery again. Open bicameral perfusion tubes were re- planted, and patients recovered well after continuous douche and negative pressure drainage. Conclusion After the sccessful treatment of 14 cases of superior position intestinal fistula patients, we find that continuous douche and negative pressure drainage using open bicameral perfusion tubes is satisfactory.%目的 探讨开放式双套管持续冲洗低负压引流应用于高位肠瘘的护理体会.方法 对14例高位肠瘘患者,经确诊后置入开放式双套管并行持续冲洗低负压引流.结果10例经持续冲洗引流后瘘口自行愈合,4例因经引流3~5 d后仍出现高热、弥漫性腹膜炎、全身中毒症状而再次手术,术中重新置开放式双套管,术后行持续冲洗引流后治愈.结论 对14例高位肠瘘患者应用开放式双套管持续冲洗低负压引流,效果满意.

  16. Sound absorption and reflection with coupled tubes

    OpenAIRE

    2000-01-01

    This paper describes a special sound absorbing technique with an accompanying efficient numerical design tool. As a basis pressure waves in a single narrow tube or pore are considered. In such a tube the viscosity and the thermal conductivity of the air, or any other fluid, can have a significant effect on the wave propagation. An important aspect is that due to the viscothermal wave propagation sound energy is being dissipated. This has been applied to configurations consisting of a manifold...

  17. Endotracheal Tube Cuff Management at Altitude

    Science.gov (United States)

    2014-02-05

    model study of endotracheal intubation including mechanical ventilation and four methods of cuff pressure management during ascent and descent...AFRL-SA-WP-SR-2014-0007 Endotracheal Tube Cuff Management at Altitude SSgt Tyler J. Britton, RRT1; Richard D. Branson, RRT2...REPORT TYPE Special Report 3. DATES COVERED (From – To) June 2012 – December 2013 4. TITLE AND SUBTITLE Endotracheal Tube Cuff Management

  18. A miniature high repetition rate shock tube.

    Science.gov (United States)

    Tranter, R S; Lynch, P T

    2013-09-01

    A miniature high repetition rate shock tube with excellent reproducibility has been constructed to facilitate high temperature, high pressure, gas phase experiments at facilities such as synchrotron light sources where space is limited and many experiments need to be averaged to obtain adequate signal levels. The shock tube is designed to generate reaction conditions of T > 600 K, P shock waves with predictable characteristics are created, repeatably. Two synchrotron-based experiments using this apparatus are also briefly described here, demonstrating the potential of the shock tube for research at synchrotron light sources.

  19. Air flow exploration of abrasive feed tube

    Science.gov (United States)

    Zhang, Shijin; Li, Xiaohong; Gu, Yilei

    2009-12-01

    An abrasive water-jet cutting process is one in which water pressure is raised to a very high pressure and forced through a very small orifice to form a very thin high speed jet beam. This thin jet beam is then directed through a chamber and then fed into a secondary nozzle, or mixing tube. During this process, a vacuum is generated in the chamber, and garnet abrasives and air are pulled into the chamber, through an abrasive feed tube, and mixes with this high speed stream of water. Because of the restrictions introduced by the abrasive feed tube geometry, a vacuum gradient is generated along the tube. Although this phenomenon has been recognized and utilized as a way to monitor nozzle condition and abrasive flowing conditions, yet, until now, conditions inside the abrasive feed line have not been completely understood. A possible reason is that conditions inside the abrasive feed line are complicated. Not only compressible flow but also multi-phase, multi-component flow has been involved in inside of abrasive feed tube. This paper explored various aspects of the vacuum creation process in both the mixing chamber and the abrasive feed tube. Based on an experimental exploration, an analytical framework is presented to allow theoretical calculations of vacuum conditions in the abrasive feed tube.

  20. Heat transfer analysis of consolidated dry storage system for CANDU spent fuel considering environmental conditions of Wolsong site

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y. [Korea Hydraulic and Nuclear Power Company, Taejon (Korea, Republic of)

    2004-07-01

    The purpose of the present paper is to perform heat transfer analysis of the MACSTOR/KN-400 dry storage system for CANDU spent fuel in order to predict maximum concrete temperatures and temperature gradients. This module has twice the capacity of the existing MACSTOR-200, which is in operation at Gentilly-2. In the thermal design of the MACSTOR/KN-400, Thermal Insulation Panels(TIP) were introduced to reduce concrete temperatures and temperature gradients in the module caused by the high fuel heat loads. Environmental factors such as solar heat, daily temperature variations and ambient temperatures in summer and winter at Wolsong site and the assumed presence of hot baskets were taken into consideration in the simulations. Two cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter. The maximum local concrete temperatures were predicted to be 63 .deg. C for the off-normal case. The temperature gradients in the concrete walls and roof are predicted to be 28C and 25C for off-normal operation in summer, incorporating a 3C uncertainty. In conclusion, this paper shows that the maximum temperature for the module is expected to meet the temperature limitations of ACI 349.

  1. Development of finite element models for the study of ageing effects in CANDU 6 concrete containment buildings

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Y.; Jaffer, S., E-mail: Yuqing.Ding@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    In nuclear power plants (NPPs), concrete containment buildings (CCBs) provide the final physical barrier against the release of radioactive materials into the environment and protect the nuclear structures housed within the containment building. CCBs have to be maintained to ensure leak tightness and sound structural integrity for the safe operation throughout the life of NPPs. However, the integrity of CCBs may be affected by the ageing of its concrete, post-tensioning cables and reinforcing bars (rebars). Finite element models (FEMs) of CANDU 6 CCBs have been developed using 2 independent finite element programs for the study of the effect of ageing of CCBs. These FEMs have been validated using multiple-source data and have been used for preliminary analyses of the effect of thermal load and ageing degradation on the concrete structure. The modelling assumptions and simplifications, approach, and validation are discussed in this paper. The preliminary analyses for temperature effects and potential applications to the study of ageing degradation in CCBs using the FEMs are briefly introduced. (author)

  2. Failure analysis of retired steam generator tubings

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-15

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

  3. Feeding tube insertion - gastrostomy

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/002937.htm Feeding tube insertion - gastrostomy To use the sharing features on this page, please enable JavaScript. A gastrostomy feeding tube insertion is the placement of a feeding ...

  4. Tube Feeding Troubleshooting Guide

    Science.gov (United States)

    Tube Feeding Troubleshoot ing Guide This guide is a tool to assist you, and should not replace your doctor’s ... everyone. table of contents Going Home with Tube Feedings....................................................2 Nausea and ... ...

  5. Neural Tube Defects

    Science.gov (United States)

    Neural tube defects are birth defects of the brain, spine, or spinal cord. They happen in the ... that she is pregnant. The two most common neural tube defects are spina bifida and anencephaly. In ...

  6. Dynamic Response and Fracture of Composite Gun Tubes

    Directory of Open Access Journals (Sweden)

    Jerome T. Tzeng

    2001-01-01

    Full Text Available The fracture behavior due to dynamic response in a composite gun tube subjected to a moving pressure has been investigated. The resonance of stress waves result in very high amplitude and frequency strains in the tube at the instant and location of pressure front passage as the velocity of the projectile approaches a critical value. The cyclic stresses can accelerate crack propagation in the gun tube with an existing imperfection and significantly shorten the fatigue life of gun tubes. The fracture mechanism induced by dynamic amplification effects is particularly critical for composite overwrap barrels because of a multi-material construction, anisotropic material properties, and the potential of thermal degradation.

  7. Static stability of collapsible tube conveying non-Newtonian fluid

    CERN Document Server

    Yushutin, V S

    2014-01-01

    The global static stability of a Starling Resistor conveying non-Newtonian fluid is considered. The Starling Resistor consists of two rigid circular tubes and axisymmetric collapsible tube mounted between them. Upstream and downstream pressures are the boundary condition as well as external to the collapsible tube pressure. Quasi one-dimensional model has been proposed and a boundary value problem in terms of nondimensional parameters obtained. Nonuniqueness of the boundary value problem is regarded as static instability. The analytical condition of instability which defines a surface in parameter space has been studied numerically. The influence of fluid rheology on stability of collapsible tube is established.

  8. An Evaluation of a General Venting Strategy in CANDU 6 Reactor Building

    Energy Technology Data Exchange (ETDEWEB)

    Kim, See Darl; Kim, Dong Ha; Park, Soo Yong; Song, Yong Man; Jin, Young Ho

    2006-03-15

    If the reactor building sprays or local air coolers are not available, depressurization by reactor building venting is considered as a useful mitigation strategy for a severe accident management of the Wolsong plants. As the CFVS is not established in the Wolsong Units, the reactor building isolation system can be a substitute for reactor building venting. The D{sub 2}O Vapour recovery system which has a 30' diameter penetration is expected to meet the NRC requirements. To investigate the effectiveness of the Reactor Building Venting Strategy, three kinds of accidents are analyzed: a SBO, a SLOCA and a Large LOCA. The reactor building pressure behavior was analyzed with ISAAC 2.0.2 for four different cases: without venting, 55psig/50psig, 50psig/40psig and 50psig/30psig valve open/close pressures. It was found that applying venting for a SBO reduces the mass fraction of the CsI released to the environment by 67.8% (valve open/close pressure of 55psig/50psig), by 64.4% (valve open/close pressure of 50psig/40psig) and by 63.5% (valve open/close pressure of 50psig/30psig). For a SLOCA, venting strategy reduces the mass fraction of the CsI by 58.3% (valve open/close pressure of 55psig/50psig), by 55.0% (valve open/close pressure 50psig/40psig) and by 48.3% (valve open/close pressure 50psig/30psig). For a LLOCA, reactor building venting reduces the mass fraction of the CsI released to the environment by less than 10% when compared to that without reactor building venting. When the reactor building spray or local air coolers can not be operated, a depressurization strategy by using the D{sub 2}O Vapour Recovery System could prevent a reactor building failure and reduce the amount of CsI released to the environment. The present study shows that the operation of valves at a pressure of 55psig/50psig is safe and effective. Based on the current study, the strategy of reactor building venting is involved SAMG-5.

  9. 变频空调中翅片管蒸发器换热与压降特性%Heat transfer and pressure drop characteristics of fin - and - tube evaporators in variable frequency air- condition system

    Institute of Scientific and Technical Information of China (English)

    刘金平; 祁元龙; 邹永胜

    2011-01-01

    通过试验研究了变频空调系统中翅片管蒸发器管内外侧换热与压降特性,分析了压缩机频率对波纹翅片和百叶窗翅片管蒸发器管内沸腾换热系数、空气侧换热因子和摩擦因子的影响.结果表明:随着压缩机频率的增加,两种翅片的管内沸腾换热系数hi均增加;摩擦因子f都减小,百叶窗翅片的摩擦因子f是波纹翅片的2倍多;换热因子j随着压缩机频率增加而减小,波纹翅片的j因子随Re数变化明显,百叶窗翅片的则变化不大.%The heat transfer and pressure drop characteristics of outside and inside of the fin - and - tube evaporators in variable frequency air - condition system were studied. The effects of compressor frequency on inside boiling heat transfer coefficient, airside heat transfer factor and friction factor of wavy fin and louver fin were investigated. The results indicated that with the increase of compressor frequency, the boiling heat transfer coefficients in two - phase region of two fins increased. The friction factors f of both fins decreased, and the factor f of louver fin almost was two times higher than that of wavy fin. The factors j of the two fins also decreased and the factor j of wavy fin changed obviously than louver fin with the variety of Re number.

  10. 吹气管长度对脉冲萃取柱柱重瞬间压降信号测量的影响%Effect of Tube Length on Measurement of Instantaneous Pressure Drop of Column Weight in Pulsed Extraction Column

    Institute of Scientific and Technical Information of China (English)

    李少伟; 曾鑫; 景山; 刘继连; 吴秋林

    2014-01-01

    The theoretical and experimental study on the purge tube length effect in the air purge method was carried out .Two problems including the effect of the purge tube length on the pressure amplitude and on the pressure phase delay were mainly investiga-ted .The results show that the tube length has little effect on the pressure amplitude when it is no bigger than 17 m .Obvious attenuation is found out when the tube length is bigger than 17 m and the attenuation coefficient calculated by the theoretical model shall be used to calculate the real pressure amplitude . The tube length has obvious delay effect on the pressure wave .The delay time increases with the tube length .The pres-sure wave delay has big effect in a pressure difference measurement when the lengths of the two purge tubes are not equal .The gas velocity in the purge tube calculated by the theoretical model provides a good method for the evaluation of the purge cup volume .A purge cup with volume of 50 mL could satisfy the requirement of the air purge method . The theoretical and experimental results in this article provide fundamental for the application of the air purge method .%本文从理论和实验两方面对吹气法测量中吹气管长度的影响问题进行系统研究,主要包括吹气管长度对柱重压力波动幅值测量的影响和对柱重压力波动的相位延迟作用两个方面。结果表明,在吹气管长度不大于17 m时,柱重压力波动幅值的变化可忽略,而大于17 m时,则有明显的衰减,需通过模型计算的衰减系数进行实际波动幅值的计算;吹气管长度对压力波动的相位有明显的延迟作用,延迟时间随吹气管长度的增大而增大,这个延迟作用在压差测量中如果吹气管长度不一致,会对测量结果有很大影响。理论计算的气体流速为吹气杯体积的估算提供了一个重要方法,计算表明,本实验中50 m L的吹气杯可满足吹气法要求。本文的理

  11. Experiments with micro-fin tube in single phase

    Energy Technology Data Exchange (ETDEWEB)

    Copetti, J.B.; Macagnan, M.H.; De Souza, D.; Oliveski, R.D.C. [Universidade do Vale do Rio dos Sinos, Sao Leopoldo (Brazil). Department of Mechanical Engineering

    2004-12-01

    This work shows heat transfer and friction characteristics for water single-phase flow in micro-fin tubes. The analysis of thermal and hydraulic behavior from a laminar to a turbulent flow was carried out in an experimental setup with a 9.52 mm diameter micro-fin tube. The tube was wrapped up with an electrical resistance tape to supply a constant heat flux to its surface. Different operational conditions were considered in the heating tests. The inlet and outlet temperatures, differential wall temperatures along the tube, pressure drop and flow rate were measured. The relationships of heat flux and flow rate with heat transfer coefficient and pressure drop were analyzed. Under the same conditions, comparative experiments with an internally smooth tube were conducted. The micro-fin tube provides higher heat transfer performance than the smooth tube (in turbulent flow h{sub micro-fin}/h{sub smooth}=2.9). In spite of the increase in pressure drop ({delta}p{sub micro-fin}/{delta}p{sub smooth}=1.7) the heat transfer results were significantly higher (about 80%). This shows the advantages of this enhanced configuration in thermal performance related to conventional tubes. The smooth tube results were validated by the comparison with the Dittus-Boelter and Gnielinski correlations. For the micro-fin tube an empirical correlation to the heat transfer coefficient adjusted from the set of measured data is proposed. The values obtained are in conformity with experimental results. (author)

  12. Experimental study on a simple Ranque Hilsch vortex tube

    Science.gov (United States)

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

    2005-03-01

    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 built and nitrogen is used as its working fluid. A special Pitot tube is used for the measurement of the pressure and velocity. This Pitot tube consists of a capillary which has only one hole in the cylinder wall. With this Pitot tube, the pressure and velocity fields inside the tube were measured. In the same way, the temperature field was measured with a thermocouple. The results of three different entrance conditions are compared here. With the measurements results, the analysis based on the two thermodynamic laws has been made. It is found that rounding off the entrance has influence on the performance of the vortex tube. The secondary circulation gas flow inside the vortex tube can be enhanced and enlarged, the performance of the Ranque-Hilsch vortex tube improved.

  13. Improvement of pump tubes for gas guns and shock tube drivers

    Science.gov (United States)

    Bogdanoff, D. W.

    1990-01-01

    In a pump tube, a gas is mechanically compressed, producing very high pressures and sound speeds. The intensely heated gas produced in such a tube can be used to drive light gas guns and shock tubes. Three concepts are presented that have the potential to allow substantial reductions in the size and mass of the pump tube to be achieved. The first concept involves the use of one or more diaphragms in the pump tube, thus replacing a single compression process by multiple, successive compressions. The second concept involves a radical reduction in the length-to-diameter ratio of the pump tube and the pump tube piston. The third concept involves shock heating of the working gas by high explosives in a cyclindrical geometry reusable device. Preliminary design analyses are performed on all three concepts and they appear to be quite feasible. Reductions in the length and mass of the pump tube by factors up to about 11 and about 7, respectively, are predicted, relative to a benchmark conventional pump tube.

  14. Mechanical Instabilities of Biological Tubes

    CERN Document Server

    Hannezo, Edouard; Prost, Jacques; 10.1103/PhysRevLett.109.018101

    2012-01-01

    We study theoretically the shapes of biological tubes affected by various pathologies. When epithelial cells grow at an uncontrolled rate, the negative tension produced by their division provokes a buckling instability. Several shapes are investigated : varicose, enlarged, sinusoidal or sausage-like, all of which are found in pathologies of tracheal, renal tubes or arteries. The final shape depends crucially on the mechanical parameters of the tissues : Young modulus, wall-to-lumen ratio, homeostatic pressure. We argue that since tissues must be in quasistatic mechanical equilibrium, abnormal shapes convey information as to what causes the pathology. We calculate a phase diagram of tubular instabilities which could be a helpful guide for investigating the underlying genetic regulation.

  15. A study of swirl flow in draft tubes

    Energy Technology Data Exchange (ETDEWEB)

    Dahlhaug, Ole Gunnar

    1997-12-31

    This thesis presents measurements performed inside conical diffuser and bend, draft tubes of model hydro turbines, and draft tube of a prototype hydro turbine. Experimental results for swirling flow in conical diffuser and bend are presented in three different geometries. The axial velocity decreases at the centre of the tube at high swirl numbers because of an axial pressure gradient set up by the downstream frictional damping of the tangential velocities and the pressure increase downstream of the diffuser. Analytical models of the tangential velocity profiles are found and the radial pressure distribution calculated. Good correlation to the measured pressure distribution was achieved. Diffuser efficiency was calculated based on the equations for velocity and pressure profiles, which gave a qualified estimate of the diffuser hydraulic performance. The calculation shows that the bend reduces the efficiency by more than 30%. For a straight tube followed by a diffuser, numerical calculations were done, using K{epsilon}, RNG and RSM turbulence models for all measured swirl numbers. The K{epsilon} model gave best results for the forced vortex profile at low swirl numbers, while the RSM model gave best results at high swirl number. The turbulent kinetic energy at high swirl numbers gave the largest difference between the calculated and the measured values. Measurements on draft tubes in model turbines show the importance of good draft tube design. Prototype measurements on a Francis turbine show how the outlet draft tube flow should be measured for prototype draft tube evaluation. 54 refs., 118 figs., 2 tabs.

  16. Replacement of radiography with ultrasonic phased array for feeder tubes in CANDU reactors using ASME code case N-659-2

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, R.; Bower, Q.; Arseneau, S., E-mail: bsimmons@metalogicinspection.com, E-mail: qbower@metalogicinspection.com, E-mail: sarseneau@metalogicinspection.com [Metalogic Inspection Services, Edmonton, Alberta (Canada)

    2013-07-01

    In this paper we will discuss phased array technology for the replacement of radiography on new construction projects in the nuclear industry. Specifically, through the implementation of A.S.M.E. code N-659-2 and MetaPhase phased array services. Phased Array is not considered a new technique on in service welds in the nuclear industry; however it was unprecedented on new construction welds and required significant investment in regulatory approval (C.N.S.C.), technology research and development, regulatory, client and technician training for successful service implementation. This paper will illustrate the abilities and limitations associated in replacing radiography with MetaPhase, as well as the substantial benefits relative to increased production, improved weld quality, enhanced safety and overall project cost savings. (author)

  17. Turbulent flow in longitudinally finned tubes

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.P.; Hirsa, A.; Jensen, M.K. [Rensselaer Polytechnic Inst., Troy, NY (United States). Dept. of Mechanical Engineering, Aeronautical Engineering and Mechanics

    1996-09-01

    An experimental investigation of fully developed, steady, turbulent flow in longitudinally finned tubes has been performed. A two-channel, four-beam, laser-Doppler velocimeter was used to measure velocity profiles and turbulent statistics of air flow seeded with titanium dioxide particles. Mean velocities in axial, radial, and circumferential directions were measured over the tube cross sections and pressure drop in the tubes was measured at six stations along the test section length in order to calculate the fully developed friction factor. Four experimental tube geometries were studied: one smooth tube; two 8-finned tubes (fin height-to-radius ratios of 0.333 and 0.167), and one 16-finned tube (fin height-to-radius ratio of 0.167); detailed measurements were taken at air flow rates corresponding to Reynolds numbers of approximately 5,000, 25,000, and 50,000. Friction factor data were compared to literature results and showed good agreement for both smooth and finned tubes. The wall shear stress distribution varied significantly with reynolds number, particularly for Reynolds numbers of 25,000 and below. Maximum wall shear stress was found at the fin tip and minimum at the fin root. Four secondary flow cells were detected per fin (one in each interfin spacing and one in each core region for each fin); secondary flows were found to be small in comparison to the mean axial flow and relative magnitudes were unaffected by axial flow rate at Reynolds numbers above 25,000. The fluctuating velocities had a structure similar to that of the smooth tube in the core region while the turbulence in the interfin region was greatly reduced. The principal, primary shear stress distribution differed considerably from that of the smooth tube, particularly in the interfin region, and the orientation was found to be approximately in the same direction as the gradient of the mean axial velocity, supporting the use of an eddy viscosity formulation in turbulence modeling.

  18. Ultimate Temperature of Pulse Tube Cryocoolers

    Science.gov (United States)

    Kittel, P.

    2010-04-01

    An ideal pulse tube cryocooler using an ideal gas can operate at any temperature. This is not true for real gasses. The enthalpy flow resulting from the real gas effects of 3He, 4He, and their mixtures in ideal pulse tube cryocoolers puts limits on the operating temperature of pulse tube cryocoolers. The discussion of these effects follows a previous description of the real gas effects in ideal pulse tube cryocoolers and makes use of models of the thermophysical properties of 3He and 4He. Published data is used to extend the analysis to mixtures of 3He and 4He. The analysis was done for pressures below 2 MPa and temperatures below 2.5 K. Both gasses and their mixtures show low temperature limits for pulse tube cryocoolers. These limits are in the 0.5-2.2 K range and depend on pressure and mixture. In some circumstances, even lower temperatures may be possible. Pulse tube cryocoolers using the two-fluid properties of dilute 3He in superfluid 4He appear to have no limit.

  19. Analysis of Hydrostatic Pressure Bursting Test on β "-Al2O3 Ceramics Tubes for Energy-storage Sodium-sulfur Batteries%储能钠硫电池β"-Al2O3陶瓷管静水压爆破试验分析

    Institute of Scientific and Technical Information of China (English)

    张建平; 朱翔宇; 刘芳; 刘宇; 祝铭

    2013-01-01

    为了确定钠硫电池关键部件β"-Al2 O3陶瓷管的极限强度,利用直尺、游标卡尺及壁厚仪测量了陶瓷管外形尺寸,通过静水压爆破试验得到了爆破极限压力,采用薄壁容器受压后应力近似公式和第一强度理论设计准则计算出陶瓷管的极限应力,并分析了影响极限应力的主要因素,这对陶瓷管的强度校核和优化设计具有很强的指导意义.%In order to determine the ultimate strength of β "-Al2 O3 ceramics tubes which are the key components of sodium-sulfur batteries,the size of ceramic tubes was measured by ruler,slide caliper and wall thickness gauge,and the blasting limit pressure was obtained by the hydrostatic pressure bursting test.The ultimate stress of the ceramic tube was calculated by using the approximate formulas of the thinwalled container subjected to the pressure and the design criteria of the first strength theory,and the main factors affecting the ultimate stress were analyzed,which provides some significant guideline to the strength check and the optimization design of the ceramic tube.

  20. Applications of a thru-tubing cement retainer conveyed on coiled tubing

    Energy Technology Data Exchange (ETDEWEB)

    Willems, T. (Baker Oil Tools, Houston, TX (Canada)); Tudor, E.H. (Canadian Fracmaster Ltd., Calgary, AB (Canada)); Cooke, J.A. (Chevron Canada Resources, Fox Creek, AB (Canada))

    1994-01-01

    A thru-tubing inflatable permanent cement retainer has been developed to selectively squeeze cement without pulling the production tubing or killing the well. The system has been run successfully on coiled tubing through 73 mm tubing and set in 177.8 mm casing. Due to the performance limitations of the inflatable cement retainer, unconventional methods were used to inflate the tool and perform a hesitation cement squeeze. Equipment and procedures used to selectively squeeze off water production in two wells in the Kaybob south field in central Alberta are described. Both wells are gas producers from a sour, underpressured carbonate formation. In order to control hydrostatic pressure and obtain the squeeze, nitrogen gas was used as the circulation and displacement medium. Workover design considerations, cement retainer differential pressure ratings, depth control, inflation procedure, cement slurry design, cement squeeze operation, and job procedure are described. 2 refs., 3 figs.

  1. Design of water shock tube for testing shell materials

    OpenAIRE

    Ji, Hongjuan; Mustafa, Mohamad; Khawaja, Hassan Abbas; Ewan, Bruce C.; Moatamedi, Mojtaba

    2014-01-01

    This paper presents design considerations for a shock tube experimental rig used to investigate the dynamic failure mechanisms of shell geometries subjected to water shock impact loading. In such setup, it is desirable that the drive pressure used within the tube can provide a wide range of impulsive loads on the test structures and some flexibility can be achieved on the applied pulse durations. With this aim a review of various existing shock tube experimental setup is presented and choi...

  2. Iatrogenic velopharyngeal insufficiency caused by neonatal nasogastric feeding tube.

    Science.gov (United States)

    Pollack, Aron Z; Ward, Robert F; DeRowe, Ari; April, Max M

    2014-08-01

    Complications from a prolonged nasogastric tube intubation, though seldom reported, are well described. Herein we describe the first two reported cases of velopharyngeal insufficiency secondary to velopharyngeal scarring and immobility from repetitive nasogastric tube insertions and prolonged use. Differing only in location, the proposed pathophysiologic mechanism of injury is identical to that of the nasogastric tube syndrome, a rare and serious, well described entity consisting of bilateral vocal fold paralysis due to pressure-induced ulceration of the posterior cricoarytenoid musculature.

  3. Filmwise condensation of steam on low integral-finned tubes

    OpenAIRE

    Georgiadis, Ioannis V.

    1984-01-01

    Approved for public release; distribution is unlimited Filmwise condensation heat-transfer measurements if steam were made on horizontal tubes under vacuum and near-atmospheric pressures. Data were taken for a smooth tube and for 21 tubes which contained rectangularly-shaped, low integral fins. The fin geometry was systematically varied in order to investigate the dependence of the steam-side heat-transfer coefficient on fin spacing, thickness and height. The condensation process was fo...

  4. Development of CANDU advanced fuel fabrication technology - A development of amorphous alloys for the solder of nuclear reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jai Young;