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Sample records for candu pressure tubes

  1. Pressure tube life management in CANDU-6 nuclear plant

    International Nuclear Information System (INIS)

    Operating parameters of pressure tube in CANDU-6 reactor, the relation between pressure tube life and plant life improvement of pressure tube by AECL in past years were summarized, and the factors affecting pressure tube life, idea and main measures of pressure tube life management in QINSHAN CANDU-6 power plant introduced

  2. Ballooning of CANDU pressure tubes. Model assessment

    International Nuclear Information System (INIS)

    The transient creep equations used to analyze the possible ballooning and failure of Zr-2.5% Nb pressure tubes during a loss-of-coolant accident (LOCA) were developed and verified using as-received Zr-2.5% Nb pressure tube material. But in a CANDU reactor, the pressure tubes absorb deuterium and are exposed to a continuous neutron fluence. Consequently, a literature survey was done to determine how irradiation damage and deuterium might affect the creep rate and ductility of Zr-2.5% Nb pressure tubes in the temperature range from 600 to 800 degrees C. It was found that irradiation damage, dissolved deuterium and deuteride blisters could possibly affect the creep rate and ductility of ZR-2.5% Nb pressure tubes in this temperature range, but deuteride platelets are expected to have little effect. Further tests are required to determine the effect of irradiation damage and deuterium on the creep rate and ductility of pressure tubes

  3. Structural integrity evaluations of CANDU pressure tubes

    International Nuclear Information System (INIS)

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

  4. Highlights of the metallurgical behaviour of CANDU pressure tubes

    International Nuclear Information System (INIS)

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

  5. The pressure tubes in the CANDU power reactor

    International Nuclear Information System (INIS)

    Nuclear power reactors using zirconium alloy pressure tubes generate electricity in several countries. In Ontario CANDU reactors generate about 30 percent of the electricity produced in the province. The pressure tubes of the first five CANDU reactors were made of cold-worked Zircaloy-2, an alloy of zirconium and tin developed by the US Navy. In 1958 the USSR published information on a Zr-2.5 wt percent Nb alloy, in which the Nb promotes stabilization of the β phase, thus presenting opportunities of exploiting metallurgically strong pressure tubes analogous to the heat-treatable α-β titanium alloys. After two reactors using Zr-2.5 wt percent Nb in a quenched and aged condition were constructed, an extensive development program on cold-worked Zr-2.5 wt percent Nb pressure tubes resulted in their becoming the reference tubes for all future CANDU reactors. Pressure tubes of Zr-3.3 wt percent Sn-0.8 wt percent Nb-0.8 wt percent Mo (Excel) are in an advanced state of development. These tubes will be used in an annealed condition; projections show that they will have improved dimensional stability over the lifetime of the reactors. These improvements result from experimental programs leading to an understanding of the relationship between microstructures and fabrication variables and effects of the environment during service in nuclear reactors. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  7. High-temperature transient creep properties of CANDU pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Fong, R.W.L.; Chow, C.K

    2002-06-01

    During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)

  8. High-temperature transient creep properties of CANDU pressure tubes

    International Nuclear Information System (INIS)

    During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate, and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al. [1], based on creep equations derived using uniaxial tensile specimens [2]. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. This work was funded by the CANDU Owners Group (COG) R and D

  9. Ultrasonic crack-tip diffraction in CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    Currently there is no reliable method of measuring defect depths in CANDU reactor pressure tubes. The demonstrated success of crack-tip diffraction (or time-of-flight-testing) in round-robins on thick components has promoted an interest in this technique. In CANDU reactors, pressure tubes are effectively accessible only from the inside. Development work has concentrated on outside surface defects using 45 degree shear waves in contrast to the longitudinal waves usually used for testing thick components with this technique. Due to the small wall thickness of the pressure tubes (4.2 mm) and the typical sizes of defects of interest (0.15 mm or greater), frequencies of the order of 20 MHz are being used. A further complication comes from the orientation of the defects, which may be at any angle in pressure tubes. Initial studies have been performed on a series of outside surface notches and slots, plus a real fatigue crack. This crack was on the inside surface, so the technique required measuring this defect's depth from the outside. Initial results are encouraging. Even without signal processing, crack-tip diffracted signals were detectable from all but very large (2.5 mm) and very small (less than 0.076 mm) notches. Errors in estimates of defect depths were typically less than 0.1 mm for all the notches, and the results were consistent. Measurements on the fatigue crack showed similar random errors, though there appeared to be a deterministic error of about 0.1 mm as well

  10. Ballooning of CANDU pressure tubes - experiments with degraded tube material

    International Nuclear Information System (INIS)

    Three as-received Zr-2.5% Nb pressure tube specimens and three specimens with eight 0.5 mm deep defects machined on the inside surface were tested in the ballooning test rig at Stern Laboratories Inc. The temperature ramp rate was controlled between 28 K s-1 and 35 K s-1. Temperatures on the outside and inside surfaces of the specimens, and circumferential and longitudinal strains were recorded during the transients. Post-test longitudinal, circumferential and wall thickness strains were measured. All as-received specimens ruptured full-length near the top, i.e., the hottest point. All defected specimens failed at either or both upper defects, one rupture being full-length and the others limited to one to three times the length of the defect. (author). 4 refs., 2 tabs., 15 figs

  11. Mechanistic modeling of thermal-mechanical deformation of CANDU pressure tube under localized high temperature condition

    International Nuclear Information System (INIS)

    Thermal strain deformation is a pressure tube failure mechanism. The main objective of this paper is to develop mechanistic models to evaluate local thermal-mechanical deformation of a pressure tube in CANDU reactor and to investigate fuel channel integrity under localized contact between fuel elements and pressure tube. The consequence of concern is potential creep strain failure of a pressure tube and calandria tube. The initial focus will be on the case where a fuel rod contacts the pressure tube at full power with highly cooling condition

  12. Assessment and management of ageing of major nuclear power plant components important to safety: CANDU pressure tubes

    International Nuclear Information System (INIS)

    The report documents the current practices for assessment and management of the ageing of the pressure tubes in CANDU reactors and Indian PHWTRs. Chapter headings are: fuel channel and pressure tube description, design basis for the fuel channel and pressure tube, degradation mechanisms and ageing concerns for pressure tubes, inspection and monitoring methods for pressure tubes,assessment methods and fitness-for-service guidelines for pressure tubes, mitigation methods for pressure tubes, and pressure tube ageing management programme

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

    International Nuclear Information System (INIS)

    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

  14. A Development of Preliminary Evaluating Model of Crept Pressure Tube Diameter for CANDU Reactor (2)

    International Nuclear Information System (INIS)

    Pressure tube of CANDU reactor can be expanded toward both radial and axial directions due to irradiation under the high pressure and temperature condition. As the irradiation period increases, the radial expansion due to creep of the pressure tube increases. The radial expansion of the pressure tube comes out the reduction of the coolability and it results in the power deration. Although the radial expansion of the pressure tube directly affect the safety and economy of the currently operated CANDU reactor, there is no domestic evaluation model to predict the pressure tube diameter. Accordingly, it is necessary to develop the prediction model of the pressure tube diameter and the is the motivation of this study. The objectives of the current work is to develop the basic evaluation model of the pressure tube diameter for CANDU reactor especially for Wolsong NPP (Nuclear Power Plant). In order to develop the diameter evaluation model, measured data for total 86 channels were collected from Wolsong NPP 1, 2, 3 and 4 and analyzed. Based on the provided data, the operational conditions such as a temperature, pressure and neutron flux along the axial direction were derived. All data were analysed to derive the correlation between the pressure tube diameter and the other operation parameters. The evaluation model of pressure tube diameter was modeled by using the neural network algorithm. Neural network algorithm has been widely used to derive the non-linear relation between the input and output data. The developed neural network model was learned based on the data from Wolsong NPP 2, 3, and 4 and was tested by using data from Wolsong NPP 1. The current project will be carried out by IAEA CRP in which all CANDU nations are going to participate

  15. Wet channel measurement of pressure tube to calandria tube spacing in CANDU reactors

    International Nuclear Information System (INIS)

    The pressure tube (PT) to calandria tube (CT) spacing in CANDU reactors is an important parameter that relates to the general condition of the fuel channels. The measurement system that was developed to measure this parameter during the wet channel inspections of Pickering Units 1 and 2 is described in this paper. A send-receive eddy current probe was designed which is primarily sensitive to variations in PT/CT spacing but is also affected by pressure tube wall thickness. A computer simulation showed that the phase angles of the response to these variables are similar for all usable frequencies, thus eliminating the possibility of multifrequency compensation. A marriage of technologies was proposed involving the ultrasonic measurement of wall thickness values which are then used to extract the spacing information from the eddy current signal. The accuracy of the system is approximately ±(30% +.1mm) which has been sufficient to determine if and where any of the pressure tubes have come in contact with their calandria tube. Field experience with the new system is discussed and areas for development are also outlined

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

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

    International Nuclear Information System (INIS)

    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

  18. The blister phenomena in relation to pressure tube integrity in CANDU reactor

    International Nuclear Information System (INIS)

    Zirconium alloy pressure tubes in CANDU type PHWR reactors are exposed to aqueous conditions embracing high temperature, fast neutron flux and high pressure. Two properties, dimension and hydrogen concentration, represent the main properties where changes are important to the life of a pressure tube. Rupture of a cold worked Zircaloy-2 pressure tube in Pickering Unit 2 in 1983 occurred when a crack developed from an array of hydride blisters. These have been observed on the outside surface of the pressure tube where it contacted the surrounding calandria tube. The contact of the pressure tube with the calandria tube can occur during the operation time and produces in the pressure tube a localized cooling. The consequence of the local heating of the calandria tube is some localized hydride precipitation. Under certain conditions, hydrogen will migrate down the temperature gradient and accumulate in the coldest region. Such precipitation, when it occurs under operating conditions, is considered to be the start of blister formation. When the blister cracking threshold is reached, the blister cracking can initiate a crack on the tube body. The failure mechanisms in zirconium alloy pressure tubes involve the presence of hydrogen in the initiation process and then a propagation process. If crack, originating from a hydride blister on the outside of the pressure tube, is developed then crack growth is possible in the axial direction to a partial thickness unstable length. Unstable pressure tube rupture is an event in a CANDU reactor that has potentially serious economic and safety consequences and reactor operation under conditions which entail the risk of such failure should be avoided. (authors)

  19. Critical heat flux in CANDU moderator following a pressure tube to calandria tube contact - part I

    International Nuclear Information System (INIS)

    Heavy water moderator surrounding each fuel channel is one of the important features in CANDU reactors that act as a heat sink for the fuel in the situations where other means of heat removal fail. In the critical break LOCA scenario, fuel cooling becomes severely degraded due to rapid flow reduction in the affected flow pass of the heat transport system. This can result in pressure tubes experiencing significant heat-up while coolant pressure is still high, thereby causing uniform thermal creep strain (ballooning) of the pressure tube (PT) into contact with its calandria tube (CT). The contact of the hot PT with the CT causes rapid redistribution of stored heat from the PT to CT and a large spike in heat flux from the CT to the moderator fluid. For lower subcooling conditions of the moderator, this heat flux spike can cause dryout of the CT. This can detrimentally affect channel integrity if the CT post-dryout temperature becomes sufficiently high to result in thermal creep strain deformation. The focus of this research is to develop a mechanistic model to predict Critical Heat Flux (CHF) on the CT surface following a contact with its pressure tube. A COMSOL multi-physics model using a two-dimensional transient fluid-thermal analysis of the CT surface undergoing heat up is used to predict flow and temperature profile on the CT surface. A mechanistic CHF model is to be proposed based on a concept of wall dry patch formation, prevention of rewetting and subsequent dry patch spreading. (author)

  20. Remote metallurgical investigations on pressure tubes removed from CANDU power reactors

    International Nuclear Information System (INIS)

    As part of the periodic in-service inspection program for CANDU reactors, pressure tubes are periodically removed for destructive examination. The procedures, equipment, and facilities used to perform metallurgical examinations on these highly irradiated components are described. The initial examinations of the tubes from the generating station are performed underwater in inspection bays. Detailed visual examination and metallography are subsequently performed in shielded hot-cell facilities; a description of the remote metallographic equipment and preparation techniques used is given. Examinations of two recently removed Zr-2.5%Nb pressure tubes containing fretting-wear flaws and a lamination flaw are used to highlight the techniques employed

  1. Analysis of the effects of irradiation on the stress-strain behavior in CANDU pressure tubes

    International Nuclear Information System (INIS)

    After commissioning of the Cernavoda Nuclear Power Plant - Unit 1, Romania ranges among the users of CANDU-PHWR (Canadian Deuterium Uranium - Pressurized Heavy Water Reactor) type reactor, adopted and developed in Canada, using natural uranium as nuclear fuel and D2O (heavy water) as moderator and coolant. The main components of the reactor core are the fuel channels pressure tubes. These tubes are made of Zr-2.5%Nb alloy and during the normal operating conditions their mechanical properties could be modified due to irradiation. There are four damage mechanisms responsible for the limiting lifetime of CANDU pressure tubes: circumferential expansion, irradiation growth, creep sag and hydrogen increase. The paper presents the experimental methods developed at the Institute for Nuclear Research Pitesti (INR) in order to obtain the influence of the irradiation on CANDU pressure tubes stress-strain behavior. The tensile test methodology has been developed using the Hot Cells facilities from INR. The irradiation was performed on Zr-2.5%Nb alloy samples in the Romanian TRIGA Reactor. The results will be used in the structural integrity assessments, performed in accordance with R6/rev.4 British Energy procedure. (authors)

  2. Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable β-Zr that is present as a thin layer encasing the predominant α-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded β-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the β-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

  3. Assessment of aging of Zr-2.5Nb pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    In modern CANDU nuclear generating stations, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the fuel bundles and the heavy water (D2O) coolant. The pressure tubes operate at an internal pressure of about 10 MPa and temperatures ranging from about 250oC at the inlet to about 310oC at the outlet. Over the expected 30 year lifetime of these tubes they will be subjected to a total fluence of approximately 3 x 1026 n m-2. In addition, these tubes gradually pick up deuterium as a result of a slow corrosion process. When the hydrogen plus deuterium concentration in the tubes exceeds the hydrogen-deuterium solvus, the tubes are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC). If undetected, such a cracking mechanism could lead to unstable rupture of the pressure tube. A fitness-for-service methodology has been developed which assures that this will not happen. A key element in this methodology is the acquisition of data and understanding-from surveillance and accelerated aging testing-to assess and predict changes in the DHC initiation threshold, the DHC velocity and the fracture toughness (critical crack length) as a function of service time. The most recent results of the DHC and fracture toughness properties of CANDU pressure tubes as a function of time in service are presented and used to suggest procedures for mitigation and life extension of the pressure tubes. (author)

  4. Corrosion mechanisms of tubes from the Candu high pressure feed water heaters

    International Nuclear Information System (INIS)

    This study has been carried out to investigate the corrosion mechanism developed on stainless steel tubes from High Pressure Feed Water Heater in a Candu NPP. With this goal in mind, specimens of the failed tubes removed from High Pressure Feed Water Heater were analyzed in laboratory by different methods (visual examination, scanning electron microscopy (SEM) and optical microscopy). The corrosion deposits from the surfaces of tubes were examined by energy dispersive X-ray spectrometry method (EDX) to identify the chemical composition. The results of the laboratory tests showed that the tubes failed by OD (outside diameter) chloride induced stress corrosion cracking mechanism. Stress corrosion cracking (SCC) is a form of slow crack growth that occurs when a susceptible alloy is stressed in a specific corrosive environment. Cracks were independent of tube support locations and welds. The corrosion mechanism and possible causes identified are presented, followed by conclusions and recommendations for corrosion minimization. (authors)

  5. A probabilistic approach to leak-before-break in CANDU pressure tubes

    International Nuclear Information System (INIS)

    In the CANDU reactor, the coolant passes through the core in zirconium alloy pressure tubes. A few of these pressure tubes have leaked at cracks near the rolled joint where the pressure tube is attached to the end fitting. A probabilistic methodology, and associated computer code (called MARATHON), has been developed to calculate the time from first leakage to unstable fracture in a probabilistic format. The methodology explicitly uses material property distributions, and allows the risk associated with leak-before-break to be estimated. A model of the leak detection system is included to calculate the time between leak detection and unstable fracture. The sensitivity of the risk to changing reactor conditions allows the optimization of reactor management after leak detection. Preliminary material property distributions show the probability of unstable fracture is very low, and that ample time is available to shut down the unit and locate the leaking tube. (author)

  6. Elasto-plastic behaviour of the pressure tube (Zr-2.5Nb%) of CANDU reactor

    International Nuclear Information System (INIS)

    To ensure the structural integrity and meet the safety conditions in the pressure tubes of Cernavoda CANDU type reactor, subjected to a severe thermodynamic operational environment (operation temperature 560 K - 585 K, tangential stress coefficient σe = 110-130 MPa), to the corroding ambient and to the radiation field (a fast neutron flux of 1017 n/m2s), the knowledge of mechanical parameter evolution, along the lifetime span (about 30 - 40 years), is needed. The present work reports the materials data of samples extracted from a pressure tube of Cernavoda reactor (alloy-Zr97.5Nb2.5), axially submitted, to ambient and 300 deg. C temperatures. The thermodynamic stress monitoring and the experimental data acquisition and processing were carried out by an analog-to-digital converter. The experimental data obtained by this procedure were correlated rather well by means of the Hsu-Bertels law, adequate to describe the mechanical fatigue of elasto-plastic materials. The parameters determined are used in computing and predicting the behaviour of CANDU pressure tube in normal operation conditions

  7. Numerical analysis of zirconium hydride blisters in CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    CANDU nuclear reactors use zirconium alloy pressure tubes for primary containment of fuel and coolant. The 1983 failure of a pressure tube in Unit 2 of the Pickering Nuclear Generating Station was attributed to the formation of large precipitates of zirconium hydride, referred to as blisters. These blisters formed at localized cold spots on the pressure tube surface where it had come into contact with the colder calandria tube. The high hydrogen concentrations in the Zircaloy-2 pressure tubes used only in the first two Pickering Units were a major contributing factor to blister formation and the ultimate failure. In an effort to better understand the mechanism of crack initiation at a blister, a program was undertaken to use finite element methods to model the stresses generated by the formation of a blister in a tube. The preliminary results in this work have been published elsewhere. This paper summarizes the recent refinements to the model and our present understanding of the development of stresses in and around hydride blisters. (orig./GL)

  8. Suppressing hydrogen ingress during aqueous corrosion of CANDU Zr-2.5 Nb pressure tube material

    International Nuclear Information System (INIS)

    As a result of their special properties, including low neutron cross-section and intrinsic corrosion resistance, Zr alloys are used in the fabrication of nuclear core components, particularly fuel cladding (in most reactor types) and also Zr-2.5 Nb pressure tubes in CANDU trademark (Canada Deuterium Uranium) reactors. Corrosion and H uptake during service can limit the life of these components. Therefore, remedial action may be appropriate to slow the H uptake rate and prolong the working life of these reactor components. This work has explored the possibility of reducing H uptake in pressure tube material by incorporating an inhibiting agent into the corrosion environment. Two approaches have been tested, depositing a thin metallic film on the initial oxide surface and adding an inhibiting agent to the solution. The latter approach appears more practical. Screening experiments were conducted in short-term (∝30 day) exposures in high temperature (340 C) aqueous out-reactor environments, simulating the CANDU trademark heat transport coolant with various chemistries. Compounds tested included aluminum acetate, aluminum nitrate, lithium nitrate, rhodium nitrate and yttrium nitrate. Comparison of results from the aluminum nitrate additives and aluminum acetate additives suggests that the nitrate anion is the effective ingredient for H ingress inhibition. The nitrate anion appears to reduce the rate of H ingress regardless of the associated cation. However, each cation appears to affect the rate of corrosion differently. These cations were found to be incorporated in the oxide film. (authors)

  9. Sizing cracks in thin-walled CANDU reactor pressure tubes using crack-tip diffraction

    International Nuclear Information System (INIS)

    The most practical nondestructive means of measuring the depth of cracks approximately 0.4 mm deep in CANDU reactor pressure tubes is the ultrasonic crack-tip diffraction method. Initially, optimum ultrasonic parameters for wave mode, transducer frequency, main-bang pulse characteristics, incident and diffracted angles were obtained on three fatigue cracks, based on the criteria of maximum signal amplitude and accuracy in determination of crack depth. In addition, three signal processing techniques, auto and cross-correlation, rectification and smoothing and the magnitude of the analytic signal, were used to obtain time measurements. The results of these measurements are presented. Except for the first fatigue crack, the depth calculations were accurate to within the specified range of ± 0.1 mm

  10. Flaw tolerance of the AISI 403 end fittings of CANDU pressure tubes

    International Nuclear Information System (INIS)

    The fuel channel assemblies in a CANDU nuclear reactor locate and support the fuel bundles in the reactor core and form part of the Primary Heat Transport System. Heavy water coolant flows through each fuel channel and over the fuel bundles to remove the heat generated by the fission reaction. The pressure tube is rolled into the end fitting at each end by a mechanically rolled joint. At the other end of the end fitting there is a seal which is removed to allow on-power refuelling. Thus the material must have a yield strength sufficient to make a leak tight rolled joint and adequate corrosion resistance at the seal face. End fittings are made of AISI 403, a 12% Cr-0.1% steel, which has the combination of properties needed. However, at the strength needed to make the rolled joint the mils lateral expansion (MLE) measured in Charpy tests does not consistently meet toughness requirements of Section III of the ASME Boiler and Pressure Vessel Code. A program was undertaken to demonstrate, by bursting end fittings containing machined flaws, that there was a safety factor of at least 3 at its end-of-life condition. The crack shape used was based on that specified in Appendix G of Section III but the crack was made deeper so that it was located in the region of high residual stress in the rolled joint. The partial thickness defect, a machined slot 17 mm deep and 75 mm long was fatigue sharpened before bursting

  11. Mechanistic modeling of bearing pad to pressure tube contact under localized high temperature conditions in a CANDU fuel channel

    International Nuclear Information System (INIS)

    During a postulated critical break LOCA (loss of coolant accident) in a CANDU reactor the coolant flow rates in the fuel channels of the flow pass of the reactor core downstream of the pipe break can rapidly reduce to very low values and remain very low for a period of tens of seconds following the break. Under the sustained low flow conditions, the fuel sheath (cladding) temperature in the affected channels rapidly increases and the coolant in the channels becomes significantly voided. The pressure tubes in the affected pass heat up under a combination of convection heat transfer from the low flows of superheated seam and thermal radiation heat transfer from the hot fuel. Additionally, hot spots may develop on the inner surface of pressure tubes at locations where the fuel bearing pads are in direct contact with the pressure tube. Localized thermal creep strain deformation at the hot spots is a potential pressure tube failure mechanism which could challenge fuel channel integrity. This paper evaluates the local thermal-mechanical deformation of a pressure tube in a CANDU reactor under critical break LOCA conditions tube using a coupled thermal-mechanical finite element COMSOL multi-physics model and investigates the conditions resulting in fuel channel failure due to localized contact between bearing pad and pressure. The mechanistic models are validated against data obtained from COG funded experiments performed at WRL (Whiteshell Research Laboratory). Multiphysics calculations are performed in which the heat transfer, thermal-mechanical and creep strain equations are solved, simultaneously. Heat conduction from bearing pads to the inner surface of the pressure tube is modeled with appropriate convective and radiation heat transfer boundary conditions. Thermal creep strain deformation of the Zr-2.5%Nb pressure tube is modeled using correlations derived from separate uniaxial tests that are reported in the literature. Contact conductance models based on

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

    Energy Technology Data Exchange (ETDEWEB)

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

    International Nuclear Information System (INIS)

    Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

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

    International Nuclear Information System (INIS)

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

  15. Eddy measurements on CANDU pressure tubes (I). The influence of the metallurgical states

    International Nuclear Information System (INIS)

    The experimental results, concerning the hydrogen content into the CANDU pressure tube by eddy current measurements, are described.The absorption and precipitation of hydrogen in zirconium alloys are affected by the macroscopic conditions (temperature, mechanical stresses etc.) and by the microstructural characteristics of alloy (grains structure, oxygen content etc.). The experimental study of electrical conductivity of Zr-2.5%Nb depending on the hydrogen content, include the influence of other processes, which affected the metallurgical state of samples. Consequently, the experimental study consists of the electrical conductivity measurements on the samples with the following characteristics: samples as received (CH0 = 9 ppm; CO0 = 1000 ppm), hydride samples by gas-metal reaction at thermodynamic equilibrium (CHmax = 155 ppm), oxidized samples (COmax = 1200 ppm), recrystallized samples by heating in gauge in the temperature range from 3000 deg. C to 8000 deg. C and in the time interval from 4 hours to 56.5 hours.The rectangular channel with axial orientation on the internal surface of the samples (reference defects) was processed for all the samples. The reference defects were dimensionally characterized before and after thermal treatments. No dimensional modification of the defects was observed. The influences of the metallurgical state on the defect signals are determined by electrical conductivity changes of samples. The values determined by eddy current method were compared with the electrical conductivity determined by four-probes method. The hydriding process induced decreases of the resistivity with Δρ≅8μΩ·cm. These results were obtained by both methods. For the oxidized samples, the results are the same for the eddy current and four-probes methods, but these are in contradiction with other results reported in the literature. It is possible that another effects associated with thermal treatments could induce the decrease of the electrical

  16. Update of operating experience with cold-worked Zr-2.5%Nb pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Zr-2.5 Nb pressure tubes are now used in all CANDU reactors. To ensure they perform reliably, their performance is carefully monitored. Both in situ inspection and sampling and testing techniques for tubes periodically removed from reactors have been developed. The data from these inspections and tests, together with models developed from research programs give confidence that pressure tubes will function effectively and safely for their design life. This presentation will describe how service life affects changes in the major material parameters in pressure tubes and the resulting maintenance activities resulting from those changes. Thermal creep, irradiation creep and irradiation growth change the dimensions during service, and axial elongation due to growth and sag due to creep in the older reactors have resulted in major maintenance programs. However, the dimensional changes continue to follow the behaviour predicted by the design equations and in the newer reactors should not limit service life. Extensive in situ sampling and the analysis of the tubes recently removed from Pickering Unit 3 indicate that hydrogen ingress into the pressure tubes from corrosion on the inside surface is very low and tests on irradiated material indicate that it should continue to remain low. The ingress rate from the annulus gas side can be significant if the integrity of the oxide on the outside surface is not maintained as a barrier. To maintain te integrity of the autoclave oxide, the recommended annulus gas is carbon dioxide, with oxygen addition, and adequate flow must be ensured. An explanation of the cause of relatively high hydrogen concentrations in a few Pickering A Zr-2.5% Nb pressure tubes has been developed defining the role of annulus side ingress. The model developed to predict the time and conditions to initiate blisters in pressure tubes that are in contact with their calandria tubes has been validated by the inspection, removal and examination of tubes and gives

  17. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Y. J.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2008-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes.

  18. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes

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

    International Nuclear Information System (INIS)

    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)

  20. Development of a CANDU fuel channel model to assess the effect of a pressure tube creep on the safety related parameters

    International Nuclear Information System (INIS)

    Recently the effect of pressure tube creep on the reactor safety in CANDUs emerges as an important issue of safety analysis due to a need for an extended operation. The accident analysis for the aged plants needs to incorporate major degradations of the plant performance in the safety analysis. In this paper, a CATHENA fuel channel model for studying the effects of the vertical offset of the fuel bundles in a crept pressure tube on the fuel and pressure tube cooling is developed. The current practice of the CANDU safety analysis assumes that the fuel bundles stay in a manner concentric to the pressure tube centerline even in the crept pressure tubes, whereas in reality the bundles sit at the bottom of the pressure tube. With this point in mind, 37-pin models with and without vertical offset of the bundle in the crept fuel channel are developed and tested for Reactor Outlet Header (ROH) 100% break LOCA accident, and results compared. As a result, it was found that the difference between the uncrept fuel channel model and the two crept fuel channel models, a concentric one and another vertically offset one, is quite significant, whereas the difference between the two crept fuel channel models is insignificant. Therefore it is concluded that the use of the concentric crept fuel channel model for the aged CANDU-6 safety analysis is justifiable for the first 200 sec into an accident. (author)

  1. Characterization of excel alloy pressure tube material for CANDU SCW reactors

    International Nuclear Information System (INIS)

    The phase transformation temperatures, aging response, and creep rupture strength of Zr alloy Excel (Zr- 3.5%Sn- 0.8%Nb- 0.8%Mo) pressure tube material were investigated. The α → α+β and α+β → β transus temperatures were found to be in the range of 600-690 °C and 962-975 °C respectively. Precipitation hardening was observed in the microstructures water-quenched from high in the α+β or β regions followed by aging at 400-500 °C for 1 hr. The results of creep-rupture experiments at 400 °C suggest that a fully martensitic and aged microstructure has better creep properties at high stress levels (>700 MPa) and a microstructure obtained by air-cooling from high in the α+β region shows good creep properties at lower stresses (<560 MPa). (author)

  2. Silicon carbide TRIPLEX materials for CANDU fuel cladding and pressure tubes

    International Nuclear Information System (INIS)

    Ceramic Tubular Products has developed a superior silicon carbide (SiC) material TRIPLEX, which can be used for both fuel cladding and other zirconium alloy materials in light water reactor (LWR) and heavy water reactor (CANDU) systems. The fuel cladding can replace Zircaloy cladding and other zirconium based alloy materials in the reactor systems. It has the potential to provide higher fuel performance levels in currently operating natural UO2 (NEU) fuel design and in advanced fuel designs (UO2(SEU), MOX thoria) at higher burnups and power levels. In all the cases for fuel designs TRIPLEX has increased resistance to severe accident conditions. The interaction of SiC with steam and water does not produce an exothermic reaction to produce hydrogen as occurs with zirconium based alloys. In addition the absence of creep down eliminates clad ballooning during high temperature accidents which occurs with Zircaloy blocking water channels required to cool the fuel. (author)

  3. Prediction of Axial and Radial Creep in CANDU 6 Pressure Tubes

    International Nuclear Information System (INIS)

    Status and proposals: 1. A review of literature concerning on the in-reactor deformation of PTs has been carried ouţ. 2. A model based on MFNN has been proposed to assess the radial and axial creep of CANDU 6 PTs. 3. Preliminary discussion with Cernavoda NPP (Romania) has been lunched, and now the preparation of official documents (collaboration in providing the inspection data from fuel channel in Unit 1 and 2) are in progress. 4. Further activities: • Improvement MFNN to accommodate complex data base (eventually with many variables) for radial and axial in-reactor deformation PT, and to satisfy the requirements from NPP Cernavoda and hopefully from present CRP database; • To build-up a database by running the creep equations (if the creep constants are provided by AECL); training of MFNN on them and to qualify it as a tool for PT in-reactor deformation prediction

  4. Selection of instruments used for vibration measurement of fuel bundles in a pressure tube under CANDU reactor operating conditions

    International Nuclear Information System (INIS)

    Vibration characteristics of CANDU fuel bundle and fuel elements is a key parameter considered in the design of a fuel bundle. Out-reactor frequency and temperature sweep tests, under reactor operating conditions, are performed to verify vibration characteristics of CANDU fuel bundles. Several options have been considered in the selection of vibration instrumentation to perform out-reactor frequency and temperature sweep tests. This paper compares the benefits and disadvantages of various vibration instruments and summarizes the rationale behind the selection of instruments used for vibration measurements over a range of temperature and pressure pulsation frequencies. The conclusions are presented from the bench tests performed, which confirm the use of the selected instruments. (author)

  5. Ultrasonic measurement method of calandria tube sagging in CANDU reactor

    International Nuclear Information System (INIS)

    Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor (calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the calandria tube (made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, it possibly contacts with liquid injection nozzle crossing beneath the calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, calandria tube and liquid inject ion tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here

  6. Delayed by hydrogen cracking: Cantilever beam test to determine radial direction KiH in CANDU Zr-2.5% Nb pressure tubes

    International Nuclear Information System (INIS)

    As part of the project for the development of the fabrication technology for pressure tubes for CANDU reactors the Cantilever beam test is being optimized in order to determine the KIH intensity factor. This determination establishes the fractal mechanical conditions for detention a fissure from developing by delayed by hydrogen cracking (DHC) in the tube's radial direction. This test is performed at temperature in a previously hydride test piece. For a given intensification factor of KI stresses the fissure will propagate depending on the precipitation of the hydrides in the pre-fissure. To determine the KIH factor, by definition the stress value (KI) by which the growth of the fissure in a radial direction is stopped, a flector momentum is applied that produces KI factors higher than the specified value of 7 MPa.m-0.5 to 250oC very gradually reducing the momentum until the fissure is arrested where the KI value that appears when the fissure is stopped is by definition KIH. This work detects the propagation of the fissure by AE (acoustic emission) which estimates a propagation time interval. The progress made to date on the detection of the propagation of a fissure and the microstructural studies that ascertain how it spreads are presented

  7. Improving the calandria tubes for CANDU reactors

    International Nuclear Information System (INIS)

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

  8. Ultrasonic measurement of gap between calandria tube and liquid injection nozzle in CANDU reactor

    International Nuclear Information System (INIS)

    Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor(Calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the Calandria tube(made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, ti possibly contacts with liquid injection nozzle crossing beneath the Calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, Calandria tube and liquid injection tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here. The gap between LIN and CT at site was measured using by this ultrasonic device at site

  9. Fuel for CANDU pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Unique properties, performance and evolution of CANDU fuel are described. The manufacturing conditions, uranium requirements, and fuel costs are discussed. The in-service performance of the fuel has been excellent and defect mechanisms and operating criterion are described. Evolutionary improvements in CANDU fuel and new fuel cycles such as plutonium and thorium are being explored to insure that the CANDU reactor remains competitive in the future. (author)

  10. Development of failure maps for integrity assessment of pressure tubes

    International Nuclear Information System (INIS)

    The issue of integrity of pressure tubes in CANDU reactors in accident conditions is examined based on the potential for local failure. The significance of the main parameters that influence the local strain failure are studied by a detailed parametric analysis. Based on this study a set of failure maps are presented for assessment of pressure tube integrity

  11. Bearing pad to pressure tube contact simulation

    International Nuclear Information System (INIS)

    Thermal creep strain deformation is a very important pressure tube failure mechanism. During a postulated LOCA (loss of coolant accident) with failure of emergency core injection sys- tem (ECIS), the fuel cladding temperature rapidly increases and the pressure tube becomes completely dry in a few seconds after flow stagnation occurs. Subsequently, the pressure tube circumference is heated by thermal radiation except at the spots where the bearing pads are in direct contact with the pressure tube. Therefore, the localized hot spots are developed on the pressure tube's inner surface under the bearing pads. The main objective of this paper is to evaluate the local thermal-mechanical deformation of a pressure tube in a CANDU reactor and to investigate the fuel channel integrity under localized contact between bearing pad and pressure tube. Furthermore, the mechanistic models are validated against the experimental works per- formed at WRL (Whiteshell research laboratory). Calculations are performed using the finite element method in which the heat, thermal mechanical and creep strain equations are solved, simultaneously. According to the experimental set up, the heat conduction from bearing pads to the inner surface of the pressure tube with appropriate convective and radiation boundary conditions has been simulated. Furthermore, the thermal creep strain deformation has been obtained for when the pressure tube is still under operational condition. It is observed that the pressure tube thermal strain will occur if sufficient high temperature is reached however, depending on the severity of flow degradation in the fuel channel, these localized hot spots could represent a potential creep strain failure of the pressure tube. Whether the pressure tube would fail at these hot spots before contacting the calandria tube depends on the localized temperature and experienced pressure transients. Sensitivity analysis is performed in order to evaluate the contact conductance, the

  12. Deformation of in-service pressure tubes

    International Nuclear Information System (INIS)

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

  13. Analysis of transient dry patch behavior on CANDU reactor calandria tubes in a LOCA with late stagnation and impaired ECI

    International Nuclear Information System (INIS)

    An analytical method to describe the behavior of transient dry patches on CANDU reactor calandria tubes has been developed. Dry patches may form following the sagging of a pressure tube onto a calandria tube in certain low-probability scenarios in which a loss-of-coolant accident occurs with subsequent failure or impairment of the emergency cooling injection function. Results of the analysis show that the dry patches will not grow beyond a few degrees on each side of the bottom of the calandria tube and will rewet within a few tens of seconds, with the values depending on the specific CANDU reactor design and the mechanism of dry patch formation and rewetting. Maximum local calandria tube temperatures reached during the transient will be about 5500C to 7000C. There will be no significant effects (0C) on fuel, sheath and maximum pressure tube temperatures. The analytical results provide confidence that pressure tube and calandria tube integrity will not be threatened by dry-patch formation in the LOCA scenarios studied

  14. Prediction of pressure tube ballooning under non-uniform circumferential temperature gradients and high internal pressure

    International Nuclear Information System (INIS)

    In some accident scenarios in CANDU reactors the pressure tube is expected to reach sufficiently high temperature at high internal pressure such that the pressure tube expands radially, i.e., the pressure tube balloons.Under these conditions it is of importance to the assessment of fuel channel integrity to be able to accurately predict the timing and extent of pressure tube ballooning. If the circumferential temperature gradient on the pressure tube is non-uniform, the resulting transverse hoop stress is non-uniform and the pressure tube experiences a non-uniform ballooning. This could result in a failure of the pressure tube before it balloons into contact with the surrounding calandria tube. The fuel channel integrity code SMARTT (Simulation Method for Azimuthal and Radial Temperature Transients) is used to predict the ballooning of CANDU Zr-2.5wt%Nb pressure tubes. The pressure tube strain rate calculation in SMARTT was extracted and used as the basis for the code PTSTRAIN which was constructed to model pressure tube ballooning with the temperature of the pressure tube and the internal pressure specified as the boundary conditions for the calculation. The main objectives of this paper are to describe the comparison of the predictions of this code against two different sets of experiments which were performed with defected and non-defected pressure tubes, and to provide further validation of the pressure tube ballooning model against independent experiments. (author)

  15. Rejection index for pressure tubes

    International Nuclear Information System (INIS)

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

  16. Designing and calculating the pressure loses for different geometries of CANDU type fuel clusters

    International Nuclear Information System (INIS)

    It is well known that circulation of the coolant through the pressure tube of a CANDU type reactor must ensure, through its flow rate values, the optimal conditions of heat transfer from the fuel clusters towards the heavy water. The flow rate through fuel channels differs from one another (up to 24 kg/s) depending on the fuel element sheath temperature, the latter depending in turn one the channels/clusters positions in the calandria vessel. In these conditions, one of the main problem of design in the CANDU type reactor plants is related to the hydraulic resistance represented by the fuel clusters loading the pressure tube or, in other words, the problem of pressure losses (pressure drops) over the length of the fuel cluster column. More precisely, this hydraulic resistance should not exceed a given value imposed by the performance calculations for the pumps used. A sustained activity of analysing comparatively the different geometry types of the fuel clusters was developed at INR Pitesti, a special attention being paid to their behavior as hydraulic resistances. The paper presents a set of computation programs devoted on one hand to the design of fuel clusters of different types and to an estimating computation of the pressure losses resulting from loading these clusters into a specific fuel channel of the CANDU type reactor, on the other hand. During the presentation of the work, different computing codes will be run for demonstration

  17. Innovation in pressure tube life assessment

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Proposal for a Coordinated Research Project: Prediction of Axial and Radial Creep in Pressure Tubes

    International Nuclear Information System (INIS)

    Participation of Argentina: • hydrogen charge of 90 samples for hydrogen determination: IGF, HVEMS, DSC, Resistivity, DTA; • 17 samples for non destructive techniques; • 6 blisters in CANDU pressure tube sections for NDT evaluation

  20. Pressure tube type reactor

    International Nuclear Information System (INIS)

    Heretofore, a pressure tube type reactor has a problem in that the evaluation for the reactor core performance is complicate and no sufficient consideration is made for the economical property, to increase the size of a calandria tank and make the cost expensive. Then, in the present invention, the inner diameter of a pressure tube is set to greater than 50% of the lattice gap in a square lattice like arrangement, and the difference between the inner and the outer diameters of the calandria tube is set smaller than 20% of the lattice gap. Further, the inner diameter of the pressure tube is set to greater than 40% and the difference between the inner and the outer diameters of the calandria tube is set smaller than 30% of the lattice gap in a triangle lattice arrangement. Then, heavy water-to-fuel volume ratio can be determined appropriately and the value for the coolant void coefficient is made more negative side, to improve the self controllability inherent to the reactor. In particular, when 72 to 90 fuel rods are arranged per one pressure tube, the power density per one fuel rod is can be increased by about twice. Accordingly, the number of the pressure tubes can be reduced about to one-half, thereby enabling to remarkably decrease the diameter of the reactor core and to reduce the size of the calandria, which is economical. (N.H.)

  1. Advancing the technologies of CANDU

    International Nuclear Information System (INIS)

    CANDU standard product design will continue to evolve, building upon the success of current operating units. Progressive improvements and enhancements will continue to be made to the CANDU system with heavy water moderated, pressure tube reactor technology of high neutron efficiency, based on the results of advanced technology R and D and operational experience from operating CANDU stations. The directions of development will respond to customer's requirements for economical, reliable and safe generating stations

  2. CANDU 9 - Overview

    International Nuclear Information System (INIS)

    The CANDU 9 plants are single unit versions of the very successful four unit Bruce B design, incorporating relevant technical advances made in the CANDU 6 and the newer Dalington and CANDU 3 designs. The CANDU 9 plant described in this paper is the CANDU 9 480/SEU with a net electrical output in the range of 1050 MW. In this designation 480 refers to the number of fuel channels, and SEU refers to slightly enriched uranium. Emphasis is placed on evolutionary design and the use of well-proven design features to ensure minimum financial risk to utilities choosing a CANDU 9 plant by assuring regulatory licensability and reliable operation. In addition, the CANDU 9 power plants reflect the important lessons learned by utilities in the construction and operation of CANDU units and, indeed, relevant experience gained by the world nuclear community in its operation of over 400 reactors of a variety of types. As a results, the CANDU 9 plants offer a high level of investment security to the owner, together with relatively low energy costs. The latter results from reduced specific capital cost, reduced operation and maintenance cost, and reduced radiation exposure to plant staff. A high level of standardization has always been a feature of CANDU reactors. This theme is emphasized in the CANDU 9 plants; all key components (steam generators, heat transport pumps, pressure tubes, fuelling machines, etc.) are of the same design as those proven in-service on operating CANDU power stations. The CANDU 9 power plants are readily adaptable to the individual requirements of different utilities and are suitable for a range of site conditions. (author). 12 figs

  3. Aspects regarding the behaviour in operation of the fuel channel and spacer ring between the calandria tube and the pressure tube

    International Nuclear Information System (INIS)

    This work presents the applied solutions for achievement of pressure tube support, motivation for choosing solutions of fitting fuel channels in the case of CANDU-6 from Cernavoda NPP and some experimental test results. (Author) 5 Figs

  4. CANDU bundle junction. Misalignment probability and pressure-drop correlation

    International Nuclear Information System (INIS)

    The pressure drop over the bundle junction is an important component of the pressure drop in a CANDU (Canada Deuterium Uranium) fuel channel. This component can represent from ∼ 15% for aligned bundles to ∼ 26% for rotationally misaligned bundles, and is dependent on the degree of misalignment. The geometry of the junction increases the mixing between subchannels, and hence improves the thermal performance of the bundle immediately downstream. It is therefore important to model the junction's performance adequately. This paper summarizes a study sponsored by COG (CANDU Owners Group) and an NSERC (National Science and Engineering Research Council) Industrial Research Grant, undertaken, at CRL (Chalk River Laboratories) to identify and develop a bundle-junction model for potential implementation in the ASSERT (Advanced Solution of Subchannel Equations in Reactor Thermalhydraulics) subchannel code. The work reported in this paper consists of two components of this project: an examination of the statistics of bundle misalignment, demonstrating that there are no preferred positions for the bundles and therefore all misalignment angles are equally possible; and, an empirical model for the single-phase pressure drop across the junction as a function of the misalignment angle. The second section of this paper includes a brief literature review covering the experimental, analytical and numerical studies concerning the single-phase pressure drop across bundle junctions. 32 refs., 9 figs

  5. Mechanistic modeling of pool film-boiling and quench on a Candu calandria tube following a critical break LOCA

    International Nuclear Information System (INIS)

    Following a postulated critical LBLOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a Candu CT (approximately 13 cm). A second order, non-linear and non-homogeneous ODE for vapour film thickness has been derived. The variation of steady state vapour film thickness prior to quench as a function of subcooling temperature, wall superheat, and incident heat flux is examined. The CT outer surface heatup rate and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (authors)

  6. Aging of elastomers in CANDU pressure boundary service

    International Nuclear Information System (INIS)

    This report describes the properties and aging of elastomers, and examines the performance of major elastomeric components in CANDU pressure boundary service. The components examined are vacuum building roof seals, pressure relief duct seals, airlock door seals, fuelling machine hoses, and cable penetrations. For each of these components, the design requirements, technical specifications and component testing procedures are compared with applicable standards. Information on actual and recommended monitoring and maintenance methods is presented. Operational and environmental stressors are identified. Component failure modes, causes and frequencies are described, as well as the remedial action taken. Many different elastomers are used in CANDU plants, for many different applications. Standards and manufacturers' recommendations are not consistent and may vary from one component to another. Accordingly, the monitoring, maintenance and replacement practices tend to vary from one application to another, and may also be different at different stations. Recommendations are given in this report for improved monitoring and maintenance, in an attempt to provide more consistency in approach. A summary of some experiences with elastomers from non-Canadian sources is contained in the last section. 125 refs

  7. A CFD Model for High Pressure Liquid Poison Injection for CANDU-6 Shutdown System No. 2

    International Nuclear Information System (INIS)

    In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a CFD code, to simulate the formation of the poison jet curtain inside the moderator tank. For validation, an attempt was made to validate this model against a poison injection experiment performed at BARC. As conclusion this set of models is judged to be appropriate. (authors)

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

    International Nuclear Information System (INIS)

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

  9. The pressure-tube sampler - communication theory in mechanical engineering

    International Nuclear Information System (INIS)

    I became involved in the mechanical engineering of the pressure-tube sampler indirectly after I was assigned related reactor-physics calculations. I discovered a problem in the proposed method of operating a machine inside an operating CANDU reactor, considered it as a problem in communication theory and thus created a number of conceptual solutions to the problem. The role of mathematics in this work has been questioned because the same concepts could have been created in other ways. (author). 5 refs

  10. Heat transfer parameters for glass-peened calandria tube in pressure tube and calandria tube contact conditions

    International Nuclear Information System (INIS)

    During a postulated event of large LOCA in CANDU reactors, the pressure tube may balloon to contact with its surrounding calandria tube to transfer heat to the moderator. To confirm the integrity of the fuel channel after the contact with a given moderator subcooling, many experiments have been performed in the last three decades by applying different pressure tube heatup rates, different pressure tube pressures and different moderator subcoolings for calandria tubes with smooth outer surface and glass-peened surface. A concept of Equivalent Moderator Subcooling (EMS) has been put forward to determine integrity of fuel channel upon pressure tube/calandria tube contact based on the existing experiment results. This concept has been presented in another work. In this work, the contact thermal conductance between pressure tube and calandria tube, critical heat flux, minimum film boiling temperature, empirical methods for nucleate boiling and film boiling heat transfer coefficient on the glass-peened calandria tube surface are discussed and estimated based on some experimental results and the EMS concept. These parameters are confirmed by simulating the existing experiments using a computer code. The estimated results may help detailed analyses on fuel channel integrity upon PT/CT contact if necessary. (author)

  11. Pressure tubes cracking due to DHC mechanism

    International Nuclear Information System (INIS)

    Zr-2.5wt%Nb alloy, used in fabrication CANDU and RMBK pressure tubes, fulfils the requirements of a material to be used under specific thermal, mechanical, irradiation and corrosion environment conditions in a nuclear reactor. Despite these advantages, the structural integrity of this assembly can be affected under certain conditions (stress, temperature, hydrogen concentration above the terminal limit of solubility), the crack initiation and propagation process being the mechanism responsible of this behaviour. During their operation the pressure tubes are susceptible to a stable cracking process referred to as Delayed Hydride Cracking (DHC). This phenomenon is one of the most important factors responsible for the degradation of these reactor components. The hydrogen concentration and the stress distribution are the parameters affecting this mechanism, leading to an embrittlement effect of the material, to a loss in the ductility and in the fracture toughness. Therefore, the structural integrity and the in-service lifetime are affected. The pressure tubes fabricated from zirconium alloy occlude exothermically hydrogen during manufacture and during operational service. When the concentration in solution exceeds the TLS (Terminal Limit of Solubility), the excess hydrogen precipitates as platelets of hydride. These hydrides are generally brittle in nature and therefore they cause a structural embrittlement, a loss of ductility and fracture toughness. Under certain thermo-mechanical conditions, the hydrides oriented after the fabrication process in the circumferential direction, tend to reorient. This phenomenon is responsible for a time-dependent failure through the hydride zone and it will arrest in the ductile zirconium matrix. This paper presents the experimental results obtained as a part of the IAEA Co-ordinated Research Project on 'Hydrogen and Hydride Induced Degradation of the Mechanical and Physical Properties of Zirconium-based Alloys', included like a

  12. General requirements for pressure-retaining systems and components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    This standard specifies the general requirements for the design, fabrication and installation of pressure-retaining systems, components, and their supports in CANDU nuclear power plants. (16 figs., 2 tabs., 25 refs.)

  13. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    International Nuclear Information System (INIS)

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D2O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D2O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

  14. Eddy Currents Inspection of CANDU Steam Generator Tubes using Zetec's ZR-1 Robot. Experience in Romania

    International Nuclear Information System (INIS)

    Full text of publication follows: The commercial operation of Unit 1 of Cernavoda NPP started on 2 December, 1996. The unit's reactor type is PHWR-CANDU 6 (electrical capacity 706 MWe), using natural uranium. The nuclear fuel is manufactured in Romania. The Cernavoda nuclear power plant has four CANDU - design steam generators that have been in service since 1996. The paper introduces the new ZR-1 Robot System for Inspection and Maintenance/Repair from Zetec that combines the newest state-of-the-art robotics technology with Zetec experience - based innovation to address the needs for inspection and repair of steam generators. The multipurpose ZR-1 can be easily installed to perform the necessary eddy current inspection and remain installed ready for follow-up maintenance and repair. It has superior technical performances and a modular three axis motion of arm that enables 100% coverage of tube sheet. Automated, repeatable, and precise positioning of tool heads ensures accurate delivery and reducing costly rework and reduces inspection time by 30%. The modular, light weight, and portable design permits easy assembly and disassembly through small openings and it reduces setup/tear down time by 30%. The first deployment of the new ZR-1 Robot was made in September 2004 at the Cernavoda NPP inspection outage. The unit's reactor type is PHWR-CANDU 6 (electrical capacity 706 MWe), using natural uranium; the nuclear fuel is manufactured in Romania. The Cernavoda nuclear power plant Unit 1 has four CANDU - design steam generators that have been in service since 1996. The paper presents also the Zetec's field experience and customer experience with this system. It describes the equipment setup in Cernavoda's steam generators mock-up, functional tests and calibration. Finally, provides details on the execution of the inspection, options for standardizing the inspection techniques and conclusions. (authors)

  15. Delayed hydrogen cracking test design for pressure tubes

    International Nuclear Information System (INIS)

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

  16. Numerical simulation of cross-flow in tube-bundles to model flow circulation of the moderator in CANDU-6

    International Nuclear Information System (INIS)

    The knowledge of external wall temperature distributions around calandria tubes is a major concern during normal and off-normal operating conditions of CANDU power reactors. To this aim, the use of Computational Fluid Dynamics (CFD) techniques to model moderator local flow velocities and temperatures can largely help in performing nuclear safety analyses. However, present numerical codes applied for this purpose makes use of the well known porous media approach. This method necessitates a previous knowledge of distributed hydraulic resistances that must be obtained from appropriate scaled experiments. Within this framework, this paper presents a set of 2D CFD simulations of incompressible cross-flows along in-line and staggered tube bundles. The numerical results are validated against experimental data obtained from the open literature. Calculations are performed using FLUENT-6 code. The Reynolds-Average Navier Stokes (RANS) equations are used in conjunction with several turbulence models and both the SIMPLE (Semi-Implicit Pressure Linked Equation) as well as the coupled pressure-based algorithm. In general, it is observed that two-equation turbulence models are able to reproduce mean velocities. Even though reasonably good predictions of flow distributions along staggered tube set-ups are obtained, the predictions of the pressure drop along in-line tubes are in general not satisfactory. In most cases, the coupled pressure-based algorithm seems to perform better but requires longer computation time. In general, the standard κ-ε is superior to others κ-ε models. The κ-ω model behaves better for fairly well developed flows. (author)

  17. The susceptibility of Candu steam generator tubing alloys to IGA/IGSCC in acid sulphate environments

    International Nuclear Information System (INIS)

    Constant extension rate tests (CERT) were carried out to assess the susceptibility of CANDU steam generator (SG) tube materials to intergranular stress corrosion cracking (IGSCC) in acidified sulphate solutions calculated to exist in SG crevices following sulphuric acid and Lake Huron water ingress. The results indicate significant susceptibility of Alloy 600 tubing (i.e. Bruce A) to IGSCC following sulphuric acid ingress and some susceptibility to intergranular attack (IGA) following Lake Huron water ingress. Alloy 800 was slightly susceptible to IGA in sulphuric acid ingress crevice chemistries but not at all to Lake Huron water ingress crevice chemistry. Alloy 690 was not susceptible to IGSCC or IGA in any of the chemistries tested. Preliminary results suggest that lead contamination of 1000 ppm does not increase the susceptibility of any of the Alloys tested to IGSCC following sulphuric acid ingress. (authors). 3 figs., 2 tabs., 6 refs

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

    International Nuclear Information System (INIS)

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

  19. Determination of dislocation density in Zr-2.5Nb pressure tubes by x-ray

    International Nuclear Information System (INIS)

    For X-ray determination of the dislocation density in CANDU Zr-2.5%Nb pressure tubes, a program was developed, using the Fourier analysis of X-ray line profiles and calculation of dislocation density by values of the coherent block size and the lattice distortion. The coincidence of obtained values of c- and a-dislocations with those, determined by the X-ray method for the same tube in AECL, was assumed to be the main criterion of validity of the developed program. The final variant of the program allowed to attain a rather close coincidence of calculated dislocation densities with results of AECL. The dislocation density was determined in all the zirconium grains with different orientations based on the texture of the stree-relieved CANDU tube. The complete distribution of c-dislocation density in -Zr grains depecding on their crystallographic orientations was constructed. The distribution of a-dislocation density within the texture maximum at L-direction, containing prismatic axes of all grains, was constructed as well. The analysis of obtained distributions testifies that -Zr grains of the stree-relieved CANDU tube significantly differ in their dislocation densities. Plotted diagrams of correlation between the dislocation density and the pole density allow to estimate the actual connection between texture and dislocation distribution in the studied tube. The distributions of volume fractions of all the zirconium grains depending on their dislocation density were calculated both for c- and a-dislocations. The distributions characterizes quantitatively the inhomogeneity of substructure conditions in the stress-relieved CANDU tube. the optimal procedure for determination of Nb content in β-phases of CANDU Zr-2.5%Nb pressure tubes was also established

  20. Determination of dislocation density in Zr-2.5Nb pressure tubes by x-ray

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Isaenkova, Perlovich; Cheong, Y. M.; Kim, S. S.; Yim, K. S.; Kwon, Sang Chul

    2000-11-01

    For X-ray determination of the dislocation density in CANDU Zr-2.5%Nb pressure tubes, a program was developed, using the Fourier analysis of X-ray line profiles and calculation of dislocation density by values of the coherent block size and the lattice distortion. The coincidence of obtained values of c- and a-dislocations with those, determined by the X-ray method for the same tube in AECL, was assumed to be the main criterion of validity of the developed program. The final variant of the program allowed to attain a rather close coincidence of calculated dislocation densities with results of AECL. The dislocation density was determined in all the zirconium grains with different orientations based on the texture of the stree-relieved CANDU tube. The complete distribution of c-dislocation density in -Zr grains depecding on their crystallographic orientations was constructed. The distribution of a-dislocation density within the texture maximum at L-direction, containing prismatic axes of all grains, was constructed as well. The analysis of obtained distributions testifies that -Zr grains of the stree-relieved CANDU tube significantly differ in their dislocation densities. Plotted diagrams of correlation between the dislocation density and the pole density allow to estimate the actual connection between texture and dislocation distribution in the studied tube. The distributions of volume fractions of all the zirconium grains depending on their dislocation density were calculated both for c- and a-dislocations. The distributions characterizes quantitatively the inhomogeneity of substructure conditions in the stress-relieved CANDU tube. the optimal procedure for determination of Nb content in {beta}-phases of CANDU Zr-2.5%Nb pressure tubes was also established.

  1. Development of delayed hydride cracking resistant-pressure tube

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  2. Development and validation of a model for high pressure liquid poison injection for CANDU-6 shutdown system no.2

    International Nuclear Information System (INIS)

    In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the calandria tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, an AEA Technology CFD code, to simulate the formation and growth of the poison jet curtain inside the moderator tank. For validation, the current model is validated against a poison injection experiment performed at BARC, India and another poison jet experiment for Generic CANDU-6 performed at AECL, Canada. In conclusion this set of models is considered to predict the experimental results in a physically reasonable and consistent manner. (author)

  3. Design features of Candu 9

    International Nuclear Information System (INIS)

    Thirty-two nuclear generating units with an aggregate installed capacity of 19,119 MWe worldwide are equipped with heavy water moderated and cooled pressure tube reactors of the Canadian Candu line. The list includes nine reactors of the 700 MWe category, and twelve reactors of the 900 MWe category in the Candu 6 series. On the basis of the 900 MWe units, Atomic Energy of Canada Ltd. (AECL) developed the advanced Candu 9 series by evolution. This series has been designed for a service life of sixty years. The use of modular, simplified units and systems in the Candu 9 design is to shorten the planning and construction phase, increase safety, and improve plant operation. AECL will offer this reactor on the world market, first to its customers in (South) Korea, which is one of the reasons why the safety parameters have been chosen especially under the aspect of seismic characteristics. (orig.)

  4. Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

    International Nuclear Information System (INIS)

    During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D2O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

  5. Analysis of the pressure tube failure at Pickering NGS A unit 2

    International Nuclear Information System (INIS)

    The failure of a Zircaloy-2 pressure tube in Pickering Unit 2 in August 1983 has been found to have been caused by an accelerated pickup rate of deuterium, contact between the pressure tube and its surrounding calandria tube, and rapid growth of zirconium hydride blisters. The pressure tubes in all later CANDU reactors are made from zirconium- niobium alloy. Examination of several Zircaloy-2 and zirconium niobium pressure tubes from different reactors has clearly shown that the deuterium pickup rate of the two materials is significantly different. The zirconium niobium pressure tubes have absorbed very little deuterium and they do not appear to be susceptible to the type of failure experienced at Pickering Unit 2

  6. The influence of heating rate on the pressure tube microstructure

    International Nuclear Information System (INIS)

    The aim of this paper is the study of the influence of heating rate on the microstructure morphology developed in pressure tube (Zr-2, 5% Nb alloy) at temperatures in the thermal transient conditions similar to LOCA (Loss of Coolant Agent) accident. The samples were tested using some thermal transient scenarios with different rates of heating. The thermal transients are performed at temperatures between 300 deg. C and 900 deg. C with the following heating rates: 5 deg. C/s, 10 deg. C/s and 20 deg. C/s. The average grain size and microhardness determination were performed in cross and longitudinal sections on the blank and tested samples. The analysis of microphotographs and data is presented as figures and diagrams. Some results of these experiments can be used in the study of degradation of the mechanical and micro-structural properties of the pressure tube materials in CANDU reactors. (authors)

  7. CANDU fuel bundle skin friction factor

    International Nuclear Information System (INIS)

    Single-phase, incompressible fluid flow skin friction factor correlations, primarily for CANDU 37-rod fuel bundles, were reviewed. The correlations originated from curve-fits to flow test data, mostly with new fuel bundles in new pressure tubes (flow tubes), without internal heating. Skin friction in tubes containing fuel bundles (noncircular flow geometry) was compared to that in equivalent diameter smooth circular tubes. At Reynolds numbers typical of normal flows in CANDU fuel channels, the skin friction in tubes containing bundles is 8 to 15% higher than in equivalent diameter smooth circular tubes. Since the correlations are based on scattered results from measurements, the skin friction with bundles may be even higher than indicated above. The information permits over- or under-prediction of the skin friction, or choosing an intermediate value of friction, with allowance for surface roughnesses, in thermal-hydraulic analyses of CANDU heat transport systems. (author) 9 refs., 2 figs

  8. Use of CATHENA to model calandria-tube/moderator heat transfer after pressure-tube/calandria-tube ballooning contact

    International Nuclear Information System (INIS)

    A study was performed to assess the effect of the calandria-tube/moderator heat transfer after pressure-tube/calandria tube ballooning contact using CATHENA. Results of this study indicated that the analytical tool, CATHENA, can be applied for pool boiling heat transfer on the external surface of a large diameter tube, such as the calandria tube used in CANDU reactors. The methodology in such CANDU-generic study can be used to simulate the tube surface with multiple boiling regimes and to assess the benefits of closely coupling thermalhydraulics modelling and fuel/fuel channel behaviour modelling. CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a one-dimensional, two-fluid thermalhydraulic simulation code designed by AECL to analyse two-phase flow and heat transfer in piping networks. The detailed heat transfer package in CATHENA allows a connection to be established from the multiple solid surfaces of tubes to the surrounding large amount of moderator water, which acts as a heat sink during a postulated loss of coolant event. The generalized heat transfer package within CATHENA allows the tube walls to be divided into several layers in the radial direction and several sectors in the circumferential direction, to account for heat transfer conditions in these two directions. The CATHENA code with the generalized heat transfer package is capable of capturing key pool-boiling phenomena such as nucleate, transition and film boiling heat transfer as well as an ability to model the rewet phenomenon to some extent. A CATHENA input model was generated and used in simulations of selected contact boiling experiment test cases. The transient wall temperatures have been calculated in different portions of the calandria tube. By using this model an adequate agreement was achieved between CATHENA calculation and experimental measurement The CATHENA code enables one to investigate the transient and local thermal-mechanical behaviour of the calandria tube

  9. Innovative Pressure Tube Light Water Reactor with Variable Moderator Control

    International Nuclear Information System (INIS)

    The features of a reactor based on multiple pressure tubes, rather than a single pressure vessel, provide the reactor with considerable flexibility for continuous design improvements and developments. This paper presents the development of innovative pressure tube light water reactor, which has the ability to advance the current pressure tubes reactors. The proposed design is aimed to simplify the pressure tubes reactors by: - replacing heavy water by a light water as a coolant and moderator, - adopting batch refueling instead of on-line refueling. Furthermore, the design is based on proven technologies, existing fuel and structure materials. Therefore, it is reasonable to expect significant capital cost savings, short licensing and introduction period of the proposed concept into the power production grid. The basic novelty of the proposed design is based on an idea of variable moderator content in the core and 'breed and burn' mode of operation. Both concepts were extensively investigated and reported in the past (2) (3) (4). In order to evaluate a practical reactor design build on proven technology, several features of the advanced CANDU reactor (ACR-1000) were adopted. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The proposed design is basically pressure tube light water reactor with variable moderator Control (PTVM LWR). This paper presents a detailed description of the PTVM core design and demonstrates the reactivity control and the 'breed and burn' mode of operation, which are implemented by the variation of the moderator in the core, from a

  10. Some aspects of the thorium fuel cycle in heavy-water-moderated pressure tube reactors

    International Nuclear Information System (INIS)

    The use of thorium fuel cycles in heavy-water-moderated pressure tube (CANDU) reactors will allow much more energy to be extracted from a given amount of fuel than is possible with the present natural uranium cycle. The extent to which various factors affect thorium fuel cycle economics and resource consumption with equilibrium 233U levels in the fuel is considered. Resource consumption in growing nuclear power systems is also considered, and it is shown that considerable savings can be achieved even under conditions of rapid growth. The main elements of the development program necessary to provide the technological base for thorium fuel cycles in CANDU reactors are discussed. (author)

  11. Some aspects of the thorium fuel cycle in heavy-water-moderated pressure tube reactors

    International Nuclear Information System (INIS)

    The use of thorium fuel cycles in heavy-water-moderated pressure tube (CANDU) reactors will allow much more energy to be extracted from a given amount of fuel than is possible with the present natural uranium cycle. The extent to which various factors affect thorium fuel cycle economics and resource consumption with equilibrium 233U levels in the fuel is considered. Resource consumption in growing nuclear power systems is also considered, and it is shown that considerable savings can be achieved even under conditions of rapid growth. The main elements of the development program necessary to provide the technological base for thorium fuel cycles in CANDU reactors are discussed

  12. Assessing CANDU requirements for irradiation - Research facilities

    International Nuclear Information System (INIS)

    The Canadian nuclear program needs ongoing access to irradiation-research facilities to support the safe operation of existing CANDU reactors and the evolutionary development of CANDU components and design features. The irradiation-research program must facilitate the testing of unique CANDU technology such as the fuel bundle, on-power refueling, the pressure tube, and the heavy-water coolant and moderator. Since 1957, NRU has operated as Canada's principal irradiation facility; however, it has become clear that NRU needs costly refurbishing if its lifetime is to be significantly extended. Accordingly, AECL has reviewed the requirements for CANDU irradiation research and is presently assessing alternatives for providing the necessary future access to irradiation-research facilities. Various options are under consideration, including renting space in existing research reactors, performing irradiations in CANDU power reactors, and building a new indigenous materials testing reactor specifically to meet essential program requirements

  13. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    The CANDU-Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper discusses the cost of producing electricity from CANDU, presents actual cost experience of CANDU and coal in Ontario, presents projected CANDU and coal costs in Ontario and compares CANDU and Light Water Reactor cost estimates in Ontario

  14. Reactor physics studies for a pressure tube supercritical water reactor (PT-SCWR)

    International Nuclear Information System (INIS)

    Preliminary lattice physics and full core neutronic analysis have been performed for the pressure-tube supercritical water reactor (PT-SCWR). Current CANDU reactor physics codes (WIMS-AECL and RFSP) were used for modeling this reactor. A key challenge in the physics design of this reactor is the optimization of lattice parameters to achieve the appropriate balance between coolant void reactivity (CVR) and fuel utilization. A vertically-oriented, batch-fuelled reactor is considered, with an insulated pressure tube to accommodate the high coolant temperatures and pressures. The analysis shows the reactor physics conceptual feasibility of the design, although further optimization is required. (author)

  15. Next generation CANDU plants

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water Reactors systems featuring horizontal fuel channels and heavy water moderator will continue to evolve, supported by AECL's strong commitment to comprehensive R and D programs. There are three key CANDU development strategic thrusts: improved economics, fuel cycle flexibility, and enhanced safety operation based on design feedback. Therefore, CANDU reactor products will continue to evolve by incorporating further improvements and advanced features that will be arising from our CANDU Technology R and D programs in areas such as heavy water and tritium, control and instrumentation, fuel and fuel cycles, systems and equipment and safety and constructability. Progressive CANDU development will continue in AECL to enhance the medium size product - CANDU 6, and to evolve the larger size product - CANDU 9. The development of features for CANDU 6 and CANDU 9 is carried out in parallel. Developments completed for one reactor size can then be applied to the other design with minimum costs and risk. (author)

  16. Eddy currents inspection of CANDU steam generator' tubes using Zetec's ZR-1 Robot: experience in Romania

    International Nuclear Information System (INIS)

    'Full text:' The paper introduces the new ZR-1 Robot System for Inspection and Maintenance/Repair from Zetec that combines the newest state-of-the-art robotics technology with Zetec experience-based innovation to address the needs for inspection and repair of steam generators. The multipurpose ZR-1 can be easily installed to perform the necessary eddy current inspection and remain installed ready for follow-up maintenance and repair. It has superior technical performances and a modular three axis motion of arm that enables 100% coverage of tube sheet. Automated, repeatable, and precise positioning of toolheads, ensures accurate delivery and reducing costly rework and reduces inspection time by 30%. The modular, lightweight, and portable design permits easy assembly and disassembly through small openings and it reduces setup/tear down time by 30%. The first deployment of the new ZR-1 Robot was made in September 2004 at the Cernavoda NPP inspection outage. The Cernavoda plant has four Advanced 600 MW CANDU-design generators that have been in service since 1996. The paper presents also the Zetec's filed experience and customer experience with this system. It describes the equipment setup in Cernavoda's generator mock-up, functional testes and calibration. Finally, provides details on the execution of the inspection, options for standardizing the inspection techniques and conclusions. (author)

  17. Two-phase natural circulation experiments in a pressurized water loop with CANDU geometry

    International Nuclear Information System (INIS)

    To provide information on two-phase natural circulation in a CANDU-type coolant circuit a series of tests has been performed in the RD-12 loop at the Whiteshell Nuclear Research Establishment. RD-12 is a 10-MPa pressurized-water loop containing two active boilers, two pumps, and two, or four, heated horizontal channels arranged in a symmetrical figure-of-eight configuration characteristic of the CANDU reactor primary heat-transport system. In the tests, single-phase natural circulation was established in the loop and void was introduced by controlled draining, with the surge tank (pressurizer) valved out of the system. The paper reviews the experimental results obtained and describes the evolution of natural circulation flow in particular cases as voidage is progressively increased. The stability behaviour is discussed briefly with reference to a simple stability model

  18. CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations both to the workers and the public

  19. Bleed condenser pressure relief valves for CANDU reactors

    International Nuclear Information System (INIS)

    Four test programs of Bleed Condenser pressure relief valves (RVs) have been completed. This work was initiated following two events in which Bleed Condenser REs were damaged. During one event, the piping connecting one of the RV's to the Bleed Condenser also failed. The author describes the investigations of these events and subsequent testing of several different RV designs. Based on the test results, pressure relief valve designs suitable for this service have been determined. Lessons learned are applicable to pressure relief valves in liquid services

  20. In-reactor deformation of a pilgered cold-worked Zr-2.5 wt% Nb pressure tube

    International Nuclear Information System (INIS)

    Zr-2.5 wt% Nb pressure tubes in CANDU reactors are cold-drawn to achieve the desired mechanical strength and tube dimensions. As part of an assessment of possible alternative fabrication processes, a pressure tube was cold-worked by pilgering and its dimensional stability monitored during service in the U-1 loop in the NRU test reactor. Strain rates have been determined for the pilgered tube from measured diameter and length changes after 22,060 h of reactor operation. These rates were compared with those for a cold-drawn tube also tested in NRU, and with the calculated behaviour of current power reactor pressure tubes extrapolated to the NRU test conditions. The deformation rates of the pilgered tube were found to be only marginally inferior to those of cold-drawn tubes

  1. Inert medium (helium) irradiation testing of pressure tube samples

    International Nuclear Information System (INIS)

    Irradiation tests currently performed in C-5 capsule aim at obtaining data and information concerning behavior to irradiation of pressure tubes of CANDU type fuel channel, to evidence the factors limiting operation life span. A calculation code for analysis and prediction of pressure tube behavior should be based upon periodical inspection results, post irradiation examination of the removed from reactor pressure tubes as well as on the experimental results obtained with materials subjected to irradiation conditions identical with the operational ones. Mechanical behavior analysis should focus both complex thermal-mechanical type stresses and mechanical properties alteration under irradiation. The experimental results should be applied: - to evaluate the irradiation effects upon mechanical properties of Zr-2.5% Nb exposed to fluences up to 1021 n·cm-2; - to gather data concerning the real stress / real deformation characteristic from which characteristic quantities can be deduced as, for instance, elasticity modulus, plasticity modulus, exponent of stress term in the Tsu-Berteles relation, to be used within the CANTUP simulation code describing pressure tube behavior, currently developed at INR Pitesti; - to develop prediction methods of pressure tube behavior and merging with in-service inspection procedure in order to forecast the life span and the proper timing for replacement before major failures occur. The samples irradiated in C-5 capsule were extracted from the ends of Zr-2.5% Nb pressure tubes resulting from Cernavoda NPP Unit 1. The samples for tensile tests were extracted on longitudinal and transversal directions of the pressure tube. The tests were carried out under following conditions: - test environment temperature, 260 - 280 deg.C; - testing medium, helium at 1 - 6 b pressure; - neutron flux (En > 1 MeV), 1 - 2 · 1013 ncm-2s-1; - neutron fluence (En > 1 MeV), 4 · 1020 ncm-2. The following characteristics were obtained from tensile test: - real

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

    International Nuclear Information System (INIS)

    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

  3. 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.

  4. 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

  5. Pressure tube rupture in a closed tank

    International Nuclear Information System (INIS)

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

  6. Computer simulation of oxide formation in ZR-NB pressure tubes

    International Nuclear Information System (INIS)

    Zr-2.5% Nb alloy pressure tubes are used in the core of CANDU nuclear reactors. During the operation of the reactor, oxidation takes place at both the interior and exterior surfaces of the pressure tube. The oxide film formed on the pressure tube surface serves as a protective barrier against hydrogen ingress. Therefore, it is important to predict and control the oxide texture, microstructure, and grain boundary distribution which ensures safe operation of the reactor. Our earlier research work has been focused on understanding the oxidation mechanism and oxide texture and microstructure development. Recently, efforts have been made to develop software that will eventually allow the prediction of oxide structure development which would result in optimizing the parameters that control the oxide structure. (author)

  7. A Feasibility Study of Surveillance of Pressure Tubes in Heavy Water Reactors

    International Nuclear Information System (INIS)

    All the Zr-2.5Nb pressure tubes used in all CANDU 6 plants and Wolsong nuclear power plants were made according to AECL's design to have a strong tangential texture for high creep resistance. Nevertheless, they are to be replaced before reaching their design lifetime due to higher growth and diametral creep than expected, the latter of which causes a reduction of the plant power to below the 100% full power after 15 years of operation. These instances of operational experience show that AECL's design of pressure tubes is invalid. Besides, comparison of in-reactor creep between pressure tubes with radial and tangential textures was made not on the condition of all variables being constant except texture but with both texture and the microstructure being varied. In other words, the in-reactor creep tests carried out by AECL turns out to be ineffective to single out the effect of texture on creep. In contract, comparison of in-reactor creep between CANDU Zr-2.5Nb tube and Russian Zr-2.5Nb tubes demonstrate that creep of Zr-2.5Nb tubes are governed not by texture but by another factor. Given that Russian TMT-2 tube with the lower degree of tangential texture shows two times lower creep rate than CANDU Zr-2.5Nb tube, it is suggested that the stability of Nb dissolved in the α-Zr grains is a creep controlling factor. Fortunately, given our improved pressure tubes being developed in cooperation with Russia that were made similar to the TMT-2 tube's manufacturing process but optimized for a better control of fine precipitates and texture, it is evident that our improved tubes will be better due to excellent creep resistance, higher fracture toughness and zero axial growth when compared to TMT-2 tube. Since the 2005 version of the CSA N285.4 stipulates surveillance of pressure tubes, material examination of Wolsong Unit-2 pressure tubes should be conducted since 2013. Korea Hydro Nuclear Power (KHNP)'s strategy is to prepare an alternative instead of material examination

  8. Economics of CANDU

    International Nuclear Information System (INIS)

    The cost of producing electricity from CANDU reactors is discussed. The total unit energy cost of base-load electricity from CANDU reactors is compared with that of coal-fired plants in Ontario. In 1980 nuclear power was 8.41 m$/kW.h less costly for plants of similar size and vintage. Comparison of CANDU with pressurized water reactors indicated that the latter would be about 26 percent more costly in Ontario

  9. Development of a CATHENA Fuel Channel Analysis Model for a Fuel Channel with Axial Variation of Radial Pressure Tube Creep in a Stratified Two-Phase Flow Condition

    International Nuclear Information System (INIS)

    A two-phase heat transfer phenomena in the fuel bundle strings located in a horizontal pressure tube with an axial variation of the radial creep, especially under a low stratified two-phase flow condition such as encountered in the CANDU reactor under the later stage of the blowdown phase of a LBLOCA, involves a complex heat transfer nature. This includes the conduction in the fuel rods, pressure tube, convection in the vapor and liquid regions, and radiation between the fuel rods exposed in the steam and the pressure tube, pressure tube and calandria tube. As these three modes of heat transfer has to be treated in a combined way, modeling the heat transfer phenomena inside the fuel bundle under the stratified flow during the later stage of LBLOCA blowdown has been one of the most challenging tasks in the CANDU safety analyses. The main reason for this hot attention is that it closely related to the integrity of the pressure tube. In this study a heat transfer model for handling this situation is developed, implemented and under preliminary testing of the analysis results. The analysis result up to now is encouraging and the validation of the model developed is ongoing. The major motivation of this study is to evaluate the conservatism of the current CANDU safety analysis methodology for a fuel channel with an axial variation of the radial creep of the pressure tube as easily experienced in the aged CANDU plant as it assumes the centerline of the fuel bundle string is the same as that of the pressure tube

  10. The CANDU Reactor System: An Appropriate Technology.

    Science.gov (United States)

    Robertson, J A

    1978-02-10

    CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity. PMID:17788102

  11. Probabilities of failure of a seam-welded calandria tube after a spontaneous pressure tube rupture

    International Nuclear Information System (INIS)

    This paper describes a methodology for calculating probabilities of seam-welded calandria tube (CT) failure after a sudden pressure tube (PT) rupture for operating conditions of CANDU reactors. Such a calculation is required in certain analyses or design option assessments such as those related to the issue of PT failure with consequential loss of moderator coincident with loss of emergency coolant injection (ECI). This accident scenario is the subject of Generic Action Item (GAl) 95G02 that was raised by the Canadian Nuclear Safety Commission (CNSC) in 1995. The CT failure probabilities were required as part of the resolution process for GAl-95G02. Two modes of CT failure considered are the prompt failure and the delayed creep failure. During the first half-second of CT pressurization by a spontaneous PT rupture, the plausible failure mechanism is the CT circumferential strain caused by a water-hammer type overpressure transient. The probability of prompt CT failure is calculated using a distribution of the measured failure strains for irradiated CTs and the calculated maximum CT strains resulting from water-hammer overpressure transients. If the CT survives the initial transient loading, the CT becomes the temporary pressure boundary to the primary heat transport system. Under certain pressure and temperature conditions the CT can experience slow-strain-rate plastic deformation and eventually fail by plastic strain. A time-to-rupture model is used to calculate the probability of this delayed creep failure within 5 to 15 minutes after a PT rupture. Then, the CT failure probability is calculated by combining prompt failure with delayed creep failure. (author)

  12. Cracking in hydride blisters in Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    When the pressure tubes contact to the calandria tubes in the CANDU reactor, temperature gradient in the Zr-2.5Nb pressure tube causes the thermal diffusion of hydrogen and formation of hydride blisters. This surface shape change is a result of the volume expansion associated with the transformation from pressure tube matrix to δ-phase hydride. Cracking in the hydride blisters may cause a direct failure of pressure tubes or develope to the delayed hydride cracking. The Zr-2.5Nb pressure tube specimen are hydrided by an electrolytic method and homogenized considering the temperature and time of hydrogen diffusion. The hydride blisters are formed on the outer surface of the specimen by a thermal diffusion between a heat bath maintained at the temperature of 415 deg C and an aluminum cold finger cooled with the flowing water of 15 deg C. An optical microscopy and 3-dimensional profilometry were used to characterize the hydride blisters with different hydrogen concentrations and thermal diffusion times. It reveals higher possibility of cracking for higher hydrogen concentration and longer time for thermal diffusion. The mechanism of cracking in the hydride blister is discussed

  13. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

  14. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations for the workers and the public

  15. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on worker and public safety, operating performance and costs, and reliability of system components

  16. Modelling of pressure tube Quench using PDETWO

    International Nuclear Information System (INIS)

    Transient two-dimensional heat conduction calculations have been carried out to determine the time-dependent temperature distribution in an overheated pressure tube during quenching with water. The purpose of the calculations is to provide input for evaluation of thermal (secondary) stresses in the pressure tube due to quench. The quench phenomenon in pressure tubes could occur in several hypothetical accident scenarios, including incidents involving intermittent buoyancy-induced flow during outages. In these scenarios, there will be two (radial and axial) or three dimensional temperature gradients, resulting in thermal stresses in the pressure tube, as the water front reaches and starts to cool down the hot pressure tube. The transient, two-dimensional heat conduction equation in the pressure tube during quench is solved using a FORTRAN package called PDETWO, available in the open literature for solving time-dependent coupled systems of non-linear partial differential equations over a two-dimensional rectangular region. This routine is based on finite difference solution of coupled, non-linear partial differential equations. Temperature gradient in the circumferential gradient is neglected for conservatism and convenience. The advancing water front is not modelled explicitly, and assumed to be at a uniform temperature and moving at a constant velocity inferred from experimental data. For outer surface and both ends of the pressure tube in the axial direction, a zero-heat flux boundary condition is assumed, while for the inner surface a moving water-quench front is assumed by appropriately varying the fluid temperature and the heat transfer coefficient. The pressure tube is assumed to be at a uniform temperature of 400oC initially, to represent conditions expected during an intermittent buoyancy-influenced flow scenario. The results confirm the expectations that axial temperature gradients and associated heat fluxes are small in comparison with those in the radial

  17. Development of the safety regulatory guides on the refurbishment for the CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. W.; Chin, T. E.; Rho, H. Y.; Park, H. B.; Yeom, H. G.; Hwang, G. M.; Hwang, B. G.; Seo, Y. H.; Lee, J. W. [Korea Power Engineering Co. Inc., Yongin (Korea, Republic of)

    2007-02-15

    In this study, requirements and standards concerned with safety performance for CANDU type reactors and review guidelines for facilities and performance concerned with refurbishment of major facilities such as pressure tubes, calandria tubes, and feeder popes were developed. To develop review guidelines for facilities and performance review concerned with refurbishment of CANDU reactors, review activities related with refurbishment and performance were categorized into designing and planning of equipments, removal and refurbishment of equipment, and confirmation of installation and inspection. As a result, following detailed review guidelines concerned with refurbishment of pressure tubes, calandria tubes, and feeder pipes in directly or indirectly referring to FSAR, design manual, startup-test manual were developed.

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

    International Nuclear Information System (INIS)

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

  19. Physics aspects of the pressure tube type SCWR preconceptual design

    International Nuclear Information System (INIS)

    One of the best options for meeting growing global energy needs is nuclear energy, since it is both emissions free and has the potential to be a sustainable energy source. The key areas for the development of future reactors are safety, sustainability, economics and security. Heavy water moderated reactors are an appealing option because of their improved neutron efficiency, which is advantageous from a sustainability standpoint. In an evolution of the current CANDU reactors, the pressure tube structure and heavy water moderator are retained in an advanced reactor design, cooled by supercritical light water. The use of supercritical light water as the coolant enables a large increase in thermal efficiency and therefore provides improved economic benefits. The use of thorium-based fuel provides improved safety, and a non-proliferative and sustainable fuel cycle. The optimization of the SCWR (Super Critical Water-cooled Reactor) is determined in part through reactor physics calculations, with respect to fuel utilization and safety (e.g. coolant void reactivity). These and other aspects of SCWR physics will be discussed in this paper. (author)

  20. Sample summary report for ROM-2 pressure tube sample

    International Nuclear Information System (INIS)

    The scope of this document is to describe the constructive structure, flaws type, and flaw dimensions for the pressure tube reference sample ROM-2 achieved by the Nuclear NDT Research and Services Company - Romania, and to present the results of the non-destructive examinations performed on this sample by Bhabha Atomic Research Center of India as investigating laboratory of the IAEA Coordinated Research Program (CRP) I3.30.10. For comparison purposes, the report contains the results of the non-destructive examinations performed on the sample ROM-2 by the Nuclear NDT Research and Services Company. This CRP is targeted to improve the structural integrity assessment of the fuel channels for CANDU nuclear reactors. ROM-2 sample inspection results, and sample summary results are reported in accordance with the new 'Terms of Reference' established at the consultancy held at IAEA Headquarters, Vienna, in November 2003. Specifically, the new form of the 'SAMPLE INSPECTION TABLE', which synthesizes all the inspection results, was implemented. The revision 1 of the ROM-2 sample summary report contains some corrections of the flaws detection results, because a number of eight intentional flaws of the ROM-2 sample were detected by the India inspection, but these flaws were characterized as non-intentional. Also, the revision 1 contains a more detailed analysis of the flaws sizing results

  1. CANDU-PHW fuel channel replacement experience

    International Nuclear Information System (INIS)

    One of the main characteristics of the CANDU pressurized heavy water reactor is the use of pressure tubes rather than one large pressure vessel to contain the fuel and coolant. This provides an inherent design capability to permit their replacement in an expeditious manner, without seriously affecting the high capacity factors of the reactor units. Of th eight Ontario Hydro commercial nuclear generating units, the lifetime performance places seven of them (including two that have had some of their fuel channels replaced), in the top ten positions in the world's large nuclear-electric unit performance ranking. Pressure tube cracks in the rolled joint region have resulted in 70 fuel channels being replaced in three reactor units, the latest being at the Bruce Nuclear Generating Station 'A', Unit 2 in February 1982. The rolled joint design and rolling procedures have been modified to eliminate this problem on CANDU units subsequent to Bruce 'A'. This paper describes the CANDU pressure tube performance history and expectations, and the tooling and procedures used to carry out the fuel channel replacement

  2. Delayed hydride cracking in Zr-2.5% wt Nb pressure tubes

    International Nuclear Information System (INIS)

    During service, pressure tubes of CANDU nuclear power reactor are prone to suffer crack growth by delayed hydride cracking (DHC). For a given H2 plus D2 concentration there is a critical temperature (Tc) below which DHC may occur. In this work, Tc was measured for CCT specimens cut from Zr-2.5 Wt % Nb pressure tubes. Hydrogen was added to the specimens to get concentrations of 40, 59 and 72 ppm. It was found that Tc is higher than the corresponding precipitation temperature. The axial crack velocity (Vp) was also measured. Decreasing temperature from Tc makes Vp increase until a maximum is attained at a temperature close to precipitation temperature. At lower temperatures, in the presence of precipitated hydrides, decreasing temperature implies lower velocities, following an Arrhenius law: Vp=Aexp(-Q/RT), with an activation energy Q= 66 KJ/mol K. (author)

  3. Advancing the CANDU reactor: From generation to generation

    International Nuclear Information System (INIS)

    Emphasizing safety, reliability and economics, the CANDU reactor development strategy is one of continuous improvement, offering value and assured support to customers worldwide. The Advanced CANDU Reactor (ACR-1000) generation, designed by Atomic Energy of Canada Limited (AECL), meets the new economic expectation for low-cost power generation with high capacity factors. The ACR is designed to meet customer needs for reduced capital cost, shorter construction schedule, high plant capacity factor, low operating cost, increased operating life, simple component replacement, enhanced safety features, and low environmental impact. The ACR-1000 design evolved from the internationally successful medium-sized pressure tube reactor (PTR) CANDU 6 and incorporates operational feedback from eight utilities that operate 31 CANDU units. This technical paper provides a brief description of the main features of the ACR-1000, and its major role in the development path of the generations of the pressure tube reactor concept. The motivation, philosophy and design approach being taken for future generation of CANDU pressure tube reactors are described

  4. Development of Regulatory Requirements and Inspection Guides for CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. W.; Kim, K.; Ryu, Y. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Ro, H. Y.; Jin, T. E. [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2009-05-15

    The first domestic CANDU power reactor, Wolsong unit 1, has been operated for about twenty years since commercial operation in 1983, and has been raised common aging issues of CANDU reactors in pressure tubes, calandria tubes, feeder pipes, etc. To solve these aging issues, utility is promoting the refurbishment activities for these major components. Therefore, confirmation and improvement for insufficient requirements considering the CNSC regulatory documents, regulatory principles between regulatory body and utilities related with refurbishment activities are required. These review contents are described herein, and representative review results are presented.

  5. 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.

  6. Eddy Currents Inspection of CANDU Steam Generators' Tubes using Zetec's ZR-1 Robot. Experience in Romania

    International Nuclear Information System (INIS)

    This is a PowerPoint presentation on behalf of COMPCONTROL ING, a Romanian private company established in 1997 the main services of which are enlisted. It is stressed that the most suitable type of inspection in terms of safety and reliability for the steam generator tubes is eddy current (EC) method. The advantages of EC testing include the following: - Extremely fast; - Accurate in detection and sizing of discontinuities; - Very good method for baseline screening; - Very high detection sensitivity to physical-chemical variations of the test specimen; - Easy setup and application for automated inspection; - Portable equipment designing; - Use of multiple channels and multi-frequencies for a better screening of signals and efficiency; - High capability to store the data for future review and comparison (using data history to evaluate the rate of degradation and life assessment studies). Between 2003 and 2005 ECT was applied to Cernavoda NPP U1 SGs as follows: - in 2003, SG-4; - in 2004, SG-2; - in 2005, SG-1; - in 2005, SG-3; - in 2005, SG-4. The purpose of inspection with eddy currents of SGs tubes was: - Detection, sizing and evaluation of possible degradations of the tubes and at the interface tube/support structures (tubesheet, tube support plates and baffles); - Completion of the baseline data for future review and comparison. The software used for acquisition and analysis of eddy current data and for inspection management were: - ZETEC Eddynet-R Zetec Acquisition Control-ZAC; - ZETEC Eddynet-R Data Analysis (bobbin and MRPC); - ZETEC Eddynet-R Data Management. The equipment ZR-1 is described and its advantages as well. Advantages of the automated scanning system are highlighted as follows: - Repeatability; - High resolution mapping; - Accurate indexing; - Minimize changes in lift-off resulting from probe wobble, eccentricity of the tube and surface irregularities; - 3-part design makes each component lighter and more compact for easier, faster installation

  7. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10-6/year. 95 refs, 3 tabs

  8. Modelling nuclear fuel vibrations in horizontal CANDU reactors

    International Nuclear Information System (INIS)

    Flow-induced fuel vibrations in the pressure tubes of CANDU reactors are of vital interest to designers because fretting damage may result. Computer simulation is being used to study how bundles vibrate and to identify bundle design features which will reduce vibration and hence fretting. (author)

  9. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost

    International Nuclear Information System (INIS)

    A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (∼49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC

  10. CANDU steam generator life management

    Energy Technology Data Exchange (ETDEWEB)

    Tapping, R.L. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Nickerson, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Spekkens, P.; Maruska, C. [Ontario Hydro, Toronto, Ontario (Canada)

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDUutilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

  11. 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)

  12. Research and development for CANDU fuel channels and fuel

    International Nuclear Information System (INIS)

    The CANDU nuclear reactor is distinctly different from BWR and PWR reactors in that it uses many small pressure tubes rather than one large pressure vessel to contain the fuel and coolant. To exploit the advantages of the natural uranium fuel, the pressure tubes, like other core components, are manufactured from zirconium alloys which have low neutron capture cross sections. Also, because natural uranium fuel only achieves a modest burnup, a simple and inexpensive fuel design has been developed. The present paper reviews the features and the research that have led to the very satisfactory performance of the pressure tubes and the fuel in CANDU reactors. Reference is made to current research and development that may lead to further economies in the design and operation of future power reactors. (author)

  13. CANDU improvement

    International Nuclear Information System (INIS)

    The evolution of the CANDU family of nuclear power plants is based on a continuous product development approach. Proven equipment and system concepts from operating stations are standardized and used in new products. Due to the modular nature of the CANDU reactor concept, product features developed for CANDU 9 can easily be incorporated in other CANDU products such as CANDU 6. Design concepts are being developed for advanced CANDU 6 or larger advanced CANDU, depending on the number of fuel channels and the fuel cycle selected. This paper provides a description of the design improvements being incorporated in CANDU 9 and further design enhancements being studied for future incorporation in CANDU 6 or larger advanced CANDU meeting the requirements of future CANDU owners. The design enhancement objectives are: To improve operational simplicity by applying modern information technology; to improve safety in a cost effective way; to improve system and component reliability and to increase plant life; to improve economics and to reduce owners' risks during all phases of a project using up-front licensing, an improved engineering process and project tools during design, construction and operation; to continue to exploit the neutron economy of CANDU with the development of advanced fuels and fuel cycles. (author)

  14. INR experience concerning the assessment of the CANDU steam generator tubing material degradation

    International Nuclear Information System (INIS)

    Steam generator degradation has caused substantial losses of power generation, resulted in large repair and maintenance costs. Institute for Nuclear Research has carried out an extensive R and D program focused on the understanding of the degradation processes especially for the tubing material and on developing remedial actions in the purpose to prevent and diminish the ageing process of which evolution supposes some considerable economic costs. Because of the huge impact of corrosion, it is imperative to have a systematic approach to recognizing and mitigating corrosion problems as soon as possible after they become apparent. A proper failure analysis includes collection of pertinent background data and service history, followed by visual inspection, photographic documentation, material evaluation, data review and conclusion procurement. In analyzing corrosion failures, one must recognize the wide range of common corrosion mechanisms. The features of any corrosion failure give strong clues as to the most likely cause of the corrosion. The principal steps of analysis and diagnosis of the steam generator tubes degradations consist in: visual inspection, chemical analysis, cross section examination by optical and scanning electron microscopy and RDX, data review, conclusions and recommendations. This paper details a proven approach to properly determining the root cause of a failure, and includes metallographic illustrations of the most common corrosion mechanisms, including general corrosion, pitting, crevice corrosion, corrosion fatigue and intergranular corrosion. (author)

  15. Burn up Analysis for Fuel Assembly Unit i n a Pressurized Heavy Water CANDU Reactor

    International Nuclear Information System (INIS)

    MCNPX code has been used for modeling a nd simulation of an assembly of CANDU Fuel bundle . The assembly is composed of a heterogeneous lattice of 37-element natural Uranium fuel, heavy water moderator and coolant. The fuel bundle is burnt in normal operation conditions of CANDU reactors. The effective multiplication factor (Keff ) of the bundle is calculated as a function of fuel burnup. The flux and power distribution are determined. Comparing t he concentrations of both Uranium and Plutonium isotopes are analyzed in the bundle. The results of the present model with the results of a benchmark problem, a good agreement was found PWR

  16. Advances in fuel channel technology for CANDU reactors

    International Nuclear Information System (INIS)

    The components of the CANDU fuel channels are being developed to have service lives of over 30 years with large margins of safety. Information from research programs and the examination of components removed from reactors has enable improvements to be made to pressure tubes, spacers, calandria tubes and end fittings. Improvements have also been made to the channel design to facilitate planned retubing. (author). 22 refs., 5 tabs., 31 figs

  17. CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU (CANada Deuterium Uranium) Pressurized Heavy Water (PHW) type of nuclear-electric generating station was developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper summarizes Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components, and nuclear safety considerations to both the workers and the public

  18. CANDU 300

    International Nuclear Information System (INIS)

    The CANDU nuclear power system is under continuous review by AECL in order to advance the CANDU concept in a manner that will assure competitiveness in both current and future markets. Over the past three years development effort has featured the CANDU 300, a CANDU nuclear generating station with a net output in the range of 320 MW9e) to 380 MW(e). At the outset AECL recognized that coal-fired power plants would be the primary competition for the CANDU advantages such as the use of natural uranium fuel and on-power refuelling, while enhancing capacity factor, reducing man-rem exposure, reducing capital cost, and minimizing construction schedules. AECL believes that the resulting CANDU 300 nuclear generating station will have substantial appeal to many utilities, in both developed and developing countries. The key features of the CANDU 300 are presented here, with particular attention to the station layout, construction methods, and construction schedules

  19. Evolution of CANDU vacuum building and pressure relief structures from Pickering NGS A to Darlington NGS A

    International Nuclear Information System (INIS)

    The vacuum building (VB) and pressure relief structures (PRS) are the unique features of multiple unit CANDU containments. In case of loss-of-coolant accident, the released radionuclides are drawn through the PRS into the subatmospheric VB, doused and contained without being released to the environment. This paper describes the differences in design, configuration and layout of the VB and PRS from Pickering NGS A to Darlington NGS A due to new developments in design concepts and to requirements which have proceeded from the experience gained in both the design and operation of the nuclear stations. (orig.)

  20. Characterization of elastic properties of Zr-2.5% Nb pressure tube by measurements of sound velocity

    International Nuclear Information System (INIS)

    The cold-worked Zr-2.5% Nb alloy is used as material for the pressure tubes of CANDU (CANadian Deuterium Uranium) nuclear reactors. During the service life in reactor, diffusion of hydrogen and/or deuterium in the pressure tubes wall will occur. Below a certain temperature, a stable hydride of zirconium will form, as a brittle phase which could lead to catastrophic failure. In the present paper, the influence of hydrogen on the acoustic-elastic properties of Zr-2.5% Nb alloy will be investigated using non-destructive method based on measurements of ultrasonic velocity. In order to obtain the most usual elastic coefficients on a given direction (axial and circumferential) of the tube, both longitudinal VL and transversal VT phase velocities have been experimentally determined. (authors)

  1. Simulation and analysis of bearing pad to pressure tube contact heat transfer under large break LOCA conditions

    International Nuclear Information System (INIS)

    In some postulated loss-of-coolant accidents (LOCAs) in a CANDU reactor, localized 'hot spots' can develop on the pressure tube as a result of decay heat dissipation by conduction through bearing pad/pressure tube contact locations. Depending on the severity of flow degradation in the channel, these 'hot spots' could represent a potential threat to fuel channel integrity. The most important parameter in the simulation of BP/PT contact is the contact conductance. Since BP/PT thermal contact conductance is a complex parameter which depends upon the thermal and physical characteristics of the material junction and the surrounding environment, contact conductance is determined from experiments relevant to the reactor conditions. A series of twelve full scale integrated BP/PT contact experiments have been conducted at AECL-WRL under CANDU Owner Group (COG). The objective of the experiments was to investigate the effect of BP/PT contact on PT thermal-mechanical behaviour. This paper presents the simulation of BP/PT interaction integrated experiments using SMARTT and MINI-SMARTT computer codes and subsequent derivation of the BP/PT contact conductance by best fitting of the experimental pressure tube temperature measurements. (author)

  2. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

    The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book Canada Enters the Nuclear Age. The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

  3. Eccentric pressurized tube for measuring creep rupture

    International Nuclear Information System (INIS)

    Creep rupture is a long term failure mode in structural materials that occurs at high temperatures and moderate stress levels. The deterioration of the material preceding rupture, termed creep damage, manifests itself in the formation of small cavities on grain boundaries. To measure creep damage, sometimes uniaxial tests are performed, sometimes density measurements are made, and sometimes the grain boundary cavities are measured by microscopy techniques. The purpose of the present research is to explore a new method of measuring creep rupture, which involves measuring the curvature of eccentric pressurized tubes. Theoretical investigations as well as the design, construction, and operation of an experimental apparatus are included in this research

  4. CANDU fuel cycle flexibility

    International Nuclear Information System (INIS)

    High neutron economy, on-power refuelling, and a simple bundle design provide a high degree of flexibility that enables CANDU (Canada Deuterium Uranium; registered trademark) reactors to be fuelled with a wide variety of fuel types. Near-term applications include the use of slightly enriched uranium (SEU), and recovered uranium (RU) from reprocessed spent Light Water Reactor (LWR) fuel. Plutonium and other actinides arising from various sources, including spent LWR fuel, can be accommodated, and weapons-origin plutonium could be destroyed by burning in CANDU. In the DUPIC fuel cycle, a dry processing method would convert spent Pressurized Water Reactor (PWR) fuel to CANDU fuel. The thorium cycle remains of strategic interest in CANDU to ensure long-term resource availability, and would be of specific interest to those countries possessing large thorium reserves, but limited uranium resources. (author). 21 refs

  5. Advanced heavy water reactor pressure tube-easy replaceability

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is a 300 MWe vertical pressure tube type reactor. A coolant channel consists of pressure tube, made of Zr-2.5 % Nb, which is separated from cold calandria tube using garter spring spacers. The principal function of pressure tube is to support and locate the fuel assembly and allows light water coolant through fuel assembly by natural circulation. Since AHWR is designed for life of 100 years, it necessitates the replacement of pressure tubes during service life. Easy replaceability of pressure tube, along with surveillance requirements, has major bearing on the design of coolant channel assembly. The several systems and tools have been conceptualised to cater the needs for easy and quick replacement of a pressure tube during reactor shut down. This paper gives the highlights of the innovative design features of coolant channel, preliminary design and pre-requisites for replacement, and experimental programme for demonstration of easy replaceability. (author)

  6. The transformation behaviour of the β-phase in Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    In a CANDU reactor, the fuel bundles and primary coolant are contained within Zr-2.5Nb pressure tubes that are approximately 6.3 m in length, have an internal diameter of 104 mm and a wall thickness of 4.2 mm. The Zr-2.5Nb pressure tubes are nominally extruded at 815 deg. C, cold-worked 27%, and stress relieved at 400 deg. C for 24 h, resulting in a structure consisting of elongated grains of hexagonal-close-packed α-Zr, partially surrounded by a thin network of filaments of body-centred-cubic β-Zr. These β-Zr filaments are metastable and contain about 20% Nb. The stress-relief treatment results in partial decomposition of the β-Zr filaments with the formation of hexagonal-close-packed ω-phase particles that are low in Nb, surrounded by a Nb-enriched β-Zr matrix. A temperature-time-transformation (TTT) diagram has been developed for the β-phase in Zr-2.5 wt%Nb pressure tubes. The results show that the morphology and/or physical state of the β-phase has a significant effect on the transformation behaviour compared with a bulk Zr-20 wt%Nb alloy. This means that a specific TTT-diagram is required to describe the behaviour of these engineering components

  7. Candu 6: versatile and practical fuel technology

    International Nuclear Information System (INIS)

    CANDU reactor technology was originally developed in Canada as part of the original introduction of peaceful nuclear power in the 1960s and has been continuously evolving and improving ever since. The CANDU reactor system was defined with a requirement to be able to efficiently use natural uranium (NU) without the need for enrichment. This led to the adaptation of the pressure tube approach with heavy water coolant and moderator together with on-power fuelling, all of which contribute to excellent neutron efficiency. Since the beginning, CANDU reactors have used [NU] fuel as the fundamental basis of the design. The standard [NU] fuel bundle for CANDU is a very simple design and the simplicity of the fuel design adds to the cost effectiveness of CANDU fuelling because the fuel is relatively straightforward to manufacture and use. These characteristics -- excellent neutron efficiency and simple, readily-manufactured fuel -- together lead to the unique adaptability of CANDU to alternate fuel types, and advancements in fuel cycles. Europe has been an early pioneer in nuclear power; and over the years has accumulated various waste products from reactor fuelling and fuel reprocessing, all being stored safely but which with passing time and ever increasing stockpiles will become issues for both governments and utilities. Several European countries have also pioneered in fuel reprocessing and recycling (UK, France, Russia) in what can be viewed as a good neighbor policy to make most efficient use of fuel. The fact remains that CANDU is the most fuel efficient thermal reactor available today [NU] more efficient in MW per ton of U compared to LWR's and these same features of CANDU (on-power fuelling, D2O, etc) also enable flexibility to adapt to other fuel cycles, particularly recycling. Many years of research (including at ICN Pitesti) have shown CANDU capability: best at Thorium utilization; can use RU without re-enrichment; can readily use MOX. Our premise is that

  8. Uses of Plutonium Fuel in Pressure-Tube-Type, Heavy-Water-Moderated Thermal Reactors

    International Nuclear Information System (INIS)

    In 1962, a feasibility study was begun in the JAERI on the uses of various nuclear fuels for pressure-tube-type, heavy-water-moderated thermal reactors. This study began with analysis of the use of uranium in heavy-water-moderated thermal reactors such as the CANDU-PHW, CANDU-BLW, SGHW, EL-4, and Ref. 15, D and E lattices, which is designed in the JAERI, from the standpoint of the core design. Then, the ways of using plutonium fuel in the same types were investigated using WATCHTOWER, FLARE and VENUS codes, including: (1) direct substitution of the plutonium from light-water reactors or Magnox reactors, (2) recycle use of the plutonium from heavy-water-moderated reactors, (3) plutonium self-sustaining cycle, and (4) plutonium phoenix fuel. The following conclusions are reported: (1) In the direct substitution of plutonium, somewhat depleted plutonium is more suitable for core design than the plutonium from Magnox reactors or light-water reactors, because the increase in the initial reactivity due to large plutonium absorption cross-section must be prevented. (2) In the plutonium self-sustaining cycle, the fuel burn-up of about 15 000 ∼20000 MWd/t would be expected from natural uranium, and the positive void reactivity which always occurs in the uraniumloaded SGHW or CANDU-BLW lattices is greatly reduced, the latter property giving some margin to bum-out heat flux. (3) It may be concluded from the fuel cycle analysis that the plutonium self-sustaining cycle is equivalent to using slightly enriched uranium (about 1.0 at.%). It may be concluded that the use of plutonium in heavy-water-moderated reactors is technologically feasible and economically advantageous. (author)

  9. Ultrasonic testing of the fracture toughness of Zr-Nb pressure tubes

    International Nuclear Information System (INIS)

    Bulk elastic properties were measured, using ultrasound, in thickness and circumferential directions of 13 Zr-Nb pressure tubes samples from CANDU nuclear reactors in the hope of finding a nondestructive means to evaluate fracture toughness. The longitudinal wave velocity in the thickness direction are found especially sensitive to changes in the α-Zr single-crystal c axis orientation distribution. This is verified by comparing measured values to predictions based on neutron diffraction measurements of the crystallographic orientation distribution and on the single-crystal elastic constants of α-Zr. Moreover, those velocities most sensitive to texture correlate best with the crack growth toughness of the pressure tubes. This led to the discovery of a correlation between the degree of alignment of the crystallographic c axes along the tube circumferential direction and crack growth toughness. The better is the alignment, the lower is crack growth toughness. Because of the measurement simplicity, the ultrasonic technique could be developed into a rugged industrial sensor

  10. The pressure tube inspection and integrity evaluation in Fugen

    International Nuclear Information System (INIS)

    Two hundred and twenty four pressure tubes are installed vertically in the reactor of the Fugen. Each pressure tube accommodates a fuel assembly and forms the pressure boundary of the primary cooling system. The operating pressure is 6.8 MPa and temperature 280degC. The pressure tube, made of Heat Treated Zr-2.5%Nb, is approximately five meter long, with 117.8 mm inner diameter and 4.3 mm wall thickness. The pressure tube is connected to upper and lower extension tubes of stainless steel with the rolled-joint technique. The soundness of the pressure tube, rolled joints, and upper/lower extension tubes must be checked in an appropriate and systematic manner. To satisfy the requirements, in-service inspections (ISIs) and post irradiation examinations (PIEs) of the pressure tubes have been carried out during the 24 year operation of the Fugen. Development of pressure tube inspection equipment started in 1971 for the ISIs. The first model of the equipment was developed and applied to the pre-service inspection of the pressure tubes in 1977. The measurement accuracy of the equipment was sufficient but the weight and size were too large to be set and handled in an irradiated environment. Thus, the design was modified to smaller the equipment in size, lighten to approx. 1/100 in weight, and realize to be handled with the refueling machine. The improved equipment was used in the 4th annual inspection in March 1984. Ultrasonic flaw detections, inner diameter measurements and inner surface visual inspection of pressure tubes were conducted. Up to the 17th annual inspection in 2002, 146 inspections in total were executed. The ultrasonic inspection detected no defect on the pressure tubes. The measured strain due to the irradiation creep of pressure tubes corresponded with the design values. To conduct the PIEs of the pressure tube materials, surveillance specimens were set in the special fuel assemblies and irradiated from the beginning of the reactor operation. Five PIEs

  11. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Hydrogen atom has two isotopes: deuterium 1H2 and tritium 1H3. The deuterium oxide D2O is called heavy water due to its density of 1105.2 Kg/m3. Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D2O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D2O management is required to preserve it. (author)

  12. Pickering NGS A: Assessment of calandria tube integrity following a sudden pressure tube failure

    International Nuclear Information System (INIS)

    The issue of calandria tube integrity following a sudden rupture of the pressure tube in Pickering NGS A reactor is addressed. Based on operating experience, only fish-mouth ruptures of the pressure tube are considered to be credible. The calandria tube response to the pressure tube break is delineated into three distinct stages, i.e. the initial transient response during the annulus filling stage, transient overpressurization and the final steady-state loading after bellows failure. The annulus response in the second stage is dominated by a waterhammer type overpressure transient with attenuation of this transient due to plastic straining of the calandria tube. The annulus pressure transients for various breaks and the sensitivity of the results to various parameters are presented. The strength margins of the calandria tube are evaluated to be relatively large. (author). 7 refs., 6 tabs., 6 figs

  13. CATHENA Code Assessment for Pressure Tube and Calandria Tube Contact Phenomena

    International Nuclear Information System (INIS)

    Canadian Algorithm for THErmalhydraulic Network Analysis (CATHENA), has been validated against full-scale Contact Boiling Experiments conducted using specific channel power, pressure, and moderator subcooling as pre-test conditions. The pressure tube (PT) and calandria tube (CT) temperatures, the extent of dryout and failures of the pressure tube or the calandria tube (if any) are the outcome of these experiments. Recently, an IAEA International Collaborative Standard Problem (ICSP) to provide contact boiling experimental data to participants for assessing the subcooling requirements for a heated pressure tube, plastically deforming into contact with the calandria tube during a postulated large break LOCA condition has been performed. The CATHENA code assessment results against the experimental data distributed for the ICSP are provided in this paper. The CATHENA code is used to simulate the experiment on pressure tube ballooning conducted at the AECL. The overall code's predictions show good agreements with the experimental data. The contact timing by the pressure tube ballooning is predicted accurately, however, it is found that the code largely underpredict the peak temperature at the pressure tube and the calandria tube. This discrepancy seems to be induced from multi-dimensional flow effects in the water tank. For more accurate calculations, detailed modeling of the water tank is required

  14. Inspection of Candu Nuclear Reactor Fuel Channels

    International Nuclear Information System (INIS)

    The Channel Inspection and Gauging Apparatus of Reactors (CIGAR) is a fully atomated, remotely operated inspection system designed to perform multi-channel, multi-task inspection of CANDU reactor fuel channels. Ultrasonic techniques are used for flaw detection, (with a sensitivity capable of detecting a 0.075 mm deep notch with a signal to noise ratio of 10 dB) and pressure tube wall thickness and diameter measurements. Eddy currrent systems are used to detect the presence of spacers between the coaxial pressure tube and calandria tube, as well as to measure their relative spacing. A servo-accelerometer is used to estimate the sag of the fuel channels. This advanced inspection system was commissioned and declared in service in September 1985. The paper describes the inspection systems themselves and discussed the results achieved to-date. (author)

  15. Advanced CANDU reactor, evolution and innovation

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed the ACRTM (Advanced CANDU(1) ReactorTM) to meet today's market challenges. It is a light water tube type pressurized water reactor and is the latest evolution of CANDU technology. The design was launched to be cost-competitive with other generating sources, while building on the unique safety and operational advantages of the CANDU design. The ACR is an evolutionary design that retains the proven CANDU features delivered at Qinshan Phase III, while incorporating a set of innovative features and proven state-of-the-art technologies that have emerged from AECL's ongoing Research, Development and Demonstration programs. This approach ensures that key design parameters are well supported by existing reactor experience and R and D. The result is a design that delivers a new threshold in safety, performance and economics while retaining ample design margin. AECL has developed the enabling technologies and components for the ACR design, and has applied them to two plant sizes, ACR-700 and ACR-1000. The ACR integrates hallmark characteristics of traditional CANDU plants (e.g. horizontal pressure tubes, on power fuelling, automated reactor control systems, and dual independent shutdown systems), new innovations (e.g. state-of-the- art control room, extensive use of modular construction techniques, smaller reactor core, enriched uranium fuel), and certain PWR features (e.g. light water coolant, negative void reactivity). The ACR is designed for a high capacity factor and low operation and maintenance costs. It fully exploits the construction techniques that contributed to the impressive schedule accomplishments at Qinshan Phase III and therefore features a very short construction schedule, 40 months construction schedule (First Concrete to Fuel Loading ) for the first unit with improvements to 36 months for later units. The ACR is a true Gen-III plus product with a broad application. It has been proven to be an ideal

  16. Resistance welding of tubes at low regidual pressure jn tube cavity

    International Nuclear Information System (INIS)

    The procedure of butt resistance welding of boilers in diameter of 32 mm at low residual pressure in tube cavities has been studied. It is shown that the creation of low residual pressure in tube cavity makes it possible to produce qualitative joints of tubes of the 20, 12Kh1MF, 12Kh18N12T steels. The maximum relative deformation in the butt zone should be in the range of 0.5...0.6

  17. Influence of aqueous environment pH on the corrosion behaviour of the CANDU steam generator tubing material

    International Nuclear Information System (INIS)

    The generalized corrosion is an undesirable process because it is accompanied by deposition of the corrosion products which affect the steam generator performances. It is very important to understand the generalized corrosion mechanism in order to evaluate the amounts of corrosion products which exist in the steam generator after a determined period of operation. The purpose of the experimental research consists in the assessment of corrosion behavior of the tube material (Incoloy-800) at normal secondary circuit parameters (temperature - 260 deg. C, pressure - 5.1 MPa). The testing environment was the demineralized water without impurities, at different pH values regulated with morpholine and cycloheyilamine (all volatile treatment). The results are presented as micrographs and graphics representing loss of metal by corrosion, corrosion rate, the total corrosion products, the adherent corrosion product, the released corrosion products and the release of the metal. (authors)

  18. Prandtl pressure tube for non-steady speed measurement

    International Nuclear Information System (INIS)

    The pressure tube consists of a long body favourable to flow with one opening on the front and one or more openings on the sides. The openings are connected to a differential pressure measurement device, a small differential pressure transducer, which is integrated between the front opening and side openings in the long body. Piezo-resistive pressure transducers are preferred. (orig./HP)

  19. Severe accident development modeling and evaluation for CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Negut, Gheorghe [National Agency for Radioactive Waste, 1, Campului Str., 115400 Mioveni (Romania)], E-mail: gheorghe.negut@andrad.ro; Catana, Alexandru [Institute for Nuclear Research Pitesti, 1, Campului Str., Mioveni P.O. Box 78, 0300 Pitesti (Romania); Prisecaru, Ilie; Dupleac, Daniel [Politehnica University Bucharest, 313, Splaiul Independentei, Sect. 6, 060042 Bucharest (Romania)

    2009-09-15

    Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents. Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

  20. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - II: DUPIC Fuel-Handling Cost

    International Nuclear Information System (INIS)

    The Direct Use of spent Pressurized water reactor fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel-handling technique has been investigated through a conceptual design study to estimate the unit cost that can be used for the DUPIC fuel cycle cost calculation. The conceptual design study has shown that fresh DUPIC fuel can be transferred to the core following the existing spent-fuel discharge route, provided that new fuel-handling equipment, such as the manipulator, opening/sealing tool of shipping casks, new fuel magazine, new fuel ram, dryer, gamma-ray detector, etc., are installed. The reverse path loading option is known to minimize the number of additional pieces of equipment for fuel handling, because it utilizes the existing spent-fuel handling equipment, and the discharge of spent DUPIC fuel can be done through the existing spent-fuel handling system without any modification. However, because the decay heat of spent DUPIC fuel is much higher than that of spent natural uranium fuel, the extra cooling capacity should be supplemented in the spent-fuel storage bay. Based on the conceptual design study, the capital cost for DUPIC fuel handling and extra storage cooling capacity was estimated to be $3 750 000 (as of December 1999) per CANDU plant. The levelized unit cost of DUPIC fuel handling was then obtained by considering the amount of fuel that will be required during the lifetime of a plant, which is 5.13 $/kg heavy metal. Compared with the other unit costs of the fuel cycle components, it is expected that DUPIC fuel handling has only a minor effect on the overall fuel cycle cost

  1. The performance of T-pad bearing pads, as a remedy against pressure tube crevice corrosion, on bundles irradiated at Bruce and Point Lepreau

    International Nuclear Information System (INIS)

    Crevice corrosion in CANDU reactors can occur between the standard design fuel bundle bearing pads and the pressure tube when the element operates at a sufficiently high power to create the crevice boiling condition necessary for the concentration of lithium hydroxide leading to enhanced oxidation of the bearing pad and pressure tube. Since crevice corrosion was discovered in Pickering pressure tubes, a concerted effort has been made on design changes to the standard bearing pads in order to minimize/elirninate crevice corrosion. This development program led to the T-Pad bearing pad design. Recent demonstration irradiations of prototype bundles, fitted with T-Pad bearing pads, were conducted in Bruce and Point Lepreau Nuclear Generating Stations. The subsequent post-irradiation examinations indicated, that except for increased hydrogen and deuterium pickup in the T-Pads, the performance of the T-Pads and bundles is consistent with standard bearing pad bundles. (author)

  2. Overview of blister phenomena in relation to pressure tube integrity

    International Nuclear Information System (INIS)

    Unstable pressure tube rupture in a pressure tube PHWR has potentially serious economic and safety consequences. Reactor operation under conditions which entail the risk of such failure should be avoided. Mostly, failure mechanisms in zirconium alloy pressure tubes involve hydrogen either in the initiation process or the propagation process. The slow propagation of such cracks by delayed hydride cracking (dhc) are well characterized experimentally and by post service observation allowing a reasonable prediction of the time for a crack to grow to an unstable length after leaking. Action can then be taken to shutdown the reactor before the crack grows to an unstable length. If there is evidence that the above sequence can be met, the design is commonly described as meeting a leak-before-break criterion. For one mode of failure, i.e. a crack originating from a hydride blister on the outside of the pressure tube, leak-before-break cannot be relied upon because the conditions in the pressure tube that led to the formation of the hydride blister, inhibit crack growth through the wall, and crack growth in the axial direction to a partial thickness unstable length is possible. The root cause of a ruptured pressure tube at Pickering unit 2 in August 1983 was the significantly displaced outlet end spacer in that channel exacerbated by the high rate of hydriding of the Zircaloy-2 pressure tube. It was imperative that the conditions which can cause blister induced failure be adequately quantified so that operation can continue with no risk of rupture. The Canadian nuclear industry has thus made a large investment in research and development related to hydrogen in pressure tubes to determine the criterion for initiation, growth and crack propagation from blisters, so as to ensure satisfactory and safe performance of the reactors. It is important to describe the essential features of blister behaviour as researchers and analysts in Canada understand them, and how the phenomenon

  3. The measurement of the deuterium concentration distributions in deuteride blisters on zirconium-alloy pressure tube material

    International Nuclear Information System (INIS)

    Deuterium concentrations in deuteride blisters on zirconium-alloy pressure tube material have been measured as a function of depth with a lateral resolution of a few tens of μm using a nuclear reaction technique with a ≅ 1300 keV 3He-ion probe. These measurements were carried out in air and results are presented for three types of blisters. The first was on a Zircaloy-2 pressure tube which had failed in a CANDU nuclear reactor. The second was formed in the laboratory on a Zr-2.5 wt% Nb pressure tube by producing cold spots on the outside of an internally heated tube (constant temperature at 3500C). Type 3 is similar to type 2 except that the internal temperature was cycled between 3500 and 1000C. A comparison is given with measurements carried out in vacuum, which show that except when deuterium concentrations are small - where count rates are low and surface concentrations are significant - the external-beam method provides a reliable analysis. This microbeam technique is at present the only procedure available with a lateral resolution better than 100 μm for measuring deuterium distributions which extend over several millimeters. The external beam is advantageous since radioactive specimens from reactors are much easier to handle outside a vacuum chamber. The measurement of an average deuteride composition of ZrD1.5 in the centre of a blister suggests that this region consists of mainly δ-phase deuterides. (orig.)

  4. The development of a remote gauging and inspection capability for fuel channels in Candu reactors

    International Nuclear Information System (INIS)

    Equipment under development for the inspection and gauging of pressure tubes in CANDU (Canadian Deuterium Uranium) type reactors is described. A brief overview of the mechanical scanning system is presented followed by a detailed description of the measurement and data processing systems for the gauging of diameter and wall thickness, volumetric inspection of the tube wall and gauging of the annular gap between the pressure tube and the calandria tube. Experience of testing ultrasonic transducers in very high (106 Roentgens/hour)(R/h) radiation fields is reviewed. (author)

  5. Development and Validation of Pressure Tube Deformation and Subcooled Boiling Models of the MARS Code for Safety Analyses of the Wolsong NPP's

    International Nuclear Information System (INIS)

    The MARS code is being considered by KINS(Korea Institute of Nuclear Safety) as a thermal-hydraulic regulatory auditing tool for nuclear power plants in South Korea. Because Korea currently has four operating units of the CANDU(Canadian Deuterium Uranium)-type reactor in Wolsong, analytic models such as the Wolsong pump model, the off-take model for arbitrary-angled branch pipes, the radiation heat transfer input model, and the subcooled boiling model have been implemented into the MARS code to extend its applicability into CANDU reactors as well as PWR's. This part of the research series presents verification and validation of the pressure tube deformation model and the Podowski subcooled boiling model

  6. Boussignac continuous positive airway pressure for weaning with tracheostomy tubes

    NARCIS (Netherlands)

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

    2008-01-01

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

  7. Thermal and hydraulic behaviour of CANDU cores under severe accident conditions

    International Nuclear Information System (INIS)

    This report describes work performed for the Atomic Energy Control Board on a) Formation and rewetting of dry patches on CANDU reactor calandria tubes during a Loss-of-Coolant Accident, and b) Analysis of accident sequence S11: Loss-of-Coolant Accident plus Loss-of-Emergency Core Cooling plus loss of moderator cooling system. For part (a), it is concluded that any dry patches which form on calandria tubes as a result of local heating to the critical heat flux will rewet in a short time (10 to 30 seconds for a Bruce-type reactor, 90 seconds for a Douglas Point-type reactor), with negligible effects on fuel sheath and maximum pressure tube temperatures. Pressure tube integrity is not predicted to be threatened. For part (b), preliminary analysis of the S11 accident sequence is presented. The complete analysis follows in the final report on the effects of severe accidents on CANDU cores

  8. Leak detection capability in CANDU reactors

    International Nuclear Information System (INIS)

    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

  9. 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.

  10. Comparison of the intracuff pressures of three different tracheostomy tubes.

    Science.gov (United States)

    Nishiyama, Tomoki

    2005-01-01

    The purpose of this study was to compare the cuff pressures of three tracheostomy tubes, MERA sofit CLEAR, Blue Line Tracheostomy Tube, and Tracheosoft. Each tracheostomy tube with an internal diameter of 7.0 mm was put into a plastic column. The cuff was then inflated with air to seal the column, and the column was filled with water. The air in the cuff was withdrawn gradually and the cuff pressure at the point of water leakage was measured. Six columns of different size were used. In columns with an internal diameter of 18-21 mm, the water leakage pressure was lower in the following order: MERA sofit CLEAR sofit CLEAR was found to maintain most safely the lowest intracuff pressure to seal the trachea among the three tracheostomy tubes tested. PMID:16032458

  11. Absorber materials in CANDU PHWRs

    International Nuclear Information System (INIS)

    In a CANDU reactor the fuel channels are arranged on a square lattice in a calandria filled with heavy water moderator. This arrangement allows five types of tubular neutron absorber devices to be located in the relatively benign environment of low pressure, low temperature heavy water between neighbouring rows or columns of fuel channels. This paper will describe the roles of the devices and outline the design requirements of the absorber component from a reactor physics viewpoint. Nuclear heating and activation problems associated with the different absorbers will be briefly discussed. The design and manufacture of the devices will be also discussed. The control rod absorbers and shut off materials are cadmium and stainless steel. In the tubular arrangement, the cadmium is sandwiched between stainless steel tubes. This type of device has functioned well, but there is now concern over the availability and expense of cadmium which is used in two types of CANDU control devices. There are also concerns about the toxicity of cadmium during the fabrication of the absorbers. These concerns are prompting AECL to study alternatives. To minimize design changes, pure boron-10 alloyed in stainless steel is a favoured option. Work is underway to confirm the suitability of the boron-loaded steel and identify other encapsulated absorber materials for practical application. Because the reactivity devices or their guide tubes span the calandria vessel, the long slender components must be sufficiently rigid to resist operational vibration and also be seismically stable. Some of these components are made of Zircaloy to minimize neutron absorption. Slow irradiation growth and creep can reduce the spring tension, and periodic adjustments to the springs are required. Experience with the control absorber devices has generally been good. In one instance liquid zone controllers had a problem of vibration induced fretting but a redesigned back-fit resolved the problem. (author). 3 refs, 8

  12. Consequences of pressure tube rupture on in-core components

    International Nuclear Information System (INIS)

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

  13. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    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

  14. 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

  15. candu fuel bundle fabrication

    International Nuclear Information System (INIS)

    This paper describes works on CANDU fuel bundle fabrication in the Fuel Fabrication Development and Testing Section (FFDT) of AECL's Chalk River Laboratories. This work does not cover fuel design, pellet manufacturing, Zircaloy material manufacturing, but cover the joining of appendages to sheath tube, endcap preparation and welding, UO2 loading, end plate preparation and welding, and all inspections required in these steps. Materials used in the fabrication of CANDU fuel bundle are: 1)Ceramic UO2 Pellet 2)Zircaloy -4. Fuel Bundle Structural Material 3) Others (Zinc stearate, Colloidal graphite, Beryllium and Heium). Th fabrication of fuel element consist of three process: 1)pellet loading into the sheats, 2) endcap welding, and 3) the element profiling. Endcap welds is tested by metallography and He leak test. The endcaps of the elements are welded to the end plates to form the 37- element bundle assembly

  16. Measurement of internal diameter of pressure tubes in pressurized heavy water reactors using ultrasonics

    International Nuclear Information System (INIS)

    The Pressure Tube in Pressurized Heavy Water Reactors (PHWRs) undergoes dimensional changes due to the effects of creep and growth as it is subjected to high pressure and temperature, which causes Pressure Tubes to permanently increase in length and diameter and to sag because of weight of fuel and coolant (heavy water) contained in it. These dimensional changes are due to prolonged stresses under high temperature and radiation. Pressure Tube stresses are evaluated for both beginning and end of life for accounting the Pressure Tube dimensional changes that occur during its design life. At the beginning of life, the initial wall thickness and un-irradiated material properties are applied. At the end of life, Pressure Tube diameter and length increases, while wall thickness decreases. Material strength also increases during that period. The increase in Pressure Tube diameter results in squeezing of garter spring spacer between the pressure and calandria Tubes. It also causes unacceptable heat removal from the fuel due to an increased amount of primary coolant that bypasses the fuel bundles. This reduces the critical channel power at constant flow. Hence the periodic monitoring of pressure Tube diameter is important for these reasons. This is also required as per the applicable codes and standards for In-Service Inspection of PHWRs. Mechanical measurement from ID of the Tube during periodic monitoring is not practically feasible due to high radiation and inaccessibility. This necessitates the development of NDT technique using Ultrasonics for periodic in-situ measurement of ID of pressure Tubes with a BARC made remotely operated drive system called BARCIS (BARC Channel Inspection system). The development of Ultrasonic based ID measurement techniques and their actual applications in PHWRs Pressure tubes are being discussed in this paper. (author)

  17. Fuel condition in Canadian CANDU 6 reactors

    International Nuclear Information System (INIS)

    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 UO2 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 discuss our

  18. Formation of hydride blisters in zirconium alloy pressure tubes

    International Nuclear Information System (INIS)

    The fracture of the Zircaloy-2 pressure tube in the Pickering Unit 2 power reactor was associated with the growth of hydride blisters at points of contact between the pressure tube and the cooler calandria tube surrounding it. Similar blisters have been observed in a Zr-2.5 wt% Nb pressure tube in WR-1, an organic-cooled research reactor. These hydride blisters were formed and grew as a result of the thermal diffusion of hydrogen in the zirconium, a mechanism whereby hydrogen diffuses down a temperature gradient. If the terminal solid solubility of hydrogen is exceeded in the cooler regions, hydride will precipitate. In this paper, the time required to grow these hydride blisters will be estimated from the blister size and the hydrogen distribution in its neighborhood, by using simple equations derived from thermal diffusion theory

  19. CANDU passive shutdown systems

    International Nuclear Information System (INIS)

    CANDU incorporates two diverse, passive shutdown systems (Shutdown System No. 1 and Shutdown System No. 2) which are independent of each other and from the reactor regulating system. Both shutdown systems function in the low pressure, low temperature, moderator which surrounds the fuel channels; the shutdown systems do not penetrate the heat transport system pressure boundary. The shutdown systems are functionally different, physically separate, and passive since the driving force for SDS1 is gravity and the driving force for SDS2 is stored energy. The physics of the reactor core itself ensures a degree of passive safety in that the relatively long prompt neutron generation time inherent in the design of CANDU reactors tend to retard power excursions and reduces the speed required for shutdown action, even for large postulated reactivity increases. All passive systems include a number of active components or initiators. Hence, an important aspect of passive systems is the inclusion of fail safe (activated by active component failure) operation. The mechanisms that achieve the fail safe action should be passive. Consequently the passive performance of the CANDU shutdown systems extends beyond their basic modes of operation to include fail safe operation based on natural phenomenon or stored energy. For example, loss of power to the SDS1 clutches results in the drop of the shutdown rods by gravity, loss of power or instrument air to the injection valves of SDS2 results in valve opening via spring action, and rigorous self checking of logic, data and timing by the shutdown systems computers assures a fail safe reactor trip through the collapse of a fluctuating magnetic field or the discharge of a capacitor. Event statistics from operating CANDU stations indicate a significant decrease in protection system faults that could lead to loss of production and elimination of protection system faults that could lead to loss of protection. This paper provides a comprehensive

  20. Candu technology: the next generation now

    International Nuclear Information System (INIS)

    We describe the development philosophy, direction and concepts that are being utilized by AECL to refine the CANDU reactor to meet the needs of current and future competitive energy markets. The technology development path for CANDU reactors is based on the optimization of the pressure tube concept. Because of the inherent modularity and flexibility of this basis for the core design, it is possible to provide a seamless and continuous evolution of the reactor design and performance. There is no need for a drastic shift in concept, in technology or in fuel. By continual refinement of the flow and materials conditions in the channels, the basic reactor can be thermally and operationally efficient, highly competitive and economic, and highly flexible in application. Thus, the design can build on the successful construction and operating experience of the existing plants, and no step changes in development direction are needed. This approach minimizes investor, operator and development risk but still provides technological, safety and performance advances. In today's world energy markets, major drivers for the technology development are: (a) reduced capital cost; (b) improved operation; (c) enhanced safety; and (d) fuel cycle flexibility. The drivers provide specific numerical targets. Meeting these drivers ensures that the concept meets and exceeds the customer economic, performance, safety and resource use goals and requirements, including the suitable national and international standards. This logical development of the CANDU concept leads naturally to the 'Next Generation' of CANDU reactors. The major features under development include an optimized lattice for SEU (slightly enriched uranium) fuel, light water cooling coupled with heavy water moderation, advanced fuel channels and CANFLEX fuel, optimization of plant performance, enhanced thermal and BOP (balance of plant) efficiency, and the adoption of layout and construction technology adapted from successful on

  1. Measurements of elastic modulus in Zr alloys for CANDU applications

    International Nuclear Information System (INIS)

    Measurements of elastic modulus as a function of temperature from 20 to 400°C were carried out on specimens of Zr-2.5Nb, Zircaloy-4, Zircaloy-2 and Excel Zr alloy using an ultrasonic resonance technique. The specimens were machined from CANDU pressure tubes, a calandria tube and commercial sheet material. Effects of crystallographic texture, neutron irradiation and hydrogen on elastic modulus were investigated. The results show that elastic modulus of the Zr alloys (1) decreases with increasing temperature, (2) depends strongly on crystallographic texture, and (3) increases slightly with neutron irradiation. (author)

  2. 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.

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

    International Nuclear Information System (INIS)

    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

  4. TEM examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

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

  5. Methodology Improvement of Reactor Physics Codes for CANDU Channels Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun; Choi, Geun Suk; Win, Naing; Aung, Tharndaing; Baek, Min Ho; Lim, Jae Yong [Kyunghee University, Seoul (Korea, Republic of)

    2010-04-15

    As the operational time increase, pressure tubes and calandria tubes in CANDU core encounter inevitably a geometrical deformation along the tube length. A pressure tube may be sagged downward within a calandria tube by creep from irradiation. This event can bring about a problem that is serious in integrity of pressure tube. A measurement of deflection state of in-service pressure tube is, therefore, very important for the safety of CANDU reactor. In this paper, evaluation of impacts on nuclear characteristic due to fuel channel deformation were aimed in order to improve nuclear design tools for concerning the local effects from abnormal deformations. It was known that sagged pressure tube can cause the eccentric configuration of fuel bundles in pressure tube by O.6cm maximum. In this case, adverse pin power distribution and reactivity balance can affect reactor safety under normal and accidental condition. Thermal and radiation-induced creep in pressure tube would expand a tube size. It was known that maximum expansion may be 5% in volume. In this case, more coolant make more moderation in the deformed channel resulting in the increase of reactivity. Sagging of pressure tube did not cause considerable change in K-inf values. However, expansion of the pressure tube made relatively large change in K-inf. Modeling of eccentric and enlarged configuration is not easy in preparation of input geometry at both HELlOS and MCNP. On the other hand, there is no way to consider this deformation in one-dimensional homogenization tool such as WIMS code. The way of handling this deformation was suggested as the correction method of expansion effect by adjusting the number density of coolant. The number density of heavy water coolant was set to be increased as the rate of expansion increase. This correction was done in the intact channel without changing geometry. It was found that this correction was very effective in the prediction of K-inf values. In this study, further

  6. X-ray diffraction residual stress measurement in the rolled-joint zone of Zr - 2.5 % Nb pressure tube

    International Nuclear Information System (INIS)

    The in-service experience of Zr - 2.5 % Nb pressure tubes in CANDU-type nuclear reactors has demonstrated very good performance over a long period of time. However, analyses done by AECL specialists on most failure cases, showed that a big percentage of defects are manufacturing defects, which appear mostly at the beginning of the rolled-joint zone. It has been observed that a correct rolling ensures an acceptable distribution of residual stress, but an incorrect one leads to an accumulation of big values of residual stress. This determines a preferential radial orientation of hydrides, which during operation in the reactor can produce DHC. To ensure a suitable performance of the Zr - 2.5 % Nb pressure tubes in the CANDU reactor, it is very important to have a correct rolling as mentioned in the procedure. This work presents a methodology for the measurement of the stressing state in the surfaces layers of the rolled-joint zone. The X-ray diffraction method can also be used for establishing the residual stress distribution across the tub wall, in order to ensure a good performance at Cernavoda nuclear plant. The results obtained for the investigated tube have led to the conclusion that the rolling process was correctly applied in this case, the values obtained for the residual stress being in good agreement with those accepted in literature. (Author) 2 Figs., 2 Tabs

  7. Deadly pressure pneumothorax after withdrawal of misplaced feeding tube

    DEFF Research Database (Denmark)

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

    2016-01-01

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

  8. Creep-rupture tests of internally pressurized Inconel 702 tubes

    Science.gov (United States)

    Gumto, K. H.

    1973-01-01

    Seamless Inconel 702 tubes with 0.375-in. outside diameter and 0.025-in. wall thickness were tested to failure at temperatures from 1390 to 1575 F and internal helium pressures from 700 to 1800 psi. Lifetimes ranged from 29 to 1561 hr. The creep-rupture strength of the tubes was about 70 percent lower than that of sheet specimens. Larson-Miller correlations and photomicrographs of some specimens are presented.

  9. Pressure drop of subcooled flow boiling in narrow tube

    International Nuclear Information System (INIS)

    The pressure drop of subcooled flow boiling in a narrow tube was investigated experimentally using water as a coolant. Experiments were conducted at nearly ambient pressure under the conditions: tube inside diameter: 1 and 3mm, tube length: 10∼100mm, and water mass velocity: 7000∼20000kg/m2s. The friction pressure drop ratio of subcooled flow boiling to non-heating water flow was examined by increasing the heat flux. The ratio begins to increase at the heat flux proposed by the Saha-Zuber correlation that the bubble begins to detach for 3mm inside diameter tube, though the heat flux is higher than the Saha-Zuber heat flux for 1mm tube. The ratio was further compared with the Bergles-Dormer correlation. The two phase friction multiplier of subcooled flow boiling was examined assuming the Ahmad void fraction and applied to the Lockhart-Martinelli (L-M) correlation. The abnormarity of the subcooled flow boiling in the case of 1mm inside diameter tube was confirmed in these discussions. (author)

  10. Size determinations, by ultrasonic techniques, of cracks in hydride blisters formed in Zr-2.5 % Nb pressure tubes

    International Nuclear Information System (INIS)

    Non destructive techniques (NDT) are very useful in the detection of flaws produced in structural components in service. During the service of CANDU nuclear power reactors, it is possible that pressure tubes (PT) may contact calandria tubes (CT). After the PT/CT contact, zirconium hydride blisters may form at the point of contact depending on the concentration of hydrogen/deuterium. Zirconium hydride is brittle and is therefore prone to cracking under stress. Ultrasonic NDT is routinely use during PT in service inspection. In order to be able of detecting cracked blisters, it is of great importance the development of standards to calibrate the employed equipment. On this purpose, hydride blisters were grown, in laboratory, on sections of pressure tube. The cracks in the blisters were detected and measured by ultrasonic techniques. The obtained results were compared with measurements carried out in optic microscope, on successive sections of the samples. The crack tip diffraction technique was found to be the more effective for the mentioned ends. (author)

  11. Nuclear Archeology for CANDU Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, Bryan L [ORNL

    2011-01-01

    The goal of this work is the development of so-called 'nuclear archeology' techniques to predict the irradiation history of both fuel-related and non-fuel-related materials irradiated in the CANDU (CANada Deuterium Uranium) family of nuclear reactors. In this application to CANDU-type reactors, two different scenarios for the collection of the appropriate data for use in these procedures will be assumed: the first scenario is the removal of the pressure tubes, calandria tubes, or fuel cladding and destructive analysis of the activation products contained in these structural materials; the second scenario is the nondestructive analysis (NDA) of the same hardware items via high-resolution gamma ray scans. There are obvious advantages and disadvantages for each approach; however, the NDA approach is the central focus of this work because of its simplicity and lack of invasiveness. The use of these techniques along with a previously developed inverse capability is expected to allow for the prediction of average flux levels and irradiation time, and the total fluence for samples where the values of selected isotopes can be measured.

  12. CANDU development

    International Nuclear Information System (INIS)

    Evolution of the 950 MW(e) CANDU reactor is summarized. The design was specifically aimed at the export market. Factors considered in the design were that 900-1000 MW is the maximum practical size for most countries; many countries have warmer condenser cooling water than Canada; the plant may be located on coastal sites; seismic requirements may be more stringent; and the requirements of international, as well as Canadian, standards must be satisfied. These considerations resulted in a 600-channel reactor capable of accepting condenser cooling water at 320C. To satisfy the requirement for a proven design, the 950 MW CANDU draws upon the basic features of the Bruce and Pickering plants which have demonstrated high capacity factors

  13. 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. PMID:24089964

  14. The enhanced CANDU 6 reactor - Generation III CANDU medium size global reactor

    International Nuclear Information System (INIS)

    Full text: The Enhanced CANDU 6TM (EC6TM) is a Generation III 700 class, heavy water moderated pressure tube reactor, designed to provide safe, reliable, nuclear power. The EC6TM has evolved from the proven CANDU 6 plants licensed and operating in five countries (four continents) with over 150 reactor years of safe operation around the world. In recent years. this global CANDU 6 fleet, with over 92% average gross capacity factor has ranked in the world's top performing reactors. The EC6 reactor builds on this success of the CANDU 6 fleet by using the operation, experience and project feedback to upgrade the design and construction techniques. A key objective of the EC6 has been to review and incorporate design improvements in the CANDU 6 to meet current safety standards. The key characteristics of the highly successful CANDU 6 reactor design include: - Powered by natural Uranium; - Ease of installation with modular, horizontal fuel channel core; - Separate low-temperature, low-pressure moderator providing inherently passive heat sinks; Reactor vault filled with light water surrounding the core; - Two independent safety shutdown systems; - On-power fuelling; - The CANDU 6 plant has a highly automated control system, with plant control computers that adjust and maintain the reactor power for plant stability (which is particularly beneficial in less developed power grids-where fluctuations occur regularly and capacities are limited). The major improvements incorporated in the EC6 design include, - More robust containment and increased passive features e.g., thicker walls, steel liner; - Enhanced severe accident management with additional emergency heat removal systems; - Improved shutdown performance for improved Large LOCA margins; - Upgraded fire protection systems to meet current Canadian and International standards; - Additional design features to improve environmental protection for workers and public- ALARA principle; - Automated and unitized back-up standby

  15. Development of Zr-2.5% Nb pressure tubes

    International Nuclear Information System (INIS)

    Initially Zr-2 was chosen for use as pressure tubes in PHWR type of reactors. During service, a proportion of the hydrogen produced by the corrosion reaction is absorbed by the pressure tubes. This produces embrittling effect, particularly under impact loading. Even greater neutron economy can be obtained if advantage is taken of the strength increase resulting from alloy additions. Of the available alloys, the one considered as having the best combination of mechanical and corrosion properties is zirconium-2.5% niobium alloy. This alloy has also superior creep resistance properties. With this alloy, improved tensile properties can be obtained by either cold working or heat treatment. As a result of increase in strength it is possible to reduce the wall thickness by about 18%. In view of the above, NPC(Nuclear Power Corporation) decided to use cold worked Zr-Nb pressure tubes from KAPP-II onwards. In order to produce these tubes, extensive trials were carried out to achieve the specified mechanical and metallurgical properties. In association with other working groups, parameters like extrusion ratio, intermediate cold work, temperature of intermediate annealing etc. were decided and frozen. With the established flow sheet, one reactor charge of Zr-Nb pressure tubes have already been delivered to KAPP-II and the work is in progress to meet the requirement of other reactors. The paper deals with the efforts which have gone into the development of Zr-Nb pressure tubes. The various quality checks exercised during processing right from the ingot stage are also described. (author). 3 figs

  16. Shock tubes: compressions in the low pressure chamber

    International Nuclear Information System (INIS)

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

  17. Propagation of atmospheric pressure fronts in long vacuum tubes

    International Nuclear Information System (INIS)

    This experimental work was undertaken during the development of a system using fast acting valves to protect the Intersecting Storage Rings (ISR) vacuum chamber at CERN against damage from the implosion of thin wall vacuum chambers. A 30 m cylindrical tube with a diameter of 130 mm and similar to that used on the ISR, was evacuated to 10-2 torr. Following the sudden entry of atmospheric pressure at one end the pressure versus time diagram was observed a several points along the tube. These diagrams show a characteristic 'staircase' function which permits the determination of the propagation velocity. There is an initial weak pressure front of a few torr, propagated at 950 m s-1, which presents little mechanical danger, even to delicate components such as ionisation gauges. After a formation time of 0.1 s, one or more large amplitude (several tens of torr) and potentially dangerous pressure fronts are propagated with a velocity of 770 m s-1

  18. Ballooning of pressure tubes - Construction of a test facility

    International Nuclear Information System (INIS)

    The test facility has been built to enable creep testing of reactor pressure tube specimens under conditions which represent those likely to be encountered in a reactor loss-of-coolant accident. The facility has been designed to be capable of specimen heating rates up to 30 K.s-1, temperatures up to 1200 C and internal pressurization up to 6 MPa with either argon or steam. Pressure tube temperature, strain rate, and pressure instrumentation have been provided for collection of data required for analysis of creep behaviour. The facility has been designed to be suitable for testing irradiated specimens in a hot cell. The report provides a detailed description of the test rig and results from two commissioning ballooning tests. (author). 2 refs., 1 tab., 4 figs

  19. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in PWRs. Canadian Deuterium Uranium (CANDU trademark) steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have resulted in a decrease in steam generator-related station unavailability of Canadian CANDU reactors. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development (R and D) work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for speciality tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service (FFS) guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. This paper will also show how recent advances in cleaning technology are integrated into a life management strategy. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New steam generator designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce-A/B, Pickering-A/B) and strategic plans to ensure that good future operation is ensured. (orig.)

  20. Pressure heat pumping in the orifice pulse-tube refrigerator

    International Nuclear Information System (INIS)

    The mechanism by which heat is pumped as a result of pressure changes in an orifice pulse-tube refrigerator (OPTR) is analyzed thermodynamically. The thermodynamic cycle considered consists of four steps: (1) the pressure is increased by a factor π1 due to motion of a piston in the heat exchanger at the warm end of the regenerator; (2) the pressure is decreased by a factor π2 due to leakage out of the orifice; (3) the pressure is further decreased due to motion of the piston back to its original position; (4) the pressure is increased to its value at the start of the cycle due to leakage through the orifice back into the pulse tube. The regenerator and the heat exchangers are taken to be perfect. The pressure is assumed to be uniform during the entire cycle. The temperature profiles of the gas in the pulse tube after each step are derived analytically. Knowledge of the temperature at which gas enters the cold heat exchanger during steps 3 and 4 provides the heat removed per cycle from this exchanger. Knowledge of the pressure as a function of piston position provides the work done per cycle by the piston. The pressure heat pumping mechanism considered is effective only in the presence of a regenerator. Detailed results are presented for the heat removed per cycle, for the coefficient of performance, and for the refrigeration efficiency as a function of the compression ratio π1 and the expansion ratio π2. Results are also given for the influence on performance of the ratio of specific heats. The results obtained are compared with corresponding results for the basic pulse-tube refrigerator (BPTR) operating by surface heat pumping

  1. CATASTROPHE FRACTURE OF THIN-WALL PRESSURE TUBES

    Institute of Scientific and Technical Information of China (English)

    魏德敏; 杨桂通

    2002-01-01

    Catastrophe theory was used to investigate the fracture behavior of thin-wall cylindrical tubes subjected to nternal explosive pressure. Based on the energy theory and catastrophe theory, a cusp catastrophe model for the fracture was established, and a critical condition associated with the model is given.

  2. Requirements for class 1C, 2C, and 3C pressure-retaining components and supports in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    This Standard applies to pressure-retaining components of CANDU nuclear power plants that have a code classification of Class 1C, 2C or 3C. These are pressure-retaining components where, because of the design concept, the rules of the ASME Boiler and Pressure Vessel Code do not exist, are not applicable, or are not sufficient. The Standard provides rules for the design, fabrication, installation, examination and inspection of these components and supports. It provides rules intended to ensure the pressure-retaining integrity of components, not the operability. It also provides rules for the support of fueling machines. The Standard applies only to new construction prior to the plant being declared in service

  3. A study of intermittent buoyancy induced flow phenomena in CANDU fuel channels

    International Nuclear Information System (INIS)

    The present work focuses on two-phase flow behavior called 'Intermittent Buoyancy Induced Flow' (IBIF) resulting from the loss of coolant circulation in a CANDU reactor core. The main objectives are to study steam bubble formation and migration through the pressure tube into the feeder tubes and headers, and to study the effect of pressure tube sagging on the two-phase flow behavior during IBIF. This paper will describe the experiments conducted with a set up simulating a pressure tube in a CANDU reactor. The test section was a 9.0-m long, 101.6-mm ID horizontal acrylic tube. Two vertical tubes simulating the feeder pipes in the reactor were attached at the end sections of the horizontal tube. Each vertical tube was connected to an open top cylindrical water tank simulating the header tank. Experiments were conducted using air and water at atmospheric pressure to qualitatively examine the IBIF phenomena. Air bubbles injected into the horizontal tube were observed to rapidly rise towards the top forming a continuous layer rather than flowing as discrete bubbles. The front of the air-water interface then moved horizontally towards the end sections at constant speeds and vented into the feeder tubes forming a slug flow. At low air injection rates an oscillating periodic behavior in the void fraction was observed in the two feeder tubes as the air vented alternately through each of these tubes. Small sagging of the pressure tube in the middle induced the injected air to move towards the end faster reducing the venting time. (author)

  4. Supporting CANDU operators-CANDU owners group

    International Nuclear Information System (INIS)

    The CANDU Owners Group (COG) was formed in 1984 by the Canadian CANDU owning utilities and Atomic Energy of Canada limited (AECL). Participation was subsequently extended to all CANDU owners world-wide. The mandate of the COG organization is to provide a framework for co-operation, mutual assistance and exchange of information for the successful support, development, operation, maintenance and economics of CANDU nuclear electric generating stations. To meet these objectives COG established co-operative programs in two areas: 1. Station Support. 2. Research and Development. In addition, joint projects are administered by COG on a case by case basis where CANDU owners can benefit from sharing of costs

  5. CANDU: Shortest path to advanced fuel cycles

    International Nuclear Information System (INIS)

    Full text: The global nuclear renaissance exhibiting itself in the form of new reactor build programs is rapidly gaining momentum. Many countries are seeking to expand the use of economical and carbon-free nuclear energy to meet growing electricity demand and manage global climate change challenges. Nuclear power construction programs that are being proposed in many countries will dramatically increase the demand on uranium resources. The projected life-long uranium consumption rates for these reactors will surpass confirmed uranium reserves. Therefore, securing sufficient uranium resources and taking corresponding measures to ensure the availability of long-term and stable fuel resources for these nuclear power plants is a fundamental requirement for business success. Increasing the utilization of existing uranium fuel resources and implementing the use of alternate fuels in CANDU reactors is an important element to meet this challenge. The CANDU heavy water reactor has unequalled flexibility for using a variety of fuels, such as Natural Uranium (NU), Low Enriched Uranium (LEU), Recycled Uranium (RU), Mixed Oxide (MOX), and thorium. This CANDU feature has not been used to date simply due to lack of commercial drivers. The capability is anchored around a versatile pressure tube design, simple fuel bundle, on-power refuelling, and high neutron economy of the CANDU concept. Atomic Energy of Canada Limited (AECL) has carried out theoretical and experimental investigations on various advanced fuel cycles, including thorium, over many years. Two fuels are selected as the subject of this paper: Natural Uranium Equivalent (NUE) and thorium. NUE fuel is developed by combining RU and depleted uranium (DU) in such a manner that the resulting NUE fuel is neutronically equivalent to NU fuel. RU is recovered from reprocessed light water reactor (LWR) fuel and has a nominal 235U concentration of approximately 0.9 wt%. This concentration is higher than NU used in CANDU reactors

  6. Temperature dependence of the anisotropic deformation of Zr-2.5%Nb pressure tube material during micro-indentation

    Science.gov (United States)

    Bose, B.; Klassen, R. J.

    2011-12-01

    The effect of temperature on the anisotropic plastic deformation of textured Zr-2.5%Nb pressure tube material was studied using micro-indentation tests performed in the axial, radial, and transverse directions of the tube over the temperature range from 25 to 400 °C. The ratio of the indentation stress in the transverse direction relative to that in the radial and axial directions was 1.29:1 and 1.26:1 at 25 °C but decreased to 1.22:1 and 1.05:1 at 400 °C. The average activation energy of the obstacles that limit the rate of indentation creep increases, from 0.72 to 1.33 eV, with increasing temperature from 25 to 300 °C and is independent of indentation direction. At temperature between 300 °C and 400 °C the measured activation energy is considerably reduced for indentation creep in the transverse direction relative to that of either the axial or radial directions. We conclude that, over this temperature range, the strength of the obstacles that limit the time-dependent dislocation glide on the pyramidal slip system changes relative to that on the prismatic slip system. These findings provide new data on the temperature dependence of the yield stress and creep rate, particularly in the radial direction, of Zr-2.5%Nb pressure tubes and shed new light on the effect of temperature on the operation of dislocation glide on the prismatic and pyramidal slip systems which ultimately determines the degree of mechanical anisotropy in the highly textured Zr-2.5Nb pressure tube material used in CANDU nuclear reactors.

  7. Enhancing the moderator effectiveness as a heat sink during loss of coolant accidents in CANDU-PHW reactors using glass-peened surfaces

    International Nuclear Information System (INIS)

    The horizontal fuel channel concept is a distinguishing feature of the CANDU-PHW reactor. Each fuel channel consists of a Zr-2.5Nb pressure tube and a Zircaloy-2 calandria tube, separated by a gas filled annulus. The calandria tube is surrounded by heavy-water moderator that also provides a backup heat sink for the reactor core. This heat sink (about 10 mm away from the hot pressure tube) ensures adequate cooling of fuel in the unlikely event of a loss-of-coolant accident (LOCA). One of the ways of enhancing the use of the moderator as a heat sink is to improve the heat-transfer characteristics between the calandria tube and the moderator. This enhancement can be achieved through surface modifications to the calandria tube which have been shown to increase the tube's critical heat flux (CHF) value. An increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels or increase the margin to dryout. A series of experiments was conducted to assess the benefits provided by glass-peening the outside surface of calandria tubes for postulated LOCA conditions. In particular, the ability to increase the tube's CHF, and thereby reduce moderator subcooling requirements was assessed. Results from the experiments confirm that glass-peening the outer surface of a tube increases its CHF value in pool boiling. This increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels by at least 5 deg. C. (author)

  8. A comparison of the passive oxide films formed on CANDU steam generator tubing alloy 600 and alloy 800

    International Nuclear Information System (INIS)

    Alloy 600 (A600) steam generator (SG) tubing has been shown to be susceptible to stress corrosion cracking (SCC). Alloy 800 (A800) was developed as a replacement, though it has shown susceptibility to corrosion under certain conditions. The properties of the passive oxide films on both alloys were extensively analyzed to determine why the performance of A800 is superior to that of A600. Surface analysis to determine oxide composition was performed using X-ray photoelectron spectroscopy (XPS) and Auger Electron spectroscopy (AES). Electrochemical measurements were made using anodic polarization and electrochemical impedance spectroscopy (EIS). The oxide films on A600 and A800 were shown to have different electrochemical and compositional properties. (author)

  9. Temperature effect of dynamic anisotropic elastic constants of Zr-2.5Nb pressure tube by resonant ultrasound spectroscopy

    International Nuclear Information System (INIS)

    Dynamic anisotropic elastic constants of CANDU Zr-2.5Nb pressure tube materials were determined by high temperature resonant ultrasound spectroscopy (RUS). The resonance frequencies were measured using a couple of alumina waveguides and wide-band ultrasonic transducers in a small furnace. The rectangular parallelepiped specimens were fabricated along with the longitudinal, radial and transverse direction of the pressure tube. The initial estimates for RUS were obtained from the orientation distribution function by X-ray pole figure and elastic stiffness of single crystal zirconium. A nine elastic stiffness tensor for orthotropic symmetry was determined in the range of room temperature ∼500 deg. C. As the temperature increases, the elastic constant tensor, cij gradually decreases. Higher elastic constants along the transverse direction compared to those along the longitudinal or radial direction are similar to the case of Young's modulus or shear modulus. A crossing of elastic constants along the longitudinal direction and radial direction was observed near 120-150 deg. C. This fact could correlate to the crossing characteristics of c44 and c66 of a zirconium single crystal in the temperature range

  10. Effect of Candu Fuel Bundle Modeling on Sever Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Dupleac, D.; Prisecaru, I. [Power Plant Engineering Faculty, Politehnica University, 313 Splaiul Independentei, 060042, sect. 6, Bucharest (Romania); Mladin, M. [Institute for Nuclear Research, Pitesti-Mioveni, 115400 (Romania)

    2009-06-15

    In a Candu 6 nuclear power reactor fuel bundles are located in horizontal Zircaloy pressure tubes through which the heavy-water coolant flows. Each pressure tube is surrounded by a concentric calandria tube. Outside the calandria tubes is the heavy-water moderator contained in the calandria itself. The moderator is maintained at a temperature of 70 deg. C by a separate cooling circuit. The moderator surrounding the calandria tubes provides a potential heat sink following a loss of core heat removal. The calandria vessel is in turn contained within a shield tank (or reactor vault), which provides biological shielding during normal operation and maintenance. It is a large concrete tank filled with ordinary water. During normal operation, about 0.4% of the core's thermal output is deposited in the shield tank and end shields, through heat transfer from the calandria structure and fission heating. In a severe accident scenario, the shield tank could provide an external calandria vessel cooling which can be maintained until the shield tank water level drops below the debris level. The Candu system design has specific features which are important to severe accidents progression and requires selective consideration of models, methods and techniques of severe accident evaluation. Moreover, it should be noted that the mechanistic models for severe accident in Candu system are largely less well validated and as the result the level of uncertainty remains high in many instances. Unlike the light water reactors, for which are several developed computer codes to analyze severe accidents, for Candu severe accidents analysis two codes were developed: MAAP4-Candu and ISAAC. However, both codes started by using MAAP4/PWR as reference code and implemented Candu 6 specific models. Thus, these two codes had many common features. Recently, a joint project involving Romanian nuclear organizations and coordinated by Politehnica University of Bucharest has been started. The purpose

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

    International Nuclear Information System (INIS)

    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.

  12. CANDU refurbishment - managing the life cycle

    International Nuclear Information System (INIS)

    All utilities that operate a nuclear power plant have an integrated plan for managing the condition of the plant systems, structures and components. With a sound plant life management program, after about 25 years of operation, replacement of certain reactor core components can give an additional 25 to 30 years of operation. This demonstrates the long-term economic strength of CANDU technology and justifies a long-term commitment to nuclear power. Indeed, replacement of pressure tubes and feeders with the most recent technology will also lead to increased capacity factors - due to reduced requirements for feeder inspections and repair, and eliminating the need for fuel channel spacer relocation which have caused additional and longer maintenance outages. Continuing the operation of CANDU units parallels the successful life extensions of reactors in other countries and provides the benefits of ongoing reliable operation, at an existing plant location, with the continued support of the host community. The key factors for successful, optimum management of the life cycle are: ongoing, effective plant life management programs; careful development of refurbishment scope, taking into account system condition assessments and a systematic safety review; and, a well-planned and well-executed retubing and refurbishment outage, where safety and risk management is paramount to ensure a successful project The paper will describe: the benefits of extended plant life; the outlook for refurbishment; the life management and refurbishment program; preparations for retubing of the reactor core; and, enhanced performance post-retubing. Given the potential magnitude of the program over the next 10 years, AECL will maintain a lead role providing overall support for retubing and plant Life Cycle Management programs and the CANDU Owners Group will provide a framework for collaboration among its Members. (author)

  13. Life Assurance Strategy for CANDU NPP

    International Nuclear Information System (INIS)

    include design provisions to replace fuel channels and steam generators. Difficult to replace components such as reactor building structures and calandria/shield tank assembly are designed for much beyond 40 years. Given the performance of CND's to date and the successfully completed rehabilitations and the lessons learned from older plants, a newly committed CANDU will have an economic service life significantly longer than 40 years. The CANDU design life was initially set at thirty years. The key components of a CANDU nuclear steam plant are the calandria vessel, the fuel channels, the reactivity control mechanisms, and the primary heat transport components including piping and steam generators. The calandria vessel, a large stainless steel tank, experiences conditions of relatively low temperature and pressure and is designed for a very long life. Experience to date shows that of the remaining components, fuel channels and reactivity control mechanisms are replaceable. Given that other refurbishments and/or replacements can be done to existing plants, a minimum of 40 year operating life can be achieved. Large scale fuel channel replacement was dictated by Station Life Assurance rather than Life Extension considerations. This major rehabilitation program has been successfully implemented for three of the Pickering A reactors to achieve a minimum 40 year operating life. In this program steady flow of successful design and process improvements have contributed to the knowledge base and know how of the CANDU industry. Over the next few years, retuning of the fourth Pickering A unit and the first of the Bruce A units will be undertaken providing the opportunity for Life extension of these units. Steam Generators in most CANDU plants continue to perform, with relatively low tube failures and plugging rates. Remedial measures are being taken, with solutions being evaluated by Ontario Hydro to address current degradation problems due to tube fouling and sludge deposition. R

  14. 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

  15. Hydride blister formation simulation in Candu type reactors

    International Nuclear Information System (INIS)

    We have developed a computer code for the probability study of hydride blister formation in pressure tubes named BLIFO. The basic hypothesis of the model are: the pressure tube is divided into five areas according to the existence of four garter springs. For each area the probability of blister formation is the probability of the hydrogen content exceeding a critical threshold when contact tube is present; the probability of a blister in a tube is the OR combination of the probabilities of a blister in each area; the tube contact is a function of the garter springs location, and the time; the critical hydrogen threshold is sorted over the areas within the pressure tube; hydrogen pick-up rate was sorted with a Gaussian distribution; the initial hydrogen content values for each tube were measured before the ensamble and they are used in the code. For Embalse evaluation, we build up a subroutine that simulate Gaussian distribution using the parameters of a typical nuclear power Candu reactor garter spring distribution. (author)

  16. CANDU fuel behaviour under transient conditions

    International Nuclear Information System (INIS)

    The Canadian R and D program to understand CANDU fuel behaviour under transient conditions is described. Fuel sheath behaviour studies have led to the development of a model of transient plastic strain in inert gas, which integrates the deformation due to several mechanisms. Verification tests demonstrated that on average the model overpredicts strain by 20%. From oxidation kinetics studies a sheath failure embrittlement criterion based on oxygen distribution has been developed. We have also established a rate equation for high-temperature stress-dependent crack formation due to embrittlement of the sheath by beryllium. An electric, simulated fuel element is being used in laboratory tests to characterize the behaviour of fuel in the horizontal. In-reactor, post-dryout tests have been done for several years. There is an axially-segmented, axisymmetric fuel element model in place and a fully two-dimensional code is under development. Laboratory testing of bundles, in its early stages, deals with the effects of geometric distortion and sheath-to-sheath interaction. In-reactor, post-dryout tests of CANDU fuel bundles with extensive central UO2 melting did not result in fuel fragmentation nor damage to the pressure tube. (author)

  17. Some design features of SCW pressure-tube nuclear reactor

    International Nuclear Information System (INIS)

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33-35% to about 43-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625degC), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems. (author)

  18. Detection of blister formation and evaluation of pressure tube/calandria tube contact location by ultrasonic velocity ratio measurement technique

    International Nuclear Information System (INIS)

    Presence of hydrogen in zircaloy pressure tube affects the velocity of ultrasound propagation. Both longitudinal wave velocity (VL) and shear wave velocity (VS) are affected depending on the concentration of hydrogen. Velocity ratio (VL/VS) changes as per the concentrations of hydrogen in different locations along the length of pressure tube. A hydride blister which forms at the pressure tube and calandria tube contact point is a distinct zone containing hydrogen 2-3 order of magnitude more than the parent matrix and hence, can be detected by sharp change in velocity ratio. (author)

  19. ASSERT/NUCIRC commissioning for CANDU 6 fuel channel CCP analysis

    International Nuclear Information System (INIS)

    CANDU PHWR fuel channel pressure tubes will expand or creep under long-term (aging process) influence of temperature, pressure, and neutron flux. This diametral pressure tube creep will influence the critical channel power (CCP), or conditions that lead to dryout. In order to provide safety analysis models to quantify the effect of diametral pressure tube creep on CCP, a COG (AECL/NBP/HQ) project is underway to commission the ASSERT and NUCIRC codes to establish reliable production tools for the assessment of CANDU6 CCP in nominal (uncrept) and crept pressure tube fuel channels. This paper gives an overview of the background and objectives of the project along with a brief introduction into the subchannel analysis code ASSERT and the 1-D thermalhydraulics code NUCIRC. This project is a multistage endeavour, for which the first stage results are presented. A detailed cross-comparison of the 1-D (NUCIRC) and subchannel (ASSERT) models of pressure drop (ΔP) and critical heat flux (CHF) has been undertaken and has led to several enhancements and refinements to the respective models. These results are presented in addition to results of ASSERT commissioning against NUCIRC for a matrix of ΔP and dryout cases in a nominal pressure tube, which are based upon Gentilly 2 and Point Lepreau site area. Additionally, the initial results of an assessment, using ASSERT, of the effects of creep on ΔP are presented. In concluding, the status and future directions for ASSERT/NUCIRC CANDU 6 CCP analysis project are summarized. (author). 2 refs., 12 figs

  20. Development of pressure tube inspection equipment for the Fugen (ATR)

    International Nuclear Information System (INIS)

    A remote-controlled in-service inspection device has been developed for inspecting the pressure tubes of the Fugen, which is a heavy-water-moderated, boiling light-water-cooled pressure-tube-type reactor. The equipment is capable of performing three kinds of inspection: ultrasonic flaw detection, measurement of inside diameter and visual inspection of the internal surface. To reduce the radiation exposure of inspectors, the three kinds of detectors, with their associated electronics and drive mechanisms for vertical and rotating movements, are housed in the inspection tool assembly, which can be mounted on or removed from the pressure tubes by remote control using a refuelling machine. The ultrasonic technique has been adopted for measurement of the internal diameter in order to shorten the inspection time. The detectors, TV camera and electronic components used in the inspection tool assembly were selected on the basis of irradiation test results. Before inspection of the Fugen reactor, the total system was tested on a mock-up pressure tube to confirm its functions, performance, durability and reliability. The test results were: (1) the ultrasonic flaw detector can detect an artificial flaw of 2.0 mm in length, 0.1 mm in width and 0.1 mm in depth with S/N=7 dB; (2) the inside diameter measurement system can measure the inside diameter, ranging from 117.5 to 119.5 mm, with an accuracy of +-20 μm; (3) an artificial flaw of 2.0 mm in length, 0.1 mm in width and 0.1 mm in depth can be observed by the internal surface observation system. The equipment was used for the inspection of ten pressure tubes of the Fugen reactor during the May 1984 annual inspection. No degradation of the performance of the equipment was observed even after 55 hours of inspection under a maximum dose rate of 2.5x105 R/h. Based on these results, the functions and performance of the equipment in practical use were fully confirmed. (author)

  1. Sectional replacement of high pressure feedwater heater tubing

    Energy Technology Data Exchange (ETDEWEB)

    Bolton, J.A.; Bowes, P.D. [TransAlta Utilities Corp., Duffield, Alberta (Canada). Plant Engineering Services

    1994-12-31

    TransAlta Utilities is a Canadian Corporation which owns and operates the coal fired Sundance Generating Station located in central Alberta. Sundance is fitted with vertical channel down, carbon steel tubed, high pressure feedwater heaters. The primary mode of failure of these HP feedwater heaters on the six generating units is steam inlet area tube erosion and vibration damage. This damage is initiated with the deterioration of the desuperheating inlet shroud and backing plate, primarily due to thermal fatigue, thus allowing direct impingement of high velocity steam and entrained condensate upon the tubing. Topics discussed are: review of the design and conditions of the heater which allowed re-conditioning; cutting, lifting and supporting of the shell at an elevation sufficient to allow free access of the entire desuperheating zone; damage observed within the desuperheating and drains cooler zones; bundle reconditioning through damage tube section replacement and support plate repair techniques; design/installation of the desuperheating, drains-cooling zone shrouds and backing plates; benefits that this type of approach may offer; conclusions.

  2. The CANDU 9 distributed control system design process

    International Nuclear Information System (INIS)

    Canadian designed CANDU pressurized heavy water nuclear reactors have been world leaders in electrical power generation. The CANDU 9 project is AECL's next reactor design. Plant control for the CANDU 9 station design is performed by a distributed control system (DCS) as compared to centralized control computers, analog control devices and relay logic used in previous CANDU designs. The selection of a DCS as the platform to perform the process control functions and most of the data acquisition of the plant, is consistent with the evolutionary nature of the CANDU technology. The control strategies for the DCS control programs are based on previous CANDU designs but are implemented on a new hardware platform taking advantage of advances in computer technology. This paper describes the design process for developing the CANDU 9 DCS. Various design activities, prototyping and analyses have been undertaken in order to ensure a safe, functional, and cost-effective design. (author)

  3. A statistical method for draft tube pressure pulsation analysis

    International Nuclear Information System (INIS)

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

  4. A statistical method for draft tube pressure pulsation analysis

    Science.gov (United States)

    Doerfler, P. K.; Ruchonnet, N.

    2012-11-01

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

  5. 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.

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

    International Nuclear Information System (INIS)

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

  7. Resistance of metallic uranium tubes to external pressure

    International Nuclear Information System (INIS)

    After giving certain indications obtained from a simplified theoretical analysis of the the deformation to be expected in a closed tube subjected to an exterior pressure, the present report presents the out-of-pile experimental results obtain on fuel elements based on metallic uranium; Experiments carried out on tubes und on the end-pellets which enclose them have made it possible to show the importance of the parameters: time, temperature, thermal cycling, geometry, nature of the alloy, radial thermal gradient; they have also made it possible to determine the shapes of projected fuel elements. Systematic tests are now being carried out both in-pile and out-of-pile with a view to define tubular elements having any shape whatever. (authors)

  8. The Enhanced CANDU 6TM Reactor - Generation III CANDU Medium Size Global Reactor

    International Nuclear Information System (INIS)

    The Enhanced CANDU 6TM (EC6TM) is a 740 MWe class heavy water moderated pressure tube reactor, designed to provide safe, reliable, nuclear power. The EC6TM has evolved from the proven eleven (11) CANDU 6 plants licensed and operating in five countries (four continents) with over 150 reactor years of safe operation around the world. In recent years, this global CANDU 6 fleet has ranked in the world's top performing reactors. The EC6 reactor builds on this success of the CANDU 6 fleet by using the operation, experience and project feedback to upgrade the design and incorporate design improvements to meet current safety standards.The key characteristics of the highly successful CANDU 6 reactor design include: Powered by natural Uranium; Ease of installation with modular, horizontal fuel channel core; Separate low-temperature, low-pressure moderator providing inherently passive heat sinks; Reactor vault filled with light water surrounding the core; Two independent safety shutdown systems; On-power fuelling; The CANDU 6 plant has a highly automated control system, with plant control computers that adjust and maintain the reactor power for plant stability (which is particularly beneficial in less developed power grids-where fluctuations occur regularly and capacities are limited). The major improvements incorporated in the EC6 design include: More robust containment and increased passive features e.g., thicker walls, steel liner; Enhanced severe accident management with additional emergency heat removal systems; Improved shutdown performance for improved Large LOCA margins; Upgraded fire protection systems to meet current Canadian and International standards; Additional design features to improve environmental protection for workers and public-ALARA principle; Automated and unitized back-up standby power and water systems; Other improvements to meet higher safety goals consistent with Canadian and International standards based on PSA studies; Additional reactor trip

  9. CANDU 6 - the highly successful medium sized reactor

    International Nuclear Information System (INIS)

    The CANDU 6 Pressurized Heavy Water Reactor system, featuring horizontal fuel channels and heavy water moderator continues to evolve, supported by AECL's strong commitment to comprehensive R and D programs. The initial CANDU 6 design started in the 1970's. The first plants went into service in 1983, and the latest version of the plant is under construction in China. With each plant the technology has evolved giving the dual advantages of proveness and modern technology. CANDU 6 delivers important advantages of the CANDU system with benefit to small and medium-sized grids. This technology has been successfully adopted by, and localized to varying extents in, each of the CANDU 6 markets. For example, all CANDU owners obtain their fuel from domestic suppliers. Progressive CANDU development continues at AECL to enhance this medium size product CANDU 6. There are three key CANDU development strategic thrusts: improved economics, fuel cycle flexibility, and enhanced safety. The CANDU 6 product is also enhanced by incorporating improvements and advanced features that will be arising from our CANDU Technology R and D programs in areas such as heavy water and tritium, control and instrumentation, fuel and fuel cycles, systems and equipment and safety and constructability. (author)

  10. CANDU nuclear power system

    International Nuclear Information System (INIS)

    This report provides a summary of the components that make up a CANDU reactor. Major emphasis is placed on the CANDU 600 MW(e) design. The reasons for CANDU's performance and the inherent safety of the system are also discussed

  11. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

  12. Criticality computations for CANDU - 6 cell utilizing the SCALE system

    International Nuclear Information System (INIS)

    The investigated CANDU - 6 cell was defined in the frame of a benchmark approach; the fuel pellets were of sintered UO2 (with natural uranium), the coolant and moderator are D2O of different purities, while the fuel element can, the pressure tube and the calandria are made of different zirconium based alloys. Utilising the modular system of SCALE codes for criticality computation in CANDU reactor is challenging from both the computation method (Monte Carlo) used and the nuclear data used, which are not specific to the CANDU reactor. For criticality computation the SAS1 (NITAWL, BONAMI, and KENO - VI modules) sequence of the SCALE system. The geometrical model used was is the one indicated in benchmark approach corresponding to the Final Safety Report for Cernavoda NPP Unit 1. Separate calculation with three nuclear data libraries were performed in order to evidence the influence of energy cutting (27, 44 and 238 groups, respectively) and of the nuclear data provenance. Sensitivity calculations aimed at investigating the influence of small variations in geometrical parameters and composition on the effective multiplication factor. Also, different variants of treating the leaks, ranging from the completely reflected cell up to the infinite lattice were studied

  13. Explaining the absence of Co-58 radiation fields around CANDU reactor primary circuit

    International Nuclear Information System (INIS)

    Radiation fields from Co-58 are rarely detected in CANDU plants. For example, Ge(Li) surveys of the Inconel 600 steam generators at some CANDU plants may show radiation attributed to Co-58 only early in plant life, and most artefacts removed from the primary circuit later in plant operation show no Co-58 present. However, Pressurized Water Reactor plants experience relatively large fields from Co-58 on their isothermal piping, e.g., steam generator channel head, and steam generators tube sampling programs do show deposits in the tubes with significant Co-58 compared to other radionuclides such as Co-60. CANDU reactors have high concentrations of dissolved iron due to the extensive use of carbon steel for the isothermal piping, e.g., feeders, headers, and steam generator channel heads. A dissolved iron transport diagram that was proposed recently for the primary circuit of CANDU plants has been validated by comparison of predicted deposit weights with plant deposit data from various components. One feature of the diagram is dissolved iron precipitation inside the steam generators tubes. An hypothesis is advanced here in which precipitating dissolved iron is proposed to occlude dissolved nickel. This removal mechanism may prevent the solubility of dissolved nickel from being exceeded anywhere around the primary circuit. In particular, this mechanism could avoid NiO precipitation in the core and the generation of large quantities of Co-58. Using this mechanism along with the known solubility behaviour of NiO with temperature, a dissolved nickel transport diagram has been proposed for CANDU plants. (authors)

  14. Remote dismantling of the pressure tube reactor from NPP Niederaichbach

    International Nuclear Information System (INIS)

    The pressure tube reactor of NPP Niederaichbach will be dismantled, segmented and packaged by remote operation, using a rotary manipulator, a cutting manipulator and a crane manipulator. With help from a number of remote controlled tools the rotary manipulator disassembles and lifts the reactor parts to a hot cell installed upon the upper reactor floor. Handling, crushing and packaging of those parts is performed with help from the crane manipulator. The cutting manipulator serves for segmenting of the moderator tank and the neutron shield tank

  15. Numerical study of the thermo-hydraulic behavior for the Candu type fuel channel

    International Nuclear Information System (INIS)

    Candu type reactors use fuel channel in a horizontal lattice. The fuel bundles are positioned in two Zircaloy tubes: the pressure tube surrounded by calandria tube. Inside the pressure tube the coolant heavy water flows. The coolant reaches high temperatures and pressures. Due to irregular neutron spatial distribution, the fuel channel stress differs from one channel to other. In one improbable event of severe accident, the fuel channel behaves differently according to its normal function history. Over the years, there have been many research projects trying to analyze thermal hydraulic performance of the design and to add some operational improvements in order to achieve an efficient thermal hydraulic distribution. This paper discusses the thermo hydraulic behavior (influence of the temperature and velocity distribution) of the most solicited channel, simulated with Fluent 6.X. Code. Moreover it will be commented the results obtained using different models and mesh applied. (authors)

  16. Intergranular susceptibility in failures of high pressure tubes

    International Nuclear Information System (INIS)

    This work addresses the influence of metallurgical susceptibility to intergranular cracking on the repeated cracking and failure of thick wall curved steel tubes from a petrochemical reactor. These tubes are made of HP-4 steel, bent and heat treated, and then subjected to autofrettage. Internal pressure is around 250 MPa. All failures are characterized by strongly branched, mostly circumferential multiple intergranular cracks. Most cracks initiated in the outer surface, in contact with steam; these were related to stress corrosion cracking (SCC). Some cracks initiated in pre defects in the inner surface, in contact with a polymer, and in the mid thickness of the tube wall. This study includes the assessment of deformation and temperature induced embrittlement mechanisms, measurement of longitudinal residual stresses, and mechanical testing included tensile, Charpy and SCC tests. Susceptibility to intergranular cracking was experimentally assessed by recreating conditions of embrittlement by thermal treatments and tensile testing. Samples with 0, 3 and 5% plastic deformations were subjected to 24 h thermal treatments between 300 and 400 deg. C. Under the conditions of previous plastic deformation due to bending and autofrettage it was possible to recreate intergranular embrittlement at service temperatures, a phenomenon similar to temper embrittlement. The process of forming the bent created localized yielding and large longitudinal residual stresses. Recovery measures, mostly relying on thermal treatments, were defined

  17. 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.

  18. CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

    This report is based on informal lectures and presentations made on CANDU Advanced Fuel Cycles over the past year or so, and discusses the future role of CANDU in the changing environment for the Canadian and international nuclear power industry. The changing perspectives of the past decade lead to the conclusion that a significant future market for a CANDU advanced thermal reactor will exist for many decades. Such a reactor could operate in a stand-alone strategy or integrate with a mixed CANDU-LWR or CANDU-FBR strategy. The consistent design focus of CANDU on enhanced efficiency of resource utilization combined with a simple technology to achieve economic targets, will provide sufficient flexibility to maintain CANDU as a viable power producer for both the medium- and long-term future

  19. Plant life management strategies for pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Ahn, Sang Bok; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This technical report reviewed aging mechanism of the major components of CANDU 6 reactor such as pressure tubes, calandria tube, end fitting, fuel channel spacer and calandria. Furthermore, the surveillance methodology was described for monitoring and inspection of these core components. Based on the in-reactor performances data such as delayed hydride cracking, leak-before-break, enhanced deformation-creep and growth, the life management of pressure tubes was illustrated in this report. (author). 19 refs., 11 figs., 2 tabs.

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

    International Nuclear Information System (INIS)

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

  1. Long-term passive CANDU containment response after a design-basis accident

    International Nuclear Information System (INIS)

    A passive CANDU reg-sign containment system, currently being developed, is aimed at limiting the consequences of a postulated accident, by ensuring the structural integrity of the containment building and limiting fission-product release to within siting dose limits, without operator action or reliance on ac power for up to 3 d. All main functions of the containment system, i.e. energy removal, hydrogen mitigation, and fission-product retention, are to be accomplished passively. The passive CANDU containment relies on the passive emergency water system (PEWS) for energy removal after an accident and on passive autocatalytic recombiners (PAR) for hydrogen removal. The key feature of this concept, is a recirculating, buoyancy-driven flow through the recombiners and the tube banks of the PEWS. This paper presents preliminary design calculations for the PEWS tank and tube banks and a simulation of the long-term passive containment response, based on the current CANDU-6 containment, to a large loss-of-coolant/loss-of- emergency coolant injection (LOCA/LOECI) using the GOTHIC code. It is shown that a 1500-M3 PEWS tank, connected to tube banks with a total surface area of 1800 m2, can limit the second pressure peak to about 300 kPa(a) if a recirculating flow is established in the containment building. The PEWS tank water is boiling in the long term, and the peak containment temperature is 114 degrees C. 6 refs., 4 figs

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

    International Nuclear Information System (INIS)

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

  3. Candu fuel and fuel cycles

    International Nuclear Information System (INIS)

    reactor designs, allowing operation today on currently available fuels and switching to other fueling options as market conditions change. This establishes an important freedom from future resource constraints without depending on future commercialization of challenging and expensive technologies such as fast breeder reactors, yet, once these are commercially available, CANDU and fast breeder fuel cycles are complementary and can achieve a highly advantageous synergism. This paper examines the fuel cycle option which CANDU reactor technology can accommodate, including the use of slightly enriched uranium direct use of spent pressurized water reactor fuel in CANDU (dupic), burning recovered uranium, mixed plutonium and uranium oxides or actinides and the use of thorium based fuel cycles. These options provide CANDU reactors with the most flexible fuelling of any reactor type, which are readily adaptable to meeting future variations in energy markets, regardless of what these may be. (author)

  4. Pressurized drift tubes scintillating fiber hadron calorimetry. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bromberg, C.; Huston, J.; Miller, R. [and others

    1995-03-22

    Under this contract members of the MSU high energy physics group constructed a full-scale Pressurized Drift Tube Chamber intended for the GEM muon system at the SSC. They achieved a position resolution of <90 {mu} over the full 5 m{sup 2} area of the detector. This resolution satisfied the GEM resolution requirements of <100 {mu} by a comfortable margin. Based on their SSC work they developed a new technique for creating wire supports in drift tubes with an overall placement accuracy of <20 {mu}. This technique requires only simple jigging and can be duplicated and operated at low cost. Also, they participated in the design and testing of a hadron calorimeter prototype for GEM. This work lead the authors to develop a semi-automatic welding machine to fuse together two plastic optical fibers. Copies of this machine are currently in use in the CDF endplug upgrade at Fermilab and additional copies are used widely in calorimeter and fiber-tracker construction.

  5. Heavy-water-moderated pressure-tube reactor safety

    International Nuclear Information System (INIS)

    Several countries have heavy-water-moderated, pressure-tube reactors either in commercial operation or in late prototype stages. The supporting safety research and development includes such areas as the thermohydraulics of circuit depressurization, heat transfer from the fuel, heat rejection to the moderator from dry fuel, fuel and sheath behaviour, and fuel channel integrity. We review the work done in Canada, Great Britain, Italy and Japan, and describe some of the experimental tests underlaying the methods of accident analysis. The reactors have safety systems which, in the event of an accident, are able to shut down the reactor, keep the fuel cooled, and contain any released radioactivity. We summarize the characteristics of these safety systems (shutdown, emergency core cooling, and containment) in the various reactors, and discuss other reactor characteristics which either prevent accidents or reduce their potential demand on the safety systems. (author)

  6. Pressure loads and structural response of the BNL high-temperature detonation tube

    OpenAIRE

    Shepherd, Joseph E.

    1992-01-01

    The high-temperature detonation tube facility being designed at Brookhaven National Laboratory must withstand dynamic pressure loads. These loads are associated with both detonations and deflagration-to-detonation transition (DDT). The present report documents the results of computations of the pressure loads and structural response. Structural response considerations indicate that radial motion of the tube is sufficiently rapid that the tube actualkly responds to the peak pressure behi...

  7. Application of acoustic emission to the testing pressure tubing materials

    International Nuclear Information System (INIS)

    Acoustic emission is one of the promising techniques for the detection of embrittlement. The Zr-2.5 Nb alloy used as pressure tubing material shows slightly low stress intensity factor when it absorbs hydrogen. In this paper, the relationship between acoustic emission count N and stress intensity factor K in the tensile test of edge-notched specimens is described. The K value is proportional to the square root of crack opening displacement phi in the elastic region. The double-notched specimens were cut from pressure tubes, and the single-notched specimens were cut from extruded bars. The crack opening displacement was measured with a clip gauge recommended by ASTM STP 410 Appendix, and the acoustic emission was measured with a Nortec AEMS-4 system and PZT-5 type sensors. The sensors were bonded on the surfaces of the specimens with epoxy adhesive or rubber contact. A peak of the acoustic emission count rate was observed at the yield point of each specimen similarly to many other metals. The N values and the size of plastic zone showed the theoretical relationship following 4th power law, on the other hand, the size of plastic zone depended linearly on the K values in elastic region. The slope and the intersection point of the regression curves for the total N count vs the square root of phi corresponding to the elastic field of stress-strain curves showed almost same values for the specimens of same shape. The influence of the bonding methods was not observed. (Kako, I.)

  8. Modelling and simulation of dynamic characteristics of CANDU-SCWR

    International Nuclear Information System (INIS)

    Owing to the thermal properties of supercritical water and features of heat transfer correlation under supercritical pressure, a detailed thermal-hydraulic model with movable boundary of is developed for CANDU-SCWR (Supercritical Water-Cooled Reactor). Steady-state results of the model agree well with the design data. The dynamic responses of CANDU-SCWR to different disturbances are simulated and characteristics are analyzed. A dynamic model for ACR is also developed using CATHENA. Differences between dynamic characteristics of CANDU-SCWR and those of ACR are highlighted and investigated. It is concluded that CANDU-SCWR has a larger time constant, but with a higher response amplitude. (author)

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

    DEFF Research Database (Denmark)

    Gaihede, Michael; Jacobsen, Henrik; Tveterås, Kjell;

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

  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

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

  11. Measurement and utility of fracture toughness properties of irradiated pressure tube from the ring tension test

    International Nuclear Information System (INIS)

    Results of fracture toughness computed from the transverse tensile properties of reactor operated Zircaloy pressure tube J-07 from Unit 1 of Madras Atomic Power Station (MAPS # I) are summarized. The pressure tube had experienced 9.5 effective full power years (EFPY) of reactor operation and had hydrogen equivalent (Heq) content upto 200 ppm. Fracture toughness and critical crack lengths have been evaluated from tube sections from inlet end, outlet end and mid length of the irradiated pressure tube. The pressure-temperature operating envelopes obtained using the variation of fracture toughness are presented. (author)

  12. CANDU-3: Features of next generation CANDU

    International Nuclear Information System (INIS)

    CANDU 3, with a net electrical output of 450 MW, is the latest and smallest version of the CANDU power system. Significant innovation built on proven CANDU reactor technology is the basis of the CANDU 3 design. This, coupled with the commitment to reduce plant cost, increase performance capacity factor, enhance safety features and incorporate technological improvements, makes CANDU 3 an advanced, world class product. This paper describes the following CANDU 3 features: A station layout to provide a flexible construction sequence, good system separation and ease of maintenance and operation. An up-front engineering and licensing process prior to beginning construction. Enhancement and simplification of safety features with extended time scales for response, thus limiting reliance on operator action for accident mitigation. Enhanced design capabilities through the use of the latest Computer Aided Design and Drafting (CADD) technology. A 38 month construction schedule achieved by using modularization and open-top construction and installation techniques. A more passive containment system incorporating a steel liner and eliminating the need for active spray. A grouping and separation philosophy for maximum protection of redundant safety systems. Ease of equipment qualification and maximum protection of critical components. Replacement of centralized control and monitoring computers with a redundant distributed control system and modern plant display system. A consistent, logical approach to control room design founded on human factors, automation and event management. (author). 6 refs, 3 figs

  13. CANDU 9 safety improvements

    International Nuclear Information System (INIS)

    The CANDU 9 is a family of single-unit Nuclear Power Plant designs based on proven CANDU concepts and equipment from operating CANDU plants capable of generating 900 MWe to 1300 MWe depending on the number of fuel channel used and the type of fuel, either natural uranium fuel or slightly enriched uranium fuel. The basic design, the CANDU 9 480/NU, uses the 480 fuel channel Darlington reactor and employs Natural Uranium (NU) fuel Darlington, the latest of the 900 MWe Class CANDU plants, consists of four integrated units with a total output of approximately 3740 MWe located in Ontario, Canada. AECL has completed the concept definition engineering for this design, and will be completing the design integration engineering by the end of 1996. AECL's design philosophy is to build-in product improvements in evolutionary from the initial prototype plants, NPD and Douglas Point, to today's operating CANDU's construction projects and advanced designs. CANDU 9 safety design follows the evolutionary path, including simple improvements based on existing well-proven CANDU safety concepts. The CANDU 9 builds on the experience base for the Darlington reference plant, and on AECL's extensive safety design experience with single unit CANDU 6 power plants. The latest CANDU 6 plants are being built in Korea by KEPCO at Wolsong 2,3 and 4. The Safety improvements for the CANDU 9 power plant are intended to provide the owner-operator with increased assurance of reliable, trouble-free operation, with greater safety margin, with improved public acceptance, and with ease of licensibility

  14. Application of welded tubes in the manufacturing of pressure equipment

    International Nuclear Information System (INIS)

    The standard EN 10217 concerning welded tubes being also accepted for domestic implementation includes seven parts covering the tubes made of different steel groups using HFW and SAW methods. This paper contains a review of the requirements in the field of manufacturing, inspection and testing as well as application of those tubes. (author)

  15. Burst Test of Stress Corrosion Cracked Stream Generator Tubes under Internal Pressure

    International Nuclear Information System (INIS)

    Outside diameter stress corrosion cracking (ODSCC) has been observed on steam generator (SG) Alloy 600HTMA tubing during in-service inspection. There is tendency for the cracking to be parallel to the axis of the tube. To prevent ODSCC tube burst due to internal pressure and maintain structural integrity, robust model to estimate burst pressure is required. These models should be validated on the basis of burst test data. This paper presents experimental burst test results with stress corrosion cracked SG tubing. The results were compared with the existing burst pressure models.

  16. CANDU fuel performance

    International Nuclear Information System (INIS)

    The paper presents a review of CANDU fuel performance including a 28-element bundle for Pickering reactors, a 37-element bundle for the Bruce and Darlington reactors, and a 37-element bundle for the CANDU-6 reactors. Special emphasis is given to the analysis of fuel defect formation and propagation and definition of fuel element operating thresholds for normal operation and accident conditions. (author)

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

    International Nuclear Information System (INIS)

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

  18. Characterization of magnetically impelled arc butt welded T11 tubes for high pressure applications

    OpenAIRE

    R. Sivasankari; V. Balusamy; P.R. Venkateswaran; G. Buvanashekaran; K Ganesh Kumar

    2015-01-01

    Magnetically impelled arc butt (MIAB) welding is a pressure welding process used for joining of pipes and tubes with an external magnetic field affecting arc rotation along the tube circumference. In this work, MIAB welding of low alloy steel (T11) tubes were carried out to study the microstructural changes occurring in thermo-mechanically affected zone (TMAZ). To qualify the process for the welding applications where pressure could be up to 300 bar, the MIAB welds are studied with variations...

  19. Mathematical simulation of the RBMK reactor pressure tubes ruptures during accidents: Computer code and verification

    International Nuclear Information System (INIS)

    The multiple rupture of the pressure tubes is the most dangerous accident of the channel reactors. There are about 2,000 channels in the RBMK. There exist two potential scenarios: (1) the case of accident when a group of channels becomes overheated; and (2) the case of accident with a rupture of one tube and shock loads on several adjacent channels. The described model considers the prediction technique for potential ruptures according to the first scenario. The probabilistic approach was applied due to existing of substantial scatter and uncertainties in parameters determining pressure tubes deformations and failure in accidents. It was founded on the randomization of the deterministic solution for pressure tube-graphite system deformation and rupture for varied values of chosen chance characters. The mathematical model for the deterministic solution considers the deformation of the system consisting of the pressure tube from the zirconium alloy containing 2.5% of niobium, graphite hard contact rings and graphite blocks. It was solved the common plane strain boundary task. Tube deformation includes three stages: tube deformation until the radial clearance between the tube and graphite disappears; tube deformation with metal flow into the vertical clearance in hard contact rings slits after disappearing of the radial clearance; deformation of the pressure tube-graphite system after closure of the radial clearance up to graphite failure. The mathematical model for the 1st scenario is described. The approach for code verification is also described

  20. The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

    International Nuclear Information System (INIS)

    On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

  1. Ultrasonic C-Scan Parameters for Detection of Hydride Blisters in Zirconium Pressure Tube

    International Nuclear Information System (INIS)

    EMAT Since Zr-2.5Nb pressure tubes have a high risk for the formation of blisters during their operation in pressurized heavy water reactors, there has been a strong incentive to develop a method for the non-destructive detection of blisters grown on the tube surfaces. However, because there is little mismatch in acoustic impedance between the hydride blisters and zirconium matrix, it is not easy to distinguish the boundary between the blister and zirconium matrix wit h the conventional methods. This study focused on the development of the ultrasonic method to detect the hydride blisters formed on Zr-2.5Nb pressure tubes. Hydride blisters were grown on the outer surface of the zirconium pressure tubes using a cold finger attached to steady state thermal diffusion equipment. An ultrasonic velocity ratio method as well as conventional ultrasonic parameters with immersion technique was developed to detect smaller hydride blisters on the zirconium pressure tube.

  2. Towards a sustainable future using pressure tube reactor technology

    International Nuclear Information System (INIS)

    We describe the contribution nuclear energy will make to global energy needs based on the sound foundation of existing technology, infrastructure, natural resources and human knowledge, while meeting the requirements of security of supply (energy independence) and growing demand. Currently all reactors internationally operate on an unsustainable once-through nuclear fuel cycle using uranium fuel. Future decisions will be increasingly based on strategic considerations involving the complete nuclear fuel cycle, including requirements related to supply assurances, resource utilization, proliferation resistance and radioactive waste disposal. Pressure tube reactor (PTR) technology using fuel channels is uniquely suited to respond to the future needs because of its inherent technical characteristics and associated fuel cycle flexibility. PTR channel technology concepts have also continued to advance based on 50 years of continuous development and improvement, with strategic considerations involving the complete nuclear fuel cycle related to: Fuel Availability and Supply Assurances, Uranium, Plutonium and Thorium utilization, Waste Minimization, Proliferation Resistance (Safeguards) ,Assured Licensability, Improved Safety Cost, Competitiveness. We show how nuclear technology development and global sustainability is determined by R and D progress, with challenging technology goals for nuclear energy systems in the four areas of sustainability, economics, safety and reliability, and proliferation resistance and physical protection, leading naturally to the next phase of PTR channel development, namely the high efficiency Supercritical Water Reactor (SCWR). Aggressive targets have been set for R and D and advanced concepts, complementary to the approaches taken in India, which support enhanced safety, cost reduction, resource sustainability, and economical and efficient operation. (author)

  3. Photomultiplier tube failure under hydrostatic pressure in future neutrino detectors

    International Nuclear Information System (INIS)

    Failure of photomultiplier tubes (PMTs) under hydrostatic pressure is a concern in neutrino detection, specifically, in the proposed Long-Baseline Neutrino Experiment project. Controlled hydrostatic implosion tests were performed on prototypic PMT bulbs of 10-inch diameter and recorded using high speed filming techniques to capture failures in detail. These high-speed videos were analyzed frame-by-frame in order to identify the origin of a crack, measure the progression of individual crack along the surface of the bulb as it propagates through the glass, and estimate crack velocity. Crack velocity was calculated for each individual crack, and an average velocity was determined for all measurable cracks on each bulb. Overall, 32 cracks were measured in 9 different bulbs tested. Finite element modeling (FEM) of crack formation and growth in prototypic PMT shows stress concentration near the middle section of the PMT bulbs that correlates well with our crack velocity measurements in that section. The FEM model predicts a crack velocity value that is close to the terminal crack velocity reported. Our measurements also reveal significantly reduced crack velocities compared to terminal crack velocities measured in glasses using fracture mechanics testing and reported in literature

  4. Photomultiplier tube failure under hydrostatic pressure in future neutrino detectors

    International Nuclear Information System (INIS)

    Failure of photomultiplier tubes (PMTs) under hydrostatic pressure is a concern in neutrino detection, specifically, in the proposed Long-Baseline Neutrino Experiment (LBNE) project. Controlled hydrostatic implosion tests were performed on prototypic PMT bulbs of 10-inch diameter and recorded using high-speed filming techniques to capture failures in detail. These high-speed videos were analyzed frame-by-frame in order to identify the origin of a crack, measure the progression of individual crack along the surface of the bulb as it propagates through the glass, and estimate crack velocity. Crack velocity was calculated for each individual crack and an average velocity was determined for all measurable cracks on each bulb. Overall, 32 cracks were measured in 9 different bulbs tested. Finite element modeling (FEM) of crack formation and growth in prototypic PMT shows stress concentration near the middle section of the PMT bulbs that correlates well with our crack velocity measurements in that section. The FEM model predicts a crack velocity value that is close to the terminal crack velocity reported. Our measurements also reveal significantly reduced crack velocities compared to terminal crack velocities measured in glasses using fracture mechanics testing and reported in literature

  5. Sample summary report for IND 1 pressure tube sample

    International Nuclear Information System (INIS)

    This report is the Summary of the Sample Inspection Reports on IND 1 sample submitted to India by the participating laboratories. The Sample Summary Report includes the following: 1. Flaw Characterization Table of IND 1 sample giving the 'Flaw Truth' in terms; - Flaw Position (axial and rotary with respect to punch mark); - Flaw Location (ID or OD); - Flaw Orientation (Axial, Circumferential or Inclined); - Flaw Dimensions (Length, Width and Depth). 2. Photographs of the 'Intentional Flaws' in IND 1 sample. 3. Sample Inspection Table from six laboratories that examined IND 1 sample. 4. Cross Reference Tables for each laboratory, which is the sample inspection table that cross reference to the Flaw Characterization Table. 5. Sample Summary Table that compares the 'True Flaw Dimensions' with the dimensions report by NDE methods by individual laboratories. 6. Sample Inspection Reports for IND 1 sample from individual laboratories. The following Sections of this reports describes the IND 1 pressure tube sample including the nature and number of different kinds of flaws, NDE Methods employed by participating laboratories, their effectiveness in detection and characterization of flaws performance of participating laboratories and the overall analysis of results of investigations

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  9. 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.

  10. Some characteristics of high pressure Ne-He-CH4 filled flash tubes

    International Nuclear Information System (INIS)

    Tests have been carried out in the recently developed high pressure Ne-He-CH4 filled flash tubes. The efficiency of the tubes and the digitisation pulse heights obtained from the tubes were extensively studied for various operating conditions. The layer efficiency of the tubes was found to be very dependent on the rise time of the applied high voltage pulse. An increase in rise-time of the applied high voltage pulse from 0.1 μs to 0.5 μs resulted in a decrease in layer efficiency of the tubes from 88% to 35%. It was found that for satisfactory operation of the flash tubes an applied high voltage pulse of at least 3 μs RC time was required. The digitisation pulse heights obtained from the tubes were found to be critically dependent on the separation of the flash tubes and the ht electrode. (Auth.)

  11. EC6TM - Enhanced Candu 6TM reactor safety characteristics

    International Nuclear Information System (INIS)

    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) CANDUR 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 structure is

  12. Enhancing the moderator effectiveness as a heat sink during loss-of-coolant accidents in CANDU-PHW reactors using glass-peened surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Tiede, R.W.; Sanderson, D.B. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Fong, R.W.L.; Coleman, C.E

    1998-08-01

    The horizontal fuel channel concept is a distinguishing feature of the CANDU-PHW reactor. Each fuel channel consists of a Zr-2.5Nb pressure tube and a Zircaloy-2 calandria tube, separated by a gas filled annulus. The calandria tube is surrounded by heavy-water moderator that also provides a backup heat sink for the reactor core. This heat sink (about 10 mm away from the hot pressure tube) ensures adequate cooling of fuel in the unlikely event of a loss-of-coolant accident (LOCA). One of the ways of enhancing the use of the moderator as a heat sink is to improve the heat-transfer characteristics between the calandria tube and the moderator. This enhancement can be achieved through surface modifications to the calandria tube which have been shown to increase the tube's critical heat flux (CHF) value. An increase in CHIF could be used to reduce moderator subcooling requirements for CANDU fuel channels or increase the margin to dryout. A series of experiments was conducted to assess the benefits provided by glass-peening the outside surface of calandria tubes for postulated LOCA conditions. In particular, the ability to increase the tube's CHF, and thereby reduce moderator subcooling requirements was assessed. Results from the experiments confirm that glass-peening the outer surface of a tube increases its CHF value in pool boiling. This increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels by at least 5 degrees C. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  15. Next generation CANDU plants

    International Nuclear Information System (INIS)

    Future CANDU designs will continue to meet the emerging design and performance requirements expected by the operating utilities. The next generation CANDU products will integrate new technologies into both the product features as well as into the engineering and construction work processes associated with delivering the products. The timely incorporation of advanced design features is the approach adopted for the development of the next generation of CANDU. AECL's current products consist of 700MW Class CANDU 6 and 900 MW Class CANDU 9. Evolutionary improvements are continuing with our CANDU products to enhance their adaptability to meet customers ever increasing need for higher output. Our key product drivers are for improved safety, environmental protection and improved cost effectiveness. Towards these goals we have made excellent progress in Research and Development and our investments are continuing in areas such as fuel channels and passive safety. Our long term focus is utilizing the fuel cycle flexibility of CANDU reactors as part of the long term energy mix

  16. CANDU 9 design

    International Nuclear Information System (INIS)

    AECL has made significant design improvements in the latest CANDU nuclear power plant (NPP) - the CANDU 9. The CANDU 9 operates with the energy efficient heavy water moderated reactor and natural uranium fuel and utilizes proven technology. The CANDU 9 NPP design is similar to the world leading CANDU 6 but is based upon the single unit adaptation of the 900 MWe class reactors currently operating in Canada as in integrated four-unit configurations. The evolution of the CANDU family of heavy water reactors (HAIR) is based on a continuous product improvement approach. Proven equipment and systems from operating stations are standardized and used in new products. As a result of the flexibility of the technology, evolution of the current design will ensure that any new requirements can be met, and there is no need to change the basic concept. This paper will provide an overview for some of the key features of the CANDU 9 NPP such as nuclear systems and equipment, advanced control and computer systems, safety design and protection features, and plant layout. The safety enhancements and operability improvements implemented in this design are described and some of the advantages that can be expected by the operating utility are highlighted. (author)

  17. CANFLEX - an advanced fuel bundle for CANDU

    International Nuclear Information System (INIS)

    The performance of CANDU pressurized heavy-water reactors, in terms of lifetime load factors, is excellent. More than 600 000 bundles containing natural-uranium fuel have been irradiated, with a low defect rate; reactor unavailability due to fuel incidents is typically zero. To maintain and improve CANDU's competitive position, Atomic Energy of Canada Limited (AECL) has an ongoing program comprising design, safety and availability improvements, advanced fuel concepts and schemes to reduce construction time. One key finding is that the introduction of slightly-enriched uranium (SEU, less than 1.5 wt% U-235 in U) offers immediate benefits for CANDU, in terms of fuelling and back-end disposal costs. The use of SEU places more demands on the fuel because of extended burnup, and an anticipated capability to load-follow also adds to the performance requirements. To ensure that the duty-cycle targets for SEU and load-following are achieved, AECL is developing a new fuel bundle, termed CANFLEX (CANdu FLEXible), where flexible refers to the versatility of the bundle with respect to operational and fuel-cycle options. Though the initial purpose of the new 43-element bundle is to introduce SEU into CANDU, CANFLEX is extremely versatile in its application, and is compatible with other fuel cycles of interest: natural uranium in existing CANDU reactors, recycled uranium and mixed-oxides from light-water reactors, and thoria-based fuels. Capability with a variety of fuel cycles is the key to future CANDU success in the international market. The improved performance of CANFLEX, particularly at high burnups, will ensure that the full economic benefits of advanced fuels cycles are achieved. A proof-tested CANFLEX bundle design will be available in 1993 for large-scale commercial-reactor demonstration

  18. 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

  19. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

    AECL has over 40 years of experience in the nuclear field. Over the past 20 years, this unique Canadian nuclear technology has made a worldwide presence, In addition to 22 CANDU reactors in Canada, there are also two in India, one in Pakistan, one in Argentina, four in Korea and five in Romania. CANDU advancements are based on evolutionary plant improvements. They consist of system performance improvements, design technology improvements and research and development in support of advanced nuclear power. Given the good performance of CANOU plants, it is important that this CANDU operating experience be incorporated into new and repeat designs

  20. 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.

  1. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

  2. Corrosion of steams generator tubes in pressurized water reactors

    International Nuclear Information System (INIS)

    The paper summarizes the various corrosion phenomena which have impaired the reliability of PWR steam generators, and reviews the remedies adopted against them: relaxation of residual stresses in the tubes, improved specifications for water chemistry, selection of new materials

  3. 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 volume. At body temperature, the leakage rates of these tracheostomy tubes were 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. PMID:27073395

  4. Computation and measurement of calandria tube sag in pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Calandria tubes and liquid injection shutdown system nozzles in a pressurized heavy water reactor are to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with liquid injection shutdown system tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measurement probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sag between both tubes in the reactor. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted

  5. Effect of tube diameter on heat transfer to supercritical pressure fluid

    International Nuclear Information System (INIS)

    The tightened fuel rod pitch will be adopted in the design of fuel assembly for a Super-critical water Cooled Power Reactor (SCPR). In order to clarify the heat transfer of supercritical fluid flowing in a small diameter tube, experiment was carried out with a smooth tube of 4.4 mm ID in which HCFC-22 fluid at a supercritical pressure flows vertically upward and, experimental data were compared with those in larger diameter tubes. (author)

  6. Tube Plugging Criteria for the High-pressure Heaters of Ulchin NPP 3 and 4

    International Nuclear Information System (INIS)

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

  7. Marketing CANDU internationally

    International Nuclear Information System (INIS)

    The market for CANDU reactor sales, both international and domestic, is reviewed. It is reasonable to expect that between five and ten reactors can be sold outside Canada before the end of the centry, and new domestic orders should be forthcoming as well. AECL International has been created to market CANDU, and is working together with the Canadian nuclear industry to promote the reactor and to assemble an attractive package that can be offered abroad. (L.L.)

  8. CANDU fuel compression tests at elevated temperatures

    International Nuclear Information System (INIS)

    An inlet header large break loss of coolant accident (LOCA) in CANDU reactors with fuelling against flow can cause the fuel to shift in the channels with a consequent reactivity insertion. This results in an increased fuel power transient, and a potential increase in the mialyzed consequences for such events. As the reactor's age and the channel axial gaps increase, the magnitude of the predicted power u-dmient increases. A design solution to reduce the power transient is to limit the amount of fuel movement by reducing the channel axial gap. This solution was implemented into Ontario Hydro's Bruce B and Darlington reactors. A consequence of a reduced channel axial gap is the potential for the fuel column axial expansion to become constrained by the channel end components in large break LOCAs. This experimental program investigated the effects of pellet cracking and elevated sheath temperatures on the ability of the fuel elements, of the 37-element bundle design, to sustain axial loads. The unirradiated fuel elements tested were either in the as-received condition or with the U02 fuel pellets cracked in a mechanical process to simulate the effect of inufflation. The load deformation characteristics demonstrated that, for a given amount of axial compression. the loads sustainable by the elements at elevated sheath temperatures were low. As a result. excess axial expansion would be easily accommodated without further challenge to pressure tube integrity. (author)

  9. Eddy current detection of spacers in the fuel channels of CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Garter Spring (GS) spacers in the fuel channels of CANDU nuclear reactors maintain separation between the hot pressure tube and surrounding moderator cooled calandria tube. Eddy current detection of the four GSs provides assurance that spacers are at or close to design positions and are performing their intended function of maintaining a non-zero gap between pressure tube and calandria tube. Pressure tube constrictions, resulting from relatively less diametral creep at end-of-fuel bundle locations, also produce large eddy current signals. Large constrictions, present in higher service pressure tubes, can produce signals that are 10 times larger than GS signals, reducing GS detectability to 30% in standard GS-detect probes. The introduction of field-focussing elements into the design of the standard GS detection eddy current probe has been used to recover the detectability of GS spacers by increasing the signal amplitude obtained from GSs relative to that from constrictions by a factor of 10. The work presented here compares laboratory, modelling and in-reactor measurements of GS and constriction signals obtained from the standard probe with that obtained from field-focussed eddy current probe designs. (author)

  10. A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDU{sup R} 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

    Energy Technology Data Exchange (ETDEWEB)

    Mostofian, Sara; Boss, Charles [AECL Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga Ontario L5K 1B2 (Canada)

    2008-07-01

    In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

  11. 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

  12. Characterisation of hydride blister in reactor operated zircaloy-2 pressure tube

    International Nuclear Information System (INIS)

    Zircaloy-2 pressure tubes pickup Hydrogen species (H and D) during in-reactor service. The hydrogen pickup leads to hydride precipitation and in the event of a contact between the pressure tube and the calandria tube, hydrogen migrates to the cold spot leading to the formation of hydride blister. One such hydride blister location in an operated Zircaloy-2 pressure tube of RAPS-2 was subjected to metallographic studies and mapping of the microstructure across the tube thickness. Mapping of the hydrogen concentration across the tube thickness was carried out by careful sampling and H estimation by DSC technique. The H profile across the tube thickness, up to the blister boundary, was generated. The hydride blister region was found to be made up of microstructurally different regions starting from dense massive hydride at the outer surface and followed in sequence by a region with dense and thick platelets oriented parallel to the blister boundary and radial platelet region, which subsequently merged with the background platelet distribution appropriate for the average hydrogen content of the pressure tube. The equivalent blister depth corresponding to H content of 16,000 w/ppm has been estimated from the H profile at the blister location. In the case of a hydride blister with measured thickness of 0.4mm the equivalent blister thickness was found to be 0.414mm. Mapping of the hardness of the massive hydride and the adjoining microstructurally different regions was carried out by microhardness measurements at room temperature. (author)

  13. New flux detectors for CANDU 6 reactors

    International Nuclear Information System (INIS)

    CANDU reactors utilize large numbers of in-core self-powered detectors for control and protection. In the original design, the detectors (coaxial cables) were wound on carrier tubes and immersed in the heavy water moderator. Failures occurred due to corrosion and other factors, and replacement was very costly because the assemblies were not designed with maintenance in mind. A new design was conceived based on straight detectors, of larger diameter, in a sealed package of individual 'well' tubes. This protected the detectors from hostile environments and enabled individual failed sensors to be replaced by inserting spares in vacant neighbouring tubes. The new design was made retrofittable to older CANDU reactors. Provision was made for on-line scanning of the core with a miniature fission chamber. The modified detectors were tested in a lengthy development program and found to exhibit superior performance to that of the original detectors. Most of the CANDU reactors have now adopted the new design. In the case of the Gentilly-2 and Point Lepreau reactors, advantage was taken of the opportunity to redesign the detector layout (using better codes and the increased flexibility in positioning detectors) to achieve better coverage of abnormal events, leading to higher trip setpoints and wider operating margins

  14. Fracture toughness of irradiated Zr–2.5Nb pressure tube from Indian PHWR

    International Nuclear Information System (INIS)

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

  15. Falling film flow, heat transfer and breakdown on horizontal tubes

    International Nuclear Information System (INIS)

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

  16. In-service inspection of zircaloy pressure tube of CIRUS reactor

    International Nuclear Information System (INIS)

    The pressurized water loop (PWL) of Cirus uses a 10 meter long zircaloy tube of 57.9 mm ID and 5.4 mm wall thickness. The loop has been used for irradiation testing of various experimental fuel pins since 1972. As part of the refurbishment programme, the condition of Zircaloy-2 pressure tube of the pressurized water loop was investigated by Eddy current and ultrasonic testing. The eddy current probe was balanced over a portion of the tube and the differential signals were recorded for the entire length of the tube. For ultrasonic flaw scanning, gadgets were fabricated and scanning was carried out to evaluate the condition of irradiated pressure tube. For ultrasonic testing an annular probe holder matching to the internal diameter of the zircaloy tube was used for immersion scanning. The probe holder fitted with 10 MHz line focused ultrasonic probes inclined at 28 deg in axial and circumferential directions. A normal spot focused probe was also used to measure wall thickness and detection of laminar flaws. Axial and circumferential grooves of 3% wall thickness depth on ID and OD were used as standard calibration defects. The eddy current and ultrasonic tests did not detect any defect of unacceptable size in the zircaloy pressure tube. (author)

  17. 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...

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

    OpenAIRE

    Daniel Leite Cypriano Neves; Jansen Renato de Carvalho Seixas; Ediberto Bastos Tinoco; Adriana da Cunha Rocha; Ibrahim de Cerqueira Abud

    2004-01-01

    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...

  19. Enhanced CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Full text: The CANDU 6 power reactor is visionary in its approach, remarkable for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Ltd, the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980's as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first CANDU 6 plants- Gentilly 2 and Point Lepreau in Canada, Embalse in Argentina and Wolsong- Unit 1 in Korea have been in service for more than 21 years and are still producing electricity at peak performance and to the end of 2004, their average lifetime Capacity Factor was 83.2%. The newer CANDU 6 units in Romania (Cernavoda 1), Korea (Wolsong-Units 2, 3 and 4) and Qinshan (Phase III- Units 1 and 2) have also been performing at outstanding levels. The average lifetime Capacity Factor of the 10 CANDU 6 operating units around the world has been 87% to the end of 2004. Building on these successes, AECL is committed to the further development of this highly successful design, now focussing on meeting customer's needs for reduced costs, further improvements to plant operation and performance, enhanced safety and incorporating up-to-date technology as warranted. This has resulted in AECL embarking on improving the CANDU 6 design through an upgraded product termed as the 'Enhanced CANDU 6' (EC6)- which incorporates several attractive but proven features that will make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that will be incorporated in the EC6 include increasing the plant's power output, shortening the overall project schedule, decreasing the capital cost, dealing with obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the

  20. Zirconium oxide layer thickness measurement on irradiated PHWR pressure tube by eddy current technique

    International Nuclear Information System (INIS)

    Hydriding is one of the life limiting factors in zircaloy pressure tubes in PHWRs. Hydrogen pick-up in the pressure tube is a direct consequence of the corrosion and oxidation of the internal surface of the pressure tube. Accelerated hydrogen pick-up starts after the oxide layer reaches a critical thickness. Hence development of a non-destructive method for measurement of oxide layer thickness in the pressure tube is very essential in monitoring the condition of the pressure tube in the reactor. Oxide layer thickness can be measured non-destructively using an eddy current technique. The probe contains an eddy current coil driven by a high frequency current which produces a varying magnetic field around the coil. The high frequency electromagnetic field produced by the coil penetrates the non conductive oxide layer and induces eddy currents in the conductive substrate. The eddy currents produce an opposing magnetic field that affects the impedance of the coil. Since the impedance variations are strongly dependent on the distance from the coil to the conducting base metal, the probe produces a voltage signal proportional to the thickness of the non-conductive oxide layer. The results from the development and use of this technique for oxide layer measurement in irradiated pressure tube is presented in this paper. (author)

  1. Analysis of the Internal Pressure in Tube Hydroforming and Its Experimental Investigation

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The internal pressure of the process was studied theoretically and experimentally. The external load character and internal stress character of tube hydroforming were discussed first. Then, according to the characters,the function and classification of internal pressure were presented in general. Base on the stress analysis, its effect on the yield criterion and calculation formula were also researched and derived. To verify the correction of the theoretical analysis and derived formula, experiments with different internal pressures were carried out and the result was compared and discussed. It demonstrates that internal pressure plays an important role in tube hydroforming and theory and formula discussed and derived by this paper are feasible in practice.

  2. The CANDU contribution to environmentally friendly energy production

    International Nuclear Information System (INIS)

    National prosperity is based on the availability of affordable, energy supply. However, this need is tempered by a complementary desire that the energy production and utilization will not have a major impact on the environment. The CANDU energy system, including a next generation of CANDU designs, is a major primary energy supply option that can be an important part of an energy mix to meet Canadian needs. CANDU nuclear power plants produce energy in the form of medium pressure steam. The advanced version of the CANDU design can be delivered in unit modules ranging from 400 to 1200 MWe. This Next Generation of CANDU designs features lower cost, coupled with robust safety margins. Normally this steam is used to drive a turbine and produce electricity. However, a fraction of this steam (large or small) may alternatively be used as process steam for industrial consumption. Options for such steam utilization include seawater desalination, oil sands extraction and heating. The electricity may be delivered to an electrical grid or alternatively used to produce quantities of hydrogen. Hydrogen is an ideal clean transportation fuel because its use only produces water. Thus, a combination of CANDU generated electricity and hydrogen distribution for vehicles is an available, cost-effective route to dramatically reduce emissions from the transportation sector. The CANDU energy system contributes to environmental protection and the prevention of climate change because of its very low emission. The CANDU energy system does not produce any NOx, SOx or greenhouse gas (notably CO2) emissions during operation. In addition, the CANDU system operates on a fully closed cycle with all wastes and emissions fully monitored, controlled and managed throughout the entire life cycle of the plant. The CANDU energy system is an environmentally friendly and flexible energy source. It can be an effective component of a total energy supply package, consistent with Canadian and global climate

  3. Internal heat transfer and pressure drop measurements in a variously baffled shell and tube heat exchanger

    Science.gov (United States)

    Galindo, P.

    1984-06-01

    Heat transfer coefficients, pressure distributions, and fluid flow patterns on the shell side of shell and tube heat exchangers are discussed. The main focus was to quantify the effect of the size of the baffle window on the heat transfer coefficient, which was measured at each tube in the bundle and at three Reynolds numbers. Pressure drops were obtained by measuring detailed pressure distributions within the exchangers. The flow visualizations provided fluid flow patterns adjacent to the shell wall, to the baffle plates, and at each tube of the array. Performance comparisons among the exchangers were carried out holding the heat transfer surface area fixed together with either the pumping power, the mass flow rate, or the pressure drop. Numerical evaluations of commonly employed design procedures are presented using the present data as a means for rank ordering their validity. Tinker's design method provided the best predictions of the present heat transfer and pressure drop results, which are unaffected by leakage and bypass.

  4. 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.

  5. Steam generator tube failures

    International Nuclear Information System (INIS)

    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

  6. Experimental and computational thermalhydraulics research related to CANDU reactor operation and safety

    International Nuclear Information System (INIS)

    This paper describes recent, ongoing and planned research projects at the University of Ottawa, whose objective is to enhance our knowledge of flow and heat transfer in CANDU rod bundles and header/feeder systems and to assist the Canadian nuclear industry in the analysis of operation and safety of CANDU components as well as in designing improved ones. Several experimental facilities are being developed, including a refractive-index matching flow loop for detailed measurements of flows in eccentric annuli and rod bundles, a large-scale, heated rod-bundle facility with air as medium, matching the Reynolds number of the CANDU core and suitable for the study of the effects of geometrical distortions (e.g., pressure tube creep, spacers and fuel element bow) and transients, and an air-water loop for the testing of the operation of wire-mesh sensors and the study of two-phase flows in simple header/feeder vessels. Extensive CFD work on similar topics is also been conducted in parallel with the experiments using the experimental results for its validation. (author)

  7. A Preliminary Assessment of the Adjuster Rod Depletion Effect in the CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yonghee; Roh, Gyuhong; Kim, Won Young; Kim, Hak Sung; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    Lifetime of the Wolsong-1 CANDU reactor, which will be shutdown in April, 2009. Major reactor components such as the pressure tube are to be replaced and it is expected that the CANDU reactor can be operated for additional 25-30 years. Meanwhile, all the reactivity devices including the adjuster rods (ADJ) are supposed to be continuously used without any change. In the CANDU reactor, 21 stainless steel (SS) ADJs are used to control the core power distribution and compensate for some reactivity loss during several abnormal cases. The ADJs are normally fully inserted and the SS absorber should undergo a slow depletion through neutron irradiation for a long time. In April, 2009, the accumulated FPY (Full Power Day) of Wolsong-1 is about 23 years. Depletion of ADJs should result in a smaller ADJ worth and a higher fuel burnup and the core power distribution should also be affected by the ADJ depletion. In this work, the effects of the ADJ depletion have been assessed in terms of ADJ worth, time-average core characteristics.

  8. A study on CANDU model assessment of RELAP5/CANDU using RD-14M B9401 multi-channel RIH break experiment

    International Nuclear Information System (INIS)

    B9401 experiment, performed in RD-14M[1] multi-channel facility, was analyzed using RELAP5/MOD3 and RELAP5/CANDU and compared with experiment results. The RELAP5/CANDU code has been developed since 1998, based on RELAP5, in order to have auditing tool of CANDU NPP. The RELAP5/CANDU code is under developing and they have not been assessed much for a CANDU reactor. Therefore, this study has been initiated with an aim to identify the code applicability in a CANDU reactor by simulating some of the tests performed in the RD-14M facility and to get the assessment results for RELAP5/CANDU code. The RD-14M test facility at Whiteshell Nuclear Research Establishment is a full-scale multi-channel pressurized-water loop. The RELAP5/MOD3 and RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the main phenomena occurred during the transient, in qualitative view. In quantitative view, the RELAP5/CANDU[4] predicted better than that of RELAP5. In the case of experiment that the stratification in fuel channel is dominant, it is expected that RELAP5/CANDU can give more accurate result than RELAP5

  9. Trends in CANDU licensing

    International Nuclear Information System (INIS)

    Modern utilities view nuclear power more and more as a commodity - it must compete 'today' with current alternatives to attract their investment. With its long construction times and large capital investment, nuclear plants are vulnerable to delays once they have been committed. There are two related issues. Where the purchaser and the regulator are experienced in CANDU, the thrust is a very practical one: to identify and resolve major licensing risks at a very early stage in the project. Thus for a Canadian project, the designer (AECL) and the prospective purchaser would deal directly with the AECB. However CANDU has also been successfully licensed in other countries, including Korea, Romania, Argentina, India and Pakistan. Each of these countries has its own regulatory agency responsible for licensing the plant. In addition, however, the foreign customer and regulator may seek input from the AECB, up to and including a statement of licensability in Canada; this is not normally needed for a ''repeat'' plant and/or if the customer is experienced in CANDU, but can be requested if the plant configuration has been modified significantly from an already-operating CANDU. It is thus the responsibility of the designer to initiate early discussions with the AECB so the foreign CANDU meets the expectations of its customers

  10. Ultrasonic testing of pressure contact welded joints of heterogeneous tubes

    International Nuclear Information System (INIS)

    A method of ultrasonic testing of welded joints of tubes of heterogeneous 12Kh1MF and 1Kh18N12T steels is described. The tubes are 32 to 57 mm in diameter with the walls 4 to 6 mm thick. A prism of a serial inclined converter rated at 5 MHz has been used for testing. The testing has been conducted by a singly - and doubly reflected beam at the incident angle of 50 deg. The sensitivity margin of the converter is 35 dB at a 6 to 9 dB signal/noise ratio. 25 specimens have been tested. The test results have shown that amplide of echosignal in a defective sample is by 2 dB higher as compared with the reference signal. Criteria according to which a sample is considered to be defective are given

  11. Advanced eddy current technique for measurement of annular gap between pressure tube and calandria tube in Indian Pressurized Heavy Water Reactors (PHWRs)

    International Nuclear Information System (INIS)

    In Indian Pressurised Heavy Water Reactors (PHWRs), the PT (pressure tube) is designed to be nominally concentric with the encircling CT (calandria tube). Due to various factors PT becomes eccentric with respect to CT over the life of reactor. If this becomes excessive, hot PT will come in contact with cold CT. Such a cold spot could act as potential location for initiating blister formation and premature failure of PT. Hence it is important to periodically measure annular gap between PT and CT. An advanced eddy current technique has been successfully developed and incorporated in BARCIS (BARC Channel Inspection System) for measurement of PT-CT gap. (author). 4 refs., 3 figs

  12. Design and performance of CANDU steam generators

    International Nuclear Information System (INIS)

    The recirculating U-tube steam generators at Ontario Hydro's Pickering and Bruce CANDU Nuclear Generating Stations have demonstrated excellent reliability over many years of operation. Tube failures have been rare, contributing to high plant capacity factors. Of the approximately 390,000 steam generator U-tubes at the Pickering and Bruce plants, one tube leak has occurred at Pickering to date and 12 tubes have leaked at Bruce in a total of 89 reactor years of operation. The success of these units is attributed to an age old respect for steam generation equipment, an ongoing pursuit of research and design advancements, and extensive cooperative efforts on the part of the utility, the system designer, and the supplier. The supplier's involvement began with the steam generator design and manufacture for the very first CANDU plant. The utility's involvement began with their direct participation in the earliest stages of nuclear plant design. The utilities' contribution to the success of these units relates to the rigorous approach used in definition of requirements, in understanding the supplier's design in detail and in making every effort to operate, monitor, and service the equipment with appropriate care. This paper presents the supplier's and utility's approach to achieving these remarkable results. In order to present these viewpoints separately and in some detail, this paper is divided into two parts: Part 1 - The Equipment Supplier Perspective, and Part 2 - The Utility Perspective

  13. 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%.

  14. Retrieval, volume reduction and storage of pressure tubes - operating experiences. Contributed Paper PE-04

    International Nuclear Information System (INIS)

    In Unit-I of Madras Atomic Power Station (MAPS), all the coolant channels were replaced en masse during the year 2005 due to various ageing factors. The Pressure Tubes and Calandria Tubes received as high active waste were stored in an SS lined pool under DM water at CWMF for further processing. It was decided to reduce the volume of the tubes and store in Tile Holes. A detailed process flow sheet for volume reduction was prepared; sub-systems for the individual operating steps were designed, fabricated and erected. Trials with non-active zircaloy tubes were carried out first to validate operating and maintenance procedures. Subsequently, operation with the active pressure tubes with surface dose rate of few hundreds of R/h was demonstrated and clearance obtained for regular operations. The campaign involving the volume reduction and storage of 298 Pressure Tubes and 2 Calandria Tubes was then started and has been successfully completed recently. This paper summarizes the experiences gained from the campaign which was executed with minimum man-rem expenditure conserving precious disposal space. (author)

  15. 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%.

  16. Radiological Characteristics of decommissioning waste from a CANDU reactor

    International Nuclear Information System (INIS)

    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 1016 Bq, 2.09 x 103 W, 5.31 x 1014 m3-water, 4.69 x 105 kg, and 7.38 x 101 m3, 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.

  17. The simulation of CANDU fuel channel behavior in thermal transient conditions

    International Nuclear Information System (INIS)

    In certain LOCA conditions into the CANDU fuel channel, is possible the ballooning of the pressure tube and the contact with the calandria tube. After the contact moment, a radial heat transfer to the moderator through the contact area is occurs. When the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. Thus, the fuel channel could lose its integrity. This paper present a computer code, DELOCA, developed in INR, which simulate the transient thermo-mechanical behaviour of CANDU fuel channel before and after contact. The code contains few models: alloy creep, heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. It was verified step by step by Contact1 and Cathena codes. In this paper, the results obtained at different temperature increasing rates are presented. Also, the contact moment for a RIH 5% postulated accident was presented. The input data was furnished by the Cathena thermo-hydraulic code. (author)

  18. Pressure and wall heat transfer behind a hydrogen/azide detonation wave in narrow tubes

    International Nuclear Information System (INIS)

    The reported study is concerned with the pressure evolution behind the detonation wave in tubes with an interior diameter in the range from 1 to 10 mm. Hydrogen azide in tubes with length-to-diameter ratios greater than 375 was detonated. The initial pressures were in the range from 1 to 20 Torr. The pressure behind the leading shock was measured with piezoelectrical transducers made of lead titanate and lead zirconate. It was found that the detonation velocity depends on wall heat losses. The wall heat flux observed behind the wave front was not in agreement with that calculated for constant flow parameters. In the diameter and pressure range considered, the wall heat flux varies strongly with tube diameter. This observation can be related to flow deviations regarding the Chapman-Jouguet parameters

  19. Simulation-based reactor control design methodology for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Kattan, M.K.; MacBeth, M.J. [Atomic Energy of Canada Limited, Saskatoon, Saskatchewan (Canada); Chan, W.F.; Lam, K.Y. [Cassiopeia Technologies Inc., Toronto, Ontario (Canada)

    1996-07-01

    The next generation of CANDU nuclear power plant being designed by AECL is the 900 MWe CANDU 9 station. This design is based upon the Darlington CANDU nuclear power plant located in Ontario which is among the world leading nuclear power stations for highest capacity factor with the lowest operation, maintenance and administration costs in North America. Canadian-designed CANDU pressurized heavy water nuclear reactors have traditionally been world leaders in electrical power generation capacity performance. This paper introduces the CANDU 9 design initiative to use plant simulation during the design stage of the plant distributed control system (DCS), plant display system (PDS) and the control centre panels. This paper also introduces some details of the CANDU 9 DCS reactor regulating system (RRS) control application, a typical DCS partition configuration, and the interfacing of some of the software design processes that are being followed from conceptual design to final integrated design validation. A description is given of the reactor model developed specifically for use in the simulator. The CANDU 9 reactor model is a synthesis of 14 micro point-kinetic reactor models to facilitate 14 liquid zone controllers for bulk power error control, as well as zone flux tilt control. (author)

  20. Trends in the capital costs of CANDU generating stations

    International Nuclear Information System (INIS)

    This paper consolidates the actual cost experience gained by Atomic Energy of Canada Limited, Ontario Hydro, and other Canadian electric utlities in the planning, design and construction of CANDU-PHWR (CANada Deuterium Uranium-Pressurized Heavy Water Reactor) generating stations over the past 30 years. For each of the major CANDU-PHWR generating stations in operation and under construction in Canada, an analysis is made to trace the evolution of the capital cost estimates. Major technical, economic and other parameters that affect the cost trends of CANDU-PHWR generating stations are identified and their impacts assessed. An analysis of the real cost of CANDU generating stations is made by eliminating interest during construction and escalation, and the effects of planned deferment of in-service dates. An historical trend in the increase in the real cost of CANDU power plants is established. Based on the cost experience gained in the design and construction of CANDU-PHWR units in Canada, as well as on the assessment of parameters that influence the costs of such projects, the future costs of CANDU-PHWRs are presented

  1. An integrated CANDU system

    International Nuclear Information System (INIS)

    Twenty years of experience have shown that the early choices of heavy water as moderator and natural uranium as fuel imposed a discipline on CANDU design that has led to outstanding performance. The integrated structure of the industry in Canada, incorporating development, design, supply, manufacturing, and operation functions, has reinforced this performance and has provided a basis on which to continue development in the future. These same fundamental characteristics of the CANDU program open up propsects for further improvements in economy and resource utilization through increased reactor size and the development of the thorium fuel cycle

  2. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

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

    International Nuclear Information System (INIS)

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

  4. 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.

  5. Pressure tests to assess the significance of defects in boiler and superheater tubing

    International Nuclear Information System (INIS)

    Internal pressure tests on 9 per cent Cr-1 per cent Mo steel tubing containing artificial defects demonstrated that the resultant loss of strength was less than a simple calculation based on the reduced tube thickness would suggest. Bursting tests on tubes containing longitudinal defects of varying length, depth and acuity showed notch strengthening at ambient temperature and at 5500C. A flow stress concept developed for simple bursting tests was shown to apply to creep conditions at 5500C. Results of creep and short-term bursting tests show that the length as well as the depth of the defect is an important factor affecting the life of bursting strength of the tubes. Defects less than 10 per cent of the tube thickness were found to have an insignificant effect. (author)

  6. R and D in support of CANDU plant life management

    International Nuclear Information System (INIS)

    One of the keys to the long-term success of CANDUs is a high capacity factor over the station design life. Considerable R and D in underway at AECL to develop technologies for assessing, monitoring and mitigating the effect of plant ageing and for improving plant performance and extending plant life. To achieve longer service life and to realize high capacity factor from CANDU stations, AECL is developing new technologies to enhance fuel channel and steam generator inspection capabilities, to monitor system health, and to allow preventive maintenance and cleaning (e.g., on-line chemical cleaning processes that produce small volumes of wastes). The life management strategy for fuel channels and steam generators requires a program to inspect components on a routine basis to identify mechanisms that could potentially affect fitness-for-service. In the case of fuel channels, the strategy includes inspections for dimensional changes, flaw detection, and deuterium concentration. New techniques are been developed to enhance these inspection capabilities; examples include accurate measurement of the gap between a pressure tube and its calandria tube and rapid full-length inspections of steam generator tubes for all known flaw types. Central to life management of components are Fitness-for-Service Guidelines (FFSG) that have been developed with the CANDU Owners Group (COG) that provide a standardized method to assess the potential for propagation of flaws detected during in-service inspections, and assessment of any change in fracture characteristics of the material. FFSG continue to be improved with the development of new technologies such as the capability to credit relaxation of stresses due to creep and non-rejectable flaws in pressure tubes. Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that system health is continually monitored and managed. AECL has developed a system Health Monitor

  7. Steady state 3-D simulation of CANDU6 moderator circulation under the normal operating condition

    International Nuclear Information System (INIS)

    Moderator circulation inside the CANDU-6 reactor vessel of Wolsong 2/3/4 (KOREA) under normal operating condition is analyzed by using computational fluid dynamics for predicting the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by internal heating is accounted for by the Boussinesq approximation. The standard k-ε turbulence model associated with logarithmic wall treatment is applied to predict the turbulent jet flows from eight inlet nozzles. The matrix of calandria tubes in the core region is simplified as porous media, in which anisotropic hydraulic impedance is modeled using an empirical correlation of frictional pressure loss. The governing equations are solved by CFX-4.4, a commercial CFD code developed by AEA technology. The CFD model has been successfully validated against experimental data obtained in the Stern Laboratories Inc. (SLI) in Hamilton, Ontario. Steady-state 3-D prediction of the CANDU-6 moderator circulation under normal operating conditions gives a maximum moderator temperature of 82.9oC at the upper core region, which corresponds to the minimum subcooling temperature of 24.8oC considering hydrostatic pressure increase. The flow pattern on the normal plane to z-axis is determined as 'mixed-type'. (author)

  8. CANDU development: the next 25 years

    International Nuclear Information System (INIS)

    CANDU Pressurized Heavy Water Reactors have three main characteristics that ensure viability for the very long term. First, great care has been taken in designing the CANDU reactor core so that relatively few neutrons produced in the fission process are absorbed by structural or moderator materials. The result is a reactor with high neutron economy that can burn natural uranium and a core that operates with 2-3 times less fissile content than other, similarly-sized reactors. In addition to neutron economy, the use of a simple bundle design and on-power fuelling augment the ability of CANDU reactors to burn a variety of fuels with relatively low fissile content with high efficiency. This ensures that fuel supply will not limit the applicability of the technology over the long term. Second, the presence of large water reservoirs ensures that even the severest postulated accidents are mitigated by passive means. For example, the presence of the heavy water moderator, which operates at low pressure and temperature, acts as a passive heat sink for many postulated accidents. Third, the modular nature of the core (e.g., fuel channels) means that components can be relatively easily replaced for plant life extension and upgrading. Since these factors all influence the long-term sustainability of CANDU nuclear technology, it is logical to build on this base and to add improvements to CANDU reactors using an evolutionary approach. This paper reviews AECL's product development directions and shows how the above characteristics are being exploited to improve economics, enhance safety, and ensure fuel cycle flexibility for sustainable development. (author). 21 refs., 9 figs

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

    International Nuclear Information System (INIS)

    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

  10. Creep-rupture tests of internally pressurized Hastelloy-X tubes

    Science.gov (United States)

    Gumto, K. H.; Colantino, G. J.

    1973-01-01

    Seamless Hastelloy-X tubes with 0.375-in. outside diameter and 0.025-in. wall thickness were tested to failure at temperatures from 1400 to 1650 F and internal helium pressures from 800 to 1800 psi. Lifetimes ranged from 58 to 3600 hr. The creep-rupture strength of the tubes was from 20 to 40 percent lower than that of sheet specimens. Larson-Miller correlations and photomicrographs of some specimens are presented.

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

    DEFF Research Database (Denmark)

    Lindgreen, Britta; Tvergaard, Viggo; Needleman, Alan

    2008-01-01

    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...... stiffness against further expansion of the tube. The rate dependence of the necking behavior gives noticeable differences in neck development for slow loading versus fast loading....

  12. Replacement of CANDU reactivity Control Devices

    International Nuclear Information System (INIS)

    Ontario Hydro operates 20 AECL designed CANDU nuclear power reactors, some of which have been in service for 20 years. These pressurized heavy water, natural uranium fuelled reactors, ranging in size from 540 to 900 MWe, have continuously provided high capacity factors and low total electricity production costs, among the world's leaders in performance. CANDU's inherently have large cores and utilize a large number of diverse types of Reactivity Control Devices (RCDs) for fully automatic, continuous measurement and regulation of bulk power level as well as spatial uniformity of fission power in the core. The devices also control start-up and power manoeuvering. Other RCDs provide two independent systems of measurement and neutron absorber insertion for fast reactor shutdown. The continuous proper operation of RCDs obviously has significant influence on plant performance and availability, yet Ontario Hydro (OH) experience is that no significant loss of capacity factor has been attributed to the RCDs. This paper focuses on these Ontario Hydro replacement practices as they apply to RCD equipment in CANDU plants. The particular practices described relate to some extent to the unique aspects of CANDU plants, but the concepts of thorough planning, operational quality and teamwork are universally valid. Practicing safe, efficient component replacements contributes to reliable, cost effective plant operation. (author)

  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 com......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......) 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...... the model and are compared with results of experimental pressure wave propagation in the liquid-filled, flexible tube. A good agreement between experimental data and theoretical predictions is found....

  14. Advanced CANDU control centre

    International Nuclear Information System (INIS)

    The CANDU 9 design is based upon the 900 MWe class Darlington station in Canada, which is among the world leading nuclear power stations for capacity factor with low operation, maintenance and administration costs. The CANDU 9 design provides an advanced control centre with enhanced operations features. The advanced AECL control centre design includes the proven functionality of existing CANDU control centres, those implementable characteristics identified by systematic design combined with a human factors analysis of operations requirements and features needed to improve station operability which are made possible by the application of current technology. The design strategy is to preserve the general main control room operations staff work area as unchanged as possible to facilitate the inclusion of past features and operational experience while incorporating operability improvements. The author will present those features of the advanced CANDU control centre which facilitates improved operability capabilities. As well, aspects of the design process utilized, application of simulation technology and conclusions regarding this design approach will be reviewed

  15. CANDU, building the future

    International Nuclear Information System (INIS)

    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

  16. CANDU market prospects

    International Nuclear Information System (INIS)

    This 1994 survey of prospective markets for CANDU reactors discusses prospects in Turkey, Thailand, the Philippines, Korea, Indonesia, China and Egypt, and other opportunities, such as in fuel cycles and nuclear safety. It was concluded that foreign partners would be needed to help with financing

  17. 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.

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

    International Nuclear Information System (INIS)

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

  19. Analysis of the pressure tube failure at Pickering NGS A unit 2

    International Nuclear Information System (INIS)

    About noon on the 1st August 1983, the pressure tube in fuel channel G16 of the Pickering NGS A unit 2 reactor developed a critical through-wall crack and failed by fast fracture after 342 days of continuous full power operation. Following removal of the fuel, a TV inspection inside the fuel channel revealed an axial crack in the bottom of the pressure tube approximately 2 metres long, in which two fuel pencils were lodged. After extracting the fuel pencils, the fuel channel was removed and shipped to the Atomic Energy of Canada Limited's Chalk River Nuclear Laboratories for detailed examination to determine the cause of failure. Examination of failures normally takes a course of looking at the fracture and gradually refining the work into finer detail to determine the actual origin of the failure. In this case, several other aspects also needed to be examined. The position of the garter spring was very important, as was examination of the calandria tube, which was subsequently removed. During the inspection several other fuel channels in Pickering A, Bruce A and NPD reactors were inspected and some removed for further assessment at CRNL. All these aspects came together to outline the cause and mechanism of failure. The following gives a very brief review of the salient features of the examination of Zircaloy 2 and zirconium-niobium pressure tubes and the implication for operation of subsequent reactors which have zirconium-niobium pressure tubes

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

    International Nuclear Information System (INIS)

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

  1. 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.

  2. 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.

  3. Development of poison injection code-COPJET for high pressure liquid poison injection in pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Shut Down System-2 (SDS-2) in advanced vertical pressure tube type reactor, provides rapid reactor shutdown by high pressure injection of a neutron absorbing liquid called poison, into the moderator in the calandria. Poison inside the calandria is distributed by poison jets issued from holes provided in the injection tubes. Effectiveness of the system depends on the rate and spread of the poison in the moderator. In this study, a transient one-dimensional (1-D) hydraulic code, COPJET is developed, to predict the performance of system by predicting poison jet length with time. Validation of the COPJET is done with the data available in literature. Thereafter, it is applied for poison jet length prediction of advanced vertical pressure type reactor. (author)

  4. Experimental investigation on heat transfer characteristics of water in vertical upward tube under supercritical pressure

    International Nuclear Information System (INIS)

    Within the range of pressure from 22.5 to 30 MPa, mass flux from 1009 to 1626 kg/(m2·s), and inner wall heat flux from 216 to 822 kW/m2, an in-depth experiment was conducted under supercritical pressure to investigate the heat transfer characteristics of water in vertical upward smooth tube. The heat transfer characteristics of water under supercritical pressure were obtained in the experiment. The effects of pressure, inner wall heat flux and mass flux on the heat transfer coefficient and inner wall temperature were analyzed, the heat transfer mechanism was discussed, and the corresponding empirical correlations were also presented. The experimental results show that when the bulk fluid temperature is near the pseudo-critical temperature, the tube wall temperature increases slowly with the fluid enthalpy, the heat transfer coefficient gets larger abruptly and the heat transfer enhancement phenomenon occurs in the smooth tube. Otherwise, the tube wall temperature increases obviously with the increasing fluid enthalpy and the heat transfer coefficient is low. With the increase of pressure and inner wall heat flux, and with the decrease of mass flux, the inner wall temperature increases, the heat transfer coefficient decreases and the heat transfer enhancement is weakened. With the increase of inner wall heat flux, the maximum of heat transfer coefficient appears ahead. (authors)

  5. Development and validation of the 3-D CFD model for CANDU-6 moderator temperature predictions

    International Nuclear Information System (INIS)

    A computational fluid dynamics model for predicting the moderator circulation inside the CANada Deuterium Uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the Calandria tubes. The buoyancy effect induced by internal heating is accounted for by Boussinesq approximation. The standard κ-ε turbulence model associated with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the Calandria tubes in the core region is simplified to porous media, in which an-isotropic hydraulic impedance is modeled using an empirical correlation of the frictional pressure loss. The governing equations are solved by CFX-4.4, a commercial CFD code developed by AEA technology. The CFD model has been successfully verified and validated against experimental data obtained in the Stern Laboratories Inc. (SLI) in Hamilton, Ontario

  6. Evaluation and analysis of critical crack length of irradiated pressure tubes from Indian pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Results of fracture toughness KJic were computed from transverse tensile properties of reactor operated pressure tubes, and axial critical crack length values derived from KJic are presented. Similarly fracture resistance curves derived from tensile properties of reactor operated pressure tubes and axial critical crack length values computed therefrom are presented. Under normal operating condition of the reactors the pressure tubes experience temperatures ranging from 250 deg C to 300 deg C. In occurrence of contact between pressure tube and calandria tube, the contact region may not be expected to have a mean through wall temperature below 200 deg C. The axial critical crack length of three reactor operated pressure tubes, therefore were evaluated in the temperature range 200 deg C to 300 deg C. The significance of the magnitude of the evaluated critical crack length is discussed. (author)

  7. The Method of Drainage Tube Pulling under Negative Pressure in Patients with Open Heart Surgery

    Directory of Open Access Journals (Sweden)

    Ozcan Gur

    2012-06-01

    Full Text Available Purpose: Postoperative complications is to impress negatively patient mortality. One of this complications relate to respiratory system. In our study, we aimed to present our clinical experience related to negative pressure practise while pulling thorax drainage tube. Methods: 448 patients (337 male, 111 female taken to open heart surgery december 2007 and november 2011 in our clinic. The mean age was 57,33±3,6 in patients. Drainage tubes were pulled on postoperative 1 day except much drainage and pneumothorax. Negative pressure was applied thorax drainage tubes between 70 and 100 mm Hg while pulling thorax drainage tubes. Results: Total 1008 number drainage tubes were placed. Thorasyntesis was applied because of liquid collection in left hemithorax in 3 patients. Pneumothorax was seen in 1 patient. Conclusion: We consider the rate of pneumothorax and pulmonary infection due to liquid collection in postoperative period may be reduced with modification while pulling drainage tube after heart surgery. [Cukurova Med J 2012; 37(3.000: 146-149

  8. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    The Three Mile Island accident spurred a world-wide interest in severe accidents. The initial reaction was to increase the preventative measures in existing designs, followed by development of predictive capabilities to improve the management of severe accidents. Recently, emphasis has been placed in new designs on mitigative measures which slow down or contain the progression of a severe accidents. U.S. requirements for Advanced Light Water Reactor designs must now: provide reactor cavity floor space to enhance debris spreading; provide a means to flood the reactor cavity to assist in the cooling process. The paper describes how CANDU Pressurized Heavy Water Reactors (PHWRs) have severe accident prevention and mitigation inherent in the design; in particular, the U.S. severe accident requirements can be met without significant change to the design of current CANDUs. (author). 32 refs, 7 figs, 1 tab

  9. Estimation of impact pressure due to rupture in beam-tube for research reactor

    International Nuclear Information System (INIS)

    Neutrons have been used for studies in material sciences of physics, chemistry, metals and alloys, ceramics, polymers, and biological sciences. This application leads to build up research reactor all over the world. JRTR (Jordan Research and Training Reactor) which plans to build up in Jordan is multipurpose research reactor which is developed entirely with domestic technology to overseas. Thermal power is 5MW upgradable 10MW. JRTR have four horizontal beam tubes, 3 ST(Standard) and 1NR (Neutron radiography). The beam tube's cavities are filled with helium, purged regularly to prevent a build-up of radioactive gases and moisture. They are highly reliable because they have no moving parts. The beam tube embedded part is aligned with its corresponding beam tube in the reflector. Objective of this study is to describe water hammer phenomenon in beam tube and determine an impact pressure charged in end film of beam tube for accomplishing nuclear safety function of research reactor while beam tube is ruptured due to some accident such as earthquake. The water hammer was experimentally and analytically studied by Lai, Saruba, Ballanco, and Watters

  10. Creep properties of electric resistance welded boiler tubes under internal pressure

    International Nuclear Information System (INIS)

    Creep rupture tests on electric resistance welded (ERW) tubular specimens of carbon steel and 1% Cr-0.5% Mo steel and burst tests on thickness-deviated tubular specimens of carbon steel are described. Also, changes of structures and mechanical properties of 1% Cr-0.5% Mo steel tubes after exposure to 5500C for up to 10,000 hours under a tensile hoop stress of 108 MPa are described. The creep rupture properties of ERW boiler tubes were proved to be quite comparable to those of seamless tubes, and the slightest deviation in wall thickness was shown to affect the internal pressure rupture behavior. Changes of structures at welded portion of ERW 1% Cr-0.5% Mo steel tubes were as same as those of base metal

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

    International Nuclear Information System (INIS)

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

  12. Hydride blister features in Zircaloy-2 pressure tubes removed from Pickering units 1 and 2: A selection of micrographs

    International Nuclear Information System (INIS)

    Following the failure of the pressure tubes in channel G16 in Pickering Unit 2 (P2G16) sections of it and other pressure tubes removed from Pickering Units 1 and 2 were examined to study the blisters that led to the failure. Several tubes were removed for examination before the decision was made to replace all of the Zircaloy-2 tubes with Zr-2.5Nb tubes. Additional work on the Zircaloy-2 tubes was stopped at this time. The examination carried out on the tubes at AECL CRL included a visual inspection of the outside surface looking for blisters, followed by optical metallography of the transverse and axial sections that contained the blisters. The micrographs are of blisters found in pressure tubes removed from Pickering units 1 and 2. The micrographs document principal features found associated with the blisters in the pressure tubes not to cover all the details of blister growth and of crack initiation and growth. The pictures found within the report were to give an appreciation of the different stages that happen after the pressure tubes come into contact with the calandria tubes. 32 figs

  13. A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDUR 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

    International Nuclear Information System (INIS)

    In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

  14. Francis turbine draft tube modelling for prediction of pressure fluctuations on prototype

    Science.gov (United States)

    Alligné, S.; Landry, C.; Favrel, A.; Nicolet, C.; Avellan, F.

    2015-12-01

    The prediction of pressure fluctuations amplitudes on Francis turbine prototype is a challenge for hydro-equipment industry since it is subjected to guarantees to ensure smooth and reliable operation of the hydro units. The European FP7 research project Hyperbole aims to setup a methodology to transpose the pressure fluctuations measured on the reduced scale model to the prototype generating units. This paper presents this methodology which relies on an advanced modelling of the draft tube cavitation flow, and focuses on the transposition to the prototype of the draft tube model parameters identified on the reduced scale model. Different modelling assumptions of the draft tube are considered and their influence on the eigenmodes and the forced response of the system are presented.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Estimation of fracture resistance curve of pressure tube from ring tension test

    International Nuclear Information System (INIS)

    For the estimation of through-wall axial critical crack length of zirconium alloy pressure tube in-residence in pressurized heavy water reactors, fracture resistance curves are needed. A method developed to derive the curve from tensile properties is elaborated. The critical crack length derived from the curve was compared with the critical crack length derived from fracture toughness KJic obtained from the tensile properties utilising another method. (author)

  18. Influence of temperature and pressure in adsorption of uranium hexafluoride on carbon nano tube by Monte Carlo simulation study

    International Nuclear Information System (INIS)

    Uranium hexafluoride physical adsorption on armchair carbon nano tubes is studied with Monte Carlo Simulations in a wide range of temperatures and pressures. All of the particle particle interaction are modeled with Lennard-Jones potential. We have written a FORTRAN program for Monte Carlo simulation and then calculated the inside density of carbon nano tube, out side density of carbon nano tube and total density of carbon nano tube. The conclusions drown from theses calculations are then contrasted and compared. Result show that the total amount of uranium hexafluoride adsorption on single-walled carbon nano tube, increases with high pressure and low temperature

  19. The CANDU 3 containment structure

    International Nuclear Information System (INIS)

    The design of the CANDU 3 nuclear power plant is being developed by AECL CANDU's Saskatchewan office. There are 24 CANDU nuclear power units operating in Canada and abroad and eight units are under construction is Romania and South Korea. The design of the CANDU 3 plant has evolved on the basis of the proven CANDU design. The experiences gained during construction, commissioning and operation of the existing CANDU plants are considered in the design. Many technological enhancements have been implemented in the design processes in all areas. The object has been to develop an improved reactor design that is suitable for the current and the future markets worldwide. Throughout the design phase of CANDU 3, emphasis has been placed in reducing the cost and construction schedule of the plant. This has been achieved by implementing design improvements and using new construction techniques. Appropriate changes and improvements to the design to suit new requirements are also adopted. In CANDU plants, the containment structure acts as an ultimate barrier against the leakage of radioactive substances during normal operations and postulated accident conditions. The concept of the structural design of the containment structure has been examined in considerable detail. This has resulted in development of a new conceptual design for the containment structure for CANDU 3. This paper deals with this new design of the containment structure

  20. Incentives for improvement of CANDU

    International Nuclear Information System (INIS)

    CANDU is a relatively young technology which has demonstrated many achievements as an electrical power generation system. These achievements include an unsurpassed safety record, high annual and lifetime capacity factors, low electricity cost and a broad range of other performance strengths which together indicate that the CANDU technology is fundamentally sound. Known capabilities not yet fully exploited, such as advanced fuel cycle options, indicate that CANDU technology will continue to pay strong dividends on research, development and design investment. This provides a strong incentive for the improvement of CANDU on a continuing basis

  1. Pre and post garter spring repositioning ultrasonic inspection of pressure tubes

    International Nuclear Information System (INIS)

    This paper present a description of the ultrasonic cracked hydride blister detections system used for pre and post inspection of pressure tubes during garter spring repositioning in CNE (Embalse Nuclear Power Station). Ultrasonic system setup configuration, transducers characteristics, blister detection head, calibration of parameters, operating procedure, records of ultrasonic inspections and evaluation. (author)

  2. Detection of Hydride Blisters in Zirconium Pressure Tubes using Ultrasonic Mode Conversion and Velocity Ratio Method

    International Nuclear Information System (INIS)

    When the pressure tubes(f are in contact with the calandria tube(CT) in the pressurized heavy water reactor(PHWR), the temperature difference between inner and outer wall of W results in a thermal diffusion of hydrogen (deuterium) and hydride blisters are formed on the outer surface of PT. Because the hydride blisters and zirconium matrix are acoustically continuous, it is not easy to distinguish the blisters from the matrix with conventional ultrasonic method. An ultrasonic velocity ratio method was developed to detect small hydride blisters on the zirconium pressure tube. Hydride blisters were grown in the PT specimen using a steady state thermal diffusion device. The flight times of longitudinal echo and reflected shear echo from the outer surface were measured accurately. The velocity ratio of the longitudinal wave to the shear wave was calculated and displayed using contour plot. Compared to the conventional flight time method of longitudinal wave, the velocity ratio method shows superior sensitivity to detect smaller blisters as well as better images for the blister shapes. Detectable limit of the outer shape of the hydride blisters was conservatively estimated as 50μm, with the same specifications of ultrasonic transducer used in the actual PHWR pressure tube inspection

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

    International Nuclear Information System (INIS)

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

  4. Irradiation Effect on the Mechanical Property for Wolsong 1 Pressure Tube

    International Nuclear Information System (INIS)

    Summary: Need for accurate prediction model for PT dimensional change - One of key data to determine power decrease and lifetime; • Limit for PT measured data - Necessity for collaboration of HWR countries; • Irradiation Effect on PT Mechanical Property - Basic data of pressure tube behavior for the irradiation - More research on the microstructure change for the irradiation

  5. Evaluated Plan Stress Of Weld In Pressure Tube Using X Ray Diffraction Technique

    International Nuclear Information System (INIS)

    X ray diffraction is a fundamental technique measuring stress, this technique has determined crystal strain in materials, from that determined stress in materials. This paper presents study of evaluating plane stress of weld in pressure tube, using modern XRD apparatus: X Pert Pro. (author)

  6. Proof testing of CANDU concrete containment structures

    International Nuclear Information System (INIS)

    Prior to commissioning of a CANDU reactor, a proof pressure test is required to demonstrate the structural integrity of the containment envelope. The test pressure specified by AECB Regulatory Document R-7 (1991) was selected without a rigorous consideration of uncertainties associated with estimates of accident pressure and conatinment resistance. This study was undertaken to develop a reliability-based philosophy for defining proof testing requirements that are consistent with the current limit states design code for concrete containments (CSA N287.3).It was shown that the upodated probability of failure after a successful test is always less than the original estimate

  7. CANDU at the crossroads

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1990-11-01

    ''Ready for the challenge of the 90s'' was the theme of this year's gathering of the Canadian Nuclear Association held in Toronto, 3-6 June. What that challenge really entails is whether the CANDU system will survive as the last remaining alternative to the light water reactor in the world reactor market, or whether it will decline into oblivion along with the Advanced Gas Cooled reactor and so many other technically excellent systems which have fallen along the way. The fate of the CANDU system will not be determined by its technical merits, nor by its impeccable safety record. It will be determined by public perceptions and by the deliberations of an Environmental Assessment Panel established by the Government of Ontario. The debate at the Association meeting is reported. (author).

  8. Study on the Pressure Drop in the Helical Wire Inserted Tube Using FLUENT Code

    International Nuclear Information System (INIS)

    Bubble departure from the surface is an important physical phenomenon in flow boiling CHF especially in subcooled and low quality region. It is necessary to remove the bubbles near the wall surface because relevant escape of bubble from the heated wall makes the maximum heat removal, the CHF increase. It is possible to remove bubbles on the inner surface of the tube wall by the centrifugal force due to the swirl flow using swirl flow generators such as wire coils, ribs and twisted tapes. It is known that the heat transfer coefficient and the critical heat flux are improved when swirl flow generators are used. However, at the same time, an increase of the energy consumption in the entire system is generated, because the flow resistance also increases. In the turbulent flow inside a wire inserted tube, the laminar sub-layer is disturbed by the wire, which increases the friction and thereby the pressure drop. In this context, only when the increased CHF is profitable compared to the increased pressure drop, the use of swirl generator is justifiable. The purpose of this study is to clarify the effect of an inserted helical wire tube on the pressure drop in the tube and optimize the effect of swirl generator in balancing pressure drop and swirl generation. Pressure drop is calculated using computational flow dynamics code, FLUENT, as the pitches and the wire diameters are varied. The working fluid is water with 0.01m inner diameter tube, which is the similar one with the actual annular fuel size. The calculated results are compared with the correlation based on experimental data

  9. Qinshan CANDU NPP outage performance improvement through benchmarking

    International Nuclear Information System (INIS)

    With the increasingly fierce competition in the deregulated Energy Market, the optimization of outage duration has become one of the focal points for the Nuclear Power Plant owners around the world. People are seeking various ways to shorten the outage duration of NPP. Great efforts have been made in the Light Water Reactor (LWR) family with the concept of benchmarking and evaluation, which great reduced the outage duration and improved outage performance. The average capacity factor of LWRs has been greatly improved over the last three decades, which now is close to 90%. CANDU (Pressurized Heavy Water Reactor) stations, with its unique feature of on power refueling, of nuclear fuel remaining in the reactor all through the planned outage, have given raise to more stringent safety requirements during planned outage. In addition, the above feature gives more variations to the critical path of planned outage in different station. In order to benchmarking again the best practices in the CANDU stations, Third Qinshan Nuclear Power Company (TQNPC) have initiated the benchmarking program among the CANDU stations aiming to standardize the outage maintenance windows and optimize the outage duration. The initial benchmarking has resulted the optimization of outage duration in Qinshan CANDU NPP and the formulation of its first long-term outage plan. This paper describes the benchmarking works that have been proven to be useful for optimizing outage duration in Qinshan CANDU NPP, and the vision of further optimize the duration with joint effort from the CANDU community. (authors)

  10. Status of advanced technologies for CANDU reactors

    International Nuclear Information System (INIS)

    The future development of the CANDU reactor is a continuation of a successful series of reactors, the most recent of which are nine CANDU 6 Mk 1* units and four Darlington units. There are three projects underway that continue the development of the CANDU reactor. These new design projects flow from the original reactor designs and are a natural progression of the CANDU 6 Mk 1, two units of which are operating successfully in Canada, one each in Argentina and Korea, with five more being built in Rumania. These new design projects are known as: CANDU 6 Mk 2, an improved version of CANDU 6 Mk 1; CANDU 3, a small, advanced version of the CANDU 6 Mk 1; CANDU 6 Mk 3, a series of advanced CANDU reactors. A short description of modified versions of CANDU reactors is given in this paper. 5 figs

  11. Advancing CANDU experience to the world steam generator market

    International Nuclear Information System (INIS)

    Tube degradation in certain recirculating nuclear steam generators has provided a market for steam generator replacement. Prior to this need, B and W supplied over 200 steam generators for CANDU nuclear plants. With this experience, and implementing extensive research and development improvements in material selection, design enhancements, and new manufacturing and analytical methods, B and W has supplied or secured orders for the replacement of 26 steam generators. Along with plans for new replacement orders, B and W will continue to supply steam generators for future CANDU plants. This paper will review the progression of B and W's CANDU experience to meet the replacement steam generator market, and examine the continuous improvements required for today's increasingly demanding nuclear specifications. (author). 1 tab., 4 figs

  12. Joint studies on large CANDU

    International Nuclear Information System (INIS)

    CANDU PHWRs have demonstrated generic benefits which will be continued in future designs. These include economic benefits due to low operating costs, business potential, strategic benefits due to fuel cycle flexibility and operational benefits. These benefits have been realized in Korea through the operation of Wolsong 1, resulting in further construction of PHWRs at the same site. The principal benefit, low electricity cost, is due to the high capacity factor and the low fuel cost for CANDU. The CANDU plant at Wolsong has proven to be a safe, reliable and economical electricity producer. The ability of PHWR to burn natural uranium ensures security of fuel supply. Following successful Technology Transfer via the Wolsong 2,3 and 4 project, future opportunity exists between Korea and Canada for continuing co-operation in research and development to improve the technology base, for product development partnerships, and business opportunities in marketing and building PHWR plants in third countries. High reliability, through excellent design, well-controlled operation, efficient maintenance and low operating costs is critical to the economic viability of nuclear plants. CANDU plants have an excellent performance record. The four operating CANDU 6 plants, operated by four utilities in three countries, are world performance leaders. The CANDU 9 design, with higher output capacity, will help to achieve better site utilization and lower electricity costs. Being an evolutionary design, CANDU 9 assures high performance by utilizing proven systems, and component designs adapted from operating CANDU plants (Bruce B, Darlington and CANDU 6). All system and operating parameters are within the operating proven range of current plants. KAERI and AECL have an agreement to perform joint studies on future PHWR development. The objective of the joint studies is to establish the requirements for the design of future advanced CANDU PHWR including the utility need for design improvements

  13. The next generation CANDU 6

    International Nuclear Information System (INIS)

    AECL's product line of CANDU 6 and CANDU 9 nuclear power plants are adapted to respond to changing market conditions, experience feedback and technological development by a continuous improvement process of design evolution. The CANDU 6 Nuclear Power Plant design is a successful family of nuclear units, with the first four units entering service in 1983, and the most recent entering service this year. A further four CANDU 6 units are under construction. Starting in 1996, a focused forward-looking development program is under way at AECL to incorporate a series of individual improvements and integrate them into the CANDU 6, leading to the evolutionary development of the next-generation enhanced CANDU 6. The CANDU 6 improvements program includes all aspects of an NPP project, including engineering tools improvements, design for improved constructability, scheduling for faster, more streamlined commissioning, and improved operating performance. This enhanced CANDU 6 product will combine the benefits of design provenness (drawing on the more than 70 reactor-years experience of the seven operating CANDU 6 units), with the advantages of an evolutionary next-generation design. Features of the enhanced CANDU 6 design include: Advanced Human Machine Interface - built around the Advanced CANDU Control Centre; Advanced fuel design - using the newly demonstrated CANFLEX fuel bundle; Improved Efficiency based on improved utilization of waste heat; Streamlined System Design - including simplifications to improve performance and safety system reliability; Advanced Engineering Tools, -- featuring linked electronic databases from 3D CADDS, equipment specification and material management; Advanced Construction Techniques - based on open top equipment installation and the use of small skid mounted modules; Options defined for Passive Heat Sink capability and low-enrichment core optimization. (author)

  14. 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.

  15. Control of reactor inlet header temperature (RIHT) rise in CANDU

    International Nuclear Information System (INIS)

    Fouling of tube surfaces in a CANDU steam generator is analyzed using a mathematical model and is shown to account for a major portion of the observed Reactor Inlet Header Temperature (RIHT) rise. First, a detailed heat transfer model is made to account for tube wall temperature at every point along a tube, then the solubility of magnetite is calculated at that wall temperature to check for primary side fouling. Once fouling can occur by magnetite crystal growth on the wall, the rate of fouling is determined by the mass transfer of dissolved iron from the water to the tube surface. The fouling deposit increases heat transfer resistance and thus the heavy water outlet temperature (RIHT) rises. This rise is followed through time and a detailed prediction of deposit weight profiles expected in the tubes is made. The inlet dissolved iron concentration to the boiler is calculated by using a simple flow-assisted corrosion model of outlet feeders. Primary side fouling of the boiler tubes is predicted to be a major contributor to RIHT rise. Coolant pH has a strong effect on flow-assisted corrosion of the outlet feeders and thus on the amount of iron entering the boilers. Deposit weights are predicted accurately for both Gentilly-2 (G-2) and Pickering-1 steam generators using solubility data from Sweeton and Baes at pH 10.3 or 1.4 mg/kg Li. Operation at the low end of the specified pH range seems desirable, e.g. 0.35 mg Li/kg, to reduce the fouling rate of the boiler tubes. Secondary side fouling is predicted to be equally important, but deposit data specific to each plant are needed to assess the precise contribution of this to RIHT rise. A gradual rise in primary side steam quality at the boiler inlet due to RIHT rise will itself generate an RIHT rise simply due to a reduction in boiler area available for sensible heat transfer. Finally, mechanical effects such as divider plate leakage and changes in primary side flow due to pressure tube creep and to increased surface

  16. Cantilever beam test of Zr-2.5Nb pressure tubes with hydride blisters

    International Nuclear Information System (INIS)

    The hydride blisters can be formed by the temperature gradient in the Zr-2.5Nb pressure tube if the pressure tubes contact to the calandria tubes. A volume expansion due to hydride blister causes steep stress gradient in the region of blister-matrix interface, possibly develops to delayed hydride cracking (DHC). After the rupture of pressure tubes due to hydride blisters in Pickering unit 2, many investigations concluded that the probability of blister to DHC may be low because the numerical analysis shows high compressive stresses are developed in the region of blister-matrix interface. This paper investigated fracture behavior of blister and possibility of DHC through cantilever beam test of blistered specimen produced by thermal diffusion processes in laboratory. The fractured surface after cantilever beam test shows a brittle fracture in the region of blister, typical DHC behavior in the region of Zr-2.5Nb matrix, and brittle fracture of crowded circumferential hydrides in the region of blister-matrix interface, where a steep stress gradient is expected

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

    International Nuclear Information System (INIS)

    The complexity of two-phase flow boiling on a tube bundle presents many challenges to the understanding of the physical phenomena taking place. It is important to quantify these numerous heat flow mechanisms in order to better describe the performance of tube bundles as a function of the operational conditions. In the present study, the bundle boiling facility at the Laboratory of Heat and Mass Transfer (LTCM) was modified to obtain high-speed videos to characterise the two-phase regimes and some bubble dynamics of the boiling process. It was then used to measure heat transfer on single tubes and in bundle boiling conditions. Pressure drop measurements were also made during adiabatic and diabatic bundle conditions. New enhanced boiling tubes from Wolverine Tube Inc. (Turbo-B5) and the Wieland-Werke AG (Gewa-B5) were investigated using R134a and R236fa as test fluids. The tests were carried out at saturation temperatures Tsat of 5 °C and 15 °C, mass flow rates from 4 to 35 kg/m2s and heat fluxes from 15 to 70 kW/m2, typical of actual operating conditions. The flow pattern investigation was conducted using visual observations from a borescope inserted in the middle of the bundle. Measurements of the light attenuation of a laser beam through the intertube two-phase flow and local pressure fluctuations with piezo-electric pressure transducers were also taken to further help in characterising the complex flow. Pressure drop measurements and data reduction procedures were revised and used to develop new, improved frictional pressure drop prediction methods for adiabatic and diabatic two-phase conditions. The physical phenomena governing the enhanced tube evaporation process and their effects on the performance of tube bundles were investigated and insight gained. A new method based on a theoretical analysis of thin film evaporation was used to propose a new correlating parameter. A large new database of local heat transfer coefficients were obtained and then utilised

  18. Effect of tube length on measurement of instantaneous pressure drop of column weight in pulsed extraction column

    International Nuclear Information System (INIS)

    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 investigated. 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 pressure 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. (authors)

  19. An advanced CANDU reactor with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    AECL is studying an advanced CANDU reactor concept, with supercritical steam as coolant. The coolant, being a high density gas, at a pressure above 22 MPa and temperatures above 370 deg C, does not encounter the two-phase region with its associated fuel-dryout and flow-instability problems. Increased coolant temperature leads directly to increased plant thermodynamic efficiency, thereby reducing unit energy cost through reduced specific capital cost and reduced fueling cost. The reduced coolant in-core density leads to sufficiently reduced void reactivity, so that light water becomes a coolant option. The use of supercritical water coolant also opens up the possibility of enhanced safety with a natural circulation primary flow, taking advantage of the gas expansion coefficient. To preserve neutron economy, especially at high coolant temperatures, a fuel channel that is currently being developed has a pressure tube that is thermally insulated from high-temperature coolant and is in contact with the cold heavy-water moderator. Two stages of development of a supercritical-cooled CANDU reactor were identified. The first uses conventional or near-conventional zirconium-alloy fuel cladding with coolant core-mean temperatures near 400 deg C, and the second uses advanced high-temperature fuel cladding at coolant core-mean temperatures near 500 deg C. A first-stage cost reduction of 20% from the CANDU 6 design is estimated as a result of improved thermodynamic efficiency. A large change in coolant density across the core leads to a factor 3 or 4 reduction in heavy-water inventory and a corresponding reduction in coolant void reactivity. The latter leads to improved fuel burnup and reduced demands on the safety shutdown systems. (author)

  20. Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Catana, Alexandru [RAAN, Institute for Nuclear Research, Str. Campului Nr. 1, Pitesti, Arges (Romania); Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel [University POLITEHNICA of Bucharest (Romania)

    2008-07-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D{sub 2}O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D{sub 2}O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

  1. CANDU 9 fuelling machine carriage

    International Nuclear Information System (INIS)

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs

  2. Endotracheal tube cuff pressure before, during, and after fixed-wing air medical retrieval.

    Science.gov (United States)

    Brendt, Peter; Schnekenburger, Marc; Paxton, Karen; Brown, Anthony; Mendis, Kumara

    2013-01-01

    Abstract Background. Increased endotracheal tube (ETT) cuff pressure is associated with compromised tracheal mucosal perfusion and injuries. No published data are available for Australia on pressures in the fixed-wing air medical retrieval setting. Objective. After introduction of a cuff pressure manometer (Mallinckrodt, Hennef, Germany) at the Royal Flying Doctor Service (RFDS) Base in Dubbo, New South Wales (NSW), Australia, we assessed the prevalence of increased cuff pressures before, during, and after air medical retrieval. Methods. This was a retrospective audit in 35 ventilated patients during fixed-wing retrievals by the RFDS in NSW, Australia. Explicit chart review of ventilated patients was performed for cuff pressures and changes during medical retrievals with pressurized aircrafts. Pearson correlation was calculated to determine the relation of ascent and ETT cuff pressure change from ground to flight level. Results. The mean (± standard deviation) of the first ETT cuff pressure measurement on the ground was 44 ± 20 cmH2O. Prior to retrieval in 11 patients, the ETT cuff pressure was >30 cmH2O and in 11 patients >50 cmH2O. After ascent to cruising altitude, the cuff pressure was >30 cmH2O in 22 patients and >50 cmH2O in eight patients. The cuff pressure was reduced 1) in 72% of cases prior to take off and 2) in 85% of cases during flight, and 3) after landing, the cuff pressure increased in 85% of cases. The correlation between ascent in cabin altitude and ETT cuff pressure was r = 0.3901, p = 0.0205. Conclusions. The high prevalence of excessive cuff pressures during air medical retrieval can be avoided by the use of cuff pressure manometers. Key words: cuff pressure; air medical retrieval; prehospital. PMID:23252881

  3. Characterization of magnetically impelled arc butt welded T11 tubes for high pressure applications

    Directory of Open Access Journals (Sweden)

    R. Sivasankari

    2015-09-01

    Full Text Available Magnetically impelled arc butt (MIAB welding is a pressure welding process used for joining of pipes and tubes with an external magnetic field affecting arc rotation along the tube circumference. In this work, MIAB welding of low alloy steel (T11 tubes were carried out to study the microstructural changes occurring in thermo-mechanically affected zone (TMAZ. To qualify the process for the welding applications where pressure could be up to 300 bar, the MIAB welds are studied with variations of arc current and arc rotation time. It is found that TMAZ shows higher hardness than that in base metal and displays higher weld tensile strength and ductility due to bainitic transformation. The effect of arc current on the weld interface is also detailed and is found to be defect free at higher values of arc currents. The results reveal that MIAB welded samples exhibits good structural property correlation for high pressure applications with an added benefit of enhanced productivity at lower cost. The study will enable the use of MIAB welding for high pressure applications in power and defence sectors.

  4. CFX analysis of the CANDU moderator thermal-hydraulics in the Stern Lab. Test Facility

    International Nuclear Information System (INIS)

    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. It is shown that the present CFD prediction without the empirical correlation based on the pressure drop test is in good agreement with the test results. The prediction becomes more accurate, as the flow conditions become more turbulent with a higher Reynolds number. However, the temperature fluctuation is observed during iteration steps for a steady-state simulation of the thermal-hydraulic test. This result shows that the flow and temperature distribution inside the moderator tank may not be stable in the actual test

  5. Determination of interaction effect between cracked tube and eggcrate support plate on the burst pressure

    International Nuclear Information System (INIS)

    Abstracts: Steam generator is one of the major components in the nuclear power plant comprising pressure-retaining boundary. Typically, there are thousands of thin-walled tubes as well as several support plates in a steam generator. According to the operating experience of the steam generators, a lot of cracks have been found in the tubes. Therefore, an accurate integrity assessment of the tubes is crucial for maintaining safety as well as reliability of a nuclear power plant. The steam generator tubes are supported by several support plates, and deformations of the tubes are partially restrained depending on the crack location and the gap between the tube and the support plate. In the authors' previous study, it has been reported that burst pressures for circumferentially cracked tubes are significantly affected by the support plate and existing solutions differ from the actual burst pressure. However, this interaction effect for axially cracked tubes has not been fully investigated while those are frequently occurred during operation. In this paper, therefore, a number of elastic-perfectly plastic finite element analyses were performed considering the contact interaction between the tube and the support plate. The burst pressure is then evaluated in accordance with the lower bounding limit theorem and the support-induced interaction effects on the burst pressure were determined

  6. Next Generation CANDU Performance Assurance

    International Nuclear Information System (INIS)

    AECL is developing a next generation CANDU design to meet market requirements for low cost, reliable energy supplies. The primary product development objective is to achieve a capital cost substantially lower than the current nuclear plant costs, such that the next generation plant will be competitive with alternative options for large-scale base-load electricity supply. However, other customer requirements, including safety, low-operating costs and reliable performance, are being addressed as equally important design requirements. The main focus of this paper is to address the development directions that will provide performance assurance. The next generation CANDU is an evolutionary extension of the proven CANDU 6 design. There are eight CANDU 6 units in operation in four countries around the world and further three units are under construction. These units provide a sound basis for projecting highly reliable performance for the next generation CANDU. In addition, the next generation CANDU program includes development and qualification activities that will address the new features and design extensions in the advanced plant. To limit product development risk and to enhance performance assurance, the next generation CANDU design features and performance parameters have been carefully reviewed during the concept development phase and have been deliberately selected so as to be well founded on the existing CANDU knowledge base. Planned research and development activities are required only to provide confirmation of the projected performance within a modest extension of the established database. Necessary qualification tests will be carried out within the time frame of the development program, to establish a proven design prior to the start of a construction project. This development support work coupled with ongoing AECL programs to support and enhance the performance and reliability of the existing CANDU plants will provide sound assurance that the next generation

  7. Disposal costs for advanced CANDU fuel cycles

    International Nuclear Information System (INIS)

    The CANDU reactor can 'burn' a wide range of fuels without modification to the reactor system, including natural uranium, slightly enriched uranium, mixed oxide and spent LWR fuels. The economic feasibility of the advanced fuel cycles requires consideration of their disposal costs. Preliminary cost analyses for the disposal of spent CANDU-SEU (Slightly Enriched Uranium) and CANDU-DUPIC (Direct Use of spent PWR fuel In CANDU) fuels have been performed and compared to the internationally published costs for the direct disposal of spent CANDU and LWR fuels. The analyses show significant economic advantages in the disposal costs of CANDU-SEU and CANDU-DUPIC fuels. (author)

  8. Pressure drop across a tube-bundle flow rectifier. Sekiso koshi no teiko keisu ni kansuru kenkyu

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, T.; Yoshino, F. (Tottori Universtiy, Tottori (Japan). Faculty of Engineering)

    1991-01-25

    The pressure drop coefficient of a tube-bundle flow rectifier made by piling tubes in parallel in the flow direction was obtained with experiments. A tube-bundle rectifier forms optional pressure drop distribution, has rectification effects, and is fit for the formation of the field of the homogeneous flow with large velocity gradient. Tubes may be piled in a staggered or a side-by side arrangement. As a duct, an acrylic pipe of 130 outer diameter, 119 inner diameter and 8,069 mm length was used. A tube-bundle rectifier is a cartridge of integral construction in which polypropylene straws of 6.53 inner diameter and 0.19 mm wall thickness are piled. The ratio of the flow passage sectional area against the whole sectional area is 9.3% in the staggered arrangement and 21.5% in the side-by side arrangement. While the opening ratio is small in the staggered arrangement, the pressure drop coefficient is not necessarily large in this arrangement; the coefficient depends on the tube length and the Reynolds number. In some cases, on the contrary, the pressure drop coefficient is larger in the side-by-side arrangement. It was also indicated that the approximation of the wire mesh equivalent pressure drop coefficient in the extremity, where the tube wall is as thick as the length of the tube-bundle rectifier, can be obtained with the pressure drop coefficient of plain weave wire mesh. 11 refs., 6 figs.

  9. Effect of texture on the anisotropic creep of Zr-2.5Nb tubes

    International Nuclear Information System (INIS)

    Zr-2.5Nb is used as the pressure-tube material in CANDU reactors. Under normal operating conditions, pressure tubes undergo dimensional changes due to the high temperature, stress and fast neutron flux. Thermal creep contributes to the deformation. In this project, the thermal creep at various temperatures, stress states, crystallographic texture and microstructures is being investigated. The research is being carried out on Zr-2.5Nb tubes with basal plane normals concentrated in the radial, axial and transverse directions, respectively. In this paper we report the results of scoping studies to determine the temperature and stress dependence of the diametral creep of tubes with a predominantly radial texture and the preparation of 'micro-pressure tubes' with a variety of textures suitable for investigating creep anisotropy. (author)

  10. High temperature deformation and burst behavior of internally pressurized zircaloy-4 tubes

    International Nuclear Information System (INIS)

    To investigate ballooning and burst behavior of the fuel claddings under hypothetical LWR loss-of-coolant accident conditions, internally pressurized Zircaloy-4 tubes were heated to burst and their deformation progresses were observed. Two methods of heating were used, i.e. direct and internal heating. In the former, current flows through the wall, and in the latter, through a coaxial insulated electric heater inside the tube. Followings are the results: (1) Circumferential rupture strain depends strongly on the burst temperature, a maximum at about 8200C and a minimum at about 9200C. (2) The direct heating method generally produces larger deformation than the indirect, i.e. internal heating method. (3) Due possibly to azimuthal temperature difference and/or wall thickness variation, the circumferential rupture strains fluctuate in the internally heated tubes. (4) The appearances of burst tubes vary with their burst temperatures in such a way as it be almost common to the tubes in the both heating methods. These results should serve for the analysis and prediction of fuel assembly deformation behavior. (author)

  11. Experimental investigation of pressure fluctuations caused by a vortex rope in a draft tube

    International Nuclear Information System (INIS)

    In the last years hydro power plants have taken the task of power-frequency control for the electrical grid. Therefore turbines in storage hydro power plants often operate outside their optimum. If Francis-turbines and pump-turbines operate at off-design conditions, a vortex rope in the draft tube can develop. The vortex rope can cause pressure oscillations. In addition to low frequencies caused by the rotation of the vortex rope and the harmonics of these frequencies, pressure fluctuations with higher frequencies can be observed in some operating points too. In this experimental investigation the flow structure and behavior of the vortex rope movement in the draft tube of a model pump-turbine are analyzed. The investigation focuses on the correlation of the pressure fluctuation frequency measured at the draft tube wall with the movement of the vortex rope. The movement of the vortex rope is analyzed by the velocity field in the draft tube which was measured with particle image velocimetry. Additionally, the vortex rope movement has been analyzed with the captures of high-speed-movies from the cavitating vortex rope. Besides the rotation of the vortex rope due to pressure fluctuation with low frequencies the results of the measurement also show a correlation between the rotation of the elliptical or deformed rope cross-section and the higher frequency pressure pulsation. An approximation shows that the frequencies of the pressure fluctuation and the movement of the vortex rope are also connected with the velocity of the flow. Taking into account the size and position of the cavitating vortex core as well as the velocity at the position of the surface of the cavitating vortex core the time-period of the rotation of the vortex core can be approximated. The results show that both, the low frequency pressure fluctuation and the higher frequency pressure fluctuation are correlating with the vortex rope movement. With this estimation, the period of the higher frequency

  12. The CANDU 6

    International Nuclear Information System (INIS)

    The CANDU 6 is a modem nuclear power plant designed and developed under the aegis of Atomic Energy of Canada, Limited (AECL) for domestic use and for export to other countries. This design has successfully met criteria for operation and redundant safety features required by Canada and by the International Atomic Energy Agency (IAEA) and has an estimable record of performance in all applications to date. Key to this success is a defined program of design enhancement in which changes are made while retaining fundamental features proven by operating experience. Basic design features and progress toward improvements are presented here. (author)

  13. CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU-PHW program is based upon 38 years of heavy water reactor experience with 35 years of operating experience. Canada has had 72 reactor years of nuclear-electric operations experience with 10 nuclear units in 4 generating stations during a period of 18 years. All objectives have been met with outstanding performance: worker safety, public safety, environmental emissions, reliable electricity production, and low electricity cost. The achievement has been realized through total teamwork involving all scientific disciplines and all project functions (research, design, manufacturing, construction, and operation). (auth)

  14. 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.

  15. New inverted hydride fuel design concept for pressure tube type super critical water reactors

    International Nuclear Information System (INIS)

    In this study, an innovative core design having inverted configuration has been proposed for pressure tube type supercritical water reactors. In this design the relative positions of fuel and coolant have been inverted and U-Th-Zr-hydride fuel has been used. A coupled neutronics and thermal hydraulics analysis was done for the proposed Inverted Pressure Tube Type (IPTT) SCWR. The neutronics analysis was carried out by using a 3D fine mesh diffusion theory code and thermal hydraulics calculations were done by using single channel model. These two codes were coupled with each other by a link code. The average outlet temperature for the proposed IPTT-SCWR was found to be 625degC with maximum clad surface temperature (MCST) under the design limits i.e. below 850degC. Moreover a core loading pattern has also been proposed to achieve uniform radial power distribution and lower cladding surface temperature. (author)

  16. Formation and growth of hydride blisters in Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    Hydride blisters were formed on the outer surface of Zr-2.5Nb pressure tube by a nonuniform steady thermal diffusion process. A thermal gradient was applied to the pressure tube with a heat bath kept at a temperature of 415 .deg. C and an aluminum cold finger cooled with flowing water of 15 .deg. C. Optical microscopy and three-dimensional laser profilometry were used to characterize the hydride blisters with different hydrogen concentrations and thermal diffusion time. Hydride blisters were expected to start at a hydrogen concentration of 30 - 70 ppm and a thermal diffusion time of 4 - 6x105 sec. The hydride blister size increases with higher hydrogen concentrations and longer thermal diffusion time. Some of the samples revealed cracks on the hydride blisters. The ratio of hydride blister depth to height was estimated as approximately 8:1

  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

    Gaseous hydrogen as an automotive fuel is reaching the point of commercial introduction. Development of hydrogen fuelling stations considering an acceptable fuelling time by cooling the hydrogen to -40 C has started. This paper presents a design study of coaxial tube ammonia evaporators for three...... different concepts of hydrogen cooling, one onestage and two two-stage processes. An exergoeconomic optimization is imposed to all three concepts to minimize the total cost. A numerical heat transfer model is developed in Engineer Equation Solver, using heat transfer and pressure drop correlations from the...... open literature. With this model the optimal choice of tube sizes and circuit numbers are found for all three concepts. The results show that cooling with a two-stage evaporator after the pressure eduction valve yields the lowest total cost, 45 % lower than the highest, which is with a one...

  18. A general computing code devoted to the analysis of bending vibrations specific to the CANDU type fuel channel

    International Nuclear Information System (INIS)

    It is known that circulation of the coolant through the pressure tube of a CANDU type reactor initiates and maintains bending vibrations in: individual fuel elements, fuel cluster, cluster column and in the pressure tube. The driving forces are either aleatory, due to turbulent flow, or harmonical due to the pressure pulsations from the circulation pumps. The vibrations induced by laminar flow in case of excessive intensities may induce both a acceleration of the fretting wear phenomena in the fuel elements and pressure tubes and a premature aging of the latter. In these conditions an important problem in the cluster design is that of obtaining, based on knowledge of laminar flow frequency structure, the eigenfrequencies for the four categories of oscillatory systems mentioned above and thus to avoid by construction the resonance phenomenon or at least to diminish its impairing effects. An activity of comparative analysis in different fuel cluster types is underway at INR Pitesti, a special attention being of course directed toward their vibrational behavior. The paper presents a general computational code devoted to characterization of bending vibration for: individual fuel elements, fuel element cluster, pressure tube loaded or not with fuel clusters and filled or not with coolant; fuel channel. During the presentation of the work the computing code will be run for demonstration

  19. Detecting Nonlinearity in Pressure Data Inside the Draft Tube of a Real Francis Turbine

    OpenAIRE

    Sello, S.

    1995-01-01

    A general method for testing nonlinearity in time series is described and applied to measurements of different pressure data inside the draft tube surge of a real Francis turbine. Comparing the current original time series to an ensemble of surrogates time series, suitably constructed to mimic the linear properties of the original one, we was able to distinguish a linear stochastic from a nonlinear deterministic behaviour and, moreover, to quantify the degree of nonlinearity present in the re...

  20. Exact Solution of a Cylinder Tube Made of Metallic Foam Under Inner Pressure

    Institute of Scientific and Technical Information of China (English)

    ZHU Ai-yu; FAN Tian-you

    2008-01-01

    Exact solution of the stress and velocity fields of a cylinder tube of metallic foams under inner Pressure is given in which the Triantafillou and Gibson constitutive law(TG model)for the material is taken as a basis of the calculation.The nonlinear equation is turned linear equation by introducing a kinematics parameter.The differences between the full condensed materials and the effect of the relative densitv are also discussed.

  1. The formation and characteristics of hydride blisters in c.w. Zircaloy-2 pressure tubes

    International Nuclear Information System (INIS)

    Under the auspices of the IAEA, a consultants' meeting was arranged in Vienna, 1994 July 25-29, at which a Canadian delegation, consisting of AECL and Ontario Hydro Technologies personnel, presented information on their knowledge of the behaviour of hydride blisters in Zircaloy-2 pressure tubes. This document contains the 10 papers presented by the Canadian delegation to the meeting. It is believed that they represent a good reference document on hydride blister phenomena

  2. Some characteristics of the digitization pulses from high pressure neon-helium flash tubes

    International Nuclear Information System (INIS)

    Characteristics of the digitization output pulses from high pressure neon-helium flash tubes were studied under various operation conditions using square ultra-high voltage pulses. Properties reported by previous workers were compared. Two discharge mechanisms, the Townsend avalanche discharge and the streamer discharge, were observed to occur in sequence in some events. The output waveforms for both discharge mechanisms were studied in detail. The charge induced on a detecting probe was also estimated from the measured data. (Auth.)

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

    International Nuclear Information System (INIS)

    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.

  4. Break flow modeling for a steam generator tube rupture (SGTR) incident in a pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    The design-basis steam generator tube rupture (SGTR) scenario for the pressurized water reactor (PWR) postulates an instantaneous double-ended break of a steam generator (SG) U-tube. The flow rate through the broken U-tube depends on the primary-to-secondary side differential pressure in the affected SG, the primary coolant subcooling, and the break location along the U-tube. In this report, the RELAP5/MOD2 code's capability in predicting the SGTR break flow rate is assessed against experiments conducted on the Large Scale Test Facility (LSTF). The code is then used to predict break flow rate in the PWR for typical SGTR situations. It is shown that the code simulates well the break flow rates in the LSTF experiments for both single-phase and two-phase discharges, including two-phase critical flow discharge. The calculated PWR break flow rate takes a maximum for a break occurring at the lower end of the U-tube, on its cold leg side, because of the combined influence of tube-inlet fluid subcooling and frictional pressure drop along the broken tube. Modeling the tube frictional pressure drop is important to predict the break flow rate dependence on inlet fluid sub-cooling; simplified break flow modeling which applies a constant discharge coefficient less than unity, instead of modeling explicitly the tube frictional length, fails to predict the change in break flow rate accurately if the inlet subcooling varies for a wide range. (author)

  5. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

    construction costs through more efficient work planning and use of materials, through reduced re-work and through more precise configuration management. Full-scale exploitation of AECL's electronic engineering and project management tools enables further reductions in cost. The Candu fuel-channel reactor type offers inherent manufacturing and construction advantages through the application of a simple, low-pressure low-temperature reactor vessel along with modular fuel channel technology. This leads to cost benefits and total project schedule benefits. As a result, the targets which AECL has set for replication units - overnight capital cost of $1000 US/kW and total project schedule (engineering/manufacturing/construction/commissioning) of 48 months, have been shown to be achievable for the reference NG Candu design. (authors)

  6. Very High Pressure Single Pulse Shock Tube Studies of Aromatic Species

    Energy Technology Data Exchange (ETDEWEB)

    Brezinsky, K.

    2006-11-28

    The principal focus of this research program is aimed at understanding the oxidation and pyrolysis chemistry of primary aromatic molecules and radicals with the goal of developing a comprehensive kinetic model at conditions that are relevant to practical combustion devices. A very high pressure single pulse shock tube is used to obtain experimental data over a wide pressure range in the high pressure regime, 5-1000 bars, at pre-flame temperatures for fuel pyrolysis and oxidation over a broad spectrum of equivalence ratios. Stable species sampled from the shock tube are analyzed using standard chromatographic techniques using GC/MS-PDD and GC/TCD-FID. Experimental data from the HPST (stable species profiles) and data from other laboratories (if available) are simulated using kinetic models (if available) to develop a comprehensive model that can describe aromatics oxidation and pyrolysis over a wide range of experimental conditions. The shock tube has been heated (1000C) recently to minimize effects due to condensation of aromatic, polycyclic and other heavy species. Work during this grant period has focused on 7 main areas summarized in the final technical report.

  7. Detection of hydride blister in PHWR pressure tubes using ultrasonic velocity ratio method

    International Nuclear Information System (INIS)

    When the pressure tubes(PT) contact to the calandria tube(CT) in the pressurized heavy water reactor(PHWR), the temperature difference between inner and outer wall of PT results in a thermal diffusion of hydrogen (deuterium) and hydride blisters are formed on the outer surface of PT. Because the hydride blisters are acoustically continued to zirconium matrix, it is not easy to detect the blisters with conventional ultrasonic method. An ultrasonic velocity ratio method was developed to detect small hydride blisters on the zirconium pressure tube. Hydride blisters were grown in the PT specimen with a steady state thermal diffusion device. Ultrasonic velocity ratio method were developed for detection of hydride blisters. The flight time of longitudinal echo and reflected shear echo from the outer surface were measured and calculated to the parameter of velocity ratio of longitudinal wave to shear wave. The velocity ratio was plotted to modified c-scan display and converted to contour plot. The plots shows the capability that the blisters could be detected as well as imaged the shapes.

  8. Study of hydride blisters grown on Zr-2.5Nb pressure tube spool piece under simulated condition of in-reactor pressure and temperature

    International Nuclear Information System (INIS)

    Indian Pressurised Heavy Water Reactor (PHWR) have pressure tubes, made from zirconium alloy. These pressure tubes undergo corrosion with the high temperature (300 deg C) heavy water coolant under the reactor environment and pick up a part of hydrogen generated as result of this corrosion reaction. This hydrogen affects the integrity of pressure tubes in many ways; nucleation and growth of hydride blisters being one of them. The present study has been carried out to understand the mechanisms of nucleation and growth of hydride blisters and their effect on the serviceability of the component in the reactor environment. (author)

  9. Draft tube pressure pulsation predictions in Francis turbines with transient Computational Fluid Dynamics methodology

    International Nuclear Information System (INIS)

    An automatic Computational Fluid Dynamics (CFD) procedure that aims at predicting Draft Tube Pressure Pulsations (DTPP) at part load is presented. After a brief review of the physics involved, a description of the transient numerical setup is given. Next, the paper describes a post processing technique, namely the separation of pressure signals into synchronous, asynchronous and random pulsations. Combining the CFD calculation with the post-processing technique allows the quantification of the potential excitation of the mechanical system during the design phase. Consequently it provides the hydraulic designer with a tool to specifically target DTPP and thus helps in the development of more robust designs for part load operation of turbines

  10. Draft tube pressure pulsation predictions in Francis turbines with transient Computational Fluid Dynamics methodology

    Science.gov (United States)

    Melot, M.; Nennemann, B.; Désy, N.

    2014-03-01

    An automatic Computational Fluid Dynamics (CFD) procedure that aims at predicting Draft Tube Pressure Pulsations (DTPP) at part load is presented. After a brief review of the physics involved, a description of the transient numerical setup is given. Next, the paper describes a post processing technique, namely the separation of pressure signals into synchronous, asynchronous and random pulsations. Combining the CFD calculation with the post-processing technique allows the quantification of the potential excitation of the mechanical system during the design phase. Consequently it provides the hydraulic designer with a tool to specifically target DTPP and thus helps in the development of more robust designs for part load operation of turbines.

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

    International Nuclear Information System (INIS)

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

  12. CANDU-BLW-250

    International Nuclear Information System (INIS)

    The plant 'La Centrale nucleaire de Gentilly' is located between Montreal and Quebec City on the south shore of the St. Lawrence River and start-up is scheduled for 1971. A CANDU-BLW reactor is the nuclear steam generator. his reactor utilizes a heavy water moderator, natural uranium oxide fuel, and a boiling light water coolant. To be economic, this type of plant must have a minimum light water inventory in the reactor core. A minimum inventory is obtained (a) by reducing the cross-sectional area for coolant flow to a minimum, and (b) by operating at a low-coolant density. In CANDU-BLW-250, this is accomplished by operating a closed spaced fuel rod bundle at high steam quality. These features and others in the BLW concept lead to a number of areas of concern and they are summarized below: (1) Heat Transfer: It is intended that under normal operating conditions the fuel sheaths will always be wetted with coolant. (ii) Hydrodynamic Stability: Experiments and analysis indicate that the plant has a considerable over-power capacity before instability is predicted. (iii) Control: This plant does have a positive power coefficient and the transient performance with various disturbances are detailed. (iv) Safety: The positive power coefficient leads to concern over the loss of coolant accident. The results of some accident analysis are presented. (author)

  13. Passive safety features for next generation CANDU power plants

    International Nuclear Information System (INIS)

    CANDU offers an evolutionary approach to simpler and safer reactors. The CANDU 3, an advanced CANDU, currently in the detailed design stage, offers significant improvements in the areas of safety, design simplicity, constructibility, operability, maintainability, schedule and cost. These are being accomplished by retaining all of the well known CANDU benefits, and by relying on the use of proven components and technologies. A major safety benefit of CANDU is the moderator system which is separate from the coolant. The presence of a cold moderator reduces the consequences arising from a LOCA or loss of heat sink event. In existing CANDU plants even the severe accident - LOCA with failure of the emergency core cooling system - is a design basis event. Further advances toward a simpler and more passively safe reactor will be made using the same evolutionary approach. Building on the strength of the moderator system to mitigate against severe accidents, a passive moderator cooling system, depending only on the law of gravity to perform its function, will be the next step of development. AECL is currently investigating a number of other features that could be incorporated in future evolutionary CANDU designs to enhance protection against accidents, and to limit off-site consequences to an acceptable level, for even the worst event. The additional features being investigated include passive decay heat removal from the heat transport system, a simpler emergency core cooling system and a containment pressure suppression/venting capability for beyond design basis events. Central to these passive decay heat removal schemes is the availability of a short-term heat sink to provide a decay heat removal capability of at least three days, without any station services. Preliminary results from these investigations confirm the feasibility of these schemes. (author)

  14. Thorium utilization in ACR (Advanced CANDU) and CANDU-6 reactors

    International Nuclear Information System (INIS)

    It is the main objective of this study to investigate fuel composition options for CANDU type of reactors that are capable of using a mixture of U-Th as fuel. A homogenous mixture of (U-Th)O2 was used in all elements of fuel bundles. The core of CANDU-6 and ACR (Advanced CANDU) were modeled using MCNP5. In equilibrium core, using MONTEBURNS2 code (coupled with MCNP5 and ORIGENS) for once-through uranium and once-through uranium-thorium fuel cycle of CANDU-6 and ACR, discharge burnups and spent fuel compositions were computed. For various enrichments of uranium and different fractions of thorium in a uranium-thorium fuel mixture, performing burnup calculations, relevant relations were derived; in addition, conversion ratio, fuel requirement, uranium resource utilization, and natural uranium savings were determined, and their changes with burnup were observed. Appropriate fuel compositions were discussed.

  15. Tracheal Rupture due to Diffusion of Nitrous Oxide to Cuff of High-Volume, Low-Pressure Intubation Tube

    OpenAIRE

    Atalay, Canan; AYKAN, Şeyda; CAN, Abdullah; Doğan, Nazım

    2009-01-01

    Tracheal rupture is a rare complication of endotracheal intubation. Risk factors include short neck, repeated attempts due to failed intubation, inappropriate stylus, over-inflation of the cuff, poor positioning of the tube, inappropriate tube size, weakened membrane structure due to steroid use, chronic obstructive pulmonary disease, tracheomalacia, kyphosis, and use of nitric oxide during the operation. In this article, we suggest that high-volume, low-pressure tubes may not always provide ...

  16. Experimental study of heat transfer and pressure drop characteristics on shell-side of pin-fin tube oil cooler

    International Nuclear Information System (INIS)

    The comparative experimental study for one smooth tube oil cooler and three pin-fin tube oil coolers was performed by using lubricating oil as heat transfer medium. The experimental results indicate that in the range of experimental study, total heat transfer coefficient of pin-fin tube oil coolers is about 1.4-2 times higher than that of the smooth tube oil cooler. The heat transfer and pressure drop characteristics are greatly different for different structures of pin-fin tube oil coolers. The effects of the structure of pin-fin tube and shell-side flow path number are dominant to influence heat transfer and pressure drop characteristics of oil coolers. In the range of experimental study, large pin-fin height is conducive to the oil flow disturbance, but not conducive to the heat transfer on the tube-base heat transfer surface of pin-fin tube; single-pass pin-fin tube oil cooler offers high total heat transfer coefficient and volumetric heat transfer capacity, the global heat transfer performance and the friction characteristics are better than that of two-pass pin-fin tube oil cooler. (authors)

  17. Deformation behaviour of Zr-2.5 wt percent Nb pressure tubes under internal pressurization and ramp temperature conditions

    International Nuclear Information System (INIS)

    The deformation behaviour of 0.6 m long Zr-2.5 wt percent Nb pressure tube specimens deformed under internal pressurization and ramp temperature conditions has been studied. Both inert gas and steam have been used as the pressurizing medium. Controlled heating rates in the range 1-25 K/s and internal pressures in the range 0-12 MPa have been investigated. It was observed that, except at the very high test pressures, the samples deformed to circumferential strains well in excess of the nominal 'contact' strain of 0.18 before possible rupture. Over the time scale of the biaxial tests carried out, the effect of steam on deformation behaviour appeared to be minimal. It was also found that the major features of the biaxial deformation behaviour can be investigated, at least qualitatively, by carrying out uniaxial tensile tests. However, detailed analysis of uniaxial and biaxial data showed that there were several major differences between the two types of data. These differences are discussed in considerable detail in the present report

  18. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

  19. Health monitoring requirements for CANDU steam generators

    International Nuclear Information System (INIS)

    AECL is developing an equipment health monitoring (EHM) module as part of SMART CANDU development to provide station maintenance personnel with the information required to assess the condition of critical station equipment and to predict when maintenance is required. SMART CANDU is a suite of software applications that is being developed by AECL to help station staff to efficiently and effectively implement their health monitoring and ageing management programs. The EHM application integrates information from all relevant station sources (e.g., on-line instruments, local 'smart' field components, walk-down data, and inspection and monitoring software) on the station local area network and presents the user with a snapshot of the current health of the component of interest. The EHM application also permits staff to be proactive by alerting them to the early warning signs of degraded equipment functionality and to potential equipment failure. User requirements for a steam generator EHM display described in the present work are being developed in consultation with station staff and subject matter experts. The required information is being collected from multiple sources and may include, for example, on-line and grab sample process and chemistry data, inspection results from eddy current and ultrasonic measurements and maintenance information (e.g., chemical cleaning, plugged tubes). These data can be combined with AECL predictive models for steam generator fouling, crevice chemistry, flow induced vibration, etc. to provide station staff with an assessment of current steam generator conditions and help mitigate key degradation mechanisms throughout the life of the CANDU station. (author)

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

    International Nuclear Information System (INIS)

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

  1. Vertical laryngeal position and oral pressure variations during resonance tube phonation in water and in air. A pilot study.

    Science.gov (United States)

    Wistbacka, Greta; Sundberg, Johan; Simberg, Susanna

    2016-10-01

    Resonance tube phonation in water (RTPW) is commonly used in voice therapy, particularly in Finland and Sweden. The method is believed to induce a lowering of the vertical laryngeal position (VLP) in phonation as well as variations of the oral pressure, possibly inducing a massage effect. This pilot study presents an attempt to measure VLP and oral pressure in two subjects during RTPW and during phonation with the free tube end in air. VLP is recorded by means of a dual-channel electroglottograph. RTPW was found to lower VLP in the subjects, while it increased during phonation with the tube end in air. RTPW caused an oral pressure modulation with a bubble frequency of 14-22 Hz, depending mainly on the depth of the tube end under the water surface. The results indicate that RTPW lowers the VLP instantly and creates oral pressure variations. PMID:26033381

  2. CANDU 9 enhancements and licensing

    International Nuclear Information System (INIS)

    The CANDU 9 design has followed the evolutionary product development approach that has characterized the CANDU family of nuclear power plants. In addition to utilizing proven equipment and systems from operating stations, the CANDU 9 design has looked ahead to incorporate design and safety enhancements necessary to meet evolving utility and licensing requirements. With the requirement that the CANDU 9 design should be licensable for both domestic and foreign potential users, the pre-project Basic Design Engineering program included a two year formal extensive review by the Canadian Regulatory Agency, the Atomic Energy Control Board . Documentation submitted for the licensing review included the licensing basis, safety requirements and safety analyses necessary to demonstrate compliance with regulations and to assess system design and performance

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

    International Nuclear Information System (INIS)

    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)

  4. 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.

  5. In vitro estimation of pressure drop across tracheal tubes during high-frequency percussive ventilation

    International Nuclear Information System (INIS)

    Tracheal tubes (TT) are used in clinical practice to connect an artificial ventilator to the patient's airways. It is important to know the pressure used to overcome tube impedance to avoid lung injury. Although high-frequency percussive ventilation (HFPV) has been increasingly used, the mechanical behavior of TT under HFPV has not yet been described. Thus, we aimed at characterizing in vitro the pressure drop across TT (ΔPTT) by identifying the model that best fits the measured pressure–flow (P– V-dot ) relationships during HFPV under different working pressures (PWork), percussive frequencies and mechanical loads. Three simple models relating ΔPTT and flow ( V-dot ) were tested. Model 1 is characterized by linear resistive [Rtube ⋅  V-dot (t)] and inertial [I ⋅ V¨(t)] terms. Model 2 takes into consideration Rohrer's approach [K1 ⋅  V-dot (t) + K2 ⋅  V-dot 2(t)] and inertance [I ⋅ V¨(t)]. In model 3 the pressure drop caused by friction is represented by the non-linear Blasius component [Kb ⋅  V-dot 1.75(t)] and the inertial term [I ⋅ V¨(t)]. Model 1 presented a significantly higher root mean square error of approximation than models 2 and 3, which were similar. Thus, model 1 was not as accurate as the latter, possibly due to turbulence. Model 3 presented the most robust resistance-related coefficient. Estimated inertances did not vary among the models using the same tube. In conclusion, in HFPV ΔPTT can be easily calculated by the physician using model 3. (paper)

  6. Development and Validation of the 3-D Computational Fluid Dynamics Model for CANDU-6 Moderator Temperature Predictions

    International Nuclear Information System (INIS)

    A computational fluid dynamics (CFD) model for predicting the moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the Calandria tubes. The buoyancy effect induced by internal heating is accounted for by Boussinesq approximation. The standard k-[curly epsilon] turbulence model associated with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the Calandria tubes in the core region is simplified to porous media, in which anisotropic hydraulic impedance is modeled using an empirical correlation of the frictional pressure loss. The governing equations are solved by CFX-4.4, a commercial CFD code developed by AEA Technology. The CFD model has been successfully verified and validated against experimental data obtained at Stern Laboratories Inc. in Hamilton, Ontario, Canada

  7. CANDU in Korea, present and future - Operating performance of Wolsong 1 and construction of Wolsong 2, 3 and 4

    International Nuclear Information System (INIS)

    The excellent performance of Wolsong-1 reactor, attributable to the excellence of CANDU features and facilities, and to traditional Korean diligence and dedication, encouraged the purchase of Units 2, 3, and 4. Units 2 and 3 were under construction at the time of the conference, while Unit 4 was about to be started; some information on contracts and scheduling is provided. Modifications to the design of Wolsong-1 included an additional separator in the steam generators, and conversion of the raw water service system and the recirculated cooling water system to dual loop design. Concerns with regard to maintenance of Wolsong-1 included pressure tube integrity, delayed neutron tube maintenance, and inadequate supply of spare parts and system components (particularly for the plant control computer and the fuelling machines). The expectation was that these problems would be remedied in time for the later units. 9 tabs., 2 figs

  8. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    CANDU has a tradition of incorporating passive systems and passive components whenever they are shown to offer performance that is equal to or better than that of active systems, and to be economic. Examples include the two independent shutdown systems that employ gravity and stored energy respectively, the dousing subsystem of the CANDU 6 containment system, and the ability of the moderator to cool the fuel in the event that all coolant is lost from the fuel channels. CANDU 9 continues this tradition, incorporating a reserve water system (RWS) that increases the inventory of water in the reactor building and profiles a passive source of makeup water and/or heat sinks to various key process systems. The key component of the CANDU 9 reserve water system is a large (2500 cubic metres) water tank located at a high elevation in the reactor building. The reserve water system, while incorporating the recovery system functions, and the non-dousing functions of the dousing tank in CANDU 6, embraces other key systems to significantly extend the passive makeup/heat sink capability. The capabilities of the reserve water system include makeup to the steam generators secondary side if all other sources of water are lost; makeup to the heat transport system in the event of a leak in excess of the D2O makeup system capability; makeup to the moderator in the event of a moderator leak when the moderator heat sink is required; makeup to the emergency core cooling (ECC) system to assure NPSH to the ECC pumps during a loss of coolant accident (LOCA), and provision of a passive heat sink for the shield cooling system. Other passive designs are now being developed by AECL. These will be incorporated in future CANDU plants when their performance has been fully proven. This paper reviews the passive heat removal systems and features of current CANDU plants and the CANDU 9, and briefly reviews some of the passive heat removal concepts now being developed. (author)

  9. Technology transfer in CANDU marketing

    International Nuclear Information System (INIS)

    The author discusses how the CANDU system lends itself to technology transfer, the scope of CANDU technology transfer, and the benefits and problems associated with technology transfer. The establishment of joint ventures between supplier and client nations offers benefits to both parties. Canada can offer varying technology transfer packages, each tailored to a client nation's needs and capabilities. Such a package could include all the hardware and software necessary to develop a self-sufficient nuclear infrastructure in the client nation

  10. Effect of superficial velocity on vaporization pressure drop with propane in horizontal circular tube

    Science.gov (United States)

    Novianto, S.; Pamitran, A. S.; Nasruddin, Alhamid, M. I.

    2016-06-01

    Due to its friendly effect on the environment, natural refrigerants could be the best alternative refrigerant to replace conventional refrigerants. The present study was devoted to the effect of superficial velocity on vaporization pressure drop with propane in a horizontal circular tube with an inner diameter of 7.6 mm. The experiments were conditioned with 4 to 10 °C for saturation temperature, 9 to 20 kW/m2 for heat flux, and 250 to 380 kg/m2s for mass flux. It is shown here that increased heat flux may result in increasing vapor superficial velocity, and then increasing pressure drop. The present experimental results were evaluated with some existing correlations of pressure drop. The best prediction was evaluated by Lockhart-Martinelli (1949) with MARD 25.7%. In order to observe the experimental flow pattern, the present results were also mapped on the Wang flow pattern map.

  11. Nonlinear vacuum gas flow through a short tube due to pressure and temperature gradients

    International Nuclear Information System (INIS)

    The flow of a rarefied gas through a tube due to both pressure and temperature gradients has been studied numerically. The main objective is to investigate the performance of a mechanical vacuum pump operating at low temperatures in order to increase the pumped mass flow rate. This type of pump is under development at CEA-Grenoble. The flow is modelled by the Shakhov kinetic model equation, which is solved by the discrete velocity method. Results are presented for certain geometry and flow parameters. Since according to the pump design the temperature driven flow is in the opposite direction than the main pressure driven flow, it has been found that for the operating pressure range studied here the net mass flow rate through the pump may be significantly reduced

  12. Crystallographic Phases, Texture and Dislocation Densities of ZR2.5%NB Pressure Tubes at Different Stages of Manufacturing

    International Nuclear Information System (INIS)

    Neutron diffraction experiments have been performed on specimens produced from Zr2.5%Nb pressure tubes, in order to characterize the crystallographic phases, texture and dislocation densities at different stages of a new manufacturing schedule developed in Argentina. Experiments were performed on ENGIN-X a time-of-flight neutron strain scanner at the Isis Facility, UK, using an optimized measurement strategy that exploits the well-known crystallographic texture of the pressure tubes. (author)

  13. Is sealing cuff pressure, easy, reliable and safe technique for endotracheal tube cuff inflation?: A comparative study

    OpenAIRE

    Al-metwalli, Roshdi R.; Abdulmohsen A Al-Ghamdi; Mowafi, Hany A.; Sayed Sadek; Mohammed Abdulshafi; Mousa, Wesam F.

    2011-01-01

    Objective: To compare the three common methods of endotracheal tube cuff inflation (sealing pressure, precise standard pressure or finger estimation) regarding the effective tracheal seal and the incidence of post-intubation airway complications. Methods: Seventy-five adult patients scheduled for N 2 O free general anesthesia were enrolled in this study. After induction of anesthesia, endotracheal tubes size 7.5 mm for female and 8.0 mm for male were used. Patients were randomly assigned into...

  14. Application of Deformable Templates for Recognizing Tracks Detected with High Pressure Drift Tubes

    International Nuclear Information System (INIS)

    The modification of the deformable template method (DTM) application to the problem of track finding and track parameter determination for data detected with high pressure drift tubes (HPDT) in the design of ATLAS for the muon spectrometer experiment is proposed. Our DTM applications consist of two parts, according to two stages of the study. The first part relates to the stage of HPDT study on the CERN muon beam (BEAM-TEST) with the simplest one-prong events without noise signals, where the main obstacle is the left-right ambiguities for each tube. In the second part more complicated HPDT data are to be handled with noise signals. It was shown that the suggested DTM development solves the problem of track recognition and track parameter determination for both noiseless and noise data. Results are obtained on the real beam test data and on data simulating the muon spectrometer on the basis of HPDT. 14 refs., 10 figs

  15. The CANDU 9 fuel transfer system

    International Nuclear Information System (INIS)

    The CANDU 9 fuel transfer system is based on the CANDU 6 and the Ontario Hydro Darlington NGD designs, modified to suit the CANDU 9 requirements. The CANDU 9 new fuel transfer system is very similar to the CANDU 6, with modifications to allow new fuel loading from outside containment, similar to Darlington. The CANDU 9 irradiated fuel transfer system is based on the Darlington irradiated fuel transfer system, with modifications to meet the more stringent containment requirements, improve performance, and match station layout. (author). 2 refs., 6 figs

  16. Comparative results for benchmark test problems in CANDU lattices

    International Nuclear Information System (INIS)

    The paper presents comparative results for the main types of lattice cell calculation performed by using the available versions of WIMS and DRAGON codes: WIMSD5B and DRAGON 3.05. The lattice cell calculations main goal is to obtain the optimal input parameters combination that gives closer results to IAEA measurements. The comparison was made for IAEA benchmark problems applied to CANDU lattices. IAEA nuclear data libraries updated in WLUP (WIMS Libraries Updates Project) project were used. The input data have been set from test problems description. The comparisons have been performed for the following heavy water cell configuration: square lattice, natural uranium (NU) 37 fuel rods bundle with 0.72% enrichment in U-235, 28.58 cm lattice pitch, 0.5965 cm for central rod radius and 0.6050 cm for the other fuel rods, Zircaloy-4 as cladding material and 1050 Al alloy for the pressure and calandria tubes, respectively. Lattice calculations were effectuated for observing which combination gave the closer results to IAEA measurements. (author)

  17. Improved CANDU fuel performance

    International Nuclear Information System (INIS)

    The fuel defect rate in CANDU power reactors has been very low (0.06 percent) since 1972. Most defects were caused by power ramping. The two measures taken to reduce the defect rate, by about an order of magnitude, were changes in the fuelling schemes and the introduction of thin coatings of graphite on the inside surface of the Zircaloy fuel cladding. Power ramping tests have demonstrated that graphite layers, and also baked poly-dimethyl-siloxane layers, between the UO2 pellets and Zircaloy cladding increase the tolerance of fuel to power ramps. These designs are termed graphite CANLUB and siloxane CANLUB; fuel performance depends on coating parameters such as thickness and wear resistance and on environmental and thermal conditions during the curing of coatings. (author)

  18. Corrosion of the tube-tubesheet joint in the Candu steam generator at operation resuming after the chemical cleaning of depositions occurred due to the contamination of condense by the cooling water

    International Nuclear Information System (INIS)

    The purpose of this paper is to establish the corrosion behaviour of the Iy-800 (material of the tube) and SA 508 cl.2 (material of the tube sheet) after the chemical cleaning of the steam generator on the secondary circuit specific components. On the surface of these materials there are initial deposits consisting of oxides and corrosion products. The chemical cleaning was made at temperatures between 60-80oC. The substances included in the chemical cleaning solution are citric acid, Na2EDTA, HEDTA, ammonium citrate, hydrazine hydrate and thiourea. After the chemical cleaning the samples were passivated and corrosion-tested at the typical parameters of the steam generator secondary circuit (demineralized water, pH=9.5, all volatile treatment, temperature 265oC and pressure 5.1 MPa). The results of metallographic and X-ray diffraction examinations are presented as well as the reasons for selecting the best solution for the chemical cleaning. (author)

  19. The CANDU experience in Romania

    International Nuclear Information System (INIS)

    The CANDU program in Romania is now well established. The Cernavoda Nuclear Station presently under construction will consist of 5-CANDU 600 MWE Units and another similar size station is planned to be in operation in the next decade. Progress on the multi-unit station at Cernavoda was stalled for 18 months in 1982/83 as the Canadian Export Development Corporation had suspended their loan disbursements while the Romanian National debt was being rescheduled. Since resumption of the financing in August 1983 contracts worth almost 200M dollars have been placed with Canadian Companies for the supply of major equipment for the first two units. The Canadian design is that which was used in the latest 600 MWE CANDU station at Wolsong, Korea. The vast construction site is now well developed with the cooling water systems/channels and service buildings at an advanced stage of completion. The perimeter walls of the first two reactor buildings are already complete and slip-forming for the 3rd Unit is imminent. Many Romanian organizations are involved in the infrastructure which has been established to handle the design, manufacture, construction and operation of the CANDU stations. The Romanian manufacturing industry has made extensive preparations for the supply of CANDU equipment and components, and although a major portion of the first two units will come from Canada their intentions are to become largely self-supporting for the ensuing CANDU program. Quality assurance programs have been prepared already for many of the facilities

  20. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10-6/year. CANDU nuclear plant designers and owner/operators share information and operational experience nationally and internationally through the CANDU Owners' Group (COG). The research program generally emphasizes the unique aspects of the CANDU concept, such as heat removal through the moderator, but it has also contributed significantly to areas generic to most power reactors such as hydrogen combustion, containment failure modes, fission product chemistry, and high temperature fuel behaviour. Abnormal plant operating procedures are aimed at first using event-specific emergency operating procedures, in cases where the event can be diagnosed. If this is not possible, generic procedures are followed to control Critical Safety Parameters and manage the accident. Similarly, the on-site contingency plans include a generic plan covering overall plant response strategy, and a specific plan covering each category of contingency

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. Ultrasonic estimation of hydride degradation of zirconium pressure tubes of RBMK fuel channel

    International Nuclear Information System (INIS)

    Fuel channels of nuclear reactors, which are major structural elements of a reactor core, have to meet strict requirements in terms of operational reliability. The middle part of the fuel channel, located in a graphite stack, is a tube made of a zirconium-2.5% niobium alloy. However, zirconium alloys can pick up hydrogen during operation as a consequence of corrosion reaction with water. Hydrogen redistributes easily at elevated temperatures migrating down a temperature or concentration gradient and up a stress gradient. When the terminal solid solubility is exceeded in a component such as a pressure tube that is highly stressed for long periods of time, delayed hydride cracking failures may occur. To estimate degradation of the zirconium alloy in the presence of hydrides, predetermined amounts of hydrogen were added to the sections of the fuel channel tubes by electrolytic deposition of a layer of hydride on the surface of the pressure tube material followed by dissolving the hydride layer by diffusion annealing at an elevated temperature. For estimation of the concentration of zirconium hydride platelets in the zirconium alloy test samples ultrasonic testing methods were proposed. The first method is based on precise measurement of velocity of longitudinal and shear wave at different directions and the second is based on the investigation of high frequency ultrasonic signals backscattered in a focal zone of an ultrasonic transducer. The experimental investigations were performed on the zirconium alloy samples of different concentration of hydrides in the immersion tank at a room temperature. The results obtained on testing samples using different excitation conditions and different types of ultrasonic waves are presented. (orig.)

  4. Physics analysis on the NRU core for an accident scenario of a loop pressure tube crack

    International Nuclear Information System (INIS)

    The Nuclear Research Universal (NRU) reactor loops are high temperature, high pressure test facilities, designed for power reactor fuel development and materials testing within the core of the NRU reactor. The loops allow test material to be subject to neutron flux, temperature and pressure conditions typical of a power reactor. This paper describes the physics analysis on the NRU core for an accident scenario of a loop pressure tube crack with a concurrent liner tube failure. After the crack has occurred, thermal-hydraulic analysis predicts the formation of a steam bubble of 50 cm radius in the D2O moderator/coolant around the loop test section. The steam displaces the D2O moderator and has a negative reactivity effect. This negative reactivity effect is large enough to overcome the positive loop void reactivity such that the reactor is shut down and reactor safety is not compromised. The paper also describes the sensitivity of steam bubble densities on the reactivity effect and presents results for subsequent reductions of fluxes and channel powers around the loop site. (author)

  5. Physics analysis on the NRU core for an accident scenario of a loop pressure tube crack

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C., E-mail: leungt@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2010-07-01

    The Nuclear Research Universal (NRU) reactor loops are high temperature, high pressure test facilities, designed for power reactor fuel development and materials testing within the core of the NRU reactor. The loops allow test material to be subject to neutron flux, temperature and pressure conditions typical of a power reactor. This paper describes the physics analysis on the NRU core for an accident scenario of a loop pressure tube crack with a concurrent liner tube failure. After the crack has occurred, thermal-hydraulic analysis predicts the formation of a steam bubble of 50 cm radius in the D{sub 2}O moderator/coolant around the loop test section. The steam displaces the D{sub 2}O moderator and has a negative reactivity effect. This negative reactivity effect is large enough to overcome the positive loop void reactivity such that the reactor is shut down and reactor safety is not compromised. The paper also describes the sensitivity of steam bubble densities on the reactivity effect and presents results for subsequent reductions of fluxes and channel powers around the loop site. (author)

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

    International Nuclear Information System (INIS)

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

  7. 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.

  8. Ear tube insertion

    Science.gov (United States)

    Myringotomy; Tympanostomy; Ear tube surgery; Pressure equalization tubes; Ventilating tubes; Ear infection - tubes; Otitis - tubes ... trapped fluid can flow out of the middle ear. This prevents hearing loss and reduces the risk ...

  9. Enhanced candu 6 reactor: status

    International Nuclear Information System (INIS)

    The CANDU 6 power reactor is visionary in its approach, renowned for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Limited (AECL), the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980s as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first set of CANDU 6 plants - Gentilly 2 and Point Lepreau in Canada, Embalse in Argentina and Wolsong- Unit 1 in Korea - have been in service for more than 22 years and are still producing electricity at peak performance; to the end of 2004, their average Lifetime Capacity Factor was 83.2%. The newer CANDU 6 units in Romania (Cernavoda 1), Korea (Wolsong-Units 2, 3 and 4) and Qinshan (Phase III- Units 1 and 2) have also been performing at outstanding levels. The average lifetime Capacity Factor of the 10 CANDU 6 operating units around the world has been 87% to the end of 2004. Building on these successes, AECL is committed to the further development of this highly successful design, now focussing on meeting customers' needs for reduced costs, further improvements to plant operation and performance, enhanced safety and incorporating up-to-date technology, as warranted. This has resulted in AECL embarking on improving the CANDU 6 design through an upgraded product termed the ''Enhanced CANDU 6'' (EC6), which incorporates several attractive but proven features that make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that are being incorporated into the EC6 include increasing the plant's power output, shortening the overall project schedule, decreasing the capital cost, dealing with obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the

  10. MARATHON - a computer code for the probabilistic estimation of leak-before-break time in CANDU reactors

    International Nuclear Information System (INIS)

    The presence of high levels of moisture in the annulus gas system of a CANDU reactor indicates that a leaking crack may be present in a pressure tube. This will initiate the shutdown of the reactor to prevent the possibility of fuel channel damage. It is also desirable, however, to keep the reactor partially pressurized at hot shutdown for as long as it is necessary to unambiguously identify the leaking pressure tube. A premature full depressurization may cause an extended shutdown while the leaking tube is being located. However, fast fracture could occur during an excessively long hot shutdown period. A probabilistic methodology, together with an associated computer code (called MARATHON), has been developed to calculate the time from first leakage to unstable fracture in a probabilistic format. The methodology explicitly uses distributions of material properties and allows the risk associated with leak-before-break to be estimated. A model of the leak detection system is integrated into the methodology to calculate the time from leak detection to unstable fracture. The sensitivity of the risk to changing reactor conditions allows the optimization of reactor management after leak detection. In this report we describe the probabilistic model and give details of the quality assurance and verification of the MARATHON code. Examples of the use of MARATHON are given using preliminary material property distributions. These preliminary material property distributions indicate that the probability of unstable fracture is very low, and that ample time is available to locate the leaking tube

  11. Failure assessment and evaluation of critical crack length for a fresh Zr-2 pressure tube of an Indian PHWR

    International Nuclear Information System (INIS)

    Fracture analysis of Zr-2 pressure tubes having through wall axial crack was done using finite element method. The analysis was done for tubes in as received condition. During reactor operation the mechanical properties of Zr-2 undergo changes. The analysis is valid for pressure tubes of newly commissioned reactors. The main aim of the study was to determine critical crack length of pressure tubes in normal operating conditions. Elastic plastic fracture analysis was done for different crack lengths to determine applied J-integral values. Tearing modulus instability concept was used to evaluate critical crack length. One of the important parameter studied was, the effect of crack face pressure, which leaking fluid exert on the crack faces/lips of through wall axial crack. Its effect was found to be significant for pressure tubes. It increases the applied J-integral values. Approximate analytical solutions which takes into account the plasticity ahead of crack tip, are available and widely used. These formulae do not take into account the crack face pressure. Since, for the present situation the effect of crack face pressure is significant hence, detailed finite analysis was necessary. Detailed 3D finite element analysis gives an insight into the variation of J-integral values over the thickness of pressure tube. It was found that J values are maximum at the middle layer of the tube. A peak factor on J values was defined and evaluated as ratio of maximum J to average J across the thickness, crack opening area for each length was also evaluated. The knowledge of crack opening area is useful for leak before break studies. The failure assessment was also done using Central Electricity Generating Board (CEGB) R-6 method considering the ductile tearing. The reserve factors (or safety margins) for different crack lengths was evaluated using R-6 method. (author). 30 refs., 21 figs., 34 tabs

  12. A model for analyzing CANDU-6 SDS No.2 poison injection system

    International Nuclear Information System (INIS)

    In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called. ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a CFD code, to simulate the formation of the poison jet curtain inside the moderator tank. For the validation, an attempt was made to validate this model against a poison injection experiment performed at Bhabha Atomic Research Center (BARC) of India. The interim progress will be presented and the validation analysis result is discussed

  13. Second International Conference on CANDU Fuel

    International Nuclear Information System (INIS)

    Thirty-four papers were presented at this conference in sessions dealing with international experience and programs relating to CANDU fuel; fuel manufacture; fuel behaviour; fuel handling, storage and disposal; and advanced CANDU fuel cycles. (L.L.)

  14. Use of pressurized eccentric tubes to study the effect of hydrostatic stress on swelling

    International Nuclear Information System (INIS)

    A technique for measuring the effect of hydrostatic stress on radiation-induced swelling is presented. This technique is based on the nonuniform hydrostatic stress that arises when an eccentric tube (a tube with inner and outer surfaces having dissimilar centers of revolution) is internally pressurized. The elastic analyses of the thin- and thick-walled eccentric tube are given. The elastic stress state is allowed to relax plastically, based on a constitutive law for deformation during neutron irradiation. In this case, the constitutive law contains a linearly stress-dependent deviatoric strain rate and a dilatation rate that is linearly dependent on hydrostatic stress. Emphasis is placed on the specimen design and experimental procedure for in-reactor experiments in which the coefficient relating hydrostatic stress and swelling is sought. It is shown that, for the 316L stainless steel specimens placed in EBR-II, we may expect that any appreciable effect of hydrostatic stress on swelling will be observable through changes in specimen curvature

  15. Elastic-plastic fracture mechanics analyses of cracked steam generator tubes under internal pressure

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeong Keun; Ahn, Min Yong; Moon, Seong In; Chang, Yoon Suk; Kim, Young Jin [Sungkyunkwan Univ., Suwon (Korea, Republic of); Hwang, Seong Sik; Kim, Joung Soo [KAERI, Taejon (Korea, Republic of)

    2005-07-01

    The structural and leakage integrity of steam generator tube should be maintained during operation even though a crack is existed on it. During the past three decades, several limit load solutions have been proposed to resolve the integrity issue. However, for exact load carrying capacity estimation of specific components under different conditions, these solutions have to be modified by using lots of experimental data. The purpose of this paper is to introduce a new burst pressure estimation scheme based on fracture mechanics analyses for steam generator tube with an axial or circumferential through-wall crack. To do this, closed-form engineering equations were derived to get relevant parameters from three dimensional elastic-plastic finite element analyses combined with reference stress method. Also, a series of structural integrity analyses were carried out using the calculated J-integral from engineering equations and fracture toughness data. Thereby, in comparison with the experimental data as well as corresponding estimation results from limit load solutions, it was proven that the proposed estimation scheme can be used as an efficient tool for integrity evaluation of cracked steam generator tubes.

  16. Quality Products - The CANDU Approach

    International Nuclear Information System (INIS)

    The prime focus of the CANDU concept (natural uranium fuelled-heavy water moderated reactor) from the beginning has economy, heavy water losses and radiation exposures also were strong incentives for ensuring good design and reliable equipment. It was necessary to depart from previously accepted commercial standards and to adopt those now accepted in industries providing quality products. Also, through feedback from operating experience and specific design and development programs to eliminate problems and improve performance, CANDU has evolved into today's successful product and one from which future products will readily evolve. Many lessons have been learned along the way. On the one hand, short cuts of failures to understand basic requirements have been costly. On the other hand, sound engineering and quality equipment have yielded impressive economic advantages through superior performance and the avoidance of failures and their consequential costs. The achievement of lifetime economical performance demands quality products, good operation and good maintenance. This paper describes some of the basic approaches leading to high CANDU station reliability and overall excellent performance, particularly where difficulties have had to be overcome. Specific improvements in CANDU design and in such CANDU equipment as heat transport pumps, steam generators, valves, the reactor, fuelling machines and station computers, are described. The need for close collaboration among designers, nuclear laboratories, constructors, operators and industry is discussed. This paper has reviewed some of the key components in the CANDU system as a means of indicating the overall effort that is required to provide good designs and highly reliable equipment. This has required a significant investment in people and funding which has handsomely paid off in the excellent performance of CANDU stations. The close collaboration between Atomic Energy of Canada Limited, Canadian industry and the

  17. CANDU: The fuel conserving reactor

    International Nuclear Information System (INIS)

    Because of their high neutron economy and unique design features, CANDU heavy water moderated reactors are the only established commercial reactors able to use directly low fissile content fuels such as natural uranium or uranium recovered from spent light water reactor fuel (RU). These features also help them to achieve the highest fuel utilization of all commercially available reactors, whether the fuel is based on natural uranium or mixed oxides of plutonium, uranium or thorium. As nuclear capacity growth increases demands on the world's finite uranium resources, AECL envisages near term use in CANDU reactors of a fuel incorporating RU and fuels containing thorium, with either plutonium or low enriched uranium (LEU) as the fissile 'driver' fuel. In the long term, AECL proposes the use of future 'Generation X' CANDU reactors with enhanced neutron economy to achieve a near-Self-Sufficient Equilibrium Thorium (SSET) fuel cycle. This CANDU SSET would have a conversion ratio of unity and be able to produce power indefinitely, with the need for little additional fissile material once equilibrium is reached (the amount of 233U needed in the fresh fuel is the same as is present in the discharged fuel, including processing losses.) This would also enable a CANDU-Fast Breeder Reactor (FBR) synergism that would allow each fuel-generating, though expensive, FBR to supply the initial fissile requirements of several less-expensive, CANDU SSET reactors operating on the thorium cycle. The closer the approach to an SSET that CANDUs can achieve, the higher the ratio of CANDUs to breeders in an economically optimized reactor fleet. CANDU reactors thereby become natural partners of both light water-cooled thermal reactors and fast breeder reactors, in both cases making optimum use of their spent fuel components and enhancing the overall sustainability of nuclear power. (authors)

  18. Hydraulic performance evaluation of pressure compensating (pc) emitters and micro-tubing for drip irrigation system

    International Nuclear Information System (INIS)

    Drip irrigation system is necessary for those areas, where the water scarcity issues are present. The present study was conducted at the field station of Climate Change, Alternate Energy and Water Resources Institute (CAEWRI), National Agricultural Research Center (NARC), Islamabad, during 2013, regarding drip irrigation system. Drip irrigation system depends on uniform emitter application flow. All the emitters were tested and replicated thrice at pressure head (34 to 207Kpa) with an increment of 34 Kpa. The minimum and maximum discharges were 1.32 - 3.52, 3.36 - 5.42, and 43.22 - 100.99 Lph, with an average of 2.42, 4.63 and 73.66 Lph, for Bow Smith, RIS and Micro-tubing, respectively. It indicates that more than 90% of emission uniformity (EU) and uniformity coefficient (CU) for all Emitters, which shows excellent water application with least standard deviation, ranging 0.12 to 2.37, throughout the operating pressure heads in all emitters. An average coefficient of variation (CV) of all emitters were behaving less than 0.07, indicating an excellent class at all operating pressure heads between 34 to 207 Kpa. Moreover, the relationship of discharge and pressure of emitters indicates that discharge increased with the increase of pressure head. The Q-H curve plays key role in the selection of emitters. (author)

  19. The failure of the pressure tube in fuel channel NO6 of Bruce NGS-A unit 2 in March 1986

    International Nuclear Information System (INIS)

    The events leading up to the failure of Bruce NGS-A Unit 2 pressure tube NO6 are described. Based on the subsequent examination of the tube, reactor operating history and pressure tube manufacturing information, the failure sequence is deduced. Remedial actions which have been adopted for avoiding similar occurrences are listed

  20. Development of advanced CANDU PHWR -Development of the advanced CANDU technology-

    International Nuclear Information System (INIS)

    The target of this project is to assess the feasibility of improving PHWR and to establish the parameter of the improved concept and requirements for developing it. To set up the requirements for the Improved Pressurized Heavy Water Reactor: (1) Design requirements of PHWR main systems and Safety Design Regulatory Requirements for Safety Related System i.e. Reactor Shutdown System, Emergency Core Cooling System and Containment System were prepared. (2) Licensing Basis Documents were summarized and Safety Analysis Regulatory. Requirements were reviewed and analyzed. To estimate the feasibility of improving PHWR and to establish the main parameters of the concept of new PHWR in future: (1) technical level/developing trend of PHWR in Korea through Wolsong 2, 3 and 4 design experience and Technical Transfer Program was investigated to analyze the state of basic technology and PHWR improvement potential. (2) CANDU 6 design improvement tendency, CANDU 3 design concept and CANDU 9 development state in other country was analyzed. (3) design improvement items to apply to the reactors after Wolsong 2, 3 and 4 were selected and Plant Design Requirements and Conceptual Design Description were prepared and the viability of improved HWR was estimated by analyzing economics, performance and safety. (4) PHWR technology improving research and development plan was established and international joint study initiated for main design improvement items

  1. Development of advanced CANDU PHWR -Development of the advanced CANDU technology-

    Energy Technology Data Exchange (ETDEWEB)

    Seok, Ho Cheon; Na, Yeong Hwan; Seok, Soo Dong; Lee, Bo Uk; Kwak, Ho Sang; Kim, Bong Ki; Kim, Seok Nam; Min, Byeong Joo; Park, Jong Ryunl; Shin, Jeong Cheol; Lee, Kyeong Ho; Lee, Dae Hee; Lee, Deuk Soo; Lee, Yeong Uk; Lee, Jeong Yang; Jwon, Jong Seon; Jwon, Chang Joon; Ji, Yong Kwan; Han, Ki Nam; Kim, Kang Soo; Kim, Dae Jin; Kim, Seon Cheol; Kim, Seong Hak; Kim, Yeon Seung; Kim, Yun Jae; Kim, Jeong Kyu; Kim, Jeong Taek; Kim, Hang Bae; Na, Bok Kyun; Namgung, In; Moon, Ki Hwan; Park, Keun Ok; Shon, Ki Chang; Song, In Ho; Shin, Ji Tae; Yeo, Ji Won; Oh, In Seok; Jang, Ik Ho; Jeong, Dae Won; Jeong, Yong Hwan; Ha, Jae Hong; Ha, Jeong Koo; Hong, Hyeong Pyo; Hwang, Jeong Ki [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    The target of this project is to assess the feasibility of improving PHWR and to establish the parameter of the improved concept and requirements for developing it. To set up the requirements for the Improved Pressurized Heavy Water Reactor: (1) Design requirements of PHWR main systems and Safety Design Regulatory Requirements for Safety Related System i.e. Reactor Shutdown System, Emergency Core Cooling System and Containment System were prepared. (2) Licensing Basis Documents were summarized and Safety Analysis Regulatory. Requirements were reviewed and analyzed. To estimate the feasibility of improving PHWR and to establish the main parameters of the concept of new PHWR in future: (1) technical level/developing trend of PHWR in Korea through Wolsong 2, 3 and 4 design experience and Technical Transfer Program was investigated to analyze the state of basic technology and PHWR improvement potential. (2) CANDU 6 design improvement tendency, CANDU 3 design concept and CANDU 9 development state in other country was analyzed. (3) design improvement items to apply to the reactors after Wolsong 2, 3 and 4 were selected and Plant Design Requirements and Conceptual Design Description were prepared and the viability of improved HWR was estimated by analyzing economics, performance and safety. (4) PHWR technology improving research and development plan was established and international joint study initiated for main design improvement items.

  2. 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. PMID:26075510

  3. Future generations of CANDU: advantages and development with passive safety

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) advances water reactor and CANDLT technology using an evolutionary development strategy. This strategy ensures that innovations are based firmly on current experience and keeps our development programs focused on one reactor concept, reducing risks, development costs, and product development cycle times. It also assures our customers that our products will never become obsolete or unsupported, and the continuous line of water reactor development is secure and supported into the future. Using the channel reactor advantage of modularity, the subdivided core has the advantage of passive safety by heat removal to the low- pressure moderator. With continuous improvements, the Advanced CANDU Reactor TM (ACR-1000TM) concept will likely remain highly competitive for a number of years and leads naturally to the next phase of CANDU development, namely the Generation IV CANDU -SCWR concept. This is conventional water technology, since supercritical boilers and turbines have been operating for some time in coal-fired power plants. Significant cost, safety, and performance advantages would result from the CANDU-SCWR concept, plus the flexibility of a range of plant sizes suitable for both small and large electric grids, and the ability for co-generation of electric power, process heat, and hydrogen. In CANDU-SCWR, novel developments are included in the primary circuit layout and channel design. The R and D in Canada is integrated with the Generation IV international Forum (GIF) plans, and has started on examining replaceable insulating liners that would ensure channel life, and on providing completely passive reactor decay heat removal directly to the moderator heat sink without forced cooling. In the interests of sustainability, hydrogen production by a CANDU- SCWR is also be included as part of the system requirements, where the methods for hydrogen production will depend on the outlet temperature of the reactor

  4. Severe accident analysis for shutdown state scenarios of CANDU6 plant using ISAAC

    International Nuclear Information System (INIS)

    This paper describes the analysis results for the shutdown state accident scenarios with ISAAC (Integrated Severe Accident Analysis code for CANDU plants) in terms of the severe core damage progression from the fuel heat up to the fuel channel failure, fuel material relocation to the calandria vessel, and to calandria/reactor building failure. The analyzed cases include the CANDU6 genetic scenarios. For the base case which represents the most pessimistic assumption for the safety systems, the important phenomena are described at the plant systems following the accident progression including the primary heat transport system, the core (fuel channel and suspended debris bed), the calandria vessel, the reactor vault, and the reactor building. Also the fission product and hydrogen behavior are analyzed. In order to see the effect of the safety systems on severe core damage accident progression, the availability of a moderator cooling system and a shield cooling system are considered for the sensitivity cases. Each scenario is analyzed up to 500,000 seconds (138.9 hours) to see the corium behavior until the reactor vault bottom concrete melt-through. For the ISAAC simulation of the CANDU6 plant, 18 representative fuel channels for 190 actual channels each loop (9 channels between steam generators) are defined for the core configuration. The reactor building is defined with 13 compartments and 22 junctions including reactor building leakage. The result of the most severe case of base case shows that the core uncovers at 2.8 hours, pressure tube ruptures at 3.3 hours due to creep, the reactor building fails at about 32.6 hours and the calandria fails at 49.9 hours after an accident initiation. In the scenarios where the moderator cooling system is available, the pressure tubes experience the creep rupture, but the fuel melt do not occur and the reactor building maintained their integrity. The end shield cooling system can't prevent the core melt and relocation but

  5. Evolution of microstructure during fabrication of Zr-2.5 wt pct Nb ally pressure tubes

    International Nuclear Information System (INIS)

    Microstructural changes occurring during the fabrication of Zr-2.5 pct Nb alloy pressure tubes by a modified route, involving hot extrusion followed by two pilgering operations with an intermediate annealing step, have been examined in detail. In the conventional fabrication route, the hot extrusion step is followed by a single cold drawing operation in which the cold work to the extent of 25 pct is imparted to the material for achieving the required mechanical properties. Tensile properties obtained at each stage of fabrication have been evaluated and compared between the two processes. The main aim of this work has been to produce a microstructure and texture which are know n to yield a lower irradiation growth. Additionally, suitable annealing conditions have been optimized for the intermediate annealing which annihilates the cold work introduced by the first cold pilgering operation without disturbing the two-phase elongated microstructure. This elongated α + βI microstructure is required for obtaining the desired level of strength at 310 C. The final microstructure and the crystallographic texture of the finished pressure tube have been compared with those reported for the conventionally processed material

  6. Bruce and Darlington power pulse and pressure tube integrity programs -status 1995

    International Nuclear Information System (INIS)

    The optimum solution to pressure tube fretting at the inlet of the Bruce and Darlington channels, a concern which became very serious following inspections in early 1992, is to remove the inlet bundle and operate with a 12 fuel bundle channel. During analysis of this operating mode a 'power pulse' was identified which could occur during an inlet header break where all the fuel in the channel moved rapidly to the inlet of the channel. The pulse was unacceptable and the units were derated until solutions could be implemented. A number of solutions were identified and each station has begun implementation of their specific solution. Implementation has not been without problems and this paper provides a status report on the progress to date of the long bundle implementation solution for Bruce B and Darlington and the fuelling with the flow solution being implemented at Bruce A. Both types of solution have a significant impact on the original concern, fretting of the pressure tube. (author). 1 ref., 6 figs

  7. Characteristics of gas-liquid two-phase flow in a vertical small diameter tube at a medium pressure

    International Nuclear Information System (INIS)

    Most of correlations for calculating two-phase flow parameters, such as flow pattern transitions, void fraction and pressure drop, have been developed based on the experimental data on tubes greater than 10 mm in diameter at near atmospheric pressures. For that, the applicability of such correlations is doubtful to the flow in small diameter tubes at a medium pressure as seen in compact heat exchangers like residential room conditioners. In this connection, the purpose of this study is to provide experimental data for gas-liquid two-phase vertical flows in a small diameter tube at medium pressures since the published data for such flows is limited to examine existing correlations and/or develop a new one. Experiments have been conducted on air-water two-phase flows in a vertical circular tube of 9.48 mm internal diameter. In the experiment, system pressure in the channel has been systematically changed from 0.2 to 0.7 MPa (absolute) to study the effect of the pressure on two-phase flow parameters, i.e., two-phase flow pattern transitions, bubble size in bubble flow, void fraction, interfacial shear force, frictional pressure drop and static pressure fluctuations. Furthermore, the respective data obtained have been compared with existing correlations. (author)

  8. Evolutionary CANDU 9 plant improvements

    International Nuclear Information System (INIS)

    The CANDU 9 is a 935 MW(e) nuclear power plant (NPP) based on the multi-unit Darlington and Bruce B designs with additional enhancements from our ongoing engineering and research programs. Added to the advantages of using proven systems and components, CANDU 9 offers improvement features with enhanced safety, improved operability and maintenance including a control centre with advanced man-machine interface, and improved project delivery in both engineering and construction. The CANDU 9 NPP design incorporated safety enhancements through careful attention to emerging licensing and safety issues. The designers assessed, revised and evolved such systems as the moderator, end shield, containment and emergency core cooling (ECC) systems while providing an integrated final design that is more passive and severe-accident-immune. AECL uses a feedback process to incorporate lessons learned from operating plants, from current projects experiences and from the implementation or construction phase of previous projects. Most of the requirements for design improvements are based on a systematic review of current operating CANDU stations in the areas of design and reliability, operability, and maintainability. The CANDU 9 Control Centre provides plant staff with improved operability and maintainability capabilities due to the combination of systematic design with human factors engineering and enhanced operating and diagnostics features. The use of advanced engineering tools and modem construction methods will reduce project implementation risk on project costs and schedules. (author)

  9. ACR technology for CANDU enhancements

    International Nuclear Information System (INIS)

    The ACR-1000 design retains many essential features of the original CANDU plant design. As well as further-enhanced safety, the design also focuses on operability and maintainability, drawing on valuable customer input and OPEX. The engineering development of the ACR-1000 design has been accompanied by a research and confirmatory testing program. This program has extended the database of knowledge on the CANDU design. The ACR-1000 design has been reviewed by the Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) which concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada after completing three phases of the pre-project design review. The generic PSAR for the ACR-1000 design was completed in September 2009. The PSAR contains the ACR-1000 design details, the safety and design methodology, and the safety analysis that demonstrate the ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. The ACR technology developed during the ACR-1000 Engineering Program and the supporting development testing has had a major impact beyond the ACR program itself: Improved CANDU components and systems; Enhanced engineering processes and engineering tools, which lead to better product quality, and better project efficiency; and Improved operational performance. This paper provides a summary of technology arising from the ACR program that has been incorporated into new CANDU designs such as the EC6, or can be applied for servicing operating CANDU reactors. (author)

  10. 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

  11. Development of Cyclic Pressurization Fatigue Test Technique for Spent Fuel Cladding Tube

    International Nuclear Information System (INIS)

    If the utility adopts a load following operation, the cyclic changes of the diameter causing a low-cycle fatigue will occur more frequently. Although failures regarding a radial fatigue in the fuel cladding have not been reported yet, it is essential to accumulate a fatigue life database for use in a fuel design. Since Soniak's proposal for the low cycle radial fatigue under cyclic pressurization of the fuel cladding, KAERI's R and D group has also produced a lot of low cycle fatigue data for the un-irradiated fuel cladding tube using a cyclic pressurization device. However, fatigue data regarding irradiated fuel cladding under cyclic pressurization has not been obtained around the country until now. And the infrastructures and fatigue test techniques, which can produce the fatigue data on the irradiated fuel cladding, are still worse off. The objectives of this study are to develop a low cycle fatigue test techniques for irradiated fuel cladding, as well as to produce a stress-life curve of the irradiated cladding under the cyclic pressurization

  12. Development and validation of a model for CANDU-6 SDS2 poison injection analysis

    International Nuclear Information System (INIS)

    In CANDU-6 reactor there are two independent reactor shutdown systems. The shutdown system no. 2(SDS2) injects the liquid poison into the moderator tank by high pressure via small holes on the 6 nozzle pipes and stops the nuclear chain reaction. To ensure the safe shutdown of a reactor loaded with either DUPIC or SEU fuels it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. The motivation for this work arose from the fact that the computer code package for performing this task is not transfered to Korea yet. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a commercial CFD code developed by AEA technology, to simulate the formation of the poison jet curtain inside the moderator tank. For the validation, a simulation for a generic CANDU-6 SDS2 design poison jet growth experiment was made to evaluate this model's capability against experiment. As no concentration field was measured and only the growth of the poison jet height was obtained by high speed camera, the validation was limited as such. The result showed that if one assume the jet front corresponds to 200 ppm of the poison the model succeed to

  13. Examining the response pressure along a fluid-filled elastic tube to comprehend Frank's arterial resonance model.

    Science.gov (United States)

    Lin Wang, Yuh-Ying; Sze, Wah-Keung; Lin, Chin-Chih; Chen, Jiang-Ming; Houng, Chin-Chi; Chang, Chi-Wei; Wang, Wei-Kung

    2015-04-13

    Frank first proposed the arterial resonance in 1899. Arteries are blood-filled elastic vessels, but resonance phenomena for a fluid-filled elastic tube has not drawn much attention yet. In this study, we measured the pressure along long elastic tubes in response to either a single impulsive water ejection or a periodic water input. The experimental results showed the low damped pressure oscillation initiated by a single impulsive water input; and the natural frequencies of the tube, identified by the peaks of the response in the frequency domain, were inversely proportional to the length of the tube. We found that the response to the periodic input reached a steady distributed oscillation with the same period of the input after a short transient time; and the optimal pressure response, or resonance, occurred when the pumping frequency was near the fundamental natural frequency of the system. We pointed out that the distributed forced oscillation could also be a suitable approach to analyze the arterial pressure wave. Unlike Frank's resonance model in which the whole arterial system was lumped together to a simple 0-D oscillator and got only one natural frequency, a tube has more than one natural frequency because the pressure P(z,t) is a 1-D oscillatory function of the axial position z and the time t. The benefit of having more than one natural frequency was then discussed. PMID:25773589

  14. Optimization of process parameters for control of hydrogen in Zr-2.5Nb pressure tubes for PHWRs

    International Nuclear Information System (INIS)

    Hydrogen induced problems such as hydrogen embrittlement, blistering and DHC are some of the most critical life limiting factors for PHWR pressure tubes. The pressure tubes pick up on average 1 ppm hydrogen every year during reactor operation. Therefore to extend the life of the pressure tubes by countering the deleterious effects of hydrogen, the initial hydrogen content of the as manufactured pressure tube has to be brought down considerably. Earlier NFC was producing Zr-2.5Nb pressure tubes from double melted ingots with an H content of nearly 25 ppm. Owing to the above consideration, process optimization was carried out to reduce this H content from 25 ppm to 5 ppm. This paper describes the various steps adopted for reduction of the hydrogen content during a series of manufacturing operations and processes such as sponge production, melting, extrusion, pilgering, pickling, cleaning, heat treatment and final finishing operations. Intermediate product analysis and characterization has been carried out to monitor and minimize the hydrogen content at critical process steps. (author)

  15. Theoretical and experimental modeling of the multiple pressure tube rupture for RBMK reactor. Pt. 1

    International Nuclear Information System (INIS)

    Rupture of a single fuel channel (pressure tube) or several fuel channels of the RBMK may occur in service conditions on NPPs with a variety of initiating events. It is assumed in RBMK Safety Cases that the force of the escaping fluid will not cause neighboring channels to break. This assumption has not been justified. Hence, an analysis of the multiple pressure tube rupture (MPTR) possibility is needed. The analysis of the MPTR problem requires performing a series of theoretical and experimental studies of separate physical processes running in the RBMK reactor, as well as development of mathematical models and their physical equivalents. The experimental rigs concerned the MPTR problem have been designed and constructed at Electrogorsk Research and Engineering Center, Russia. Investigation of the circumstances and mechanisms of a single channel rupture at the various conditions and scenarios is one of the main stages of the MPTR problem analysis. Theoretical models of the single channel rupture under thermal and mechanical loading have been developed including a channel constrained by the graphite block. Deformation of the channel under internal pressure and localized thermal action is modeled within the framework of the nonlinear shells theory taking into account physically nonlinear material behavior. Computer program based on these models enables to describe the thermomechanical deformation of a single channel and to predict rupture moment. Theoretical studies were accompanied by experimental modeling single channel rupture by means of series experimental examinations at TKR-F test rig (Model of an Accidental Channel). This test rig represents a model of the single disrupted fuel channel in a surrounding graphite column. Experimental examinations make possible the development and verification of theoretical models and make more exact the conditions and mechanism of a single channel rupture. Theoretical and experimental modeling consolidation sets out technique

  16. Applicability of miniature specimen techniques for evaluating the mechanical properties of pressure tube and cladding material

    International Nuclear Information System (INIS)

    The Miniature Specimen Test Techniques (MSTT) namely Small Punch Test (SPT) and Ball Indentation Test (BIT) are commonly employed for evaluating the tensile properties of metallic materials. While discs of 3mm diameter with 0.25mm thickness are utilized for SPT method, the samples for BIT are of any shape with parallel polished top and bottom surfaces having thickness of at least 1mm. The SPT technique is based on driving a ball through clamped miniature disc specimens for deforming till it fractures whereas BIT involves multiple indentations, load and unload cycles at a single indentation point on a polished metallic surface by a spherical indenter. The specimens were fabricated from Zr-2.5%Nb pressure tube (PT) material that is used in pressurised heavy water reactor (PHWR). A suitable die-punch assembly was designed and developed in house to clamp the specimen to carry out the small punch test with the help of a screw-driven universal testing machine. In this work we have utilized miniature tensile specimens for evaluating and comparing the mechanical properties of PT material with that obtained from SPT and BIT. The tensile specimens were prepared using wire cut Electrical Discharge Machining (EDM) as per general guideline of ASTM standard E-8. Tests were carried out at ambient and higher temperatures. The tensile properties obtained from tension test and two MSTTs show that the tensile properties vary with orientation and temperature. In order to evaluate mechanical properties of cladding tube two techniques namely SPT and Ring Tension Test (RTT) have been used. The RTT is another technique, already established for estimation of the mechanical properties of cladding tube material in the transverse direction. Experimental results were generated at ambient and higher temperatures by preparing specimens from the same cladding tube in the form of 3mm discs and rings. The basic fixture that was used for carrying out ring tension test of the cladding tube consists

  17. Numerical Investigation of Air-Side Heat Transfer and Pressure Drop in Circular Finned-Tube Heat Exchangers

    OpenAIRE

    Mon, Mi Sandar

    2009-01-01

    A three-dimensional numerical study is performed to investigate the heat transfer and pressure drop performance on the air-side of circular finned tube bundles in cross flow. New heat transfer and pressure drop correlations for the air-cooled heat exchangers have been developed with the Reynolds number ranging from 5000 to 70000. The heat transfer and pressure drop results agree well with several existing experimental correlations. In addition, the influence of the geometric parameters on the...

  18. Improvement of high-voltage performance of acceleration tubes by cleaning the walls with a high-pressure water jet

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, S. E-mail: takeuchi@tandem.tokai.jaeri.go.jp; Nakanoya, T.; Kabumoto, H.; Yoshida, T

    2003-11-11

    We cleaned electrostatic accelerator tubes by applying a high-pressure water jet and examined their high-voltage performances at 1 and 3 MV. The cleaning was very effective in reducing discharge activities at their rated voltages. We did some experimental investigations with the tubes and their ceramic insulators. We found that removal of microparticles loosely bound on the vacuum-side ceramic surfaces had an important effect in eliminating the discharge activities.

  19. CFD Analysis of Shell and Tube Heat Exchanger to Study the Effect of Baffle Cut on the Pressure Drop

    OpenAIRE

    Avinash D Jadhav; Tushar A Koli

    2014-01-01

    The shell side design of a shell and tube heat exchanger; in particular the baffle spacing, baffle cut and shell diameter dependencies of the heat transfer coefficient and the pressure drop are investigated by numerically modelling a small heat exchanger. The flow and temperature fields inside the shell are resolved using a commercial CFD package. A set of CFD simulations is performed for a single shell and single tube pass heat exchanger with a variable number of baffles and turb...

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

    OpenAIRE

    Gustavo Boroni; Alejandro Clausse

    2011-01-01

    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...

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

    OpenAIRE

    ARGUNHAN, Zeki; Yildiz, Cengiz

    2006-01-01

    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.

  2. 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.

  3. Build your own Candu reactor

    International Nuclear Information System (INIS)

    The author discusses the marketing of Candu reactors, particularly the export trade. Future sales will probably be of the nuclear side of a station only, thus striking a compromise between licensing and 'turnkey' sales. It is suggested that AECL might have made more money in the past had it not given the right to manufacture Candu fuel away to Canadian industry. Future sales to certain potential customers may be limited by the requirement of strict safeguards, which will almost certainly never be relaxed. (N.D.H.)

  4. Pressure-driven flow past spheres moving in a circular tube

    Science.gov (United States)

    Sheard, G. J.; Ryan, K.

    A computational investigation, supported by a theoretical analysis, is performed to investigate a pressure-driven flow around a line of equispaced spheres moving at a prescribed velocity along the axis of a circular tube. This fundamental study underpins a range of applications including physiological circulation research. A spectral-element formulation in cylindrical coordinates is employed to solve for the incompressible fluid flow past the spheres, and the flows are computed in the reference frame of the translating spheres.Both the volume flow rate relative to the spheres and the forces acting on each sphere are computed for specific sphere-to-tube diameter ratios and sphere spacing ratios. Conditions at which zero axial force on the spheres are identified, and a region of unsteady flow is detected at higher Reynolds numbers (based on tube diameter and sphere velocity). A regular perturbation analysis and the reciprocal theorem are employed to predict flow rate and drag coefficient trends at low Reynolds numbers. Importantly, the zero drag condition is well-described by theory, and states that at this condition, the sphere velocity is proportional to the applied pressure gradient. This result was verified for a range of spacing and diameter ratios. Theoretical approximations agree with computational results for Reynolds numbers up to O(100).The geometry dependence of the zero axial force condition is examined, and for a particular choice of the applied dimensionless pressure gradient, it is found that this condition occurs at increasing Reynolds numbers with increasing diameter ratio, and decreasing Reynolds number with increasing sphere spacing.Three-dimensional simulations and predictions of a Floquet linear stability analysis independently elucidate the bifurcation scenario with increasing Reynolds number for a specific diameter ratio and sphere spacing. The steady axisymmetric flow first experiences a small region of time-dependent non

  5. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  6. 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. PMID:27397738

  7. Collapse of geometrically imperfect stainless steel tubes under external hydrostatic pressure

    International Nuclear Information System (INIS)

    This paper reports on an investigation into the buckling behaviour of 5 geometrically imperfect duplex stainless steel tube models, subjected to external hydrostatic pressure. The research was partly experimental and partly theoretical, where in the latter case; both analytical and numerical theoretical analyses were carried out. The experiments were carried out on five stainless steel tube models of different lengths, using two mild steel end bungs to seal the models. The experimental results showed that the Duplex stainless steel specimens behaved similarly to other isotropic materials tested by other researchers. The theoretical calculations and analyses were made by MisesNP, a DOS-based computer program, together with the ANSYS finite element structural analysis software. Combining the results of the present series of models, together with the results of other experimenters, a design chart was produced, which can be used for designing full-scale vessels. It should be emphasised that this design chart has been extended to those from previous studies, so that shorter and thicker vessels can now be designed.

  8. Control of alkaline stress corrosion cracking in pressurized-water reactor steam generator tubing

    International Nuclear Information System (INIS)

    Outer-diameter stress corrosion cracking (ODSCC) of alloy 600 (UNS N06600) tubings in steam generators of the Kori-1 pressurized-water reactor (PWR) caused an unscheduled outage in 1994. Failure analysis and remedy development studies were undertaken to avoid a recurrence. Destructive examination of a removed tube indicated axial intergranular cracks developed at the top of sludge caused by a boiling crevice geometry. A high ODSCC propagation rate was attributed to high local pH and increased corrosion potential resulting from oxidized copper presumably formed during the maintenance outage and plant heatup. Remedial measures included: (1) crevice neutralization by crevice flushing with boric acid (H3BO3) and molar ratio control using ammonium chloride (NH4Cl), (2) corrosion potential reduction by hydrazine (H2NNH2) soaking and suppression of oxygen below 20 ppb to avoid copper oxide formation, (3) titanium dioxide (TiO2) inhibitor soaking, and (4) temperature reduction of 5 C. Since application of the remedy program, no significant ODSCC has been observed, which clearly demonstrates the benefit of departing from an oxidizing alkaline environment. In addition, the TiO2 inhibitor appeared to have a positive effect, warranting further examination

  9. Detecting Nonlinearity in Pressure Data Inside the Draft Tube of a Real Francis Turbine

    CERN Document Server

    Sello, S

    1995-01-01

    A general method for testing nonlinearity in time series is described and applied to measurements of different pressure data inside the draft tube surge of a real Francis turbine. Comparing the current original time series to an ensemble of surrogates time series, suitably constructed to mimic the linear properties of the original one, we was able to distinguish a linear stochastic from a nonlinear deterministic behaviour and, moreover, to quantify the degree of nonlinearity present in the related dynamics. The problem of detecting nonlinear structure in real data is quite complicated by the influence of various contaminations, like broadband noise and/or long coherence times. These difficulties have been overcame using the combination of a suitable nonlinear filtering technique and a qualitative redundancy statistic analysis. The above investigations allow a quantitative characterization of different dynamical regimes of motion of gas cavities inside real turbines and, moreover, allow to support the reliabil...

  10. Hydride distribution around a blister in Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    Blisters were grown in Zr-2.5Nb pressure tube sections by a thermal gradient without applying external stress. The surrounding hydride distribution was analyzed. Hydride platelets were observed in the radial direction of the blister. The precipitation of these hydrides was found to be favored by low temperature of blister growth and slow cooling rate after blister formation. The misfit strain produced by hydride blister growth provides the stress necessary to promote radial precipitation. During the subsequent tensile test at 200 C (delayed hydride cracking test) the radial hydride length and thickness are increased. This increase is explained by a stress concentrator effect of the blister. When this effect vanishes, the increase of radial hydrides continues by an autocatalytic effect and stress concentrator effect of the hydride platelet. If a crack originated in the blister reaches the matrix it could propagate along a radial hydride previously precipitated. (orig.)

  11. Microstructural change of a 9Cr steel longitudinal welded tube under internal pressure creep loading

    International Nuclear Information System (INIS)

    In this study, the microstructure of a base metal, and the heat-affected zone (HAZ) and weld metal of a 9Cr steel longitudinal welded tube with various internal pressure creep damage levels was studied by transmission electron microscopy and energy dispersive X-ray spectroscopy. Each portion of the steel weldment clearly changed with damage level. The present results indicate that the recovery of the microstructure is faster in HAZ than in other portions, and thus creep deformation preferentially occurs in HAZ. Moreover, it was revealed that stress accelerates the growth of M23C6 precipitates as well as the reduction of dislocation density, consequently promoting recovery. It was also confirmed that the so-called type IV failure is reasonably explained by precipitate strengthening.

  12. Experimental Evaluation of the Burst Pressure of Steam Generator Tube with Multiple Part-through-wall Cracks

    International Nuclear Information System (INIS)

    Since steam generator (SG) tube is a pressure boundary of pressurized water reactor (PWR), the maintaining integrity of SG tube is very important. However, various types of defect caused by a mechanical and chemical degradations have been observed in the SG tube. In particular, the outer diameter stress corrosion cracking (ODSCC) in the secondary side is a dominant type of defect, which can lead to leakage of primary coolant and burst of SG tube. Thus, the integrity evaluation of SG tubes with SCC is considered to be an important issue. A number of experimental and analytical studies have been conducted to evaluate burst pressure of SG tube with defects and proposed evaluation models. But, most of the models were developed based on single cracks, although SCC initiates and grows at multiple sites on the surface of SG tube. Therefore, this study carried out burst tests using SG tube specimens containing multiple part-trough-wall (PTW) flaws at room temperature (RT) and evaluated burst pressure of SG tubes with multiple PTW flaws. The burst tests were conducted on 56 specimens and burst pressures were obtained. Also, failure mode of SG tube with multiple flaws was investigated by examining the shape of crack and tearing from post-test specimens. The reduction was more pronounced for L=25.4mm than L=6.3mm and was more pronounced when three flaws were arranged than when two flaws were arranged. Burst pressure increased with increasing axial distance between flaws for collinear multiple flaws, whereas the pressure decreased and saturated with increasing circumferential distance between non-aligned multiple flaws. For SG tubes with parallel multiple flaws, the burst pressure was influenced by circumferential distance between flaws and length of flaws. When two flaws were parallel with circumferential distance of l1=1mm, the burst pressure was higher about 2% than that of single flaw for L=6.3mm, but it was lower about 7% than that of single flaw for L=25.4mm. Thus an

  13. Numerical model for thermal and mechanical behaviour of a CANDU 37-element bundle

    International Nuclear Information System (INIS)

    Prediction of transient fuel bundle deformations is important for assessing the integrity of fuel and the surrounding structural components under different operating conditions including accidents. For numerical simulation of the interactions between fuel bundle and pressure tube, a reliable numerical bundle model is required to predict thermal and mechanical behaviour of the fuel bundle assembly under different thermal loading conditions. To ensure realistic representations of the bundle behaviour, this model must include all of the important thermal and mechanical features of the fuel bundle, such as temperature-dependent material properties, thermal viscoplastic deformation in sheath, fuel-to-sheath interactions, endplate constraints and contacts between fuel elements. In this paper, we present a finite element based numerical model for predicting macroscopic transient thermal-mechanical behaviour of a complete 37-element CANDU nuclear fuel bundle under accident conditions and demonstrate its potential for being used to investigate fuel bundle to pressure tube interaction in future nuclear safety analyses. This bundle model has been validated against available experimental and numerical solutions and applied to various simulations involving steady-state and transient loading conditions. (author)

  14. Study of pressure tube cold pressurisation of the primary heat transport system under different design basis events

    International Nuclear Information System (INIS)

    Full text: Avoidance of cold pressurisation of the pressure tubes (PTs) for PHWR is considered to be an important issue from the safety point of view. During the reactor startup/cooldown operation, a hot pressurisation practice is always followed to avoid cold pressurisation to avoid brittle failure of the pressure tube caused by hydrogen pickup. In several design basis events it has been found that system undergoes a cold pressurised conditions. Such events are identified and analysed with thermal-hydraulic safety analysis code RELAP4/MOD6 and discussed in this paper

  15. 1st RCM of IAEA CRP on Prediction of Axial and Radial Creep in HWR Pressure Tubes

    International Nuclear Information System (INIS)

    Expected outcome: Improved understanding of pressure tube creep mechanism by studying the effect of intrinsic (material response) as well as extrinsic parameters (operating conditions). • Improvement of material characterization technology: many laboratories participating in this CRP will conduct the microstructure characterization for the first time. • Recommendation for manufacturing to achieve optimal PT performance: The database will enable the identification of best pressure tube performance by comparison of data. • Improvement in aging management procedure: (channel selection for PT deformation management, etc.). • Safety enhancement for operating HWRs by reducing the uncertainty in the prediction of PT deformation

  16. Conceptual designs for very high-temperature CANDU reactors

    International Nuclear Information System (INIS)

    300 - 500 MWe). The steam cycle and coolant conditions are proposed to be the same as CANDU-X Mark I. The major difference between the reactors is that natural convection would be used to circulate the primary coolant around the heat transport system. This approach enhances cycle efficiency and safety, and is viable for reactors operating near the critical point of water because of the large increases in heat capacity and thermal expansion coefficient across the core. The third concept, CANDUal-X, is a dual cycle concept, with core conditions similar to the Mark 1 and NC. In this concept, coolant leaving the core is first expanded through a VHP turbine in a direct cycle. Employing a dual steam cycle avoids a high-pressure steam generator. The conditions of the core and the VHP expansion can be designed such that the exhaust from the turbine is used as the heat source for an indirect cycle; that is, the secondary side can be equivalent to that presently employed in conventional CANDU plants. An advantage of this concept over conventional direct cycle nuclear plants is that only one relatively small turbine is exposed to radioactive coolant, and it is located within containment. In summary, the reactors described above represent concepts that evolve logically from the current CANDU designs to higher efficiency, with only modest extensions of current technology. This paper presents a technical overview of the different conceptual designs, as well as a brief discussion of the enabling technologies that are common to each, which is the focus of current R and D. (author)

  17. Optimization of stress relief heat treatment of PHWR pressure tubes (Zr-2.5Nb alloy)

    International Nuclear Information System (INIS)

    The micro-structure of cold worked Zr-2.5%Nb pressure tube material consists of elongated grains of α-zirconium enclosed by a thin film of β-zirconium phase. This β-Zr phase is unstable and on heating, progressively decomposes to α-Zr phase and β-phase enriched with Nb and ultimately form βNb. Meta-stable ω-phase precipitates as an intermediate step during decomposition depending on the heat treatment schedule, βZr→α+βenr→α+ω+βenr→α+βenr→α+βNb Morphological changes occur in the β-zirconium phase during the decomposition. The continuous ligaments of βZr phase turn into a discontinuous array of particles followed by globulization of the β-phase. The morphological changes impose a significant effect on the creep rate and on the delayed hydride cracking velocity due to reduction in the hydrogen diffusion coefficient in αZr. If the continuity of β-phase is disrupted by heat treatment, the effective diffusion coefficient decreases with a concomitant reduction in DHC velocity. The pressure tubes for the Indian PHWRs are made by a process of hot extrusion followed by cold pilgering in two stages and an intermediate annealing. Autoclaving at 400 deg. C for 36 h ensures stress relieving of the finished tubes. In the present studies, autoclaving duration at 400 deg. C was varied from 24 h to 96 h at 12 h-steps and the micro-structural changes in the β-phase were observed by TEM. Dislocation density, hardness and the micro-structural features such as thickness of β-phase, inter-particle spacing and volume fraction of the phases were measured at each stage. Autoclaving for a longer duration was found to change the morphology of β-phase and increase the inter-particle spacing. Progressive changes in the aspect ratio of the β-phase and their size and distribution are documented and reported. These micro-structural modifications are expected to decrease DHC velocity during reactor operation

  18. CANDU 9 Control Centre Mockup

    International Nuclear Information System (INIS)

    This paper provides a summary of the design process being followed, the benefits of applying a systematic design using human factors engineering, presents an overview of the CANDU 9 control centre mockup facility, illustrates the control centre mockup with photographs of the 3D CADD model and the full scale mockup, and provides an update on the current status of the project. (author)

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

    International Nuclear Information System (INIS)

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

  20. Estimation of the hot extrusion process pressure cycle of zircaloy tubes by torsion and compression tests

    International Nuclear Information System (INIS)

    In the production of Zircaloy-4 tubes for nuclear reactors, the first semi-processed tubular form is obtained using the extrusion process. Empirical equations are normally used, which can be applied to extrusion with axial symmetry, or analytical ones are used such as Seibel's equation to evaluate the extrusion process based on the material flow tension. When we use the flow tension corresponding to the mean value of the velocity of extrusion deformation, the extrusion pressure is significantly underestimated, with relation to the experimentally measured pressure. This is because of the flow tension's heavy dependence on the velocity of deformation, which is typical of commercial zirconium alloys. Therefore, the pressure was estimated by calculating the power dissipated during the deformation assuming a velocity field of homogenous deformation in each stage of deformation but without considering friction forces between the work and the extrusion matrix. The flow tension for the torsion tests performed are compared with the results obtained by compression as reported in the literature. These results are compared with four extrusion sequences carried out with different: reduction rates, temperatures, and deformation velocities. The flow tension from the compression test presents greater tension values than those estimated by the torsion test. The origin of these differences is discussed and the conclusion is that they can be attributed to the different crystallography textures generated in both tests. Once the correction is made for the texture variation, the flow tension values evaluated with both testing types in samples of Zircaloy-4 are the same. The peculiarities of each test in relation to the extrusion process are discussed. Despite the very simplified hypotheses that were assumed, the extrusion pressures calculated with the compression and torsion flow tension results, considering their dependence on the speed of deformation and temperature variation during