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Sample records for calandrias

  1. Fabrication of seamless calandria tubes

    International Nuclear Information System (INIS)

    Full text: Calandria tube is a large diameter, thin walled zircaloy-4 tube and is an important structural component of PHWR type of reactors. These tubes are lifetime components and remain during the full life of the reactor. Calandria tubes are classified as extremely thin walled tubes with a diameter to wall thickness ratio of around 96. Such thin walled tubes are conventionally produced by seam welded route comprising of extrusion of slabs followed by a series of hot and rolling passes, shaping into O-shape and eventual welding. An alternative and superior method of fabricating the calandria tubes, the seamless route, has been developed, which involves hot extrusion of mother blanks followed by three successive cold pilger reductions. Eccentricity correction of the extruded blanks is carried out on a special purpose grinding equipment to bring the wall thickness variation within permissible limits. Predominant wall thickness reductions are given during cold pilgering to ensure high Q-factor values. The texture in the finished tubes could be closely, controlled with an average fr value of 0.65. Pilgering parameters and tube guiding system have been specially designed to facilities rolling of thin walled tubes. Seamless calandria tubes have distinct advantages over welded tubes. In addition to the absence of weld, they are dimensionally more stable, lighter in weight and possess uniform grains with superior grain size. The cycle time from billet to finished product is substantially reduced and the product is amenable to high level of quality assurance. The most significant feature of the seamless route is its material recovery over welded route. Residual stresses measured in the tubes indicate that these are negligible and uniform along the length of the tube. In view of their superior quality, the first charge of seamless calandria tubes will be rolled into the first 500 MWe Pressurised Heavy Water Reactor at Tarapur

  2. 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)

  3. 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

  4. Computation and measurement of calandria tube sag in PHWR

    International Nuclear Information System (INIS)

    Calandria tubes and liquid injection shutdown system (LISS) tubes in a pressurized heavy water reactor (PHWR) is known 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 LISS tube crossing beneath and 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 measuring probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sags of both tubes in the PHWR. 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. (author)

  5. Update on seamless calandria tube development and qualification

    International Nuclear Information System (INIS)

    AECL is undertaking the qualification of the production of seamless calandria tubes as replacement components for installation in reactors during retubing. Seamless tube prototypes made from Zircaloy-2 possessing a suitable crystallographic texture have been shown to be significantly stronger than seam-welded tubes under both rising pressure and sustained pressure conditions in a simulated reactor loading. This paper describes the seamless calandria tube development program and current status. (author)

  6. 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)

  7. Study on heat removal capability of calandria vault water from molten corium in calandria vessel during severe accident of a PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, Sumit Vishnu, E-mail: svprasad@barc.gov.in; Nayak, Arun Kumar; Kulkarni, Primal Pramod; Vijayan, Pallippattu Krishnan; Vaze, Keshav K.

    2015-04-01

    Highlights: • Scaled test set up simulating the calandria vessel and calandria vault water of PHWR. • Experiments conducted with simulant material at 1100 °C. • Investigation of heat transfer from the melt pool to the outside vault water through calandria vessel. • Numerical analysis to study the capability of calandria vault to remove decay heat from molten pool in calandria vessel. - Abstract: Recent Fukushima nuclear accident has triggered further awareness amongst reactor designers regarding enhancing the safety measures in a nuclear reactor. It has become important to analyse the capability of decay heat removal in a reactor to avoid radioactivity releases to the environment. Such a study has been carried out for an Indian PHWR. In a hypothetical severe core damage accident in PHWR, multiple failure of the core cooling system may lead to collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a debris bed. Due to decay heat generation, the debris ultimately melts down forming a molten pool inside calandria vessel. Calandria vessel is surrounded by calandria vault water that acts as heat sink. In order to study the extent of heat transfer from molten pool to surrounding water under severe accident condition, an experiment was carried out wherein a simulant material was poured inside a simulated calandria vessel immersed in the simulated calandria vault water. The amount of melt and water present in calandria vault scaled proportionately with regard to an Indian 700 MWe PHWR. Results show that as soon as the melt was poured in the vessel, a thick crust was formed on the inner calandria vessel, which reduced the heat transfer from the melt pool to vault water resulting high temperature gradient in melt. Even though the cylindrical vessel inner temperature was found to be very high, the water outside the vessel never boiled. When the cylindrical vessel was opened after the experiment, there was no

  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. CFD analysis of poison injection in AHWR calandria

    International Nuclear Information System (INIS)

    The present work intends to give details of design and performance validation of SDS-2. The performance is evaluated on the basis of dispersion of poison in calandria in a given period of time. Location of injection tube and injection holes, size of jet hole and number of holes are some of the design parameters which greatly affect dispersion of poison in calandria. A Computational Fluid Dynamic (CFD) study for axial and radial injection of poison was carried out using open source CFD code OpenFOAM. CFD benchmarking was done using experiments performed by Johari (Johari et al. 1997) to identify suitable turbulence model for this problem. An experimental facility simulating poison injection in moderator in presence of calandria tubes was used to further validate the CFD model is shown in the paper. CFD analysis was carried out for axial as well as radial injection for AHWR geometry. CFD analysis using OpenFOAM has been carried out to study high pressure poison injection for single jet of Shut Down System - 2 (SDS- 2) of Advanced Heavy Water Reactor (AHWR) for various design options. CFD model used in analysis have been validated with experimental data available in literature as well as experiments performed for AHWR specific geometry. Various turbulence models are tested and their adequacy for such flow problems has been established. The CFD model is then used to simulate poison injection for two design options for AHWR and their performance is compared. (author)

  10. Irradiation of zircaloy calandria tube samples for growth measurement in Dhruva reactor

    International Nuclear Information System (INIS)

    In our PHWRs, the calandria tubes are made of seamwelded zircaloy-2 material which is placed in service in fully annealed condition. The available literature contains considerable information on the irradiation response of seamwelded calandria tubes thereby generating enough confidence towards assuming satisfactory behaviour of seamwelded calandria tubes in the reactor. However it is known that garter spring transmits a load from the pressure tube to the calandria tube and after a few years of reactor operation, creep of the calandria tube is the controlling factor in resisting the sag of the coolant channels. There is a proposal from Nuclear Fuel Complex (NFC), Hyderabad to manufacture the seamless zircaloy calandria tubes in place of welded tubes which will result in increased production of calandria tubes at a much lower cost and will also meet the increased production demand for calandria tubes for future power programme. In view of the necessity to change over from seamwelded to seamless zircaloy calandria tubes, it is very important to generate the irradiation creep, growth and mechanical properties data for seamless and seamwelded zircaloy calandria tubes by irradiating their samples in the reactor. One such irradiation programme has been started in Dhruva reactor where the samples of seamless and seamwelded zircaloy calandria tubes have been loaded in specially designed tiers in a slug rod and the slug rod has been installed in the reactor. Dhruva being a high flux reactor, the irradiation growth behaviour of seamless and seamwelded zircaloy sample can be studied in a much shorter time. This paper describes the details of the zircaloy sample irradiation programme including the design of special zircaloy sample tiers, the zircaloy slug rod and a precision zircaloy irradiation growth measuring jig towards collecting a meaningful zircaloy irradiation growth data. (author). 4 figs

  11. 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

  12. CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

    International Nuclear Information System (INIS)

    Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated

  13. CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kansal, Anuj Kumar, E-mail: akansal@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Maheshwari, Naresh Kumar, E-mail: nmahesh@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Vijayan, Pallippattu Krishnan, E-mail: vijayanp@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2015-06-15

    Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated.

  14. 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

  15. Measurement and computation for sag of calandria tube due to irradiation creep in PHWR

    International Nuclear Information System (INIS)

    Calandria tubes and Liquid Injection Shutdown System(LISS) tubes in a Pressurized Heavy Water Reactor(PHWR) 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 LISS 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 measuring probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sags of both tubes in the PHWR. 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

  16. 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

  17. CFD analysis of flow and temperature distribution inside the calandria of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Passive systems are being examined for the future AHWR reactor designs. One of these systems is the passive moderator cooling system, which removes heat from the moderator in case of a Station Black Out (SBO). The heavy-water moderator gets heated due to the residual heat from the core structures and rises upward due to buoyancy. This is cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator (Calandria) pressure beyond safe limits. In this paper CFD investigations are carried out to study the temperature distributions and flow distribution inside the Calandria using a three-dimensional CFD code, OpenFoam 2.2.0. The results provide a band of operable mass flow rates which are safe for operation by virtue of prediction of hot spots in the Calandria. (author)

  18. 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

  19. Nuclear heat generation race on the walls and floor of calandria vault of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Nuclear heat generation rate profile along the walls and floor of calandria vault is required to analyse its strength during the life time of reactor power operation and to ascertain adequacy of cooling and shielding arrangements. Hence detailed heat generation profile on the face and across the thickness of walls and floor is generated. (author). 7 refs., 1 fig., 3 tabs

  20. Validation of impinging jet models to be used in CANDU calandria vessel CFD simulations

    International Nuclear Information System (INIS)

    The knowledge of the external wall temperature distributions on calandria tubes is of major concern in nuclear safety analysis. One of the models used by the Canadian industry consists in replacing the calandria by an equivalent porous media with appropriate anisotropic hydraulic resistances. This technique has the advantage to treat a non-connected domain as an equivalent quasi-continuous media; however, it cannot provide information about local velocity variations. Within the framework of the present study, a full-scale modeling of the moderator using a Computational Fluid Dynamic code (FLUENT) is underway. The use of a 2D model have shown that the geometry of calandria nozzles have a strong effect on the flow distribution. Some authors suggest to model the flow at the entrance of the calandria as successive flow circulation through a portion of a straight pipe, a curved pipe, and a circular nozzle placed in front of an impinging plate, and to use the results as input data in full-scale calculations. Obtaining these data requires large computational resources before performing complete flow simulations, while they do necessarily represent neither the real geometry nor the actual flow conditions. Therefore, the present study is aimed to find appropriate water-jet modeling approaches that can help in improving moderator circulation simulations. In particular, the principal interest consists in finding a semi-analytical nozzle model that can be used as a constitutive relationship in a CFD code. This approach will contribute both to increase the number of meshes in the calandria vessel as well as to decrease the computational time. (author)

  1. Stress analysis of 500 MWe PHWR calandria non-radial nozzles for external loadings

    International Nuclear Information System (INIS)

    The 500 MWe Pressurised Heavy Water Reactor (PHWR) calandria is a stainless steel horizontal cylindrical vessel which houses the core of the reactor. It has a number of vertical non-radial nozzles which locate shut down, reactivity control and over pressure relief safety devices. The calandria is designed, fabricated and inspected as per ASME Boiler and Pressure Vessel Code class one component stringent requirements. Model studies of this equipment is carried out to know the stress concentration factors for various thermal and seismic loads transferred through these nozzles. In this experiment nozzles lying in one quadrant of calandria are simulated on a 1:3.2 size model and six types of forces are applied on each nozzle. Total 136 numbers of 3 element strain rosettes have been used in this experiment. The strain data are collected through a 100 channel data recording system and stresses on prototype calandria are predicted for various load cases with the help of a BASIC language rosette analysis program implemented on the microprocessor system of the data logger. The interaction effect between various nozzles is also studied. Comparisons have been made between present experimental results and simplified analytical design calculation (based on Bijlaard's method) along with finite element analysis results of a selected group of 3 nozzles. It is shown that local stresses near an opening decay within a short distance from root of the nozzle and interaction effect on neighbouring nozzles surrounding a loaded nozzles is small. The stress concentration factor data generated can be used by the designer to calculate peak and local stresses on the calandria shell-nozzle junctions of various reactivity devices for all six kinds of nominal thermal and seismic loads. (author). 3 refs., tabs., figs

  2. Numerical simulation of poison injection in moderator inside calandria for 220 MWe PHWRs

    International Nuclear Information System (INIS)

    Full text: The motivation of the present work comes from the inability of Secondary Shutdown System (SSS) of 220 MWe PHWRs to make the reactor core subcritical for prolonged durations due to its insufficient worth. The worth and effectiveness of SSS can be increased by replacing the closed shut off tubes of the existing design with perforated shut off or LPI (Liquid Poison Injection) tubes which then facilitates the direct injection of liquid poison into the moderator of calandria. The effectiveness of the new concept depends on how the poison distributes in flowing moderator streams inside calandria within a stipulated time. Hence, the objective of the present study is to understand the evolution of poison distribution in moderator of the calandria after injection of poison through perforated tubes of the new conceptual design of SSS. To achieve the present objectives, numerical simulations were performed using Computational Fluid Dynamic (CFD) techniques. As a validation study, CFD simulations were performed for the case of a single submerged turbulent jet injected in to a stagnant medium and validated with experiments available in the literature. During normal operation, moderator circulation establishes a steady flow profile inside calandria. The new concept of SSS, when called for, injects poison into this established moderator flow profile. Hence, CFD simulations were initially performed to establish the steady state moderator flow pattern inside calandria without poison injection. These results were used as initial conditions for the unsteady CFD simulations of poison injection into the moderator. The poison in the SSS tanks gets pushed into calandria through perforated tubes using high pressure helium gas expansion above the poison surface in poison storage tanks. This process results in a time varying poison velocity at the LPI tube inlets. A one-dimensional unsteady state model was developed to predict the time varying velocity at LPI tube inlets as a

  3. Fabrication of seamless calandria tubes by cold pilgering route using 3-pass and 2-pass schedules

    Science.gov (United States)

    Saibaba, N.

    2008-12-01

    Calandria tube is a large diameter, extremely thin walled zirconium alloy tube which has diameter to wall thickness ratio as high as 90-95. Such tubes are conventionally produced by the 'welded route', which involves extrusion of slabs followed by a series of hot and cold rolling passes, intermediate anneals, press forming of sheets into circular shape and closing the gap by TIG welding. Though pilgering is a well established process for the fabrication of seamless tubes, production of extremely thin walled tubes offers several challenges during pilgering. Nuclear fuel complex (NFC), Hyderabad, has successfully developed a process for the production of Zircaloy-4 calandria tubes by adopting the 'seamless route' which involves hot extrusion of mother blanks followed by three-pass pilgering or two-pass pilgering schedules. This paper deals with standardization of the seamless route processes for fabrication of calandria tubes, comparison between the tubes produced by 2-pass and 3-pass pilgering schedules, role of ultrasonic test charts for control of process parameters, development of new testing methods for burst testing and other properties.

  4. Fabrication of seamless calandria tubes by cold pilgering route using 3-pass and 2-pass schedules

    International Nuclear Information System (INIS)

    Calandria tube is a large diameter, extremely thin walled zirconium alloy tube which has diameter to wall thickness ratio as high as 90-95. Such tubes are conventionally produced by the 'welded route', which involves extrusion of slabs followed by a series of hot and cold rolling passes, intermediate anneals, press forming of sheets into circular shape and closing the gap by TIG welding. Though pilgering is a well established process for the fabrication of seamless tubes, production of extremely thin walled tubes offers several challenges during pilgering. Nuclear fuel complex (NFC), Hyderabad, has successfully developed a process for the production of Zircaloy-4 calandria tubes by adopting the 'seamless route' which involves hot extrusion of mother blanks followed by three-pass pilgering or two-pass pilgering schedules. This paper deals with standardization of the seamless route processes for fabrication of calandria tubes, comparison between the tubes produced by 2-pass and 3-pass pilgering schedules, role of ultrasonic test charts for control of process parameters, development of new testing methods for burst testing and other properties

  5. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

    2015-10-15

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  6. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    International Nuclear Information System (INIS)

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  7. 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

  8. Improving accuracy of ET measurement of LISS nozzle to calandria tube clearance

    Energy Technology Data Exchange (ETDEWEB)

    Craig, S.T.; Krause, T.W.; Schankula, J.J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)]. E-mail: craigs@aecl.ca

    2006-07-01

    The AECL Fuel Channel Inspection System (AFCIS) has been used in an in-reactor field trial to successfully measure the clearance between Liquid Injection Shutdown System (LISS) nozzles and calandria tubes. Each measurement over the full length of a channel added only 15 minutes to the on-channel inspection time. No changes were required to the inspection heads. The only equipment changes made were the addition of a Remote Field Eddy Current (RFEC) module to the eddy current instrument, and minor wiring changes, at the instrument, to achieve a RFEC configuration. With the experience gained from the field trial, factors potentially limiting accuracy were identified. These, and other factors, were investigated and are discussed herein. The RFEC probe is delivered inside the pressure tube. Magnetic fields from the RFEC probe extend through the conducting walls of the pressure tube and calandria tube to interact with the LISS nozzle. Data acquired during the field trial showed the LISS nozzle signal is distinct and the signal-to-noise ratio is very favourable. Nevertheless, comparison of the RFEC measurements to a visual examination, made during the same outage, had the RFEC method underestimating the clearance by 2.5 mm on average. By way of laboratory tests, the following factors were investigated as potential sources of error: resistivity and geometry of LISS nozzle reference/calibration pieces, pressure-tube wall thickness, diameter and resistivity variations, pressure-tube to calandria-tube gap, and radial offsets of the probe within the pressure-tube. The sensitivity to these various noise sources was established. A model, based on fundamental electromagnetic principles, was developed and was used to normalize the effects of LISS nozzle conductivity and geometry. This enabled compensation for various sources of error, and made it possible to produce a correction factor for the field trial data, reducing the average difference from the visual inspection of LISS

  9. 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

  10. Experience in ultrasonic gap measurement between calandria tubes and liquid injection shutdown systems nozzles in Bruce Nuclear Generating Station

    International Nuclear Information System (INIS)

    The gaps between calandria tubes (CT) and Liquid Injection Shutdown System (LISS) nozzles at the Bruce Nuclear Generating Station ''A'' (Bruce A) are known to decrease with time due to radiation induced creep/sag of the calandria tubes. If this gap decreases to a point where the calandria tubes come into contact with the LISS nozzle, the calandria tubes could fail as a result of fretting damage. Proximity measurements were needed to verify the analytical models and ensure that CT/LISS nozzle contact does not occur earlier than predicted. The technique used was originally developed at Ontario Hydro Technologies (formerly Ontario Hydro Research Division) in the late seventies and put into practical use by Research and Productivity Council (RPC) of New Brunswick, who carried out similar measurements at Point Lepreau NGS in 1989 and 1991. The gap measurement was accomplished y inserting an inspection probe, containing four ultrasonic transducers (2 to measure gaps and 2 to check for probe tilt) and a Fredericks electrolytic potentiometer as a probe rotational sensor, inside LISS Nozzle number-sign 7. The ultrasonic measurements were fed to a system computer that was programmed to convert the readings into fully compensated gaps, taking into account moderator heavy water temperature and probe tilt. Since the measured gaps were found to be generally larger than predicted, the time to CT/LISS nozzle contact is now being re-evaluated and the planned LISS nozzle replacement will likely be deferred, resulting in considerable savings

  11. Corrosion survelliance for reactor materials in the calandria vault of pickering NGS a unit 1

    International Nuclear Information System (INIS)

    This paper presents selected results from an 18 month period of corrosion surveillance for structural materials inside the calandria vault of Pickering NGS Unit 'A'. Cooling pipe leaks have resulted in humidity build up inside the air-filled vault, and subsequent radiolysis of the water and air by the high gamma radiation field from the operating reactor results in the formation of nitric acid. Condensation of the nitric acid on cooler components (pipes, support brackets) promotes general corrosion and possibly localized corrosion of carbon steel structures. The corrosion surveillance system was installed to directly monitor changes in corrosion rates of selected materials. One period has been chosen in which, due to a bioshield cooling water leak, the corrosion rates increased

  12. Effect of processing on properties of thin walled calandria tubes for pressurised heavy water reactor

    Science.gov (United States)

    Kapoor, K.; Padmaprabu, C.; Ramana Rao, S. V.; Sanyal, T.; Kashyap, B. P.

    2003-02-01

    Thin walled calandria tubes for pressurised heavy water reactors are manufactured either by seam welding of Zircaloy-4 sheets or by seamless route. In the present study, the effect of processing on the critical properties such as texture, microstructure, hydriding behaviour and residual stress for both the routes as well as the mechanical anisotropy developed due to seam welding are investigated. The properties of the seam welded tube in the fusion and adjoining region are markedly different from the base material and from the seamless tube. Residual stress measurements indicate that heat affected zone (HAZ) of seam welded tubes have longitudinal tensile residual stress and the seamless tubes have uniform compressive stress along the circumference. The phase transition in the presence of residual stresses due to thermal gradient is found to modify the texture in the HAZ. The hydride orientation and mechanical anisotropy in these regions are found to be dependent on the texture of the material.

  13. Prediction of moderator temperature distribution inside CANDU calandria and its effect on criticality and safety

    International Nuclear Information System (INIS)

    A numerical study for the prediction of the moderator flow and temperature distribution inside a CANDU reactor vessel is carried out. The porous medium concept which is usually used to account for the effect of the calandria tubes matrix on the flow pattern is avoided. Instead, the tube matrix is simulated by blocking the cells containing the tubes in the finite difference mesh. A jet momentum dominated flow pattern is predicted with two main hot spots regions. The calculated moderator temperature distribution is used to credit the moderator as an ultimate heat sink, and to study the effects of the changes in the moderator temperature on reactor criticality. The multiplication factor is calculated using a refined version of the four-factor formula at different moderator temperatures. Consequently the temperature coefficient of reactivity is calculated

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

    Science.gov (United States)

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

    2015-03-01

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

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

    International Nuclear Information System (INIS)

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

  16. Effect of processing on properties of thin walled calandria tubes for pressurised heavy water reactor

    International Nuclear Information System (INIS)

    Thin walled calandria tubes for pressurised heavy water reactors are manufactured either by seam welding of Zircaloy-4 sheets or by seamless route. In the present study, the effect of processing on the critical properties such as texture, microstructure, hydriding behaviour and residual stress for both the routes as well as the mechanical anisotropy developed due to seam welding are investigated. The properties of the seam welded tube in the fusion and adjoining region are markedly different from the base material and from the seamless tube. Residual stress measurements indicate that heat affected zone (HAZ) of seam welded tubes have longitudinal tensile residual stress and the seamless tubes have uniform compressive stress along the circumference. The phase transition in the presence of residual stresses due to thermal gradient is found to modify the texture in the HAZ. The hydride orientation and mechanical anisotropy in these regions are found to be dependent on the texture of the material

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

    Energy Technology Data Exchange (ETDEWEB)

    Katchadjian, Pablo, E-mail: katcha@cnea.gov.ar; Desimone, Carlos, E-mail: katcha@cnea.gov.ar; Garcia, Alejandro, E-mail: katcha@cnea.gov.ar [Comisión Nacional de Energía Atómica, Depto. ENDE - INEND, Av. Gral. Paz 1499, Buenos Aires (Argentina); Antonaccio, Carlos; Schroeter, Fernando; Molina, Héctor [Nucleoeléctrica Argentina-SA, Arribeños 3619, Buenos Aires (Argentina)

    2015-03-31

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

  18. Coupled response of calandria shell and coolant channels in the event of a single coolant channel failure in a PHWR

    International Nuclear Information System (INIS)

    One of the important design basis accidents in a pressurized heavy water reactor (PHWR) is the failure of a coolant channel inside the calandria. This results in introduction of high pressure coolant into a low pressure moderator system. In order to demonstrate the integrity of core components and calandria shell it is important to consider the effect of this impulsive loading in a realistic manner. The problem becomes complex due to the presence of a large number of submerged tubes. Moreover it is important to consider the dynamic behavior of these submerged tubes in a coupled manner as motion of one tube may excite the other neighboring tubes due to fluid coupling effect. The other important aspect of this problem is the consideration of wave reflection effects between the neighboring tubes which results in a complex pressure gradient in the vicinity of an accident. The present paper demonstrates the capabilities of two dimensional (2-D) code FLUSOL and three dimensional (3-D) code FLUSHEL to tackle the above fluid-structure interaction problem. The transient analyses carried out with these codes bring out some important conclusions regarding the safety of channels and calandria shell near the accident site for such safety related problems

  19. Ultrasonic measurement of gap between calandria tube and liquid injection shutdown system tube in PHWR

    International Nuclear Information System (INIS)

    Sag of CT or liquid injection shutdown system tubes in pressurized heavy water reactor is known to occur due to irradiation creep and growth during plant operation. When the sag of CT is big enough, the CT tube possibly comes in contact with liquid injection shutdown system tube (LIN) crossing beneath the CT, which subsequently may prevent the safe operation. It is therefore necessary to check the gap between the two tubes in order to confirm no contacts when using a proper measure periodically during the plant life. An ultrasonic gap measuring probe assembly which can be fed through viewing port installed on the calandria was developed and utilized to measure the sags of both tubes in a pressurized heavy water reactor in Korea. It was found that the centerlines of CT and LIN can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. But the measured gap data observed at the viewing port were actually not the data at the crossing point of CT and LIN. To get the actual gap between two tubes, mathematical modeling for the deflection curves of two tubes was used. The sags of CT and LIN tubes were also obtained by comparison of the present centerlines with the initial elevations at the beginning of plant operation. The gaps between two tubes in the unmeasurable regions were calculated based on the measurement data and the channel power distribution

  20. A study on the effect of rise in calandria vault cooling water and concrete temperature during system degraded cooling and station black out conditions in 540 MWe Indian PHWRs

    International Nuclear Information System (INIS)

    Full text of publication follows: The calandria vault houses the calandria vessel which contains heavy water moderator and natural UO2 in number of coolant tubes. Demineralized water is circulated through calandria vault housing in Indian PHWRs to remove part of nuclear heat produced in calandria and to cool the concrete structure. It also provides shielding against neutron and gamma radiation. The calandria vault concrete structure acts as biological shield. The calandria vault cooling system consists of 3 pumps (normally 2 working and 1 standby) and 2 plate type heat exchangers to remove the heat generation due to attenuation of gamma rays. The calandria vault wall is also subjected to heat load, the contribution being mostly from attenuation of gammas. It is of interest to estimate the temperature distribution of the wall to evaluate whether concrete peak temperature as well as differential temperature are within acceptable limits. An analysis was done to estimate the temperature rise of the calandria vault water and calandria vault concrete during degraded cooling (1 pump and 1 heat exchanger)of the system and Station Black Out (Off site and On site power failure). During Station Black Out forced circulation in calandria vault cooling system is not available, so the heat loss is to the Reactor Building atmosphere through calandria vault wall concrete. For estimation of the concrete temperature, two-dimensional finite difference explicit scheme was used with time dependent and spatial variation in heat generation and convective boundary conditions were considered. Calandria vault concrete is in contact with the calandria vault water through stainless steel liner and 3 mm air gap. Calandria vault water temperature distribution in the flow direction was predicted in the above mentioned events by considering the heat gain from the neighbouring systems and heat generation in water. Outer surface of calandria vault wall is in contact with Reactor building atmosphere

  1. 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)

  2. 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)

  3. 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

  4. 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.

  5. 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)

  6. Sensitivity study on a CFD model for the analysis of moderator fluid flow and heat transfer inside CANDU calandria

    International Nuclear Information System (INIS)

    In this paper, sensitivity study on a CFD model for the accurate analysis of moderator fluid flow and heat transfer inside calandria is conducted. Two main items, i.e. porous medium assumption and turbulence model, are considered for in-line tube bank and Sheridan Park Engineering Laboratory (SPEL) experiment. Using the commercial flow solver, FLUENT, the prediction to consider the real geometry of fuel channels is compared to previous results conducted with the porous medium assumption using the isotropic pressure loss model. Also, the prediction performance of various turbulence models (e.g. k-ε model, Reynolds stress model, etc.) is assessed. (author)

  7. The transition criteria of circulating flow pattern of moderator in the calandria tank of CANDU nuclear power plant

    International Nuclear Information System (INIS)

    The moderator cooling system to the Calandria tank of CANDU nuclear power plant provides an alternative pass of heat sink during the hypothetical loss of coolant accident. Also, the neutron population in the CANDU plant can be affected by the moderator temperature change which strongly depends on the circulating flow pattern in the Calandria tank. It has been known that there are three distinguished flow patterns: the buoyancy dominated flow, the momentum dominated flow, and the mixed type flow. The Canadian Nuclear Safety Commission (CNSC) recommended that a series of experimental works should be performed to verify the three dimensional codes. Two existing facilities, SPEL (1982) and STERN (1990), have produced experimental data for these purposes. The present work is also motivated to build up a new scaled experimental facility named HGU for the same purposes. CANDU-6 was selected as the target plant to be scaled down. In the design for the scaled facility, the knowledge on the flow regime transitions in the circulating flow was imperative. In the present study, to pave the way for the scaling, the flow pattern maps of circulating flow were constructed based on the Reynolds number and Archimedes number. The CFX code was employed with real meshes to represent all calandria tubes in the tank. The flow pattern maps were constructed for SPEL, STERN, HGU, and CANDU6. As the key transition criterion useful for scaling law, a new Archimedes number considering the jet impingement of the feed water in the Calandria tank was found. The transition of flow patterns was made with the same Archimedes number for CANDU6, STERN and HGU. However, SPEL which has third of the modified Archimedes number showed different maps in the wider region of mixed flow pattern was observed. It was found that the Archimedes number considering the inlet nozzle velocity plays the key role in patterns classification. Also, it can be suggested that the moderator cooling system needs to be designed

  8. 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)

  9. 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

  10. 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)

  11. 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

  12. Numerical Analysis on the Calandria Tubes in the Moderator of a Heavy Water Reactor Using OpenFOAM and Other Codes

    Science.gov (United States)

    Chang, Se-Myong; Kim, Hyoung Tae

    2014-06-01

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors.

  13. Numerical analysis on the calandria tubes in the moderator of a heavy water reactor using OpenFOAM and other codes

    International Nuclear Information System (INIS)

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors. (authors)

  14. Development Of NRU Reflector Wall Inspection System

    OpenAIRE

    Simpson, N.; B.V. Luloff; N. Zahn; R. Lumsden

    2013-01-01

    In 2009 May, the National Research Universal (NRU) calandria leaked. During the next year, the calandria was inspected with six new Non-Destructive Evaluation (NDE) techniques to determine the extent of the corrosion, repaired, and finally the repair was inspected with four additional new NDE techniques before the reactor was returned to service.The calandria is surrounded by a light-water reflector vessel fabricated from the same material as the calandria vessel. Concerns that the same corro...

  15. Study of impulsive pressure response in a tank

    International Nuclear Information System (INIS)

    The Advanced Thermal Reactor (ATR) is a boiling-light-water-cooled heavy-water-moderated pressure-tube-type reactor. If simultaneous break of the pressure tube and of calandria tube were to occur resulting from a severe accident, high-pressure and high-temperature coolant would be discharged into the cool moderator. Although the moderator is cool enough to condense the evaporated vapor from the view point of a long term behavior, an impulsive pressure may occur due to the high-energy liquid ejection into the low-energy liquid in the case of no existence of water surface in the calandria vessel. The integrity of the calandria vessel was confirmed using a full-scale facility that had a water surface in the vessel. In the experiment, pressure in the vessel surface was about 0.3 MPa due to the pressure absorption by gas layer above the heavy water surface. A rupture disk is provided at the calandria vessel to prevent over-pressure. To investigate the response of the calandria vessel without any water surface, experiments with explosive compounds were conducted with a 1/6-scale calandria vessel. A calculation method with the AUTODYN-2D code has been studied to establish the evaluation method at the real situation. Main experimental parameters were the thickness of the calandria vessel and calandria tubes that simulate real flexibility or rigid cases under the condition of 1/6-scale set-up. When the calandria vessel was flexible and the calandria tube was thin to simulate rigidity of the real calandria tube of 1.9 mm in thickness made of zircaloy-2, pressure at the surface of calandria vessel was almost atmospheric pressure. It is clarified that amount of the attenuation of the impulsive pressure was dependent mainly on elastic deformation of the calandria vessel. The AUTODYN-2D code could predict the pressure attenuation characteristic by a flexible calandria vessel

  16. Dimensionamiento de una planta de digestión anaerobia, en fases separadas para su incorporación en el centro integral de tratamiento de residuos "Las Calandrias"

    OpenAIRE

    Cano Santana, Patricio Iván

    2013-01-01

    La legislación vigente en el campo del tratamiento de residuos aboga por la valorización de los residuos. El Centro Integral de Tratamiento de Residuos “Las Calandrias” es una planta de residuos que se encuentra en el término municipal de Jerez de la Frontera y que se encarga del tratamiento de los residuos de Jerez de la Frontera y su entorno. En el Centro Integral de Tratamiento de Residuos “Las Calandrias” se produce la recuperación y el reciclaje de los residuos y la valori...

  17. 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.

  18. Effect of flooding of annulus space between CT and PT with light water coolant and heavy water moderator on AHWR reactor physics parameters

    International Nuclear Information System (INIS)

    In AHWR lattice, the pressure tube (PT) contains light water coolant which carries away heat generated in the fuel pins. The pressure tube (PT) and calandria tube (CT) are separated by air (density=0.0014 g/cc) of wall thickness 1.79 cm. Air between pressure tube and calandria tube acts as insulator and minimize the heat transfer from coolant to moderator which is outside the calandria tube. In case of flooding or under any unforeseeable circumstances, the air gap between the coolant tube and calandria tube may be filled with the light water coolant or heavy water moderator. This paper gives the details of effect of filling the annulus space between CT and PT with light water or heavy water moderator on reactor physics parameters. (author)

  19. 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

  20. Effect of distribution of volumetric heat generation on moderator temperature distribution

    International Nuclear Information System (INIS)

    Three dimensional computational fluid dynamics (CFD) analysis has been performed to study the effect of distribution of volumetric heat generation on moderator temperature distribution for the moderator flow and temperature distribution inside the Calandria vessel. Moderator system of Advanced Heavy Water Reactor (AHWR) under normal operating condition is taken as case study. CFD code OpenFOAM is used in the analysis, which is validation with the related experimental data. CFD model includes the Calandria vessel, Calandria tubes, inlet header and outlet header. Analysis has been performed with uniform and spatial distribution volumetric heat generation. Studies show that the maximum temperature in moderator is lower in case of spatial distribution of heat generation as compared to uniform heat generation in Calandria. (author)

  1. 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.

  2. Studies of loss-of-coolant and loss-of-regulation accidents

    International Nuclear Information System (INIS)

    Studies of a CANDU reactor during loss of coolant with delayed emergency core cooling showed that the moderator is an effective heat sink, and that in reactors with moderator dump the calandria sprays provide effective cooling. Fuel channel melting would not occur, and a coolable geometry will be maintained. Studies on film cooling and film stability on calandria tubes and on the analysis of flow reversal in vertical feeder tubes are also reported

  3. Leak before break detection-annulus gas monitoring system evolution and operating experience at KGS

    International Nuclear Information System (INIS)

    Full text: Pressurised heavy water reactors (PHWR) at RAPS 1 and 2 and MAPS have provision for detection of pressure tube leak by indirect method. The reactor vessel (calandria) is housed in calandria vault (C/V) filled with air and C/V moisture element indicates the water leak from calandria tube or pressure tube. Further, detection of leak is a cumbersome process. From NAPS onwards, calandria is housed in C/V filled with water, annulus between calandria tube and pressure tube is filled with CO2 and annulus gas monitoring system (AGMS) is provided by design for detection of any pressure tube leak. The design was improved and AGMS for Kaiga 1 and 2 and RAPS 3 and 4 is having re-circulation mode of operation. The design provides for monitoring dew point of annulus gas (CO2) for indicating the leak and later to identify the pressure tube/calandria tube having leak. The paper deals with operating experience of AGMS at Kaiga generating station (KGS). During the commissioning and initial power operation at KGS, problems were encountered in re-circulation mode. These problems were high radiation field near AGMS piping, high temperature on blower body, blower bearing failure and system leaks. Design modifications were carried out for effective performance of the system for detecting leak before break

  4. 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)

  5. NRU light water reflector inspections

    International Nuclear Information System (INIS)

    In 2009 May, the National Research Universal (NRU) calandria leaked. In the next year the calandria was inspected with six new techniques, repaired, and the repair inspected with four additional new techniques. The calandria is surrounded by a light water reflector vessel. Concerns that the same corrosion mechanism may have damaged the reflector vessel led to plans to inspect the full circumference of the reflector wall for corrosion damage, through a 64 mm diameter port, from 10 m away, while inspecting from the corroded surface. The ultrasonic technique is intended to produce a very fine inspection grid, through the corroded surface, using a very small probe body over the approximately 5 m2 to be inspected. The calandria vessel inspections were successfully performed, in a short time period, under difficult conditions. This paper will discuss Operating Experience (OPEX) from the 2009/2010 calandria wall inspections and industry dialogue that were applied to the Reflector Wall Inspection (RWI) project. Interaction with the CANDU Inspection Qualification Bureau (CIQB) on procedures, training plans, and mock-up training, led to improvements in how these activities were performed for the RWI project. Experience with shift turnover meetings led to improvements in how the inspection team captures lessons learned, trends issues, and eliminates reoccurring design and procedure problems. Experience with collected and analyzed data, inspection shift logs, and video captured from inspections and repairs, were used to improve inspection performance. Finally, OPEX from radiation/activation and the calibration process reduced dose through tool design and procedure changes. (author)

  6. Comparative Study for Modeling Reactor Internal Geometry in CFD Simulation of PHWR Internal Flow

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Gong Hee; Woo, Sweng Woong; Cheong, Ae Ju [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    The main objective of the present study is to compare the results predicted by using either the real geometry of tubes or porous medium assumption and to assess the prediction performance of both methods. Estimating the local subcooling of the moderator in a CANDU calandria under transient conditions is one of the major concerns in the CANDU safety analysis. Therefore extensive CFD analyses have been performed for predicting the moderator temperature in a CANDU calandria or its similar shape. However most of previous studies used a porous medium assumption instead of considering the real geometry of calandria tube. A porous medium assumption has some possible weaknesses; The increased production of turbulence due to vortex shedding in the wake of the individual tubes is not considered in the turbulence model. It is difficult to identify the true effects of the outer ring of calandria tubes on the generation of the highly non-uniform flows in the reflector region. It is not clear how well the pressure loss models quantitatively represent the three-dimensional effects of the turbulent flows through the calandria tubes.

  7. Numerical analysis of debris melting phenomena during late phase CANDU 6 severe accident

    International Nuclear Information System (INIS)

    Highlights: ► We used Ansys Fluent to simulate debris melt phenomena for a CANDU reactor. ► Two models were employed: Large Eddy Simulation and PECM tool. ► The PECM approach is comparable within 10% with LES model. ► The debris can be well contained inside the calandria during a severe accident. - Abstract: The objective of this paper was to study the phase change of the debris formed on the CANDU 6 calandria bottom in a postulated accident sequence. The molten pool and crust formation were studied employing the Ansys Fluent code. Two models using Large Eddy Simulation and Phase Change Effective Convectivity Model (PECM) predict the conjugate, radiative and convective heat transfer inside and from the corium pool. LES require a very fine grid to capture the crust formation and the free convection flow. This aspect (fine mesh requirement) correlated with the long transient has imposed the use of a slice from the 3D calandria geometry in order not to exceed the computing resources. From the safety point of view it is very important to maintain a heat flux through the wall below the critical heat flux assuring the integrity of the calandria vessel, thus the paper results include temperature profiles and heat fluxes through calandria wall. The results predicted by the PECM approach are comparable within 10% with those obtained by the computational fluid dynamics method, proving that the PECM can be used to perform further sensitivity studies.

  8. Experimental investigation of thermal behavior of concentric tubes during a severe accident

    International Nuclear Information System (INIS)

    A pair of experiments were conceived and executed to provide data and a technical basis for investigating selected aspects of postulated severe accidents in a pressure tube/calandria tube configuration. The response to core damage and debris relocation within the pressure tube was investigated experimentally. The experimental objectives of the two tests were: 1) to assess the potential for failure of an unflawed pair of concentric tubes when prototypic wall stress is produced while high temperature debris is resident within the inner tube and sub-cooled water is present outside the outer tube, and 2) to assess the dynamic and energetic interaction given the rupture of the concentric tubes and the discharge of molten debris under steam pressure into the surrounding sub-cooled water pool. These experiments provide an effective demonstration of the passive cooling mechanism which can prevent calandria tube failure and of the interaction between molten debris and water if a calandria tube were to fail. (author)

  9. 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.

  10. KAPP-3 and 4 containment pressure following postulated severe accident along with SAMG implementation

    International Nuclear Information System (INIS)

    Containment is an ultimate safety barrier which is designed to enclose whole reactor systems and to prevent the spread of active air-borne fission products. Studies are done to access its performance following severe accident i.e. Loss of Coolant Accident (LOCA) along with failure of Emergency Core Cooling System (ECCS), moderator and calandria vault water cooling system. The accident progression begins with the double ended break in reactor outlet/inlet header with simultaneous failure of ECCS followed by failure of moderator and calandria vault water cooling system. Initially decay heat and metal water reaction energy are assumed to be added to moderator water resulting in boiling of moderator and re-pressurization of containment due to steam addition. Subsequent to moderator boiling, decay heat and metal water reaction energy are assumed to be added to calandria vault water resulting in boiling and re-pressurization of containment due to steam addition. After moderator and calandria vault water have completely boiled off, rapid hydrogen generation would take place due to oxidation of pressure tubes and calandria tubes. In such accident scenario, the core is severely damaged. It will also lead to release of a large quantity of radio nuclides to containment atmosphere. To arrest the progression of accident, which can result in Severe Core damage and large amount of hydrogen production, which could leads to containment failure due to hydrogen deflagration or detonation, application of Severe Accident Management Guidelines (SAMG) has been studied. SAMG involve addition of water to calandria and calandria vault. It would result the boiling of the added water and consequent pressurization of containment. This paper presents the analysis for pressure-temperature of KAPP-3 and 4 containment following the postulated accident along with the application of Severe Accident Management Guidelines (SAMG). SAMG initiated action helps in arresting the progression of core

  11. 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

  12. Welding with the TIG automatic process of the end fittings for the execution of the Embalse nuclear power plant fuel channel rechange

    International Nuclear Information System (INIS)

    The present work describes the methodology for the cutting of the existing welding and subsequent welding applied by the TIG process of the coupling composed by the shroud ring and the end fitting ring from one of Embalse nuclear power plant's fuel channels. The replacement will be previously determined by the SLAR-ETTE mechanism where a displacement operated among the Gartner Spring rings, the pressure tubes are separated from the Calandria tubes. The welding to be carried out has the function of stamping the CO2 annular gas (thermal insulator) circulating between the pressure tube and the Calandria one during the functioning of the plant. (Author)

  13. Preliminary severe accident management strategies for Wolsong nuclear power plants

    International Nuclear Information System (INIS)

    Severe accident management strategies for Wolsong 2,3,4 Nuclear Power Plants are presented. The defense in depth concept, which limits release of radioactive materials out of containment building, is applied to develop these strategies. These strategies are actions to prevent or to mitigate core damage, rupture of calandria vessel, rupture of calandria vault, rupture of containment building, and release of radioactive materials. These strategies are deduced from the results of level 2 PSA for Wolsong NPPs. These preliminary results will be assessed further and proved to be effective to Wolsong Plants. Then these severe accident management strategies can be used to develop severe accident management program for Wolsong NPPs

  14. CFD Application and OpenFOAM on the 2-D Model for the Moderator System of Heavy-Water Reactors

    International Nuclear Information System (INIS)

    The flow in the complex pipeline system in a calandria tank of CANDU reactor is transported through the distribution of heat sources, which also exerts the pressure drop to the coolant flow. So the phenomena should be considered as multi-physics both in the viewpoints of heat transfer and fluid dynamics. In this study, we have modeled the calandria tank system as two-dimensional simplified one preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD(computational fluid dynamics) methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors

  15. The impact of long-lived particulates on the Pickering NGS retubing program

    International Nuclear Information System (INIS)

    During retubing of Pickering 1 and 2, contamination by carbon-14 particulate was discovered. The source of the carbon-14 was activation of the nitrogen in the annular space between the calandria tubes and pressure tubes. This paper discusses the origin and nature of the particulate hazard, workplace hazard controls, protective equipment, contamination monitoring, and dosimetry for carbon-14

  16. Program management for spring location and repositioning (SLAR) operations at Pickering NGS

    International Nuclear Information System (INIS)

    In 1988 a major project was planned for Units 5 and 6 of Pickering NGS (Nuclear Generating Station). The project involved remotely locating and repositioning the four spacer (garter) springs separating each pressure tube from the surrounding calandria tube. This paper describes the requirements for SLAR, and the work being undertaken at Pickering NGS to implement the project

  17. Welding procedures used in the fabrication of fuel elements for the DON Reactor exponential experiment

    International Nuclear Information System (INIS)

    This exponential experiment required 74 units (37 loaded with UO2 and 37 with UC) to simulate the Reactor fuel channels. Each unit was enclosed in a tube similar to the calandria ones. It contained the pressure tube, the shroud and the 19 rods cluster. Within the pressure tube, in touch with the elements, was the organic liquid. (Author)

  18. Shock wave effects on shut-off-rod guide tubes

    International Nuclear Information System (INIS)

    Experimental and analytical investigations were made of the sudden loading of calandria and shut-off-rod guide tubes by an incident shock wave in water. The objective was to assess the possibility of tube collapse due to passage of a shock wave generated by a hypothesized pressure tube rupture in a pressure tube reactor. At the highest experimental shock strength (approximately 8 MPa) the measured deformation of zircaloy guide and calandria tubes was elastic even when the latter were air-filled. However, air-filled aluminium calandria tubes buckled at about 1 MPa shock strength. Numerical analysis, using an explicit finite difference code suitable for fluid-structure interaction was applied to unperforated fluid-filled tubes, for a shock strength of 10 MPa. This simulation showed that the shock accelerates the tube quickly to the velocity of the water so that no elastic deformation occurs. It appears that shut-off-rod guide tubes are unlikely to be collapsed by passage of a shock wave arising from a single pressure tube rupture. An adjacent calandria tube, being gas filled, would be more vulnerable to collapse though this was not demonstrated with zircaloy tubes

  19. Welding procedures used in the fabrication of fuel elements for the DON Reactor exponential experiment; La soldadura en la fabricacion de elementos combustibles destinados a una experiencia exponencial

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Beltran, A.; Jaraiz Franco, E.; Rivas Diaz, M. de las

    1965-07-01

    This exponential experiment required 74 units (37 loaded with UO{sub 2} and 37 with UC) to simulate the Reactor fuel channels. Each unit was enclosed in a tube similar to the calandria ones. It contained the pressure tube, the shroud and the 19 rods cluster. Within the pressure tube, in touch with the elements, was the organic liquid. (Author)

  20. Development of NRU reflector wall inspection system

    International Nuclear Information System (INIS)

    In 2009 May, the National Research Universal (NRU) calandria leaked. During the next year, the calandria was inspected with six new Non-Destructive Evaluation (NDE) techniques to determine the extent of the corrosion, repaired, and finally the repair was inspected with four additional new NDE techniques before the reactor was returned to service. The calandria is surrounded by a light-water reflector vessel fabricated from the same material as the calandria vessel. Concerns that the same corrosion mechanism had damaged the reflector vessel led to the development of a system to inspect the full circumference of the reflector wall for corrosion damage. The inspection region could only be accessed through 64 mm diameter ports, was 10 m below the port, and had to be inspected from the corroded surface. The ultrasonic technique was designed to produce a closely spaced wall thickness (WT) grid over an area of approximately 5 m2 on the corroded surface using a very small probe holder. This paper describes the Reflector Wall Inspection (RWI) development project and the system that resulted. (author)

  1. Development of NRU reflector wall inspection system

    Energy Technology Data Exchange (ETDEWEB)

    Lumsden, R.H.; Luloff, B.V.; Zahn, N.; Simpson, N., E-mail: lumsdenr@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-06-15

    In 2009 May, the National Research Universal (NRU) calandria leaked. During the next year, the calandria was inspected with six new Non-Destructive Evaluation (NDE) techniques to determine the extent of the corrosion, repaired, and finally the repair was inspected with four additional new NDE techniques before the reactor was returned to service. The calandria is surrounded by a light-water reflector vessel fabricated from the same material as the calandria vessel. Concerns that the same corrosion mechanism had damaged the reflector vessel led to the development of a system to inspect the full circumference of the reflector wall for corrosion damage. The inspection region could only be accessed through 64 mm diameter ports, was 10 m below the port, and had to be inspected from the corroded surface. The ultrasonic technique was designed to produce a closely spaced wall thickness (WT) grid over an area of approximately 5 m2 on the corroded surface using a very small probe holder. This paper describes the Reflector Wall Inspection (RWI) development project and the system that resulted. (author)

  2. Conceptual Study on Dismantling of CANDU Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woo-Tae; Lee, Sang-Guk [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    In this paper, we reviewed 3D design model of the CANDU type reactor and suggested feasible cutting scheme. The structure of CANDU nuclear reactor, the calandria assembly was reviewed using 3-D CAD model for future decommissioning. Through the schematic diagram of CANDU nuclear power plant, we identified the differences between PWR and CANDU reactor assembly. Method of dismantling the fuel channels from the calandria assembly was suggested. Custom made cutter is recommended to cut all the fuel channels. The calandria vessel is recommended to be cut by band saw or plasma torch. After removal of the fuel channels, it was assumed that radiation level near the calandria vessel is not very high. For cutting of the end shields, various methods such as band saw, plasma torch, CAMC could be used. The choice of a specific method is largely dependent on radiological environment. Finally, method of cutting the embedment rings is considered. As we assume that operators could cut the rings without much radiation exposure, various industrial cutting methods are suggested to be applied. From the above reviews, we could conclude that decommissioning of CANDU reactor is relatively easy compared to that of PWR reactor. Technologies developed from PWR reactor decommissioning could be applied to CANDU reactor dismantling.

  3. 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

  4. 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

  5. CFD modeling of debris melting phenomena during late phase Candu 6 severe accident

    International Nuclear Information System (INIS)

    The objective of this paper was to study the phase change of the debris formed on the Candu 6 calandria bottom in a postulated accident sequence. The molten pool and crust formation were studied employing the Ansys-Fluent code. The 3D model using Large Eddy Simulation (LES) predicts the conjugate, radiative and convective heat transfer inside and from the corium pool. LES simulations require a very fine grid to capture the crust formation and the free convection flow. This aspect (fine mesh requirement) correlated with the long transient has imposed the use of a slice from the 3D calandria geometry in order not to exceed the computing resources. The preliminary results include heat transfer coefficients, temperature profiles and heat fluxes through calandria wall. From the safety point of view it is very important to maintain a heat flux through the wall below the CHF assuring the integrity of the calandria vessel. This can be achieved by proper cooling of the tank water which contains the vessel. Also, transient duration can be estimated being important in developing guidelines for severe accidents management. The debris physical structure and material properties have large uncertainties in the temperature range of interest. Thus, further sensitivity studies should be carried out in order to better understand the influence of these parameters on this complex phenomenon. (authors)

  6. 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.)

  7. Evaluation of ductility of zircaloy-2 materials using a small ellipsoidal-shaped punch

    International Nuclear Information System (INIS)

    Seam-welded Zircaloy-2 calandria tubes are used in the fuel channel of CANDU reactors, to isolate the hot pressure tube from the cool heavy-water moderator. The mechanical properties of irradiated welds are required to assess whether they will be satisfactory to the end-of-life of the reactor. It is impractical to irradiate short sections of calandria tube for burst testing. Therefore, we have developed a small ellipsoidal-shaped punch to evaluate the mechanical properties of anisotropic Zircaloy-2 material having various microstructures such as those present in a welded calandria tube. Small punch tests were performed on four different types of calandria tubes, two welded and two seamless, and the results show that the mechanical anisotropy of the tubes could be differentiated. The fracture produced in unirradiated small specimens showed a straight-line crack, similar to a straight-line rupture on a tube during burst testing. A re-constitution of the properties from small specimens was used for correlation with the burst properties of the tubes. A consistent correlation was obtained for the punch load and displacement with the burst strength and ductility of the tubes. Based on this correlation, mechanical properties of irradiated welds using the small ellipsoidal punch method are being evaluated. (author)

  8. Loss-of-coolant accidents with impaired emergency coolant injection

    International Nuclear Information System (INIS)

    This report describes work on loss-of-coolant accidents in CANDU reactors in which emergency coolant injection is indefinitely delayed. Two situations are considered, one in which additional heat sinks (moderator, moderator cooling system, etc.) are available and effective and the other in which they are not. For the situation in which additional heat sinks are available, the computer program IMPECC has been used for the analysis. Certain corrections and modifications to IMPECC are described and recommendations on the maximum time steps to use are given. For CANDU reactors in which calandria tubes remain submerged, should local CHF occur when a pressure tube sags onto a calandria tube, which is highly improbable, the dry patch caused would not spread beyond 100 and rewetting would occur rapidly. Calandria tube and pressure tube integrity will almost certainly be maintained. This conclusion is not affected even if the contact strip between pressure and calandria tubes is very wide, which is unlikely. For the situation in which additional heat sinks are not available, a simplified event tree has been developed to examine possible accident sequences. Two accident sequences have been selected for study: loss of moderator cooling system and loss of moderator heat sink. A detailed description of the analytical steps for the loss of moderator cooling system accident sequence is given

  9. Heat exchanger for a contaminated fluid

    International Nuclear Information System (INIS)

    A heat exchanger, in particular for a contaminated fluid in the nuclear industry. The tubes forming the tube core are welded and crimped across the whole width of the tubular plate which defines the floating head together with the sealing cover, and said tubular plate is also welded and crimped to the calandria along the whole of its periphery. (author)

  10. NRU light water reflector inspections

    Energy Technology Data Exchange (ETDEWEB)

    Lumsden, R.H.; Hebert, H.; Zahn, N.; Simpson, N.W. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2011-07-01

    In 2009 May, the National Research Universal (NRU) calandria leaked. In the next year the calandria was inspected with six new techniques, repaired, and the repair inspected with four additional new techniques. The calandria is surrounded by a light water reflector vessel. Concerns that the same corrosion mechanism may have damaged the reflector vessel led to plans to inspect the full circumference of the reflector wall for corrosion damage, through a 64 mm diameter port, from 10 m away, while inspecting from the corroded surface. The ultrasonic technique is intended to produce a very fine inspection grid, through the corroded surface, using a very small probe body over the approximately 5 m{sup 2} to be inspected. The calandria vessel inspections were successfully performed, in a short time period, under difficult conditions. This paper will discuss Operating Experience (OPEX) from the 2009/2010 calandria wall inspections and industry dialogue that were applied to the Reflector Wall Inspection (RWI) project. Interaction with the CANDU Inspection Qualification Bureau (CIQB) on procedures, training plans, and mock-up training, led to improvements in how these activities were performed for the RWI project. Experience with shift turnover meetings led to improvements in how the inspection team captures lessons learned, trends issues, and eliminates reoccurring design and procedure problems. Experience with collected and analyzed data, inspection shift logs, and video captured from inspections and repairs, were used to improve inspection performance. Finally, OPEX from radiation/activation and the calibration process reduced dose through tool design and procedure changes. (author)

  11. Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report. Vol. 1

    International Nuclear Information System (INIS)

    This report gives the results of a study of the thermo-hydraulic aspects of severe accident sequences in CANDU reactors. The accident sequences considered are the loss of the moderator cooling system and the loss of the moderator heat sink, each following a large loss-of-coolant accident accompanied by loss of emergency coolant injection. Factors considered include expulsion and boil-off of the moderator, uncovery, overheating and disintegration of the fuel channels, quenching of channel debris, re-heating of channel debris following complete moderator expulsion, formation and possible boiling of a molten pool of core debris and the effectiveness of the cooling of the calandria wall by the shield tank water during the accident sequences. The effects of these accident sequences on the reactor containment are also considered. Results show that there would be no gross melting of fuel during moderator expulsion from the calandria, and for a considerable time thereafter, as quenched core debris re-heats. Core melting would not begin until about 135 minutes after accident initiation in a loss of the moderator cooling system and until about 30 minutes in a loss of the moderator heat sink. Eventually, a pool of molten material would form in the bottom of the calandria, which may or may not boil, depending on property values. In all cases, the molten core would be contained within the calandria, as long as the shield tank water cooling system remains operational. Finally, in the period from 8 to 50 hours after the initiation of the accident, the molten core would re-solidify within the calandria. There would be no consequent damage to containment resulting from these accident sequences, nor would there be a significant increase in fission product releases from containment above those that would otherwise occur in a dual failure LOCA plus LOECI

  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. Analysis of progression of severe accident in Indian PHWRs

    International Nuclear Information System (INIS)

    In India a wide variety of nuclear reactors are in operation and in different stages of construction. The main stay of Indian nuclear power programme today is 'Pressurised Heavy Water Reactors (PHWRs)'. There are 13 operating PHWRs and several others in different stages of construction. These reactors are either of 220 MWe or 540 MWe capacity. Atomic Energy Regulatory Board for authorization needs safety analysis reports, which consists of a detailed analyses of all design basis accidents. However, there is a more to carry out severe accident analysis for accident management programme. This paper describes an analysis of a severe accident caused by Loss of Coolant Accident (LOCA), co-incident with loss of emergency core cooling system and loss of moderator heat sink in 220 MWe Indian PHWR. Initially in a matter of about 60 seconds most of the coolant from primary heat transport system blows out, the reactor gets tripped, but in the absence of emergency core cooling system, the heat removal from the fuel bundles is very poor. Consequently, the fuel bundle starts getting heated up. The only mode of heat transfer is radiative heat transfer from fuel bundle to pressure tube, from pressure tube to calandria tube and them convective heat transfer from calandria tube to the relatively cold moderator in which the reactor channels are immersed. In the absence of availability of moderator heat sink the moderator gets heated up and eventually boils. And slowly the moderator level in the calandria starts falling. Soon the channel gets uncovered and the temperatures of the channel components shoot up as the temperature of the pressure tube and calandria tube rise. Mechanical properties deteriorate rapidly with temperature as structural elements of reactor channel are made of zircaloy. Under the weight of the fuel, the reactor channel gives in and falls into the remaining moderator. This process continues till all the moderator is evaporated, leading to damage to the entire

  14. ACR-1000: Enhanced response to severe accidents

    International Nuclear Information System (INIS)

    Full text: Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-TM700 (ACR-700TM) as an evolutionary advancement of the current CANDU 6R reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000TM for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life. and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The moderator heavy water in the ACR-1000 calandria vessel, as in any other CANDU-type reactor, provides ample heat removal capacity in severe accidents. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel will be designed for debris retention. Core damage termination is achieved by flooding of the core components with water and keeping them flooded thereafter. Successful termination can be achieved in the fuel channels, calandria vessel or calandria vault by water supply by the Long Term Cooling (LTC) pumps and by gravity feed from the Reserve Water System. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes. Containment

  15. 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

  16. Design and fabrication of the retube transfer cask for Bruce N.G.S

    International Nuclear Information System (INIS)

    The retubing of CANDU reactors is a complex process which involves the removal and disposal of highly activated and contaminated calandria tubes and pressure tubes. For Bruce 'A' N.G.S., old pressure tubes will be removed from the reactor by cutting them into three segments; two end fitting assemblies which measure up to 3.2 m in length, and one pressure tube segment which measures up to 6.3 m. in length. Calandria tubes are 6.2 m in length. The function of the retube transfer cask is to provide for shielded transfer of these components between the reactor face and the in-ground disposal facility. This paper describes the design and fabrication of this cask. (author) 1 tab., 7 figs

  17. Analytical model for performance verification of liquid poison injection system of a nuclear reactor

    International Nuclear Information System (INIS)

    Highlights: • One-dimensional modelling of shut down system-2. • Semi-empirical correlation poison jet progression. • Validation of code. - Abstract: 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 (1D) hydraulic code, COPJET is developed, to predict the performance of system by predicting progression of poison jet with time. Validation of the COPJET is done with the data available in literature. Thereafter, it is applied for advanced vertical pressure type reactor

  18. 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)

  19. 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

  20. Uncertainty Analysis of the Potential Hazard of MCCI during Severe Accidents for the CANDU6 Plant

    Directory of Open Access Journals (Sweden)

    Sooyong Park

    2015-01-01

    Full Text Available This paper illustrates the application of a severe accident analysis computer program to the uncertainty analysis of molten corium-concrete interaction (MCCI phenomena in cases of severe accidents in CANDU6 type plant. The potential hazard of MCCI is a failure of the reactor building owing to the possibility of a calandria vault floor melt-through even though the containment filtered vent system is operated. Meanwhile, the MCCI still has large uncertainties in several phenomena such as a melt spreading area and the extent of water ingression into a continuous debris layer. The purpose of this study is to evaluate the MCCI in the calandria vault floor via an uncertainty analysis using the ISAAC program for the CANDU6.

  1. 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

  2. Ageing of zirconium alloy components

    International Nuclear Information System (INIS)

    India has two types (pressurized heavy water reactors (PHWRs) and boiling water reactors (BWRs)) of commercial nuclear reactors in operation, in addition to research reactors. Many of the life limiting critical components in these reactors are fabricated from zirconium alloys. The progressive degradation of these components caused by the cumulative exposure of high energy neutron irradiation with increasing period of reactor operation was monitored to assess the degree of ageing. The components/specimens examined included fuel element claddings removed from BWRs, pressure tubes and garter springs removed from PHWRs and calandria tube specimens used in PHWRs. The tests included tension test (for cladding, garter spring), fracture toughness test (for pressure tube), crush test (for garter spring), and measurement of irradiation induced growth (for calandria tube). Results of various tests conducted are presented and applications of the test results are elaborated for residual life estimation/life extension of the components

  3. 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

  4. Experience of detecting blisters in irradiated coolant channels of Indian Pressurised Heavy Water Reactors

    International Nuclear Information System (INIS)

    A number of irradiated pressure tubes which were in contact with calandria tube during reactor operation have been subjected to detailed examination. In case of contact, calandria tube/ pressure tube (CT/PT) contact hydrogen absorbed in the pressure tube migrates and keeps accumulating in the contact region cold spot under thermal gradient. Over a length of time, accumulated hydrogen at the contact zone forms localized massive concentration of δ-phase zirconium hydride, which is termed as Blister. Blister grows in size with time in the reactor and reaches a critical size when it can crack. Presence of a cracked blister is a matter of concern for the safety of pressure tubes. Ultrasonic velocity ratio measurement technique has been developed and applied to evaluate formation of hydride blisters in irradiated pressure tube during the course of post irradiation examination. (author)

  5. CIGAR

    International Nuclear Information System (INIS)

    CIGAR (channel inspection and gauging apparatus for reactors) is an automated system that has been developed for the inspection of fuel channels in CANDU PHWR reactors. A reactor contains several hundred horizontally mounted channels each consisting of a Zirconium alloy pressure tube, which holds the natural uranium fuel bundles, surrounded by a calandria tube. Garter spring spacers separate the two tubes. The inspection system performs ultrasonic flaw detection, diameter and wall thickness measurement, and sag profile measurements of the pressure tube. It also determines the location of the inter-tube spacers, and measures the gap between the pressure tube and the calandria tube. The CIGAR system is fully operational and has successfully completed inspections of a number of in-service fuel channels. This paper describes the equipment and it's operation, and gives examples of typical inspection results

  6. A review of computer codes MODTURC-CLAS and PHOENICS

    International Nuclear Information System (INIS)

    This report provides a review of computer codes MODTURC-CLAS and PHOENICS, as applied to simulating the moderator flow inside the calandria of a CANDU nuclear reactor. It is concluded that the mathematical formulations of the codes account for the dominant physics of the moderator flows. However, weaknesses in these formulations include the pressure loss model for the calandria tube effects; the turbulence models which currently do not account for buoyancy effects, streamline curvature effects and low Reynolds number effects; and the resolution of the computational grids used: Two-dimensional simulations are in relatively good qualitative agreement with experimental data, although some quantitative differences warrant further investigation. It is recommended that additional verification of both two-and three-dimensional simulations be carried out with both codes. The problems identified with PHOENICS should, however, be corrected prior to testing it on other moderator flow situations. (author) 26 refs., 3 tabs., 16 figs

  7. 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

  8. 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

  9. 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)

  10. Salida de campo a Villarmentero de Esgueva (Valladolid) el 18 de junio de 1956

    OpenAIRE

    Valverde Gómez, José Antonio, 1926-2003

    2008-01-01

    Salida de campo a Villarmentero de Esgueva (Valladolid) el 18 de junio de 1956, de la que se anotaron observaciones sobre el anfibio Epidalea calamita (Sapo corredor, llamado Bufo calamita por el autor), y las siguientes aves: Calandrella sp. (Terrera), Carduelis cannabina (Pardillo común, llamada Colorín y Acanthis cannabina por el autor), Coturnix coturnix (Codorniz común), Falco tinnunculus (Cernícalo vulgar), Galerida malabarica (Cogujada malabar), Melanocorypha calandra (Calandria), Mili...

  11. Salida de campo a las cuestas de Villanubla, Fuensaldaña y el Cabildo, en Valladolid, el 15 de agosto de 1953

    OpenAIRE

    Valverde Gómez, José Antonio, 1926-2003

    2008-01-01

    Salida de campo a las Cuestas de Villanubla, Fuensaldaña y el Cabildo, en Valladolid, el 15 de agosto de 1953, de la que se anotaron observaciones sobre las siguientes aves: Calandrella brachydactyla (Terrera común), Circus aeruginosus (Aguilucho lagunero occidental), Coturnix coturnix (Codorniz común), "Culialbo", Melanocorypha calandra (Calandria), Merops apiaster (Abejaruco europeo), Perdiz (Alectoris sp. o Perdix sp.), Petronia petronia (Gorrión chillón, también llamada Jiria), Pterocles ...

  12. Humid scraping method to obtain samples for the analysis of D2 incorporated in the pressure tubes of Embalse Nuclear Power Plant

    International Nuclear Information System (INIS)

    From ten fuel channels of the CNE reactor four samples of each channel were taken by means of the Humid Scraping method in order to evaluate the equivalent hydrogen content by incorporating deuterium in the pressure tubes. With these data, it is possible to make a list of priorities of channels for future replacement of spacer rings between pressure and calandria tubes, using Slarette equipment. (author)

  13. 1978 annual report

    International Nuclear Information System (INIS)

    In fiscal 1978 efforts continued to be made to increase New Brunswick's energy independence through research into and development of indigenous power sources including coal, peat, tidal, and hydroelectric resources. At Point Lepreau nuclear generating station the reactor building was completed, the concrete vault was prepared, and the calandria was installed. Work continued within the reactor building. Excavation of the cooling water tunnels and riser shafts was completed. (LL)

  14. Current safety issues of CANDU licensing

    International Nuclear Information System (INIS)

    As requested by Korea Institute of Nuclear Safety(KINS), the status of five generic licensing issues has been examined and their potential impact on a new plant that would be constructed in Canada has been evaluated. The results and conclusions of this evaluation are summarized as follows: steam explosion in calandria, hydrogen explosion in containment, use of PSA in reactor licensing, human factors, safety critical software

  15. Study of sources, dose contribution and control measures of Argon-41 at Kaiga Generating Station

    International Nuclear Information System (INIS)

    Air is used as a medium for cooling calandria vault and thermal shield systems in the earlier Pressurised Heavy Water Reactors (Rajasthan Atomic Power Station and Madras Atomic Power Station) in India. This leads to production of significant quantity of 41Ar in calandria vault and thermal shield cooling systems due to neutron activation of 40Ar present in air (∼1% v/v). The presence of 41Ar in reactor building contributes significant external doses to plant personnel during reactor operation and the release of this radionuclide to the environment result in dose to the public in the vicinity of the plants. An attempt is made to eliminate Argon-41 production in Indian standard Pressurised Heavy Water Reactors (Narora Atomic Power Station, Kakrapar Atomic Power Station, Kaiga Generating Station -1 and 2 and Rajasthan Atomic Power Station-3 and 4), by filling the calandria vault with demineralized water and providing a separate Annulus Gas Monitoring System (AGMS) for detecting leaks from calandria tube or pressure tube using Carbon dioxide as a medium. However, 41Ar is produced in the Annulus Gas Monitoring System, Primary Heat Transport cover gas system and moderator cover gas system due to ingress of air into the systems during operational transients or due to trace quantity of air present as an impurity in the gases used for the above systems. A study was conducted to identify and quantify the sources of 41Ar in the work areas. This report brings out the sources of 41Ar, reasons for 41Ar production and the results of the measures incorporated to reduce the presence of 41Ar in the above systems. (author)

  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. Safety benefits from CANDU reactor replacement. A case study

    International Nuclear Information System (INIS)

    Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

  18. Safety benefits from CANDU reactor replacement - a case study

    International Nuclear Information System (INIS)

    Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

  19. Fuel and fuel channel behaviour in loss of coolant accident without the availability of the emergency coolant injection system

    International Nuclear Information System (INIS)

    Safety Analysis of CANDU reactors assesses fuel and fuel channel behaviour under high temperature transient accident conditions. The basic purpose of the analysis is to establish the channel integrity (a sufficient, but not necessary condition) even when the Emergency Cooling Injection System is presumed to be unavailable. For such severe accident conditions, the channel is heated to temperatures where it deforms and creates a heat removal path from the fuel through the pressure tube and calandria tube to the moderator. The moderator in CANDU reactor is a separate system and can provide heat removal for the heat produced within the channel. This occurs through pressure tube deformation either by circumferential strain or sag whereby the pressure tube contacts the calandria tube and allowing heat to be conducted directly to the moderator. It is found that the heat generated within the channel is transported to the moderator, and that the implied modes of channel (pressure tube) deformation are physically possible, and do not lead to failure of the pressure tube (i.e. of the pressure boundary). This paper considers the fuel and channel thermal and mechanical behaviour at very high temperatures. It discusses modelling of fission product release from fuel, deformation of the pressure tube and calandria tube, and hydrogen production insofar as it affects the fuel analysis and the containment analysis. (author)

  20. Validation of severe accident management guidance for the wolsong plants

    International Nuclear Information System (INIS)

    Full text: Full text: The severe accident management(SAM) guidance has been developed for the Wolsong nuclear power plants in Korea. The Wolsong plants are 700MWe CANDU-type reactors with heavy water as the primary coolant, natural uranium-fueled pressurized, horizontal tubes, surrounded by heavy water moderator inside a horizontal calandria vessel. The guidance includes six individual accident management strategies: (1) injection into primary heat transport system (2) injection into calandria vessel (3) injection into calandria vault (4) reduction of fission product release (5) control of reactor building condition (6) reduction of reactor building hydrogen. The paper provides the approaches to validate the SAM guidance. The validation includes the evaluation of:(l) effectiveness of accident management strategies, (2) performance of mitigation systems or components, (3) calculation aids, (4) strategy control diagram, and (5) interface with emergency operation procedure and with radiation emergency plan. Several severe accident sequences with high probability is selected from the plant specific level 2 probabilistic safety analysis results for the validation of SAM guidance. Afterward, thermal hydraulic and severe accident phenomenological analyses is performed using ISAAC(Integrated Severe Accident Analysis Code for CANDU Plant) computer program. Furthermore, the experiences obtained from a table-top-drill is also discussed

  1. Moderator analysis of the Wolsong Units 2/3/4 for the case of 35% RIH breaks with a loss of ECC injection using CFX-4.4

    International Nuclear Information System (INIS)

    A 3-D CFD model was developed for the transient simulation of the moderator circulation inside the Calandria vessel, using a commercial code CFX-4.4. To reduce the discretization errors in the reflector region, a set of butterfly-shaped grid structures were generated and used. The porous media approach was applied for the core region containing 380 Calandria tubes. A 3-dimensional hydraulic model for the porous core region has been developed based on the empirical pressure drop correlations. The transient moderator analysis was performed for the 35% RIH(Reactor Inlet Header) break with a loss of ECC(Emergency Core Cooling) injection, which gives the largest heat load to the moderator out of all the DBA's. During the initial 1200 seconds in the LOCA transient, the local subcooling near any PT/CT contact location does not go down below the minimum subcooling margin of 28oC, the experimentally derived threshold. It is concluded that film boiling on the Calandria tube surfaces would not occur during the selected LOCA scenario and thus the fuel channel integrity is maintained in the CANDU-6 reactor. (author)

  2. Correlation between the critical heat flux and the fractal surface roughness of zirconium alloy tubes

    Energy Technology Data Exchange (ETDEWEB)

    Fong, R.W.L.; McRae, G.A.; Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Nitheanandan, T.; Sanderson, D.B. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

    1999-10-01

    In CANDU fuel channels, Zircaloy calandria tubes isolate the hot pressure tubes from the cool heavy water moderator. The heavy-water moderator provides a backup heat sink during some postulated loss-of-coolant accidents. The decay heat from the fuel is transferred to the moderator to ensure fuel channel integrity during emergencies. Moderator temperature requirements are specified to ensure that the transfer of decay heat does not exceed the critical heat flux (CHF) on the outside surface of the calandria tube. An enhanced CHF provides increases in safety margin. Pool boiling experiments indicate the CHF is enhanced with glass-peening of the outside surface of the calandria tubes. The objective of this study was to evaluate the surface characteristics of glass-peened tubes and relate these characteristics to CHF. The micro-topologies of the tube surfaces were analysed using stereo-pair micrographs obtained from scanning electron microscopy (SEM) and photogrammetry techniques. A linear relationship correlated the CHF as a function of the 'fractal' surface roughness of the tubes. (author)

  3. Experimental investigation of transient behaviour of IPHWR under heat up condition

    Energy Technology Data Exchange (ETDEWEB)

    Dutt, Nitesh, E-mail: mech.nitesh@gmail.com [Mechanical & Industrial Engineering Department, IIT Roorkee, Roorkee (India); Sahoo, P.K., E-mail: sahoofme@iitr.ac.in [Mechanical & Industrial Engineering Department, IIT Roorkee, Roorkee (India); Mukhopadhyay, Deb, E-mail: dmukho@barc.gov.in [Mechanical & Industrial Engineering Department, IIT Roorkee, Roorkee (India); Bhabha Atomic Research Centre, Anushakti Nagar, Mumbai (India)

    2015-08-15

    Loss of coolant accident (LOCA) in nuclear reactor has a low probability to occur, but may have significant consequences due to decaying heat of the nuclear fuel rods. During LOCA with un-mitigated station blackout (SBO) in Indian Pressurized Heavy Water Reactor (IPHWR), there is a mismatch between heat generation and heat removed from the fuel channels, that results in structural failure of the fuel channels. These channels settle down at the bottom of the Calandria vessel, forming debris bed. Decay heat is absorbed by the moderator present in the Calandria vessel. The moderator level decreases due to continuous evaporation, which exposes fuel channels gradually. An experimental setup has been designed and fabricated to simulate the debris bed scenario. Experiments are conducted at different moderator levels, in the Calandria vessel and at different rated power. The temperature of the exposed rods gets stabilized with time due to steam cooling. It is observed that, the temperature is well below the melting point of the cladding, PT and CT material.

  4. Coolability of severely degraded CANDU cores. Revised

    Energy Technology Data Exchange (ETDEWEB)

    Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Nijhawan, S

    1996-01-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually resolidify. Thus, the calandria vessel would act inherently as a 'core-catcher' as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author)

  5. Some studies on fluid structure interaction problems

    International Nuclear Information System (INIS)

    In this report a review of fluid structure interaction problems has been made. The areas of development work on liquid container design, seismic response of pressure supression pool in a nuclear power plant, analysis of submerged structures and pulse induced vibration in coupled fluid shell systems are discussed. The areas of application for analysis of moderator filled calandria of a pressurised heavy water reactor (PHWR) and analysis of calandria tube/pressure tube/guide tube in case of overpressurisation of calandria are emphasised. A brief review of various methods for solving fluid structure interaction problem is then undertaken. A formulation for two dimensional and axi-symmetric problems based on finite element method using pressure field in fluid and displacement field in structure is presented along with a number of problems solved using this method. Implementation of implicit explicit mixed time integration scheme for solution of equations of the coupled field is discussed. A comparison has been made between present approach and structure mechanics priority approach. (author). 8 figs., 36 refs

  6. A flashing driven moderator cooling system for CANDU reactors: Experimental and computational results

    International Nuclear Information System (INIS)

    A flashing-driven passive moderator cooling system is being developed at AECL for CANDU reactors. Preliminary simulations and experiments showed that the concept was feasible at normal operating power. However, flow instabilities were observed at low powers under conditions of variable and constant calandria inlet temperatures. This finding contradicted code predictions that suggested the loop should be stable at all powers if the calandria inlet temperature was constant. This paper discusses a series of separate-effects tests that were used to identify the sources of low-power instabilities in the experiments, and it explores methods to avoid them. It concludes that low-power instabilities can be avoided, thereby eliminating the discrepancy between the experimental and code results. Two factors were found to be important for loop stability: (1) oscillations in the calandria outlet temperature, and (2) flashing superheat requirements, and the presence of nucleation sites. By addressing these factors, we could make the loop operate in a stable manner over the whole power range and we could obtain good agreement between the experimental and code results. (author)

  7. Development of channel inspection and gauging apparatus for 235 MWe PHWRs

    International Nuclear Information System (INIS)

    Channel inspection and gauging apparatus is being developed to enable in-service channel inspection and gauging. Phase I apparatus to measure annular gap between pressure tube and calandria tube in a dry channel has been developed. The apparatus consists of a gauging head and a drive mechanism. The gauging head utilities an eddy current probe to measure the annular gap between pressure tube and calandria tube and an ultrasonic sensor to measure the wall thickness of the pressure tube. The output signal of the eddy current probe needs to be corrected for the effect of pressure tube wall thickness variation. This paper gives the details of the above apparatus. The results of calibration tests at mock-up station are presented. The paper outlines the program for the phase-wise development of Channel Inspection and Gauging Apparatus for use in heavy water filled channels without their isolation from PHT and draining. The final apparatus will have the facilities for ultrasonic flaw detection, ultrasonic gauging to measure pressure tube diameter and wall thickness, an inclinometer to measure slope and sag of pressure tube and eddy current probe for the measurement of annular gap between pressure tube and calandria tube. (author). 6 figs

  8. 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

  9. Correlation between the critical heat flux and the fractal surface roughness of zirconium alloy tubes

    International Nuclear Information System (INIS)

    In CANDU fuel channels, Zircaloy calandria tubes isolate the hot pressure tubes from the cool heavy water moderator. The heavy-water moderator provides a backup heat sink during some postulated loss-of-coolant accidents. The decay heat from the fuel is transferred to the moderator to ensure fuel channel integrity during emergencies. Moderator temperature requirements are specified to ensure that the transfer of decay heat does not exceed the critical heat flux (CHF) on the outside surface of the calandria tube. An enhanced CHF provides increases in safety margin. Pool boiling experiments indicate the CHF is enhanced with glass-peening of the outside surface of the calandria tubes. The objective of this study was to evaluate the surface characteristics of glass-peened tubes and relate these characteristics to CHF. The micro-topologies of the tube surfaces were analysed using stereo-pair micrographs obtained from scanning electron microscopy (SEM) and photogrammetry techniques. A linear relationship correlated the CHF as a function of the 'fractal' surface roughness of the tubes. (author)

  10. Video inspection of lattice tube rolled joint area with structured lighting for enhanced surface profiling

    International Nuclear Information System (INIS)

    Retube Tubesheet Bore Dressing is a series of operations being performed by the Bruce Retube Project between calandria tube removal and installation of the new calandria tubes. Specialized remotely-operated tooling has been designed and developed to perform the task to ensure proper tubesheet bore conditions prior to making the new calandria tube rolled joint. A key part of the tool is the video inspection module. During operation, the module scans the area to ensure proper tool location and to assess surface condition. The tool may then be used to perform either rotary brushing or milling to improve the groove profile. Following dressing, the system re-inspects the bore to verify the operation. The tool may also perform a video inspection of the inboard section of the lattice tube and verify the condition of the inboard bearing. In order to enhance the operator's ability to evaluate the machined surface, structured lighting has been used, consisting of an axially projected laser light stripe. This was included in the design to better define the surface profile during scanning. Surface inspection images are shown to demonstrate the performance of the unit. By analysing a series of captured images from the surface scan, both a surface profile and diametral measurements of the tubesheet bore can be calculated, extending the technique from a qualitative to a quantitative method. (author)

  11. 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

  12. Seismic re-evaluation of Madras Atomic Power Stations 1 and 2

    International Nuclear Information System (INIS)

    a combination of analysis, testing and by walkdown of the plant. The calandria, end-shields and the steam generators in these units are hanging from the ceiling by way of hanger rods. The end-shields are bolted to the calandria vault by 164 numbers of spring loaded bolts. The calandria receives support from the bolted end-shields in E-W and Vertical direction through the 306 numbers of coolant tubes, which pass through the calandria tubes of the calandria and the lattice tubes of the end-shield. In the N-S direction, there is a gap of 38 mm between the calandria and the end-shield in the N-S direction. In the seismic re-evaluation of the MAPS- l and 2, all the major equipment, viz., calandria, end-shields, steam generators and the PHT piping have been coupled with the primary civil structure and analysed by response spectrum method using free-field ground response spectra instead of qualifying them as decoupled equipment and using floor response spectra. Other pipings have been qualified by floor response spectra. The paper brings out the details of the work accomplished during seismic re-evaluation of the MAPS- l and 2. (author)

  13. PIV Measurement of Isothermal Flow in the Moderator Circulation Test (MCT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Im, Sunghyuk; Sung, Hyung Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Seo, Han; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, Hyoung Tae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    One of the important design features of a CANDU reactor (a pressurize heavy water reactor) is the use of moderator as a heat sink during some postulated accidents such as a large break Loss Of Coolant Accident (LOCA). If the moderator available subcooling at the onset of a large LOCA is greater than the subcooling requirements, a sustained calandria tube dryout is avoided. The subcooling requirements are determined from a set of experiments known as the fuel channel contact boiling experiments. The difference between available subcooling and required subcooling is called subcooling margins. The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the local temperature in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. In the present work the test vessel is equipment with 380 acrylic pipes instead of the heater rods and a preliminary measurement of velocity field using PIV is performed under the iso-thermal test conditions. The 2D velocity is measured on the cross-sectional plane normal to the axial direction of the tank. The PIV measurement results could capture the same flow pattern as that expected in the CANDU6 calandria tank under momentum dominant flow condition, where the inlet jets penetrate to the top of the tank and produce a downward flow through the center of the tube columns towards the outlet nozzle and the flow fields are in symmetric distributions. The measurements of downward velocities are performed at different locations. The velocity is shown to be axially uniform. The velocity is rapidly decreased as the measurement location is far from the center of tank, since the downward flow is dominant along the center of the tube columns. More experimental works for the iso-thermal conditions as well as the heating conditions will be performed using PIV measurement in the

  14. 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)

  15. A Component-Scale Thermal-Hydraulic Analysis of the Flow in a CANDU Moderator Tank

    International Nuclear Information System (INIS)

    In this regard, to estimate the local subcooling of the moderator inside the Calandria vessel is very important. However, the local temperature is measured at the inlet and the outlet of the vessel only. Therefore, we need to accurately predict the local temperature inside the Calandria vessel. In this study the component-scale thermal-hydraulic analysis of a real CANDU moderator tank is performed by using the CUPID code, which is based on the lessons from the previous results. The porous media approach is used for an efficient simulation at a component scale. The inlet nozzles are also modeled using a simplified mesh. The validity of the results is checked and effects of the inlet nozzle modeling on the results are discussed. To examine the effects of the inlet nozzle modeling, three inlet nozzle models are proposed. The Case 1 using a uniform velocity at the nozzles presented a thermal stratification in the upper region of the Calandria tank, which is caused by the lack of the flow momentum injected into the inlet nozzles. The results are not realistic. The Case 2 using a more realistic velocity distribution instead of the uniform velocity, predicted an asymmetric mixed flow regime, maintaining with a balance between the momentum and buoyancy force, which was experimentally shown the STERN test facility in the nominal operating conditions. For a further improvement, the Case 3 was attempted, which aims at the preservation of both the mass and momentum flow at the inlet nozzles. The results of the Case 3 seem the best among the three cases. The local maximum temperature was very close to that of other calculations and the flow pattern was also very similar to the experimental and computational results

  16. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  17. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    International Nuclear Information System (INIS)

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days

  18. Remotely controlled repairs at Douglas Point NGS

    International Nuclear Information System (INIS)

    In September, 1977, leakage of heavy water at a rate of 125 kg/hr was detected in an area of the Douglas Point NGS reactor vault below the calandria known as the lower labyrinth. Radiation in the area ranges up to 5000 R/hr and the only ready access was through four 75 mm inspection ports that open into the moderator room. Remote-controlled equipment was designed and built to diagnose the problems and carry out repairs. All damaged piping was fixed, supports were replaced as needed, and system vibration was reduced. The work was done with no injuries and little radiation dose

  19. 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.

  20. 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

  1. Post irradiation examination of tight fit garter springs from Indian PHWR

    International Nuclear Information System (INIS)

    Garter springs play an important role in maintaining the annulus gap between hot pressure tubes and cold calandria tubes of PHWRs. Post irradiation examination (PIE) was carried out on the garter springs removed from Indian PHWR after around 8 and 15 Hot Operating Years (HOY). PIE studies included visual examination, dimensional measurements, metallographic examination and relevant mechanical tests. The girdle wires of these garter springs were also examined and subjected to the tension and bend tests. This paper gives the results of the PIE investigations and discusses its relevance for continued performance of garter springs in PHWRs

  2. Eddy current and ultrasonic fuel channel inspection at Karachi Nuclear Power Plant

    International Nuclear Information System (INIS)

    In November of 1993 and in-service inspection was performed on eight fuel channels in the Karachi Nuclear Power Plant (KANUPP) reactor. The workscope included ultrasonic and eddy current volumetric examinations, and eddy current measurement of pressure-to calandria tube gap. This paper briefly discusses the planning strategy of the ultrasonic and eddy current examinations, and describes the equipment developed to meet the requirements, followed by details of the actual channel inspection campaign. The presented nondestructive examinations assisted in determining fitness for service of KANUPP reactor channels in general, and confirmed that the problems associated with channel G12 were not generic in nature. (author)

  3. 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)

  4. An experimental study of a flashing-driven CANDU moderator cooling system

    International Nuclear Information System (INIS)

    The results of an experimental study to investigate the feasibility of using a passive flashing-driven natural circulation loop for CANDU-reactor moderator heat rejection are presented. A scaled loop was constructed and tested at conditions approximating those of a CANDU calandria cooling system. The results showed that stable loop operation was possible at simulated powers approaching normal full power. At lower powers, flow oscillations occurred as the flow in the hot-leg periodically changed from two-phase to single-phase. The results from earlier numerical predictions using the CATHENA thermalhydraulics code showed good qualitative agreement with the experimental results. (author). 6 refs., 11 figs

  5. 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)

  6. Enhancing the seismic capability of the on-power refueling system of the CANDU reactor

    International Nuclear Information System (INIS)

    The CANDU reactor assembly includes several hundred horizontal fuel channels, each containing twelve fuel bundles, arranged in a square lattice, and supported by the reactor structures. CANDU operates on natural uranium or other low fissile content fuel, and is refueled on-power, with either four or eight fuel bundles in a channel being replaced during each refueling operation. The fueling machines clamp onto the opposite ends of the fuel channel to be refueled. The seismic capacity of this refueling system is evaluated in terms of its dynamic response during an earthquake. This paper describes the approach adopted to enhance the seismic capability of the fueling machine and calandria assembly for earthquakes of O.3g ground acceleration covering a broad range of soil conditions ranging from soft to hard. A detailed, 3-D finite element seismic model of the fueling machine and calandria assembly system is developed to calculate the seismic responses of the structure. Some relatively simple hardware design changes have been considered to increase the seismic capacity of the CANDU 6 reactor. These changes in the fueling machine and calandria assembly of the CANDU 6 reactor are briefly described. They have been incorporated into the finite element seismic model of the system. Most of these design changes have already been considered and implemented in other CANDU reactor projects. The current CANDU 6 reactor design fully meets the requirements of seismic qualification for sites with potential for O.2g ground acceleration where the seismic loads need to be combined with the other design loads for the support and pressure boundary components to demonstrate compliance with the applicable Code requirements. In the present study it is demonstrated that, with relatively simple hardware changes, the fueling machine and calandria assembly of the CANDU 6 reactor can withstand earthquakes of O.3g ground acceleration. Based on the current study and some preliminary analysis of the

  7. Surveillance and maintenance tasks with tele-operated mobile robot in nuclear power plants

    International Nuclear Information System (INIS)

    Tele-operated mobile robot has been developed in KAERI to develop robotic technologies for the preventive maintenance and remote inspection of nuclear power plants such as remote maintenance work in reactor coolant system at PWR and inspection of pressure tubes in calandria at PHWR. The extendable inspection mast attached on the mobile platform enables human eyes to look into reactor face during full power plant operation. These robotic systems have been developed for keeping human workers from high radiation exposure in hostile nuclear environments. (author)

  8. Radionuclide Release after Channel Flow Blockage Accident in CANDU-6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hoon; Jun, Hwang Yong [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The channel flow blockage accident is one of the in core loss of coolant accidents, the release path of radionuclide is very different from conventional loss of coolant accidents. The large amount of radionuclide released from broken channel is being washed during it passes through the moderator in Calandria. The objective of containment behavior analysis for channel flow blockage event is to assess the amount of radionuclide release to the ambient atmosphere. Radionuclide release rates in case of channel flow blockage with all safety system available, that is containment building is intact, as well as with containment system impairment are analyzed with GOTHIC and SMART code

  9. Advantages of butterfly valves for power plants

    International Nuclear Information System (INIS)

    Butterfly valves are increasingly used in nuclear power plants. They are used in CANDU reactors for class 2 and 3 service, to provide emergency and tight shutoff valves for all inlets and outlets of heat exchangers and all calandria penetrations. Guidelines for meeting nuclear power plant valve specifications are set out in ASME Section 3, Nuclear Power Plant Components. Some details of materials of construction, type of actuator, etc., for various classes of nuclear service are tabulated in the present article. The 'fishtail' butterfly valve is an improved design with reduced drag, as is illustrated and explained. (N.D.H.)

  10. Fuel elements assembling for the DON project exponential experience

    International Nuclear Information System (INIS)

    It is described the fuel unit used in the DON exponential experience, the manufacturing installments and tools as well as the stages in the fabrication.These 74 elements contain each 19 cartridges loaded with synterized urania, uranium carbide and indium, gold, and manganese probes. They were arranged in calandria-like tubes and the process-tube. This last one containing a cooling liquid simulating the reactor organic. Besides being used in the DON reactor exponential experience they were used in critic essays by the substitution method in the French reactor AQUILON II. (Author) 6 refs

  11. Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

    International Nuclear Information System (INIS)

    As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed

  12. 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)

  13. 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)

  14. Qualification of control rod guide tube assembly for flow induced vibration of 700 MWe Indian PHWR-KAPP 3 and 4

    International Nuclear Information System (INIS)

    The vibrations induced in elastic structures by vortex shedding are of practical importance because of its potentially destructive effect on various power plant components in terms of displacement and stress. In 700 MWe PHWR higher thermal power is produced by allowing partial boiling of the coolant at the end of the channels compared to 540 MWe reactors of same core design. It leads to increase in heat content of the moderator due to neutron moderation and capture, attenuation of gamma radiation. To maintain equilibrium moderator temperature, flow of the moderator in the calandria is increased through the diffusers. Qualification of various internal components in the calandria is essential for increased loads generated due to fluid induced vibration and other loads. In-core component of Control Rod Mechanism called Guide Tube Assembly is one important safety components which will be subjected to high flow of moderator. Finite element model of the assembly has been developed to evaluate the natural frequency. Vortex shedding frequency is evaluated based on the moderator velocity profile in the vicinity of guide tube assembly at different elevations. Guide tube assembly is checked for its functionality to ensure free movement of control rod. Stress analysis is carried out for the additional forces generated due to flow induced vibration. (author)

  15. 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.

  16. 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)

  17. 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)

  18. Biomass and Soil Carbon Stocks in Wet Montane Forest, Monteverde Region, Costa Rica: Assessments and Challenges for Quantifying Accumulation Rates

    Directory of Open Access Journals (Sweden)

    Lawrence H. Tanner

    2016-01-01

    Full Text Available We measured carbon stocks at two forest reserves in the cloud forest region of Monteverde, comparing cleared land, experimental secondary forest plots, and mature forest at each location to assess the effectiveness of reforestation in sequestering biomass and soil carbon. The biomass carbon stock measured in the mature forest at the Monteverde Institute is similar to other measurements of mature tropical montane forest biomass carbon in Costa Rica. Local historical records and the distribution of large trees suggest a mature forest age of greater than 80 years. The forest at La Calandria lacks historical documentation, and dendrochronological dating is not applicable. However, based on the differences in tree size, above-ground biomass carbon, and soil carbon between the Monteverde Institute and La Calandria sites, we estimate an age difference of at least 30 years of the mature forests. Experimental secondary forest plots at both sites have accumulated biomass at lower than expected rates, suggesting local limiting factors, such as nutrient limitation. We find that soil carbon content is primarily a function of time and that altitudinal differences between the study sites do not play a role.

  19. Zirconium hydride blister modelling and the application to the P2-G16 failure

    International Nuclear Information System (INIS)

    A CANDU pressure tube failure in Pickering GS 'A', Unit 2, occurred in August, 1983, during reactor operation. It resulted from the propagation of a crack which initiated at solid zirconium hydride concentrations (or blisters) on the outer surface of the pressure tube where there was contact with the much cooler surrounding calandria tube. Numerical modelling capabilities have been developed [1-5] to predict the history of the pressure tube/calandria tube contact, the redistribution of temperature due to contact, the hydrogen migration, precipitation and hydride blister formation and the changes in geometry, thermal boundary conditions and material properties during the growth of blisters. This paper shows the application of these capabilities to simulate the conditions that led to the P2-G16 pressure tube failure. Best estimates for boundary conditions, controlling parameters and deuterium ingress rates are applied and the results are compared with measurements from the failed pressure tube. The excellent correlation between predicted and observed blisters confirms the predictive capability of the model. The effect of variations in boundary conditions on blister growth rates is also shown. (author)

  20. Effect Of Forced Mixing and Condensation On Hydrogen Distribution During Postulated Accident In MAPS Containment

    International Nuclear Information System (INIS)

    Madras Atomic Power Station (MAPS) of 220 MWe is a pressure tube type reactor with heavy water as moderator and coolant and natural Uranium Dioxide as fuel. It consists of 306 horizontal fuel channel assemblies and surrounded by three separate water systems i.e. primary coolant, moderator and calandria vault water system. The whole reactor is enclosed in the Containment Building. In the present study, coincident failure of Loss of Coolant Accident (LOCA) and Emergency Core Cooling System (ECCS) is postulated as dual failures accident. Such accident scenario, results in production of steam, hydrogen, along with release of fission products to containment environment. In such accident scenario pressure tubes would deform and get in contact with calandria tube, eventually transferring decay heat to moderator system. To limit the local hydrogen concentration forced mixing of the Dome Area air with the FMVs environment is considered as a suitable local hydrogen management system for the present design. Flows of mixing were evaluated to limit down the hydrogen concentration within acceptable limits. The paper brings out these aspects for MAPS. (authors)

  1. Theory Manual for ISAAC Computer Code

    International Nuclear Information System (INIS)

    Major models adopted in ISAAC are introduced briefly. The primary heat transport system of two independent figure-of-eight loops are represented in ISAAC. All four steam generators and 4 pumps are also modeled individually. PHTS model tracks the masses and energy in one gas space and in multiple water pools. The pressurizer model is similar to the PWR model in MAAP4, except two independent surge lines which are connected to each PHTS loop. The two-region steam generator model was newly implemented into the ISAAC code. In each region, masses and energies of water, steam, and non-condensable gases are tracked. Then, the pressure, water temperature, and gas temperature in each region are calculated based on the masses and energies, using non-equilibrium thermodynamic model. The core heatup module calculates the thermal-hydraulic response and fission product transport, including the rates-of-change of dynamic variables, within the core region. The key quantities of calandria tank model include thermal-hydraulic variables, modeling of corium, water, and calandria tank wall heat transfer, corium debris bed in the bottom of the tank, failure mechanisms, and fission product transport. ISAAC also models main safety features in Wolsong plants as well as fission product behavior. As described, ISAAC has a fundamental models to capture the main phenomena during the severe accident in Wolsong plants

  2. Strategies for accelerating the SLARette process

    International Nuclear Information System (INIS)

    The SLARette (Spacer Location and Repositioning) process is continuing on several CANDU reactors, where loose fitting garter springs (spacers) were used, to prevent contact between the calandria tube and the pressure tube for the target life. With time, the sag in the fuel channel is increasing and consequently increasing the potential for contact between the pressure tube and the calandria tube. Also, due to increasing sag in the pressure tubes and increasing magnitude of the fuel channel constrictions on the eddy current detection system, the Spacer Location and Repositioning activities are becoming more time consuming and difficult. For CANDU owners, during the SLARette campaigns, station outage time is the most expensive item. Therefore, it is beneficial to complete the SLARette process as early as possible and as fast as possible. New SLARette strategies can substantially accelerate the overall SLARette process and thus minimize the outage time. There are several strategies to perform the SLARette process. These strategies include: using the SLARette Mark II Delivery System; using the SLARette Advanced Delivery System; implement creative fuel handling technique; operate from both sides of the reactor using Mark II Delivery Systems; operate both sides using Advanced Delivery Systems. Each strategy offers different benefits, rate of fuel channel processing (SLARette Activity), and schedule constraints. This paper provides the details of each strategy and compare them in terms of outage time, man-rem consumption, and constraints. (author)

  3. 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)

  4. Simulation model for jet flow in liquid injection system of CANDU-6 SDS 2

    International Nuclear Information System (INIS)

    For the performance analysis of the secondary shutdown system (SDS-2), a computational fluid dynamics (CFD) model for the poison jet flow is being developed to analyze the flow and poison concentration fields formed inside the moderator tank. As the ratio between Calandria shell and the nozzle hole diameter of the injection system is so big as 1055, it is impractical to develop a full size model encompassing the whole Calandria tank. To reduce the model to a manageable size, a quarter of the five-lattice-pitch length segment of the tank was modeled by using the symmetric nature of the jet and the injected jet was treated as source term to remove the limit caused by the small diameter of the injection nozzle hole, when the grid of the calculation domain was generated. A half model calculation was performed to show the symmetricity of the quarter model. For the validation of the source treatment of the inlet flow condition, the simulation result was compared with the experimental data of the gas jet. The symmetricity was confirmed by the results of simulation the half model calculation on the symmetric line and the result of simulation for the source treatment well agreed with the experiment when a fine mesh grid structure was used near the inlet

  5. Severe accident analysis to verify the effectiveness of severe accident management guidelines for large pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gokhale, O.S., E-mail: onkarsg@barc.gov.in; Mukhopadhyay, D., E-mail: dmukho@barc.gov.in; Lele, H.G., E-mail: hglele@barc.gov.in; Singh, R.K., E-mail: rksingh@barc.gov.in

    2014-10-15

    Highlights: • The progression of severe accident initiated from high pressure scenario of station black out has been analyzed using RELAP5/SCDAP. • The effectiveness of SAMG actions prescribed has been established through analysis. • The time margin available to invoke the SAMG action has been specified. - Abstract: The pressurized heavy water reactor (PHWR) contains both inherent and engineered safety features that help the reactor become resistant to severe accident and its consequences. However in case of a low frequency severe accident, despite the safety features, procedural action should be in place to mitigate the accident progression. Severe accident analysis of such low frequency event provides insight into the accident progression and basis to develop the severe accident management guidelines (SAMG). Since the order of uncertainty in the progression path of severe accident is very high, it is necessary to study the consequences of the SAMG actions prescribed. The paper discusses severe accident analysis for large PHWRs for multiple failure transients involving a high pressure scenario (initiation event like SBO with loss of emergency core cooling system and loss of moderator cooling). SAMG actions prescribed for such a scenario include water injection into steam generator, calandria vessel or calandria vault at different stages of accident. The effectiveness of SAMG actions prescribed has been investigated. It is found that there is sufficient time margin available to the operator to execute these SAMG actions and the progression of severe accident is arrested in all the three cases.

  6. Severe accident analysis to verify the effectiveness of severe accident management guidelines for large pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Highlights: • The progression of severe accident initiated from high pressure scenario of station black out has been analyzed using RELAP5/SCDAP. • The effectiveness of SAMG actions prescribed has been established through analysis. • The time margin available to invoke the SAMG action has been specified. - Abstract: The pressurized heavy water reactor (PHWR) contains both inherent and engineered safety features that help the reactor become resistant to severe accident and its consequences. However in case of a low frequency severe accident, despite the safety features, procedural action should be in place to mitigate the accident progression. Severe accident analysis of such low frequency event provides insight into the accident progression and basis to develop the severe accident management guidelines (SAMG). Since the order of uncertainty in the progression path of severe accident is very high, it is necessary to study the consequences of the SAMG actions prescribed. The paper discusses severe accident analysis for large PHWRs for multiple failure transients involving a high pressure scenario (initiation event like SBO with loss of emergency core cooling system and loss of moderator cooling). SAMG actions prescribed for such a scenario include water injection into steam generator, calandria vessel or calandria vault at different stages of accident. The effectiveness of SAMG actions prescribed has been investigated. It is found that there is sufficient time margin available to the operator to execute these SAMG actions and the progression of severe accident is arrested in all the three cases

  7. 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

  8. Thermo-mechanical behaviour of coolant channels for heavy water reactors under accident conditions

    International Nuclear Information System (INIS)

    The objective of nuclear safety research programme is to develop and verify computer models to accurately predict the behavior of reactor structural components under operating and off normal conditions. Indian Pressurised Heavy Water Reactors (PHWRs) are tube type of reactors. The coolant channel assemblies, being one of the most important components, need detailed analysis under all operating conditions as well as during postulated conditions of accidents for its thermo-mechanical behaviour. One of the postulated accident scenarios for heavy water moderated pressure tube type of reactors i.e. PHWRs is Loss Of Coolant Accident (LOCA) coincident with Loss Of Emergency Core Cooling System (LOECCS). In this case, even though the reactor is tripped, the decay heat may not be removed adequately due to low or no flow condition and inventory depletion of primary side. Since the emergency core cooling system is presumed to be not available, the cooling of the fuel pins and the coolant channel assembly depends on the moderator cooling system, which is assumed to be available. Moderator cooling system is a separate system in PHWRs. In PHWRs, the fuel assembly is surrounded by pressure tube, an annulus insulating environment and a concentric calandria tube. In this postulated accident scenario, a structural integrity evaluation has been carried out to assess the modes of deformation of pressure tube-calandria tube assembly in a tube type nuclear reactor. The loading of pressure and temperature causes the pressure tube to sag/balloon and come in contact with the outer cooler calandria tube. The resulting heat transfer could cool and thus control the deformation of the pressure tube thus introducing inter-dependency between thermal and mechanical contact behaviour. The amount of heat thus expelled significantly depends on the thermal contact conductance and the nature of contact between the two tubes. Deformation of pressure tube creates a heat removal path to the relatively

  9. 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

  10. Design of cobalt adjuster rods for new Pickering NGS A core

    International Nuclear Information System (INIS)

    The core of Pickering NGS A was redesigned to acheive a better shutdown system without increasing the number of calandria penetrations. The shutdown system has 73 percent more reactivity than before. The average discharge burnup has been increased by about 5 percent at the expense of some reduced ability to override xenon transients and operation without fuelling. Past experience at Pickering NGS A, has shown that there is less than one forced outage per year per unit where the xenon override time capability is needed. For such a lwo outage frequency, a large xenon override time at the expense of average discharge burnup is not economical. Also, a 36 day of operation without fuelling is quite satisfactory

  11. Radiolytic generation of gases in reactors

    International Nuclear Information System (INIS)

    Water or heavy water is used in different circuits in a reactor. Their most common use is as a moderator and/or as a coolant. Light water is used at other places such as in end shield, calandria vault etc., In the process they are exposed to intense ionizing radiation and undergo radiolytic degradation. The molecular produts of radiolysis are hydrogen, hydrogen peroxide and oxygen. As is commonly known if hydrogen is formed beyond a certain level, in the presence of oxygen it may lead to combustion or even explosion. Thus one should comprehend the basic principles of radiolysis and see whether the concentration of these gases under various conditions can be worked out. This report attempts to analyse in depth the radiolytic generation of gases in reactor systems. (author). 3 tabs

  12. 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)

  13. Development and field application of a leak sealant for the NRU water reflector

    International Nuclear Information System (INIS)

    The development and successful application of a unique leak sealant formulation comprised of a mixture of graded, hollow ceramic microspheres, surface oxidized aluminum powder and saturated gibbsite suspension is described. The project was undertaken to address the escalating leakage from up to 15 small weld defects in the water reflector vessel, an integral component of the NRU (National Research Universal) reactor calandria. The reflector surrounds the reactor core with a neutron reflecting blanket of light water. Injection of the sealant is typically done with the reactor shutdown and the water reflector system operating normally, but can also be performed with the reactor at full power. The procedure is simple and effective. Individual treatments of as little as 125 ml of sealant (10 ppm in the 12,500 L system) have yielded leak reductions exceeding 2000 L/day. The accumulated leak reductions of several treatments now total 90% of the historic high of 5,400 L/day. (author)

  14. The status of improved pressurized heavy water reactor development - A new option for PHWR -

    International Nuclear Information System (INIS)

    Currently, the 900 MWe class Improved Pressurized Heavy Water Reactor (PHWR), which is a type of CANDU reactor based on the systems and components of operating CANDU plants, is under development. The improved PHWR has a 480 fuel channel calandria, uses 37 element natural uranium fuel bundles and has a single unit containment. Adaptation of a steel-lined containment structure and improved containment isolation systems permit a reduced exclusion area boundary (EAB) compared to the existing larger capacity CANDU reactors (Darlington, Bruce B). The improved PHWR buildings are arranged to achieve minimum spacing between reactor units. Plant safety and economy are increased through various design changes based on the operating experience of existing CANDU plants. 4 refs. (Author)

  15. Separating rings detection in fuel channels of Embalse NPP

    International Nuclear Information System (INIS)

    The design specifications of Embalse Nuclear Power Plants (CANDU Type Reactor 600Mw) define the positions to be taken by 4 separating rings of the fuel channels. Experience has demonstrated the displacement possibility of the above mentioned rings. It means a risk of contact between pressure tube and calandria tube. In order to determine the position of separating rings, an inspection system based on Eddy Currents technique was developed by CNEA personnel. Detection is performed through two special probes operating according the ''emitter-receiver'' principle. Obtained signals and its relative position are recorded in a video tape and registered in paper. The probe is telecommanded by an automatic equipment. In this paper the construction and calibration of the detection equipment is described, as well as the propulsion. Final results are also outlined in the inspection carried out in November 1986 when an effective displacement of separating rings was verified from its design position in most of the inspected tubes

  16. CANDU heat sinks improvements as a follow up to Fukushima Daiichi accident ''the regulator perspective''

    Energy Technology Data Exchange (ETDEWEB)

    Mesmous, Noreddine; Harwood, Chris [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-06-15

    The purpose of this paper is to provide a summary of the Canadian Nuclear Safety Commission (CNSC) recommendations related to improving the heat sink strategy as a follow up to the Fukushima Daiichi Accident (FDA). As a follow up to FDA, CNSC staff tasked the Nuclear Power Plant (NPP) licensees to review the lessons learned from the FDA and re-examine the NPP safety cases. The reviews have examined the CANDU defence-in-depth strategy and considered events more severe than those that have historically been regarded as credible, and evaluated their impact on the NPPs safety. Availability of emergency equipment was shown to be crucial during the FDA and its availability could have arrested the accident progression early enough to minimize any radioactive release to the environment. As a result, licensees presented appropriate evaluations of the means to provide coolant make-up to the primary Heat Transport System (HTS), boilers, moderator, calandria vault, and irradiated fuel pools.

  17. Inspection procedures for assessment of blister formation/cracking

    International Nuclear Information System (INIS)

    There are various types of inspection that can provide information about a pressure tube's condition with respect to blister formation and cracking. These include spacer location, pressure tube/calandria tube (PT/CT) gap measurement, detection of PT/CT contact, detection of cracked blisters, detection of uncracked blisters, measurement of hydrogen by taking scrape samples and measurement of hydrogen using non-destructive methods. Detection of PT/CT contact and detection of cracked and uncracked blisters are performed using ultrasonic methods. India has developed an ultrasonic inspection capability,/1/, but it has been operated at what we believe is too low a sensitivity,/2/. Current ultrasonic methods and experience related to contact and blister detection are described. 13 refs., 11 figs

  18. Simulating the behaviour of zirconium-alloy components in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, C.E

    2001-12-01

    To prevent failure in nuclear components one needs to understand the interactions between adjacent materials and the changes in their physical properties during all phases of reactor operation. Three examples from CANDU reactors are described to illustrate the use of simulations that imitate complicated reactor situations. These are: swelling tests that led to a method for increasing the tolerance or Zircaloy fuel cladding to power ramps; observations of the behaviour of leaking cracks in Zr-2.5Nb pressure tubes that provide confidence in the use of leak-before-break as part of the defence against flaw development; and contact boiling tests on modifications to the surfaces of Zircaloy calandria tubes that enhance the ability of the heavy water moderator to act as a heat sink after a postulated loss-of-coolant accident. (author)

  19. 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

  20. 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)

  1. Analysis of composite tube sheets with a square array of holes under uniformly distributed load. Part 1

    International Nuclear Information System (INIS)

    The end closures of calandria of reactor generally consists of two thick tube sheet having a square array of holes. At the location of these holes, the tube sheets are connected by lattice tubes. When such a structure is subjected to external loads, restraint is exerted by lattice tube, the shell connecting the tube sheets. The paper summarises different methods of solution adopted by a designer to find out the elastic deformation, stress distribution when the structure is subjected to uniformly distributed load. Assumptions and limitations of different methods which include some classical methods and others developed by the authors have been highlighted. The results obtained for some specific geometries by different methods have been compared. (author)

  2. Evaluation of the mechanical properties of SA 333 Gr.6, AISI 304 and Zr-2.5% Nb through Automated Ball Indentation (ABI) technique

    International Nuclear Information System (INIS)

    Automated Ball Indentation (ABI) technique has been employed in evaluating the tensile property data on three materials, namely SA333 Gr.6 carbon steel (used as PHT piping), AISI 304 (used as calandria vessel) and Zr-2.5% Nb (used as coolant tube) in Pressurised Heavy Water Reactors (PHWRs) with a view to exploring the applicability of ABI technique in providing reliable mechanical property data. The exercise was carried out in cooperation with a second laboratory where conventional tension tests alone were conducted such that the output of the study could be independently monitored and evaluated in an unbiased manner. The results generated in the authors' laboratory were found to be fully in agreement with what were obtained through conventional tension tests. Thus the study has been successful in establishing the reliability of the data obtained through miniature route especially in the case of coolant tube which has immense applications. (author)

  3. 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)

  4. MAPLE: a Canadian multipurpose reactor concept for national nuclear development

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited, following an investigation of Canadian and international needs and world-market prospects for research reactors, has developed a new multipurpose concept, called MAPLE (Multipurpose Applied Physics Lattice Experimental). The MAPLE concept combines H2O- and D2O-moderated lattices within a D2O calandria tank in order to achieve the flux advantages of a basic H2O-cooled and moderated core along with the flexibility and space of a D2O-moderated core. The SUGAR (Slowpoke Uprated for General Applied Research) MAPLE version of the conept provides a range of utilization that is well suited to the needs of countries with nuclear programs at an early stage. The higher power MAPLE version furnishes high neutron flux levels and the variety of irradiation facilities that are appropriate for more advanced nuclear programs

  5. Molten fuel-moderator interaction

    International Nuclear Information System (INIS)

    A critical review of the current understanding of vapor explosions was carried out. It was concluded that, on the basis of actual industrial accidents and large scale experiments, energetic high yield steam explosion cannot be regarded as an improbable event if large quantities of molten fuel and coolant are mixed together. This study also reviewed a hydrodynamic transient model proposed by Henry and Fauske Associates to assess a molten fuel-moderator interaction event. It was found that the proposed model negates a priori the possibility of a violent event, by introducing two assumptions: 1) fine fragmentation of the molten fuel, and ii) rapid heat transfer from the fine fragments to form steam. Using the Hicks and Menzies thermodynamic model, maximum work potential and pressure rise in the calandria were estimated. However, it is recommended that a more representative upper bound model based on an underwater explosion of a pressurized volume of steam be developed

  6. Simulation of hydrogen migration and blisters formation in zirconium alloys

    International Nuclear Information System (INIS)

    The phenomenon of hydrogen migration and hydride blister growth after pressure tube/calandria tube contact in CANDU reactors is addressed. This phenomenon is by now regarded as an important factor limiting reactors lifetime, since it originated Pickering incident in 1983. Numerical results of thermally-assisted diffusion in excellent agreement with quasi-analytical solutions of the mathematical model were obtained. A sensitivity analysis was performed to assess the accuracy of these results. Some two-dimensional calculations are also included to demonstrate the capabilities of the numerical methods. The main outcomes of the work are the following: a through understanding of the mathematics and physics involved in hydrogen migration under thermal gradients. The validation of a numerical procedure based on a regularization of the constitutive equations. Blister growth rates in slab geometries for initial concentrations that span the full range of technological interest. Some preliminary two-dimensional results allow the design of future developments. (Author)

  7. Candu heat sinks improvements as a follow up to Fukushima Daiichi accident 'the regulator perspective'

    International Nuclear Information System (INIS)

    The purpose of this paper is to provide a summary of the Canadian Nuclear Safety Commission (CNSC) recommendations related to improving the heat sink strategy as a follow up to the Fukushima Daiichi Accident (FDA). As a follow up to FDA, CNSC staff tasked the Nuclear Power Plant (NPP) licensees to review the lessons learned from the FDA and re-examine the NPP safety cases. The reviews have examined the CANDU defence-in-depth strategy and considered events more severe than those that have historically been regarded as credible, and evaluated their impact on the NPPs safety. Availability of emergency equipment was shown to be crucial during the FDA and its availability could have arrested the accident progression early enough to minimize any radioactive release to the environment. As a result, licensees presented appropriate evaluations of the means to provide coolant make-up to the primary Heat Transport System (HTS), boilers, moderator, calandria vault, and irradiated fuel pools (authors)

  8. 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)

  9. Simulating the behaviour of zirconium-alloy components in nuclear reactors

    International Nuclear Information System (INIS)

    To prevent failure in nuclear components one needs to understand the interactions between adjacent materials and the changes in their physical properties during all phases of reactor operation. Three examples from CANDU reactors are described to illustrate the use of simulations that imitate complicated reactor situations. These are: swelling tests that led to a method for increasing the tolerance or Zircaloy fuel cladding to power ramps; observations of the behaviour of leaking cracks in Zr-2.5Nb pressure tubes that provide confidence in the use of leak-before-break as part of the defence against flaw development; and contact boiling tests on modifications to the surfaces of Zircaloy calandria tubes that enhance the ability of the heavy water moderator to act as a heat sink after a postulated loss-of-coolant accident. (author)

  10. Development of liquid poison injection system (SDS-2) for 500 MWe PHWRs

    International Nuclear Information System (INIS)

    A secondary shut-down system (SDS-2) in the form of a mecahnism for introducing poison into the moderator of the PHWR is under development in Reactor Engineering Division of BARC. The system, as conceived, consists of a tank containing pressurised helium connected to poison tanks through quick opening solenoid valves. The tanks are connected to horizontal injection tubes in the calandria. On system actuation, gadolinium nitrate solution from the tanks passes to the injection tubes which have a number of holes through which the poison enters the moderator. This report details the development work being done on this poison injection system. An experimental facility was set up to measure the poison jet growth rate and the jet spread after injection, and mathematical models were developed to convert the observed jets into reactivity worth values. A description of the work and the computed results are presented. (author). 21 graphs. , 15 tabs

  11. Experimental study of poison moderator interface movement for shut down system #2(SDS#2) of 540 MWe PHWR

    International Nuclear Information System (INIS)

    The poison solution and the moderator in Secondary Shutdown System (SDS-2) of 500 MWe PHWR, are separated by their own liquid in liquid interface, termed as poison moderator interface (PMI). During normal operation of the reactor, the interface moves towards the calandria, mainly because of molecular diffusion from poison to moderator. Other reasons for movement are mixing of poison and moderator due to physical disturbances in the moderator level and to some extent due to temperature difference between the two liquids. The electrical conductivity of these liquids was found to be the most reliable parameter indicating interface movement. For this purpose, two on-line high-pressure conductivity probes have been installed on moderator side for each one of the six poison tanks. During normal operation of reactor, the interface moves slowly towards the calandria over a period of time and gives rise to increase in conductivity. To study the interface pattern and factors affecting the same, a full-scale experimental setup was developed and series of experiments carried out. The experimental results showed that the interface is quite stable and annunciation can be placed around 100 micro siemens/cm before back flushing is initiated. One dimensional diffusion analysis of the obtained experimental data showed that the derived model for PMI setup with diffusion parameter of 900 cm2/hr is able to predict the interface movement quite satisfactorily. This report gives an insight into the experiments carried out for estimation of the effective diffusion parameter for the poison moderator interface, model formulation and its prognostic behavior. (author)

  12. 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

  13. Dynamic analysis of a reactor building on alluvial soil

    International Nuclear Information System (INIS)

    The reactor building consists of reinforced concrete internal framed structure enclosed in double containment shells of prestressed and reinforced concrete all resting on a common massive raft. The external cylindrical shell is capped by a spherical dome while the internal shell carries a cellular grid slab. The building is partially buried under ground. The soil consists of alluvial going to 1000 m depth. The site lies in a moderate seismic zone. The paper presents the dynamic analysis of the building including soil-structure interaction. The mathematical model consists of four parallel, suitably interconnected structures, namely inner containment, outer containment, internal frame and the calandria vault. Each one of the parallel structures consists of lumped-mass beam elements. The soil below the raft and on the sides of outer containment shell is represented by elastic springs in both horizontal and vertical directions. The various assumptions required to be made in developing the mathematical model are briefly discussed in the paper. Transfer matrix technique has been used to determine the frequencies and mode shapes. The deformations due to bending, shear and effect of the rotary inertia have been included. Various alternatives of laterally interconnecting the internals and the shells have been examined and the best alternative from earthquake considerations has been obtained. In the study, the effect of internal structure flexibility and Calandria vault flexibility on the whole building have been studied. The resulting base raft motion and the structural timewise response of all floors have been determined for the design basis (safe shutdown) earthquake by mode superposition

  14. 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)

  15. The Effect of Containment Filtered Venting System on the Severe Accident Management Strategies of the CANDU6 Plant

    International Nuclear Information System (INIS)

    In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout and severe core damages and released a large amount of radioactive materials outside of the plants. After this accident Nuclear Safety and Security Commission (NSSC) decided to install a filtered containment venting system (CFVS) at all the operating nuclear power plants in Korean. To comply with NSSC's request, Wolsong Unit 1 has installed a CFVS. Current severe accident management guidance, which does not consider a CFVS has 6 severe accident management strategies for CANDU6 plant. These strategies are inject in to the primary heat transport system (PHTS), inject in to the calandria, inject into the calandria vault, reduce fission product releases, control containment conditions, reduce containment hydrogen. The CFVS is designed to open and to close isolation valves by an operator. An operator opens the CFVS isolation valve when the containment pressure exceeds the design pressure (124 kPa(g)) and closes isolation valves when the containment pressure decreases below 50 kPa(g). The operation of the CFVS not only influences the current strategies (adds a means of controlling containment conditions) but also requires the new strategies. This paper discusses the necessity of the new strategies, such as the prevention of containment vacuum and the injection into the containment. The necessity of the additional severe accident management strategies for CANDU6 plants which installed a CFVS is evaluated. The operation of a CFVS affects the water inventory in the basement also, but not significantly. The SBO accident requires the water injection into the containment at least 4 days after an accident initiation if a passive spray system fails. If a spray system operates, then the injection into the containment is required more than 10 days after an accident initiation even though a CFVS operates

  16. Study on protective layer for severe accident conditions for EC6 reactor vault structure

    International Nuclear Information System (INIS)

    The Enhanced CANDU 6 (EC6) is designed both for the prevention and mitigation of Design Basis Accidents (DBAs) as well as Beyond Design Basis Accidents (BDBAs). The foremost objective, in accordance with the safety goals specified in the CNSC Regulatory Document (RD-337), is to prevent the occurrence of any accident that could jeopardize nuclear safety, and, if an accident should occur, to limit the radiological releases resulting from the accident and minimize the impact on nearby communities. During a postulated severe core accident, Molten Core-Concrete Interaction (MCCI) may occur when molten core debris breaches the calandria vessel and contacts concrete surfaces, whereby the thermal and chemical properties of the melt contribute to the potential degradation of the concrete. The earliest phase of MCCI is characterized by very-high-temperature molten metal and oxide pouring from the calandria vessel and settling as a pool on the concrete surfaces of the vault floor. The molten material can result in spalling or fragmentation of the concrete near where the corium first contacts the concrete. As the corium settles on the concrete surface, the melt begins to react chemically with the concrete through the penetrating cracks and fragments produced on the initial contact, generating various gases including carbon monoxide and combustible hydrogen. In order to control and mitigate MCCI, a protective layer (refractory material) with suitable material properties and sufficient thickness was proposed to protect the reactor vault concrete floor. To further enhance vault floor protection and mitigate the conditions under severe accidents a special concrete composition in the upper layer of the vault floor concrete is to be provided in case the refractory material is breached. This special concrete should minimize the generation of various gases including combustible hydrogen and carbon monoxide during MCCI. As a part of research and development program an experimental

  17. Station Blackout Initiated Event Chronology in LWR/HWR NPP

    International Nuclear Information System (INIS)

    Since the crisis at Fukushima nuclear power plants, a severe accident progression has been recognized as a very important area for an accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of a severe accident progression among the typical pressurized water reactor (PWR), boiling water reactor (BWR) and pressurized heavy water reactor (PHWR). The OPR 1000-like (ABB-CE type PWR), Peach Bottom-like (BWR/4 RCS with a MARK I Containment), and Wolsong1-like (CANDU6 type) plants are selected as reference plants of typical 1000 MWe PWR, 1140MWe BWR, and 600 MWe PHWR, respectively. The design parameters of these plants are quite different. Some of the major different design features of CANDU6 plant from other light water reactors, in terms of a severe accident, are that the plant adopts a duel primary heat transport system and has an additional amount of cooling water in the calandria vessel (calandria tank, CT) and calandria vault (CV). Another feature is that the CT is always submerged in water because the CV is flooded during normal operation. The containment (reactor building, R/B) failure pressure of the CANDU6 plant is considerably lower than that of the typical PWR or BWR4/MARK-I. The containment vessel free volume of MARK-I is much smaller than that of the PWR or CANDU6 plant. Since there is no steam generator (SG) or passive cooling system, the amount of cooling water inventory in BWR4 is relatively less than other plants. Meanwhile the minimum available time of battery power against station blackout (SBO) accident is different among plant types: six hours for BWR4 and four hours for 1000MWe PWR. Therefore, plant responses against the severe core damage scenarios like Fukushima accident are expected to be much different. By identifying plant response signatures, the appropriate correction actions can be developed as part of severe accident management. A SBO scenario, where all off-site power is lost

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  19. Station Blackout Initiated Event Chronology in LWR/HWR NPP

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Since the crisis at Fukushima nuclear power plants, a severe accident progression has been recognized as a very important area for an accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of a severe accident progression among the typical pressurized water reactor (PWR), boiling water reactor (BWR) and pressurized heavy water reactor (PHWR). The OPR 1000-like (ABB-CE type PWR), Peach Bottom-like (BWR/4 RCS with a MARK I Containment), and Wolsong1-like (CANDU6 type) plants are selected as reference plants of typical 1000 MWe PWR, 1140MWe BWR, and 600 MWe PHWR, respectively. The design parameters of these plants are quite different. Some of the major different design features of CANDU6 plant from other light water reactors, in terms of a severe accident, are that the plant adopts a duel primary heat transport system and has an additional amount of cooling water in the calandria vessel (calandria tank, CT) and calandria vault (CV). Another feature is that the CT is always submerged in water because the CV is flooded during normal operation. The containment (reactor building, R/B) failure pressure of the CANDU6 plant is considerably lower than that of the typical PWR or BWR4/MARK-I. The containment vessel free volume of MARK-I is much smaller than that of the PWR or CANDU6 plant. Since there is no steam generator (SG) or passive cooling system, the amount of cooling water inventory in BWR4 is relatively less than other plants. Meanwhile the minimum available time of battery power against station blackout (SBO) accident is different among plant types: six hours for BWR4 and four hours for 1000MWe PWR. Therefore, plant responses against the severe core damage scenarios like Fukushima accident are expected to be much different. By identifying plant response signatures, the appropriate correction actions can be developed as part of severe accident management. A SBO scenario, where all off-site power is lost

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

    International Nuclear Information System (INIS)

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

  1. Theoretical aspects concerning the aging behaviour of the loop of clamp spacer string in the fuel channel

    International Nuclear Information System (INIS)

    Aging of the pressure tube, a principal component of the CANDU fuel channel is a well established fact. Also known are the effects of aging upon the increase of the diameter as well as residual elongation of the pressure tube under temperature and intense radiation field influence (which after 15 years operation amount Δd ≅ 1.5 mm and Δl ≅ 75 mm, respectively). In conditions of conservation of the material volume the exterior diameter of the pressure tube increases by about 0.08 mm/year what leads effectively to an additional effort on the elastic support and, consequently, to an additional compression of the clamp string loops contacting the pressure and the calandria tubes. A theoretical computation showed that the force statically distributed upon the contact loop increases up to a value which can exceed several times the initial value after a number of years of operation. Tacking into account the fact that in running fuel channels there are accepted as admissible amplitudes of the displacement signal under the vibrations, induced by the primary coolant flow, of about 300 μm (peak-peak), the peak value of the force applied upon contact loops increases correspondingly. Consequently, the dynamical spectrum of the pressure tube vibrations will be superimposed upon the stressed clamp string contact loops. Contact loop after twisting will deform at bending. When after twisting the contact loops remain under load, the loop cross section is stressed at deformation under the action of shear axial forces and bending moment, respectively. In this work the issue was theoretically approached by means of Mohr-Maxwell method for a loop quarter. From the theoretical computation it was obtained that the contact string loop acquires in time a continuous residual deformation (some μm/year) so that one can anticipate that the contact loop will become oval-shaped. At that moment the loop will be not able to twist anymore, the stress variation will be entirely absorbed by

  2. Theoretical aspects regarding behaviour at aging of the garter-string spacer coil in the fuel channel

    International Nuclear Information System (INIS)

    Aging of the pressure tube, the main component of the CANDU fuel channel, is an established fact. There are also known the effects of aging on the diameter's growth as well as on the permanent elongation of the pressure tube under the influence of temperature and of the strong radiation field (after fifteen years of operation Δd ≅ 1.5 mm, Δl ≅ 75 mm, respectively). The outer diameter of the pressure tube grows by ≅ 0.08 mm/year under material volume preservation conditions what effectively leads to an additional load in the elastic support, and to an additional compression of the garter string coils in contact with the pressure and the calandria tubes, respectively. After a few years of operation it comes out, by theoretical calculus, that the static load distributed on the contact coil grows to a value that can exceed a few times the initial value. Taking into account the fact that in the operation of the fuel channel, amplitudes of the signals from movement at vibrations induced by the primary coolant flow of about 300 μm (peak to peak) are accepted as allowable, the peak value of the load distributed on the contact coil, grows accordingly. Therefore, on the garter string coils statically stressed in the contact area will overlap the dynamic spectrum of the pressure tube vibrations. After torsion, the contact coil will support a torsion deflection. After torsion of the contact coils, in condition of keeping them under stress, the coil's section is stressed for deformation under the influence of the cutting axial loads and of the torsion torque, respectively. In the paper, the issue is theoretically approached using Mohr-Maxwell method on the coil's quarter. It is assessed, by theoretical calculation, that in time the contact string's coil supports a continuously remanent deflection (few μm/year). Anticipating that after a time the contact coil will be such oval so that it could not be twisted. At this moment the variation of the stress is absorbed

  3. Validation of a CATHENA fuel channel model for the post blowdown analysis of the high temperature thermal-chemical experiment CS28-1, I - Steady state

    International Nuclear Information System (INIS)

    To form a licensing basis for the new methodology of the fuel channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown fuel channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal-chemical experiment CS28-1. As the major concerns of the post-blowdown fuel channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H2 at how high a fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought. The result shows that once the pressure tube temperature is predicted correctly by the CATHENA heat transfer model between the pressure tube and the calandria tube through a gap thermal resistance adjustment, all the remaining temperatures of the inner ring, middle ring and outer ring FES temperatures can be predicted quite satisfactorily, say to within an accuracy range of 20-25 deg. C, which is comparable to the reported accuracy of the temperature measurement, ±2%. Also the analysis shows the choice of the emissivity of the solid structures (typically, 0.80, 0.34, 0.34 for FES, PT, CT), and the thermal resistance across the CO2 annulus are factors that significantly affect the steady state prediction accuracy. A question on the legitimacy of using 'transparent' assumption for the CO2 gas annulus for the radiation heat transfer between the pressure tube and the calandria tube in CATHENA code

  4. 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

  5. Development of new MMS modules for CANDU reactors, using CompGen program

    International Nuclear Information System (INIS)

    Modelling of thermodynamic processes in stationary and dynamic regimes are very important for nuclear power plants from both design and operation point of view. Since safety requirements for NPPs are higher than the ones for classical power plants simulation analyses are necessary in this field. In the last years there where developed and improved many computing codes which can be used in such analyses. One of these codes is MMS. MMS package was developed by Framatome Technologies, Bacok and Wilcox and nHance Technologies Inc. It uses a modular architecture based on ACSL program. MMS is provided with a rich library of modules including control, electromechanical, gas, water, boundary, fossil and nuclear modules. Unfortunately, the water and nuclear modules were developed initially only for light LWR NPPs. The main problem of using MMS for CANDU NPPs was the lack of heavy water properties. Recently nHance Technologies improved some of existing modules allowing to work also with heavy water systems. However, development of new modules appropriate to CANDU reactors is necessary to simulate various systems in CANDU NPPs. New MMS modules developed by the authors, characteristic for CANDU NPPs concern: Calandria, heat exchanger, delay tank, end shield., fuel storage pool, calandria vault. These modules were built with means of CompGen program, which is included in the MMS package. They were used at Center of Technology and Engineering for Nuclear Projects - CITON, at Bucharest to simulate operational regimes of the primary circuit in Cernavoda NPP. In this present paper the capabilities of MMS package regarding thermal-hydraulic simulation of nuclear systems are shown . Also, presented are the results of a joint research of Power Plants Simulation Laboratory of University 'Politehnica' at Bucharest and CITON on several MMS modules, developed by using CompGen. These results serve for safety analyses for Cernavoda 2 NPP. As an illustration, a 'real' safety analysis for the

  6. 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

  7. 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

  8. The effect of ballooning model in ISAAC

    International Nuclear Information System (INIS)

    The purpose of this paper is to analyze the ballooning effect in the fuel channel with the ISAAC (Integrated Severe Accident Analysis code for CANDU Plants) computer code which was developed for the severe accident analysis at CANDU plants. Cladding/PT ballooning model parameter in the ISAAC code is used to analyze its effect mainly on the hydrogen production from the fuel channel. According to the current ISAAC version, cladding failure occurs when the cladding temperature exceeds a user-specified failure temperature, resulting in a fission products release from the fuel rod to the primary system. The typical accident sequences of a loss of feed water (LOFW) and a large LOCA (LLOCA) scenario are selected. Unlike the PWRs, the ballooning effect in CANDU plants is not so clear during the low pressure sequences. Instead, pressure tube ballooning during the high pressure sequences decreases the amount of hydrogen generated from the fuel when pressure tube-calandria tube contact occurs, resulting in a cladding and pressure tube temperature decrease. (authors)

  9. The Assessment of Advanced HWR with Dual Moderator Concept

    International Nuclear Information System (INIS)

    The advantages of existing Candu-PHWR reactor over other types are, better safety performance because of the low excess reactivity by the use of natural U oxide, the separation of D2O as a moderator and as a coolant, and better neutron economy. The result in the long period power excursion behavior, cool moderator that can act as long term heat sink, and the flexibility in fuel option. The problem of positive feedback in the Candu design is to be overcome by the use of the so called dual moderator concept, in which two moderator systems are used, i.e. a main moderator system outside calandria tube and an annular moderator system inside annular space. The result of this calculation showed that, this concept can achieve not only a negative local power feedback but also promises a greater simplicity in the power regulating system by the possibility of elimination of adjuster rod and incorporation of more than one passive shutdown systems. The boiling of annular moderator can also be used as a power regulating system as well as a core excess reactivity compensation system

  10. low dose irradiation growth in zirconium

    International Nuclear Information System (INIS)

    Low dose neutron irradiation growth in textured and recrystallized zirconium, is studied, at the Candu Reactors Calandria temperature (340 K) and at 77 K. It was necessary to design and build 1: A facility to irradiate at high temperatures, which was installed in the Argentine Atomic Energy Commission's RA1 Reactor; 2: Devices to carry out thermal recoveries, and 3: Devices for 'in situ' measurements of dimensional changes. The first growth kinetics curves were obtained at 365 K and at 77 K in a cryostat under neutron fluxes of similar spectra. Irradiation growth experiments were made in zirconium doped with fissionable material (0,1 at %235U). In this way an equivalent dose two orders of magnitude greater than the reactor's fast neutrons dose was obtained, significantly reducing the irradiation time. The specimens used were bimetallic couples, thus obtaining a great accuracy in the measurements. The results allow to determine that the dislocation loops are the main cause of irradiation growth in recrystallized zirconium. Furthermore, it is shown the importance of 'in situ' measurements as a way to avoid the effect that temperature changes have in the final growth measurement; since they can modify the residual stresses and the overconcentrations of defects. (M.E.L.)

  11. Artificial neural networks in CT-PT contact detection in a PHWR

    International Nuclear Information System (INIS)

    In a pressurized heavy water reactor (PHWR), contact between calandria tubes (CT) and pressure tubes (PT) makes them susceptible to delayed hydrogen cracking. Periodic inspection of the channels must be carried out to detect the contact. As the number of channels in a PHWR is very large (306 in a 230 MW plant) periodic in-service inspection of all the channels leads to an unacceptable downtime. A non-intrusive technique that employs a system identification method is presently used for contact detection. The technique tends to overpredict the number of channels in contact, i.e. they diagnose many channels as contacting while the channels are in fact not in contact. This puts a large number of healthy channels on the at risk list reducing the efficacy of the method. This paper demonstrates the power of artificial neural networks (ANNs) in diagnosing the CT-PT contact. A counterpropagation neural network consisting of a Kohonen layer and a Grossberg layer has been employed. The noise tolerance of the network has been demonstrated. (orig.)

  12. SLARette Mark 2 system

    International Nuclear Information System (INIS)

    The SLAR (Spacer Location and Repositioning) program has developed the technology and tooling necessary to locate and reposition the fuel channel spacers that separate the pressure tube from the calandria tube in a CANDU reactor. The in-channel SLAR tool contains all the inspection probes, and is capable of moving spacers under remote control. The SLAR inspection computer system translates all eddy currents and ultrasonic signals from the in-channel tool into various graphic displays. The in-channel SLAR tool can be delivered and manipulated in a fuel channel by either a SLAR delivery machine or a SLARette delivery machine. The SLAR delivery machine consists of a modified fuelling machine, and is capable of operating under totally remote control in automatic or semi-automatic mode. The SLARette delivery machine is a smaller less automated version, which was designed to be quickly installed, operated, and removed from a limited number of fuel channels during regular annual maintenance outages. This paper describes the design and operation of the SLARette Mark 2 system. 5 figs

  13. NRU vessel repair and return to service: enhancing a Canadian R and D asset for the future

    Energy Technology Data Exchange (ETDEWEB)

    Cox, D.S.; Arnold, J.B.; Lee, J.K., E-mail: coxd@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    The National Research Universal (NRU) reactor was successfully returned to high power operation in 2010 August, after completing extensive inspections and repairs of the calandria vessel, in response to a small leak of heavy water that was discovered in 2009 May. The aluminum alloy vessel material had corroded from the outside surface over many years, and required application of weld build-up and plated weld repair at ten locations, executed by remote tooling deployed from inside the reactor vessel. Many specialized remotely operated tools were developed for inspection, sampling and repair operations. In parallel with the repair efforts, important maintenance activities were performed, enabled by the de-fuelled state with the heavy water drained from the vessel. Two annual inspection cycles have now been completed since the reactor was returned to high power operation, confirming the fitness of the vessel for continued operation. The NRU reactor is now entering its 56th year of operation and the current operating licence extends to 2016. Based on the outcome of a comprehensive Integrated Safety Review, AECL is continuing to implement major equipment upgrades and process improvements to support safe and reliable operation through 2021, for the benefit of Canadians and the world. (author)

  14. NRU vessel repair and return to service: enhancing a Canadian R and D asset for the future

    International Nuclear Information System (INIS)

    The National Research Universal (NRU) reactor was successfully returned to high power operation in 2010 August, after completing extensive inspections and repairs of the calandria vessel, in response to a small leak of heavy water that was discovered in 2009 May. The aluminum alloy vessel material had corroded from the outside surface over many years, and required application of weld build-up and plated weld repair at ten locations, executed by remote tooling deployed from inside the reactor vessel. Many specialized remotely operated tools were developed for inspection, sampling and repair operations. In parallel with the repair efforts, important maintenance activities were performed, enabled by the de-fuelled state with the heavy water drained from the vessel. Two annual inspection cycles have now been completed since the reactor was returned to high power operation, confirming the fitness of the vessel for continued operation. The NRU reactor is now entering its 56th year of operation and the current operating licence extends to 2016. Based on the outcome of a comprehensive Integrated Safety Review, AECL is continuing to implement major equipment upgrades and process improvements to support safe and reliable operation through 2021, for the benefit of Canadians and the world. (author)

  15. In-situ radiation damage in Mg

    International Nuclear Information System (INIS)

    The creep rate in the low flux regions at the ends of the fuel channels is important for predicting fuel channel sag and pressure tube-to-calandria tube contact. This creep rate is partly determined by the accumulation of radiation damage, which is a function of flux. There appears to be a threshold flux, below which no dislocation loop nucleation occurs in the matrix of the material. This may or may not correspond to a condition of maximum hardening. We are attempting to verify this observation by examining the effect of dose rate on radiation damage in Mg. These observations will be useful in interpreting the power reactor data for pressure tubes. Magnesium specimens of 99.9% purity were prepared by electropolishing and irradiated in a transmission electron microscope (TEM) at 300 kV. Images were collected using a charge coupled device (CCD) camera and used to compare equivalent accumulated damage for low and high flux conditions. Experiments show that a threshold flux exists, with generally no observation of new dislocation loops below the threshold flux, and nucleation of new dislocation loops above the threshold flux. Nucleation of dislocation loops occurs heterogeneously both at pre-existing network dislocations and at other sites within the matrix. Although preferential nucleation at dislocations is anticipated, the heterogeneous nucleation within the matrix appears to be related to the presence of small precipitates. Very high purity Mg will be used to verify this hypothesis. (author)

  16. 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)

  17. Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Young; Jung, Yun Sik; Hwang, Gi Suk; Kim, Nam Seok [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2004-02-15

    The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items the one is the experimental study of entrainment at horizontal pipe with {+-} 36 .deg. C , {+-} 72 .deg. C branch pies, the other is the model improvement of the moderator heat sink in the Calandria. The off-take experiments on onset of entrainment and branch quality were investigated by using water and air as working fluid, and the experimental data were compared by the previous correlations. The previous correlations could not expect experimental results, thus the weak points of the previous correlations were investigated. The improvement of the previous model continues as the next year research. The thermal hydraulic scaling analysis of SPEL, STERN and ideal linear scaling analysis have been studied. As a result, a new scaling method were needed to design a new experimental facility (HGU). A new scaling method with 1/8 length scale was applied. From these results, the thermal hydraulic model for CFD code simulation was designed and test apparatus has been made. The moderator temperature distribution experiments and CFD code simulation will be continued in next year.

  18. ACR-1000® end-temperature peaking analysis under postulated accident conditions

    International Nuclear Information System (INIS)

    This paper presents a novel and systematic approach to conduct end-temperature peaking analysis under accident conditions for an ACR-1000 reactor, using a two-dimensional (radial and axial) finite-element computer code FEAT. In the past, end-flux peaking effects were overly conservatively assessed by including power increase in the fuel end region without accounting for heat transfer enhancement due to flow disturbance at the bundle end region, especially at the down-stream of a bundle junction. The current analysis determines the end-flux-peaking induced increase in fuel sheath and fuel centreline temperatures while accounting for all relevant key phenomena such as end-flux peaking and heat transfer characteristics including the effects of flow/thermal boundary layer redeveloping at the bundle end region. Using this method significantly reduces the fuel sheath temperature increase caused by end-flux peaking in comparison with the conservative analysis. The postulated accident events considered in this analysis include large break loss-of-coolant accident (LOCA), small break LOCA, and pressure tube rupture within an intact calandria tube. The determined temperature increases relative to the case without end-flux peaking are required to be quantitatively included in detailed safety analyses for postulated accidents. (author)

  19. 1982-83 annual report

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd. continued to improve on its operating results and financial position with record net earnings of $22.6 million, an increase of 15 percent. Commercial revenue from exports was 53 percent. Revenue from nuclear supply and services declined to $218.9 million as reactor projects neared completion or were deferred. Radiation equipment and isotope sales increased by 25 percent to $84.3 million. This revenue included the first shipment of a Canadian cancer therapy accelerator, isotope production from the new cyclotron, and radiopharmaceutical products. Pickering Generating Station revenue was $38.1 million, an increase of 28 percent. Heavy water production increased by 21 percent and unit operating costs decreased by 9 percent due to increased productivity. Operating profit from commercial operations of $25.8 million was slightly lower, due to increased product development costs and marketing expenses. Studies were made of the long-term behaviour of pressure and calandria tubes in CANDU reactors. Cesium and iodine released from fuel in a hypothetical accident would be retained within the system to a much greater degree than was previously believed, according to other basic research studies. Comprehensive safeguards systems and equipment developed in collaboration with the IAEA were installed on the four 600MW CANDU units which began operating. Experiments testing a process to extract useful by-products from used CANDU fuel were completed using facilities in Italy. Future developments using by-products in combination with uranium or thorium will ensure CANDU's long-term viability

  20. 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)

  1. 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

  2. Effects of the plastic deformation and thermal cycles on the mechanical properties of fully recrystallized Zircaloy-4

    International Nuclear Information System (INIS)

    The development of crystallographic texture in a product depends, for a given material, of its fabrication history. In our case, the evolution of that texture results from a combination of cold working and thermal cycles applied together or separately. In the present work, cold working levels ranging from 50 % to approximately 90 % and different heat treatment cycles has been applied to Zircaloy-4 sheets and tubes. Using X-ray diffraction techniques and the direct pole figure method, the evolution of crystallographic texture has been analyzed for each fabrication route. We observed that cold working levels up to 90 % without intermediate annealing heat treatment do not change significantly the classic angle between basal pole and the normal/radial direction of the product (φ ≅ ± 25 degrees). Furthermore, the application of intermediate cold working levels (50 % - 60 %) and more than two intermediate annealing heat treatments exhibits a marked modification of the basal pole orientation. The basal poles appear now parallel to the normal direction (φ ≅ 0 degrees) of the product. Additionally, the crystallographic texture change observed with X-ray procedures was evaluated by the measure of anisotropic parameters R and P. The results here obtained will be use in the future as a basis for the design of a fabrication route capable to obtain in a HPTR process, seamless calandria tubes strengthened by crystallographic texture. (author)

  3. Development of demonstration advanced thermal reactor

    International Nuclear Information System (INIS)

    The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

  4. Strain model and zircaloy-steam reaction model in the TUF code and their application in large LOCA

    International Nuclear Information System (INIS)

    As an integral part of a generic study of the Emergency Coolant Injection System effectiveness in Ontario Hydro reactors during a large break Loss of Coolant Accident (LOCA), the TUF (Two-Unequal-Fluids) code has been developed to enhance safety analysis capability. Recent enhancement to the TUF code includes the pressure tube transverse strain model and the zircaloy-steam reaction model. These models are employed to predict thermal-mechanical response of fuel channels and determine the thermal heat load to the moderator during postulated large LOCA scenarios. Presented in this paper are the description of the models, the cross-code comparison of the predictions between the TUF code and the SMARTT code and the discussion of parameters that may affect the pressure tube strain and the effect of pressure tube ballooning into contact with the calandria tube on the system response simulations. The modified TUF code is employed to quantify the extent of pressure tube ballooning and to calculate the thermal heat load to moderator. (author) 14 refs., 4 tabs., 16 figs

  5. 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

  6. 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.

  7. 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

  8. 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)

  9. 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

  10. 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

  11. 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)

  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. Advanced in fuel channel gauging tool - instrumenting a SLAR tool for dual purpose

    International Nuclear Information System (INIS)

    This paper describes the latest inspection technology to be implemented on a SLARette tool. In 2002, a gauging module was developed and qualified to replace the SLARette tool's blister module. This gauging module had five ultrasonic transducers for diameter creeping and wall thickness measurements. The results of the 2002 SLARette campaign were excellent; the data obtained came within ten microns of that collected with a CANDE tool in 2003. Gentilly-2 decided to continue developing gauging techniques and apparatus to be mounted on a SLARette tool. The new front-end module incorporates both sag measurement and revolutionary pressure tube (PT)/calandria tube (CT) gap modules. Advancements were made along several lines: (1) selection of a radiation-resistant sag module, (2) development of a sag simulator, (3) improvement of AECL gap measurement technology and finally, (4) design of a front-end encompassing module with motorized lift-off capability. This front-end module is only 19 centimeters long and is capable of performing all of the gauging measurements required for fuel channel life-cycle management. This paper will detail the development efforts of Hydro-Quebec, IREQ and AECL in improving fuel channel gauging technology, as well as the implementation and field results of the 2005 Gentilly-2 inspection campaign. (author)

  14. Repairs in 104 Gy/h?

    International Nuclear Information System (INIS)

    In 1989 it was found that in each unit of the MAPS Candu nuclear power station in India the centre portion of the heavy water inlet manifold opposite the 300 mm inlet pipe had torn away. Equipment for remote inspection, repair and removal of debris in an area with constricted and difficult access and radiation fields of about 104 Gy/h was developed by Ricardo Hitec of the United Kingdom. This consists of a work performing manipulator, TV viewing systems and a posting tube manipulator for insertion of tools and debris containers into the calandria. A variety of special end effector and tools were also developed jointly with AEA Technology. A first repair campaign was carried out on Unit 1 in 1991. Following a detailed TV survey of the damage a reappraisal of the situation was undertaken and a programme of equipment enhancement carried out. In July 1992 a second repair campaign took place on Unit 2. The difficulties encountered and the degree of success achieved are described. Work proceeded at an intensive level for 14 days when the campaign was ended in view of the exhaustion both of personnel and the equipment. Although more work could have been done a major improvement had been achieved. (UK)

  15. A layman's guide to radiation-induced deformation processes in zirconium alloys

    International Nuclear Information System (INIS)

    The fuel channel (comprising a pressure tube and a calandria tube fabricated from zirconium alloys) in a CANDU reactor undergoes shape changes because of radiation-induced deformation. This is a consequence of the microstructural modification arising from radiation damage produced by the fast-neutron flux. This report summarizes our current understanding of the physical processes responsible for the deformation. With the non-specialist reader in mind, the underlying mechanisms are described in a manner that avoids much of the associated technical terminology. Thus, the basic concepts of plasticity in a crystalline material are introduced and related to the various microstructural defects created during irradiation. In particular, the mechanisms of creep (a time-dependent strain activated by an applied stress) and growth (a time-dependent strain occurring in the absence of stress) are discussed in a non-technical language assisted by simple diagrams. Reference is made to both theoretical investigations (avoiding mathematical complexity) and experimental measurements. It is shown how the qualitative and quantitative knowledge can be used to derive a predictive model for reactor designers and operators. The current status of such a model is evaluated and suggestions for future improvements made

  16. Progress report, Chemistry and Materials Division, 1 April - 30 June, 1981

    International Nuclear Information System (INIS)

    The work of the Division in the areas of solid state science, radiation, physical and analytical chemistry, and materials science during the quarter is described. Measurements of ion stopping power have emphasized the importance of axial symmetry and may be used to show the contribution of nuclear inelastic events to stopping processes. Enhancement of ion scattering at 180 degrees can occur even in the first few layers of a single crystal of gold implanted with heavy atoms. Agreement has been obtained between experimental and calculated rates for dechanneling of protons in gold. The rate of decomposition of HOI in aqueous solutions has been determined. The effects of radiation on dithiothreitol is being studied. Laser photochemistry work includes investigations of multiphoton dissociation and of laser-induced zirconium isotope separation. A method has been found for the preparation of oxygen gas samples for the determination of oxygen isotope ratios in water, and high-performance liquid chromatography has been applied to metals in ground water. Sputtered coatings of stainless steel on the surface of zircaloy fuel cladding reduce the oxidation rate in steam. A theoretically-based design equation for irradiation growth of pressure tubes has been developed. Studies on the effect of small strains on zircaloy-2 tubing show the need to avoid even small amounts of compressive deformation of calandria tubes

  17. Development and qualification of liquid poison injection system (SDS-2) for 500 MW(e) PHWRs

    International Nuclear Information System (INIS)

    The Shutdown System No. 2 (SDS-2) in the form of a mechanism for introducing poison into the moderator of the PHWR is under development in RED. The system, as conceived, consists of a tank containing pressurised helium connected to poison tanks through quick opening solenoid valves. The tanks are connected to horizontal perforated poison injection tubes in the calandria. On system actuation, gadolinium nitrate solution from the tanks passes to the injection tubes which have a number of holes through which the poison enters the moderator in the form of high speed sub-merged jets. This paper details the experimental studies being carried out on the development of high speed colour water (representing poison) jets into plain water (representing moderator). An experimental facility (Phase-I) was set up to measure the circular and slit jet growth rate, jet spread, discharge rate and interaction of multiple jets. Based on the results, a jet growth equation was formulated. Later a full scale mock-up for one injection unit (Phase-2) was commissioned and various experiments up to injection pressure of 100 kg/sq cm were conducted to verify predicted results and check the performance of full integrated system. A description of the work and the computed results are presented here. (author)

  18. Assessment of the integrity of KANUPP fuel channels

    International Nuclear Information System (INIS)

    The KANUPP reactor first produced power in 1972. In 1983 one of the fuel channels, G12 was difficult to refuel due to the South end fitting being retracted relative to the adjacent channels. In November 1993, eight fuel channels were inspected non-destructively for defects. The dimensions and shape of the pressure tubes were measured, and channel G12 was removed for examination and testing at the Chalk River Laboratories. Examination of channel G12 showed that the South end calandria tube rolled joint had been leaking moderator heavy water into the annulus. The inboard bearing was completely corroded away which resulted in the South end fitting seizing in the debris. The Inconel X750 garter springs were both completely broken into small rings due to Stress Corrosion Cracking (SCC) in acids produced by radiolysis of the moderator heavy water. The inspection of the eight fuel channels did not reveal any serious defects. The length and diameter changes in the pressure tubes were both small. The deuterium ingress had been low with the result that the hydrogen isotope concentration of the pressure tubes was below Terminal Solid Solubility (TSS) even close to the end fittings. Irradiation had increased the strength and reduced the ductility and fracture toughness of the tubes and the Delayed Hydride Cracking (DHC) properties were similar to irradiated cold worked Zr-2.5Nb pressure tubes. The examination and testing showed that the KANUPP pressure tubes are all probably in good condition and fit for several more years service. (author)

  19. Effect of extrusion variables on crystallographic texture of Zr-2.5 wt% Nb

    International Nuclear Information System (INIS)

    In this paper, we describe the effects of extrusion variables on the crystallographic texture of Zr-2.5% Nb tubes extruded in the α + β phase field. We compare the results from small experimental extrusions (16 mm diameter) and full size pressure tube and calandria tube extrusions (50 to 125 m diameter) made using commercial presses. For both small and large tubes increasing the extrusion ratio increases the proportion of basal plane normals in the transverse direction. Superimposed on this are several other effects observed on the small tubes: increasing extrusion ratio reduces the proportion of basal plane normals in the longitudinal direction; increasing temperature produces a small concentration of basal plane normals in the longitudinal direction and moves the majority of basal plane normals farther towards the transverse direction; increasing the billet grain size moves the basal plane normals towards the radial direction. These effects can be explained by considering competing systems of deformation mechanism acting simultaneously. One system of deformation mechanism, similar to that during cold forming, rotates the basal plane normals towards the direction of major compressive strain. A second system rotates the basal plane normals away from both the direction of major compressive strain and the extrusion direction. (orig.)

  20. Reactivity initiated accidents and loss of shutdown - 20 years later

    International Nuclear Information System (INIS)

    A review of the safety of Ontario's nuclear power reactors was conducted in 1987 after the Chernobyl accident. As part of this review an analysis was performed of a Loss of Coolant Accident in a Pickering A unit with coincident failure to shutdown. This analysis showed that the power excursion was halted by channel and calandria vessel failures leading to moderator fluid displacement. The containment structure did not fail and, at worst, might suffer minor cracking at the top of the dome of the reactor building. Overall the dose consequences of such an accident were no worse than the limiting design basis dual failure event. In the intervening twenty years following this analysis, significant experimental information has been obtained that relates to power pulse behaviour. This information, together with conservatisms in the original analysis, are reviewed and assessed in this paper. In addition, the issue of reactivity initiated events in other reactor types is reviewed to identify the reactor design characteristics that are of importance in these events. Contrary to popular belief the existence of positive coolant void reactivity is not as significant a factor as it is sometimes stated to be. On balance, with appropriate design measures, no one reactor type can be claimed to be 'more safe' than another. The underlying basis for this statement is articulated in this paper. (author)

  1. Dynamic analysis of containment building of 500 MWe pressurised heavy water reactor using finite element method

    International Nuclear Information System (INIS)

    The reactor building essentially comprises of a double containment system. The two containments viz. inner and outer containments (ICW and OCW), are axi-symmetric structures along with the raft and therefore an axi-symmetric finite element analysis of these should suffice. However, the inner containment is connected with the reactor internals (i.e. internal structures and calandria vault) at EL-130.0 meter elevation and thereby will have an interaction with the reactor internals. Exact modelling of this interaction effect is a formidable task since the reactor internals are not axi-symmetric. Hence, an equivalent axi-symmetric model of the reactor internals was evolved in such a way that the dynamic characteristics of the interaction effect are preserved. Analysis of the containment building has been carried out for two mutually perpendicular horizontal directions (N-S) and E-W) and the vertical direction using response spectrum and time history technique. Due credence was given to soil-structure interaction. This report presents the results and conclusions arrived at for these analyses. (author). 18 refs., 81 figs., 68 tabs., 3 appendixes

  2. Development of high pressure conductivity probe (HPCP) for secondary shut down system (SDS-2) of 500 MWe PHWR

    International Nuclear Information System (INIS)

    The poison solution and the moderator in Secondary Shutdown System (SDS-2) of 500 MWe PHWR, are separated by their own liquid in liquid interface. This interface moves towards the calandria because of molecular diffusion, temperature difference and physical disturbances in the moderator level. It is proposed to install two numbers of high pressure conductivity probes (HPCP) to monitor the interface movement as well as to provide the safe annunciation value for interface location. On actuation of the SDS-2 signal, high-pressure helium will inject the poison into the moderator to shutdown the reactor. During poison injection, these probes will experience high pressure of nearly 85 kg/sq.cm. Global market survey indicated that conductivity probes having built in temperature sensor are available for a maximum pressure rating of 35 kg/sq.cm. Hence in order to meet the process requirement of SDS-2, the development of HPCP suitable for a pressure of 85 kg/sq.cm. was taken up. Two numbers of such probes were successfully designed, fabricated and evaluated for their performance. The developed conductivity probes fully meet the laid design and performance criteria. The aforesaid development work was a successful endeavour towards indigenisation of high-pressure conductivity probe for future applications. This report deals with the design aspects, fabrication technique, material and performance evajuation criteria and test results of HPCP. (author)

  3. Contributions to the research programs in nuclear and industrial electronics, domestic production of instrumentation, safety and control systems and equipment for nuclear reactors and auxiliary installations

    International Nuclear Information System (INIS)

    Domestic production of component system and equipment for the control and safety of nuclear facilities was one of the priority objective of the Nuclear Research Institute Pitesti. The problems addressed were particularly related to design and production of analog and digital equipment for measurements, triggering and display of the values of process parameters as well as to regulating complex functions of this equipment. Associated to this effort were the research works concerning: - reliability and in-service life-time of the electronic components and equipment in the safety and control systems for nuclear processes; - radiation endurance of industrial electronic components; utilization of whirling currents in calandria tube testing; - expert systems and applications in nuclear reactor control and safety; design and testing methods of process real time software packages for safety in control critical systems for nuclear domain. There are presented characteristics of the following equipment: 1. amplifier for ionization chambers with triggering comparator circuits for the CANDU 600 reactor shut down system; 2. amplifier for ionization chambers without triggering comparator circuits for power regulating system; 3. safety and regulating computerized system for C9 and C5 cans; 4. acquisition system for dosimetric data in nuclear facilities; 5. program able digital comparator for the reactor shut down system; 6. stationary gamma areal monitors for CANDU 600 reactors and other nuclear facilities

  4. Thermal-hydraulic analysis of the CANDU moderator tank using the CUPID code

    International Nuclear Information System (INIS)

    The component-scale thermal-hydraulic analysis code, CUPID, has been developed for steady-state and transient analyses of single- and two-phase flows in light water reactors. As an application to CANDU nuclear reactor, the single-phase natural circulation flow inside the moderator tank has been analyzed by assessing the experiments conducted at the 1/4-scaled test facility. A porous media approach was applied for the tube bundles to avoid computational complexity. This resulted in a good agreement with the experimental data. In this study, the analysis is extended to the prototype geometry of a CANDU reactor. A similar coarse mesh model was adopted for the Calandria tube bundle region. The flow at the inlet nozzles was complicated. However, for simplicity, each inlet nozzles was not modeled in detail and, instead, three different boundary conditions for the inlet nozzle flow were examined; A momentum-weighted flow distribution obtained from a fine-mesh CFD calculation was implemented to preserve both the mass and momentum flow rates at the boundary. This approach resulted in a more realistic temperature distribution in the tank. In conclusion, the CUPID code can cost-effectively predict the thermal-hydraulics in the moderator tank by using the porous media approach and an appropriate inlet flow modeling. (author)

  5. 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)

  6. 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

  7. 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.

  8. X-ray measurement of near surface residual stress in textured cold worked stress relieved zirconium alloy components for nuclear applications

    International Nuclear Information System (INIS)

    Zirconium alloys are commonly used in a number of nuclear reactor core components due to their various suitable physical, chemical and mechanical properties. Residual stress has great effect on the performance and lifetime of nuclear reactor core components. The x-ray residual-stress measurement using the standard multi-exposure technique can result in errors for textured material like Zircaloys, having highly anisotropic hexagonal close packed crystal structure. Diffraction peak intensity depends on factors like inclination, rotation of the beam and position of detectors. The low intensity due to texture results in the observed errors. Variation of elastic constants with direction in the case of textured materials also leads to errors. In the present study, firstly the non-linearity in the d vs. sin2Ψ plot has been explained and then a technique has been standardized to measure residual stress in textured materials with minimum % error. Firstly the determination of texture independent path with similar intensity in both the detectors is done by selection of inclination and rotation angles from pole figure. Next, directional x-ray elastic constants are introduced into the calculation using the single crystal elastic data and texture data. Following this standard procedure, the residual stress in critical PHWR components like, pressure tube, seamless and seam welded calandria tube and garter spring have been estimated. (author)

  9. Numerical Analysis of CANDU-6 Moderator System Using OpenFOAM

    International Nuclear Information System (INIS)

    On the moderator of CANDU-6 reactor, thanks to the rapid development of CFD (Computational Fluid Dynamics), the 1-D model code can be substituted to the 3-D simulation codes. The three-dimensional computation becomes not so expensive that now we can enjoy the benefit of innovation about CFD technology. In this study, we have modeled the Calandria tank system as simplified models preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors. The use of OpenFOAM is a very important point for the present study. The OpenFOAM is based on the object-oriented programming using C++ language. The solvers and libraries of physical properties, for example, are declared as classes to produce a new code with the reproduction from the existing classes. As this code is fully open to the public, the development of CFD code with OpenFOAM should be very prospective to the future design of system codes, not just restricted in the area of hydro-thermal system concerning atomic reactors

  10. Gadolinium depletion event in a CANDU® moderator - causes and recovery

    International Nuclear Information System (INIS)

    Gadolinium nitrate is added to the moderator of CANDU units to maintain the reactor in a guaranteed shutdown state (GSS). In April 2008, after being in stable GSS for over 30 hours, one of Ontario Power Generation's Pickering-B units showed a gradual depletion of the dissolved gadolinium, despite purification being isolated. Further additions of gadolinium stabilized the moderator gadolinium concentration, however, since the root cause of the depletion was not immediately identified, the unit was placed in the drained shutdown state, per established procedures. The cumulative gadolinium depletion amounted to about 3200 grams, the equivalent of about 12 ppm. Analysis showed the presence of oxalate in the moderator water. It is well-known that gadolinium forms a very insoluble oxalate (log Ksp = -29.1). Although sub-micron filtration of water samples did not show the presence of gadolinium particulate, the measured levels of oxalate, 1.2 to 2 ppm, were sufficient to react with 1.4 to 2.4 ppm of gadolinium. The source of oxalate was traced to radiolysis of dissolved CO2 species. This unit had been experiencing chronic low-level ingress of CO2 from the Annulus Gas System. Free oxalate ion is normally susceptible to radiolytic breakdown back to CO2, but Gd3+ provides a stable sink for radiogenic oxalate, 2 Gd3+ + 3 C2O42- → Gd2(C2O4)3. Subsequent testing confirmed that gadolinium oxalate is quite stable with respect to gamma irradiation. Inspections showed well-crystallized gadolinium oxalate deposited on moderator system surfaces. Estimates indicated that about 1200 grams of gadolinium could have deposited on in-core surfaces, including the outside of the calandria tubes. That amount of negative reactivity was a concern, since it would prevent re-start of the unit. OPG, with support from AECL-Chalk River and Kinectrics, embarked on a two-pronged chemistry recovery program aimed at 1) developing solvents for dissolution of the gadolinium oxalate deposits and 2

  11. Assessment of In-Core Damage for Feeder Stagnation Break in CANDU

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-05-15

    A feeder break is a single channel accident while the other channels remain intact in the CANDU core. For some ranges of feeder break size, a flow in the channel can become stagnate due to a force balance between the upstream and the downstream ends. In the extreme, this can lead to a rapid fuel heat up and fuel damage, and the failure of a fuel channel. This break scenario is called a feeder stagnation break. Following the feeder stagnation break, the fuel and pressure tube in the affected channel heat up quickly. The channel fails due to overheating and the channel contents begin to discharge into the moderator. The discharge is composed of steam, some hydrogen produced by possible metal-water reaction, and solid fuel elements of fuel fragments with molten material. The severity of the transient is primarily determined by the amount of molten material discharged into the moderator, and by the interaction between the molten material and the moderator, which determines the rate of energy release. After a channel rupture (pressure tube and calandria tube) some SOR (Shut-Off Rod) guide tubes, which are located in the vicinity of the break in the core, may be damaged. If the damage to the guide tube is substantial, some SORs may not be able to descend into the moderator, and therefore, not contribute to the shut down of the reactor. The increase in system reactivity, due to factors such as poison dilution from discharging coolant and void formation, may challenge the reactivity worth of the available undamaged SORs. Therefore, an analysis of the reactivity worth of the partially impaired SDS 1 (Shut-Down System 1) is required to determine that it can compensate for the increase in reactivity and shut down the reactor. In this study, the hydrodynamic transient, due to the dispersed molten material and the discharged steam, was calculated following the feeder stagnation break. The timing of the channel failure and the mass of the molten material were provided from the

  12. An innovative method for on-power radiometry of end-shields of nuclear power plants

    International Nuclear Information System (INIS)

    Every lndian PHWR reactor calandria is sandwiched within a pair of shield on either side. These shields are perpendicular to the coaxial axis of calandria and are called end-shields. These provide shielding from leakage radiation from reactor core in escaping out to Fuelling Machine vault, thereby significantly reducing the dose rates in the vaults. This has got a direct impact on radiation field in accessible areas. By maintaining low dose rates in accessible areas, the individual and collective doses of radiation workers can be effectively controlled well within the stipulated limits. Thus, it is of utmost importance to ensure adequacy of shielding provided by end-shields. In this context, a limited radiometry exercise is executed after filling of end-shields with steel balls and prior to their installation at designated place. This exercise provides limited inputs along the periphery of end-shield due to limited strength of radiation source, its handling provisions and dose constraints to the individual. In order to ascertain an in-depth analysis of shielding adequacy on-power, different methodologies have been adopted and have certain limitation in precisely pinpointing the affected area/location besides limitation on number of locations that can be monitored at a single stretch. To overcome these important anomalies, a computer based setup has been indigenously designed. The setup essentially comprises of a radiation monitor with wide energy, measuring, temperature and humidity range; a custom designed 25 m long compatible cable with suitable connectors; a laptop with additional cooling arrangement; a configurable interfacing software; thermal shielding for the detector and tying/fixing provisions. The radiation monitor after being properly shielded for thermal impacts is installed on the head of Fuelling Machine. It is connected through long cable to a laptop kept at Fuelling Machine service area with due cooling provisions (as temperature in the area will

  13. Effect of spray parameter on containment depressurization during LOCA in KAPP 3 and 4, 700 MWE IPHWR

    International Nuclear Information System (INIS)

    KAPP 3 and 4 is an Indian Pressurized Heavy Water Reactor (IPHWR) of 700 MWe capacities. It is a pressure tube type reactor with heavy water as moderator and coolant and natural Uranium Dioxide as fuel. It consists of 392 horizontal fuel channel assemblies and surrounded by three separate water systems i.e. primary coolant, moderator and calandria vault water system. Containment of Indian PHWR is an ultimate barrier, which is designed to envelope whole reactor systems, to prevent the spread of active air-borne fission products in accident condition. Containment Spray System has been provided for energy as well as activity removal from the Containment system. This paper discusses about the studies done to assess the effect of spray parameters such as spray flow rate, droplets diameter and height of fall on containment peak pressure and temperature, long term containment depressurization and energy removal from the containment during Loss of Coolant Accident (LOCA). The spray flow rate and droplets diameter play an important role in removing residual energy from containment atmosphere, which influences depressurization of containment. It is obvious that faster depressurization of containment during postulated LOCA helps in limiting radiological consequences. From radiological considerations, droplets diameter is required to be kept to the lowest practically possible value and flow rate of spray should be high. Spray water droplets fall height governs the exposure time of droplets, which is the direct indication of energy removal rate. However, it is observed from the sensitivity studies that for a height of spray droplet fall more than 16.5 m, for the range of spray water flow rate and droplets sizes considered in the analyses, there is no significant change in heat removal. (authors)

  14. 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

  15. Engineered safety in development of liquid poison injection system (shut down system-2) for 500 MWe PHWR

    International Nuclear Information System (INIS)

    Full text: The provision of shut down systems (SDS) is a mandatory requirement for safety of any nuclear reactor. The SDS shall be capable of making and holding the core adequately subcritical in the event of any anticipated operational occurrence and postulated accident conditions. The shut down function will perform as intended when its design and components are thoroughly evaluated for their reliability and effectiveness. A full scale mock up for one injection unit was designed and developed at Hall No.7, BARC. Experimental studies were carried out to qualify the design and evolve process parameters such as gas tank pressure, poison discharge rate and poison injection time. In liquid poison injection system i.e. shutdown system -2, there is no physical barrier, between the two liquids i.e. the poison and the moderator. A liquid in liquid interface, called poison moderator interface (PMI) separates these fluids. Extensive lab scale studies have been carried out on PMI movement study i.e. the interface movement due to molecular diffusion and due to process disturbances under simulated reactor condition. On the basis of lab scale results, a full-scale PMI setup has been designed and developed to generate plant data. From reactor safety consideration, the floating ball in poison tank is designed in such a way that it prevents the over pressurisation of calandria. For this purpose a non-intrusive ultrasonic ball detection system (U-BDS) has been developed. This paper covers the PMI system for 500 MWe PHWR with relevant safety aspects and describes in detail, the experimental results of PMI study. The engineered safety in design, methodology and qualification of U-BDS and its role intended in performance of SDS-2 have been also discussed in the paper

  16. Zirconium - an imported mineral commodity

    International Nuclear Information System (INIS)

    This report examines Canada's position in regard to the principal zirconium materials: zircon; fusion-cast zirconium-bearing refractory products; zirconium-bearing chemicals; and zirconium metal, master alloys, and alloys. None of these is produced in Canada except fused alumina-zirconia and certain magnesium-zirconium alloys and zirconium-bearing steels. Most of the 3 000-4 000 tonnes of the various forms of zircon believed to be consumed in Canada each year is for foundry applications. Other minerals, notably chromite, olivine and silica sand are also used for these purposes and, if necessary, could be substituted for zircon. Zirconium's key role in Canada is in CANDU nuclear power reactors, where zirconium alloys are essential in the cladding for fuel bundles and in capital equipment such as pressure tubes, calandria tubes and reactivity control mechanisms. If zirconium alloys were to become unavailable, the Canadian nuclear power industry would collapse. As a contingency measure, Ontario Hydro maintains at least nine months' stocks of nuclear fuel bundles. Canada's vulnerability to short-term disruptions to supplies of nuclear fuel is diminished further by the availability of more expensive electricity from non-nuclear sources and, given time, from mothballed thermal plants. Zirconium minerals are present in many countries, notably Australia, the Republic of South Africa and the United States. Australia is Canada's principal source of zircon imports; South Africa is its sole source of baddeleyite. At this time, there are no shortages of either material. Canada has untapped zirconium resources in the Athabasca Oil Sands (zircon) and at Strange Lake along the ill-defined border between Quebec and Newfoundland (gittinsite). Adequate metal and alloy production facilities exist in France, Japan and the United States. No action by the federal government in regard to zirconium supplies is called for at this time

  17. 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

  18. Alternative concept for a fast energy amplifier accelerator driven reactor

    International Nuclear Information System (INIS)

    Recently Rubbia et al. introduced a conceptual design of a Fast Energy Amplifier (EA) as an advanced innovative reactor which utilizes a neutron spallation source induced by protons as an external source in a subcritical array imbibed a molten lead coolant which, besides being breeder and waste burner, generates energy. This paper introduces some qualitative changes in Rubbia's concept such as more than one point of spallation, in order to reduce the requirement in the energy and current of the accelerator, and mainly to make a more flat neutron distribution. The subcritical core which in Rubbia's concept is an hexagonal array of pins immersed in a molten lead coolant is replaced by a concept of a solid lead calandria with the fuel elements in channels cooled by helium, allowing on line refueling or shuffling, and the utilization of a direct thermodynamic cycle (Brayton), which is more efficient than a vapor cycle. Although the calculations to demonstrate the feasibility of the EA alternative concept are underway and not yet finished, these ideas do not violate the basic physics of the EA, as showed in this paper, with evident advantages in the fuel cycle (on line refueling); reduced requirements in the accelerator complex, which is more realistic and economical in today accelerators technology; and finally the utilization of He as coolant compared with molten Pb is more close to the proved technology given the know how of gas cooled reactors and more efficient from the thermodynamic point of view, allowing simplification and the utilization in other process, besides electricity generation, as hydrogen generation. (author)

  19. Effect of spray parameter on containment depressurization during LOCA in KAPP 3 and 4, 700 MWE IPHWR

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, S. K.; Bhartia, D. K.; Mohan, N.; Malhotra, P. K.; Ghadge, S. G. [Directorate of Reactor Safety and Analysis, Nuclear Power Corporation of India Limited, C-3,Nabhikiya Urja Bhavan, Anushaktinagar,, Mumbai 400094 (India)

    2012-07-01

    KAPP 3 and 4 is an Indian Pressurized Heavy Water Reactor (IPHWR) of 700 MWe capacities. It is a pressure tube type reactor with heavy water as moderator and coolant and natural Uranium Dioxide as fuel. It consists of 392 horizontal fuel channel assemblies and surrounded by three separate water systems i.e. primary coolant, moderator and calandria vault water system. Containment of Indian PHWR is an ultimate barrier, which is designed to envelope whole reactor systems, to prevent the spread of active air-borne fission products in accident condition. Containment Spray System has been provided for energy as well as activity removal from the Containment system. This paper discusses about the studies done to assess the effect of spray parameters such as spray flow rate, droplets diameter and height of fall on containment peak pressure and temperature, long term containment depressurization and energy removal from the containment during Loss of Coolant Accident (LOCA). The spray flow rate and droplets diameter play an important role in removing residual energy from containment atmosphere, which influences depressurization of containment. It is obvious that faster depressurization of containment during postulated LOCA helps in limiting radiological consequences. From radiological considerations, droplets diameter is required to be kept to the lowest practically possible value and flow rate of spray should be high. Spray water droplets fall height governs the exposure time of droplets, which is the direct indication of energy removal rate. However, it is observed from the sensitivity studies that for a height of spray droplet fall more than 16.5 m, for the range of spray water flow rate and droplets sizes considered in the analyses, there is no significant change in heat removal. (authors)

  20. 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.

  1. Monitoring parameters for fuel channel safety in AHWR - an overview

    International Nuclear Information System (INIS)

    Full text: Advanced heavy water reactor (AHWR) is a vertical pressure tube type reactor having boiling light water coolant circulating by thermosyphon in reactor coolant channels. The fuel channel monitoring system instrumentation becomes very important for maintaining channel safety and integrity. The various monitoring parameters for fuel channel monitoring system can be classified as (a) continuous monitoring and (b) sample monitoring. The continuous monitoring parameters for fuel channel safety and operation are i) channel inlet pressure ii) channel temperature iii) channel flow and iv) channel power. Coolant activity and Annulus gas monitoring are sample-monitoring parameters for ensuring channel integrity. The inlet pressure to the core being same for all the channels and AHWR being a pressurized boiling water system, pressure and temperature measurement cannot be considered as monitoring parameter for channel safety. Hence the continuously monitoring parameter for fuel channel safety becomes flow and power monitoring only. Since channel DT is negligible, the void fraction measurement is required for getting the channel power. Hence in AHWR the measurement of flow at the single-phase inlet section and measurement of two-phase flow/void fraction at the tail pipe section becomes important for estimating channel flow and channel power for reactor safety and operation. Measurement of two-phase flow in natural circulation system is complex and various attempts were made to understand and develop instrumentation so that an effective channel monitoring can be carried out. The channel integrity is monitored by sampling methods namely annulus gas monitoring and activity monitoring. Since the annulus gap is partially open, the moisture/activity in the annulus air need to be measured accurately to identify and quantify the leak rate and in turn the pressure tube/calandria tube integrity. This paper describes in detail the various aspects of fuel channel monitoring

  2. Intelligent instrument for precise detection of spacer wires in radioactive coolant channel of Indian PHWR

    International Nuclear Information System (INIS)

    Pressure tube in Indian pressurized heavy water reactor is known to have sag due to dislocation of loose spacer wires (called garter spring spacers), irradiation creep and growth during operating life of the nuclear power plant. When the sag of the pressure tube is large enough, the pressure tube comes in contact with calandria tube, resulting in formation of hydride blisters over a period of time, which may eventually lead to failure of pressure tube. A system has been developed for precise detection of Garter spring spacers (SPGs), which play a vital role in relocation of Garter spring for extending the creep limited life of pressure tube. The system consists of eddy current based sensor, hardware conditioning module, PC-field Interface Unit, software conditioning module and ISA bus based add on cards to acquire the Garter spring data in conditioned form. The system also consists of precise displacement optical sensor and a quadrature pulse counter card for accurate position of the sensor head. The In-house developed automation software drive the tool with the help of motor driven linear mechanism for online recording of sensor signal along with the position information data along the length of pressure tube. All the input output signal between field and PC are optically and magnetically isolated to ohmic noise. The recorded data is passed through surge filters, digital filters, smoothing filter, interpolator, calibrators, slope finder to indicate the exact position of Garter spring spacers in automatic and semiautomatic mode. The analysis mode also takes care of data saving on storage disk, data security and audio-visual error messages while operating the system. All software and hardware modules are indigenously developed at RED/BARC and this system has been successfully used in numbers of Indian PHWRs. The same system with a minor modification in software and hardware can be used in inspection of tubular components in other industries also. (author)

  3. GOTHIC code evaluation of alternative passive containment cooling features

    International Nuclear Information System (INIS)

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

  4. 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)

  5. Design Details of the CGE Vertical Heavy-Water Pressure-Tube Reactor. Some Factors Influencing the Design

    International Nuclear Information System (INIS)

    The vertical HWR basic design criteria were to simplify existing designs to improve reliability and availability and to do this with existing proven technology. It was also a requirement to maintain access to operating areas at all times and to provide adequate shutdown shielding of the reactor to permit contact maintenance. The simplified fuel channel is characterized by; (a) Single-ended fuel changing on power using only one fuelling machine. (b) End-fitting closure which stores the seal. (c) Complete factory assembly and testing of the fuel channel before on-site installation. (This is permitted because of the use of only one end fitting.) (d) Fuel channel bolted to calandria permitting easy replacement of the assembly. This feature offers the possibility of retubing the reactor when better pressure-tube materials are developed (that is, the use of thinner pressure tubes to improve burnup which could be economically attractive if the price of uranium increases) . The simplified on-power refuelling is characterized by: (a) A simplified single fuelling machine using only mechanical drives and stops. (b) The capability of reshuffling the fuel into any desired order in the string by means of an accessible fuel handling mechanism. (c) The option of direct operator control of the fuelling machine. Improvements that have been made in reducing the building volume, dry vault concept and construction procedures by using the vertical design are described. In addition, the paper describes how the station can operate at base load to take advantage of low fuelling cost or how it may be used to control system frequency. The use of reactivity control mechanisms and steam by-pass to enable the power plant to operate with a continuously varying output over a wide range is described. (author)

  6. Experience in the application of the IAEA QA code and guides to the manufacture of nuclear reactor components

    International Nuclear Information System (INIS)

    India has made considerable progress in the indigenous manufacture of 'Quality' nuclear reactor components. All activities associated with the development of atomic energy from mining of strategic minerals to the design, construction, and operation of nuclear power plants including supporting research and development efforts are mainly carried out by the Department of Atomic Energy (DAE). Through the sustained efforts of DAE, the major industries, both in public and private sectors supplying nuclear components have now adopted the practice of systematic quality assurance (QA). The stringent QA steps are mandatory for achieving the desired quality in the manufactured nuclear components. Control blades for BWRs are now indigenously manufactured by the Atomic Fuels Division (AFD) of Bhabha Atomic Research Centre (BARC), a constituent unit of DAE. For the Project Dhruva, a 100 MW(th) nuclear reactor, constructed at BARC, Trombay, Bombay, an independent cell was formed to carry out quality audit on the manufactured components. The components were designed, fabricated, inspected and tested to the desired quality level. The QA activities were enforced from the procurement of raw materials to the audit of the completed component for monitoring the manufacturer's continued compliance with the design. The major components of Dhruva, viz. calandria, end-shield, coolant channels, heat exchangers, etc., were covered under these quality audit activities. The paper highlights the QA programme implemented in the manufacture of control blades for BWRs, illustrated with a typical example, the end-shield for Dhruva. The authors consider that the recommendations and guidelines provided in the documents 50-SG-QA3, 50-SG-QA8, 50-SG-QA10, etc., were useful in providing a formal and systematic framework, under which various quality assurance functions have been carried out

  7. The CIRENE program: experience gained in development of technology

    International Nuclear Information System (INIS)

    The construction of the Cirene 40 MW prototype electrical power plant at Latina is the main objective of the Cirene development program. A plant contract was given to NIRA at the end of 1976, and the plant completion is foreseen by the end of 1984. The Cirene is a heavy water-moderated, natural uranium-fueled pressure-tube reactor using boiling light water as the primary coolant. The design of Cirene is outlined. The numbers of pressure tubes, calandria tubes, liquid rod tubes and regulating rod tubes are 60, 60, 10 and 4, respectively, all made of Zircaloy-2. The equivalent diameter, length and lattice pitch of the core are 236 cm, 400 cm and 27 cm, respectively. The heavy water tank has the central zone of I.D. 369 cm and the dump annulus of I.D. 520 cm, made of AISI 304 L. These key parameters are shown. The means of regulating radioactivity during the normal operation of the reactor are: the two-phase rods, the level of the moderator and the concentration of boron in moderator. The function of two-phase rod regulation system is to introduce a variable density two-phase mixture consisting of boric acid solution and nonabsorbent gas in the core. The R and D was conducted to assess the functional and dynamic features of this control system. The plant is equipped with two independent fast-acting scram systems, discharging of moderator and liquid-rod system. The pressure-tube rupture test was carried out. (Nakai, Y.)

  8. 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)

  9. Effective utilization of maintenance staff in design and implementation of major project work

    International Nuclear Information System (INIS)

    The reorganization of Pickering Nuclear Division some 2 years ago resulted in the formation of the Projects and Modifications department. This department takes an integrated approach to manage all aspects of large projects at Pickering. The integration of Design, Drafting, Procurement, Construction and Operations functions into project teams represents a fundamental change to project management at Pickering. The development of integrated teams has great potential for reducing both the time and cost associated with project implementation, while at the same time improving the quality, and maintainability of the commissioned in service project. The Pickering Rehab organization 1989-1993, established to perform the rehab / retube of Units 3 and 4 had proven that a team environment will produce effective results. The outcome was astounding, critical categories such as Safety, Quality of Work, and Timeliness, had proven the team's effectiveness. The integration of operations maintenance staff into the project work activities is still evolving, and has probably required the most adaptation to change for both the former Construction and Operations organizations. Maximizing the utilization of the maintenance staff in the design and implementation of major project work will prove to be a key to a long term operating success of these projects. This paper will focus in on the effective usage of Maintenance staff in the design and implementation phases of major project work at Pickering, and on the benefits realized using this approach. It will be divided into 5 sections as indicated. 1. Past Project Shortfalls. 2. Benefits of the inclusion of Maintenance staff in the Calandria Vault Rehab Project. 3. Maintenance involvement in the Pickering 'A' Shutdown System Enhancement (SDSE) Project. 4. Challenges resulting from the inclusion of Maintenance staff project teams. 5. Summary. (author)

  10. Assessment of the integrity of KANUPP fuel channels

    International Nuclear Information System (INIS)

    The KANUPP reactor first produced power in 1972. In 1983 one of the fuel channels, G12, was difficult to refuel due to the South end fitting being retracted relative to the adjacent channels. In 1993 November, eight fuel channels were inspected non-destructively for defects. The dimensions and shape of the pressure tubes were measured, and channel G12 was removed for examination and testing at the Chalk River Laboratories. Examination of channel G12 showed that the South end calandria tube rolled joint had been leaking moderator heavy water into the annulus. The inboard bearing was completely corroded away, which resulted in the South end fitting seizing in the debris. The Inconel X750 garter springs were both completely broken into small rings, due to Stress Corrosion Cracking (SCC) in acids produced by radiolysis of the moderator heavy water. The inspection of the eight fuel channels did not reveal any serious defects. The length and diameter changes in the pressure tubes were both small. The deuterium ingress had been low, with the result that the hydrogen isotope concentration of the pressure tubes was below Terminal Solid Solubility (TSS) even close to the end fittings. Irradiation had increased the strength and reduced the ductility and fracture toughness of the tubes, and the Delayed Hydride Cracking (DHC) properties were similar to irradiated cold-worked Zr-2.5Nb pressure tubes. The examination and testing showed that the KANUPP pressure tubes are all probably in good condition and fit for several more years of service. (author). 3 tabs., 12 figs

  11. 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.

  12. Addition of oxygen in the inlet of recombiner unit in moderator cover gas system to facilitate recombination of deuterium and oxygen to bring deuterium concentration in safe limits

    International Nuclear Information System (INIS)

    In moderator system of PHWR, radiolytic decomposition of Heavy Water takes place in the Calandria and D2 and O2 are formed. Since the mixture of D2 and O2 is explosive in nature, there is a limit and various action levels for concentration of Oxygen and deuterium in moderator cover gas. The maximum percentage limit of deuterium is 4% v/v in presence of Oxygen present in stoichiometric ratio. In March -2013, the deuterium concentration in moderator cover gas of RAPS-2 increased to 3.0 % v/v and the oxygen concentration was only 0.93% v/v which was much less than the stoichiometric value (1.5% v/v). So it was decided to add oxygen in the inlet of recombiner unit of moderator cover gas system. As in year 1999 there was a fire incident in Darlington unit -3 due to combustion of EDPM which was used as seat material in the isolating valve in the inlet of oxygen pressure regulating valve. EDPM has low ignition temp. So Oxygen addition was not practiced in any other reactors. Oxygen addition was carried out in RAPS-2 with all precaution and with proper planning. Initially oxygen was added with very slow rate of 1 SCFH (Standard Cubic Feet per Hour) intermittently and the process was repeated to see any harmful effect. After qualifying the procedure, oxygen addition was done for 20 hrs at the rate of 2.5 SCFH and D2 concentration came down to 1.95 % v/v. This paper will consists of radiolytic decomposition of D2O, the qualification plan for O2 addition, the data of moderator cover gas and moderator system parameters before, during and after Oxygen addition. (author)

  13. Helios: resonance capture in zirconium

    International Nuclear Information System (INIS)

    Recently, natural Zr with resonance-shielding data has been added as a new nuclide to the nuclear-data library of the lattice code HELIOS. This has made it possible to study the importance of resonance shielding by Zr in lattice calculation. Usually, resonance shielding by Zr is neglected because it is a weak absorber. Many lattice codes cannot even treat resonance capture in the clad and even less so in the shroud a WWER, the channel of a BWR, the pressure and calandria tubes of a CANDU, or the pressure tube of an RBMK. This paper shows for some lattice types the effect of resonance shielding by Zr and presents a detailed analysis for SVEA and WWER-440 fuel. Because resonance shielding reduces neutron capture, it increases reactivity. In Zr it occurs mainly in the Zr-91 resonances at 293 and 681 e V, and the Zr-96 resonance at 301 eV, with peaks of 250 b, and 1050 b. Its effect increases when the spectrum hardens, e.g. for SVEA fuel the reactivity increase depends on the void and on whetter the control blades are inserted or not--for uncontrolled at 40% steam void it varies from 160 to 190 pcm between 0 and 50 M Wd/kg. For WWER fuel with ppm B it varies from 180 to 260 pcm. In CANDU fuel, which has a soft spectrum but contains much Zr, the reactivity increase is about 230 pcm. For RBMK fuel it is about 340 pcm, and for an old uncontrolled 8 x 8 BWR assembly it is about 200 and 380 pcm at 0 and 70% steam void (Authors)

  14. The application of neutron diffraction to materials science problems in the Canadian nuclear industry

    International Nuclear Information System (INIS)

    The main advantage of neutron diffraction over X-ray diffraction is that thermal neutrons easily pass through, for example, 25 mm of steel, so that measurements can be made at depth in engineering components. A program at Chalk River to investigate the industrial applications of neutron diffraction began with measurement on over-rolled Zr-2.5Nb pressure tubes, a topic of major concern in the eighties. It was quickly realized that neutrons could provide measurements of residual stress accurate enough to be of real interest. Over the ensuing period, major contributions have been made in measuring stresses and crystallographic texture in components for the nuclear industry including end-fittings, steam generator tubing, pressure tubes and calandria tubes, and weldments. In addition to work for the nuclear industry, there have been many applications in the aerospace, automotive, defence and pipeline industries in Canada and throughout the world. Residual stresses arise because of inhomogeneous plastic deformation of the material. Inhomogeneous plastic deformation not only occurs on a macroscopic scale but also on the scale of the grain size. The stresses that occur on this scale are called intergranular of type=II stresses. These intergranular effects, taken with the strong crystallographic alignment in zirconium alloy tubing, determine the growth of components in the reactor environment. Systematic studies of the origin of intergranular residual stresses arising from thermal effects and plasticity effects were carried out on Zircaloy-2 and Zr-2.5Nb alloys which have led to a theoretical understanding of component growth. Finally, a very recent texture scanning technique was able to shed light on the microstructure of zirconium alloy components. (author) 17 refs., 14 figs

  15. Stress-induced reorientation of hydride precipitates in Zr-2.5Nb-0.5Cu garter springs under complex loading

    International Nuclear Information System (INIS)

    Zr-2.5Nb-0.5Cu garter springs which are placed between coolant and calandria tubes in PHWRs experience complex loading due to simultaneous application of tension, compression and torus bending moment due to coolant tubes. The gradual pick up of hydrogen by the garter springs during service is likely to have hydride platelets reoriented under the applied stresses. In the present paper, the magnitudes and the directions of the principal stresses under the complex loading condition obtained have been calculated and the extent of hydride reorientation predicted. Simulation experiments consisting of simulated loading of hydrogen (upto 400 ppm) precharged springs at the service temperature (300degC) and also in-situ hydrogen charging of the springs under simulated loading conditions have been carried out. In addition, hydrogen precharged springs have been subjected to temperature cycling between 50 and 300degC under complex loading conditions, to evaluate the influence of temperature variation on hydride reorientation. Metallographic examination of the hydride platelets in the above springs has shown an excellent agreement with the analytical prediction. Torus bending moment values appear to play a significant role in reorienting the hydride platelets. It has been observed that under normal torus bending moment corresponding to 90 mm dia coolant tubes hydrogen platelets close to the outer rim of the spiral get reoriented in the radial direction. However, on application a torus bending moment corresponding to 30 mm dia tubes, hydride platelets get reoriented along the radial direction, irrespective of the magnitude of tensile and compression loading. (author). 9 refs., 15 figs., 1 appendix

  16. 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

  17. Development of transducers for integrated garter spring repositioning system

    International Nuclear Information System (INIS)

    In order to reposition the dislocated garter springs in active channels of 235 MW Pressurised Heavy Water Reactors (PHWRs), a tool named as Integrated Garter Spring Repositioning System (INGRES) has been developed. The tool consists of transducers to detect the concentricity between the Pressure Tube (P/T) and Calandria Tube (C/T) and also to detect garter springs in the channel besides different modules for correcting the eccentricity between P/T and C/T and garter spring repositioning. The transducers used in the system namely Concentricity Detection Probe (CDP) and Garter Spring Detection Probe (GSDP) are based on the eddy current techniques. The CDP makes use of four eddy current bobbin probes separated 90 degrees apart in cross sectional plane of channel assembly. The transducer gives output signal in proportional to the air gap between P/T and C/T in two axes (X and Y) which are designed for the purpose. The output of the unit is obtained on the Cathode Ray Oscilloscope (CRO) screen in the form of illuminated dot. The dot position on the CRO screen gives the information about mismatch in concentricity between P/T and C/T of the channel. The GSDP meant for detecting garter springs in PHWR channel uses two sets of primary and secondary coils connected in differential mode. The output signals from the transducers are processed through a signal processing unit devised for the purpose to obtain output from it as a horizontal beam on the CRO screen. The garter spring presence in the channel is indicated by a change in the voltage level of beam and also by audio-visual indication in the form of buzzer and LED illumination on the processing unit. This paper gives general design and development aspects of the CDP and GSDP transducers of the INGRES tool. (author). 3 figs

  18. Numerical modelling of the impedance plane for simultaneous determining, by eddy currents, the oxide thickness and conductivity in Zircaloy

    International Nuclear Information System (INIS)

    During service at high temperature or in aggressive media, metallic structures and components may suffer different types of changes or degradation. As an example, phase transformations may occur, second phases may precipitate, and in consequence the mechanical and chemical properties of the material may change. Their behavior will therefore differ from that considered for the design of the component. The knowledge of the amount of a precipitated second phase in a component should be an important tool in the hands of the maintenance engineer. And it would be very important to obtain this knowledge nondestructively and reliably. The objective of this project is to evaluate by eddy currents the amount of the hydrogen incorporated during service in structural zirconium base materials. For this purpose, a series of Zircaloy-4 specimens with oxide layers of different thickness and different concentration of hydrogen obtained by controlled autoclave treatments were used. These specimens were tested with eddy current equipment. The information produced by an eddy current test is the superposition of many variables, e.g.: thickness of oxide layers, conductance, thickness of specimen, etc. In order to sort out this information, an analytical model of the impedance plane was programmed in a PC, with which this information was processed, permitting, in this way, to evaluate the conductivity of materials, taking into account the effect of oxide layers thickness. A linear relationship between the conductivity and the hydrogen content in the range of hydrogen concentrations of technological interest was observed. Therefore, the calculated electrical conductivity may be transformed to the amount of hydrogen content, using a suitable calibration curve. This process will allow for the nondestructive assessment of the amount of hydrogen in reactor components, such as pressure and calandria tubes, a knowledge which will enable the experts to predict the degree of fragility of those

  19. 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)

  20. Full-scale seismic verification test for 600MWe CANDU fuelling machine and its correlation analysis

    International Nuclear Information System (INIS)

    One of the unique features of the CANDU reactor is its on power refuelling capability, which is facilitated by the two fuelling machines moving over the reactor (calandria) faces and engaging in the refuelling process. The F/M system has a number of geometric and mechanical non-linearities. To predict both the linear and non-linear dynamic response and to evaluate the seismic capability of the F/M system during an earthquake, it requires the development of a new non-linear seismic analysis code. This paper describes first the representative results of the seismic verification test for F/M. The test rig, which is composed of an actual F/M, carriage system, one fuel channel (F/C) and the support structure such as a bridge, is forced-vibrated by the hydraulic exciter installed at various portions of the test rig. As to both F/M system itself and F/M-F/C coupled system, sinusoidal swept test, steady state sinusoidal excitation test and free vibration test by blocking the input of exciter were carried out. Furthermore, Coulomb damping test for F/C was performed to grasp the sliding vibration characteristics of F/C in the axial direction which is thought to have an influence on the dynamic behavior of F/M coupled to F/C. Based on these test results the fundamental vibration characteristics such as resonance frequency, vibration mode shape, equivalent modal damping factor and so on, and the influence of gap, friction and non-linear spring between each component on the dynamic vibration characteristics of F/M are clarified. Also, the results of non-linear correlation analysis conducted for representative test results by using the new code are presented here

  1. Plant condition assessments as a requirement before major investment in life extension for a CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Full text: Since, to extend the life of a CANDU-6 reactor beyond its original design life requires the replacement of reactor components (380 pressure and calandria tubes), a major investment will have to be done. After a preliminary technical and economical feasibility study, Hydro- Quebec, owner of the Gentilly-2 NPP, has decided to perform a more detailed assessment to: 1. Get assurance that it is technically and economically viable to extend Gentilly-2 for another 20 years beyond the original design life; 2. Identify the detailed work to be done during the refurbishment period planned in 2008-2009; 3. Define the overall cost and the general schedule of the refurbishment phase; 4. Ensure an adequate licensing strategy to restart after refurbishment; 5. Complete all the Environmental Impact Studies required to obtain the government authorizations. The business case to support the refurbishment of Gentilly-2 has to take in consideration the reactor core components, which will be the major work to be completed during refurbishment. In summary the following main component will have to be changed or refreshed: The pressure and calandria tubes and the feeders (partial replacement only) (ageing mechanisms); The control computers (obsolescence); The condenser tubes (tubes plugging); The turbine control and electric-governor (obsolescence). An extensive campaign is under way to assess the 'health' of the station systems, structures and components (SSC). Two processes have been used for this assessment: Plant Life Management Studies (PLIM) for approximately 10 critical SSC or families of SSC (PLIM Studies); Condition Assessment Studies for other SSC with a lower impact on the Plant production or safety). The PLIM Studies are done on SSC's, which were judged critical because they are not replaceable (Reactor Building, Calandria), or that their failure could have a significant impact on safety or production (electrical motors, majors pumps, heat exchangers and pressure

  2. 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

  3. Specific aspects for Cernavoda - Unit 1 NPP life assurance

    International Nuclear Information System (INIS)

    Full text: The main scope of a Plant Life Management Program is to operate the NPP in a safe manner and at a competitive cost during the reactor life. To achieve this goal, it is important to continuously evaluate the degradation of the main structures and components of the NPP. Background -Cernavoda NPP design life is 30 years. Compared with this target, the operation history is not long (Unit 1 is in commercial operation since 1997). It is still important to begin a plant life management program early to identify the critical components and structures, to establish the data needed for their monitoring and to find methods to mitigate their degradation. A specific aspect for Cernavoda NPP - Unit 1 is the long delay between the fabrication of the main components and the start-up. Most components were procured 10-15 years before start-up. First criticality was achieved in 1996, but the containment perimeter wall sliding was complete in 1983, the Calandria vessel was installed in 1985, the Steam Generators were in position in 1987, the fuel channels were installed in 1989. In evaluating the history of these components, the preservation period must be observed. For Unit 2, which will be in service around 2005, the delay will be longer. For this reason, CNCAN (the Romanian Regulatory Authority) imposed, as a condition to resume the work, to evaluate the ageing of the existing components and structures in order to establish their acceptability for use in the plant. The results of this evaluation can be used as references for subsequent evaluations. Plant Life Assurance Programme - The first step of a PLIM programme is to identify the components and structures that are important for the plant life management. Critical components and structures selection is done using the following criteria: safety criteria - components and structures whose failure can cause a release of radioactivity or which have to mitigate the release of radioactivity in case of a failure of other

  4. Hydrogen distribution and management during postulated accident in KAPP - 3 and 4 700 MWe, IPHWR containment

    International Nuclear Information System (INIS)

    KAPP-3 and 4 Nuclear Power Project consist of 2 X 700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) units. These Units are pressure tube type reactor with heavy water as moderator and coolant and natural Uranium Dioxide as fuel. It consists of 392 horizontal fuel channel assemblies and surrounded by three separate water systems i.e. primary coolant, moderator and calandria vault water system. The whole reactor is enclosed in the Containment Building. Generation and accumulation of hydrogen in containment atmosphere during postulated accident scenario could pose a potential threat to the integrity of the containment as the hydrogen can form flammable or even explosive mixture with air in the containment. In case of reactor header break, the hydrogen generated due to metal water reaction is expected to be released to Break Compartment (Fuelling Machine Vault) and mix uniformly with air and steam in the vault. Subsequently, additional hydrogen is expected to be released to containment sump at a slower rate due to radiolysis of coolant, moderator and sump water water. This paper discusses the results of the studies carried out by NPCIL computer code 'PACSR' to estimate the hydrogen distribution in KAPP - 3 and 4 containment following its release due to metal water reaction, and radiolysis of the coolant, moderator and sump water during postulated accident condition, to ascertain if accumulation of hydrogen could pose a potential threat to the integrity of the containment. As total amount of hydrogen generation is not much, the global concentration of hydrogen would not reach at flammability limit of 4% even after 10 days of accident. The local concentration in break compartment (FMV) reaches to 3.5 % and does not cross the flammability limit during initial period of accident though hydrogen generation rate would be very high due to metal water reaction. Provisions of passive vertical opening between pump room and FM vaults exist in the design of KAPP - 3 and 4

  5. NUCLEARMONTAJ - an essential component of the Romanian nuclear power program

    International Nuclear Information System (INIS)

    NUCLEARMONTAJ represents a special and successful case of a company tightly related to the national nuclear power program. NUCLEARMONTAJ was involved in the most important works developed at Cernavoda Nuclear Power Plant such as Cernavoda NPP Unit 1, Intermediate Spent Fuel Storage project and Cernavoda NPP Unit 2 project. As main provider of building and and mounting services at national level in the nuclear field, NUCLEARMONTAJ has supported the most part of the expenses for the development of new technologies and formation of high skilled specialists. Even if the company's profile and structure evolved and extended following to market demands the attention paid to the nuclear energy division was constantly increased. Also, the company established special partnerships with major international companies in order to bring in Romania not only new technologies but elusive and necessary know-how elements, a special case being the joint venture associating NUCLEARMONTAJ with ANSALDO - Italy. At Cernavoda NPP, NUCLEARMONTAJ executed the installment work for both 700 MW turbine conventional equipment but also electric generators, condensers, pre-heater systems for high and low pressure as well as the works for heavy equipment like: calandria, feeders, main pumps, pressurizers, degassing systems, fuel handling machines for spent and fresh fuels and fuel storing equipment. In parallel, the company performed specific works in the frame of the other four units. The success in matching the company's profile to the market demands entailed signing contracts with organizations in Senegal, Nigeria, Brazil, Chile, Bahamas, Saudi Arabia, Indonesia, Jordan, Iran, United Arab Emirates, Kuwait, Qatar, Czech Republic, Germany, Italy and Russia. Lately, NUCLEARMONTAJ acts both as principal contractor and subcontractor with prestigious companies as, for instance, ABB, SNAM PROGETTI, F. FOCHI, FOSTER WHEELER ITALIANA or AECL. Recently, NUCLEARMONTAJ began works of constructions

  6. PC based transducer system for precise detection of garter spring spacers in coolant channels of Indian PHWR's

    International Nuclear Information System (INIS)

    In early generation of Indian PHWRs, loose fit Garter Spring (GS) spacers have been used to prevent Pressure Tube (PT) and Calandria Tube (CT) contact. Garter spring spacers have tendency to shift from their designed locations due to flow induced vibrations in pressure tube during hot conditioning. The shifting of GS spacers defeats the very purpose of serving as spacers to prevent PT -CT contact. Excessive span between the spacers lead to gap reduction between PT and CT due to pressure tube sag. The PTCT contact should be avoided to prevent formation of cold spot followed by hydride blisters on PT. An Integrated Garter spring Reposition System (INGRES) designed in RED require precise location and unambiguous detection of GS spacers for positioning of Garter Spring Repositioning Device (GSRD) for generating an optimum force on Garter Spring to relocate the Garter spring to new optimum position. Thus, a PC based transducer system, as a part of INGRES, has been developed for precise detection of garter spring spacers in coolant channels of PHWRs. The system consists of eddy current based sensor, Signal conditioning module, Software module and ISA bus based PC add on cards to acquire the Garter spring signal data in conditioned form. The system makes use of precise displacement optical sensor and quadrature pulse counter card for locating accurate position of the GS sensor head. The in-house developed automation software drives the transducer module in to coolant channel with the help of motor driven linear positioning mechanism for online recording of transducer signal along with the GS sensor head position information data along the length of pressure tube. All the input output signals between field and PC are optically and magnetically isolated to ohmic noise. The recorded data is passed through surge filters, digital filters, smoothing filter, interpolator, calibrators, slope finder to indicate the exact position of Garter spring spacers in automatic and

  7. Possible refurbishment of Point Lepreau

    International Nuclear Information System (INIS)

    In February 2000, the NB Power Board of Directors approved Phase one of a project to produce a business case including a detailed scope and estimate associated with the possible refurbishment of the Point Lepreau Generating Station (PLGS). The Preliminary plan for refurbishment projects an 18-month outage starting as early as the spring of 2006. If the station were to be refurbished, then it would be run for another 25 to 30 years. The decision on whether or not to refurbish PLGS has not been made and is not expected until the summer of 2002. The results of the first phase of the project will be used to prepare a detailed business case that will be presented to the NB Power board of directors in January of 2002. At that time a decision will be made as to whether to refurbish the unit, or obtain other means of replacing the energy produced by PLGS. The station currently produces about a third of the power generated within the province. If the business case is approved, all-380 Pressure Tubes and Calandria Tubes, along with their related End Fittings and Feeders would be replaced. This material would be stored in new storage vaults to be constructed at the existing on-site Waste Management Facility. Replacement of other station components will be performed as required, as determined from the results of a comprehensive Plant Condition Assessment. The condition assessments build on work done under the Plant Life Management Program. Point Lepreau Generating Station has operated well since start of commercial operation in early 1983. With a lifetime capacity factor of about 84% (up to the end of 2000), it has proven to be an economic and environmentally sound electricity provider. The station has also had a significant positive economic impact in Southern New Brunswick, employing over 600 people. However the Pressure Tubes and Feeders are nearing the point in time in which they will exceed their fitness for service criteria. Although tubes can be replaced on an

  8. 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

  9. Probabilistic evaluations for CANTUP computer code analysis improvement

    International Nuclear Information System (INIS)

    Structural analysis with finite element method is today an usual way to evaluate and predict the behavior of structural assemblies subject to hard conditions in order to ensure their safety and reliability during their operation. A CANDU 600 fuel channel is an example of an assembly working in hard conditions, in which, except the corrosive and thermal aggression, long time irradiation, with implicit consequences on material properties evolution, interferes. That leads inevitably to material time-dependent properties scattering, their dynamic evolution being subject to a great degree of uncertainness. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods in order to predict the structural component response. This work initiates the possibility to extend the deterministic thermomechanical evaluation on fuel channel components to probabilistic structural mechanics approach starting with deterministic analysis performed with CANTUP computer code which is a code developed to predict the long term mechanical behavior of the pressure tube - calandria tube assembly. To this purpose the structure of deterministic calculus CANTUP computer code has been reviewed. The code has been adapted from LAHEY 77 platform to Microsoft Developer Studio - Fortran Power Station platform. In order to perform probabilistic evaluations, it was added a part to the deterministic code which, using a subroutine from IMSL library from Microsoft Developer Studio - Fortran Power Station platform, generates pseudo-random values of a specified value. It was simulated a normal distribution around the deterministic value and 5% standard deviation for Young modulus material property in order to verify the statistical calculus of the creep behavior. The tube deflection and effective stresses were the properties subject to probabilistic evaluation. All the values of these properties obtained for all the values for

  10. Results from the fourth high-pressure melt ejection test completed in the molten fuel moderator interaction facility at Chalk River Laboratories

    International Nuclear Information System (INIS)

    The fourth high-pressure melt ejection test using prototypical corium was completed at Chalk River Laboratories. This test was one of four tests planned by Atomic Energy of Canada Limited to study the potential for energetic interaction between molten fuel and water. The experiments were designed to address one of the very low probability postulated accident events considered for Candu Pressurized Heavy Water Reactors (PHWRs). The accident event considered is the severe reduction in the coolant flow to a single channel. This reduction could result from a blockage in the flow or a break in the inlet piping to a fuel channel. If the reduction in the flow is severe (approaching complete cessation of the flow), the fuel channel will overheat and fail. Such a failure is not predicted to propagate to other fuel channels; the scenario is terminated with the emergency coolant injection. Under severely restricted flow blockage conditions, the temperature excursion could result in fuel melting. Conservative safety analysis assessments consider the implications of the worst-case scenario, which can involve the ejection of the molten material from the fuel channel into the heavy-water moderator. The predictions are that the melt will be finely fragmented and will transfer energy to the moderator as it is dispersed, creating a modest pressure pulse in the calandria vessel. The high-pressure melt ejection experiments funded by the Candu Owners Group have been performed to confirm these predictions and to show that a highly energetic 'steam explosion, ' and associated high-pressure pulse, is not possible. The high-pressure melt ejection test described here consisted of heating 12.5 kg of a thermite mixture U, UO2, Zr, and CrO3, representing the molten material in a fuel channel, inside an insulated pressure tube. When the molten material reached the desired temperature of ∼2400 deg.C, the pressure inside the tube was raised to about 10.5 MP a, and the pressure tube failed due

  11. Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR

    International Nuclear Information System (INIS)

    .0 MPa and 4 MPa. In order to simulate partially voided conditions inside PT, asymmetric heating has been carried out by injecting power to selected heater pins of the upper section of the 19 element fuel bundle simulator housed in a PT. This simulates nearly a stratification level of a half filled reactor channel. Through this technique an expected maximum circumferential temperature gradient of around 440 °C, has been attended from top to bottom periphery of PT. Tests also cover a power range of 8–11 kW which simulates different decay power levels. An asymmetric ballooning over eighty percent of PT length is observed for all the experiments and the deformation is mostly located to the upper part of the PT. The PT integrity is observed for lower internal pressure tests however a local failure has been observed for the test at 4.0 MPa. This is found to be due to excessive local strain prior to establishment of contact with Calandria Tube

  12. 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

  13. Clean energy for a new generation. Steam generator life cycle management and Bruce restart

    International Nuclear Information System (INIS)

    In the mid to late 1990s, Ontario Hydro decided to lay-up and write-down the Bruce A Nuclear Reactors. Upon transition to Bruce Power L.P., Canada's first and only private nuclear operator, new life and prospects were injected into the site, local economy and the provincial energy portfolio. The first step in this provincial power recovery initiative involved restart of Bruce Units 3 and 4 in the 2003/04 time-frame. Units 3 and 4 have performed beyond expectation during the last five-year operating interval. A combination of steam generator and fuel channel issues precluded a similar restart of Units 1 and 2. Enter the refurbishment of Bruce Units 1 and 2. This first-of-a-kind undertaking within the Canadian nuclear power industry is testament to the demonstrated industry leadership by Bruce Power L.P., their investors and the significant vendor community contribution that is supporting this major power infrastructure enhancement. Initiated as a 'turn-key' project solution separated from the operating units, this major refurbishment project has evolved to a fully managed in-house refurbishment project with the continued support from the broader vendor community. As part of this first-of-kind undertaking, Bruce Power L.P. is in the process of accomplishing such initiatives as a complete fuel channel re-tube (i.e. full core calandria and pressure tube replacement), replacement of all boilers (i.e. 16 in total) and the majority of feeder pipe replacement. Complimentary major upgrades and replacement of the remainder of plant equipment including both nuclear and non-nuclear valves, heat exchangers, electrical infrastructure, service water systems and components, all while meeting a parallel evolving/maturing regulatory environment related to achieving compliance with IAEA derived modern codes and standards. Returning to ground level, boiler replacement is a key part of the refurbishment undertaking and this further reflected a meeting of the 'old' and the 'new'. Pre

  14. Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, Ashwini K., E-mail: ashwinikumaryadav@gmail.com [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kumar, Ravi, E-mail: ravikfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Chatterjee, B., E-mail: barun@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gupta, Akhilesh, E-mail: akhilfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Mukhopadhyay, D., E-mail: dmukho@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2013-02-15

    .0 MPa and 4 MPa. In order to simulate partially voided conditions inside PT, asymmetric heating has been carried out by injecting power to selected heater pins of the upper section of the 19 element fuel bundle simulator housed in a PT. This simulates nearly a stratification level of a half filled reactor channel. Through this technique an expected maximum circumferential temperature gradient of around 440 °C, has been attended from top to bottom periphery of PT. Tests also cover a power range of 8–11 kW which simulates different decay power levels. An asymmetric ballooning over eighty percent of PT length is observed for all the experiments and the deformation is mostly located to the upper part of the PT. The PT integrity is observed for lower internal pressure tests however a local failure has been observed for the test at 4.0 MPa. This is found to be due to excessive local strain prior to establishment of contact with Calandria Tube.

  15. Fundamentos ecológicos y biogeográficos de la rareza de la avifauna madrileña: Una propuesta de modificación del catálogo regional de especies amenazadas

    Directory of Open Access Journals (Sweden)

    Carrascal, L. M.

    2006-05-01

    ático – ‘Sensible a la Alteración de su Hábitat’; Calandria – descatalogarla; Mirlo Acuático - ‘Sensible a la Alteración de su Hábitat’; Colirrojo Real - ‘Sensible a la Alteración de su Hábitat’ o ‘Vulnerable’; Papamoscas Gris - ‘De Interés Especial’; Alcaudón Real Meridional – descatalogarla; Picogordo – ‘De Interés Especial’.