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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. 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

  3. 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)

  4. 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)

  5. 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)

  6. 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

  7. 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

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

    International Nuclear Information System (INIS)

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

  9. 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.

  10. 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

  11. 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

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

    International Nuclear Information System (INIS)

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

  13. 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)

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. 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)

  12. 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)

  13. 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)

  14. 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)

  15. 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

  16. The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes

    Directory of Open Access Journals (Sweden)

    Hyoung Tae Kim

    2016-01-01

    Full Text Available The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor, has been modeled in multidimension for the computation based on CFD (computational fluid dynamics technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchmark problem of STERN laboratory experiment with a precise modeling of tubes, compared with each other as well as the measured data and a porous model based on the experimental correlation of pressure drop. Also the effect of turbulence model is discussed for these low Reynolds number flows. As a result, they are shown to be successful for the analysis of three-dimensional numerical models related to the calandria system of CANDU reactors.

  17. 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)

  18. 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.

  19. The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes

    OpenAIRE

    Kim, Hyoung Tae; Chang, Se-Myong; Shin, Jong-Hyeon; Kim, Yong Gwon

    2016-01-01

    The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor), has been modeled in multidimension for the computation based on CFD (computational fluid dynamics) technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchm...

  20. 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

  1. 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...

  2. 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)

  3. 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)

  4. 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

  5. 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.

  6. 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

  7. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  9. Measuring velocity profile in scaled CANDU6 moderator tank using particle image velocimetry

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Han; Bang, In Cheol [UNIST, Ulsan, (Korea, Republic of); Kim, Hyoung Tae; Cha, Jae Eun [KAISt, Daejeon (Korea, Republic of)

    2012-10-15

    The Calandria vessel, which is called as CANDU6 moderator tank, is the actual reactor core including fuel channels and moderator in PHWR. The understanding of circulation patterns in Calandria vessel is important because the cooling capability is related to the moderator tank. Therefore, measuring velocity and temperature patterns in the Calandria vessel is the key factor for the safety of CANDU reactor. Various experimental and numerical efforts to predict and analyze the thermal hydraulic characteristic of Calandria vessel have been made. Yoon et al. developed a CFD model for the CANDU6 moderator analyzing the velocity profile and the temperature distribution. Khartabil et al. did an experiment to measure 3 dimensional velocity and temperature distribution in moderator circulation tests at a 1/4 scaled down facility. Laser Doppler Anememetry (LDA) was used to detect the velocity profile and thermocouples detect the temperature distribution. Previous study presented the velocity profile in a 1/40 scaled-down Calandria vessel and compared with CFD analysis. In the present work, Particle Image Velocimetry (PIV) technique is used to obtain velocity profiles in a 1/8 scaled down Calandria vessel.

  10. OpenFOAM Analysis of CANDU-6 Moderator Flow

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chang, Se-Myong [Kunsan National University, Gunsan (Korea, Republic of)

    2015-10-15

    In this study OpenFOAM (Open Field Operation and Manipulation), an open source CFD solver, is used to simulate the three-dimensional moderator flow in calandria tank of CANDU-6 reactor improving the computational efficiency by parallel computing which does not need any proprietary license. A prototype of CANDU-6 reactor is numerically analyzed about three-dimensional moderator flow in calandrian tank with OpenFOAM, an open source CFD code. The horizontal fuel channels in a CANDU-6 reactor (a pressurized heavy water reactor) are submerged in the heavy water (D{sub 2}O) pool which is contained by a cylindrical tank, calandria. Each fuel channel consists of concentric tubes: a Pressure Tube (PT) and a Calandria Tube (CT). And the CO{sub 2} gas is filled between these tubes. Consequently, a heat flux is rapidly transferred to the outer CT so that a film boiling may occur in CT. As a result, it is important to keep the subcooling in the moderator. It is one of the major concerns in the CANDU safety analyses to estimate the local subcooling margin of the moderator inside the calandria tank. Previous experimental studies showed that the film boiling would be unlikely to occur if the local moderator subcooling is sufficient. Therefore, an accurate prediction of the moderator temperature distribution in the calandria tank is needed to confirm the channel integrity. There have been numerous computational efforts to estimate the thermal hydraulics in the calandria tank using CFD codes. Hadaller et al. obtained a tube bank pressure drop model for tube bundle region of the calandria tank and implemented it into the MODTURC{sub C}LAS code. Yoon et al. used the CFX code to develop a CFD model with a porous media approach for the core region. However, it is known that porous media modeling provide only average values of flow velocities and temperatures and do not give any information about local flow variables near tube solid walls, which are necessary to implement accurate heat

  11. 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.

  12. Thermal-hydraulic analysis of a heavy-water reactor moderator tank using the CUPID Code

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Su Ryong; Jeong, Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Kim, Hyoung Tae; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, a preliminary analysis is performed for the CANDU moderator tank. The calculation results using the basic case input showed a unrealistic, thermal stratification in the upper region, which was caused by the lack of the momentum of the cooling water from the inlet nozzle. To increase the flow momentum from the inlet nozzle, the cross-section area of each inlet nozzle was reduced by half and, then, the calculation showed very realistic results. It is clear that the modeling of the inlet nozzle affects the calculation result significantly. Further studies are needed for a realistic and efficient simulation of the flow in the Calandria tank. When the core cooling system fails to remove the decay heat from the fuel channels during a loss of coolant accident (LOCA), the pressure tube (PT) could strain to contact its surrounding Calandria tube (CT), which leads to sustained CTs dry out, finally resulting in damages to nuclear fuel. This situation can occur when the degree of the subcooling of the moderator inside the Calandria vessel is insufficient. 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 outlet of the vessel only. Therefore, we need to accurately predict the local temperature inside the Calandria vessel.In this study, the thermal-hydraulic analysis of the real-scale heavy-water reactor moderator is carried out using the CUPID code. The applicability of the CUPID code to the analysis of the flow in the Calandria vessel has been assessed in the previous studies.

  13. Studies on flow induced vibration of reactivity devices of 700 MWe Indian PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Prabhakaran, K.M., E-mail: kmprabha@yahoo.com [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Goyal, P.; Dutta, Anu; Bhasin, V.; Vaze, K.K.; Ghosh, A.K. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Pillai, Ajith V.; Mathew, Jimmy [Nuclear Power Corporation of India Ltd., Mumbai 400 094 (India)

    2012-03-15

    Highlights: Black-Right-Pointing-Pointer FIV studies on internals of heavy water filled calandria of 700 MWe Indian PHWR is presented. Black-Right-Pointing-Pointer This includes CFD and structural dynamic analysis to predict the dynamic behavior of component lying inside calandria. Black-Right-Pointing-Pointer Results of these calculations as well as conclusions from this investigation are presented. Black-Right-Pointing-Pointer It is established that FIV is not a concern in the present design of calandria internals. - Abstract: Component failures due to excessive flow-induced vibration are still affecting the performance and reliability of nuclear power stations. Tube failures due to fretting-wear in nuclear steam generators, and vibration related damage of reactor internals are of particular concern. In the Indian nuclear industry, flow induced vibrations are assessed early in the design process and the results are incorporated in the design procedures. In this paper the details of flow induced vibration studies on internals like liquid zone control unit and poison injection units of heavy water filled calandria of 700 MWe Indian pressurized heavy water reactor is given. This includes computational fluid dynamics studies from which the velocities are extracted for the components lying inside the calandria. With these velocities as input, further studies are performed to predict the dynamic behavior of these components. Results of these calculations as well as conclusions derived from this investigation are presented. Based on the studies it has been established that flow induced vibration is not a concern in the present design of 700 MWe calandria internals.

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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)

  1. 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)

  2. 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.)

  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. 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.

  5. Review of the Safety Concern Related to CANDU Moderator Temperature Distribution and Status of KAERI Moderator Circulation Test (MCT) Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, Bo W.; Kim, Hyoung T. [Severe Accident and PHWR Safety Research Division, Daejeon (Korea, Republic of); Kim, Tongbeum [University of the Witwatersrand, Johannesburg (South Africa); Im, Sunghyuk [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Following a large break LOCA and before Emergency Coolant Injection (ECI) initiation, pressure tubes (PT) significantly heat up as a result of the initial power pulse and degraded coolant flow. Consequently, some pressure tubes balloon and come into contact with the calandria tubes (CT). Following the PT/CT contact, the pressure tubes cool as they transfer some of the absorbed heat to the moderator via conduction at contact locations. As long as sustained calandria tube dryout does not occur, the calandria tube surface temperature remains below the creep threshold temperature and no further deformation is expected. Consequently, a sufficient condition to ensure fuel channel integrity following a large LOCA, is the avoidance of sustained calandria tubes dryout. 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 fuel channel contact experiments. The difference between available subcooling and required subcooling is called subcooling margins. The moderator flow circulation patterns are complicated slow flows that significantly vary from buoyancy dominated to inertia dominated patterns. Accurate predictions of flow patterns are essential for accurate calculation of moderator temperature distributions and the related moderator subcooling. Following a large break LOCA and before Emergency Coolant Injection (ECI) initiation, pressure tubes (PT) significantly heat up as a result of the initial power pulse and degraded coolant flow. Consequently, some pressure tubes balloon and come into contact with the calandria tubes (CT). Following the PT/CT contact, the pressure tubes cool as they transfer some of the absorbed heat to the moderator via conduction at contact locations. As long as sustained calandria tube dryout does not occur, the calandria tube surface temperature remains below the creep

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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.

  12. 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

  13. 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

  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. 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)

  16. 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)

  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. 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

  19. 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)

  20. 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)

  1. 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

  2. Preliminary study of CANDU moderator thermal hydraulics using the CUPID code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gi; Jeong Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Lee, Jae Ryong; Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    When the moderator cooling system fails, moderator may act as to remove decay heat which occurs in fuel. During loss of coolant accident (LOCA), the film boiling occurs in the Calandria tube (CT) because the hot pressure tube would deform into contacting with the calandria tube. And lower subcooling would decrease the margin of the CT to dryout. So, it is important to estimate a local subcooling of the moderator inside the Calandria vessel. However, in order to predict the internal temperature the study of empirical experiments and calculations are needed because only the inlet/outlet temperature can be measured in real reactor. In this study, the internal flow of the moderator was predicted by using the CUPID code, which has been developed in KAERI. The CUPID adopts three dimensional, transient, two phase and three field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two fluid model. The CUPID code shows single phase and two phase flow through two phase flow calculations of virtual can be applied.

  3. 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)

  4. 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.

  5. 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)

  6. 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.

  7. 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

  8. 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

  9. Fuel elements assembling for the DON project exponential experience; Montaje de los elementos combustibles para la experiencia exponencial del proyecto DON

    Energy Technology Data Exchange (ETDEWEB)

    Anca Abati, R. de

    1966-07-01

    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.

  10. 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.

  11. 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)

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. 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

  17. Moderator Circulation Simulation for 35% Reactor Inlet Header Break in the Wolsong Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    The objective of this study is present the results of moderator circulation simulations by the CFD code MODTURC{sub C}LAS V2.9-IST for the refurbished Wolsong unit 1. The present simulations were performed for a loss of Class IV power during a Large Break Loss Of Coolant Accident (35% inlet header break) without ECC injection and steam generator crash cool-down (LOCA/LOECC/LOCC). The analysis was performed to facilitate the assessment of fuel channel integrity following pressure tube (PT) and calandria tube (CT) contact by estimating the subcooling available for the inlet header break scenario

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

    2000-04-01

    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.

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  1. 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)

  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. 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

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

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

    Full Text Available The scope of this paper is to achieve the device functioning steps for the commissioning of the horizontal fuel channels of calandria vessel. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. For the decommissioning operation design shall be taken to ensure all aspects of security, environmental protection during decommissioning operation steps and creating and implementing work procedures resulting from developed decommissioning plan. The fuel channel decommissioning device is designed for dismantling and extraction of the fuel channel and its components. The decommissioning operation consists of following major steps: platform with device positioning to the fuel channel to be dismantled; coupling and locking the device at the fuel channel; unblock, extract and store the channel closure plug; unblock, extract and store the channel shield plug; block and cut the middle and the end of the pressure tube; block, extract and store the end fitting; block, extract and store the half of pressure tube; mounting of the extended closing plug. The operations steps are performed by the Cutting and Extraction Device and by the extraction actuator from the device handling elements assembly. After each step of dismantling is necessary the confirmation its finalization in order to perform the next operation step. The dismantling operation steps of the fuel channel components are repeated for all the 380 channels of the reactor, from the front of calandria side (plane R as well as the rear side (plane R'.

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  7. 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

  8. 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

  9. 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)

  10. 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)

  11. 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)

  12. 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

  13. 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)

  14. Remote field eddy current testing

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Y. M.; Jung, H. K.; Huh, H.; Lee, Y. S.; Shim, C. M

    2001-03-01

    The state-of-art technology of the remote field eddy current, which is actively developed as an electromagnetic non-destructive testing tool for ferromagnetic tubes, is described. The historical background and recent R and D activities of remote-field eddy current technology are explained including the theoretical development of remote field eddy current, such as analytical and numerical approach, and the results of finite element analysis. The influencing factors for actual applications, such as the effect of frequency, magnetic permeability, receiving sensitivity, and difficulties of detection and classification of defects are also described. Finally, two examples of actual application, 1) the gap measurement between pressure tubes and calandria tube in CANDU reactor and, 2) the detection of defects in the ferromagnetic heat exchanger tubes, are described. The future research efforts are also included.

  15. 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)

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-04-15

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

  6. 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

  7. 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.)

  8. 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

  9. 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)

  10. 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)

  11. 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

  12. 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

  13. 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.)

  14. 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

  15. 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

  16. Utilization of noise analysis technique for mechanical vibrations estimation in the ATUCHA{sub 1} and Embalse Argentine NPP; Uso de la tecnica de analisis de ruido para la estimacion de vibraciones mecanicas en las centrales nucleares argentinas Atucha I y Embalse

    Energy Technology Data Exchange (ETDEWEB)

    Lescano, V.H.; Wentzeis, L.M. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes; Guevara, M.; Moreno, C. [Nucleoelectrica Argentina S.A., Cordoba (Argentina). Central Nuclear Embalse; Pineyro, J. [Nucleoelectrica Argentina S.A., Buenos Aires (Argentina). Central Nuclear Atucha I

    1996-07-01

    In Argentine, comprehensive noise measurements have been performed with the reactor instrumentation of the PHWR power plant Atucha I and Embalse. The Embalse reactor is a CANDU-600 (600 Mwe) type pressurized heavy water reactor. It's a heavy water moderator and heavy water cooled natural uranium fueled pressure tube system. Signal of vanadium and platinum type in core-self power neutron detectors of ex-core ion chambers and of a moderator pressure sensor have been recorded and analysed. The vibration of reactor internals as vertical and horizontal in-core neutron flux detectors units and the coolant channels systems, consisting of calandria and pressure tubes with fuel bundles, have been identified and monitored during normal reactor operation. Atucha I, is a PHWR reactor natural uranium fueled, and heavy water moderated and cooled. Neutron noise techniques using of ex-core ionization chambers and in-core Vanadium SPND's were implemented, among others, in order to produce early detection of anomalous vibrations in the reactor internals. Noise analysis was successfully performed to identify normal and peculiar vibrations in particular reactor internals. (author)

  17. 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

  18. 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

  19. 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.

  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. 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

  2. CANDU safety analysis system establishment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo; Rhee, B. W.; Park, J. H.; Kim, H. T.; Choi, H. B.; Shim, J. I.; Yoon, C.; Yang, M. K

    2002-03-01

    To develop CANDU safety analysis system, methodology, and assessment technology, GAIs from CNSC and GSIs drived by IAEA are summarized. Furthermore, the following safety items are investigated in the present study. - It is intended to secure credibility of the void reactivity in the stage of nuclear design and analysis. The measurement data concerned with the void reactivity were reviewed and used to assess the physics code such as POWDERPUFS-V/RFSP, and the lattice code such as WIMS-AECL and MCNP-4B. - Reviewing the Final Safety Analysis Report for Wolsong-2/3/4 Units, the safety analysis methodology, classification for accident scenarios, safety analysis codes, their interface, etc. were examined. - The development of 3D CFD transient analysis model has been performed to predict local subcooling of the moderator in the vicinity of Calandria tubes in a CANDU-6 reactor in the case of Large LOCA transient. - The trip coverage analysis methodology based on CATHENA code is developed. The simulation of real plant transient showed good agreement. The trip coverage map was generated successfully for two typical depressurization and pressurization event. - The multi-dimensional analysis methodology for hydrogen distribution and hydrogen burning phenomena in PHWR containment is developed using GOTHIC code. The multi-dimensional analysis predicts the local hydrogen behaviour compared to the lumped parameter model.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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)

  8. Bearing pad to pressure tube contact simulation

    Energy Technology Data Exchange (ETDEWEB)

    Talebi, F.; Behdadi, A.; Luxat, J.C., E-mail: farshat@mcmaster.ca, E-mail: behdada@mcmaster.ca, E-mail: luxatj@mcmaster.ca [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)

    2011-07-01

    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

  9. 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

  10. Propiedades estructurales de ejes huecos y sólidos con una grieta plana//Structural proprieties of hollow and whole axles, with a flat crack

    Directory of Open Access Journals (Sweden)

    Jon‐Ariza De‐Miguel

    2014-05-01

    Full Text Available En este trabajo se estudiaron las diversas ventajas constructivas que los ejes huecos presentan respecto a los sólidos, como la reducción de masa, mejor control de calidad, menores defectos en producción y facilidad de acoplamiento. De ahí su profusión en aplicaciones industriales, como ejes de ferrocarril, de reductores o en calandrias. Dichas ventajas se ven alteradas cuando el eje trabaja a fatiga en flexión rotativa y aparece una grieta sobre el mismo. Se comprobaron dichas ventajas y sus variaciones sobre un modelo inicial de una grieta plana superficial, y se ha observado que los ejes huecos resultan más perjudicados por estas grietas para ciertos valores del diámetro interior, mientras que para otros son ventajosos. Se concluye del presente estudio que puede establecerse un rango de valores para el diámetro interior de ejes huecos en el cual su aprovechamiento seríaóptimo.Palabras claves: ejes huecos y sólidos, propiedades estructurales, grieta plana.______________________________________________________________________________AbstractThis work studies the several advantages hollow axles show when compared to whole ones, such as lightness, better quality control, less manufacturing defects, coupling simplicity. Thence their being profusely employed in industry, in applications like railway axles, gearboxes or calenders. Theseadvantages are altered when the axle works under bending with rotation under fatigue and a crack appears on its surface. An analytical research upon a flat surface crack model has been made of these advantages and their variations and it has been observed that hollow axles suffer more from these cracks than whole ones for certain inner diameter values, whereas for others even less than whole ones. It can be deduced that a gap for the inner diameter can be ascertained, where the axle´s rendering would be optimal.Key words: hollowand whole axles, structural properties, flat crack.

  11. 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)

  12. 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

  13. 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

  14. 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

  15. Radiological Characteristics of decommissioning waste from a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  16. 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

  17. 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

  18. 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

  19. 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.

  20. 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’.

  1. 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