Performance of pressure tubes in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

1989-01-01

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

Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

International Nuclear Information System (INIS)

Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

1978-12-01

This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

CANDU pressure tube leak detection by annulus gas dew point measurement. A critical review

International Nuclear Information System (INIS)

Greening, F.R.

2017-01-01

In the event of a pressure tube leak from a small through-wall crack during CANDU reactor operations, there is a regulatory requirement - referred to as Leak Before Break (LBB) - for the licensee to demonstrate that there will be sufficient time for the leak to be detected and the reactor shut down before the crack grows to the critical size for fast-uncontrolled rupture. In all currently operating CANDU reactors, worldwide, this LBB requirement is met via continuous dew point measurements of the CO_2 gas circulating in the reactor's Annulus Gas System (AGS). In this paper the historical development and current status of this leak detection capability is reviewed and the use of moisture injection tests as a verification procedure is critiqued. It is concluded that these tests do not represent AGS conditions that are to be expected in the event of a real pressure tube leak.

CANDU pressure tube leak detection by annulus gas dew point measurement. A critical review

Energy Technology Data Exchange (ETDEWEB)

Greening, F.R. [CTS-NA, Tiverton, ON (Canada)

2017-03-15

In the event of a pressure tube leak from a small through-wall crack during CANDU reactor operations, there is a regulatory requirement - referred to as Leak Before Break (LBB) - for the licensee to demonstrate that there will be sufficient time for the leak to be detected and the reactor shut down before the crack grows to the critical size for fast-uncontrolled rupture. In all currently operating CANDU reactors, worldwide, this LBB requirement is met via continuous dew point measurements of the CO{sub 2} gas circulating in the reactor's Annulus Gas System (AGS). In this paper the historical development and current status of this leak detection capability is reviewed and the use of moisture injection tests as a verification procedure is critiqued. It is concluded that these tests do not represent AGS conditions that are to be expected in the event of a real pressure tube leak.

Delayed hydride cracking and elastic properties of Excel, a candidate CANDU-SCWR pressure tube material

International Nuclear Information System (INIS)

Pan, Z.L.

2010-01-01

Excel, a Zr alloy which contains 3.5%Sn, 0.8%Nb and 0.8%Mo, shows high strength, good corrosion resistance, excellent creep-resistance and dimension stability and thus is selected as a candidate pressure tube material for CANDU-SCWR. In the present work, the delayed hydride cracking properties (K IH and the DHC growth rates), the hydrogen solubility and elastic modulus were measured in the irradiated and unirradiated Excel pressure tube material. (author)

Evaluation of a leaking crack in an irradiated CANDU pressure tube

International Nuclear Information System (INIS)

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

1988-06-01

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

Fatigue crack initiation at complex flaws in hydrided Zr-2.5%Nb samples from CANDU pressure tubes

International Nuclear Information System (INIS)

Stoica, L.; Radu, V.

2016-01-01

The paper addresses the phenomena which occur at locations where the oxide layer of the inner surface of CANDU tube pressure is damaged by the contact with the fuel element or due to the action of hard particles at the interface between the tube pressure and bearing pad of fuel element. In such situations generate defects, which most often are defects known as ''bearing pad fretting flaws'' or ''debris fretting flaws''. In this paper the experiments are completed in a series of previous works on the mechanical fatigue phenomenon on samples prepared from the pressure tube Zr-2.5% Nb alloy. The phenomenon of variable mechanical stress (or fatigue) may lead to initiation of cracks at the tip of volumetric flaws, according to the accumulation of hydrides, which then fractures and can propagate through the tube wall pressure due to the mechanism of type DHC (Delayed Hydride Cracking). (authors)

Database and prediction model for CANDU pressure tube diameter

Energy Technology Data Exchange (ETDEWEB)

Jung, J.Y.; Park, J.H. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

2014-07-01

The pressure tube (PT) diameter is basic data in evaluating the CCP (critical channel power) of a CANDU reactor. Since the CCP affects the operational margin directly, an accurate prediction of the PT diameter is important to assess the operational margin. However, the PT diameter increases by creep owing to the effects of irradiation by neutron flux, stress, and reactor operating temperatures during the plant service period. Thus, it has been necessary to collect the measured data of the PT diameter and establish a database (DB) and develop a prediction model of PT diameter. Accordingly, in this study, a DB for the measured PT diameter data was established and a neural network (NN) based diameter prediction model was developed. The established DB included not only the measured diameter data but also operating conditions such as the temperature, pressure, flux, and effective full power date. The currently developed NN based diameter prediction model considers only extrinsic variables such as the operating conditions, and will be enhanced to consider the effect of intrinsic variables such as the micro-structure of the PT material. (author)

A Study on the Small Punch Test for Fracture Strength Evaluation of CANDU Pressure Tube Embrittled by Hydrogen

International Nuclear Information System (INIS)

Nho, Seung Hwan; Ong, Jang Woo; Yu, Hyo Sun; Chung, Se Hi

1996-01-01

The purpose of this study is to investigate the usefulness of small punch(SP) test using miniaturized specimens as a method for fracture strength evaluation of CANDU pressure tube embrittled by hydrogen. According to the test results, the fracture strength evaluation as a function of hydrogen concentration at -196 .deg. C was much better than that at room temperature, as the difference of SP fracture energy(Esp) with hydrogen concentration was more significant at -196 .deg. C than at room temperature for the hydrogen concentration up to 300ppm-H. It was also observed that the peak of average AE energy, the cumulative average AE energy and the cumulative average AE energy per equivalent fracture, strain increased with the increase of hydrogen concentration. From the results of load-displacement behaviors, Esp behaviors, macro- and micro-SEM fractographs and AE test it has been concluded that the SP test method using miniaturized specimen(10mmx10mmx0.5mm) will be a useful test method to evaluate the fracture strength for CANDU pressure tube embrittled by hydrogen

Comparison of evaluation method for planar flaw in pressure tube

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

Elmoselhi, M.B.; Donner, A.; Brennenstuhl, A.; Warr, B.D.; Ellis, P.J.; Evans, D.W.

2002-01-01

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

Improving the calandria tubes for CANDU reactors

International Nuclear Information System (INIS)

Coleman, C.E.; Fong, R.W.L.; Doubt, G.L.

1997-01-01

CANDU calandria tubes are made from annealed Zircaloy-2 sheet formed into a cylinder and welded along its length to make the tube. The current calandria tubes have given exemplary service for many years. With more stringent regulations and the need to accommodate warm cooling water in tropical countries, we started a development program to increase the margins for failure during postulated accidents. These improvements involve increasing the tube strength and optimising the heat-transfer from an excessively hot fuel channel to the cool moderator. If the postulated accident involves a pressure tube break, it would be desirable if the calandria tube withstood the full pressure of the heat-transport system. The weakest link in current calandria tubes is the weld. Thickening the weld can increase the strength by 20% while seamless tubes can be 45% stronger than current tubes. The latter tubes can hold full system pressure for many hours without failure. If during the postulated accident the fuel and pressure tube become excessively hot but do not touch the calandria tube, the radiant heat loss must be maximised. Current calandria tubes have an absorptivity (emissivity) of about 0.2. To protect the fuel and the fuel channel we have devised a finish to the inside surface of the calandria tube that increases the emissivity to 0.7. If during the postulated accident the hot pressure tube touches the cool calandria tube, the contact conductance and the critical heat flux must be optimised to ensure nucleate boiling of the moderator at the outside surface of the calandria tube and therefore efficient exploitation of the moderator as a heat sink. In laboratory tests small ridges on the inside surface and roughening of the outside surface have been shown to increase the margins against failure and increase the possible moderator temperatures thus providing the opportunity to decrease the cost of the moderator heat-exchange system and remove restrictions on reactor operation in

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-05-15

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

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

International Nuclear Information System (INIS)

Yoon, C.; Rhee, B. W.; Chung, B. D.; Cho, Y. J.; Kim, M. W.

2008-01-01

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

Critical heat flux for downward-facing pool boiling on CANDU calandria tube surface

Energy Technology Data Exchange (ETDEWEB)

Behdadi, Azin, E-mail: behdada@mcmaster.ca; Talebi, Farshad; Luxat, John

2017-04-15

Highlights: • Pressure tube-calandria tube contact may challenge fuel channel integrity in CANDU. • Critical heat flux variation is predicted on the outer surface of CANDU calandria tube. • A two-phase boundary layer flow driven by buoyancy is modeled on the surface. • Different slip ratios and flow regimes are considered inside the boundary layer. • Subcooling effects are added to the model using wall heat flux partitioning. - Abstract: One accident scenario in CANDU reactors that can challenge the integrity of the primary pressure boundary is a loss of coolant accident, referred to as critical break LOCA, in which the pressure tube (PT) can undergo thermal creep strain deformation and contact its calandria tube (CT). In such case, rapid redistribution of stored heat from PT to CT, leads to a large spike in heat flux to the moderator which can cause bubble accumulation and dryout on the CT surface. A challenge to fuel channel integrity is posed if critical heat flux occurs on the surface of the CT and results in sustained film boiling. If the post-dryout temperature becomes sufficiently high then continued creep strain of the PT and CT may lead to fuel channel failure. In this study, a mechanistic model is developed to predict the critical heat flux variations along the downward facing outer surface of CT. The hydrodynamic model considers a liquid macrolayer beneath an elongated vapor slug on the surface. Local dryout is postulated to occur whenever the fresh liquid supply to the macrolayer is not sufficient to compensate for the liquid depletion. A boundary layer analysis is performed, treating the two phase motion as an external buoyancy driven flow. The model shows good agreement with the available experimental data and has been modified to take into account the effect of subcooling.

Leak-before-break concept for evaluation of flows detected in pressure tubes in a Candu type reactor

International Nuclear Information System (INIS)

Crespi, J.C.

1992-01-01

This paper reviews the role of the Leak-Before-Break concept for evaluation of flaws detected in cold-worked Zr 2.5% Nb pressure tubes in a CANDU type reactors. The acceptance criteria are intended to prevent failure by brittle fracture, plastic collapse of the ligament and delayed hydride cracking. The methodology developed here was of great help in order to establish the operative conditions for fuel channel garter springs repositioning by means of the SLA Rette tool at Embalse Nuclear Generating Station, Cordoba, Argentina. (author)

A survey on the corrosion susceptibility of Alloy 800 CANDU steam generator tubing materials

International Nuclear Information System (INIS)

Lu, Y.C.; Dupuis, M.; Burns, D.

2008-01-01

To provide support for a proactive steam generator (SG) aging management strategy, a survey on the corrosion susceptibility of the archived Alloy 800 tubing from CANDU SGs under plausible crevice chemistry conditions was conducted to assess the potential material degradation issues in CANDU SGs. Archived Alloy 800 samples were collected from four CANDU utilities. High-temperature electrochemical analysis was carried out to assess the corrosion susceptibility of the archived SG tubing under simulated CANDU crevice chemistry conditions at both 150 o C and 300 o C. The potentiodynamic polarization results obtained from the archived CANDU SG tubes were compared to the data from ex-service tubes removed from Darlington Nuclear Generating Station (DNGS) SGs and a reference nuclear grade Alloy 800 tubing. It was found that the removed Darlington SG tubes, with signs of in-service degradation, were more susceptible to pitting corrosion than the reference nuclear grade Alloy 800 tubing. At 150 o C, under the same neutral crevice chemistry conditions, the potentiodynamic polarization curve of the ex-service Darlington SG tubing has an active peak, which is a sign of propensity to crevice/underdeposit corrosion. This active peak was not observed in any of the potentiodynamic polarization curves of all archived Alloy 800 CANDU SG tubing indicating that archived CANDU SG tubes are less susceptible to the underdeposit corrosion under SG startup conditions. The corrosion behaviour of the archived Alloy 800 tubes from CANDU SG was similar to that of the reference nuclear grade Alloy 800 tubing. The results of this survey suggest that the Alloy 800 tubing materials used in the existing CANDU utilities (other than ex-service DNGS tubing) will continue to have reliable performance under specified CANDU operating conditions. Ex-service SG tubing from DNGS, although showing lower than average corrosion resistance, still has a wide acceptable operating margin and the in

Impact of aging and material structure on CANDU plant performance

International Nuclear Information System (INIS)

Nadeau, E.; Ballyk, J.; Ghalavand, N.

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-01-15

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

In situ sampling for pressure tube deuterium concentration

International Nuclear Information System (INIS)

Harrington, A.J.; Kittmer, C.A.

1988-01-01

The present method of assessing the useful life of pressure tubes in CANDU (CANada Deuterium Uranium) reactors requires the periodic removal and examination of a tube. Special tooling was developed at Atomic Energy of Canada Limited (AECL) to obtain a sample of material from a pressure tube without removing the tube from the reactor. The sampling tool concept has been successfully used by Ontario Hydro during scheduled outages at the Pickering Nuclear Generating Station (PNGS). (author)

Innovation in pressure tube life assessment

International Nuclear Information System (INIS)

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

2003-01-01

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

Impact of aging and material structure on CANDU plant performance

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Jyrkama, Mikko I., E-mail: mjyrkama@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada); Bickel, Grant A., E-mail: grant.bickel@cnl.ca [Canadian Nuclear Laboratories, Chalk River Laboratories, Chalk River, ON, Canada K0J 1J0 (Canada); Pandey, Mahesh D., E-mail: mdpandey@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada)

2016-04-15

Highlights: • New and simple statistical model of pressure tube diametral creep. • Based on surveillance data of 328 pressure tubes from eight different CANDU reactors. • Uses weighted least squares (WLS) to regress out operating conditions. • The shape of the diametral creep profiles are predicted very well. • Provides insight and relative ranking of strain behaviour of in-service tubes. - Abstract: This paper presents the development of a simplified regression approach for modelling the diametral creep over time in Zr-2.5 wt% Nb pressure tubes used in CANDU reactors. The model is based on a large dataset of in-service inspection data of 328 different pressure tubes from eight different CANDU reactor units. The proposed weighted least squares (WLS) regression model is linear in time as a function of flux and temperature, with a temperature-dependent variance function. The model predicts the shape of the observed diametral creep profiles very well, and is useful not merely for prediction, but also for assessing tube-to-tube variability and manufacturing properties among the inspected tubes.

Evaluation of Fatigue Crack Initiation for Volumetric Flaw in Pressure Tube

International Nuclear Information System (INIS)

Choi, Sung Nam; Yoo, Hyun Joo

2005-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Advancing the CANDU reactor: From generation to generation

International Nuclear Information System (INIS)

Hopwood, Jerry; Duffey, Romney B.; Yu, Steven; Torgerson, Dave F.

2006-01-01

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

Specific aspects in the manufacturing and operating of CANDU reactor pressure tubes (P/T)

International Nuclear Information System (INIS)

Muscaloiu, C.

1997-01-01

The CANDU reactor design is based on a number of individual P/T in which nuclear fuel bundles are located. P/T are required to be operated in an environment of elevated temperature (300 o C), internal pressure (10 Mpa), fast neutron flux (E>1 MeV) and heavy water. The most suitable material which can provide the desired neutron economy and still maintain its mechanical properties along with corrosion resistance is zirconium alloys Zr+ 2.5 % Nb with the following composition: niobium, 2.5 to 2.8 weight percent; oxygen, 1,000 to 1,300 ppm; zirconium + allowed impurities - balance. A total of 380 pressure tubes are installed into reactor. Each pressure tube is attached at each end to a stainless steel end fitting by means of a grooved, expanded joint. The installation works were performed by ANM Bucuresti, under the technical support of General Electric Canada. The integrity of P/T after installation was examined as follows: - the surface of the rolled area on unrolled internal surface extending 25 mm beyond rolled area was inspected for irregularities by means of a boroscope; - all pressure tubes were subjected to the helium leak test after F/C installation. During P/T operating life periodical inspections according to Canadian Standard CSA N285.4 are performed. The selection of the P/T for inspection is based either on particular properties or on the operating conditions of the fuel channel. The inspection consists in: a) Base Line Inspection within 2 years period commencing after 7,000 EFPH of operation which will include a volumetric inspection over P/T full length and measurements of P/T sag, ID, wall thickness and F/C bearing positions; b) Periodic Inspection in the same conditions plus material surveillance (on the four most significant indication P/T detected during the Base Line Inspection). The inspection will be performed on 14 selected P/T. (author)

Bulging of pressure tubes at hot spots under LOCA conditions

International Nuclear Information System (INIS)

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

1996-01-01

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

CANDU-PHW fuel channel replacement experience

International Nuclear Information System (INIS)

Dunn, J.T.; Kakaria, B.K.

1982-09-01

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

CANDU steam generator tubing material service experience and allied development

International Nuclear Information System (INIS)

Hart, A.E.; Lesurf, J.E.

1976-01-01

This paper covers the following aspects for the tube materials in CANDU-PHW steam generators: inservice performance with respect to tube leaks and coolant activity attributable to boiler tube corrosion, selection of tube materials for use with non-boiling and boiling primary coolants, supporting development on corrosion, vibration, fretting wear, tube inspection, leak detection and plugging of defective tubes. (author)

Pressure tube replacement in Pickering NGS A units 1 and 2

International Nuclear Information System (INIS)

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

1986-10-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Development of delayed hydride cracking resistant-pressure tube

Energy Technology Data Exchange (ETDEWEB)

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

2000-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-11-01

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

Delayed hydride cracking in Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

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

2007-01-01

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

Fuel rod bundles proposed for advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Prodea, Iosif; Catana, Alexandru

2010-01-01

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

Fuel channel design improvements for large CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Villamagna, A; Price, E G; Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

From the initial designs used in NPD and Douglas point reactors, the CANDU fuel channel and its components have undergone considerable development. Two major designs have evolved: the Pickering/CANDU 6 design which has 12 fuel bundles in the core and where the new fuel is inserted into the inlet end, and the Bruce/Darlington design which has 13 bundles in the channel and where new fuel is inserted into the outlet end. In the development of a single unit CANDU reactor of the size of a Bruce or Darlington unit which would use a Darlington design calandria, the decision has been made to use the CANDU 6 fuel channel rather than the Darlington design. The CANDU 6 channel has provided excellent performance and will not encounter the degree of maintenance required for the Bruce/Darlington design. The channel design in turn influences the fuelling machine/fuel handling concepts required. The changes to the CANDU 6 fuel channel design to incorporate it in the large unit are small. In fact, the changes that are proposed relate to the desire to increase margins between pressure tube properties and design conditions or ameliorate the consequences of postulated accident conditions, rather than necessary adaptation to the larger unit. Better properties have been achieved in the pressure tube material resulting from alloy development program over the past 10 years. Pressure tubes can now he made with very low hydrogen concentrations so that the hydrogen picked up as deuterium will not exceed the terminal solid solubility for the in-core region in 30 years. The improvements in metal chemistry allow the production of high toughness tubes that retain a high level of toughness during service. A small increase in wall thickness will reduce the dimensional changes without significantly affecting burnup. Changes to increase safety margins from postulated accidents are concentrated on containing the consequences of pressure tube damage. The changes are concentrated on the calandria tube

Thin-walled large-diameter zirconium alloy tubes in CANDU reactors

International Nuclear Information System (INIS)

Price, E.G.; Richinson, P.J.

1978-08-01

The requirements of the thin-walled large-diameter Zircaloy-2 tubing used in CANDU reactors are reviewed. Strength, residual stress patterns, texture and prior deformation contribute to the stability of these tubes. The extent to which the present manufacturing route meets these requirements is discussed. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Preliminary analysis for u tube degradation in CANDU steam generator using CATHENA

Energy Technology Data Exchange (ETDEWEB)

Shin, So Eun; Lee, Jeong Hun; Park, Tong Kyu; Hwang, Su Hyun [FNC Technology Co., Seoul (Korea, Republic of); Jung, Jong Yeo [KAERI, Daejeon (Korea, Republic of)

2012-10-15

The interest in plant safety and integrity has been increasing due to long term operation of nuclear power plants (NPPs) and lots of efforts have been devoted to developing the degradation evaluation model for all the Structure, System, and Components (SSCs) of NPPs in these days. The efforts, however, were mainly concentrated on pressurized light water reactors (PWRs) in domestic. In contrast, the study for the aging degradation of counterparts of CANDU (CANada Deuterium Uranium) reactors has been rarely performed, even though Wolsong unit 1 (WS1), that is a CANDU 6 NPP in Korea, has been operating for almost 30 years. Therefore, the assessment of the aging degradation is required and the proper and exact evaluation model for the aging degradation of SCCs of CANDU, especially WS1, is urgently needed. In this study, the aging degradation of steam generators (SGs) in WS1 was mainly discussed. Based on cases of the aging degradation of SGs in overseas CANDU reactors, the major potential aging mechanisms of SGs were estimated since there has been no case of accident due to degradation in CANDU NPPs in Korea . Some core parameters which are indicators of the degree of degradation were calculated by CATHENA (Canadian algorithm for thermal hydraulic network analysis). In the result of comparing two calculation cases; core parameters for only aged SGs in fresh plant and those for all the aged component, it can be concluded that aging of SGs is a main component in the degradation assessment of CANDU NPPs, and keeping the integrity of steam generator (SG) tubes is important to guarantee the safety of the NPPs.

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

International Nuclear Information System (INIS)

Marques, Andre L.F.; Todreas, Neil E.; Driscoll, Michael J.

2000-01-01

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 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)

Deformation of in-service pressure tubes

International Nuclear Information System (INIS)

Sarce, A.L.

1993-01-01

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

Delayed hydrogen cracking test design for pressure tubes

International Nuclear Information System (INIS)

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

1999-01-01

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

The CANDU 9

International Nuclear Information System (INIS)

Hart, R.S.

1994-01-01

The CANDU 9 plants are single unit versions of the Bruce B design, incorporating relevant technical advances made in CANDU 6, and the newer Darlington and CANDU 3 designs. This paper describes the CANDU 9 480/SEU, with an electrical output of about 1050 MW. In this designation, 480 refers to the number of fuel channels, and SEU to slightly enriched uranium. Emphasis is placed on evolutionary design, and the use of well proven design features, to ensure regulatory licensability and reliable operation. Safety is enhanced through simplification and improvement of key systems and components. Relatively low energy costs result from reduced specific capital cost, reduced operating and maintenance cost, and reduced radiation exposure to personnel. Standardization is emphasized inasmuch as all key components (steam generators, heat transport pumps, pressure tubes fuelling machines etc.) ar of the same design as those in operating CANDU stations. Advanced CANDU fuel cycles are readily accommodated. 1 ref., 1 tab., 11 figs

Heat exchanger tubing materials for CANDU nuclear generating stations

International Nuclear Information System (INIS)

Taylor, G.F.

1977-07-01

The performance of steam generator tubing (nickel-chromium-iron alloy in NPD and nickel-copper alloy in Douglas Point and Pickering generating stations) has been outstanding and no corrosion-induced failures have occurred. The primary coolant will be allowed to boil in the 600 MW (electrical) CANDU-PHW reactors. An iron-nickel-chromium alloy has been selected for the steam generator tubing because it will result in lower radiation fields than the alloys used before. It is also more resistant than nickel-chromium-iron alloy to stress corrosion cracking in the high purity water of the primary circuit, an unlikely but conceivable hazard associated with higher operating temperatures. Austenitic alloy and ferritic-austenitic stainless steel tubing have been selected for the moderator coolers in CANDU reactors being designed and under construction. These materials will reduce the radiation fields around the moderator circuit while retaining the good resistance to corrosion in service water that has characterized the copper-nickel alloys now in use. Brass and bronze tubes in feedwater heaters and condensers have given satisfactory service but do, however, complicate corrosion control in the steam cycle and, to reduce the transport of corrosion products from the feedtrain to the steam generator, stainless steel is preferred for feedwater heaters and stainlss steel or titanium for condensers. (author)

Development of Realistic Safety Analysis Technology for CANDU Reactors

International Nuclear Information System (INIS)

Park, Joo Hwan; Rhee, B. W.; Rho, G. H.

2010-04-01

The following 3 research items have been studied to develop and establish the realistic safety analysis and the associated technologies for a CANDU reactor. At the first, WIMS-CANDU which is physics cell code for a CANDU has been improved and validated against the physics criticality experiment data transferred through the international cooperation programs. Also an improved physics model to take into account the pressure tube creep was developed and utilized to assess the effects of the pressure tube creep of 0%, 2.5% and 5% diametral increase of pressure tube on core physics parameters. Secondly, the interfacing module between physics and thermal-hydraulics codes has been developed to provide the enhancement of reliability and convenience of the calculation results of the physics parameters such as power coefficient which was calculated by independent code systems. Finally, the important parameters related to the complex heat transfer mechanisms in the crept pressure tubes were identified to find how to improve the existing fuel channel models. One of the important parameters such as the oxidation model of Zr-steam reaction was identified, implemented and verified with the experimental data of the high pressure and temperature fuel channel and its model was utilized for CFD analysis of the crept pressure tube effect on the reactor safety. The results were also provided to validate the CATNENA models of the crept pressure tube and the effects of the pressure tube creep on the blowdown and post-blowdown phase during LOCA was assessed. The results of this study can be used to assess the uncertainty analysis of coolant void reactivity and the effects of the creep deformed pressure tubes on physics/TH/safety issues. Also, those results will be used to improve the current design and operational safety analysis codes, and to technically support the related issues to resolve their problems

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

International Nuclear Information System (INIS)

Fan, H.Z.; Bilanovic, Z.; Nitheanandan, T.

2004-01-01

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

CANDU fuel

International Nuclear Information System (INIS)

MacEwan, J.R.; Notley, M.J.F.; Wood, J.C.; Gacesa, M.

1982-09-01

The direction of CANDU fuel development was set in 1957 with the decision to build pressure tube reactors. Short - 50 cm long - rodded bundles of natural UO 2 clad in Zircaloy were adopted to facilitate on-power fuelling to improve uranium utilization. Progressive improvements were made during 25 years of development, involving 650 man years and 180 million dollars. Today's CANDU bundle is based on the knowledge gained from extensive irradiation testing and experience in power reactors. The main thrust of future development is to demonstrate that the present bundle is suitable, with minor modifications, for thorium fuels

Database for Pressure Tube Diameter and Operation Data of Wolsong NPP

International Nuclear Information System (INIS)

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

2010-12-01

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

Pressure Tube and Pressure Vessel Reactors; certain comparisons

Energy Technology Data Exchange (ETDEWEB)

Margen, P H; Ahlstroem, P E; Pershagen, B

1961-04-15

In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

Pressure Tube and Pressure Vessel Reactors; certain comparisons

International Nuclear Information System (INIS)

Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

1961-04-01

In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

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

International Nuclear Information System (INIS)

Gyongyosi, T.

2005-01-01

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

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

International Nuclear Information System (INIS)

Gyongyosi, T.

2005-01-01

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

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

Mathematical modeling of CANDU-PHWR

Energy Technology Data Exchange (ETDEWEB)

Gaber, F.A.; Aly, R.A.; El-Shal, A.O. [Atomic Energy Authority, Cairo (Egypt)

2001-07-01

The paper deals with the transient studies of CANDU 600 pressurized Heavy Water Reactor (PHWR) system. This study involved mathematical modeling of CANDU PHWR major system components and the developments of software to study the thermodynamic performances. Modeling of CANDU-PHWR was based on lumped parameter technique.The integrated CANDU-PHWR model includes the neutronic, reactivity, fuel channel heat transfer, piping and the preheater type U-tube steam generator (PUTSG). The nuclear reactor power was modelled using the point kinetics equations with six groups of delayed neutrons and reactivity feed back due to the changes in fuel temperature and coolant temperature. The complex operation of the preheater type U-tube steam generator (PUTSG) is represented by a non-linear dynamic model using a state variable, moving boundary and lumped parameter techniques. The secondary side of the PUTSG model has six separate lumps including a preheater region, a lower boiling section, a mixing region, a riser, a chimmeny section, and a down-corner. The tube side of PUTSG has three main thermal zones. The PUTSG model is based on conservation of mass, energy and momentum relation-ships. The CANDU-PHWR integrated model are coded in FORTRAN language and solved by using a standard numerical technique. The adequacy of the model was tested by assessing the physical plausibility of the obtained results. (author)

Probabilistic modeling of material resistance to crack initiation due to hydrided region overloads in CANDU Zr-2.5%Nb pressure tubes

International Nuclear Information System (INIS)

Gutkin, L.; Scarth, D.A.

2014-01-01

Zr-2.5%Nb pressure tubes in CANDU nuclear reactors are susceptible to hydride-assisted cracking at the locations of stress concentration, such as in-service flaws. Probabilistic methodology is being developed to evaluate such flaws for crack initiation due to hydrided region overloads, which occur when the applied stress acting on a flaw with an existing hydrided region at its tip exceeds the stress at which the hydrided region is formed. As part of this development, probabilistic modeling of pressure tube material resistance to overload crack initiation has been performed on the basis of a set of test data specifically designed to study the effects of non-ratcheting hydride formation conditions and load reduction prior to hydride formation. In the modeling framework, the overload resistance is represented as a power-law function of the material resistance to initiation of delayed hydride cracking under constant loading, where both the overload crack initiation coefficient and the overload crack initiation exponent vary with the flaw geometry. In addition, the overload crack initiation coefficient varies with the extent of load reduction prior to hydride formation as well as the number of non-ratcheting hydride formation thermal cycles. (author)

Probabilistic modeling of material resistance to crack initiation due to hydrided region overloads in CANDU Zr-2.5%Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Gutkin, L.; Scarth, D.A. [Kinectrics Inc., Toronto, ON (Canada)

2014-07-01

Zr-2.5%Nb pressure tubes in CANDU nuclear reactors are susceptible to hydride-assisted cracking at the locations of stress concentration, such as in-service flaws. Probabilistic methodology is being developed to evaluate such flaws for crack initiation due to hydrided region overloads, which occur when the applied stress acting on a flaw with an existing hydrided region at its tip exceeds the stress at which the hydrided region is formed. As part of this development, probabilistic modeling of pressure tube material resistance to overload crack initiation has been performed on the basis of a set of test data specifically designed to study the effects of non-ratcheting hydride formation conditions and load reduction prior to hydride formation. In the modeling framework, the overload resistance is represented as a power-law function of the material resistance to initiation of delayed hydride cracking under constant loading, where both the overload crack initiation coefficient and the overload crack initiation exponent vary with the flaw geometry. In addition, the overload crack initiation coefficient varies with the extent of load reduction prior to hydride formation as well as the number of non-ratcheting hydride formation thermal cycles. (author)

Fuel Management Study for a CANDU reactor Using New Physics Codes Suite

International Nuclear Information System (INIS)

Kim, Won Young; Kim, Bong Ghi; Park, Joo Hwan

2008-01-01

A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. The primary reactivity control in a CANDU reactor is the on-power refueling on a daily basis and an additional reactivity control is provided through an individual reactivity device movement, which includes 21 adjusters, 6 liquid zone controllers, 4 mechanical control absorbers and 2 shutdown systems. The refueling in CANDU is carried out on power and this makes the in-core fuel management different from that in a reactor refueled during shutdowns. The objective of a fuel management is to determine a fuel loading and fuel replacement procedure which will result in a minimum total unit energy cost in a safe and reliable operation. In this article, the in-core fuel management for the CANDU reactor was studied by using the new physics code suite of WIMS-IST/DRAGON-IST/RFSP-IST with the model of Wolsong-1 NPP

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

International Nuclear Information System (INIS)

Azer, N.; Wong, H.

1994-01-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Advanced pressure tube sampling tools

International Nuclear Information System (INIS)

Wittich, K.C.; King, J.M.

2002-01-01

Deuterium concentration is an important parameter that must be assessed to evaluate the Fitness for service of CANDU pressure tubes. In-reactor pressure tube sampling allows accurate deuterium concentration assessment to be made without the expenses associated with fuel channel removal. This technology, which AECL has developed over the past fifteen years, has become the standard method for deuterium concentration assessment. AECL is developing a multi-head tool that would reduce in-reactor handling overhead by allowing one tool to sequentially sample at all four axial pressure tube locations before removal from the reactor. Four sets of independent cutting heads, like those on the existing sampling tools, facilitate this incorporating proven technology demonstrated in over 1400 in-reactor samples taken to date. The multi-head tool is delivered by AECL's Advanced Delivery Machine or other similar delivery machines. Further, AECL has developed an automated sample handling system that receives and processes the tool once out of the reactor. This system retrieves samples from the tool, dries, weighs and places them in labelled vials which are then directed into shielded shipping flasks. The multi-head wet sampling tool and the automated sample handling system are based on proven technology and offer continued savings and dose reduction to utilities in a competitive electricity market. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-07-01

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

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

International Nuclear Information System (INIS)

Mihalache, Maria; Vasile, Radu; Deaconu, Mariea

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Luncheon address: Development of the CANDU reactor

International Nuclear Information System (INIS)

Bain, A.S.

1997-01-01

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

Leak-before-break experience in CANDU reactors

International Nuclear Information System (INIS)

Price, E.G.; Moan, G.D.; Coleman, C.E.

1988-01-01

In the Canada deuterium uranium (CANDU) reactor, each of the ∼ 400 hot pressure tubes containing the fuel bundles and the pressurized heat transport water is surrounded and insulated from the cold moderator by a calandria tube. The pressure tubes are made from cold-worked Zr-2.5 Nb with a minimum wall thickness of 4.19 mm, and the calandria tubes are made from annealed Zircaloy-2 with a minimum wall thickness of 1.37 mm. The annulus between these two tubes contains an inert gas. Leak-before-break has developed into an operational tool in CANDU reactors to prevent unstable failure of pressure tubes. A procedure for leak detection and reactor response has been developed from the use of the annulus gas, whose dew point is measured to ascertain if leaks have crept into the annulus. The characteristics of the crack are used to establish the response time for leak detection. The reactor is required to be shut down before the length of the slowly growing crack has reached the critical stage. This critical crack length, determined using slit burst tests on tubes, is the crack length at which the crack growth becomes unstable. The most likely crack growth mechanism is delayed hydride cracking. This mechanism requires three conditions to occur simultaneously: the material must be sensitive to delayed hydride cracking; zirconium hydrides must be present in the material; and the tensile stress must be sufficiently great

Technologies in support of CANDU development

International Nuclear Information System (INIS)

Turner, C.; Tapping, B.

2005-01-01

Atomic Energy of Canada, Ltd. (AECL) has significant research and development (R and D) programs designed to meet the needs of both existing CANDU reactors and new and evolving CANDU plant designs. These R and D programs cover a wide range of technology, from chemistry and materials support through to inspection and life management tools. Emphasis is placed on effective technology development programs for fuel channels, feeders and steam generators to ensure their operation through design life, and beyond. This paper specifically addresses how the R and D has been applied in the production of longer-lived pressure tubes for the most recent CANDU 6 reactors, and how this technology forms the basis for the pressure tubes of the Advanced CANDU Reactor (ACR). Similarly, AECL has developed solutions for other critical components such as calandria tubes, feeder pipe and steam generators. The paper also discusses how the R and D knowledge has been integrated into aging management databases and health monitoring tools. Since 1997, AECL has been working with CANDU utilities on comprehensive and integrated CANDU Plant Life Management (PLiM) programs for successful and reliable plant operation through design life and beyond. AECL has developed and implemented an advanced chemistry monitoring and diagnostic system, called ChemAND which allows on-line access by the operators to current and past chemistry conditions enabling appropriate responses and facilitating planning of shutdown maintenance actions. An equivalent tool for monitoring, trending and diagnosing thermal and mechanical data has also been developed; this tool is called ThermAND. AECL is developing the Maintenance Information, Monitoring, and Control (MIMC) system, which provide information to the user for condition-based decision-making in maintenance. To enable more effective inspections, surveillance and data collection, AECL has developed unique one-off tooling to carry out unanticipated inspection and repair

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

International Nuclear Information System (INIS)

Bozzano, Patricia B.

2013-01-01

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

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

International Nuclear Information System (INIS)

Park, Jun Su; Jeong, Seung Ha

1996-06-01

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

Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

International Nuclear Information System (INIS)

Jiang, J.T.; Luxat, J.C.

2008-01-01

Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Jiang, J.T.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)

2008-07-01

Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

An update on CANDU fuel channel performance

International Nuclear Information System (INIS)

Price, E.G.; Holt, R.A; Wong, H.; Gautheir, P.; Ellis, P.J.; Slade, J.

1998-01-01

Fuel channel components, particularly pressure tubes, are monitored to establish changes in their condition expected from their service environment. Such monitoring takes the form of the in-service NDE, in-service sampling and periodic tube removal for destructive examination. The pressure tubes in early CANDU 6 stations have now reached mid-life. The expected trends in behaviour of the various properties of the pressure tubes are being confirmed. Diametral deformation potentially limits the life of some pressure tubes, and fuel developments, such as the CANFLEX fuel bundle, will provide additional CHF margins. Corrosion and the associated deuterium pick-up are monitored to assess if any change in rates have occurred. Both have historically continued on an essentially linear trend with time. Recent data indicate that the rates of pick-up could be gradually increasing. Changes in mechanical properties, particularly toughness, have reached saturation levels. In the early reactors, the toughness is acceptable, assuring leak-before-break. The pressure tubes in later reactors have very low concentrations of chlorine and phosphorous and the level of hydrogen are also low, and these tubes are predicted to retain relativity high fracture toughness in-service. Debris fretting wear has occurred in CANDU reactors. Laboratory testing and experience show that no cracks have initiated from debris frets to date. Further debris fretting will have to be avoided by care in maintenance procedures, and existing defects must be monitored in the future as hydrogen concentration levels increase. Spacer location and repositioning (SLAR) has been necessary in earlier reactors with loose spacers. Procedures and equipment for SLAR have been developed and successfully applied. The efficiency of spacer relocation operations has improved with experience. (authors)

Technical basis for the CANDU steam generator tube fitness-for-service guidelines

International Nuclear Information System (INIS)

Kozluk, M.J.; Scarth, D.A.; Graham, D.B.

2002-01-01

Active degradation mechanisms in steam generators and preheaters in Canadian CANDU T M generating stations are managed through Steam Generator Programs that incorporate tube inspection, maintenance (cleaning), fitness-for-service assessment, and preventative plugging as part of the overall steam generator management strategy. Steam generator and preheater tubes are inspected in accordance with the CSA Standard CAN/CSA-N285.4-94[l]. When a detected flaw indication does not satisfy the criteria of acceptance by examination, CSA-N285.4-94 permits a fitness-for-service assessment to determine acceptability. In 1999 Ontario Power Generation issued, for trial use, fitness-for-service guidelines for steam generator and preheater tubes in CANDU nuclear power plants. The main objectives of the Fitness-for-Service Guidelines are to provide reasonable assurance that tube structural integrity is maintained, and to provide reasonable assurance that there are adequate margins between estimated accumulated dose and applicable site dose limits. The Fitness-for-Service Guidelines are intended to provide industry-standard acceptance criteria and evaluation procedures for assessing the condition of steam generator and preheater tubes in terms of tube structural integrity, operational leak rate, and consequential leakage during an upset or abnormal event. This paper describes the technical basis for the minimum required safety factors specified in Table IC-1 of the Fitness-for-Service Guidelines and for the flaw models used to develop the flaw stability requirements in the nonmandatory, Appendix C of the Fitness-for-Service Guidelines. (author)

Fracture toughness behaviour using small CCT specimen of Zr-2.5Nb pressure tube materials

International Nuclear Information System (INIS)

Oh, Dong Joon; Kim, Young Suk; Ahn, Sang Bok; Im, Kyung Soo; Kwon, Sang Chul; Cheong, Yong Mu

2001-03-01

Fracture toughness of Zr-2.5Nb pressure tube is the essential data to estimate the CCL(critical crack length) for the concept of LBB(Leak-Before-Break) in PHWR. Zr-2.5Nb pressure tubes could be degraded due to the absorption of hydrogen from coolant and the irradiation. To investigate the fracture toughness behaviour such as J-resistance curves, dJ/da, and CCL of some Zr-alloys (CANDU-double, -quad, CW-E125, TMT-E125, E-635), the transverse tensile test and the fracture toughness test of small CCT (Curved Compact Tension) specimen with 17 mm width were carried out with the variation of testing temperature at different testing condition. To define the fracture mechanism of degradation, the fractographic comparison of fracture surface was performed using the stereoscope and SEM. In addition, the effect of non-uniformed pre-fatigue crack was also studied. In conclusion, CANDU double-melted was less tougher than CANDU quad-melted and the hydrogen embrittlement was found at room temperature. Finally, while the effect of non-uniformed pre-fatigue crack was considerable at room temperature, this effect was disappeared at 250-300 .deg. C

Corrosion control in CANDU nuclear power reactors

International Nuclear Information System (INIS)

Lesurf, J.E.

1974-01-01

Corrosion control in CANDU reactors which use pressurized heavy water (PHW) and boiling light water (BLW) coolants is discussed. Discussions are included on pressure tubes, primary water chemistry, fuel sheath oxidation and hydriding, and crud transport. It is noted that corrosion has not been a significant problem in CANDU nuclear power reactors which is a tribute to design, material selection, and chemistry control. This is particularly notable at the Pickering Nuclear Generating Station which will have four CANDU-PHW reactors of 540 MWe each. The net capacity factor for Pickering-I from first full power (May 1971) to March 1972 was 79.5 percent, and for Pickering II (first full power November 1971) to March 1972 was 83.5 percent. Pickering III has just reached full power operation (May 1972) and Pickering IV is still under construction. Gentilly CANDU-BLW reached full power operation in May 1972 after extensive commissioning tests at lower power levels with no major corrosion or chemistry problems appearing. Experience and operating data confirm that the value of careful attention to all aspects of corrosion control and augur well for future CANDU reactors. (U.S.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Severe accident development modeling and evaluation for CANDU

Energy Technology Data Exchange (ETDEWEB)

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

2009-09-15

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

Severe accident development modeling and evaluation for CANDU

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

2009-01-01

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

Rejection index for pressure tubes

International Nuclear Information System (INIS)

Mitchell, A.B.; Meneley, D.

1989-10-01

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

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

International Nuclear Information System (INIS)

Mohany, A.; Feenstra, P.; Janzen, V.P.; Richard, R.

2009-01-01

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

Examination of parameters affecting overload fracture behavior of flaw-tip hydrides in Zr-2.5Nb pressure tubes in Candu reactors

International Nuclear Information System (INIS)

Cui, J.; Shek, G.K.; Wang, Z.R.

2007-01-01

Service-induced flaws in Zr-2.5Nb alloy pressure tubes in Candu (Canada Deuterium Uranium Reactors) nuclear reactors are susceptible to a crack initiation and growth mechanism known as Delayed Hydride Cracking (DHC), which is a repetitive process that involves hydrogen diffusion, hydride precipitation, growth and fracture of a hydride region at the flaw-tip under a constant load. Crack initiation may also occur under another loading condition when the hydride region is subjected to an overload. An overload occurs when the hydride region at the flaw tip is loaded to a stress higher than that at which this region is formed such as when the reactor experiences a transient pressure higher than the normal operating pressure where the hydride region is formed. Flaw disposition requires justification that the hydride region overload will not fracture the hydride region, and initiate DHC. In this work, monotonically increasing load experiments were performed on unirradiated Zr-2.5Nb pressure tube specimens containing simulated debris frets (V-notch) and bearing pad frets (BPF, U-shape notch) to examine overload fracture behavior of flaw-tip hydrides formed under hydride ratcheting conditions. Hydride cracking in the overload tests was detected by the acoustic emission technique and confirmed by post-test metallurgical examination. Test results indicate that the resistance to overload fracture is affected by a number of parameters including hydride formation stress, flaw shape (V-notch vs. BPF) and flaw radius (0.015 mm vs. 0.1 mm). The notch-tip hydride morphologies were examined by optical microscopy and scanning electron microscopy (SEM) which show that they are affected by the hydride formation conditions, resulting in different overload fracture resistance. Finite element stress analyses were also performed to obtain flaw-tip stress distributions for interpretation of the test results. (authors)

OpenFOAM Analysis of CANDU-6 Moderator Flow

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Chang, Se-Myong

2015-01-01

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 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 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 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 transfer

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

Plating end fittings to reduce hydrogen ingress at rolled joints in CANDU reactors

International Nuclear Information System (INIS)

White, A.J.; Urbanic, V.F.; Bahurmuz, A.A.; Clendening, W.R.; Joynes, R.; McDougall, G.M.; Skinner, B.C.; Venkatapathi, S.

1993-10-01

Zr-2.5Nb pressure tubes in CANDU nuclear reactors absorb hydrogen at a low rate from the primary heat transport water circulated through the tubes. Extra hydrogen is picked up at the rolled joints that connect the pressure tubes to out-of-core steel piping. This enhanced ingress may contribute to pressure-tube cracking at incorrectly assembled joints. The risk of pressure-tube failure has been decreased by ensuring correct joint assembly, and could be further decreased by reducing hydrogen ingress at rolled joints. This paper reviews progress toward using plated end fittings to reduce rolled-joint hydrogen ingress

Methodology of structural integrity assesment of CANDU-6 NPP fuel channels

International Nuclear Information System (INIS)

Radu, Vasile

2004-01-01

The paper describes the methodology of assessing the structural integrity in the CANDU-6 fuel channels making use of alternative methods of evaluation. An evaluation of the structural integrity of a CANDU-6 pressure tube made of Zr-2,5%Nb presenting both sharp and blunted defects is done. These analyses are made in compliance with the Canadian guide 'Pressure Tubes Fitness-for-service', and other two Recognizing procedures internationally adopted: the British procedure R6/Rev.4 and the American procedure API 579. Previously, the data base containing the data on materials property as well as the heat and dynamical loads in normal operation in CANDU-6 pressure tubes was established. Obtaining complete diagrams for structural integrity of pressure tubes with sharp and blunted defects in conditions of normal operation and long-term irradiation is the next step of the methodology to be developed. Modelling sharp and blunted defects on the inner side of pressure tubes, which can occur in normal reactor operating conditions is achieved by means of capabilities of pre-processing of two finite element analysis codes namely FEA-Crack and FEA-Flaw. The second part of the work deals with the analyses by finite element method of the fracture mechanics by means of the FEA-Crack code and with the evaluation of sharp and blunted defects by FAD diagrams in compliance with the British procedure R6/Rev.4. For typical models of blunted defects and thermo-mechanical loads specific to normal operation finite element analyses by FEA-Flaw codes were performed. Then FAD-iDHC diagrams were constructed to evaluate the initiation of the slow cracking under hydrogen/deuterium absorption, the known phenomenon of delayed hydride cracking (DHC)

CANDU severe accident analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

2007-01-01

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

Severe core damage experiments and analysis for CANDU applications

International Nuclear Information System (INIS)

Mathew, P.M.; White, A.J.; Snell, V.G.; Bonechi, M.

2003-01-01

AECL uses the MAAP CANDU code to calculate the progression of a severe core damage accident in a CANDU reactor to support Level 2 Probabilistic Safety Assessment and Severe Accident Management activities. Experimental data are required to ensure that the core damage models used in MAAP CANDU code are adequate. In SMiRT 16, details of single channel experiments were presented to elucidate the mechanisms of core debris formation. This paper presents the progress made in severe core damage experiments since then using single channels in an inert atmosphere and results of the model development work to support the experiments. The core disassembly experiments are conducted with one-fifth scale channels made of Zr-2.5wt%Nb containing twelve simulated fuel bundles in an inert atmosphere. The reference fuel channel geometry consists of a pressure tube/calandria tube composite, with the pressure tube ballooned into circumferential contact with the calandria tube. Experimental results from single channel tests showed the development of time-dependent sag when the reference channel temperature exceeded 850 degC. The test results also showed significant strain localization in the gap at the bundle junctions along the bottom side of the channel, thus suggesting creep to be the main deformation mechanism for debris formation. An ABAQUS finite element model using two-dimensional beam elements with circular cross-section was developed to explain the experimental findings. A comparison of the calculated central sag (at mid-span), the axial displacement at the free end of the channel and the post-test sag profile showed good agreement with the experiments, when strain localization was included in the model, suggesting such a simple modelling approach would be adequate to explain the test findings. The results of the tests are important not only in the context of the validation of the analytical tools and models adopted by AECL for the severe accident analysis of CANDU reactors but

Thermo-Economic Assessment of Advanced,High-Temperature CANDU Reactors

International Nuclear Information System (INIS)

Spinks, Norman J.; Pontikakis, Nikos; Duffey, Romney B.

2002-01-01

Research underway on the advanced CANDU examines new, innovative, reactor concepts with the aim of significant cost reduction and resource sustainability through improved thermodynamic efficiency and plant simplification. The so-called CANDU-X concept retains the key elements of the current CANDU designs, including heavy-water moderator that provides a passive heat sink and horizontal pressure tubes. Improvement in thermodynamic efficiency is sought via substantial increases in both pressure and temperature of the reactor coolant. Following on from the new Next Generation (NG) CANDU, which is ready for markets in 2005 and beyond, the reactor coolant is chosen to be light water but at supercritical operating conditions. Two different temperature regimes are being studied, Mark 1 and Mark 2, based respectively on continued use of zirconium or on stainless-steel-based fuel cladding. Three distinct cycle options have been proposed for Mark 1: the High-Pressure Steam Generator (HPSG) cycle, the Dual cycle, and the Direct cycle. For Mark 2, the focus is on simplification via a Direct cycle. This paper presents comparative thermo-economic assessments of the CANDU-X cycle options, with the ultimate goal of ascertaining which particular cycle option is the best overall in terms of thermodynamics and economics. A similar assessment was already performed for the NG CANDU. The economic analyses entail obtaining cost estimates of major plant components, such as heat exchangers, turbines and pumps. (authors)

A CANDU Severe Accident Analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie

2006-01-01

As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents for CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D2O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 10000 deg C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the existing data. The results are encouraging. (authors)

Modelling nuclear fuel vibrations in horizontal CANDU reactors

International Nuclear Information System (INIS)

Jagannath, D.V.; Oldaker, I.E.

1976-01-01

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

Advances in fuel channel technology for CANDU reactors

International Nuclear Information System (INIS)

Cheadle, B.A.; Coleman, C.E.

1994-05-01

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

Application of fuel management calculation codes for CANDU reactor

International Nuclear Information System (INIS)

Ju Haitao; Wu Hongchun

2003-01-01

Qinshan Phase III Nuclear Power Plant adopts CANDU-6 reactors. It is the first time for China to introduce this heavy water pressure tube reactor. In order to meet the demands of the fuel management calculation, DRAGON/DONJON code is developed in this paper. Some initial fuel management calculations about CANDU-6 reactor of Qinshan Phase III are carried out using DRAGON/DONJON code. The results indicate that DRAGON/DONJON can be used for the fuel management calculation for Qinshan Phase III

CANDU steam generator life management

International Nuclear Information System (INIS)

Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

1998-01-01

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

Fracture Toughness Round Robin Test International in pressure tube materials

International Nuclear Information System (INIS)

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

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Jiang, J.T.; Luxat, J.C. [McMaster University, A315 JHE Building, 1280 Main St.W. Hamilton, ON, L8S 4L7 (Canada)

2008-07-01

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

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

International Nuclear Information System (INIS)

Jiang, J.T.; Luxat, J.C.

2008-01-01

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

Flooding of a large, passive, pressure-tube light water reactor

International Nuclear Information System (INIS)

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

1997-01-01

A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

Examination of core components removed from CANDU reactors

International Nuclear Information System (INIS)

Cheadle, B.A.; Coleman, C.E.; Rodgers, D.K.; Davies, P.H.; Chow, C.K.; Griffiths, M.

1988-11-01

Components in the core of a nuclear reactor degrade because the environment is severe. For example, in CANDU reactors the pressure tubes must contend with the effects of hot pressurised water and damage by a flux of fast neutrons. To evaluate any deterioration of components and determine the cause of the occasional failure, we have developed a wide range of remote-handling techniques to examine radioactive materials. As well as pressure tubes, we have examined calandria tubes, garter springs, end fittings, liquid-zone control units and flux detectors. The results from these examinations have produced solutions to problems and continually provide information to help understand the processes that may limit the lifetime of a component

The water chemistry of CANDU PHW reactors

International Nuclear Information System (INIS)

LeSurf, J.E.

1978-01-01

This review will discuss the chemistry of the three major water circuits in a CANDU-PHW reactor, viz., the Primary Heat Transport (PHT) water, the moderator and the boiler water. An important consideration for the PHT chemistry is the control of corrosion and of the transport of corrosion products to minimize the growth of radiation fields. In new reactors the PHT will be allowed to boil, requiring reconsideration of the methods used to radiolytic oxygen and elevate the pH. Separation of the moderator from the PHT in the pressure-tubed CANDU design permits better optimization of the chemistry of each system, avoiding the compromises necessary when the same water serves both functions. Major objectives in moderator chemistry are to control (a) the radiolytic decomposition of D 2 0; (b) the concentration of soluble neutron poisons added to adjust reactivity; and (c) the chemistry of shutdown systems. The boiler water and its feed water are treated to avoid boiler tube corrosion, both during normal operation and when perturbations are caused to the feed by, for example, leaks in the condenser tubes which permit ingress of untreated condenser cooling water. Development of a system for automatic analysis and control of feed water to give rapid, reliable response to abnormal conditions is a novel feature which has been developed for incorporation in future CANDU-PHW reactors. (author)

Maintenance and plugging technology for CANDU steam generator tubing

International Nuclear Information System (INIS)

Prince, J.; Nicholson, A.; Hare, J.; McGoey, L.; Stafford, T.; Gowthorpe, P.

2006-01-01

In order to keep aging steam generators in service and to successfully manage the life of these critical components, the capability must exist to perform tube plugging and other complex maintenance activities in-situ. In the early days of CANDU steam generator operation, the only option was to perform these activities manually, which had inherent safety and quality risks. The challenge was to be able to perform these activities remotely thus eliminating some of the confined space and radiological exposure risks. The additional challenge was to develop equipment and techniques which would result in significantly improved quality, particularly for the completed plug welds which would be returned to service. Over the past fifteen years, this technology has matured and has produced remarkable results in field application. Some 14000 tube plugs have been successfully installed to date using automated plugging techniques. This paper presents an overview of the development of techniques available to utilities for steam generator tube plugging as well as some highlights of other steam generator tube maintenance activities such as primary side tube removal and tube end damage repair. Aspects covered in the paper include plug and procedure development, automated equipment and manipulators for tool deployment, process controls and personnel requirements. Recently, the steam generator tube plugging performed by OPG has been incorporated into a formal quality program under the requirements of ASME NCA 4000. An overview of the quality program will be presented and details of some of the important aspects of the quality program will be discussed. (author)

Inert medium (helium) irradiation testing of pressure tube samples

International Nuclear Information System (INIS)

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

2001-01-01

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

Hydride blister formation simulation in Candu type reactors

International Nuclear Information System (INIS)

Otero, D.; Bollini, C.; Sangregorio, D.

1992-01-01

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

Thermal hydraulics of CANDU severe accident analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

2007-01-01

As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents in CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D 2 O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the available data. The results are encouraging. (authors)

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

International Nuclear Information System (INIS)

Gulshani, P.; So, C.B.

1986-10-01

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

Development of inspection equipment for fuel bundles of CANDU-PHWR using R981 underwater radiation tolerant camera

Energy Technology Data Exchange (ETDEWEB)

Koo, Dae-Seo; Cho, Moon-Sung; Jo, Chang-Keun; Jun, Ji-Su; Jung, Jong Yeob; Park, Kwang-June; Suk, Ho-Chun

2005-03-15

The inspection equipment of fuel bundles was developed, which could perform visual inspection and dimensional measurement on fuel bundles of CANDU-PHWR, to evaluate, analyze the defective behavior of fuel bundles and inner surface of pressure tubes of inherent two-phase flow over 24kg/s in CANDU-6. The R981 radiation tolerant camera system with pan and tilt function was ordered and manufactured, which was waterproof, shielding radiation in underwater 10m in depth. The performance test, of the system ,due to camera-object distance was carried out in air/underwater atmosphere. The results of performance test of R981 radiation tolerant camera system are good. The inspection equipment of fuel bundles using R981 radiation tolerant camera system and underwater-radiation tolerant LVDT sensor(D5/200AW) was fabricated, which could perform visual inspection and dimensional measurement on fuel bundles of CANDU-PHWR with measurement accuracy 10{mu}m. This equipment will be utilizable integrity evaluation of fuel bundles which are irradiated in pressure tube of CANDU-PHWR.

Characteristics of CANDU fuel bundles that caused pressure tube fretting at the bundle midplane

Energy Technology Data Exchange (ETDEWEB)

Dennier, D; Manzer, A M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Koehn, E [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

Detailed measurements on new bundles, and those that caused fretting during in- and out-reactor tests, have given insight into the factors responsible for fretting at the midplane of the inlet bundle. Bottom fuel elements that were attached near radial endplate spokes and had inboard bearing pads in the rolled joint cavity produced a significant portion of the observed fret marks. These elements are influenced by several driving forces that deflect the centre bearing pads towards the pressure tube surface. The evidence suggests that slight changes in bundle design may be possible to reduce pressure tube fretting. (author). 4 refs., 3 tabs., 8 figs.

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

International Nuclear Information System (INIS)

Cirimello, Pablo; Haddad, Roberto; Domizzi, Gladys

2003-01-01

During service, pressure tubes of CANDU nuclear power reactor are prone to suffer crack growth by delayed hydride cracking (DHC). For a given H 2 plus D 2 concentration there is a critical temperature (T c ) below which DHC may occur. In this work, T c was measured for CCT specimens cut from Zr-2.5 Wt % Nb pressure tubes. Hydrogen was added to the specimens to get concentrations of 40, 59 and 72 ppm. It was found that T c is higher than the corresponding precipitation temperature. The axial crack velocity (V p ) was also measured. Decreasing temperature from T c makes V p increase until a maximum is attained at a temperature close to precipitation temperature. At lower temperatures, in the presence of precipitated hydrides, decreasing temperature implies lower velocities, following an Arrhenius law: Vp=Aexp(-Q/RT), with an activation energy Q= 66 KJ/mol K. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

International Nuclear Information System (INIS)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

2011-01-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Ultimate pressure capacity of CANDU 6 containment structures

International Nuclear Information System (INIS)

Radulescu, J.P.; Pradolin, L.; Mamet, J.C.

1997-01-01

This paper summarizes the analytical work carried out and the results obtained when determining the ultimate pressure capacity (UPC) of the containment structures of CANDU 6 nuclear power plants. The purpose of the analysis work was to demonstrate that such containment structures are capable of meeting design requirements under the most severe accident conditions. For this concrete vessel subjected to internal pressure, the UPC was defined as the pressure causing through cracking in the concrete. The present paper deals with the overall behaviour of the containment. The presence of openings, penetrations and the ultimate pressure of the airlocks were considered separately. (author)

Visual inspection technology of the narrow and small confined area for monitoring feederpipe support of pressure tube in calandria reactor

Energy Technology Data Exchange (ETDEWEB)

Cho, Jae Wan; Lee, Nam Ho; Choi, Young Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

There are 760 feederpipes, which they are connected to inlet/outlet of the 380 pressure tube channels on the front of the calandria, in CANDU-type Reactor of Wolsung Nuclear Power Plant. As an ISI(In-Service Inspection) and PSI (Post-Service Inspection) requirements, maintenance activities of measuring the thickness of curvilinear part of feederpipe and inspecting the feederpipe support area within calandria are needed to ensure continued reliable operation of nuclear power plant. And ultrasonic probe is used to measure the thickness of curvilinear part of feederpipe, however workers are exposed to radioactivity irradiation during the measurement period. But, it is exposed to radioactivity irradiation during the measurement period. But, it is impossible to inspect feederpipe support area thoroughly because of narrow and confined accessibility, that is , an inspection space between the pressure tube channels is less than 100 mm and pipes in feederpipe support area are congested. And also, workers involved in inspecting feederpipe support area are under the jeopardy of high-level radiation exposure. Concerns about sliding home, which make the move of feederpipe connected to pressure tube channel smooth as pressure tube expands and contracts in its axial direction, stuck to feedeerpipe support and some of the structural components have made necessary the development of video inspection probe system with narrow and confined accessibility to observe and inspect feederpipe support area more close. Using video inspection probe system, it is possible to inspect and repair abnormality of feederpipe support connected to pressure tube channels of the calandria more accurate and quantative than naked eye. Therefore, that will do much for ensuring safety of CANDU-type nuclear power plant. 45 figs.,31 tabs. (Author)

Modeling of fuel bundle vibration and the associated fretting wear in a CANDU fuel channel

International Nuclear Information System (INIS)

Mohany, A.; Hassan, M.

2011-01-01

In this paper a numerical model is developed to predict the vibration response of a CANDU® fuel bundle and the associated fretting wear in the surrounding pressure tube. One excitation mechanism is considered in this model; turbulence-induced excitation caused by coolant flow inside the fuel channel. The numerical model can be easily adapted to include the effects of seismic events, fuel bundle impact during refuelling and start-up of the reactor, and the acoustic pressure pulsations caused by the primary heat transport (PHT) pumps. The simulation is performed for a typical CANDU fuel bundle with 37 fuel elements. The clearances between the buttons of the inner fuel elements, and between the bearing pads of the outer fuel elements and the pressure tube were measured from an actual fuel bundle. Some variability among the measured clearance values was observed. Therefore, probability density functions of the measured clearance values were established and the simulation was performed for the probabilistic distribution of the clearance values. The contact between the fuel bundle and the pressure tube is modeled using pseudo-force contact method. The proposed modelling technique can be used in future CANDU reactors to avoid fuel and pressure tube fretting damage due to the aforementioned excitation mechanisms. (author)

Prospects for stronger calandria tubes

International Nuclear Information System (INIS)

Ells, C.E.; Coleman, C.E.; Hosbons, R.R.; Ibrahim, E.F.; Doubt, G.L.

1990-12-01

The CANDU calandria tubes, made of seam welded and annealed Zircaloy-2, have given exemplary service in-reactor. Although not designed as a system pressure containment, calandria tubes may remain intact even in the face of pressure tube rupture. One such incident at Pickering Unit 2 demonstrated the economic advantage of such an outcome, and a case can be made for increasing the probability that other calandria tubes would perform in a similar fashion. Various methods of obtaining stronger calandria tubes are available, and reviewed here. When the tubes are internally pressurized, the weld is the weak section of the tube. Increasing the oxygen concentration in the starting sheet, and thickening the weld, are promising routes to a stronger tube

Economics of CANDU-PHW

International Nuclear Information System (INIS)

Jackson, H.A.; Woodhead, L.W.; Fanjoy, G.R.

1984-03-01

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

Economics of CANDU-PHW

International Nuclear Information System (INIS)

McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.

1982-03-01

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

Economics of CANDU-PHW

International Nuclear Information System (INIS)

Jackson, H.A.; Horton, E.P.; Woodhead, L.W.; Fanjoy, G.R.

1985-03-01

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

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

International Nuclear Information System (INIS)

Dolbey, M.P.; Kupcis, O.A.

1979-01-01

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 (10 6 Roentgens/hour)(R/h) radiation fields is reviewed. (author)

Aging of elastomers in CANDU pressure boundary service

International Nuclear Information System (INIS)

VanBerlo, C.; Leidner, J.

1987-09-01

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

Role of operator response guidelines in CANDU 9 design program

International Nuclear Information System (INIS)

Jaitly, R.K.

2000-01-01

The CANDU 9 is a large version of the CANDU Pressurized Heavy Water Reactor (PHWR) system developed in Canada. With an electrical output of approximately 935 MWe, the CANDU 9 complements the established mid-size CANDU 6 (700 MWe) and makes use of proven technology updated with state of the art features resulting from ongoing development. The CANDU 9 builds on the reactor and process system designs of the operating Darlington and Bruce B plants, and incorporates a modified CANDU 6 station layout, as well as improved construction methods and operational features. A high level of standardization has always been a feature of CANDU reactors. This theme is emphasized in the CANDU 9; all key components (reactor core, steam generators, coolant pumps, pressure tubes, etc.) are of the same design as those proven in service in the operating CANDU power stations. Including Probabilistic Safety Assessment (PSA) as part of the CANDU 9 design process from the outset of the program was seen as key to ensuring completeness of safety related requirements. The PSA work provided an in-depth understanding of the plant response to various postulated accidents. As well, the time frame for recovery and the related operator actions were identified. This information together with AECL's experience in supporting the development of Emergency Operating Procedures (EOPs) for the operating CANDU reactors are the basis for preparation of CANDU 9 Operator Response Guidelines (ORGs). Technical content, format and human factors considerations adopted for the ORGs are such that these can be readily converted to EOPs. The scope of ORGs includes generic as well as event specific ORGs. This dual approach is required to provide defense-in-depth. This paper describes the process used to prepare ORGs for the CANDU 9 reactor and discusses important benefits gained from the application of ORGs as input to the control center design and future preparation of the EOPs. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Jarron, D.; Trelinski, M.; Kretz, S. [Ontario Power Generation, Ajax, Ontario (Canada)]. E-mail: don.jarron@opg.com; mike.trelinski@opg.com; steve.kretz@opg.com

2006-07-01

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

Inspection of Candu Nuclear Reactor Fuel Channels

International Nuclear Information System (INIS)

Baron, J.; Jarvis, G.N.; Dolbey, M.P.; Hayter, D.M.

1986-01-01

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) [pt

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

International Nuclear Information System (INIS)

Lee, Jae Yong; Na, Man Gyun

2011-01-01

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

Establishment of Safety Analysis System and Technology for CANDU Reactors

International Nuclear Information System (INIS)

Park, Joo Hwan; Rhee, B. W.; Min, B. J.; Kim, H. T.; Kim, W. Y.; Yoon, C.; Chun, J. S.; Cho, M. S.; Jeong, J. Y.; Kang, H. S.

2007-06-01

The following 4 research items have been studied to establish a CANDU safety analysis system and to develop the relevant elementary technology for CANDU reactors. First, to improve and validate the CANDU design and operational safety analysis codes, the CANDU physics cell code WIMS-CANDU was improved, and validated, and an analysis of the moderator subcooling and pressure tube integrity has been performed for the large break LOCAs without ECCS. Also a CATHENA model and a CFD model for a post-blowdown fuel channel analysis have been developed and validated against two high temperature thermal-chemical experiments, CS28-1 and 2. Second, to improve the integrated operating system of the CANDU safety analysis codes, an extension has been made to them to include the core and fuel accident analyses, and a web-based CANDU database, CANTHIS version 2.0 was completed. Third, to assess the applicability of the ACR-7 safety analysis methodology to CANDU-6 the ACR-7 safety analysis methods were reviewed and the safety analysis methods of ACR-7 applicable to CANDU-6 were recommended. Last, to supplement and improve the existing CANDU safety analysis procedures, detailed analysis procedures have been prepared for individual accident scenarios. The results of this study can be used to resolve the CANDU safety issues, to improve the current design and operational safety analysis codes, and to technically support the Wolsong site to resolve their problems

Determination of delayed hydride cracking velocity of CANDU Zr-2.5Nb pressure tube

International Nuclear Information System (INIS)

Kim, Young Suk; Kim, Chan Jung; Rheem, Y. W.; Im, K. S.; Kwon, Sang Chul

2000-07-01

As agreed upon the contract with an IAEA Co-ordinated Research Project 'Hydrogen and Hydride Induced Degradation of the Mechanical and Physical Properties of Zirconium Based Alloys', we conducted DHC tests at 3 different temperatures of 144, 182 and 250 deg C on the curved compact tension specimens made from a Zr-2.5Nb pressure tube. Additional tests were carried out at 200 and 230 deg C with an aim to determine the activation energy for delayed hydride cracking. This report summarizes the results of DHC tests obtained so far. All the DHC tests were conducted in accordance with the procedures suggested by the Host Lab. 7 DHCV values determined at the same temperature such as 250 deg C show very low standard deviation, whose average values are very comparable to those reported by the participants. Thus, one of the most important results we have got is that we establish qualified DHC testing procedure through the IAEA CRP. An activation energy for DHC of unirradiated Zr-2.5Nb pressure tube was 49 KJ/mol which is very similar to the activation energy of 43 KJ/mol for irradiated Zr-2.5Nb pressure tubes. DHCV increased linearly with the hydrogen content up to around 25 ppm and then became saturated at higher hydrogen concentration

Flooding of a large, passive, pressure-tube LWR

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

DHC velocity comparison of CANDU Zr-2.5Nb pressure tube for the difference of test environments

International Nuclear Information System (INIS)

Cho, S. Y.; Kim, Y. S.; Oh, D. Z.; Yim, K. S.; Yim, Y. W.

2001-01-01

Zr-2.5Nb Pressure tube was used in the distilled water under high temperature and pressure. However, the evaluation of DHCV for pressure tube was limited in the air until now. Therefore, it was necessary for DHCV both in the air and in the distilled water under high temperature and pressure to evaluate. In advance, new DHC equipment simulating the real operating condition in the distilled water under high temperature and pressure was developed and DHCV test was conducted by this equipment. The test was carried out under simulated condition using distilled water of 250 .deg. C, 86bar and this result was compared with of DHCV in the air of 250 .deg. C. DHCV of the distilled water was ranged from 8.42x10 -8 to 9.92x 10 -8 m/s and the average value was 9.01x 10 -8 m/s. As compared with the air condition, it was found that characteristics of DHCV was not affected by the distilled water of high temperature and pressure. At the same temperature, DHCV of the irradiated pressure tube was faster than that of test result

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

International Nuclear Information System (INIS)

Nitheanandan, T.; Tiede, R.W.; Sanderson, D.B.; 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)

Development of thermal hydraulic evaluation code for CANDU reactors

International Nuclear Information System (INIS)

Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

2004-02-01

To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

Falling film flow, heat transfer and breakdown on horizontal tubes

International Nuclear Information System (INIS)

Rogers, J.T.

1980-11-01

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

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

International Nuclear Information System (INIS)

Rhee, B.-W.; Jeong, C.J.; Choi, J.H.; Yoo, S.-Y.

2002-01-01

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

Eddy current proximity measurement of perpendicular tubes from within pressure tubes in CANDU nuclear reactors

Science.gov (United States)

Bennett, P. F. D.; Underhill, P. R.; Morelli, J.; Krause, T. W.

2018-04-01

Fuel channels in CANDU® (CANada Deuterium Uranium) nuclear reactors consist of two non-concentric tubes; an inner pressure tube (PT) and a larger diameter calandria tube (CT). Up to 400 horizontally mounted fuel channels are contained within a calandria vessel, which also holds the heavy water moderator. Certain fuel channels pass perpendicularly over horizontally oriented tubes (nozzles) that are part of the reactor's liquid injection shutdown system (LISS). Due to sag, these fuel channels are at risk of coming into contact with the LISS nozzles. In the event of contact between the LISS nozzle and CT, flow-induced vibrations from within the moderator could lead to fretting and deformation of the CT. LISS nozzle proximity to CTs is currently measured optically from within the calandria vessel, but from outside the fuel channels. Measurement by an independent means would provide confidence in optical results and supplement cases where optical observations are not possible. Separation of PT and CT, known as gap, is monitored from within the PT using a transmit-receive eddy current probe. Investigation of the eddy current based gap probe as a tool to also measure proximity of LISS nozzles was carried out experimentally in this work. Eddy current response as a function of LISS-PT proximity was recorded. When PT-CT gap, PT wall thickness, PT resistivity and probe lift-off variations were not present this dependence could be used to determine the LISS-PT proximity. This method has the potential to provide LISS-CT proximity using existing gap measurement data. Obtaining LISS nozzle proximity at multiple inspection intervals could be used to provide an estimate of the time to LISS-CT contact, and thereby provide a means of optimizing maintenance schedules.

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

International Nuclear Information System (INIS)

Burrill, K.A.; Guzonas, D.A.

2002-01-01

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

CANDU safety under severe accidents

International Nuclear Information System (INIS)

Snell, V.G.; Howieson, J.Q.; Alikhan, S.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

1996-01-01

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

CANDU safety under severe accidents

Energy Technology Data Exchange (ETDEWEB)

Snell, V G; Howieson, J Q [Atomic Energy of Canada Ltd. (Canada); Alikhan, S [New Brunswick Electric Power Commission (Canada); Frescura, G M; King, F [Ontario Hydro (Canada); Rogers, J T [Carleton Univ., Ottawa, ON (Canada); Tamm, H [Atomic Energy of Canada Ltd. (Canada). Whiteshell Research Lab.

1996-12-01

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{sup -6}/year. 95 refs, 3 tabs.

CANDU operating experience

International Nuclear Information System (INIS)

McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.

1982-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

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

International Nuclear Information System (INIS)

Dinu, A.; Nedelcu, L.

1995-01-01

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

Heavy water cycle in the CANDU reactor

International Nuclear Information System (INIS)

Nanis, R.

2000-01-01

Hydrogen atom has two isotopes: deuterium 1 H 2 and tritium 1 H 3 . The deuterium oxide D 2 O is called heavy water due to its density of 1105.2 Kg/m 3 . Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D 2 O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D 2 O management is required to preserve it. (author)

Prediction of hydrogen distribution in the reactor building in CANDU6 plant

International Nuclear Information System (INIS)

Jin, Y.; Song, Y.

2008-01-01

The CANDU plants have a lot of zircaloy. The fuel cladding, calandria tubes and pressure tubes are made of zircaloy. The zircaloy can be oxidized and hydrogen is generated during severe accident progression. The detonation or deflagration to detonation transition (DDT) due to hydrogen combustion may occur if the local hydrogen concentration or global hydrogen concentration exceeds certain value. The detonation may result in the rupture of the reactor building. The inside of the reactor building of CANDU plants is complex. So prediction of hydrogen distribution in the reactor building is important. This prediction is made using ISAAC code and GOTHIC code. ISAAC code partitioned the reactor building in to 7 compartments. GOTHIC code modeled the CANDU6 reactor building using 12 nodes. The hydrogen concentrations in the various compartments in the reactor building are compared. GOTHIC code slightly underpredicts hydrogen concentration in the F/M rooms than ISAAC code, but trend is same. The hydrogen concentration in the boiler room and the moderator room shows almost same as for both codes. (author)

Pressure tube type reactors

International Nuclear Information System (INIS)

Komada, Masaoki.

1981-01-01

Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

Conceptual design of a pressure tube light water reactor with variable moderator control

International Nuclear Information System (INIS)

Rachamin, R.; Fridman, E.; Galperin, A.

2012-01-01

This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

Pressure tube reactor

International Nuclear Information System (INIS)

Susuki, Akira; Murata, Shigeto; Minato, Akihiko.

1993-01-01

In a pressure tube reactor, a reactor core is constituted by arranging more than two units of a minimum unit combination of a moderator sealing pipe containing a calandria tube having moderators there between and a calandria tube and moderators. The upper header and a lower header of the calandria tank containing moderators are communicated by way of the moderator sealing tube. Further, a gravitationally dropping mechanism is disposed for injecting neutron absorbing liquid to a calandria gas injection portion. A ratio between a moderator volume and a fuel volume is defined as a function of the inner diameter of the moderator sealing tube, the outer diameter of the calandria tube and the diameter of fuel pellets, and has no influence to intervals of a pressure tube lattice. The interval of the pressure tube lattice is enlarged without increasing the size of the pressure tube, to improve production efficiency of the reactor and set a coolant void coefficient more negative, thereby enabling to improve self controllability and safety. Further, the reactor scram can be conducted by injecting neutron absorbing liquid. (N.H.)

The influence of the preliminary garter spring spacer simulator clamping force in the pressure tube spacer -calandria tube hook-up simulator aging behaviour

International Nuclear Information System (INIS)

Gyongyosi, T.; Deloreanu, G.; Puiu, D.; Corbescu, B.; Anghel, N.; Dinu, E.

2016-01-01

The garter spring spacer is a specially constructed torsion spring used to fit-out the CANDU 6 fuel channel. The pressure tube ageing decreases the gap to the calandria tube. Continuous gap decrease directly affects the garter spring spacers behavior during fuel channel assembly operation. The preliminary clamping force value of the garter spring spacer assembly is important for its ageing behavior. This paper briefly describes the experimental technological facilities used for conducted the experiments and highlights some of the important moments during an experiment carried out in laboratory conditions, without using pressurized boiled water and irradiation working conditions. The results analysis and some conclusions are outlined at the end, pointing out that a garter spring spacer preliminary clamping force increase reduces the vibration response signal amplitude, and does not lead to its relaxation. The paper is dedicated to specialists working in research and technological engineering. (authors)

Assessment of leak detection capability of Candu 6 annulus gas system using moisture injection tests

International Nuclear Information System (INIS)

Nho, Ki Man; Kim, Wang Bae

1998-01-01

The Candu 6 reactor assembly consists of an array of 380 pressure tubes, which are installed horizontally in a large cylindrical vessel, the Calandria, containing the low pressure heavy water moderator. The pressure tube is located inside calandria tube and the annulus between these tubes, which forms a closed loop with CO 2 gas recirculating, is called the Annulus Gas System (AGS). It is designed to give an alarm to the operator even for a small pressure tube leak by a very sensitive dew point meter so that he can take a preventive action for the pressure tbe rupture incident. To judge whether the operator action time is enough or not in the design of Wolsung 2, 3, and 4, the Leak Before Break (LBB) assessment is required for the analysis of the pressure tube failure accident. In order to provide the required data for the LBB assessment of Wolsung Units 2, 3, 4, a series of leak detection capability tests was performed by injecting controlled rates of heavy water vapour. The data of increased dew point and rates of rise were measured to determine the alarm set point for dew point rate of rise of Wolsung Unit 2. It was found that the response of the dew point depends on the moisture injection rate, CO 2 gas flow rate and the leak location. The test showed that Candu 6 AGS can detect the very small leaks less than few g/hr and dew point rate of rise alarm can be the most reliable alarm signal to warn the operator. Considering the present results, the first response time of dew point to the AGS CO 2 flow rate is approximated. (author)

Moderator mixing after a pressure tube failure

International Nuclear Information System (INIS)

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

1997-01-01

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

The simulation of CANDU fuel channel behavior in thermal transient conditions

International Nuclear Information System (INIS)

Mihalache, M.; Roth, M.; Radu, V.; Dumitrescu, I.

2005-01-01

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

Fast fracture of a zirconium alloy pressure tube: cause and implications

International Nuclear Information System (INIS)

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

1985-12-01

The cause of the unstable fracture of a Zircaloy-2 pressure tube in the core of a CANDU reactor is reviewed. Failure was associated with the presence of brittle zones of zirconium hydride which developed as a result of thermal gradient induced hydrogen diffusion. Unstable fracture occurred when the partial thickness crack reached an unstable length and the crack ran 2 meters along the tube and terminated by circumferential tearing. The partial thickness defect initiated and propagated to an unstable length by delayed hydride cracking is high compared to fatigue progression and increases exponentially with temperature. Delayed hydride cracking can be prevented by reducing residual stresses to a minimum and by high standards of non-destructive testing that ensures freedom from unacceptable defects. Future prevention of fast fracture is based upon the inspection of a limited number of fuel channels for the presence of defects and for conditions which can cause hydride build-up together with the periodic removal of Zr-2.5wt% Nb tubes to monitor their condition

Economics of CANDU

International Nuclear Information System (INIS)

McConnell, L.G.; Woodhead, L.W.

1981-02-01

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

Pressure tube reactor

International Nuclear Information System (INIS)

Seki, Osamu; Kumasaka, Katsuyuki.

1988-01-01

Purpose: To remove the heat of reactor core using a great amount of moderators at the periphery of the reactor core as coolants. Constitution: Heat of a reactor core is removed by disposing a spontaneous recycling cooling device for cooling moderators in a moderator tank, without using additional power driven equipments. That is, a spontaneous recycling cooling device for cooling the moderators in the moderator tank is disposed. Further, the gap between the inner wall of a pressure tube guide pipe disposed through the vertical direction of a moderator tank and the outer wall of a pressure tube inserted through the guide pipe is made smaller than the rupture distortion caused by the thermal expansion upon overheating of the pressure tube and greater than the minimum gap required for heat shiels between the pressure tube and the pressure tube guide pipe during usual operation. In this way, even if such an accident as can not using a coolant cooling device comprising power driven equipment should occur in the pressure tube type reactor, the rise in the temperature of the reactor core can be retarded to obtain a margin with time. (Kamimura, M.)

Diagnostic Technology Development for Core Internal Structure in CANDU reactor

International Nuclear Information System (INIS)

Jung, Hyun Kyu; Cheong, Y. M.; Lee, Y. S. and others

2005-04-01

Degradation of critical components of nuclear power plants has become important as the operating years of plants increase. The necessity of degradation study including measurement and monitoring technology has increased continuously. Because the fuel channels and the neighboring sensing tubes and control rods are particularly one of the critical components in CANDU nuclear plant, they are treated as a major research target in order to counteract the possible problems and establish the counterplan for the CANDU reactor safety improvement. To ensure the core structure integrity in CANDU nuclear plant, the following 2 research tasks were performed: Development of NDE technologies for the gap measurement between the fuel channels and LIN tubes. Development of vibration monitoring technology of the fuel channels and sensing tubes. The technologies developed in this study could contribute to the nuclear safety and estimation of the remaining life of operating CANDU nuclear power plants

Assessments of long term mechanical behavior of CANDU fuel channel by means of PFEM analysis

International Nuclear Information System (INIS)

Florea, S.; Pavelescu, M.

2005-01-01

Structural analysis with finite elements method is today a usual way to evaluate and predict the behavior of structural assemblies exposed to severe conditions, in order to ensure their safety end reliability. CANDU 600 fuel channel is an example in which long time irradiation with implicit consequences on material properties evolution interfere with the corrosion and thermal aggression. A high degree of uncertainty in the evolution of the material's properties must be considered. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods, in order to predict the structural components response. In INR (Institute of Nuclear Research) in the past years, a code for thermo-mechanical analysis of the fuel channel from CANDU 600 nuclear power plant of Cernavoda was developed using finite element methods (FEM). The CANTUP code evaluate the stress and strain state of the mechanical assembly of fuel channel, considered to be divided into pressure tube, calandria tube and four spacers, supposed to be equidistantly distributed along the pressure tube. The main achievement obtained with this code was the prediction of the long-term behavior of the sag of the pressure tube, by analysis in which the creep phenomenon and the contact between the spacers and calandria tube were considered. This reason has sustained the attempt to estimate the possibility to use this code in order to perform probabilistic evaluations. (authors)

Assessments of long term mechanical behavior of CANDU fuel channel by means of PFEM analysis

International Nuclear Information System (INIS)

Florea, S.; Pavelescu, M.

2005-01-01

Full text: Structural analysis with finite elements method is today a usual way to evaluate and predict the behavior of structural assemblies exposed to severe conditions, in order to ensure their safety end reliability. CANDU 600 fuel channel is an example in which long time irradiation with implicit consequences on material properties evolution interfere with the corrosion and thermal aggression. A high degree of uncertainty in the evolution of the material's properties must be considered. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods, in order to predict the structural components response. In INR (Institute of Nuclear Research) in the past years, a code for thermo-mechanical analysis of the fuel channel from CANDU 600 nuclear power plant of Cernavoda was developed using finite element methods (FEM). The CANTUP code evaluate the stress and strain state of the mechanical assembly of fuel channel, considered to be divided into pressure tube, calandria tube and four spacers, supposed to be equidistantly distributed along the pressure tube. The main achievement obtained with this code was the prediction of the long-term behavior of the sag of the pressure tube, by analysis in which the creep phenomenon and the contact between the spacers and calandria tube were considered. This reason has sustained the attempt to estimate the possibility to use this code in order to perform probabilistic evaluations. (authors)

AECL's advanced CANDU reactor - the ACR

International Nuclear Information System (INIS)

Alizadeh, Ala; Allsop, Peter; Hedges, Ken; Hopwood, Jerry; Yu, Stephen

2003-01-01

The ACR, the next generation CANDU design, represents the next step in development of the CANDU family of designs. AECL has achieved significant incremental improvements to the mid-size CANDU 6 nuclear power plant through successive projects, both in design and in project delivery. Building on this knowledge base, AECL is continuing to adapt the CANDU design to develop the ACR. This paper summarizes the ACR design features, which include major improvements in economics, inherent safety characteristics, performance and construction methods. Aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs, the ACR is an evolutionary design based on the very successful CANDU 6 reactor. The new ACR product is specifically designed to produce power at a cost competitive with other forms of power generation while achieving short construction times, improved safety, international licensability, high investor returns, and low investor risk. It achieves these targets by taking advantage of the latest advances in both pressure-tube and pressure-vessel reactor technologies and experience. The flexibility and development potential of the fuel channel approach also enables designs to be developed that address priorities identified in international long-term specification programs such as the US Department of Energy (DOE) sponsored Generation IV program and IAEA hosted INPRO program. ACR-700 can be built in 36 months with a 48 month project duration, and deliver a lifetime capacity factor in excess of 90%. Overall, the ACR design represents a balance of proven design basis and innovations to give step improvements in safety, reliability and economics. The ACR development program, now well into the detail design stage, includes parallel formal licensing in the USA and Canada. Based on the status of the ACR design and AECL's on-going experience delivering reactor projects on-time and on-budget, the first ACR could be in service by

Leak detection capability in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

Leak detection capability in CANDU reactors

International Nuclear Information System (INIS)

Azer, N.; Barber, D.H.; Boucher, P.J.

1997-01-01

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

Improving the calandria tubes for CANDU reactors

International Nuclear Information System (INIS)

Coleman, C.E.; Fong, R.W.L.; Doubt, G.L.; Nitheanandan, T.; Sanderson, D.B.

1997-07-01

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

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

International Nuclear Information System (INIS)

Lazaro, Pavel Gabriel; Balas Ghizdeanu, Elena Nineta

2008-01-01

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

Conceptual designs for advanced, high-temperature CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Bushby, S.J. [Atomic Energy of Canada Ltd., Corrosion and Surface Science Branch, Chalk River Laboratories, Chalk River, ON (Canada); Dimmick, G.R. [Atomic Energy of Canada Ltd., Fuel Channel Thermmalhydraulics Branch, Chalk River, ON (Canada); Duffey, R.B. [Atomic Energy of Canada Ltd., Principal Scientist, Chalk River Laboratories, Chalk River, On (Canada); Spinks, N.J. [Atomic Energy of Canada Ltd., Researcher Emeritus, Chalk River Laboratories, Chalk River, ON (Canada); Burrill, K.A. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, ON (Canada); Chan, P.S.W. [Atomic Energy of Canada Ltd., Reactor Core Physics Branch, Mississauga, ON (Canada)

2000-07-01

AECL is studying advanced reactor concepts with the aim of significant cost reduction through improved thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, also incorporates enhanced safety features, and flexible, proliferation-resistant fuel cycles, whilst retaining the fundamental design characteristics of CANDU: neutron economy, horizontal fuel channels, and a separate D{sub 2}O moderator that provides a passive heat sink. Where possible, proven, existing components and materials would be adopted, so that 'first-of-a-kind' costs and uncertainties are avoided. Three reactor concepts ranging in output from {approx}375 MW(e) to 1150 MW(e) are described. The modular design of a pressure tube reactor allows the plant size for each concept to be tailored to a given market through the addition or removal of fuel channels. Each concept uses supercritical water as the coolant at a nominal pressure of 25 MPa. Core outlet temperatures range from {approx}400degC to 625degC, resulting in substantial improvements in thermodynamic efficiencies compared to current nuclear stations. The CANDU-X Mark 1 concept is an extension of the present CANDU design. An indirect cycle is employed, but efficiency is increased due to higher coolant temperature, and changes to the secondary side; as well, the size and number of pumps and steam generators are reduced. Safety is enhanced through facilitation of thermo-siphoning of decay heat by increasing the temperature of the moderator. The CANDU-X NC concept is also based on an indirect cycle, but natural convection is used to circulate the primary coolant. This approach enhances cycle efficiency and safety, and is viable for reactors operating near the pseudo-critical temperature of water because of large changes in heat capacity and thermal expansion in that region. In the third concept (CANDUal-X), a dual cycle is employed. Supercritical water exits the core and feeds directly into a very high

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

International Nuclear Information System (INIS)

Marinescu, R.; Mincu, M.

2015-01-01

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

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

International Nuclear Information System (INIS)

1991-11-01

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

DELOCA, a code for simulation of CANDU fuel channel in thermal transients

International Nuclear Information System (INIS)

Mihalache, M.; Florea, Silviu; Ionescu, V.; Pavelescu, M.

2005-01-01

Full text: In certain LOCA scenarios into the CANDU fuel channel, the ballooning of the pressure tube and the contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator arises through the contact area. If the temperature of channel walls increases, the contact area is drying, the heat transfer becomes inefficiently and the fuel channel could lose its integrity. DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after the contact between the two tubes. The code contains a few models: the creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This code was systematically verified by Contact1 and Cathena codes. This paper presents the results obtained at different temperature increasing rates. In addition, the contact moment for a RIH 5% postulated accident was calculated. The Cathena thermo-hydraulic code provided the input data. (authors)

DELOCA, a code for simulation of CANDU fuel channel in thermal transients

International Nuclear Information System (INIS)

Mihalache, M.; Florea, Silviu; Ionescu, V.; Pavelescu, M.

2005-01-01

In certain LOCA scenarios into the CANDU fuel channel, the ballooning of the pressure tube and the contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator arises through the contact area. If the temperature of channel walls increases, the contact area is drying, the heat transfer becomes inefficiently and the fuel channel could lose its integrity. DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after the contact between the two tubes. The code contains a few models: the creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This code was systematically verified by Contact1 and Cathena codes. This paper presents the results obtained at different temperature increasing rates. In addition, the contact moment for a RIH 5% postulated accident was calculated. The Cathena thermo-hydraulic code provided the input data. (authors)

Ontario Hydro CANDU operating experience

International Nuclear Information System (INIS)

Bartholomew, R.W.; Woodhead, L.W.; Horton, E.P.; Nichols, M.J.; Daly, I.N.

1987-01-01

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

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

International Nuclear Information System (INIS)

Trujillo Badillo, Giovanna; Desimone, Carlos; Domizzi, Gladys

1999-01-01

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

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

International Nuclear Information System (INIS)

Krause, T.W.; Schankula, J.; Sullivan, S.P.

2002-01-01

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

Assessment of a Pressure Tube Rupture with a Poisoned Moderator

International Nuclear Information System (INIS)

Kim, S. R.; Kim, B. G.; Kim, S. C.; Kim, E. K.

2005-01-01

The postulated in-core LOCA has been analyzed and evaluated while the reactor is operating normally with a low moderator poison concentration for CANDU. However, when the reactor is operating with a relatively large amount of boron and/or gadolinium poison in the moderator, an assessment of the fuel integrity was required for the pressure tube rupture (PTR) accident. Poisoned moderator exists mainly during a startup after a prolonged shutdown lasting for more than one day. For the case of a reactor regulating system (RRS) working, the methodology of the PTR assessment with a poisoned moderator has been developed to determine the effective trip parameters, evaluate the fuel integrity, and establish the standard reactor start-up model for the Wolsong Nuclear Power Plants recently. The developed methodology and results are presented

Molten Fuel Mass Assessment for Channel Flow Blockage Event in CANDU6

International Nuclear Information System (INIS)

Lee, Kwang Ho; Kim, Yong Bae; Choi, Hoon; Park, Dong Hwan

2011-01-01

In CANDU6, a fuel channel flow blockage causes a sudden reduction of flow through the blocked channel. Depending on the severity of the blockage, the reduced flow through the channel can result in severe heat up of the fuel, hence possibly leading to pressure tube and calandria tube failure. If the calandria tube does not fail the fuel and sheath would continue to heat up, and ultimately melting could occur. Eventually, molten material runs down onto the pressure tube. Even a thin layer of molten material in contact with the pressure tube causes the pressure tube and calandreia tube to heat up rapidly. The thermal transient is so rapid that failure temperatures are reached quickly. After channel failure, the contents of the channel, consisting of superheated coolant, fission products and possibly overheated of molten fuel, are rapidly discharged into the moderator. Fuel discharged into the moderator is quenched and cooled. The rapid discharge of hot fuel and coolant into the calandria causes the moderator pressure and temperature to increase, which may cause damage to some in-core components. Thus, the assessment results of molten fuel mass are inputs to the in-core damage analysis. In this paper, the analysis methodology and results of molten fuel mass assessment for the channel flow blockage event are presented

Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

International Nuclear Information System (INIS)

Park, S. Y.; Song, Y. M.

2015-01-01

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

Predicting the in-reactor mechanical behavior of Zr-2.5Nb pressure tubes from postirradiation microstructural examination data

International Nuclear Information System (INIS)

Griffiths, M.; Davies, P.H.; Davies, W.G.; Sagat, S.

2002-01-01

Postirradiation microstructure examinations of Zr-2.5Nb pressure tubes removed from service in CANDU reactors have shown clear trends in the dislocation structure and the state of the β-phase, as a function of operating temperature, neutron flux, and time. These microstructural parameters correlate well with changes in the mechanical properties. For example, the rapid increase in dislocation loop density in the early stages of irradiation corresponds with a rapid increase in tensile strength and DHC velocity, and a corresponding decrease in fracture toughness. There is also a strong negative correlation between the degree of β-phase decomposition and DHC velocity. In addition to the effects of microstructure evolution on the mechanical properties, changes in the a-type and c-component dislocation loop densities also affect irradiation deformation (creep and growth). Statistical analyses of the irradiation microstructure data have been used to derive empirical relationships between dislocation densities and β-phase structure with temperature, flux, and time. The relationships thus derived are useful in predicting where the mechanical properties are most affected by the in-reactor operating conditions. The predictions are compared with mechanical test data for samples from various axial and circumferential locations of 42 pressure tubes removed from operating CANDU reactors. The results are discussed in terms of the mechanisms controlling tensile strength, fracture, delayed-hydride-cracking, and in-reactor deformation. (author)

Ontario Hydro CANDU operating experience

International Nuclear Information System (INIS)

Jackson, H.A.; Woodhead, L.W.; Fanjoy, G.R.

1984-03-01

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

Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

International Nuclear Information System (INIS)

Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

1983-01-01

A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

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

International Nuclear Information System (INIS)

Ardron, K.H.; Krishnan, V.S.; McGee, G.R.; Anderson, J.W.D.; Hawley, E.H.

1984-07-01

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

Prediction of Axial and Radial Creep in CANDU 6 Pressure Tubes

International Nuclear Information System (INIS)

Radu, Vasile S.

2013-01-01

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

Improved containment isolation for CANDU plants

Energy Technology Data Exchange (ETDEWEB)

Stretch, A H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

The publication of Regulatory Policy Statement R- 7 in February 1991 by the Atomic Energy Control Board imposes new requirement for the isolation of fluid piping penetrating the containment boundary. The Appendix of R-7 describes the detailed requirements for metal extensions of the containment envelope, including the code classification qf the pressure retaining portions and isolation requirements for fluid piping and tubing.The application of these new requirements to the existing CANDU 6 design resulted in a number of design changes, including the addition of manual and automatic isolation valves, changes in code classification, and leakage monitoring considerations. (author). 3 refs., 3 figs.

Basic research and industrialization of CANDU advanced fuel

Energy Technology Data Exchange (ETDEWEB)

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

2000-04-01

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

Basic research and industrialization of CANDU advanced fuel

International Nuclear Information System (INIS)

Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su

2000-04-01

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

Candu 6: versatile and practical fuel technology

International Nuclear Information System (INIS)

Hopwood, J. M.; Saroudis, J.

2013-01-01

CANDU reactor technology was originally developed in Canada as part of the original introduction of peaceful nuclear power in the 1960s and has been continuously evolving and improving ever since. The CANDU reactor system was defined with a requirement to be able to efficiently use natural uranium (NU) without the need for enrichment. This led to the adaptation of the pressure tube approach with heavy water coolant and moderator together with on-power fuelling, all of which contribute to excellent neutron efficiency. Since the beginning, CANDU reactors have used [NU] fuel as the fundamental basis of the design. The standard [NU] fuel bundle for CANDU is a very simple design and the simplicity of the fuel design adds to the cost effectiveness of CANDU fuelling because the fuel is relatively straightforward to manufacture and use. These characteristics -- excellent neutron efficiency and simple, readily-manufactured fuel -- together lead to the unique adaptability of CANDU to alternate fuel types, and advancements in fuel cycles. Europe has been an early pioneer in nuclear power; and over the years has accumulated various waste products from reactor fuelling and fuel reprocessing, all being stored safely but which with passing time and ever increasing stockpiles will become issues for both governments and utilities. Several European countries have also pioneered in fuel reprocessing and recycling (UK, France, Russia) in what can be viewed as a good neighbor policy to make most efficient use of fuel. The fact remains that CANDU is the most fuel efficient thermal reactor available today [NU] more efficient in MW per ton of U compared to LWR's and these same features of CANDU (on-power fuelling, D 2 O, etc) also enable flexibility to adapt to other fuel cycles, particularly recycling. Many years of research (including at ICN Pitesti) have shown CANDU capability: best at Thorium utilization; can use RU without re-enrichment; can readily use MOX. Our premise is that

Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

International Nuclear Information System (INIS)

Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

2009-01-01

A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

In-situ optical profilometry of CANDU fuel channels

International Nuclear Information System (INIS)

Jarvis, G.N.; Cornblum, E.O.; Grabish, M.G.

1996-01-01

Detailed knowledge of flaw geometry is crucial in the stress analysis of flaws found in the thin-walled Zirconium alloy pressure tubes of CANDU reactors. While ultrasonic inspection can provide much of the required data, the measurement of the sharpness, or root-radius, at the bottom of a flaw has not so far been possible in-situ. This paper will describe the application of optical profilometry techniques, to measure directly the depth and root-radius of open inside-surface flaws, within a flooded reactor pressure tube. The tool uses a rad-tolerant television camera, custom optics and light stripe generators to collect digitized image data from three different views of a flaw. Software has been developed to manage the collection of the image data and provide a full range of display and automated analysis options. The tool has recently been used successfully to measure fretting flaws in the 100--250 micron deep range

Pressure tube type research reactor

International Nuclear Information System (INIS)

Ueda, Hiroshi.

1976-01-01

Object: To prevent excessive heat generation due to radiation of a pressure tube vessel. Structure: A pressure tube encasing therein a core comprises a dual construction comprising inner and outer tubes coaxially disposed. High speed cooling water is passed through the inner tube for cooling. In addition, in the outer periphery of said outer tube there is provided a forced cooling tube disposed coaxially thereto, into which cooling fluid, for example, such as moderator or reflector is forcibly passed. This forced cooling tube has its outer periphery surrounded by the vessel into which moderator or reflector is fed. By the provision of the dual construction of the pressure tube and the forced cooling tube, the vessel may be prevented from heat generation. (Ikeda, J.)

Development of CANDU advanced fuel bundle

International Nuclear Information System (INIS)

Suk, H. C.; Hwang, W.; Rhee, B. W.; Jung, S. H.; Chung, C. H.

1992-05-01

This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor for 1996 and 1997, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year include the detail design of CANFLEX fuel with natural enriched uranium (CANFLEX-NU). Based on this design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel in the CANDU Cold Test Loop to investigate the condition under which maximum pressure drop occurs and the maximum value of the bundle pressure drop. (Author)

Enhanced CANDU 6 design assist probabilistic safety assessment results and insights

International Nuclear Information System (INIS)

Torabi, T.; Bettig, R.; Iliescu, P.; Robinson, J.; Santamaura, P.; Skorupska, B.; Tyagi, A.K.; Vencel, I.

2013-01-01

The Enhanced CANDU 6(EC6) is a 700 MWe reactor, which has evolved from the well-established CANDU line of reactors, which are heavy-water moderated, and heavy-water cooled horizontal pressure tube reactors, using natural uranium fuel. The EC6 design retains the generic CANDU design features, while incorporating innovations and state-of-the-art technologies to ensure competitiveness with other design with respect to operation, performance and economics. A design assist probabilistic safety assessment (PSA) was conducted during the design change phase of the project. The purpose of the assessment was to assess internal events during at-power operation and identify the design improvements and additional features needed to comply with the latest regulatory requirements in Canada and compete with other reactor designs, internationally. The PSA results show that the EC6 plant response to the postulated initiating events is well balanced, and the design meets its safety objectives. This paper summarizes the results and insights gained during the development of the PSA models for at-power internal events. (author)

2D modeling of moderator flow and temperature distribution around a single channel after pressure tube/calandria tube contact

International Nuclear Information System (INIS)

Behdadi, A.; Luxat, J.C.

2009-01-01

A 2D computational fluid dynamics (CFD) model has been developed to calculate the moderator velocity field and temperature distribution around a single channel inside the moderator of a CANDU reactor after a postulated ballooning deformation of the pressure tube (PT) into contact with the calandria tube (CT). Following contact between the hot PT and the relatively cold CT, there is a spike in heat flux to the moderator surrounding the CT which may lead to sustained CT dryout. This can detrimentally affect channel integrity if the CT post-dryout temperature becomes sufficiently high to result in thermal creep strain deformation. The present research is focused on establishing the limits for dryout occurrence on the CTs for the situation in which pressure tube-calandria tube contact occurs. In order to consider different location of the channels inside the calandria, both upward and downward flow directions have been analyzed. The standard κ - ε turbulence model associated with logarithmic wall function is applied to predict the effects of turbulence. The governing equations are solved by the finite element software package COMSOL. The buoyancy driven natural convection on the outer surface of a CT has been analyzed to predict the flow and temperature distribution around the single CT considering the local moderator subcooling, wall temperature and heat flux. The model also shows the effect of high CT temperature on the flow and subcooling around the CTs at higher/lower elevation depending on the flow direction in the domain. According to the flow pattern and temperature distribution, it is predicted that stable film boiling generates in the stagnation region on the cylinder. (author)

New flux detectors for CANDU 6 reactors

International Nuclear Information System (INIS)

Cuttler, J.M.; Medak, N.

1992-06-01

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

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

International Nuclear Information System (INIS)

Scott Hower; Luiza Vladu; Adrian Nichisov; Mihai Cretu

2006-01-01

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

CANDU in the next century

International Nuclear Information System (INIS)

Meneley, D.A.; Torgerson, D.F.

1997-01-01

AECL's main product line is available today in two designs, designated as CANDU 6 and CANDU 9. Each of these is based on successfully operating pressurized-heavy-water nuclear plants. Several new CANDU stations are under construction or planned around the world. The author presents plant concepts which may evolve from today's products during the 21st century, indicating the particular development directions which might be followed by AECL product development depending on the future competitive environment, economics, and market circumstances. This study shows that the CANDU energy supply system is sufficiently flexible to be adapted into widely varying circumstances over the next century and beyond

CANDU steam generator life management: laboratory data and plant experience

International Nuclear Information System (INIS)

Tapping, R.L.; Nickerson, J.H.; Subash, N.; Wright, M.D.

2001-10-01

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

Localization of CANDU technology

Energy Technology Data Exchange (ETDEWEB)

Alizadeh, Ala

2010-09-15

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

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

International Nuclear Information System (INIS)

Park, T-K.; Jung, S-H.

1996-01-01

During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D 2 O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

Participation of INR to the research project initiated by the CANDU Owners Group - Canada

International Nuclear Information System (INIS)

Roth, Maria; Ciocanescu, Marin; Gheorghiu, C-tin; Ohai, Dumitru; Visinescu, Doru; Ionescu, Silviu; Man, Ion; Pitigoi, Vasile; Anghel, Dumitru; Uta, Octavian

2004-01-01

Having in view the participation of the INR's Nuclear Materials and Corrosion Departments together with Chalk River Laboratories, AECL, and other nine well known institutes, to the Research Project coordinated by the International Atomic Energy Agency - Vienna and the results obtained as well, the CANDU Owners Group (COG) - Canada has proposed an economical contract to INR - Pitesti, called the DHC Project. The goal of this project was the investigation of the slow cracking in the hydration state (Delayed Hydride Cracking phenomenon) of the pressure tubes in CANDU reactors. There were determined, in specific conditions, the DHC rate, V DHC , and threshold factor of triggering the phenomenon K HI (the factor of stress intensity in the crack) occurring in un-irradiated Zr-2.5%Nb alloy. Further extensions of the project to irradiated alloys of the pressure tubes used in CANDU reactors, provided by the Canadian partner, are planned. This paper gives an overview covering the contributions of the INR's departments to the contract provisions for the current year as concluded with COG - Canada. It is described the design and execution activity of a device for processing the samples subject to testing, the mounting of the measuring chain, developing of a system for monitoring the testing parameters, the acquisition and automated processing of experimental data up to constituting the final report. The testing report was appreciated by the COG and recognized as demonstrating the INR's capability to fulfil the tasks of the DHC Project and so opened the way to further cooperation

Verification tests for GRAD, a computer program to predict nonuniform deformation and failure of Zr-2.5 wt percent Nb pressure tubes during a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Shewfelt, R.S.W.; Godin, D.P.

1985-03-01

During a postulated loss-of-coolant accident in a CANDU reactor, the temperature of the pressure tubes could rise sufficiently so that ballooning could occur. It is also likely that there would be a variation in temperature around the tube circumference, causing the deformation to be nonuniform. Since the deformation of the pressure tube controls how the core heat is transferred to the surrounding moderator, which is a large heat sink, a computer program, GRAD, has been developed to predict this nonuniform deformation. Numerous biaxial creep tests were done, where the temperature of internally pressurized sections of Zr-2.5 wt percent Nb pressure tubes were ramped to check the ability of GRAD to predict the resulting nonuniform deformation and possible tube failure. GRAD was successful in predicting the average transverse creep strain observed during the tests and the local transverse creep strain at the end of the tests. GRAD was also able to predict the failure time and average transverse creep strain at failure for all the specimens that failed

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.

CANDU fuel behaviour under transient conditions

International Nuclear Information System (INIS)

Segel, A.W.L.

1979-04-01

The Canadian R and D program to understand CANDU fuel behaviour under transient conditions is described. Fuel sheath behaviour studies have led to the development of a model of transient plastic strain in inert gas, which integrates the deformation due to several mechanisms. Verification tests demonstrated that on average the model overpredicts strain by 20%. From oxidation kinetics studies a sheath failure embrittlement criterion based on oxygen distribution has been developed. We have also established a rate equation for high-temperature stress-dependent crack formation due to embrittlement of the sheath by beryllium. An electric, simulated fuel element is being used in laboratory tests to characterize the behaviour of fuel in the horizontal. In-reactor, post-dryout tests have been done for several years. There is an axially-segmented, axisymmetric fuel element model in place and a fully two-dimensional code is under development. Laboratory testing of bundles, in its early stages, deals with the effects of geometric distortion and sheath-to-sheath interaction. In-reactor, post-dryout tests of CANDU fuel bundles with extensive central UO 2 melting did not result in fuel fragmentation nor damage to the pressure tube. (author)

Mixed convection around calandria tubes in a ¼ scale CANDU-6 moderator circulation tank

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-15

Highlights: • A secondary jet is formed at a stagnation region and is directed towards the center of the MCT. • The secondary jet undergoes the significant dissipation and mixing due to calandria tubes (CTs). • Its cooling effectiveness is reduced on the CTs in the bottom of the MCT. • With forced convection dominance, peak heat transfer is on the upper CT surface. • With natural convection dominance, peak heat transfer is on the lower CT surface. - Abstract: This study experimentally characterizes mixed convection around calandria tubes (CTs) in a ¼ scale CANDU-6 moderator circulation tank (MCT) that uses air as the working fluid. In a full scale CANDU-6 reactor that undergoes a postulated dual failure with a loss-of-coolant accident without the emergency core cooling system available, mixed convection heat transfer occurs around the CTs. The cooling effectiveness of the moderator is diminished as an emergency heat sink if overheating eventually leads to film boiling. To prevent the onset of film boiling, local sub-cooling margins of the moderator needs to be maintained or else the critical heat flux should be increased. Circulating the moderator which interacts with the overheated CTs increases the heat transfer into the moderator which may suppress film boiling. The present experimental results demonstrate that the cooling effectiveness of the circulating moderator, in particular the secondary jet, is attenuated substantially as it is convected away from the inner wall towards the center of the MCT. The momentum of the secondary jet is diffused through the CTs. At a low jet Reynolds number, the secondary jet becomes ineffective so that some overheated CTs positioned in the other half of the MCT are cooled only by natural convection.

Mathematical modeling of CANDU-PHWR

Energy Technology Data Exchange (ETDEWEB)

Gaber, F.A.; Aly, R.A.; El-Shal, A.O. [Atomic Energy Authority, Cairo (Egypt)

2003-07-01

The paper deals with the transient studies of CANDU 600 pressurized Heavy Water Reactor (PHWR). This study involved mathematical modeling of CANDU-PHWR to study its thermodynamic performances. Modeling of CANDU-PHWR was based on lumped parameter technique. The reactor model includes the neutronic, reactivity, and fuel channel heat transfer. The nuclear reactor power was modelled using the point kinetics equations with six groups of delayed neutrons and the reactivity feed back due to the changes in the fuel temperature and coolant temperature. The CANDU-PHWR model was coded in FORTRAN language and solved by using a standard numerical technique. The adequacy of the model was tested by assessing the physical plausibility of the obtained results. (author)

Pre-service proof pressure and leak rate tests for the Qinshan CANDU project reactor buildings

International Nuclear Information System (INIS)

Petrunik, K.J.; Khan, A.; Ricciuti, R.; Ivanov, A.; Chen, S.

2003-01-01

The Qinshan CANDU Project Reactor Buildings (Units 1 and 2) have been successfully tested for the Pre-Service Proof Pressure and Integrated Leak Rate Tests. The Unit 1 tests took place from May 3 to May 9, 2002 and from May 22 to May 25, 2002, and the Unit 2 tests took place from January 21 to January 27, 2003. This paper discusses the significant steps taken at minimum cost on the Qinshan CANDU Project, which has resulted in a) very good leak rate (0.21%) for Unit 1 and excellent leak rate (0.130%) for Unit 2; b) continuous monitoring of the structural behaviour during the Proof Pressure Test, thus eliminating any repeat of the structural test due to lack of data; and c) significant schedule reduction achieved for these tests in Unit 2. (author)

Method of repairing pressure tube type reactors

International Nuclear Information System (INIS)

Asada, Takashi.

1983-01-01

Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

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

Science.gov (United States)

Wang, Xinyi; Shen, Jialong; Liu, Xinbo

2018-01-01

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

Annular gap measurement between pressure tube and calandria tube by eddy current technique

International Nuclear Information System (INIS)

Bhole, V.M.; Rastogi, P.K.; Kulkarni, P.G.

1992-01-01

In pressurised heavy water reactor (PHWR) major distinguishing feature is that there are number of identical fuel channels in the reactor core. Each channel consists of pressure tube of Zr-2.5 Nb or zircaloy-2 through which high temperature, high pressure primary coolant is passing. The pressure tube contains fuel. Surrounding the pressure tube there is low pressure, cool heavy water (moderator). The moderator is thermally separated from coolant by the tube which is nominally concentric with pressure tube called calandria tube. There are four garter springs in the annular gap between pressure tube and calandria tube. During the life of the reactor there are number of factors by which the pressure tube sags, most important factors are irradiation creep, thermal creep, fuel load etc. Because of the sag of pressure tube it can touch the calandria tube resulting in formation of cold spot. This leads to hydrogen concentration at that spot by which the material at that place becomes brittle and can lead to catastrophic failure of pressure tube. There is no useful access for measurement of annular gap either through the gas annular space or from exterior of calandria tube. So the annular gap was measured from inside surface of pressure tube which is accessible. Eddy current technique was used for finding the gap. The paper describe the details of split coil design of bobbin probe, selection of operating point on normalised impedance diagram by choosing frequency. Experimental results on full scale mock up, and actual gap measurement in reactor channel, are also given. (author). 7 figs

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

International Nuclear Information System (INIS)

Kim, Yonghee; Roh, Gyuhong; Kim, Won Young; Kim, Hak Sung; Park, Joo Hwan

2008-01-01

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

Passive heat transport in advanced CANDU containment

International Nuclear Information System (INIS)

Krause, M.; Mathew, P.M.

1994-01-01

A passive CANDU containment design has been proposed to provide the necessary heat removal following a postulated accident to maintain containment integrity. To study its feasibility and to optimize the design, multi-dimensional containment modelling may be required. This paper presents a comparison of two CFD codes, GOTHIC and PHOENICS, for multi-dimensional containment analysis and gives pressure transient predictions from a lumped-parameter and a three-dimensional GOTHIC model for a modified CANDU-3 containment. GOTHIC proved suitable for multidimensional post-accident containment analysis, as shown by the good agreement with pressure transient predictions from PHOENICS. GOTHIC is, therefore, recommended for passive CANDU containment modelling. (author)

Reactor physics assessment of modified 37-element CANDU fuel bundles

International Nuclear Information System (INIS)

Pristavu, R.; Rizoiu, A.

2016-01-01

Reducing the central element diameter in order to improve the total flow area of CANDU fuel bundle and redistribute the power density of all remaining elements was studied in Canada and Korea when considering the effect of aging pressure tube diametral creep. The aim of this paper is to study the modified bundle behavior using the transport codes WIMS and DRAGON. In calculations, a WIMS nuclear data library on 172 energy groups was used. 2-D transport calculations were performed with WIMS and DRAGON, leading to similar results in estimated cell parameters. Additionally, 3-D DRAGON calculations were carried on in order to evaluate the local flux distribution shift, as well as the incremental cross sections for supercells containing modified CANDU bundles and reactivity devices. The overall effect of using modified fuel bundles was meaningless for both cell and supercell parameters, thus ensuring this possibility of fuel improvement for thermal-hydraulic purposes only. (authors)

Recent IAEA activities on CANDU-PHWR fuels and fuel cycles

International Nuclear Information System (INIS)

Inozemtsev, V.; Ganguly, C.

2005-01-01

Pressurized Heavy Water Reactors (PHWR), widely known as CANDU, are in operation in Argentina, Canada, China, India, Pakistan, Republic of Korea and Romania and account for about 6% of the world's nuclear electricity production. The CANDU reactor and its fuel have several unique features, like horizontal calandria and coolant tubes, on-power fuel loading, thin-walled collapsible clad coated with graphite on the inner surface, very high density (>96%TD) natural uranium oxide fuel and amenability to slightly enriched uranium oxide, mixed uranium plutonium oxide (MOX), mixed thorium plutonium oxide, mixed thorium uranium (U-233) oxide and inert matrix fuels. Several Technical Working Groups (TWG) of IAEA periodically discuss and review CANDU reactors, its fuel and fuel cycle options. These include TWGs on water-cooled nuclear power reactor Fuel Performance and Technology (TWGFPT), on Nuclear Fuel Cycle Options and spent fuel management (TWGNFCO) and on Heavy Water Reactors (TWGHWR). In addition, IAEA-INPRO project also covers Advanced CANDU Reactors (ACR) and DUPIC fuel cycles. The present paper summarises the Agency's activities in CANDU fuel and fuel cycle, highlighting the progress during the last two years. In the past we saw HWR and LWR technologies and fuel cycles separate, but nowadays their interaction is obviously growing, and their mutual influence may have a synergetic character if we look at the world nuclear fuel cycle as at an integrated system where the both are important elements in line with fast neutron, gas cooled and other advanced reactors. As an international organization the IAEA considers this challenge and makes concrete steps to tackle it for the benefit of all Member States. (author)

N Reactor pressure tube 1350 postirradiation examination

International Nuclear Information System (INIS)

Cook, D.J.

1977-01-01

The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

Transient Moderator Simulation Using CFX10-CAMO, a CANDU Moderator Analysis Model Based on a Coupled Solver

International Nuclear Information System (INIS)

Yoon, Churl; Park, Joo Hwan

2007-01-01

When a PHT(Primary Heat Transfer) system fails to remove excess heat from fuel channels for some loss of coolant accidents(LOCA's) in CANDU NPP's, the fuel channel temperature could increase until the pressure tube strains (i.e., balloon or sag) to contact its surrounding Calandria tube.(PT/CT contact) Following a PT/CT contact, there is a spike in the heat flux to the moderator surrounding the Calandria tube, which may lead to a sustained CT dryout and also a failure of a fuel channel. The prevention of a CT dryout following a PT/CT contact depends on the local moderator subcooling. That is, fuel channel integrity depends on the capability of the moderator to act as the ultimate heat sink for some LOCA's in a CANDU reactor. In KAERI, Yoon et al. developed a CFD model for predicting a CANDU-6 moderator temperature on the basis of a commercial CFD code CFX-4(ANSYS Inc.). This analytic model has the strength of modelling the hydraulic resistances in the core region and accounting for a heat source term in the energy equations. But convergence difficulties and a slow computing speed are the limitations of this model, because the CFX-4 code adapts a segregated solver to resolve a moderator circulation including a strong coupled-effect. Compared to a segregated solver, a coupled-solver is highly efficient and robust especially for a flow with a strong interference between the variables such as combustion

Tensile properties of quadruple melted Zr-2.5Nb pressure tubes evaluated from pressure tube offcuts

International Nuclear Information System (INIS)

Shah, Priti Kotak; Dubey, J.S.; Anantharaman, S.

2013-12-01

Rajasthan Atomic Power Station-2 (RAPS-2) is the first Pressurised Heavy Water Reactor (PHWR) in India having quadruple melted Zr-2.5Nb pressure tubes. Front-end and back-end off-cuts of sixteen pressure tubes were selected for studying the mechanical properties in axial and transverse directions of the tube. Tension tests were carried out at room temperature and at 300℃ using miniature tensile test specimens. The report presents the experimental details and discusses the base line tensile property data for the quadruple melted pressure tubes of RAPS-2. This data will be useful for the reactor life management. (author)

Dynamics of explosively imploded pressurized tubes

Science.gov (United States)

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

2011-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Scott Hower [Zetec Inc. (Romania); Luiza Vladu; Adrian Nichisov; Mihai Cretu [COMPCONTROL ING. (Romania)

2006-07-01

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

Review of the Thermal-Chemical Experiments for CANDU Fuel Channel

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae; Min, Byung Joo; Park, Joo Hwan; Yoon, Churl; Rhee, Bo Wook

2005-08-15

In the present study, thermal-chemical experiments for CANDU channel analysis are reviewed. First, 11 experiments are identified from the present references and classified according to the number of heater rods, channel orientation, and degree of FES (Fuel Element Simulator) temperature rise during transient. The main configuration of the test rigs and the position of the measurement systems are identified. The experiments were generally conducted in three stages, a low-power, a high-power and a no-power stage. These test procedures are classified and described in this document. The experimental conditions for steam, coolant, and heat power are identified. The thermal properties of solid materials and fluids in the test apparatus are listed in the tables. From the review of the main test results, the following conclusions are to be obtained. Some of the reviewed experiments were not in the quasy-steady state conditions at a low-power stage and followed by a high-power stage. Zircaloy/steam reaction started when FES temperature were 800 .deg. C and escalated when temperature exceeded 1150 .deg. C. Uncontrolled temperature escalations due to Zircaloy/steam reaction were not observed when the FES temperature reached peak point (just below the melting point) and electric power to the test section shut off (self-sustaining Zircaloy/steam reaction). There were negligible circumferential temperature gradients in the FES bundle and pressure tube for the experiments performed in a vertical channel orientation. There were, however, noticeable circumferential gradients when the pressure tube was horizontal. These gradients were attributed to slumping of the FES bundle (sagging). Sagging of the bundle may have masked any buoyancy induced temperature gradients. Furthermore, the hot FES sagged towards the pressure tube transferring more heat to the pressure tube and increasing the temperature of the pressure tube.

Review of the Thermal-Chemical Experiments for CANDU Fuel Channel

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Min, Byung Joo; Park, Joo Hwan; Yoon, Churl; Rhee, Bo Wook

2005-08-01

In the present study, thermal-chemical experiments for CANDU channel analysis are reviewed. First, 11 experiments are identified from the present references and classified according to the number of heater rods, channel orientation, and degree of FES (Fuel Element Simulator) temperature rise during transient. The main configuration of the test rigs and the position of the measurement systems are identified. The experiments were generally conducted in three stages, a low-power, a high-power and a no-power stage. These test procedures are classified and described in this document. The experimental conditions for steam, coolant, and heat power are identified. The thermal properties of solid materials and fluids in the test apparatus are listed in the tables. From the review of the main test results, the following conclusions are to be obtained. Some of the reviewed experiments were not in the quasy-steady state conditions at a low-power stage and followed by a high-power stage. Zircaloy/steam reaction started when FES temperature were 800 .deg. C and escalated when temperature exceeded 1150 .deg. C. Uncontrolled temperature escalations due to Zircaloy/steam reaction were not observed when the FES temperature reached peak point (just below the melting point) and electric power to the test section shut off (self-sustaining Zircaloy/steam reaction). There were negligible circumferential temperature gradients in the FES bundle and pressure tube for the experiments performed in a vertical channel orientation. There were, however, noticeable circumferential gradients when the pressure tube was horizontal. These gradients were attributed to slumping of the FES bundle (sagging). Sagging of the bundle may have masked any buoyancy induced temperature gradients. Furthermore, the hot FES sagged towards the pressure tube transferring more heat to the pressure tube and increasing the temperature of the pressure tube

Development of Zirconium alloys (for pressure tubes)

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

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

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro

1994-11-01

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

Station black out analysis for CANDU 6 plant

International Nuclear Information System (INIS)

Baburajan, P.K.; Rao, R.S.; Gupta, S.K.

2011-01-01

As part of International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP), 'Benchmarking severe accident computer codes for pressurised heavy water reactor applications', thermal hydraulic analysis of severe accident station black out (SBO) is carried out for a generic CANDU 6 plant. The CRP is conducted in order to improve severe accident analysis capability for heavy water reactors (HWRs) through the benchmarking exercise. The plant simulation is carried out using RELAP5/Mod3.4 best estimate system thermal hydraulic code. The total thermal power of the plant is 2064 MW. There are 380 fuel channels in the core, 12 fuel bundles per channel and each bundle assembly has 37 fuel elements. The primary heat transport system (HTS) consists of two loops. Each loop consist of inlet and exit headers, feeder lines, fuel channels, hot leg and cold leg of steam generator, pumps, pump suction and discharge lines. Ninety five fuel channels in each pass of the loop are simulated as a single channel. The steam generator as the secondary side heat sink consists of annulus down-comer, riser, steam separator, steam drum, steam header and steam lines. Fuel channels (pressure tube) and calandria tube are simulated using SCDAPSIM to study the severe accident code behaviour. The SBO transient is initiated after obtaining the steady state conditions. Present analysis is carried out till the pressure tube failure. Analysis results show that the secondary inventory is lost in about 6500 seconds of the transient. The primary inventory is lost in 10370 seconds of the transient and subsequently the pressure tube failure is predicted as the tube wall temperature exceeded 900 K. Further analysis is to be carried out by incorporating changes in the calandria model and including the modeling of calandria vault and containment. (author)

Candu technology: the next generation now

International Nuclear Information System (INIS)

Hopwood, J.M.; Duffey, R.B.; Torgerson, D.F.

2001-01-01

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

Development of first full scope commercial CANDU-6 fuel handling simulator

Energy Technology Data Exchange (ETDEWEB)

Crawford, W., E-mail: BCrawford@atlanticnuclear.ca [Atlantic Nuclear Services Inc., Fredericton, NB (Canada); McInerney, J. M., E-mail: JMcInerney@nbpower.com [Point Lepreau Generating Station, Maces Bay, NB (Canada); Moran, E.S.; Nice, J. W.; Sinclair, D.M.; Somerville, S.; Usalp, E.C.; Usalp, M., E-mail: EMoran@atlanticnuclear.ca, E-mail: JNice@atlanticnuclear.ca, E-mail: DSinclair@atlanticnuclear.ca, E-mail: SSomerville@atlanticnuclear.ca, E-mail: ECUsalp@atlanticnuclear.ca, E-mail: MUsalp@atlanticnuclear.ca [Atlantic Nuclear Services Inc., Fredericton, NB (Canada)

2015-07-01

Unique to CANDU reactors is continuous on-power refueling. In the CANDU-6 design, the fuel bundles are contained within 380 pressure tubes. Fuelling machines, one on either side of the reactor face move on a bridge and carriage system to the appointed channel and fuel under computer control. The fuelling machine is an immensely complicated mechanical device. None of the original Canadian full scope simulators incorporated the interaction of the fuel handling system. Traditionally, the final stages of Fuel Handling Operator qualification utilizes on the job training in a production environment carried out in the station main control room. For the purpose of supporting continual improvement in fuel handling training at the Third Qinshan Nuclear Plant Company (TQNPC), Atlantic Nuclear Services in a joint project with New Brunswick Power, developed the first commercial full scope CANDU-6 Fuel Handling simulator, integrated into the existing TQNPC Full Scope Simulator framework. The TQNPC Fuel Handling simulator is capable of supporting all normal on-power and off-power refuelling procedures as well as other abnormal operating conditions, which will allow training to be conducted, based on the plant specific operating procedures. This paper will discuss its development, the importance of this tool and its advantages over past training practices. (author)

Development of first full scope commercial CANDU-6 fuel handling simulator

International Nuclear Information System (INIS)

Crawford, W.; McInerney, J. M.; Moran, E.S.; Nice, J. W.; Sinclair, D.M.; Somerville, S.; Usalp, E.C.; Usalp, M.

2015-01-01

Unique to CANDU reactors is continuous on-power refueling. In the CANDU-6 design, the fuel bundles are contained within 380 pressure tubes. Fuelling machines, one on either side of the reactor face move on a bridge and carriage system to the appointed channel and fuel under computer control. The fuelling machine is an immensely complicated mechanical device. None of the original Canadian full scope simulators incorporated the interaction of the fuel handling system. Traditionally, the final stages of Fuel Handling Operator qualification utilizes on the job training in a production environment carried out in the station main control room. For the purpose of supporting continual improvement in fuel handling training at the Third Qinshan Nuclear Plant Company (TQNPC), Atlantic Nuclear Services in a joint project with New Brunswick Power, developed the first commercial full scope CANDU-6 Fuel Handling simulator, integrated into the existing TQNPC Full Scope Simulator framework. The TQNPC Fuel Handling simulator is capable of supporting all normal on-power and off-power refuelling procedures as well as other abnormal operating conditions, which will allow training to be conducted, based on the plant specific operating procedures. This paper will discuss its development, the importance of this tool and its advantages over past training practices. (author)

Delayed Hydride Cracking Mechanism in Zirconium Alloys and Technical Requirements for In-Service Evaluation of Zr-2.5Nb Tubes with Flaws

International Nuclear Information System (INIS)

Kim, Young Suk

2007-01-01

In association with periodic inspection of CANDU nuclear power plant components, Canadian Standards Association issued CSA N285.8 in 2005 as technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors. This first version, CSA N285.8 involves procedures for, firstly, the evaluation of pressure tube flaws, secondly, the evaluation of pressure tube to calandria tube contact and, thirdly, the assessment of a reactor core, and material properties and derived quantities. The evaluation of pressure tube flaws includes delayed hydride cracking evaluation the procedures of which are stipulated based on the existing delayed hydride cracking models. For example, the evaluation of flaw-tip hydride precipitation during reactor cooldown involves a procedure to calculate the equilibrium hydrogen equivalent concentration in solution at the flaw tip, Htipas follows: Htip=Hfexp[- (VH delta no.)/RT], where Hf is the total bulk hydrogen equivalent concentration, VH partial molar volume of hydrogen in zirconium, δ a difference in hydrostatic stress between the bulk and the crack tip. When Htip ≥TSSP at temperature, then flaw-tip hydride is predicted to precipitate. Eq. (1) suggests that hydrogen concentration at the crack tip would increase due to an work energy given by the difference in the hydrostatic stress

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

International Nuclear Information System (INIS)

Krishnan, S.; Bhasin, V.; Mahajan, S.C.

1997-01-01

Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

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

Energy Technology Data Exchange (ETDEWEB)

Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

1997-04-01

Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

Selected reading on introduction to pressure tube technology

International Nuclear Information System (INIS)

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

1981-10-01

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

CANDU 6 - the highly successful medium sized reactor

International Nuclear Information System (INIS)

Hedges, K. R.; Allen, P. J.; Hopwood, J. M.

2000-01-01

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

Laser plasma generation of hydrogen-free diamond-like carbon thin films on Zr-2.5Nb CANDU pressure tube materials and silicon wafers with a pulsed high-power CO2 laser

International Nuclear Information System (INIS)

Ebrahim, N.A.; Mouris, J.F.; Hoffmann, C.R.J.; Davis, R.W.

1995-06-01

We report the first experiments on the laser plasma deposition of hydrogen-free, diamond-like carbon (DLC) films on Zr-2.5Nb CANDU pressure-tube materials and silicon substrates, using the short-pulse, high-power, CO 2 laser in the High-Power Laser Laboratory at Chalk River Laboratories. The films were (AFM). The thin films show the characteristic signature of DLC films in the Raman spectra obtained using a krypton-ion (Kr + ) laser. The Vickers ultra-low-load microhardness tests show hardness of the coated surface of approximately 7000 Kg force mm -2 , which is consistent with the hardness associated with DLC films. AFM examination of the film morphology shows diamond-like crystals distributed throughout the film, with film thicknesses of up to 0.5 μm generated with 50 laser pulses. With significantly more laser pulses, it is expected that very uniform diamond-like films would be produced. These experiments suggest that it should be possible to deposit hydrogen-free, diamond-like films of relevance to nuclear reactor components with a high-power and high-repetition-rate laser facility. (author). 7 refs., 2 tabs., 15 figs

Supporting CANDU operators-CANDU owners group

International Nuclear Information System (INIS)

Collingwood, B.R.

1997-01-01

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

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

The CANDU 9 distributed control system design process

International Nuclear Information System (INIS)

Harber, J.E.; Kattan, M.K.; Macbeth, M.J.

1997-01-01

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

N Reactor pressure tube 2566 postirradiation examination

International Nuclear Information System (INIS)

Scott, K.V.

1978-01-01

Pressure tube 2566 was removed from N Reactor in July, 1977 to initiate the postirradiation examination program required by the Technical Specifications. Destructive examination of the pressure tube, after a maximum accumulated fluence of 4.6 x 10 21 n/cm 2 (E > 1 MeV), was conducted at the Hanford Engineering Development Laboratory to determine the effects of reactor service on the mechanical properties and hydrogen absorption and corrosion characteristics of the pressure tube. Tube 2566 is the sixth tube removed for destructive examination since the initial reactor startup. Evaluation of test results reveal that no significant detrimental changes have occurred in the parameters studied, since the last tube was removed in 1974

An analysis on water hammer in liquid injection shutdown system of CANDU-9

International Nuclear Information System (INIS)

Kim, T. H.; Heo, J.; Han, S. K.; Choi, H. Y.; No, T. S.

2000-01-01

The water hammer analysis code, PTRAN, is used for computation of transient pressures and pressure differentials in the Liquid Injection Shutdown System(LISS) piping network of CANDU-9 to ensure that the design allowables for LEVEL C Service Limit are met for the water hammer loads resulting from the water hammer. The LISS piping network of CANDU-9 has incorporated design improvement in considering the water hammer, such as declining the horizontal part of helium header, and raising the elevation of the overall system piping configuration, etc. The maximum pressure in the LISS piping network is found to be 7.92 MPa(a) at the closed valve in the vent line, which is below the allowable working pressure and the valve design pressure under Level C service conditions. And it is also shown that the maximum pressure in CANDU-9 is much lower than that in CANDU-6

CFD thermal-hydraulic analysis of a CANDU fuel channel

International Nuclear Information System (INIS)

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

2009-01-01

This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational fluid dynamics) methodology approach. Limited computer power available at Bucharest University POLITEHNICA forced the authors to analyse only segments of fuel channel namely the significant ones: fuel bundle junctions with adjacent segments, fuel bundle spacer planes with adjacent segments, regular segments of fuel bundles. The computer code used is FLUENT. Fuel bundles contained in pressure tubes forms a complex flow domain. The flow is characterized by high turbulence and in some parts of fuel channel also by multi-phase flow. The flow in the fuel channel has been simulated by solving the equations for conservation of mass and momentum. For turbulence modelling the standard k-e model is employed although other turbulence models can be used as well. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Since we consider only some relatively short segments of a CANDU fuel channel we can assume, for this starting stage, that heat transfer is not very important for these short segments of fuel channel. The boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. In this paper we present results for Standard CANDU 6 Fuel Bundles as a basis for CFD thermal-hydraulic analysis of INR proposed SEU43 and other new nuclear fuels. (authors)

Potential drop crack measurement systems for CANDU components

Energy Technology Data Exchange (ETDEWEB)

Sahney, R [Carleton Univ., Ottawa, ON (Canada)

1994-12-31

A project to develop an automated crack measurement system for CANDU pressure tube burst testing is currently underway. The system will utilize either Direct Current Potential Drop (DCPD) or Alternating Current Potential Drop (ACPD) techniques for crack measurement. The preliminary stage of the project involves testing and comparison of both ACPD and DCPD methods on a Zr - 2.5% Nb alloy plate with saw cuts (used to simulate cracks). Preliminary results show that both ACPD and DCPD techniques are capable of detecting cracks; further testing is in progress to determine the ability of each of the two systems to make accurate crack depth measurements. This paper will describe the two potential drop techniques and will present test results from the experimental program. (author). 10 refs., 7 figs.

Modelling of pressure tube Quench using PDETWO

International Nuclear Information System (INIS)

Parlatan, Y.; Lei, Q.M.; Kwee, M.

2004-01-01

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

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

International Nuclear Information System (INIS)

Larson, G.C.

1980-01-01

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

Pressure tube rupture in a closed tank

International Nuclear Information System (INIS)

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

1985-06-01

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

21 CFR 868.5860 - Pressure tubing and accessories.

Science.gov (United States)

2010-04-01

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

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

International Nuclear Information System (INIS)

1991-09-01

This third edition of CAN/CSA-N285.1 supersedes the 1981 and 1975 editions. It provides the specific requirements for design, fabrication, and installation of Class 1, 2 and 3 pressure-retaining systems and components in CANDU nuclear power plants, and over pressure protection of the heat transport system. The general requirements for pressure-retaining systems and components are given in CSA Standard CAN/CSA-N285.0, with which Class 1, 2 and 3 systems and components must also comply

Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report

International Nuclear Information System (INIS)

Rogers, J.T.

1984-06-01

This volume of appendices presents listings and sample runs of the computer codes used in the study of the thermalhydraulic behaviour of CANDU reactor cores during severe loss of coolant accidents. The codes, written in standard FORTRAN, are MODBOIL, to calculate moderator temperatures, pressures and water levels; DEBRIS, to calculate the transient temperature distribution in the debris of calandria and pressure tubes and fuel pellets; MOLTENPOOL, to calculate the temperature history in a pool of molten debris; CONFILM, to calculate the behaviour of a condensing film of vaporized core debris on the calandria wall, and BLDG, to calculate the pressurization of the containment during the expulsion of moderator through pressure relief ducts. In addition there are discussions of the average condensation heat transfer coefficient for vaporized core material on the calandria wall, and of vapor explosions

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

International Nuclear Information System (INIS)

Rao, Y.F.; Cheng, Z.; Waddington, G.M.

2014-01-01

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

Absorber materials in CANDU PHWR's

International Nuclear Information System (INIS)

Price, E.G.; Boss, C.R.; Novak, W.Z.; Fong, R.W.L.

1995-03-01

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 a relatively benign environment of low pressure, low temperature heavy water between neighbouring rows of 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 designed back-fit resolved the problem. (author). 3 refs., 1

Establishment of Safety Analysis System and Technology for CANDU Reactors

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, W. Y.; Kim, H. T.; Rhee, B. W.; Yoon, C.; Kang, H. S.; Yoo, K. J.

2005-03-01

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

Pressure tube reactors

International Nuclear Information System (INIS)

Natori, Hisahide.

1981-01-01

Purpose: To improve the electrical power generation efficiency in a pressure tube reactor in which coolants and moderators are separated by feedwater heating with heat generated in heavy water and by decreasing the amount of steams to be extracted from the turbine. Constitution: A heat exchanger and a heavy water cooler are additionally provided to a conventional pressure tube reactor. The heat exchanger is disposed at the pre-stage of a low pressure feedwater heater series. High temperature heavy water heated in the core is passed through the primary side of the exchanger, while feedwater is passed through the secondary side. The cooler is disposed on the downstream of the heat exchanger in the flowing direction of the heavy water, in which heavy water from the heat exchanger is passed through the primary side and the auxiliary equipment cooling water is sent to the secondary side thereof. Accordingly, since extraction of heating steams is no more necessary, the steam can be used for the rotation of the turbine, and the electrical power generation efficiency can be improved. (Seki, T.)

Experimental investigation of coolant and poisoned moderator mixing due to a simulated pressure tube/calandria tube fishmouth rupturing an overpoisoned guaranteed shutdown state

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

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

New concepts, requirements and methods concerning the periodic inspection of the CANDU fuel channels

International Nuclear Information System (INIS)

Denis, J.R.

1995-01-01

Periodic inspection of fuel channels is essential for a proper assessment of the structural integrity of these vital components of the reactor. The development of wet channel technologies for non-destructive examination (NDE) of pressure tubes and the high technical performance and reliability of the CIGAR equipment have led, in less than 1 0 years, to the accumulation of a very significant volume of data concerning the flaw mechanisms and structural behaviour of the CANDU fuel channels. On this basis, a new form of the CAN/CSA-N285.4 Standard for Periodic Inspection of CANDU Nuclear Power Plant components was elaborated, introducing new concepts and requirements, in accord with the powerful NDE methods now available. This paper presents these concepts and requirements, and discusses the NDE methods, presently used or under development, to satisfy these requirements. Specific features regarding the fuel channel inspections of Cernavoda NGS Unit 1 are also discussed. (author)

Trends in the capital costs of CANDU generating stations

International Nuclear Information System (INIS)

Yu, A.M.

1982-09-01

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

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

International Nuclear Information System (INIS)

Shewfelt, R.S.W.

1986-01-01

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

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

International Nuclear Information System (INIS)

Cheng, Z.; Rao, Y.F.; Waddington, G.M.

2014-01-01

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

Development of heat treated Zr-2.5% Nb alloy tubes for pressure tubes

International Nuclear Information System (INIS)

Saibaba, N.; Jha, S.K.; Tonpe, S.

2011-01-01

Zr-2.5% Nb alloy is the candidate material for pressure tubes of Pressurized Heavy Water Reactors (PHWR), and are manufactured in cold working condition while heat treated pressure tubes are used in RBMK and FUGEN type of reactors. The diametral creep of these tubes is the life limiting factor. This paper presents the extensive work carried out for the optimization of process parameters to manufacture heat treated Zr-2.5% Nb pressure tubes. Extensive dilactometry study was carried out to establish the transus temperature for the alloy and the effect of soaking temperature and cooling rate on the microstructure was characterized. On the basis of the study, water quenching (at 883 deg C) in the a b region with 20-25% primary a phase was selected, further cold worked, aged and finally autoclaved. Mechanical properties of the finished tubes were found to be comparable to the cold worked route. Large number of full sized tubes of about 700 - 800 mm long was produced to establish the repeatability. (author)

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.

Development of nuclear fuel. Development of CANDU advanced fuel bundle

International Nuclear Information System (INIS)

Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan; Jung, Sung Hoon

1991-07-01

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

Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

Energy Technology Data Exchange (ETDEWEB)

Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

2012-02-15

Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

Neutronics simulations on hypothetical power excursion and possible core melt scenarios in CANDU6

International Nuclear Information System (INIS)

Kim, Yonghee

2015-01-01

LOCA (Loss of coolant accident) is an outstanding safety issue in the CANDU reactor system since the coolant void reactivity is strongly positive. To deal with the LOCA, the CANDU systems are equipped with specially designed quickly-acting secondary shutdown system. Nevertheless, the so-called design-extended conditions are requested to be taken into account in the safety analysis for nuclear reactor systems after the Fukushima accident. As a DEC scenario, the worst accident situation in a CANDU reactor system is a unprotected LOCA, which is supposed to lead to a power excursion and possibly a core melt-down. In this work, the hypothetical unprotected LOCA scenario is simulated in view of the power excursion and fuel temperature changes by using a simplified point-kinetics (PK) model accounting for the fuel temperature change. In the PK model, the core reactivity is assumed to be affected by a large break LOCA and the fuel temperature is simulated to account for the Doppler effect. In addition, unlike the conventional PK simulation, we have also considered the Xe-I model to evaluate the impact of Xe during the LOCA. Also, we tried to simulate the fuel and core melt-down scenario in terms of the reactivity through a series of neutronics calculations for hypothetical core conditions. In case of a power excursion and possible fuel melt-down situation, the reactor system behavior is very uncertain. In this work, we tried to understand the impacts of fuel melt and relocation within the pressure vessel on the core reactivity and failure of pressure and calandria tubes. (author)

R and D in support of CANDU plant life management

International Nuclear Information System (INIS)

Tapping, R.L.; Holt, R.A.

1999-01-01

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

Safety assessment of the advanced CANDU reactor in postulated LOCA/LOECC events

International Nuclear Information System (INIS)

Hazen Hezhi Fan; Zoran Bilanovic

2005-01-01

transport system, representative fuel channels, and fuel during a postulated LOCA/LOECC event; including fuel channel flow conditions, coolant reduction and refilling, heat transfer through convection, conduction and thermal radiation, heat generation and hydrogen production from zirconium/steam reactions, and pressure tube deformation. The key parameters calculated include fuel and pressure tube temperatures, amount of hydrogen generation, pressure tube strain, and heat load to the moderator. These parameters are needed to assess fuel and fuel channel integrity, fission-product release, and the response of the reactor core and containment. The assessment results show that the ACR reserve water system would inject water into the reactor core by gravity. This injection successfully helps to mitigate the consequences of postulated LOCA/LOECC events; hence, the fission-product release to the public can be minimized with large safety margins. (authors)

CANDU technology for generation III + AND IV reactors

International Nuclear Information System (INIS)

Torgerson, D.F.

2005-01-01

Atomic Energy of Canada Limited (AECL) is the original developer of the CANDU?reactor, one of the three major commercial power reactor designs now used throughout the world. For over 60 years, AECL has continued to evolve the CANDU design from the CANDU prototypes in the 1950s and 1960s through to the second generation reactors now in operation, including the Generation II+ CANDU 6. The next phase of this evolution, the Generation III+ Advanced CANDU ReactorTM (ACRTM), continues the strategy of basing next generation technology on existing CANDU reactors. Beyond the ACR, AECL is developing the Generation IV CANDU Super Critical Water Reactor. Owing to the evolutionary nature of these advanced reactors, advanced technology from the development programs is also being applied to operating CANDU plants, for both refurbishments and upgrading of existing systems and components. In addition, AECL is developing advanced technology that covers the entire life cycle of the CANDU plant, including waste management and decommissioning. Thus, AECL maintains state-of-the-art expertise and technology to support both operating and future CANDU plants. This paper outlines the scale of the current core knowledge base that is the foundation for advancement and support of CANDU technology. The knowledge base includes advancements in materials, fuel, safety, plant operations, components and systems, environmental technology, waste management, and construction. Our approach in each of these areas is to develop the underlying science, carry out integrated engineering scale tests, and perform large-scale demonstration testing. AECL has comprehensive R and D and engineering development programs to cover all of these elements. The paper will show how the ongoing expansion of the CANDU knowledge base has led to the development of the Advanced CANDU Reactor. The ACR is a Generation III+ reactor with substantially reduced costs, faster construction, and enhanced passive safety and operating

A review of recent advances in the role of leak-before-break concept in assessments of flaws detected in CANDU pressure tubes

International Nuclear Information System (INIS)

Crespi, J.C.

1994-01-01

If a crack develops in a pressure tube, the leak is detected by monitoring the moisture in the gas annulus and the reactor shutdown before it becomes unstable. Because the delayed hydride cracking has been associated to date with all pressure tube failures at a rolled joints, the delayed hydride cracking is considered to be the dominant mecanism by which the flaws can grow to a size which exceeds the critical crack length. For the delayed hydride cracking failure mode leak-before-break is used as defense in depth against unstable rupture. The methodology depends on showing than the time available to detect a delayed hydride crack is much greater that the time required to detect it in the gas annulus. The time available is estimated from measurements of: (a) axial delayed hydride crack growth rates, (b) crack lengths at penetrations of the tube wall when leakage first occurs and (c) critical crack lengths at instability when a crack is growing by the delayed hydride cracking mechanism. A review of recent advances in the experimental data used in leak-before-break assessment are presented and discussed. (author). 17 refs, 6 figs, 2 tabs

CANDU: study and review

International Nuclear Information System (INIS)

Morad, César M.; Santos, Thiago A. dos

2017-01-01

The CANDU (Canadian Deuterium Uranium) is a nuclear reactor developed by AECL (Atomic Energy of Canada Limited). The first small-scale reactor is known as NPD and was made in 1955 and commenced operation in 1962. It is a pressurized heavy water reactor and uses D2O as moderator and coolant and therefore uses natural uranium as fuel. There have been two major types of CANDU reactors, the original design of around 500 MWe that was intended to be used in multi-reactor installations in large plants, and the rationalized CANDU6 which has units in Argentina, South Korea, Pakistan, Romania and China. Throughout the 1980s and 90s the nuclear power market suffered a major crash, with few new plants being constructed in North America or Europe. Design work continued through, however, and a number of new design concepts were introduced that dramatically improved safety, capital costs, economics and overall performance. These Generation III+ and Generation IV machines became a topic of considerable interest in the early 2000s as it appeared a nuclear renaissance was underway and large numbers of new reactors would be built over the next decade. The present work aims to study the reactors of the CANDU type, exploring from its creation to studies directed to G-III and G-IV reactors. (author)

CANDU: study and review

Energy Technology Data Exchange (ETDEWEB)

Morad, César M., E-mail: cesar.morad@usp.br [Universidade de São Paulo (POLI/USP), SP (Brazil). Escola Politécnica; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

2017-07-01

The CANDU (Canadian Deuterium Uranium) is a nuclear reactor developed by AECL (Atomic Energy of Canada Limited). The first small-scale reactor is known as NPD and was made in 1955 and commenced operation in 1962. It is a pressurized heavy water reactor and uses D2O as moderator and coolant and therefore uses natural uranium as fuel. There have been two major types of CANDU reactors, the original design of around 500 MWe that was intended to be used in multi-reactor installations in large plants, and the rationalized CANDU6 which has units in Argentina, South Korea, Pakistan, Romania and China. Throughout the 1980s and 90s the nuclear power market suffered a major crash, with few new plants being constructed in North America or Europe. Design work continued through, however, and a number of new design concepts were introduced that dramatically improved safety, capital costs, economics and overall performance. These Generation III+ and Generation IV machines became a topic of considerable interest in the early 2000s as it appeared a nuclear renaissance was underway and large numbers of new reactors would be built over the next decade. The present work aims to study the reactors of the CANDU type, exploring from its creation to studies directed to G-III and G-IV reactors. (author)

Sample summary report for ARG 1 pressure tube sample

International Nuclear Information System (INIS)

Belinco, C.

2006-01-01

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

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

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

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

Distributed control system for CANDU 9 nuclear power plant

International Nuclear Information System (INIS)

Harber, J.E.; Kattan, M.K.; Macbeth, M.J.

1996-01-01

Canadian designed CANDU pressurized heavy water nuclear reactors have been world leaders in electrical power generation. The CANDU 9 project is AECL's next reactor design. The CANDU 9 plant monitoring, annunciation, and control functions are implemented in two evolutionary systems; the distributed control system (DCS) and the plant display system (PDS). The CDS implements most of the plant control functions in a single hardware platform. The DCS communicates with the PDS to provide the main operator interface and annunciation capabilities of the previous control computer designs along with human interface enhancements required in a modern control system. (author)

CFD Analysis for the Steady State Test of CS28-1 Simulating High Temperature Chemical Reactions in CANDU Fuel Channel

International Nuclear Information System (INIS)

Park, Ju Hwan; Kang, Hyung Seok; Rhee, Bo Wook

2006-05-01

The establishment of safety analysis system and technology for CANDU reactors has been performed at KAERI. As for one of these researches, single CANDU fuel bundle has been simulated by CATHENA for the post-blowdown event to consider the complicated geometry and heat transfer in the fuel channel. In the previous LBLOCA analysis methodology adopted for Wolsong 2, 3, 4 licensing, the fuel channel blowdown phase was analyzed by a CANDU system analysis code CATHENA and the post-blowdown phase of fuel channel was analyzed by CHAN-IIA code. The use of one computer code in consecutive analyses appeared to be desirable for consistency and simplicity in the safety analysis process. However, validation of the high temperature post-blowdown fuel channel model in the CATHENA before being used in the accident analysis is necessary. Experimental data for the 37-element fuel bundle that fueled CANDU-6 has not been performed. The benchmark problems for the 37-element fuel bundle using CFD code will be compared with the test results of the 28-element fuel bundle in the CS28-1 experiment. A full grid model of FES to the calandria tube simulating the test section was generated. The number of the generated mesh in the grid model was 4,324,340 cells. The boundary and heat source conditions, and properties data in the CFD analysis were given according to the test results and reference data. Thermal hydraulic phenomena in the fuel channel were simulated by a compressible flow, a highly turbulent flow, and a convection/conduction/radiation heat transfer. The natural convection flow of CO 2 due to a large temperature difference in the gap between the pressure and the calandria tubes was treated by Boussinesq's buoyancy model. The CFD results showed good agreement with the test results as a whole. The inner/middle/outer FES temperature distributions of the CFD results showed a small overestimated value of about 30 .deg. C at the entrance region, but good agreement at the outlet region. The

CANDU passive shutdown systems

Energy Technology Data Exchange (ETDEWEB)

Hart, R S; Olmstead, R A [AECL CANDU, Sheridan Park Research Community, Mississauga, ON (Canada)

1996-12-01

CANDU incorporates two diverse, passive shutdown systems, independent of each other and from the reactor regulating system. Both shutdown systems function in the low pressure, low temperature, moderator which surrounds the fuel channels. The shutdown systems are functionally different, physically separate, and passive since the driving force for SDS1 is gravity and the driving force for SDS2 is stored energy. The physics of the reactor core itself ensures a degree of passive safety in that the relatively long prompt neutron generation time inherent in the design of CANDU reactors tend to retard power excursions and reduces the speed required for shutdown action, even for large postulated reactivity increases. All passive systems include a number of active components or initiators. Hence, an important aspect of passive systems is the inclusion of fail safe (activated by active component failure) operation. The mechanisms that achieve the fail safe action should be passive. Consequently the passive performance of the CANDU shutdown systems extends beyond their basic modes of operation to include fail safe operation based on natural phenomenon or stored energy. For example, loss of power to the SDS1 clutches results in the drop of the shutdown rods by gravity, loss of power or instrument air to the injection valves of SDS2 results in valve opening via spring action, and rigorous self checking of logic, data and timing by the shutdown systems computers assures a fail safe reactor trip through the collapse of a fluctuating magnetic field or the discharge of a capacitor. Event statistics from operating CANDU stations indicate a significant decrease in protection system faults that could lead to loss of production and elimination of protection system faults that could lead to loss of protection. This paper provides a comprehensive description of the passive shutdown systems employed by CANDU. (author). 4 figs, 3 tabs.

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

Directory of Open Access Journals (Sweden)

Mukul Kumar Jain

2011-01-01

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

Characterization of the inside and outside oxide surfaces of irradiated pressure tubes of Embalse nuclear power plant

International Nuclear Information System (INIS)

Bordoni, Roberto A.; Olmedo, Ana M.

2004-01-01

The inside and outside surfaces of two pressure tubes (PT) removed from Embalse nuclear power plant (CNE) after 10 of effective full power years (EFPY) were characterized. The oxide thickness of both faces, in different zones, was also measured. The inside surfaces of both PTs, B-102 (A-14) and B-298 (L-12), were covered with a black oxide that replicates the original PT surface. A network of microcracks perpendicular to the inside surface in contact with the coolant was found. In some cases, near the outlet of the PT, some spalling of the oxide was also found. These small microcracks and spalling do not affect the protective character of the oxide since a thickness about 5 or 6 μm of an undamaged oxide is found at the metal/oxide interface side. The oxide thickness changes between approximately 6 to 12 μm for B-102 tube and around 7 to 15 μm for B-298 tube. The average corrosion rate is 1.16 μm/10 4 HH for B-102 tube and 1.35 μm/10 4 HH for B-298 tube at 5.8 m position for both PTs. These corrosion rates show good corrosion behaviour of CNE PTs. The average corrosion rate of the inside surface of the PTs depends on the coolant temperature but not on fast neutron flux. The outside oxide film is black, shiny, compact and protective, replicating also the original surface. The oxide thickness changes between 2 to 6.5 μm in B-102 tube and between 1.8 to 3.7 μm B-298 tube. These oxide thicknesses are within the values reported for PTs in CANDU Stations. (author) [es

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

International Nuclear Information System (INIS)

Tingkuan, C.; Xuanzheng, C.

1987-01-01

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

Remotised sliver sample removal from irradiated pressure tube

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Remotised sliver sample removal from irradiated pressure tube

International Nuclear Information System (INIS)

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

1994-01-01

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

Predicting diametral creep of the pressure tubes in CANDU reactors using fuzzy neural networks

International Nuclear Information System (INIS)

Lee, Jae Yong; Na, Man Gyun; Park, Jong Ho

2011-01-01

Pressure tube (PT) creep is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS) in Canada deuterium uranium reactors. PT diametral creep affects the thermal hydraulic characteristics of coolant channels and the critical heat flux (CHF). CHF is a key parameter in determining the critical channel power, which is used in the trip setpoint calculations of regional overpower protection systems. This paper aims to predict PT diametral creep using the measured signals of the HTS by applying fuzzy neural networks (FNNs) according to operating conditions. The FNN model was optimized in terms of its fuzzy rules and parameters by a genetic algorithm combined with the least-squares method. Informative data that demonstrate the system's characteristic behavior were selected to train the FNN model using the subtractive clustering method. The proposed FNN model for predicting PT diametral creep was verified using the operating data of the Wolsong Unit 1 nuclear power plant in Korea. It was known that the FNN could predict the PT diametral creep accurately. Statistical and analytical uncertainty analysis methods were applied to the models and their uncertainties were evaluated using 60 sampled training and optimization data sets, as well as two fixed test data sets. (author)

Using MCNP for in-core instrument calibration in CANDU

Energy Technology Data Exchange (ETDEWEB)

Taylor, D.C. [Point Lepreau Generating Station, NB Power, Lepreau, New Brunswick (Canada); Anghel, V.N.P.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2002-07-01

The calibration of in-core instruments is important for safe and economical CANDU operation. However, in-core detectors are not normally suited to bench calibration procedures. This paper describes the use and validation of detailed neutron transport calculations for the purpose of calibrating the response of in-core neutron flux detectors. The Monte-Carlo transport code, MCNP, was used to model the thermal neutron flux distribution in the region around self-powered in-core flux detectors (ICFDs), and in the vicinity of the calandria edge. The ICFD model was used to evaluate the reduction in signal of a given detector (the 'detector shading factor') due to neutron absorption in surrounding materials, detectors, and lead-cables. The calandria edge model was used to infer the accuracy of the calandria edge position from flux scans performed by AECL's traveling flux detector (TFD) system. The MCNP results were checked against experimental results on ICFDs, and also against shading factors computed by other means. The use of improved in-core detector calibration factors obtained by this new methodology will improve the accuracy of spatial flux control performance in CANDU-6 reactors. The accurate determination of TFD based calandria edge position is useful in the quantitative measurement of changes in in-core component dimensions and position due to aging, such as pressure tube sag. (author)

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

International Nuclear Information System (INIS)

Yoon, Churl; Rhee, Bo Wook; Min, Byung Joo

2003-03-01

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

Advanced CANDU reactors

International Nuclear Information System (INIS)

Dunn, J.T.; Finlay, R.B.; Olmstead, R.A.

1988-12-01

AECL has undertaken the design and development of a series of advanced CANDU reactors in the 700-1150 MW(e) size range. These advanced reactor designs are the product of ongoing generic research and development programs on CANDU technology and design studies for advanced CANDU reactors. The prime objective is to create a series of advanced CANDU reactors which are cost competitive with coal-fired plants in the market for large electricity generating stations. Specific plant designs in the advanced CANDU series will be ready for project commitment in the early 1990s and will be capable of further development to remain competitive well into the next century

Operating performance and reliability of CANDU PHWR fuel channels in Canada

International Nuclear Information System (INIS)

Strachan, B.; Brown, D.R.

1983-03-01

CANDU nuclear plants use many small-diameter high-pressure fuel channels. Good operating performance from the CANDU fuel channels has made a major contribution to the world-leading operating record of the CANDU nuclear power plants. As of 1982 December 31, there were 7,480 fuel channels installed in 18 CANDU reactors over 500 MW(e) in size. Eight of these reactors have been declared in-service and have accumulated 24,000 fuel channel-years of operation. The only significant operating problems with fuel channels have been the occurrence of leaking cracks in 70 fuel channels and a larger amount of axial creep on the early reactors than was originally provided for in the design. Both of these problems have been corrected on all CANDU reactors built since the Bruce GS 'A' station and the newer reactors should exhibit even better performance

Visualization of Moderator Circulation with 1/8 scaled HU-KINS for Calandria Tank of CANDU6

International Nuclear Information System (INIS)

Cho, Dae Kun; Kim, Nam Seok; Lee, Jae Young; Kim, Man Woong

2005-01-01

The moderator cooling capability of Calandria tank is important during the hypothetical incident of CANDU reactor. Even though the loss of ECCS cooling, the heat of the nuclear fuel can be cooled by the heat transfer path through the contacting location between the pressure tube and Calandria tube. The moderator cooling system removes the heat from the fuel so it pays an ultimate heat sink for the nuclear safety. Therefore, the understanding and estimating the thermal hydraulic conditions of the circulating moderator in the Calandria is of importance. However, the three dimensional nature and complexity of the Calandria tubes make it difficult to analyze. Recent progress in the the three dimensional computational fluid dynamics including the model of turbulent flow stimulate the studies on this subject. For the design or safety analysis purpose, the mesh generation and furnishing proper physical models are imperative and the experimental validation should be performed. However, the experimental facilities have been constructed in the scaled way by the Canadian research groups, AECL and COG. Two facilities: SPEL (1/10 scale by Koroyannaski, 1983) and STERN (1/4 scale by Hadaller, 1990) have been producing experimental data such as the local temperatures and velocities. Several papers has been published to report their CFD codes can be available for the CANDU analysis by comparing the calculations with the experimental results of those facilities. However, as noted by Lee et al. the previous experimental facilities were not scaled properly both in the sense of the force balance between buoyancy force and the jet inertia force. Also, the power density of the moderator was not scaled properly. Therefore, it cannot be said that the observations through SPEL and STERN may have a certain discrepancy from the real CANDU-6 plant. Lee et al (2003) developed the scaling laws for the CANDU-6 and design a 1/8 scale experimental facility named as HU-KINS. In the present paper

Advanced operator interface design for CANDU-3 fuel handling system

Energy Technology Data Exchange (ETDEWEB)

Arapakota, D [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

1996-12-31

The Operator Interface for the CANDU 3 Fuel Handling (F/H) System incorporates several improvements over the existing designs. A functionally independent sit-down CRT (cathode-ray tube) based Control Console is provided for the Fuel Handling Operator in the Main Control Room. The Display System makes use of current technology and provides a user friendly operator interface. Regular and emergency control operations can be carried out from this control console. A stand-up control panel is provided as a back-up with limited functionality adequate to put the F/H System in a safe state in case of an unlikely non-availability of the Plant Display System or the F/H Control System`. The system design philosophy, hardware configuration and the advanced display system features are described in this paper The F/H Operator Interface System developed for CANDU 3 can be adapted to CANDU 9 as well as to the existing stations. (author).

Advanced operator interface design for CANDU-3 fuel handling system

International Nuclear Information System (INIS)

Arapakota, D.

1995-01-01

The Operator Interface for the CANDU 3 Fuel Handling (F/H) System incorporates several improvements over the existing designs. A functionally independent sit-down CRT (cathode-ray tube) based Control Console is provided for the Fuel Handling Operator in the Main Control Room. The Display System makes use of current technology and provides a user friendly operator interface. Regular and emergency control operations can be carried out from this control console. A stand-up control panel is provided as a back-up with limited functionality adequate to put the F/H System in a safe state in case of an unlikely non-availability of the Plant Display System or the F/H Control System'. The system design philosophy, hardware configuration and the advanced display system features are described in this paper The F/H Operator Interface System developed for CANDU 3 can be adapted to CANDU 9 as well as to the existing stations. (author)

Development of CANDU ECCS performance evaluation methodology and guides

Energy Technology Data Exchange (ETDEWEB)

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

2003-03-15

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

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

International Nuclear Information System (INIS)

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

1994-07-01

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

Delayed hydrogen cracking of zirconium alloy pressure tubes

International Nuclear Information System (INIS)

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

1976-10-01

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

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

International Nuclear Information System (INIS)

Hu, M.H.

1998-01-01

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

Determination of representative CANDU feeder dimensions for engineering simulator

International Nuclear Information System (INIS)

Cho, S.; Muzumdar, A.

1996-01-01

This paper describes a logic for selection of representative channel groups and a methodology for determination of representative CANDU feeder dimensions and the pressure drops between inlet/outlet header and fuel channel in the primary loop. A code, MEDOC, was developed based on this logic and methodology and helps perform a calculation of representative feeder dimensions for a selected channel group on the basis of feeder geometry data (fluid volume, mass flow rate, loss factor) and given property data (pressure, quality, density) at inlet/outlet header. The representative feeder dimensions calculated based on this methodology will be useful for the engineering simulator for the CANDU type reactor. (author)

Life extension of CANDU reactor cores

International Nuclear Information System (INIS)

Millard, J.; Kerker, J.; Albert, M.

2011-01-01

Candu Energy (formerly AECL), in partnership with station operators, has developed a robust methodology for demonstrating the fitness of reactor core structures, and associated reactivity control devices, as an essential element in conducting a station life extension project. The ageing of reactors is affected by ageing mechanisms impacted by operational history and design related factors such as materials, chemistries and stress distributions. The methodology of this life extension work is based on the IAEA TECDOC 1197; which documents practices for ageing management in CANDU reactors. This paper uses the work in Bruce Units 1 and 2, conducted from 2007 through to 2011, to explain the methodology. The work started with analysis of historical operational conditions and identification of the forms of degradation that could have occurred. The assessment and related inspections considered the safety and pressure boundary significance of each item, as well as its failure modes and margins. It then moved through both general and local inspection, focused mainly inside the calandria vessel once the calandria tubes were removed. The inspection found the bulk of the hardware to be in good condition, with a small number of remediation opportunities. In the course of that remediation some foreign material was sampled and removed. The minor remediation was successful and the work was completed through formal documentation of the fitness for extended life. It has been demonstrated through these analyses and visual inspections that the reactor structures and components inspected are free of indications and active degradation mechanisms that would prevent the safe and reliable operation of Bruce A Units 1 and 2 through its next 25 years of life. (author)

CANDU advanced fuel cycles

International Nuclear Information System (INIS)

Slater, J.B.

1986-03-01

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

Fabrication of PWR fuel assembly and CANDU fuel bundle

International Nuclear Information System (INIS)

Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

1980-01-01

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

Life assurance of CANDU calandria and shield tank assemblies

Energy Technology Data Exchange (ETDEWEB)

Richards, T G; Novak, W Z [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

1996-12-31

This paper contains a re-assessment of the potential degradation mechanisms for the calandria shield tank assembly (CSTA). The assessments are made in support of the design of future stations. With few exceptions, the design of CANDU CSTA`s is such that a life of up to 60 years is readily attainable. Few degradation mechanisms have been identified that might require further analysis, design, or testing and inspection. The current calandria tube design, however, is one part of the CSTA which must be replaced before 60 years. Provisions will be made in future designs for either the replacement of calandria tubes at mid-life or the introduction of stiffer more sag resistant calandria tubes. (author). 1 tab., 3 figs.

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

International Nuclear Information System (INIS)

1977-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

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

MAAP4 CANDU analysis of a generic CANDU-6 plant: preliminary results

Energy Technology Data Exchange (ETDEWEB)

Petoukhov, S.M.; Mathew, P.M

2001-10-01

To support the generic probabilistic safety analysis (PSA) program at AECL, in particular to conduct Level 2 PSA analysis of a CANDU 6 plant undergoing a postulated severe accident, the capability to conduct severe accident consequence analysis for a CANDU plant is required. For this purpose, AECL selected MAAP4 CANDU from a number of other severe accident codes. The necessary models for a generic CANDU 6 station have been implemented in the code, and the code version 0.2 beta was tested using station data, which were assembled for a generic CANDU 6 station. This paper describes the preliminary results of the consequence analysis using MAAP4 CANDU for a generic CANDU 6 station, when it undergoes a station blackout and a large loss-of-coolant accident scenario. The analysis results show that the plant response is consistent with the physical phenomena modeled and the failure criteria used. The results also confirm that the CANDU design is robust with respect to severe accidents, which is reflected in the calculated long times that are available for administering accident management measures to arrest the accident progression before the calandria vessel or containment become at risk. (author)

Technical specifications and performance of CANDU fuel

International Nuclear Information System (INIS)

Sejnoha, R.

1997-01-01

The relations between Technical Specifications and fuel performance are discussed in terms of design limits and margins. The excellent performance record of CANDU reactor fuel demonstrates that the fuel design defined in the Technical Specifications (and with it other components of the procurement cycle, such as fuel manufacturing), satisfy the requirements. New requirements, changing conditions of fuel application and accumulating experience make periodic updates of the Technical Specifications necessary. Under the CANDU Owners Group (COG) Working Party 9, a Work Package has been conducted to support the review of the Specifications and the documentation of the rationales for their requirements. So far, the review has been completed for 4 Specifications: 1 for Zircaloy tubing, and 3 for uranium dioxide powder. It is planned to complete the review of all 11 currently used specifications by 1999. The paper summarizes the results achieved to mid 1997. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

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

International Nuclear Information System (INIS)

Catana, Alexandru; Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

2008-01-01

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

Nondestructive examination techniques on Candu fuel elements

International Nuclear Information System (INIS)

Gheorghe, G.; Man, I.

2013-01-01

During irradiation in nuclear reactor, fuel elements undergo dimensional and structural changes, and changes of surface conditions sheath as well, which can lead to damages and even loss of integrity. Visual examination and photography of Candu fuel elements are among the non-destructive examination techniques, next to dimensional measurements that include profiling (diameter, bending, camber) and length, sheath integrity control with eddy currents, measurement of the oxide layer thickness by eddy current techniques. Unirradiated Zircaloy-4 tubes were used for calibration purposes, whereas irradiated Zircaloy-4 tubes were actually subjected to visual inspection and dimensional measurements. We present results of measurements done by eddy current techniques on Zircaloy- 4 tubes, unirradiated, but oxidized in an autoclave prior to examinations. The purpose of these nondestructive examination techniques is to determine those parameters that characterize the behavior and performance of nuclear fuel operation. (authors)

CANDU 3000

International Nuclear Information System (INIS)

Keillor, Mac

1987-01-01

In this article, the CANDU 300 design, and the team that designed it, are featured. The CANDU 300 will operate at an energy cost similar to that of the larger CANDU units, but is sized for emerging markets. Ease of construction is an important feature: for example, full 360-degree access is available to each of the five buildings during construction; and the whole plant consists of about 90 modules, which can be built in separate locations, and hoisted into place

3-D CFD analysis of the CANDU-6 moderator circulation under normal operating conditions

International Nuclear Information System (INIS)

Yoon, Churl; Rhee, Bo Wook; Min, Byung Joo

2004-01-01

A computational fluid dynamics model for predicting moderator circulation inside the Canada Deuterium Uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard k-ε turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by CFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are 'mixed-type', as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is 82.9 deg. C at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is 24.8 .deg. C

CANDU 3 - Modularization

International Nuclear Information System (INIS)

McAskie, M.J.

1991-01-01

The CANDU 3 Heavy Water Reactor is the newest design developed by AECL CANDU. It has set as a major objective, the achievement of significant reductions in both cost and schedule over previous designs. The basic construction strategy is to incorporate extensive modularization of the plant in order to parallel the civil and mechanical installation works. This results in a target 38 month construction schedule from first concrete to in-service compared to 68 months for the Wolsong-1 CANDU 6 actually achieved and the 54 months envisaged for an improved CANDU 6. This paper describes the module concepts that have been developed and explains how they contribute to the overall construction program and achieve the desired cost and schedule targets set for the CANDU 3. (author). 7 figs, 2 tabs

TEM examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

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

2005-09-01

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

Next generation CANDU plants

International Nuclear Information System (INIS)

Hedges, K.R.; Yu, S.K.W.

1998-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

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

International Nuclear Information System (INIS)

Walker, J.R.

1990-02-01

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

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

International Nuclear Information System (INIS)

Lee, B. W.; Jung, C. J.; Min, B. J.; Yoon, H. J.; Choi, J. H.; Jang, D. S.

2002-01-01

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

Pressure tube reactors and a sustainable energy future: the ultra development path

International Nuclear Information System (INIS)

Duffey, R.

2008-01-01

Nuclear energy must be made available, freely and readily, to help meet world energy needs, concerns over energy price and security of supply, and alleviating the uncertainties over potential climate change. The perspective offered here is a model for others to consider, adopting and adapting using whatever elements fit their own strategies and needs. The underlying philosophy is to retain flexibility in the reactor development, deployment and fuel cycle, while ensuring the principle that customer, energy market, safety, non-proliferation and sustainability needs are all addressed. Canada is the world's largest exporter of uranium, providing about one-third of the world supply for nuclear power reactors. Pressure tube reactors (PTRs), of which CANDU is a prime example, have a major role to play in a sustainable energy future. The inherent fuel cycle flexibility of the PTR offers many technical, resource and sustainability, and economic advantages over other reactor technologies and is the subject of this paper. The design evolution and development intent is to be consistent with improved or enhanced safety, licensing and operating limits, global proliferation concerns, and waste stream reduction, thus enabling sustainable energy futures. The limits are simply those placed by safety, economics and resource availability. (author)

Pressure tube reactors and a sustainable energy future: the ultra development path

International Nuclear Information System (INIS)

Duffey, R.

2008-01-01

Nuclear energy must be made available, freely and readily, to help meet world energy needs, concerns over energy price and security of supply, and alleviating the uncertainties over potential climate change. The perspective offered here is a model for others to consider, adopting and adapting using whatever elements fit their own strategies and needs. The underlying philosophy is to retain flexibility in the reactor development, deployment and fuel cycle, while ensuring the principle that customer, energy market, safety, non-proliferation and sustainability needs are all addressed. Canada is the world's largest exporter of uranium, providing about one-third of the world supply for nuclear power reactors. Pressure tube reactors (PTRs), of which CANDU, is a prime example, have a major role to play in a sustainable energy future. The inherent fuel cycle flexibility of the PTR offers many technical, resource and sustainability and economic advantages over other reactor technologies and is the subject of this paper. The design evolution and development intent is to be consistent with improved or enhanced safety, licensing and operating limits, global proliferation concerns, and waste stream reduction, thus enabling sustainable energy futures. The limits are simply those placed by safety, economics and resource availability. (author)

Verification tests for CANDU advanced fuel -Development of the advanced CANDU technology-

International Nuclear Information System (INIS)

Chung, Jang Hwan; Suk, Ho Cheon; Jeong, Moon Ki; Park, Joo Hwan; Jeong, Heung Joon; Jeon, Ji Soo; Kim, Bok Deuk

1994-07-01

This project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year Out-of-pile hydraulic tests for the prototype of CANFLEX bundle was conducted in the CANDU-hot test loop at KAERI. Thermalhydraulic analysis with the assumption of CANFLEX-NU fuel loaded in Wolsong-1 was performed by using thermalhydraulic code, and the thermal margin and T/H compatibility of CANFLEX bundle with existing fuel for CANDU-6 reactor have been evaluated. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

Qinshan CANDU NPP outage performance improvement through benchmarking

International Nuclear Information System (INIS)

Jiang Fuming

2005-01-01

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

CANDU nuclear power system

International Nuclear Information System (INIS)

1981-01-01

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

Single-phase coolant flow CFD simulations inside the CANDU channel for the 37 and the 43 elements bundles

International Nuclear Information System (INIS)

Pauna, E.; Olteanu, G.; Catana, A.

2013-01-01

In this paper, a Computation Fluid Dynamics (CFD) simulation was performed in order to find the flow conditions in the CANDU Channel for the standard (37 elements) and the new designed bundle (43 elements) using the CFD Code S aturne software. Due to the fact that the code is a single-phase one it was considered an inlet temperature of 250 O C, a flow rate of 24.17 kg/s, an outlet pressure of 10.3 MPa and a linear power of 800 kW/m. The flow conditions were achieved by using a CFD typical chain of steps which was performed starting from preprocessing (geometry, mesh and boundary conditions), through solver and post-processing. Open Source platform (Salome-Meca geometry and mesh modules, the Code S aturne solver, Paraview and Visit for post-processing) were used as computational tool kit and an unsteady state was considered. Some simplifications were considered: the tube creep was not taken into account and all the bundles were considered aligned. The three dimensional thermal-hydraulic distributions of the temperature, pressure and velocity parameters offered information for the geometry comparison and the results were in agreement with some experimental data. CFD analysis results provided valuable data regarding the thermal-hydraulic operating conditions inside the CANDU reactor channel. (authors)

Comparison of neutron parameters between a CANDU and ACR reactors; Comparacao de parametros neutronicos entre um reator CANDU e um ACR

Energy Technology Data Exchange (ETDEWEB)

Dias, Gabriel H.P.; Silva, Clarysson A.M. da; Pereira, Claubia, E-mail: gabrielhpd@yahoo.com.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-11-01

CANDU (Canadian Deuterium Uranium) is a type of reactor that uses heavy water (D{sub 2}O) as a moderator and as a refrigerant. Despite having chemical characteristics similar to light water (H{sub 2}O), heavy water has a high moderation ratio for neutrons. This feature enables CANDU to use natural uranium as fuel. However, research has evaluated the possibility of using H{sub 2}O as a refrigerant and D{sub 2}O as a moderator aiming at reducing the volume of heavy water. Such changes would imply the use of lightly enriched uranium due to the presence of H{sub 2}O. In this context, the concept of ACR (Advanced CANDU Reactor) has been developed. This reactor has an innovative design which combines of the current CANDU with the characteristics of PWR (Pressurized Water Reactor) type reactors. Studies by AECL (Atomic Energy Canada Limited) show that compared to CANDU, the ACR presents a cost reduction in construction, improved firing performance, improved operation safety and longer life. The present work aims to evaluate, in steady state, some of the main neutron parameters of CANDU-6 and ACR-1000. The MCNPX 2.6.0 code (Monte Carlo N-Particle eXtended -version 2.6.0) was used to simulate such types of reactors. The results show that the models configured in the MCNPX adequately reproduce the neutron behavior of the studied reactors. These models may be used in future work for analysis of fuel burn and evolution.

Design and analysis of CANDU advanced fuel -Development of the advanced CANDU technology-

International Nuclear Information System (INIS)

Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Park, Kwang Seok; Kim, Bong Ki; Lee, Yeong Uk; Jeong, Chang Joon; Oh, Deok Joo; Lee, Ui Joo; Park, Joo Hwan; Lee, Sang Yong; Jeong, Beop Dong; Choi, Han Rim; Lee, Yeong Jin; Choi, Cheol Jin; Choi, Jong Ho; Lee, Kwang Won; Cho, Cheon Hyi; On, Myeong Ryong; Kim, Taek Mo; Lim, Hong Sik; Lee, Kang Moon; Lee, Nam Ho; Lee, Kyu Hyeong

1994-07-01

It has been projected that a total of 5 pressurized heavy water reactors (PHWR) including Wolsong 1 under operation and Wolsong 2, 3 and 4 under construction will be operated by 2006, and so about 500 ton of natural uranium will be consumed every year and a lot of spent fuels will be generated. Therefore, the ultimate goal of this R and D project is to develop the CANDU advanced fuel having the following capabilities compared with existing standard fuel: (1) To reduce linear heat generation rating by more than 15% (i.e., less than 50 kW/m), (2) To extend fuel burnup by more than 3 times (i.e., higher than 21,000 MWD/MTU), and (3) To increase critical channel power by more than 5%. In accordance, the followings are performed in this fiscal year: (1) Undertake CANFLEX-NU design and thermalmechanical performance analysis, and prepare design documents, (2) Establish reactor physics analysis code system, and investigate the compativility of the CANFLEX-NU fuel with the standard 37-element fuel in the CANDU-6 reactor. (3) Establish safety analysis methodology with the assumption of the CANFLEX-NU loaded CANDU-6 reactor, and perform the preliminary thermalhydraulic and fuel behavior for the selected DBA accidents, (4) Investigate reactor physics analysis code system as pre-study for CANFLEX-SEU loaded reactors

The cracking of pressure tubes in the Pickering reactor

International Nuclear Information System (INIS)

Ross-Ross, P.A.

1978-01-01

Small cracks in 17 of the 390 pressure tubes in Unit 3 of the 2056 MW (electrical) Pickering Generating Station and of 52 tubes in Unit 4, resulted in each of these units being out of service for many months. The cracks originated at areas of extremely high residual tensile stress produced by improper positioning of the rolling tool used during construction to join the pressure tube to its end-fitting. The mechanism of failure was delayed hydrogen cracking. (author)

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

International Nuclear Information System (INIS)

1989-01-01

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

The CANDU 80

International Nuclear Information System (INIS)

Hart, R.S.

1998-01-01

AECL has completed the conceptual design of a small CANDU plant with an output, in the range of 300 MWth (called the CANDU 80), suitable for a variety of electrical and co-generation applications including desalination, oil sands oil extraction and processing, and the provision of electricity and heat to areas with low demand. This paper provides a brief overview of the CANDU 80, and discusses key features contributing to safety and operational margins

Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

International Nuclear Information System (INIS)

Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

2004-01-01

As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

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

International Nuclear Information System (INIS)

Sim, Woo Gun

2013-01-01

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

CANDU lectures

International Nuclear Information System (INIS)

Rouben, B.

1984-06-01

This document is a compilation of notes prepared for two lectures given by the author in the winter of 1983 at the Institut de Genie Nucleaire, Ecole Polytechnique, Montreal. The first lecture gives a physical description of the CANDU reactor core: the nuclear lattice, the reactivity mechanisms, their functions and properties. This lecture also covers various aspects of reactor core physics and describes different calculational methods available. The second lecture studies the numerous facets of fuel management in CANDU reactors. The important variables in fuel management, and the rules guiding the refuelling strategy, are presented and illustrated by means of results obtained for the CANDU 600

Periodic inspection of CANDU nuclear power plant containment components

International Nuclear Information System (INIS)

1989-09-01

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

A statistical approach to the prediction of pressure tube fracture toughness

International Nuclear Information System (INIS)

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

2008-01-01

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

Degradation of Alloy 800 steam generator tubing and its long-term behaviour predictions for plant life management

International Nuclear Information System (INIS)

Lu, Y.C.; Tapping, R.L.; Pandey, M.D.

2009-01-01

Alloy 800 tubing has a good service record in steam generators (SGs) in both German pressurized water reactors and CANDU 6 reactors, however, a recent comprehensive examination of several ex-service SG tubes removed from Darlington Nuclear Generating Station (DNGS) found that these SG tubes (which had experienced shallow pitting in service) were more susceptible to pitting corrosion in laboratory tests than a reference nuclear grade Alloy 800 tubing under SG crevice chemistry conditions. This was an unexpected finding and has raised questions about possible effects of in-service 'aging' on SG tubing. In addition, there has also been recent evidence that a few Alloy 800 tubes have experienced stress corrosion cracking (SCC) in some German pressurized water reactors (PWRs), possibly after many years of degradation-free service, although the inspection history of these tubes is not available to confirm that the reported degradation initiated recently. These findings suggest that Alloy 800 tubing may have some aging degradation susceptibility after many years of service. To provide support for a proactive SG aging management, a survey on the corrosion susceptibility of the archived Alloy 800 tubing from CANDU SGs under plausible crevice chemistry conditions was conducted to assess the potential material degradation issues in CANDU SGs. Experimental work was also performed to investigate the root cause leading to Alloy 800 SG tubing degradation. The results from this study suggested that a combination of negative factors; aggressive chemistry resulting from impurity ingress into the secondary side of the SGs, elevated electrochemical corrosion potential (ECP) during SG transients and surface strain/plastic deformation, might have led to the degradation of the ex-service SG tubing. The studies have shown that each of these conditions in isolation does not cause degradation of Alloy 800 SG tubing; a synergistic combination of factors is required. The OPEX and experimental

Moderator circulation in CANDU reactors

International Nuclear Information System (INIS)

Fath, H.E.S.; Hussein, M.A.

1989-01-01

A two-dimensional computer code that is capable of predicting the moderator flow and temperature distribution inside CANDU calandria is presented. The code uses a new approach to simulate the calandria tube matrix by blocking the cells containing the tubes in the finite difference mesh. A jet momentum-dominant flow pattern is predicted in the nonisothermal case, and the effect of the buoyancy force, resulting from nuclear heating, is found to enhance the speed of circulation. Hot spots are located in low-velocity areas at the top of the calandria and below the inlet jet level between the fuel channels. A parametric study is carried out to investigate the effect of moderator inlet velocity,moderator inlet nozzle location, and geometric scaling. The results indicate that decreasing the moderator inlet velocity has no significant influence on the general features of the flow pattern (i.e., momentum dominant); however, too many high-temperature hot spots appear within the fuel channels

ANL/CANTIA code for steam generator tube integrity assessment

International Nuclear Information System (INIS)

Revankar, S.T.; Wolf, B.; Majumdar, S.; Riznic, J.R.

2009-01-01

Steam generator (SG) tubes have an important safety role in CANDU type reactors and Pressurized Water Reactors (PWR) because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear plant. The SG tubes are susceptible to corrosion and damage. A failure of a single steam generator tube, or even a few tubes, would not be a serious safety-related event in a CANDU reactor. The leakage from a ruptured tube is within makeup capacity of the primary heat transport system, so that as long as the operator takes the correct actions, the off-site consequences will be negligible. A sufficient safety margin against tube rupture used to be the basis for a variety of maintenance strategies developed to maintain a suitable level of plant safety and reliability. Several through-wall flaws may remain in operation and potentially contribute to the total primary-to-secondary leak rate. Assessment of the conditional probabilities of tube failures, leak rates, and ultimately risk of exceeding licensing dose limits has been used for steam generator tube fitness-for-service assessment. The advantage of this type of analysis is that it avoids the excessive conservatism typically present in deterministic methodologies. However, it requires considerable effort and expense to develop all of the failure, leakage, probability of detection, and flaw growth distributions and models necessary to obtain meaningful results from a probabilistic model. The Canadian Nuclear Safety Commission (CNSC) recently developed the CANTIA methodology for probabilistic assessment of inspection strategies for steam generator tubes as a direct effect on the probability of tube failure and primary-to-secondary leak rate Recently Argonne National Laboratory has developed tube integrity and leak rate models under Integrated Steam Generator Tube Integrity Program (ISGTIP-2). These models have been incorporated in the ANL/CANTIA code. This paper presents the ANL

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-15

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

The next generation CANDU 6

International Nuclear Information System (INIS)

Hopwood, J.M.

1999-01-01

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

Prediction of CANDU-6 moderator system response following a large break LOCA using a 3D model

Energy Technology Data Exchange (ETDEWEB)

De, T K; Collins, W M; Holmes, R W [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

CANDU nuclear reactors use D{sub 2}0 as a moderator inside the calandria vessel. Heat is generated in the calandria by neutron and gamma radiation from nuclear fission. During normal operating condition, hot moderator fluid is continuously pumped out from the bottom of the CANDU-6 calandria. After passing through a heat exchanger, the cooled moderator fluid is returned to the calandria through inlet nozzles. In the unlikely event of a loss of coolant accident (LOCA) the moderator acts as a heat sink. To predict moderator system response following a large break (reactor inlet header break) LOCA, a simulation was undertaken for Class IV power available (i.e. main moderator pump running) as well as for Class IV power unavailable during the LOCA. 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 during the inlet header break. The 3D code PHOENICS2 developed by CHAM U.K was used for the simulation. The results show an asymmetric flow pattern within the moderator both in the axial Z-direction of the calandria as well as in the X-Y plane. The temperature distribution within the moderator system shows, that hot spots are generated in areas, where the flow approaches stagnation. Hot spot temperatures are higher with Class IV power unavailable. (author). 1 ref., 12 figs.

Prediction of CANDU-6 moderator system response following a large break LOCA using a 3D model

International Nuclear Information System (INIS)

De, T.K.; Collins, W.M.; Holmes, R.W.

1995-01-01

CANDU nuclear reactors use D 2 0 as a moderator inside the calandria vessel. Heat is generated in the calandria by neutron and gamma radiation from nuclear fission. During normal operating condition, hot moderator fluid is continuously pumped out from the bottom of the CANDU-6 calandria. After passing through a heat exchanger, the cooled moderator fluid is returned to the calandria through inlet nozzles. In the unlikely event of a loss of coolant accident (LOCA) the moderator acts as a heat sink. To predict moderator system response following a large break (reactor inlet header break) LOCA, a simulation was undertaken for Class IV power available (i.e. main moderator pump running) as well as for Class IV power unavailable during the LOCA. 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 during the inlet header break. The 3D code PHOENICS2 developed by CHAM U.K was used for the simulation. The results show an asymmetric flow pattern within the moderator both in the axial Z-direction of the calandria as well as in the X-Y plane. The temperature distribution within the moderator system shows, that hot spots are generated in areas, where the flow approaches stagnation. Hot spot temperatures are higher with Class IV power unavailable. (author). 1 ref., 12 figs

Steam generator tube failures

International Nuclear Information System (INIS)

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

1996-04-01

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

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

International Nuclear Information System (INIS)

Choi, Hangbok; Ko, Won Il; Yang, Myung Seung

2001-01-01

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

Consequences of pressure tube rupture on in-core components

International Nuclear Information System (INIS)

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

1982-12-01

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

The CANDU 3 containment structure

International Nuclear Information System (INIS)

1994-01-01

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

CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

International Nuclear Information System (INIS)

Charak, I.; Kier, P.H.

1995-04-01

Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule (section 50.62); (2) station blackout (section 50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term (section 50.34(f) and section 100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements (section 50.55a, etc); (6) ECCS acceptance criteria (section 50.46)(b); (7) combustible gas control (section 50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle (section 51.51); and (11) (standards section 50.55a)

The final report of ''on-the-job training'' on the CANDU reactor

International Nuclear Information System (INIS)

Kim, D.H.; Koh, B.J.

1983-01-01

This is the final Report for the technical ''on-the-job traning'' for the Wolsung CANDU nuclear power plant which is the first Pressurized Heavy Water Reactor setting up in Korea. The technical ''on-the-job traning'' was established to increase the capability for the nuclear safety evaluation in order to contribute the future safe operation of the CANDU nuclear power plant. The training has been excuted through three level courses as elementary, intermediate and ''on-the-job training'' at Wolsung power plant. The elementary course was introduction to the CANDU basics and fundamentals. The intermediate course was the more advanced course, and the detailed concepts and engineering explanations of the CANDU system had been instructed. The third course was the ''on-the-job training'' at the Wolsung plant site, which was the most emphasized course during the project. (Author)

Thermal hydraulic simulation of the CANDU nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Carvalho, Athos M.S.S. de; Ramos, Mario C.; Costa, Antonella L.; Fernandes, Gustavo H.N., E-mail: athos1495@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Rio de janeiro, RJ (Brazil)

2017-07-01

The CANDU (Canada Deuterium Uranium) is a Canadian-designed power reactor of PHWR type (Pressurized Heavy Water Reactor) that uses heavy water (deuterium oxide) for moderator and coolant, and natural uranium for fuel. There are about 47 reactors of this type in operation around the world generating more than 23 GWe, highlighting the importance of this kind of device. In this way, the main purpose of this study is to develop a thermal hydraulic model for a CANDU reactor to aggregate knowledge in this line of research. In this way, a core modeling was performed using RELAP5-3D code. Results were compared with reference data to verify the model behavior in steady state operation. Thermal hydraulic parameters as temperature, pressure and mass flow rate were verified and the results are in good agreement with reference data, as it is being presented in this work. (author)

Status of advanced technologies for CANDU reactors

International Nuclear Information System (INIS)

Lipsett, J.J.

1989-01-01

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

Passive safety features for next generation CANDU power plants

International Nuclear Information System (INIS)

Natalizio, A.; Hart, R.S.; Lipsett, J.J.; Soedijono, P.; Dick, J.E.

1989-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

The CANDU 9 fuel transfer system

International Nuclear Information System (INIS)

Keszthelyi, Z.G.; Morikawa, D.T.

1996-01-01

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

The CANDU 9 fuel transfer system

Energy Technology Data Exchange (ETDEWEB)

Keszthelyi, Z G [Canadian General Electric Co. Ltd., Peterborough, ON (Canada); Morikawa, D T [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1997-12-31

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

Team CANDU : ready for the marketplace

International Nuclear Information System (INIS)

Howieson, J.Q.

2007-01-01

This paper outlines the partnership between AECL and a number of leading global nuclear suppliers to market the Candu power reactor. The mission of the CANDU team is to develop market opportunities for CANDU technology and deliver successful CANDU projects

Cost and schedule reduction for next-generation Candu

International Nuclear Information System (INIS)

Hopwood, J.M.; Yu, S.; Pakan, M.; Soulard, M.

2002-01-01

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

Crystalline texture study of Zr-2.5%Nb pressure pipes by DRX pole figures

International Nuclear Information System (INIS)

Buioli, C P; Banchik, A D; Vizcaino, P; Samper, R; Testone, S

2012-01-01

This work presents the study of crystalline texture of Zr-2.5%Nb pressure tubes (A, B and C); structural components of CANDU nuclear power reactors [1]. The study of texture was made using the RX diffraction technique, making measurements of direct pole figures [2], in a Phillips diffractometer with pole goniometer (IFIR-CONICET). The texture was determined calculating the coefficients of JJ Kearns [3]. The samples used correspond to transversal sections of the tubes, front and back, and were characterized making de measurements of five pole figures with Miller index (001), (100), (101), (102) and (110), in order to represent in a complete way the texture in the material. The calculated Kearns coefficients were compared with the specifications given by the designer of pressure tubes AECL [4] (author)

Effect of DUPIC cycle on CANDU reactor safety parameters

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-09-15

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

Parametric study of moderator heat exchanger for Candu 6 advanced reactor

International Nuclear Information System (INIS)

Umar, Efrizon; Vecchiarelli, Jack

2000-01-01

The passive moderator system for Candu 6 advanced reactor require moderator heat exchanger with the small size and the low resistance coefficient of the shell-side. The study is to determine the required size of moderator heat exchanger, and to calculate the shell side of resistance coefficient have been done. Using computer code CATHENA, it is concluded that the moderator heat exchanger can be used at full power-normal operation condition, especially for the cases with 3600 to 8100 number of tube and 15.90 mm tube diameter. This study show that the proposed moderator heat exchanger have given satisfactory results

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Development and performance of inspection equipment for pressure tubes in Fugen

International Nuclear Information System (INIS)

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

1984-01-01

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

Joint studies on large CANDU

International Nuclear Information System (INIS)

Lee, Ikhwan; Yu, S. K. W.

1994-01-01

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

The next generation of CANDU technologies: profiling the potential for hydrogen fuel

International Nuclear Information System (INIS)

Hopwood, J.M.

2001-01-01

This report discusses the Next-generation CANDU Power Reactor technologies currently under development at AECL. The innovations introduced into proven CANDU technologies include a compact reactor core design, which reduces the size by a factor of one third for the same power output; improved thermal efficiency through higher-pressure steam turbines; reduced use of heavy water (one quarter of the heavy water required for existing plants), thus reducing the cost and eliminating many material handling concerns; use of slightly enriched uranium to extend fuel life to three times that of existing natural uranium fuel and additions to CANDU's inherent passive safety. With these advanced features, the capital cost of constructing the plant can be reduced by up to 40 per cent compared to existing designs. The clean, affordable CANDU-generated electricity can be used to produce hydrogen for fuel cells for the transportation sector, thereby reducing emissions from the transportation sector

Fission products distributions in Candu primary heat transport and Candu containment systems during a severe accident

International Nuclear Information System (INIS)

Constantin, Marin; Rizoiu, Andrei

2005-01-01

The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code (Accident Source Term Evaluation Code). The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivations to begin with a simplified geometry in order to avoid the introducing of unmanageable errors at the level of input deck. Thus only 1/4 of the PHT circuit was simulated and a simplified FPs inventory, some simplifications in the feeders geometry and containment were used. The circuit consists of 95 horizontal fuel channels connected to 95 horizontal out-feeders, then through vertical feeders to the outlet-header (a big pipe that collects the water from feeders); the circuit continues from the outlet-header with a riser and then with the steam generator and a pump. After this pump, the circuit was broken; in this point the FPs are transferred to the containment. The containment model consists of 4 rooms connected between by 6 links. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU NPP loss of coolant accident sequence. Temperature and pressure conditions in the time of the accident were calculated by the CATHENA code and the source term of FPs introduced into the PHT was estimated by the ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species are obtained by using SOPHAEROS module of ASTEC code. The distributions into the containment are obtained by the CPA module of ASTEC code (thermalhydraulics calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts and

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

International Nuclear Information System (INIS)

Hong, Joo Bo

1994-01-01

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

Pressure heat pumping in the orifice pulse-tube refrigerator

International Nuclear Information System (INIS)

Boer, P.C.T. de

1996-01-01

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

Development of nuclear quality high pressure valve bellows in Canada

International Nuclear Information System (INIS)

Janzen, P.; Astill, C.J.

1978-06-01

Concurrent with the decision to use bellows stem sealed nuclear valves where feasible in commercial-scale CANDU plants, AECL undertook to develop an indigenous high pressure valve bellows technology. This program included developing the capability to fabricate improved high pressure valve bellows in conjunction with a Canadian manufacturer. This paper describes the evolution of a two-stage bellows fabrication process involving: (1) manufacture of discrete lengths of precision thin wall telescoping tubes - from preparation of strip blanks through edge grinding and edge forming to longitudinal welding; (2) forming of bellows from tube assemblies using a novel combination of mechanical inward forming followed by hydraulic outward forming. Bellows of Inconel 600 and Inconel 625 have been manufactured and evaluated. Test results indicate comparable to improved performance over alternative high quality bellows. (author)

The Thermal-hydraulic Analysis for the Aging Effect of the Component in CANDU-6 Reactor

International Nuclear Information System (INIS)

Bae, Jun Ho; Jung, Jong Yeob

2014-01-01

CANDU reactor consists of a lot of components, including pressure tube, reactor pump, steam generator, feeder pipe, and so on. These components become to have the aging characteristics as the reactor operates for a long time. The aging phenomena of these components lead to the change of operating parameters, and it finally results to the decrease of the operating safety margin. Actually, due to the aging characteristics of components, CANDU reactor power plant has the operating license for the duration of 30 years and the plant regularly check the plant operating state in the overhaul period. As the reactor experiences the aging, the reactor operators should reduce the reactor power level in order to keep the minimum safety margin, and it results to the deficit of economical profit. Therefore, in order to establish the safety margin for the aged reactor, the aging characteristics for components should be analyzed and the effect of aging of components on the operating parameter should be studied. In this study, the aging characteristics of components are analyzed and revealed how the aging of components affects to the operating parameter by using NUCIRC code. Finally, by scrutinizing the effect of operating parameter on the operating safety margin, the effect of aging of components on the safety margin has been revealed

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

DEFF Research Database (Denmark)

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

2010-01-01

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

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

DEFF Research Database (Denmark)

Gaihede, Michael; Jacobsen, Henrik; Tveterås, Kjell

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

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

International Nuclear Information System (INIS)

Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

1985-07-01

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

On random pressure pulses in the turbine draft tube

Science.gov (United States)

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

2017-04-01

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

Multi-purpose use of the advanced CANDU compact simulator

International Nuclear Information System (INIS)

Lam, K.Y.; MacBeth, M.J.

1997-01-01

A near full-scope dynamic model of a CANDU-PHWR (Canadian Deuterium Uranium Pressurized Heavy Water) nuclear power plant was constructed as a multi-purpose advanced Compact Simulator using CASSIM (Cassiopeia Simulation) development system. This Compact Simulator has played an integral part in the design and verification of the CANDU 900 MW control centre mock-up located in the Atomic Energy of Canada (AECL) design office, providing CANDU plant process dynamic data to the Plant Display System (PDS) and the Distributed Control System (DCS), as well as mock-up panel devices. As a design tool, the Compact Simulator is intended to be used for control strategy development, human factors studies, analysis of overall plant control performance, tuning estimates for major control loops. As a plant commissioning and operational strategy development tool, the simulation is intended to be used to evaluate routine and non-routine operational procedures, practice 'what-if' scenarios for operational strategy development, practice malfunction recovery procedures and verify human factors activities