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Sample records for calandrias

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  3. Inspection of calandria front area of Wolsung NPP using technique of mapping thermal infrared image into CCD image

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Kim, Chang Hoi; Seo, Yong Chil; Choi, Young Soo; Kim, Seung Ho [Advance Robotics Teams, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2002-11-15

    This paper describes the enhanced inspection performance of a thermal infrared camera for monitoring abnormal conditions of calandria reactor area of Wolsung nuclear power plant. Thermal infrared camera have poor image qualities compared to commercial CCD cameras, as in contrast, brightness, and resolution. To compensate the poor image quality problems associated with the thermal infrared camera, the technique of mapping thermal infrared image into real ccd image is proposed. The mobile robot KAEROT/m2, loaded with sensor head system at the mast, is entered to monitor leakage of heavy water and thermal abnormality of the calandria reactor area in overhaul period. The sensor head system is composed of thermal infrared camera and ccd camera in parallel. When thermal abnormality on observation points and areas of calandria reactor area is occurred, unusual hot image taken from thermal infrared camera is superimposed on real CCD image. In this inspection experiment, more accurate positions of thermal abnormalities on calandria reactor area can be estimated by using technique of mapping thermal infrared image into CCD image, which include characters arranged in MPOQ order.

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

    Science.gov (United States)

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

    2015-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dureja, A.K., E-mail: akdureja@barc.gov.in [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Pawaskar, D.N.; Seshu, P. [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Mumbai (India); Sinha, S.K. [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Sinha, R.K. [Department of Atomic Energy, OYC, Near Gateway of India, Mumbai (India)

    2015-04-01

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

  6. The use of OPEX (in the form of inspection data) to obtain unanticipated Calandria Tube (CT) ovality measurements

    Energy Technology Data Exchange (ETDEWEB)

    Sedran, P.J. [Canadian Power Utility Services Limited, Toronto, Ontario (Canada); Rankin, B. [NB Power Nuclear Incorporated, Saint John, New Brunswick (Canada)

    2011-07-01

    For fuel channel deformation studies, inspections are an essential source of OPEX and are used to generate specific data, such as Pressure Tube (PT) dimensions. However, additional unintended information can be extracted from raw inspection data. This paper presents examples of the generation of Calandria Tube (CT) inner diameter profiles using data from the inspections of L1F06 and G2H14 to illustrate how inspection/OPEX data can contain significant hidden information. In the CT diameter profiles for L1F06 and G2H14, it was found that spacer loading produced significant local ovality in the CT and that the resultant gap reduction is consistent with the predictions of the existing deformation models. The finding of significant local CT ovality is corroborated by CT gauging measurements reported by S.A. Donahue in 2008. (author)

  7. Experimental investigation of heat transfer during severe accident of a Pressurized Heavy Water Reactor with simulated decay heat generation in molten pool inside calandria vessel

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, Sumit Vishnu, E-mail: svprasad@barc.gov.in; Nayak, Arun Kumar, E-mail: arunths@barc.gov.in

    2016-07-15

    Highlights: • Scaled test facility simulating the calandria vessel and calandria vault water of PHWR with simulated decay heat was built. • Experiments conducted with simulant material at about 1200 °C. • Experimental result shows that melt coolability and growth rate of crust thickness are affected by presence of decay heat. • No gap was observed between the crust and vessel on opening. • Result shows that vessel integrity is intact with presence of water inside water tank in both cases. - Abstract: The present study focuses on experimental investigation in a scaled facility of an Indian PHWR to investigate the coolability of molten corium with simulated decay heat in the simulated calandria vessel. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60 kg of the molten material was poured into the test section at about 1200 °C. Decay heat in the melt pool was simulated using four high watt heaters cartridges, each having 9.2 kW. The temperature distributions inside the molten pool, across the vessel wall thickness and vault water were measured. Experimental results obtained are compared with the results obtained previously for no decay heat case. The results indicated that presence of decay heat seriously affects the coolability behaviour and formation of crust in the melt pool. The location and magnitude of maximum heat flux and surface temperature of the vessel also are affected in the presence of decay heat.

  8. 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)

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

    Science.gov (United States)

    Kim, Hyoung Tae

    2014-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

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

  13. Development of NRU reflector wall inspection system

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-06-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-15

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

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

    Science.gov (United States)

    Rochefort, P. A.

    2009-02-01

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

  16. Improvement of core degradation model in ISAAC

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Ha; Kim, See Darl; Park, Soo Yong

    2004-02-01

    If water inventory in the fuel channels depletes and fuel rods are exposed to steam after uncover in the pressure tube, the decay heat generated from fuel rods is transferred to the pressure tube and to the calandria tube by radiation, and finally to the moderator in the calandria tank by conduction. During this process, the cladding will be heated first and ballooned when the fuel gap internal pressure exceeds the primary system pressure. The pressure tube will be also ballooned and will touch the calandria tube, increasing heat transfer rate to the moderator. Although these situation is not desirable, the fuel channel is expected to maintain its integrity as long as the calandria tube is submerged in the moderator, because the decay heat could be removed to the moderator through radiation and conduction. Therefore, loss of coolant and moderator inside and outside the channel may cause severe core damage including horizontal fuel channel sagging and finally loss of channel integrity. The sagged channels contact with the channels located below and lose their heat transfer area to the moderator. As the accident goes further, the disintegrated fuel channels will be heated up and relocated onto the bottom of the calandria tank. If the temperature of these relocated materials is high enough to attack the calandria tank, the calandria tank would fail and molten material would contact with the calandria vault water. Steam explosion and/or rapid steam generation from this interaction may threaten containment integrity. Though a detailed model is required to simulate the severe accident at CANDU plants, complexity of phenomena itself and inner structures as well as lack of experimental data forces to choose a simple but reasonable model as the first step. ISAAC 1.0 was developed to model the basic physicochemical phenomena during the severe accident progression. At present, ISAAC 2.0 is being developed for accident management guide development and strategy evaluation. In

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

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N

    2000-09-01

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

  18. Current safety issues of CANDU licensing

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. [University of Ottawa, Ottawa (Canada); Natalizio, A. [ENSAC Associates, Ontario (Canada)

    1994-01-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  1. ACR-1000 design provisions for severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca

    2006-07-01

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As a further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving enhanced safety features, shorter construction schedule, high plant capacity factor, improved operations and maintenance, and increased operating life. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. The ACR-1000 design meets Canadian regulatory requirements and follows established international practice with respect to severe accident prevention and mitigation. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. While maintaining existing structures of CANDU reactors that provide inherent prevention and retention of core debris, the ACR-1000 design includes additional features for prevention and mitigation of severe accidents. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel is designed for debris retention by minimizing penetrations at the bottom periphery and by accommodating thermal and weight loads of the core debris. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes

  2. Fission-product releases from a PHWR terminal debris bed

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Bailey, D.G., E-mail: morgan.brown@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Anca Abati, R. de

    1966-07-01

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

  4. Reactor core design and characteristics of the Fugen

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Mitsuo; Kowata, Yasuki; Sugawara, Satoru; Deshimaru, Takehide

    1988-03-01

    The heavy water moderated, boiling light water cooled pressure tube type reactor Fugen uses plutonium-uranium mixed oxide as a fuel. Heavy water as the moderator and the light water of coolant are separated by the pressure tubes and calandria tubes. Thereby, the reactor core is heterogenes compared with that of LWRs. This paper describes the development of reactor core design procedure based on the feature of the Fugen type reactor, the feasibility test and the validity of nuclear and thermalhydraulic design based on the operating experience.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dubey, J.S., E-mail: jsdubey@barc.gov.in; Shriwastaw, R.S.; Kumar, Ashwini; Shah, Priti Kotak; Rath, B.N.; Kumar, Sunil; Mishra, Prerna; Alur, V.D.; Mallik, G.K.; Anantharaman, S.

    2015-07-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Che, K. M.; Yoo, S. Y. [Chungnam National Univ., Taejon (Korea, Republic of); Lee, B. W.; Choi, H. B. [KAERI, Taejon (Korea, Republic of)

    2000-10-01

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

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

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

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

  11. Fitness for service assessment of coolant channels of Indian PHWRs

    Science.gov (United States)

    Sinha, R. K.; Sinha, S. K.; Madhusoodanan, K.

    2008-12-01

    A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 × 10 17 n/m 2 s, E > 1 MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr-2.5 wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

  13. Fitness for service assessment of coolant channels of Indian PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, R.K.; Sinha, S.K. [Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Madhusoodanan, K. [Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: kmadhu@barc.gov.in

    2008-12-15

    A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 x 10{sup 17} n/m{sup 2} s, E > 1 MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr-2.5 wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-11-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-04-15

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

  16. Remote field eddy current testing

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, C.E

    2001-12-01

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

  19. Asset management program

    Energy Technology Data Exchange (ETDEWEB)

    Wison, P.; Newman, G. [Bruce Power, Tiverton, Ontario (Canada)

    2013-07-01

    In order to understand our assets we have been assessing the condition of the units in our nuclear power plants developing asset life management options on a component by component basis. We have concluded that with the right work and planning we will be able to manage the units in a way that balances capacity requirements over the long term and at the same time manage the demand on critical resources. Major component replacement outages include Installing/removing bulkheads, pressure tube and calandria tube replacement, feeder replacement, steam generator replacement, supporting facilities and infrastructure, reactor inspections and maintenance including tooling enhancements, additional non reactor systems inspection & testing and continued research and analysis. These plans will have to take into account cost, resource and capacity requirements.

  20. Electromagnetic modeling for gap measurement between nuclear fuel channel and liquid injection nozzle

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D. H.; Heo, H.; Jeong, H. G. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    Fuel channels including Pressure Tube(PT) and Calandria Tube(CT) are important components of Pressurized Heavy Water Reactor(PHWR). A sagging for fuel channel increases by heat and radiation exposure with the increasing operating time. The possibility of contact to Liquid Injection Nozzle(LIN) is thus a critical issue in power plant safety. In order to solve this safety issue, electromagnetic technique was applied to compliment the ultrasonic technology. Electromagnetic fields were investigated for the gap measurement between CT and LIN using computer modeling. We calculated the electromagnetic fields, such as, magnetic flux density, current density near the fuel channel and simulated an impedance and a phase angle in receiving coil for obtaining the optimal inspection parameters, such as, frequency, inter-coil spacing, coil size and configuration. This paper shows that the simulated eddy current signals in variance with the CT/LIN gap can be used for baseline data of experimental electromagnetic technique.

  1. Electromagnetic field analysis and RFEC signal modeling for gap measurement between liquid injection nozzle and nuclear fuel channel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Hoon; Jung, Hyun Kyu; Cheong, Yong Moo; Huh, Young; Lee, Yoon Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2004-05-15

    Fuel channels including pressure tube(Pt) and calandria tube(CT) are important components of pressurized heavy water reactor(PHWR). A sagging of fuel channel increases by heat and radiation exposure with the increasing operation time. The contact of fuel channel to liquid injection nozzle(LIN) is thus a critical issue in power plant safety. In order to solve this safety issue, the electromagnetic technique was applied to compliment the present inspection technology. Electromagnetic fields were investigated for the gap measurement between CT and LIN using FEM computer modeling. We calculated the electromagnetic fields, such as, magnetic flux density, current density near the fuel channel and checked the adaptability of RFEC technology. The RFEC Signals using the volume integral method(VIM) were simulated for obtaining the optimal inspection parameters, including frequency, inter-coil spacing, coil size and configuration. Finally, we development the remote field eddy current sensor that can CT/LIN gap measurement efficiently.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  4. Examination of Dodd and Deeds solutions for a transmit-receive eddy current probe above a layered planar structure

    Science.gov (United States)

    Luloff, Mark S.; Morelli, Jordan; Krause, Thomas W.

    2017-02-01

    Exact solutions for the electromagnetic response of a transmit-receive coil pair situated above two parallel plates separated by a gap, were developed using the recently published general model of Desjardins et al. that accounts for all electromagnetic interactions between the voltage-driven probe and the conducting samples. This model was then compared to the well-known model developed by Dodd and Deeds, which assumes a constant amplitude sinusoidal current and an open-circuit pick-up coil. Both models were compared with experimental results that measured the gap profile for a Grade 2 Titanium plate (54 μΩ.cm) over a SS-316 stainless steel second layer plate (74 μΩ.cm). These materials simulated the electromagnetic properties of a Zr 2.5% Nb pressure tube and Zr-2 calandria tube, respectively, as found in the fuel channels of CANDU® reactors. It was observed that while the Dodd and Deeds' model as applied to this work achieved a good shape agreement with experimental data for excitation frequencies at 2 kHz, at the higher frequency of 16 kHz good agreement was not achieved. In contrast, the model of Desjardin's et al., adapted for a transmit-receive probe configuration above two infinite flat plates, achieved an excellent shape agreement at both frequencies.

  5. Development of CANDU pressure tube integrity evaluation system

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-05-01

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

  6. Comparison of analytical eddy current models using principal components analysis

    Science.gov (United States)

    Contant, S.; Luloff, M.; Morelli, J.; Krause, T. W.

    2017-02-01

    Monitoring the gap between the pressure tube (PT) and the calandria tube (CT) in CANDU® fuel channels is essential, as contact between the two tubes can lead to delayed hydride cracking of the pressure tube. Multifrequency transmit-receive eddy current non-destructive evaluation is used to determine this gap, as this method has different depths of penetration and variable sensitivity to noise, unlike single frequency eddy current non-destructive evaluation. An Analytical model based on the Dodd and Deeds solutions, and a second model that accounts for normal and lossy self-inductances, and a non-coaxial pickup coil, are examined for representing the response of an eddy current transmit-receive probe when considering factors that affect the gap response, such as pressure tube wall thickness and pressure tube resistivity. The multifrequency model data was analyzed using principal components analysis (PCA), a statistical method used to reduce the data set into a data set of fewer variables. The results of the PCA of the analytical models were then compared to PCA performed on a previously obtained experimental data set. The models gave similar results under variable PT wall thickness conditions, but the non-coaxial coil model, which accounts for self-inductive losses, performed significantly better than the Dodd and Deeds model under variable resistivity conditions.

  7. ASTEC adaptation for PHWR limited core damage accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, P., E-mail: pmajum@barc.gov.in [Bhabha Atomic Research Centre, Reactor Safety Division, Mumbai 400085 (India); Chatterjee, B.; Lele, H.G. [Bhabha Atomic Research Centre, Reactor Safety Division, Mumbai 400085 (India); Guillard, G.; Fichot, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, 13115 Saint-Paul-lez-Durance (France)

    2014-06-01

    Under limited core damage accidents (LCDAs) of Pressurized Heavy Water Reactor (PHWR), coolable geometry of the channel might be retained thanks to the presence of moderator heat sink. Indeed, the pressure tube is amenable to creep deformation at high temperature due to internal pressure and fuel bundles weight. Partial or complete circumferential contact between pressure tube and calandria tube aids heat dissipation to the moderator. A new module has been developed by Bhabha Atomic Research Centre (BARC) for simulating this phenomenon which is specific to horizontal-type of reactors. It requires additional calculation of pressure tube sagging/ballooning and temperature field in the circumferential direction. The module is well validated with available experimental results concerning pressure tube deformation and the associated heat transfer in the area of contact. It is then used in analysing typical LCDAs scenarios in Indian PHWR under low and medium internal pressure conditions. This module is implemented in the ASTEC IRSN-GRS severe accident code version under development and will thus be available in the next major version V2.1.

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

    Directory of Open Access Journals (Sweden)

    Jon‐Ariza De‐Miguel

    2014-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-15

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

  10. Radiological Characteristics of decommissioning waste from a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  11. La lumière et le temps sur la scène baroque : Poetique & Pratique

    Directory of Open Access Journals (Sweden)

    Françoise Siguret

    2016-06-01

    Full Text Available Time: Aristotle, in the Poetics, recommends the playwright to confine his tragedy within «two revolutions of the sun»; the concept refers to the light perception, to the fact that greek drama is acted in the open air. The messengers and the chorus represented on the stage, in the present time, what happened outside of it. In the age of the French classical theatre, the chronological sequence of the action had to conform to the laws of the reason: the so-called rule of the twenty-four hours became an indisputable rule of the action. A time exactly measured, substituting the time of the light, cyclical and mythical. In Italy, pastorals, mythological melodramas and all that belonged to the court entertainments (ballets, operas, tournaments conformed to a cyclical time in which the four seasons constituted the scenery, linking life to the four liturgical seasons and to the four parts of the day, from noon to midnight (cfr. Endymion and the Ballet de la Nuit. Light: Need to light up the indoor playhouse for practical and moral issues. Italian craftsmen implement the technical tools; a certain difference between primary light (intended to light up the stage and the auditorium and the lumi (the supplementary lighting related to a specific performance. Buontalenti’s lighting devices (sun, moon, rainbows, divine and princely splendour will enchant the spectators. France will discover these stagecraft effects with the Calandria (1548, without subsequent developments. Afterwards, Corneille will be fascinated by the ‘baroque’ charm (Médée, 1639 and Andromède, 1650. In the second half of the XVIIth century, while machinery invades opera and tragedy in music, Racine refuses anything intended to deceive the eye, though creates a lighting that may be «listened» (Britannicus. The Allegories (the «other discourse» convey meanings on the baroque stage through the perpetual slow motion of the gods and Time, till the final glory of the Prince: Cosimo

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-02-15

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

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

    Directory of Open Access Journals (Sweden)

    Carrascal, L. M.

    2006-05-01

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