Calculation of the Thermal Radiation Benchmark Problems for a CANDU Fuel Channel Analysis Using the CFX-10 Code

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

2006-07-15

To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to whole fuel channel geometry. With assumptions of a non-participating medium completely enclosed with the diffuse, gray and opaque surfaces, the solutions of the benchmark problems are obtained by the concept of surface resistance to radiation accounting for the view factors and the emissivities. The view factors are calculated by the program MATRIX version 1.0 avoiding the difficulty of hand calculation for the complex geometries. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer.

Calculation of the Thermal Radiation Benchmark Problems for a CANDU Fuel Channel Analysis Using the CFX-10 Code

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

2006-07-01

To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to whole fuel channel geometry. With assumptions of a non-participating medium completely enclosed with the diffuse, gray and opaque surfaces, the solutions of the benchmark problems are obtained by the concept of surface resistance to radiation accounting for the view factors and the emissivities. The view factors are calculated by the program MATRIX version 1.0 avoiding the difficulty of hand calculation for the complex geometries. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer

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

CANDU fuel-cycle vision

International Nuclear Information System (INIS)

Boczar, P.G.

1999-01-01

The fuel-cycle path chosen by a particular country will depend on a range of local and global factors. The CANDU reactor provides the fuel-cycle flexibility to enable any country to optimize its fuel-cycle strategy to suit its own needs. AECL has developed the CANFLEX fuel bundle as the near-term carrier of advanced fuel cycles. A demonstration irradiation of 24 CANFLEX bundles in the Point Lepreau power station, and a full-scale critical heat flux (CHF) test in water are planned in 1998, before commercial implementation of CANFLEX fuelling. CANFLEX fuel provides a reduction in peak linear element ratings, and a significant enhancement in thermalhydraulic performance. Whereas natural uranium fuel provides many advantages, the use of slightly enriched uranium (SEU) in CANDU reactors offers even lower fuel-cycle costs and other benefits, such as uprating capability through flattening the channel power distribution across the core. Recycled uranium (RU) from reprocessing spent PWR fuel is a subset of SEU that has significant economic promise. AECL views the use of SEU/RU in the CANFLEX bundle as the first logical step from natural uranium. High neutron economy enables the use of low-fissile fuel in CANDU reactors, which opens up a spectrum of unique fuel-cycle opportunities that exploit the synergism between CANDU reactors and LWRs. At one end of this spectrum is the use of materials from conventional reprocessing: CANDU reactors can utilize the RU directly without re-enrichment, the plutonium as conventional Mixed-oxide (MOX) fuel, and the actinide waste mixed with plutonium in an inert-matrix carrier. At the other end of the spectrum is the DUPIC cycle, employing only thermal-mechanical processes to convert spent LWR fuel into CANDU fuel, with no purposeful separation of isotopes from the fuel, and possessing a high degree of proliferation resistance. Between these two extremes are other advanced recycling options that offer particular advantages in exploiting the

CANDU fuel-cycle vision

International Nuclear Information System (INIS)

Boczar, P.G

1998-05-01

The fuel-cycle path chosen by a particular country will depend on a range of local and global factors. The CANDU reactor provides the fuel-cycle flexibility to enable any country to optimize its fuel-cycle strategy to suit its own needs. AECL has developed the CANFLEX fuel bundle as the near-term carrier of advanced fuel cycles. A demonstration irradiation of 24 CANFLEX bundles in the Point Lepreau power station, and a full-scale critical heat flux (CHF) test in water are planned in 1998, before commercial implementation of CANFLEX fuelling. CANFLEX fuel provides a reduction in peak linear element ratings, and a significant enhancement in thermalhydraulic performance. Whereas natural uranium fuel provides many advantages, the use of slightly enriched uranium (SEU) in CANDU reactors offers even lower fuel-cycle costs and other benefits, such as uprating capability through flattening the channel power distribution across the core. Recycled uranium (RU) from reprocessing spent PWR fuel is a subset of SEU that has significant economic promise. AECL views the use of SEU/RU in the CANFLEX bundle as the first logical step from natural uranium. High neutron economy enables the use of low-fissile fuel in CANDU reactors, which opens up a spectrum of unique fuel-cycle opportunities that exploit the synergism between CANDU reactors and LWRs. At one end of this spectrum is the use of materials from conventional reprocessing: CANDU reactors can utilize the RU directly without reenrichment, the plutonium as conventional mixed-oxide (MOX) fuel, and the actinide waste mixed with plutonium in an inert-matrix carrier. At the other end of the spectrum is the DUPIC cycle, employing only thermal-mechanical processes to convert spent LWR fuel into CANDU fuel, with no purposeful separation of isotopes from the fuel, and possessing a high degree of proliferation resistance. Between these two extremes are other advanced recycling options that offer particular advantages in exploiting the

Fuel deposits, chemistry and CANDU reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2013-01-01

'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU reactor, the first being the Nuclear Power Demonstration-2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channel led to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5, and subsequently utilized for each CANDU unit since. The difference being that during 'hot conditioning' of CANDU heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

Candu reactors with thorium fuel cycles

International Nuclear Information System (INIS)

Hopwood, J.M.; Fehrenbach, P.; Duffey, R.; Kuran, S.; Ivanco, M.; Dyck, G.R.; Chan, P.S.W.; Tyagi, A.K.; Mancuso, C.

2006-01-01

Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 550 0 C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a

A fast-running fuel management program for a CANDU reactor

International Nuclear Information System (INIS)

Choi, Hangbok

2000-01-01

A fast-running fuel management program for a CANDU reactor has been developed. The basic principle of this program is to select refueling channels such that the reference reactor conditions are maintained by applying several constraints and criteria when selecting refueling channels. The constraints used in this program are the channel and bundle power and the fuel burnup. The final selection of the refueling channel is determined based on the priority of candidate channels, which enhances the reactor power distribution close to the time-average model. The refueling simulation was performed for a natural uranium CANDU reactor and the results were satisfactory

An optimum fuel management method based on CANDU in-core detector readings

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok

2001-01-01

In this study, a new optimal fuel management method is developed for a CANDU 600 MWe (CANDU-6) reactor. At first, an efficient power mapping method has been developed, which provides an accurate core status of an operating CANDU reactor. Secondly, an optimum refueling channel selection method has been developed by an optimization theory. For the power mapping method, the measured detector readings are used as boundary conditions of the diffusion theory calculation with the Kalman filtering (DIKAL) method. The performance of the DIKAL method was assessed for various core states and applied to the calculation of power and flux distribution in the CANDU 6 reactor. Sensitivity studies have shown that DIKAL method is insensitive to the detector random and systematic errors. An optimal refueling simulation method (OPTIMA), practically applicable to a CANDU 6 reactor, has also been developed. The objective of the optimization is to reproduce the reference core performance during refueling simulation, while satisfying the operation limits of channel and bundle powers. The optimization process consists of two stages: i) elimination of candidate refueling channels by several constraints and ii) selection of refueling channels by a direct search method that uses sensitivity coefficients of channel power generated for the reference core. The elimination process sorts out an appropriate number of fuel channels suitable for refueling, considering the channel power, bundle power and fuel burnup. The optimum refueling channels are then selected such that the difference of power distribution from the reference is minimized. In order to demonstrate the applicability of the overall fuel management methodology developed in this study, the DIKAL-OPTIMA method was applied to Wolsong-3 reactor refueling simulation, which is a typical CANDU-6 reactor. The results of refueling simulation have shown that the method can be efficiently used for the performance analysis of the operating

An optimum fuel management method based on CANDU in-core detector readings

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Choi, Hang Bok

2001-01-01

In this study, a new optimal fuel management method is developed for a CANDU 600 MWe (CANDU-6) reactor. At first, an efficient power mapping method has been developed, which provides an accurate core status of an operating CANDU reactor. Secondly, an optimum refueling channel selection method has been developed by an optimization theory. For the power mapping method, the measured detector readings are used as boundary conditions of the diffusion theory calculation with the Kalman filtering (DIKAL) method. The performance of the DIKAL method was assessed for various core states and applied to the calculation of power and flux distribution in the CANDU 6 reactor. Sensitivity studies have shown that DIKAL method is insensitive to the detector random and systematic errors. An optimal refueling simulation method (OPTIMA), practically applicable to a CANDU 6 reactor, has also been developed. The objective of the optimization is to reproduce the reference core performance during refueling simulation, while satisfying the operation limits of channel and bundle powers. The optimization process consists of two stages: i) elimination of candidate refueling channels by several constraints and ii) selection of refueling channels by a direct search method that uses sensitivity coefficients of channel power generated for the reference core. The elimination process sorts out an appropriate number of fuel channels suitable for refueling, considering the channel power, bundle power and fuel burnup. The optimum refueling channels are then selected such that the difference of power distribution from the reference is minimized. In order to demonstrate the applicability of the overall fuel management methodology developed in this study, the DIKAL-OPTIMA method was applied to Wolsong-3 reactor refueling simulation, which is a typical CANDU-6 reactor. The results of refueling simulation have shown that the method can be efficiently used for the performance analysis of the operating

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

SARAPAN—A Simulated-Annealing-Based Tool to Generate Random Patterned-Channel-Age in CANDU Fuel Management Analyses

Directory of Open Access Journals (Sweden)

Doddy Kastanya

2017-02-01

Full Text Available In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the *SIMULATE module of the Reactor Fueling Simulation Program (RFSP code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the *INSTANTAN module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the *INSTANTAN module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

SARAPAN-A simulated-annealing-based tool to generate random patterned-channel-age in CANDU fuel management analyses

Energy Technology Data Exchange (ETDEWEB)

Kastanya, Doddy [Safety and Licensing Department, Candesco Division of Kinectrics Inc., Toronto (Canada)

2017-02-15

In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium) utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the *SIMULATE module of the Reactor Fueling Simulation Program (RFSP) code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the *INSTANTAN module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the *INSTANTAN module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

1200 FPD refuelling simulation of RUFIC fuel in a CANDU 6 reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Soon Young; Jeong, Chang Joon; Min, Byung Joo; Suk, Ho Chun

2001-07-01

The refuelling strategy of RUFIC (Recovered Uranium Fuel in CANDU) fuel as a high-burnup fuel for a CANDU 6 reactor is studied to determine the achievable operation characteristics of the fuel and reactor. In this study, three refuelling schemes of 4-, 2-, and 3-bundle shift for 0.92 w/o RUFIC fuel in an CANDU 6 reactor were individually evaluated through 1200 FPD(Full Power Day)refuelling simulaltions where the 0.92 w/o RUFIC is equivalent to CANFLEX 0.9 w/o SEU(Slightly Enriched Uranium) in reactivity and burnup respects. The computer code system used for this study is WIMS-AECL/DRAGON/RFSP. The results simulated for the case of 4-bundle shift refueling scheme shows that the peak maximum channel power and peak maximum CPPF(Channel Power Peaking Factor)of 7228 kW and 1.175, respectively, seems too high to maintain the available operating margins, because some data of the maximum channel power exceed the operating limit(7070 kW based on the Technical Specifications of Wolsong 3 and 4 Units). Whereas, the results simulated for the case of 2-bundle shift refuelling scheme shows that sufficient operating margin could be secured where the peak maximum channel power and peak maximum CPPF were 6889 kW and 1.094, respectively. However, the channel refuelling rate (channels/day) of the 2-bundle shift refuelling scheme is twice that of the 4-bundle shift refuelling scheme, and hence the 2-bundle shift refuelling would not be an economical refuelling scheme for the RUFIC fuel bundles. Therefore, a 3-bundle shift refuelling scheme for the RUFIC fuel in CANDU 6 reactor was also studied by the 1200 FPD refuelling simulation. As a result, it is found that all the operating parameters in the 3-bundle shift case are achivable for the CANDU 6 reactor operation, and the channel refuelling rate of 2.88 channels/day seems to be attractive compared to the refuelling rate of 4.32 channels/day in the 2-bundle shift case.

CANDU reactor experience: fuel performance

International Nuclear Information System (INIS)

Truant, P.T.; Hastings, I.J.

1985-07-01

Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

Axial shuffling fuel-management schemes for 1.2% SEU in CANDU

International Nuclear Information System (INIS)

Younis, M.H.; Boczar, P.G.

1989-11-01

The use of slightly enriched uranium (SEU) in CANDU (CANada Deuterium Uranium) requires a different fuel-management strategy than that usually employed with natural uranium fuel. Axial shuffling is a fuel-management strategy in which some or all of the fuel bundles are removed from the channel, rearranged, and reinserted into the same channel, along with fresh fuel. An axial shuffling scheme has been devised for 1.2% SEU which results in excellent power profiles, from the perspectives of both good axial flattening and power histories. With the CANFLEX (CANdu FLEXible fuelling) advanced fuel bundle, fuel rating can be reduced to below 40kW/m, with consequent safety benefits

Homogeneous Thorium Fuel Cycles in Candu Reactors

Energy Technology Data Exchange (ETDEWEB)

Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)

2009-06-15

The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy

Application of Shuttle Remote Manipulator System technology to the replacement of fuel channels in the Pickering CANDU reactor

International Nuclear Information System (INIS)

Stratton, D.; Butt, C.

1982-04-01

Spar Aerospace Limited of Toronto was the prime contractor to the National Research Council of Canada for the design and development of the Shuttle Remote Manipulator (SRMS). Spar is presently under contract to Ontario Hydro to design and build a Remote Manipulation Control System to replace the fuel channels in the Pickering A Nuclear Generating Station. The equipment may be used to replace the fuel channels in six other early generation CANDU reactors

Advanced CANDU reactors fuel analysis through optimal fuel management at approach to refuelling equilibrium

International Nuclear Information System (INIS)

Tingle, C.P.; Bonin, H.W.

1999-01-01

The analysis of alternate CANDU fuels along with natural uranium-based fuel was carried out from the view point of optimal in-core fuel management at approach to refuelling equilibrium. The alternate fuels considered in the present work include thorium containing oxide mixtures (MOX), plutonium-based MOX, and Pressurised Water Reactor (PWR) spent fuel recycled in CANDU reactors (Direct Use of spent PWR fuel in CANDU (DUPIC)); these are compared with the usual natural UO 2 fuel. The focus of the study is on the 'Approach to Refuelling Equilibrium' period which immediately follows the initial commissioning of the reactor. The in-core fuel management problem for this period is treated as an optimization problem in which the objective function is the refuelling frequency to be minimized by adjusting the following decision variables: the channel to be refuelled next, the time of the refuelling and the number of fresh fuel bundles to be inserted in the channel. Several constraints are also included in the optimisation problem which is solved using Perturbation Theory. Both the present 37-rod CANDU fuel bundle and the proposed CANFLEX bundle designs are part of this study. The results include the time to reach refuelling equilibrium from initial start-up of the reactor, the average discharge burnup, the average refuelling frequency and the average channel and bundle powers relative to natural UO 2 . The model was initially tested and the average discharge burnup for natural UO 2 came within 2% of the industry accepted 199 MWh/kgHE. For this type of fuel, the optimization exercise predicted the savings of 43 bundles per full power year. In addition to producing average discharge burnups and other parameters for the advanced fuels investigated, the optimisation model also evidenced some problem areas like high power densities for fuels such as the DUPIC. Perturbation Theory has proven itself to be an accurate and valuable optimization tool in predicting the time between

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)

Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

CANDU-6 fuel optimization for advanced cycles

Energy Technology Data Exchange (ETDEWEB)

St-Aubin, Emmanuel, E-mail: emmanuel.st-aubin@polymtl.ca; Marleau, Guy, E-mail: guy.marleau@polymtl.ca

2015-11-15

Highlights: • New fuel selection process proposed for advanced CANDU cycles. • Full core time-average CANDU modeling with independent refueling and burnup zones. • New time-average fuel optimization method used for discrete on-power refueling. • Performance metrics evaluated for thorium-uranium and thorium-DUPIC cycles. - Abstract: We implement a selection process based on DRAGON and DONJON simulations to identify interesting thorium fuel cycles driven by low-enriched uranium or DUPIC dioxide fuels for CANDU-6 reactors. We also develop a fuel management optimization method based on the physics of discrete on-power refueling and the time-average approach to maximize the economical advantages of the candidates that have been pre-selected using a corrected infinite lattice model. Credible instantaneous states are also defined using a channel age model and simulated to quantify the hot spots amplitude and the departure from criticality with fixed reactivity devices. For the most promising fuels identified using coarse models, optimized 2D cell and 3D reactivity device supercell DRAGON models are then used to generate accurate reactor databases at low computational cost. The application of the selection process to different cycles demonstrates the efficiency of our procedure in identifying the most interesting fuel compositions and refueling options for a CANDU reactor. The results show that using our optimization method one can obtain fuels that achieve a high average exit burnup while respecting the reference cycle safety limits.

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

Energy Technology Data Exchange (ETDEWEB)

Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

This is the `94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author).

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

International Nuclear Information System (INIS)

Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh

1995-07-01

This is the '94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author)

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.

Fuel management simulation for CANFLEX-RU in CANDU 6

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

Fuel management simulations have been performed for CANFLEX-09% RU fuel in the CANDU 6 reactor. In this study, the bi-directional 4-bundle shift fuelling scheme was assumed. The lattice cell and time-average calculation were carried out. The refuelling simulation calculations were performed for 600 full power days. Time-averaged results show good axial power profile with the CANFLEX-RU fuel. During the simulation period, the maximum channel and bundle power were maintained below the licensing limit of CANDU 6 reactor. 7 refs., 4 figs. (Author)

Fuel management simulation for CANFLEX-RU in CANDU 6

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

Fuel management simulations have been performed for CANFLEX-09% RU fuel in the CANDU 6 reactor. In this study, the bi-directional 4-bundle shift fuelling scheme was assumed. The lattice cell and time-average calculation were carried out. The refuelling simulation calculations were performed for 600 full power days. Time-averaged results show good axial power profile with the CANFLEX-RU fuel. During the simulation period, the maximum channel and bundle power were maintained below the licensing limit of CANDU 6 reactor. 7 refs., 4 figs. (Author)

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)

CANDU fuel cycle options in Korea

International Nuclear Information System (INIS)

Boczar, P.G.; Fehrenbach, P.J.; Meneley, D.A.

1996-04-01

The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a twoto three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than does conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U.S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or FBR reactors. If the objective of a national fuel-cycle program is the minimization of actinide waste or destruction of long-lived fission products, then studies have shown the superiority of CANDU reactors in meeting this objective. Long-term energy security can be assured either through the thorium cycle or through a CANDU 1 FBR system, in which the FBR would be operated as a 'fuel factory,' providing the fissile material to power a number of lower-cost, high efficiency CANDU reactors. In summary, the CANDU reactor's simple fuel design, high neutron economy, and on

Thorium-Based Fuels Preliminary Lattice Cell Studies for Candu Reactors

International Nuclear Information System (INIS)

Margeanu, C.A.; Rizoiu, A.C.

2009-01-01

The choice of nuclear power as a major contributor to the future global energy needs must take into account acceptable risks of nuclear weapon proliferation, in addition to economic competitiveness, acceptable safety standards, and acceptable waste disposal options. Candu reactors offer a proven technology, safe and reliable reactor technology, with an interesting evolutionary potential for proliferation resistance, their versatility for various fuel cycles creating premises for a better utilization of global fuel resources. Candu reactors impressive degree of fuel cycle flexibility is a consequence of its channel design, excellent neutron economy, on-power refueling, and simple fuel bundle. These features facilitate the introduction and exploitation of various fuel cycles in Candu reactors in an evolutionary fashion. The main reasons for our interest in Thorium-based fuel cycles have been, globally, to extend the energy obtainable from natural Uranium and, locally, to provide a greater degree of energy self-reliance. Applying the once through Thorium (OTT) cycle in existing and advanced Candu reactors might be seen as an evaluative concept for the sustainable development both from the economic and waste management points of view. Two Candu fuel bundles project will be used for the proposed analysis, namely the Candu standard fuel bundle with 37 fuel elements and the CANFLEX fuel bundle with 43 fuel elements. Using the Canadian proposed scheme - loading mixed ThO 2 -SEU CANFLEX bundles in Candu 6 reactors - simulated at lattice cell level led to promising conclusions on operation at higher fuel burnups, reduction of the fissile content to the end of the cycle, minor actinide content reduction in the spent fuel, reduction of the spent fuel radiotoxicity, presence of radionuclides emitting strong gamma radiation for proliferation resistance benefit. The calculations were performed using the lattice codes WIMS and Dragon (together with the corresponding nuclear data

CANDU fuel cycle options in Korea

International Nuclear Information System (INIS)

Boczar, P. G.; Fehrenbach, P. J.; Meneley, D. A.

1996-01-01

There are many reasons for countries embarking on a CANDU R program to start with the natural uranium fuel cycle. Simplicity of fuel design, ease of fabrication, and ready availability of natural uranium all help to localize the technology and to reduce reliance on foreign technology. Nonetheless, at some point, the incentives for using natural uranium fuel may be outweighed by the advantages of alternate fuel cycles. The excellent neutron economy, on-line refuelling, and simple fuel-bundle design provide an unsurpassed degree of fuel-cycle flexibility in CANDU reactors. The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a two- to three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than dose conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U. S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or

Pressure drop variation as a function of axial and radial power distribution in CANDU fuel channel with standard and CANFLEX 43 bundles

International Nuclear Information System (INIS)

Catana, Alexandru; Department of Energy Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

2007-01-01

CANDU 600 nuclear reactors are usually fuelled with STANDARD (STD), 37 rods fuel bundles. Natural uranium (NU) dioxide (UO 2 ), is used as fuel composition. A new fuel bundle geometry called CANFLEX (CFX) with 43 rods is proposed and some new fuel composition are considered. Flexibility is the key word for the attempt to use some different fuel geometries and compositions for CANDU 600 nuclear reactors as well as for innovative ACR-700/1000 nuclear reactors. The fuel bundle considered in this paper is CFX-RU-0.90 that encodes the CANFLEX geometry, recycled dioxide uranium (RU) with 0.90% enrichment. The goal of this proposal is ambitious: a higher average discharge burn-up up to 14000 MWd/tU and, for the same amount of generated electric power, reduction in nuclear fuel fabrication, reduction of spent nuclear fuel radioactive waste and reduction of refueling operational work by using fewer bundles. An improved sub-channel approach for thermal-hydraulic analysis is used in this paper to compute some flow parameters, mainly the pressure drop along the CANDU 600 fuel channel when STD or CFX-RU-0.90 fuel bundles. Also an intermediate CFX-NU fuel bundle are used, for gradual comparison. For CFX-RU- 0.90 four fuel bundle shift refueling scheme is used instead of eight, that will determine different axial power distributions. At the same time radial power distribution is affected by the geometry and by the fuel composition of fuel bundle type used. Some other thermal-hydraulic flow parameters will be influenced, too. One of the most important parameter is pressure drop (PD) along the fuel channel because of its importance in drag force evaluation. We start with an axial power distribution, which is characteristic for a refueling scheme of eight or four fuel bundles on a shift. Comparative results are presented between STD37, CFX-NU CFX-RU-0.90 fuel bundles in a CANDU nuclear reactor operating conditions. Neutron flux distribution analysis shows that four bundle shift

An Evaluation of a Fission Product Inventory for CANDU Fuels

International Nuclear Information System (INIS)

Jung, Jong Yeob; Park, Joo Hwan

2007-01-01

Fission products are released by two processes when a single channel accident occurs. One is a 'prompt release' and the other is a 'delayed release'. Prompt release assumes that the gap inventory of the fuel elements is released by a fuel element failure at the time of an accident. Delayed release assumes that the inventories within the grain or at the grain boundary are released after a accident due to a diffusion through grains, an oxidation of the fuel and an interaction between the fuel and the Zircaloy sheath. Therefore, the calculation of a fission product inventory and its distribution in a fuel during a normal operating is the starting point for the assessment of a fission product release for single channel accidents. In this report, the fission product inventories and their distributions within s fuel under a normal operating condition are evaluated for three types of CANDU fuels such as the 37 element fuel, CANFLEX-NU and CANFLEX-RU fuel bundles in the 'limiting channel'. To accomplish the above mentioned purposes, the basic power histories for each type of CANDU fuel were produced and the fission product inventories were calculated by using the ELESTRES code

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

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)

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)

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

CANDU fuel bundle deformation modelling with COMSOL multiphysics

International Nuclear Information System (INIS)

Bell, J.S.; Lewis, B.J.

2012-01-01

Highlights: ► The deformation behaviour of a CANDU fuel bundle was modelled. ► The model has been developed on a commercial finite-element platform. ► Pellet/sheath interaction and end-plate restraint effects were considered. ► The model was benchmarked against the BOW code and a variable-load experiment. - Abstract: A model to describe deformation behaviour of a CANDU 37-element bundle has been developed under the COMSOL Multiphysics finite-element platform. Beam elements were applied to the fuel elements (composed of fuel sheaths and pellets) and endplates in order to calculate the bowing behaviour of the fuel elements. This model is important to help assess bundle-deformation phenomena, which may lead to more restrictive coolant flow through the sub-channels of the horizontally oriented bundle. The bundle model was compared to the BOW code for the occurrence of a dry-out patch, and benchmarked against an out-reactor experiment with a variable load on an outer fuel element.

CANDU-PHW fuel management

International Nuclear Information System (INIS)

Frescura, G.M.; Wight, A.L.

1982-01-01

This report covers the material presented in a series of six lectures at the Winter College on Nuclear Physics and Reactors held at the International Centre for Theoretical Physics in Trieste, Italy, Jan 22 - March 28, 1980. The report deals with fuel management in natural uranium fuelled CANDU-PHW reactors. Assuming that the reader has a basic knowledge of CANDU core physics, some of the reactor systems which are more closely related to fuelling are described. This is followed by a discussion of the methods used to calculate the power distribution and perform fuel management analyses for the equilibrium core. A brief description of some computer codes used in fuel management is given, together with an overview of the calculations required to provide parameters for core design and support the accident analysis. Fuel scheduling during approach to equilibrium and equilibrium is discussed. Fuel management during actual reactor operation is discussed with a review of the operating experience for some of the Ontario Hydro CANDU reactors. (author)

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

INR Recent Contributions to Thorium-Based Fuel Using in CANDU Reactors

International Nuclear Information System (INIS)

Prodea, I.; Mărgeanu, C. A.; Rizoiu, A.; Olteanu, G.

2014-01-01

The paper summarizes INR Pitesti contributions and latest developments to the Thorium-based fuel (TF) using in present CANDU nuclear reactors. Earlier studies performed in INR Pitesti revealed the CANDU design potential to use Recovered Uranium (RU) and Slightly Enriched Uranium (SEU) as alternative fuels in PHWRs. In this paper, we performed both lattice and CANDU core calculations using TF, revealing the main neutron physics parameters of interest: k-infinity, coolant void reactivity (CVR), channel and bundle power distributions over a CANDU 6 reactor core similar to that of Cernavoda, Unit 1. We modelled the so called Once Through Thorium (OTT) fuel cycle, using the 3D finite-differences DIREN code, developed in INR. The INR flexible SEU-43 bundle design was the candidate for TF carrying. Preliminary analysis regarding TF burning in CANDU reactors has been performed using the finite differences 3D code DIREN. TFs showed safety features improvement regarding lower CVRs in the case of fresh fuel use. Improvements added to the INR ELESIMTORIU- 1 computer code give the possibility to fairly simulate irradiation experiments in INR TRIGA research reactor. Efforts are still needed in order to get better accuracy and agreement of simulations to the experimental results. (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

CANDU flexible and economical fuel technology in China

Energy Technology Data Exchange (ETDEWEB)

Mingjun, C. [CNNC Nuclear Power Operation Management Co., Zhejiang (China); Zhenhua, Z.; Zhiliang, M. [CNNC Third Qinshan Nuclear Power Co., Zhejiang (China); Cottrell, C.M.; Kuran, S. [Candu Energy Inc., Mississauga, ON (Canada)

2014-07-01

Use in CANDU reactor is one good option of recycled uranium (RU) and thorium (Th) resource. It is also good economy to CANDU fuel. Since 2008 Qinshan CANDU Plant and our partners (Candu Energy and CNNC and NPIC) have made great efforts to develop the engineering technologies of Flexible and Economical Fuel (RU and Th) in CANDU type reactor and finding the CANDU's position in Chinese closed fuel cycle (CFC) system. This paper presents a proposal of developing strategy and implementation plan. Qinshan CANDU reactors will be converted to use recycled and depleted uranium based fuels, a first-of-its-kind. The fuel is composed of both recycled and depleted uranium and simulating natural uranium behavior. This paper discusses its development, design, manufacture and verification tested with success and the full core implementation plan by the end of 2014. (author)

Used fuel packing plant for CANDU fuel

Energy Technology Data Exchange (ETDEWEB)

Menzies, I.; Thayer, B.; Bains, N., E-mail: imenzies@atsautomation.com [ATS Automation, Cambridge, ON (Canada); Murchison, A., E-mail: amurchison@nwmo.ca [NWMO, Toronto, ON (Canada)

2015-07-01

Large forgings have been selected to containerize Light Water Reactor used nuclear fuel. CANDU fuel, which is significantly smaller in size, allows novel approaches for containerization. For example, by utilizing commercially available extruded ASME pipe a conceptual design of a Used Fuel Packing Plant for containerization of used CANDU fuel in a long lived metallic container has been developed. The design adopts a modular approach with multiple independent work cells to transfer and containerize the used fuel. Based on current technologies and concepts from proven industrial systems, the Used Fuel Packing Plant can assemble twelve used fuel containers per day considering conservative levels of process availability. (author)

Canadian CANDU fuel development program and recent fuel operating experience

International Nuclear Information System (INIS)

Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

1999-01-01

This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the 2-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These 2 programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

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)

Flow behaviour in a CANDU horizontal fuel channel from stagnant subcooled initial conditions

International Nuclear Information System (INIS)

Caplan, M.Z.; Gulshani, P.; Holmes, R.W.; Wright, A.C.D.

1984-01-01

The flow behaviour in a CANDU primary system with horizontal fuel channels is described following a small inlet header break. With the primary pumps running, emergency coolant injection is in the forward direction so that the channel outlet feeders remain warmer than the inlet thereby promoting forward natural circulation. However, the break force opposes the forward driving force. Should the primary pumps run down after the circuit has refilled, there is a break size for which the natural circulation force is balanced by the break force and channels could, theoretically, stagnate. Result of visualization and of full-size channel tests on channel flow behaviour from an initially stagnant channel condition are discussed. After a channel stagnation, the decay power heats the coolant to saturation. Steam is then formed and the coolant stratifies. The steam expands into the subcooled water in the end fitting in a chugging type of flow regime due to steam condensation. After the end fitting reaches the saturation temperature, steam is able to penetrate into the vertical feeder thereby initiating a large buoyancy induced flow which refills the channel. The duration of stagnation is shown to be sensitive to small asymmetries in the initial conditions. A small initial flow can significantly shorten the occurrence and/or duration of boiling as has been confirmed by reactor experience. (author)

CANDU-9/480-SEU fuel handling system assessment document

International Nuclear Information System (INIS)

Hwang, Jeong Ki; Jo, C. H.; Kim, H. M.; Morikawa, D. T.

1996-11-01

This report summarize the rationale for the CANDU 9 fuel handling system, and the design choices recommended for components of the system. Some of the design requirements applicable to the CANDU 9 480-SEU fuel handling design choices are described. These requirements imposed by the CANDU 9 project. And the design features for the key components of fuel handling system, such as the fuelling machine, the carriage, the new fuel transfer system and the irradiated fuel transfer system, are described. The carriage seismic load evaluations relevant to the design are contained in the appendices. The majority of the carriage components are acceptable, or will likely be acceptable with some redesign. The concept for the CANDU 9 fuel handling system is based on proven CANDU designs, or on improved CANDU technology. Although some development work must be done, the fuel handling concept is judged to be feasible for the CANDU 9 480-SEU reactor. (author). 2 refs

The Canadian CANDU fuel development program and recent fuel operating experience

International Nuclear Information System (INIS)

Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

1999-01-01

This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the two-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These two programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

Fuel cycles - a key to future CANDU success

International Nuclear Information System (INIS)

Kuran, S.; Hopwood, J.; Hastings, I.J.

2011-01-01

Globally, fuel cycles are being evaluated as ways of extending nuclear fuel resources, addressing security of supply and reducing back-end spent-fuel management. Current-technology thermal reactors and future fast reactors are the preferred platform for such fuel cycle applications and as an established thermal reactor with unique fuel-cycle capability, CANDU will play a key role in fulfilling such a vision. The next step in the evolution of CANDU fuel cycles will be the introduction of Recovered Uranium (RU), derived from conventional reprocessing. A low-risk RU option applicable in the short term comprises a combination of RU and Depleted Uranium (DU), both former waste streams, giving a Natural Uranium Equivalent (NUE) fuel. This option has been demonstrated in China, and all test bundles have been removed from the Qinshan 1 reactor. Additionally, work is being done on an NUE full core, a Thorium demonstration irradiation and an Advanced Fuel CANDU Reactor(AFCR). AECL is developing other fuel options for CANDU, including actinide waste burning. AECL has developed the Enhanced CANDU 6 (EC6) reactor, upgraded from its best-performing CANDU 6 design. High neutron economy, on-power refueling and a simple fuel bundle provide the EC6 with the flexibility to accommodate a range of advanced fuels, in addition to its standard natural uranium. (author)

Fuel-management simulations for once-through thorium fuel cycle in CANDU reactors

International Nuclear Information System (INIS)

Chan, P.S.W.; Boczar, P.G.; Ellis, R.J.; Ardeshiri, F.

1999-01-01

High neutron economy, on-power refuelling and a simple fuel bundle design result in unsurpassed fuel cycle flexibility for CANDU reactors. These features facilitate the introduction and exploitation of thorium fuel cycles in existing CANDU reactors in an evolutionary fashion. Detailed full-core fuel-management simulations concluded that a once-through thorium fuel cycle can be successfully implemented in an existing CANDU reactor without requiring major modifications. (author)

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

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)

Future fuel cycle development for CANDU reactors

International Nuclear Information System (INIS)

Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

1987-01-01

The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

Numerical simulator of the CANDU fueling machine driving desk

International Nuclear Information System (INIS)

Doca, Cezar

2008-01-01

As a national and European premiere, in the 2003 - 2005 period, at the Institute for Nuclear Research Pitesti two CANDU fueling machine heads, no.4 and no.5, for the Nuclear Power Plant Cernavoda - Unit 2 were successfully tested. To perform the tests of these machines, a special CANDU fueling machine testing rig was built and was (and is) available for this goal. The design of the CANDU fueling machine test rig from the Institute for Nuclear Research Pitesti is a replica of the similar equipment operating in CANDU 6 type nuclear power plants. High technical level of the CANDU fueling machine tests required the using of an efficient data acquisition and processing Computer Control System. The challenging goal was to build a computer system (hardware and software) designed and engineered to control the test and calibration process of these fuel handling machines. The design takes care both of the functionality required to correctly control the CANDU fueling machine and of the additional functionality required to assist the testing process. Both the fueling machine testing rig and staff had successfully assessed by the AECL representatives during two missions. At same the time, at the Institute for Nuclear Research Pitesti was/is developed a numerical simulator for the CANDU fueling machine operators training. The paper presents the numerical simulator - a special PC program (software) which simulates the graphics and the functions and the operations at the main desk of the computer control system. The simulator permits 'to drive' a CANDU fueling machine in two manners: manual or automatic. The numerical simulator is dedicated to the training of operators who operate the CANDU fueling machine in a nuclear power plant with CANDU reactor. (author)

CANDU fuel behaviour under LOCA conditions

International Nuclear Information System (INIS)

Kohn, E.

1989-07-01

This report summarizes the current understanding of CANDU fuel-element behaviour under loss-of-coolant (LOCA) accidents. It focuses on a key in-reactor verification experiment conducted at Idaho National Engineering Laboratory (INEL) and on three Canadian in-reactor tests. The in-reactor data, and the considerable body of supporting information developed from out-reactor tests, support the general conclusion that CANDU fuel behaviour during LOCA transients is well understood. Four elements of 37-element CANDU fuel-bundle design were tested under conditions typical of a large-break LOCA blowdown in a CANDU reactor. The purpose of the test was to confirm our current understanding of fuel behaviour under loss-of-coolant accident blowdown conditions. The test also provided data for comparison with predictions made with the steady-state and transient fuel-element performance codes ELESIM and ELOCA. Key components of typical LOCA transients were incorporated in the test: namely, a rapid depressurization rate of the hot coolant, a simultaneous power increase before decreasing to decay values (a power pulse), and prototype fuel element under pre-transient power and burnup conditions. The test was successfully completed in the Power Burst Facility (PBF) reactor at INEL under contract to Ontario Hydro and AECL. The three CANDU Owners Group LOCA tests performed at Chalk River Nuclear Laboratories measured both the thermal-mechanical response and fission-gas release resulting from exposure to a LOCA transient. Results from these three tests provided further confirmation that the behaviour of the fuel under LOCA conditions is understood

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)

CANDU fuel performance

International Nuclear Information System (INIS)

Ivanoff, N.V.; Bazeley, E.G.; Hastings, I.J.

1982-01-01

CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

CANDU channel flow verification

International Nuclear Information System (INIS)

Mazalu, N.; Negut, Gh.

1997-01-01

The purpose of this evaluation was to obtain accurate information on each channel flow that enables us to assess precisely the level of reactor thermal power and, for reasons of safety, to establish which channel is boiling. In order to assess the channel flow parameters, computer simulations were done with the NUCIRC code and the results were checked by measurements. The complete channel flow measurements were made in the zero power cold condition. In hot conditions there were made flow measurements using the Shut Down System 1 (SDS 1) flow devices from 0.1 % F.P. up to 100 % F.P. The NUCIRC prediction for CANDU channel flows and the measurements by Ultrasonic Flow Meter at zero power cold conditions and SDS 1 flow channel measurements at different reactor power levels showed an acceptable agreement. The 100 % F.P. average errors for channel flow of R, shows that suitable NUCIRC flow assessment can be made. So, it can be done a fair prediction of the reactor power distribution. NUCIRC can predict accurately the onset of boiling and helps to warn at the possible power instabilities at high powers or it can detect the flow blockages. The thermal hydraulic analyst has in NUCIRC a suitable tool to do accurate predictions for the thermal hydraulic parameters for different steady state power levels which subsequently leads to an optimal CANDU reactor operation. (authors)

Proceedings of the international conference on CANDU fuel

International Nuclear Information System (INIS)

Hastings, I.J.

1986-01-01

These proceedings contain full texts of all paper presented at the first International Conference on CANDU Fuel. The Conference was organized and hosted by the Chalk River Branch of the Canadian Nuclear Society and utilized Atomic Energy of Canada Limited's facilities at Chalk River Nuclear Laboratories. Previously, informal Fuel Information Meetings were used in Canada to allow the exchange of information and technology associated with CANDU. The Chalk River conference was the first open international forum devoted solely to CANDU and included representatives of overseas countries with current or potential CANDU programs, as well as Canadian participants. The keynote presentation was given by Dr. J.B. Slater, who noted the correlation between past successes in CANDU fuel cycle technology and the co-operation between researchers, fabricators and reactor owner/operators in all phases of the fuel cycle, and outlined the challenges facing the industry today. In the banquet address, Dr. R.E. Green described the newly restructured AECL Research Company and its mission which blends traditional R and D with commercial initiatives. Since this forum for fuel technology has proven to be valuable, a second International CANDU Fuel Conference is planned for the fall of 1989, again sponsored by the Canadian Nuclear Society

Feasibility study of CANDU-9 fuel handling system

Energy Technology Data Exchange (ETDEWEB)

Hwang, Jeong Ki; Jo, C. H.; Kim, H. M.

1996-12-01

CANDU`s combination of natural uranium, heavy water and on-power refuelling is unique in its design and remarkable for reliable power production. In order to offer more output, better site utilization, shorter construction time, improved station layout, safety enhancements and better control panel layout, CANDU-9 is now under development with design improvement added to all proven CANDU advantages or applicable technologies. One of its major improvement has been applied to fuel handling system. This system is very similar to that of CANDU-3, and some parts of the system are applied to those of the existing CANDU-6 or CANDU-9. Design concepts and design requirements of fuel handling system for CANDU-9 have been identified to compare with those of the existing CANDU and the design feasibilities have been evaluated. (author). 4 tabs., 13 figs., 9 refs.

Verification tests for CANDU advanced fuel

International Nuclear Information System (INIS)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

1997-07-01

For the development of a CANDU advanced fuel, the CANFLEX-NU fuel bundles were tested under reactor operating conditions at the CANDU-Hot test loop. This report describes test results and test methods in the performance verification tests for the CANFLEX-NU bundle design. The main items described in the report are as follows. - Fuel bundle cross-flow test - Endurance fretting/vibration test - Freon CHF test - Production of technical document. (author). 25 refs., 45 tabs., 46 figs

Study on advanced nuclear fuel cycle of PWR/CANDU synergism

International Nuclear Information System (INIS)

Xie Zhongsheng; Huo Xiaodong

2002-01-01

According to the concrete condition that China has both PWR and CANDU reactors, one of the advanced nuclear fuel cycle strategy of PWR/CANDU synergism ws proposed, i.e. the reprocessed uranium of spent PWR fuel was used in CANDU reactor, which will save the uranium resource, increase the energy output, decrease the quantity of spent fuels to be disposed and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, the transition from the natural uranium to the recycled uranium (RU) can be completed without any changes of the structure of reactor core and operation mode. Furthermore, because of the low radiation level of RU, which is acceptable for CANDU reactor fuel fabrication, the present product line of fuel elements of CANDU reactor only need to be shielded slightly, also the conditions of transportation, operation and fuel management need not to be changed. Thus this strategy has significant practical and economical benefit

CANDU fuel : design/manufacturing interaction

International Nuclear Information System (INIS)

Graham, N.A.

1999-01-01

The design of CANDU fuel has been the product of intense cooperation among fuel designers and fuel manufacturers. The developments of some of the novel processes in fuel manufacture are outlined. These include the brazed-split-spacer design, the resistance welded endcap and CANLUB coatings. (author)

Comparative evaluation of fuel temperature coefficient of standard and CANFLEX fuels in CANDU 6

International Nuclear Information System (INIS)

Kim, Woosong; Hartant, Donny; Kim, Yonghee

2012-01-01

The fuel temperature reactivity coefficient (FTC) of CANDU 6 has become a concerning issue. The FTC was found to be slightly positive for the operating condition of CANDU 6. Since CANDU 6 has unique fuel arrangement and very soft neutron spectrum, its Doppler reactivity feedback of U 238 is rather weak. The upscattering by oxygen in fuel and Pu 239 buildup with fuel depletion are responsible for the positive FTC value at high temperature. In this study, FTC of both standard CANDU and CANFLEX fuel lattice are re evaluated. A Monte Carlo code Serpent2 was chosen as the analysis tool because of its high calculational speed and it can account for the thermal motion of heavy nuclides in fuel by using the Doppler Broadening Rejection Correction (DBRC) method. It was reported that the fuel Doppler effect is noticeably enhanced by accounting the target thermal motion. Recently, it was found that the FTC of the CANDU 6 standard fuel is noticeably enhanced by the DBRC

Potential of axial fuel management strategies in thorium-fuelled CANDU's

International Nuclear Information System (INIS)

Milgram, M.S.

1978-06-01

Three axial fuel management strategies are compared for use in a CANDU-PHW reactor operating on a self-sufficient, equilibrium thorium cycle. Two of these strategies are familiar ones for uranium reactors, and the third seeks to take advantage of the nuclear characteristics of the Th 232 → U 233 transmutation chain to improve the economics of the fuel cycle by periodically removing the fuel from the reactor. This results in an approximately 50% increase in burnup and an approximately 15% decrease in heavy element fuel inventory at a channel power of 6 MW, relative to the other strategies. (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

Development of fabrication technology for CANDU advanced fuel -Development of the advanced CANDU technology-

Energy Technology Data Exchange (ETDEWEB)

Choi, Chang Beom; Kim, Hyeong Soo; Kim, Sang Won; Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Jang, Ho Il; Kim, Sang Sik; Choi, Il Kwon; Cho, Dae Sik; Sheo, Seung Won; Lee, Soo Cheol; Kim, Yoon Hoi; Park, Choon Ho; Jeong, Seong Hoon; Kang, Myeong Soo; Park, Kwang Seok; Oh, Hee Kwan; Jang, Hong Seop; Kim, Yang Kon; Shin, Won Cheol; Lee, Do Yeon; Beon, Yeong Cheol; Lee, Sang Uh; Sho, Dal Yeong; Han, Eun Deok; Kim, Bong Soon; Park, Cheol Joo; Lee, Kyu Am; Yeon, Jin Yeong; Choi, Seok Mo; Shon, Jae Moon [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

1994-07-01

The present study is to develop the advanced CANDU fuel fabrication technologies by means of applying the R and D results and experiences gained from localization of mass production technologies of CANDU fuels. The annual portion of this year study includes following: 1. manufacturing of demo-fuel bundles for out-of-pile testing 2. development of technologies for the fabrication and inspection of advanced fuels 3. design and munufacturing of fuel fabrication facilities 4. performance of fundamental studies related to the development of advanced fuel fabrication technology.

The back end of the fuel cycle and CANDU

International Nuclear Information System (INIS)

Allan, C.J.; Dormuth, K.W.

2001-01-01

CANDU reactor operators have benefited from several advantages of the CANDU system and from AECL's experience, with regard to spent fuel handling, storage and disposal. AECL has over 20 years experience in development and application of medium-term storage and research and development on the disposal of used fuel. As a result of AECL's experience, short-term and medium-term storage and the associated handling of spent CANDU fuel are well proven and economic, with an extremely high degree of public and environmental protection. In fact, both short-term (water-pool) and medium-term (dry canister) storage of CANDU fuel are comparable or lower in cost per unit of energy than for PWRs. Both pool storage and dry spent fuel storage are fully proven, with many years of successful, safe operating experience. AECL's extensive R and D on the permanent disposal of spent-fuel has resulted in a defined concept for Canadian fuel disposal in crystalline rock. This concept was recently confirmed as ''technically acceptable'' by an independent environmental review panel. Thus, the Canadian program represents an international demonstration of the feasibility and safety of geological disposal of nuclear fuel waste. Much of the technology behind the Canadian concept can be adapted to permanent land-based disposal strategies chosen by other countries. In addition, the Canadian development has established a baseline for CANDU fuel permanent disposal costs. Canadian and international work has shown that the cost of permanent CANDU fuel disposal is similar to the cost of LWR fuel disposal per unit of electricity produced. (author)

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

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors. Refs, figs, tabs.

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Nichita, E., E-mail: Eleodor.Nichita@uoit.ca; Haroon, J., E-mail: Jawad.Haroon@uoit.ca

2016-10-15

Highlights: • A 37-element fuel bundle modified for {sup 99}Mo production in CANDU reactors is presented. • The modified bundle is neutronically and thermal-hydraulically equivalent to the standard bundle. • The modified bundle satisfies all safety criteria satisfied by the standard bundle. - Abstract: {sup 99m}Tc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of {sup 99}Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of {sup 99}Mo is investigated. A modified 37-element fuel bundle, suitable for the production of {sup 99}Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of {sup 99}Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium. The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of {sup 99}Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated {sup 99}Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of {sup 99}Mo.

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)

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

Natural uranium equivalent fuel an innovative design for proven CANDU technology

Energy Technology Data Exchange (ETDEWEB)

Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S., E-mail: fabricia.pineiro@candu.com [Candu Energy Inc., Mississauga, ON (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Company, Haiyan, Zhejiang (China)

2015-07-01

The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

CANDU reactor - supporting the nuclear renaissance

International Nuclear Information System (INIS)

Oberth, R.

2010-01-01

'Full text:' The CANDU reactor has proven to be a strong performer in both the Canada, with 22 units constructed in Ontario, New Brunswick and Quebec, as well as in Argentina, Korea, Romania and China where a further nine units are operating and two in the planning stage. The average lifetime capacity factor of the CANDU reactor fleet is 89%. The last seven CANDU projects in Korea, China, and Romania have been completed on budget and on schedule. CANDU reactors have the highest uranium utilization efficiency measures as electricity output per ton of uranium mined. The CANDU fuel channel design using on-power fuelling and a heavy water moderator enables flexible fueling options - from the current natural uranium option to burning uranium recovered from used LWR reactor fuel and even a thorium-based fuel. AECL and the CANDU reactor are poised to participate in the worldwide construction at least 250 new reactors over the next 20 years. (author)

Conference proceedings of the 4. international conference on CANDU fuel. V. 1,2

International Nuclear Information System (INIS)

1995-01-01

These proceedings contain the full texts of all 65 papers presented at the 4th International Conference on CANDU fuel. As such, they represent an update on the state-of-the-art in such important CANDU fuel topics as International Development Programs and Operating Experience with CANDU fuel, Performance Assessments and Fuel Behavior Modeling, Fuel Properties, Licensing and Accident Analyses for CANDU fuel, Design, Testing and Manufacturing, and Advanced Fuel Designs. The large number of papers required the use of parallel sessions for the first time at a CANDU Fuel Conference

Conference proceedings of the 4. international conference on CANDU fuel. V. 1,2

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-31

These proceedings contain the full texts of all 65 papers presented at the 4th International Conference on CANDU fuel. As such, they represent an update on the state-of-the-art in such important CANDU fuel topics as International Development Programs and Operating Experience with CANDU fuel, Performance Assessments and Fuel Behavior Modeling, Fuel Properties, Licensing and Accident Analyses for CANDU fuel, Design, Testing and Manufacturing, and Advanced Fuel Designs. The large number of papers required the use of parallel sessions for the first time at a CANDU Fuel Conference.

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

CANDU fuel cycles - present and future

International Nuclear Information System (INIS)

Mooradian, A.J.

1976-05-01

The present commercially proven Canadian nuclear power system is based on a once-through natural uranium fuel cycle characterized by high uranium utilization and a high conversion efficiency. The cycle closes with secure retrievable storage of spent fuel. This cycle is based on a CANDU reactor concept which is now well understood. Both active and passive fuel storage options have been investigated and will be described in this paper. Future development of the CANDU system is focussed on conservation of uranium by plutonium and thorium recycle. The full exploitation of these options requires continued emphasis on neutron conservation, efficiency of extraction and fuel refabrication processes. The results of recent studies are discussed in this paper. (author)

Experiments in ZED-2 to study the physics of low-void reactivity fuel in CANDU

International Nuclear Information System (INIS)

Zeller, M.B.; Celli, A.; McPhee, G.P.

1994-01-01

Prospective CANDU clients have indicated a desire for a zero or negative coolant void reactivity. In response to this market requirement AECL Research and AECL CANDU are jointly developing and testing a Low-Void Reactivity Fuel (LVRF) bundle, which will be retrofitable to the current generation of CANDU reactors. An important component of the LVRF program is the undertaking of reactor-physics experiments in the zero-energy ZED-2 lattice test facility at Chalk River Laboratories. Preliminary void-reactivity measurements have already been performed in ZED-2 using a limited amount of the prototype fuel. These experiments were to provide a proof-of-principle for the LVRF concept. A more comprehensive set of experiments are planned for later this year. Experiments to be performed include: measuring the critical buckling of CANDU-type lattices containing LVRF, with and without coolant in the channels; measuring the reactivity effect of heating the LVRF fuel and coolant in ZED-2 hot channels; and measuring detailed reaction rates and neutron density distributions across a LVRF bundle, in voided and D 2 O-cooled channels, by the foil activation method. This paper describes the experimental approach to be used for the study and presents calculations employing transport and diffusion theory to predict the results. The codes used for the simulations are the lattice code WIMS-AECL and the core code CONIFERS. Included in the paper are results from the preliminary measurement of void coefficient for LVRF in a ZED-2 lattice and a comparison of those results to predictions based on WIMS-AECL calculations. (author). 3 refs., 1 tab., 10 figs

Development of a new bundle welding technology for CANDU fuels

International Nuclear Information System (INIS)

Kim, Soo Sung; Lee, D. Y.; Goo, D. S.

2010-01-01

The new technology of welding process for fuel bundle of CANDU nuclear fuels is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of leakage of the fission products is mostly apt to occur at the weldments of fuel bundles, and it is connected directly with the safety and life prediction of the nuclear reactor in operation. The fuel bundles of CANDU nuclear fuels are welded by the electrical resistance method, connecting the endplates and endcaps with fuel rods. Therefore, the purpose of this study of the 2nd year is to select the proper welding parameters and to investigate the characteristics of the full-sized samples using the projection endplates and make some prototype samples for the endplate welding of CANDU nuclear fuels. This study will be also provide the fundamental data for the new design and fabrications of CANDU nuclear fuel bundles

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)

A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

Radial heat transfer from fuel to moderator during LOCAs for CANDU PHW reactors

International Nuclear Information System (INIS)

Hildebrandt, J.G.; So, C.B.; Gillespie, G.E.; MacLean, G.

1983-01-01

In a postulated CANDU-PHW loss-of-coolant accident (LOCA) with coincident impaired emergency cooling, the axial transport of heat from the fuel by convection is reduced. This reduction in heat removal causes the fuel to heat up and the radial heat transfer to the moderator to become significant. This paper deals with two codes that predict the thermal response of fuel channels under LOCA conditions. New channel thermal radiation models in both RAMA, a thermalhydraulic code, and CHAN II, a fuel channel thermo-chemical code, are presented and their predictions are compared with the experimental results of an electrically heated bundle of 37 fuel pins. A second experiment, involving a single heated pin in a channel with flowing steam, is presented. The predictions of RAMA and CHAN II are compared with this experiment to verify the codes' thermo-chemical models. There is good agreement between the predictions of both codes and the experimental results

Advancing CANDU Technology Through R and D

International Nuclear Information System (INIS)

Torgerson, David F.

1993-01-01

CANDU reactors are evolving to meet future requirements using incremental changes as opposed to revolutionary design changes. The main elements for advancing the technology reducing capital and operating, increasing capacity factors, increasing passive safety, and enhancing fuel/fuel cycle flexibility. These elements are being addressed by carrying out research and development in the areas of safety, plant systems and components, heavy water production, information technology, fuel channels, and fuel/fuel cycle technology. In safety, the focus is on using the inherent features of CANDU to enhance passive or natural safety concepts, such as the use of the moderator as an effective heat sink, and the development of advanced fuels to improve critical heat flux and to reduce source terms. Plant systems and components work includes improvements to plant systems such as steam generators, heat exchangers, pump seals, and advanced control room technology. Heavy water processes are being developed that can be used with existing hydrogen production plants, or that can be used in a stand-alone mode. Information technology is being developed to cover all aspects of CANDU design, construction, and operation. Fuel channel improvements include elucidation and application of basic materials science for life extension, and the development of advanced non-destructive examination methods. Fuel and fuel cycle work is focusing on LWR/CANDU synergy, such as the use of recovered uranium and the direct use of spent PWR fuel in CANDU reactor, advanced fuels to improve burnup and economics (e. g., the joint AECB/KAERI Conflux program), and low void reactivity fuel to enhance passive safety. This paper gives an overview of some of the R and D supporting these activities, with particular emphasis on Alice's vision for advancing CANDU technology over the next 10 years

CANDU fuel - fifteen years of power reactor experience

International Nuclear Information System (INIS)

Fanjoy, G.R.; Bain, A.S.

1977-01-01

CANDU (Canada Deuterium Uranium) fuel has operated in power reactors since 1962. Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) A thorough understanding of the fundamental behaviour of CANDU fuel. (b) Data showing that the predicted high utilization of uranium has been achieved. Actual fuelling costs in 1976 at the Pickering Generating Station are 1.2 m$/kWh (1976 Canadian dollars) with the simple oncethrough natural-UO 2 fuel cycle. (c) Criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to ''CANLUB'' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases. (d) Proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). Involvement by the utility in all stages of fuel development has resulted in efficient application of this fundamental knowledge to ensure proper fuel specifications, procurement, scheduling into the reactor and feedback to developers, designers and manufacturers. As of mid-1976 over 3 x 10 6 individual elements have been built in a well-estabilished commercially competitive fuel fabrication industry and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification. Development work on UO 2 and other fuel cycles (plutonium and thorium) is continuing, and, because CANDU reactors use on-power fuelling, bundles can be inserted into power reactors for testing. Thus new fuel designs can be quickly adopted to ensure that the CANDU system continues to provide low-cost energy with high reliability

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

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)

Hydrogen in CANDU fuel elements

International Nuclear Information System (INIS)

Sejnoha, R.; Manzer, A.M.; Surette, B.A.

1995-01-01

Unirradiated and irradiated CANDU fuel cladding was tested to compare the role of stress-corrosion cracking and of hydrogen in the development of fuel defects. The results of the tests are compared with information on fuel performance in-reactor. The role of hydriding (deuteriding) from the coolant and from the fuel element inside is discussed, and the control of 'hydrogen gas' content in the element is confirmed as essential for defect-free fuel performance. Finally, implications for fuel element design are discussed. (author)

Physics characteristics of CANDU cores with advanced fuel cycles

International Nuclear Information System (INIS)

Garvey, P.M.

1985-01-01

The current generation of CANDU reactors, of which some 20 GWE are either in operations or under construction worldwide, have been designed specifically for the natural uranium fuel cycle. The CANDU concept, due to its D 2 O coolant and moderator, on-power refuelling and low absorption structural materials, makes the most effective utilization of mined uranium of all currently commercialized reactors. An economic fuel cycle cost is also achieved through the use of natural uranium and a simple fuel bundle design. Total unit energy costs are achieved that allow this reactor concept to effectively compete with other reactor types and other forms of energy production. There are, however, other fuel cycles that could be introduced into this reactor type. These include the slightly enriched uranium fuel cycle, fuel cycles in which plutonium is recycled with uranium, and the thorium cycle in which U-233 is recycled. There is also a special range of fuel cycles that could utilize the spent fuel from LWR's. Two specific variants are a fuel cycle that only utilizes the spent uranium, and a fuel cycle in which both the uranium and plutonium are recycled into a CANDU. For the main part these fuel cycles are characterized by a higher initial enrichment, and hence discharge burnup, than the natural uranium cycle. For these fuel cycles the main design features of both the reactor and fuel bundle would be retained. Recently a detailed study of the use in a CANDU of mixed plutonium and uranium oxide fuel from an LWR has been undertaken by AECL. This study illustrates many of the generic technical issues associated with the use of Advanced Fuel Cycles. This paper will report the main findings of this evaluation, including the power distribution in the reactor and fuel bundle, the choice of fuel management scheme, and the impact on the control and safety characteristics of the reactor. These studies have not identified any aspects that significantly impact upon the introduction of

Natural uranium equivalent fuel. An innovative design for proven CANDU technology

Energy Technology Data Exchange (ETDEWEB)

Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S. [Candu Energy Inc., Mississauga, Ontario (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Co., Haiyan, Zhejiang (China)

2015-09-15

The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU® reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

Assessment of DUPIC fuel compatibility with CANDU-6

Energy Technology Data Exchange (ETDEWEB)

Choi, H B; Roh, G H; Jeong, C J; Rhee, B W; Choi, J W [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-07-01

The compatibility of DUPIC fuel with the existing CANDU reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will require a quantitative analysis later. (author).

Coolant Void Reactivity Analysis of CANDU Lattice

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [UNIST, Ulsan (Korea, Republic of)

2016-05-15

Models of CANDU-6 and ACR-700 fuel lattices were constructed for a single bundle and 2 by 2 checkerboard to understand the physics related to CVR. Also, a familiar four factor formula was used to predict the specific contributions to reactivity change in order to achieve an understanding of the physics issues related to the CVR. At the same time, because the situation of coolant voiding should bring about a change of neutron behavior, the spectral changes and neutron current were also analyzed. The models of the CANDU- 6 and ACR-700 fuel lattices were constructed using the Monte Carlo code MCNP6 using the ENDF/B-VII.0 continuous energy cross section library based on the specification from AECL. The CANDU fuel lattice was searched through sensitivity studies of each design parameter such as fuel enrichment, fuel pitch, and types of burnable absorber for obtaining better behavior in terms of CVR. Unlike the single channel coolant voiding, the ACR-700 bundle has a positive reactivity change upon 2x2 checkerboard coolant voiding. Because of the new path for neutron moderation, the neutrons from the voided channel move to the no-void channel where they lose energy and come back to the voided channel as thermal neutrons. This phenomenon causes the positive CVR when checkerboard voiding occurs. The sensitivity study revealed the effects of the moderator to fuel volume ratio, fuel enrichment, and burnable absorber on the CVR. A fuel bundle with low moderator to fuel volume ratio and high fuel enrichment can help achieve negative CVR.

CANDU fuel quality and how it is achieved

International Nuclear Information System (INIS)

Gacesa, M.; Quarrington, G.R.; Tarasuk, W.R.; Carrick, I.R.; Pawliw, J.; McGregor, G.; Debnam, H.R.; Proos, L.

1980-07-01

In this three part presentation CANDU fuel quality is reviewed from the point of view of a designer/operator and a fabricator. In Part 'A' fuel performance and quality considerations are discussed from the point of view of a designer-operator. In Parts 'B' and 'C' fuel quality is reviewed from the point of view of a fabricator. The presentation was divided in this way to convey the 'team effort' attitude which exists in the Canadian program; the team effort which is an essential part of the CANDU story. (auth)

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)

The relationship between natural uranium and advanced fuel cycles in CANDU reactors

International Nuclear Information System (INIS)

Lane, A.D.; McDonnell, F.N.; Griffiths, J.

1988-11-01

CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors. 22 refs

Characteristics of used CANDU fuel relevant to the Canadian nuclear fuel waste management program

Energy Technology Data Exchange (ETDEWEB)

Wasywich, K M

1993-05-01

Literature data on the characteristics of used CANDU power reactor fuel that are relevant to its performance as a waste form have been compiled in a convenient handbook. Information about the quantities of used fuel generated, burnup, radionuclide inventories, fission gas release, void volume and surface area, fuel microstructure, fuel cladding properties, changes in fuel bundle properties due to immobilization processes, radiation fields, decay heat and future trends is presented for various CANDU fuel designs. (author). 199 refs., 39 tabs., 100 figs.

Characteristics of used CANDU fuel relevant to the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Wasywich, K.M.

1993-05-01

Literature data on the characteristics of used CANDU power reactor fuel that are relevant to its performance as a waste form have been compiled in a convenient handbook. Information about the quantities of used fuel generated, burnup, radionuclide inventories, fission gas release, void volume and surface area, fuel microstructure, fuel cladding properties, changes in fuel bundle properties due to immobilization processes, radiation fields, decay heat and future trends is presented for various CANDU fuel designs. (author). 199 refs., 39 tabs., 100 figs

Advancing CANDU technology AECL's Development program

International Nuclear Information System (INIS)

Torgerson, D.F.

1997-01-01

AECL has a comprehensive product development program that is advancing all aspects of CANDU technology including fuel and fuel cycles, fuel channels, heavy water and tritium technology, safety technology, components and systems, constructability, health and environment, and control and instrumentation. The technology arising from these programs is being incorporated into the CANDU design through an evolutionary process. This evolutionary process is focused on improving economics, enhancing safety and ensuring fuel cycle flexibility to secure fuel supply for the foreseeable future. This strategic thrusts are being used by CANDU designers and researchers to set priorities and goals for AECL's development activities. The goals are part of a 25-year development program that culminates in the 'CANDU X'. The 'CANDU X' is not a specific design - it is a concept that articulates our best extrapolation of what is achievable with the CANDU design over the next 25 years, and includes the advanced features arising from the R and D and engineering to be done over that time. AECL's current product, the 700 MWe class CANDU 6 and the 900 MWe class CANDU 9, both incorporate output from the development programs as the technology become available. A brief description of each development areas is given below. The paper ends with the conclusion that AECL has a clear vision of how CANDU technology and products will evolve over the next several years, and has structured a comprehensive development program to take full advantage of the inherent characteristics of heavy water reactors. (author)

Ninth international conference on CANDU fuel, 'fuelling a clean future'

International Nuclear Information System (INIS)

2005-01-01

The Canadian Nuclear Society's 9th International Conference on CANDU fuel took place in Belleville, Ontario on September 18-21, 2005. The theme for this year's conference was 'Fuelling a Clean Future' bringing together over 80 delegates ranging from: designers, engineers, manufacturers, researchers, modellers, safety specialists and managers to share the wealth of their knowledge and experience. This international event took place at an important turning point of the CANDU technology when new fuel design is being developed for commercial application, the Advanced CANDU Reactor is being considered for projects and nuclear power is enjoying a renaissance as the source energy for our future. Most of the conference was devoted to the presentation of technical papers in four parallel sessions. The topics of these sessions were: Design and Development; Fuel Safety; Fuel Modelling; Fuel Performance; Fuel Manufacturing; Fuel Management; Thermalhydraulics; and, Spent Fuel Management and Criticalty

Development of the CANDU high-burnup fuel design/analysis technology

Energy Technology Data Exchange (ETDEWEB)

Suk, Ho Chun; Sim, K. S.; Oh, D. J.; Park, J. H.; Jun, J. S.; Yoo, K. J.

1997-08-01

This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs.

Development of the CANDU high-burnup fuel design/analysis technology

International Nuclear Information System (INIS)

Suk, Ho Chun; Sim, K. S.; Oh, D. J.; Park, J. H.; Jun, J. S.; Yoo, K. J.

1997-08-01

This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs

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)

Design of a transportation cask for irradiated CANDU fuel

International Nuclear Information System (INIS)

Nash, K.E.; Gavin, M.E.

1983-01-01

A major step in the development of a large-scale transportation system for irradiated CANDU fuel is being made by Ontario Hydro in the design and construction of a demonstration cask by 1988/89. The system being designed is based on dry transportation with the eventual fully developed system providing for dry fuel loading and unloading. Research carried out to date has demonstrated that it is possible to transport irradiated CANDU fuel in a operationally efficient and simple manner without any damage which would prejudice subsequent automated fuel handling

Study of candu fuel elements irradiated in a nuclear power plant

International Nuclear Information System (INIS)

Ionescu, S.; Uta, O.; Mincu, M.; Anghel, D.; Prisecaru, I.

2015-01-01

The object of this work is the behaviour of CANDU fuel elements after service in nuclear power plant. The results are analysed and compared with previous result obtained on unirradiated samples and with the results obtained on samples irradiated in the TRIGA reactor of INR Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor, the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) of INR Pitesti on samples from fuel elements after they were removed out of the nuclear power plant: - dimensional and macrostructural characterization; - microstructural characterization by metallographic analyses; - determination of mechanical properties; - fracture surface analysis by scanning electron microscopy (SEM). A full set of non-destructive and destructive examinations concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the cladding was performed. The obtained results are typical for CANDU 6-type fuel. The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for the improvement of the CANDU fuel. (authors)

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)

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

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)

Nuclear power - replacement of pressure tubes in CANDU reactors

International Nuclear Information System (INIS)

Anon.

1992-01-01

The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

Fuel channel closure and adapter

International Nuclear Information System (INIS)

Cashen, W.S.

1985-01-01

This invention provides a mechanical closure/actuating ram combination particularly suited for use in sealing the ends of the pressure tubes when a CANDU-type reactor is refueled. It provides a cluster that may be inserted into a fuel channel end fitting to provide at least partial closing off of a pressure tube while permitting the disengagement of the fueling machine and its withdrawal from the closure for other purposes. The invention also provides a ram/closure combination wherein the application of loading force to a deformable sealing disk is regulated by a massive load bar component forming part of the fueling machine and being therefore accessible for maintenance or replacement

Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

International Nuclear Information System (INIS)

Clegg, L.J.; Coady, J.R.

1977-01-01

The two books of Volume 1 comprise the first in a three-volume series of compilations on the radioactive decay propertis of CANDU fuel and deal with the natural uranium fuel cycle. Succeeding volumes will deal with fuel cycles based on plutonium recycle and thorium. In Volume 1 which is divided into three parts, the computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 3 contains the data relating to the plutonium product and the high level wastes produced during fuel reprocessing. (author)

Load-following performance and assessment of CANDU fuel

Energy Technology Data Exchange (ETDEWEB)

Tayal, M.; Floyd, M.; Rattan, D.; Xu, Z.; Manzer, A.; Lau, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Power Generation, Fuel and Fuel Channel Analysis Dept., Toronto, Ontario (Canada)

1999-09-01

Load following of nuclear reactors is now becoming an economic necessity in some countries. When nuclear power stations are operated in a load-following mode, the reactor and the fuel may be subjected to step changes in power on a weekly, daily, or even hourly basis, depending on the grid's needs. This paper updates the previous surveys of load-following capability of CANDU fuel, focusing mainly on the successful experience at the Bruce B station. As well, initial analytical assessments are provided that illustrate the capability of CANDU fuel to survive conditions other than those for which direct in-reactor evidence is available. (author)

Proceedings of the fifth international conference on CANDU fuel. V.1,2

International Nuclear Information System (INIS)

Lau, J.H.

1997-01-01

The First International Conference on CANDU Fuel was held in Chalk River in 1986. The CANDU Fuel community has gathered every three years since. The papers presented include topics on international experience, CANFLEX fuel bundles, Fuel design, Fuel modelling, Manufacturing and Quality assurance, Fuel performance and Safety, Fuel cycles and Spent Fuel management. Volume One was published in advance of the conference and Volume Two was printed after the conference

Proceedings of the fifth international conference on CANDU fuel. V.1,2

Energy Technology Data Exchange (ETDEWEB)

Lau, J H [ed.

1997-07-01

The First International Conference on CANDU Fuel was held in Chalk River in 1986. The CANDU Fuel community has gathered every three years since. The papers presented include topics on international experience, CANFLEX fuel bundles, Fuel design, Fuel modelling, Manufacturing and Quality assurance, Fuel performance and Safety, Fuel cycles and Spent Fuel management. Volume One was published in advance of the conference and Volume Two was printed after the conference.

CANDU type fuel behavior evaluation - a probabilistic approach

International Nuclear Information System (INIS)

Moscalu, D.R.; Horhoianu, G.; Popescu, I.A.; Olteanu, G.

1995-01-01

In order to realistically assess the behavior of the fuel elements during in-reactor operation, probabilistic methods have recently been introduced in the analysis of fuel performance. The present paper summarizes the achievements in this field at the Institute for Nuclear Research (INR), pointing out some advantages of the utilized method in the evaluation of CANDU type fuel behavior in steady state conditions. The Response Surface Method (RSM) has been selected for the investigation of the effects of the variability in fuel element computer code inputs on the code outputs (fuel element performance parameters). A new developed version of the probabilistic code APMESRA based on RSM is briefly presented. The examples of application include the analysis of the results of an in-reactor fuel element experiment and the investigation of the calculated performance parameter distribution for a new CANDU type extended burnup fuel element design. (author)

Study on the use of slightly enriched uranium fuel cycle in an existing CANDU 6 reactor

Energy Technology Data Exchange (ETDEWEB)

Yeom, Choong Sub; Kim, Hyun Dae [Institute for Advanced Engineering, Seoul (Korea, Republic of)

1998-12-31

To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers. 8 refs., 2 figs., 1 tab. (Author)

Study on the use of slightly enriched uranium fuel cycle in an existing CANDU 6 reactor

Energy Technology Data Exchange (ETDEWEB)

Yeom, Choong Sub; Kim, Hyun Dae [Institute for Advanced Engineering, Seoul (Korea, Republic of)

1997-12-31

To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers. 8 refs., 2 figs., 1 tab. (Author)

Extension of the time-average model to Candu refueling schemes involving reshuffling

International Nuclear Information System (INIS)

Rouben, Benjamin; Nichita, Eleodor

2008-01-01

Candu reactors consist of a horizontal non-pressurized heavy-water-filled vessel penetrated axially by fuel channels, each containing twelve 50-cm-long fuel bundles cooled by pressurized heavy water. Candu reactors are refueled on-line and, as a consequence, the core flux and power distributions change continuously. For design purposes, a 'time-average' model was developed in the 1970's to calculate the average over time of the flux and power distribution and to study the effects of different refueling schemes. The original time-average model only allows treatment of simple push-through refueling schemes whereby fresh fuel is inserted at one end of the channel and irradiated fuel is removed from the other end. With the advent of advanced fuel cycles and new Candu designs, novel refueling schemes may be considered, such as reshuffling discharged fuel from some channels into other channels, to achieve better overall discharge burnup. Such reshuffling schemes cannot be handled by the original time-average model. This paper presents an extension of the time-average model to allow for the treatment of refueling schemes with reshuffling. Equations for the extended model are presented, together with sample results for a simple demonstration case. (authors)

The integrity of CANDU fuel during load following

International Nuclear Information System (INIS)

Tayal, M.; Manzer, A.M.; Sejnoha, R.; Hains, A.J.

1989-08-01

This paper summarizes data and analyses of integrity and of physics of CANDU fuel during load following. Measurements of irradiated fuel show that power cycles do not enhance release of fission gas. Data from research reactors show that the power cycles cause cyclic strains in the sheath. Finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath. The stresses and the strains are well into the plastic range. The cyclic loads 'use up' some fraction of the sheath's resistance to environmentally-assisted cracking (EAC), depending on the details of the fuel design and of then power cycles. The balance of the sheath's resistance to EAC continues to be available to counteract static loads. Thousands of fuel bundles have experienced many power cycles in research and in commercial reactors. Overall integrity of fuel bundles is well over 99%. Thus, CANDU fuel continues to show good performance in both base-load and load-following reactors

Preliminary assessment on compatibility of DUPIC fuel with CANDU-6

International Nuclear Information System (INIS)

Choi, Hang-Bok; Roh, G.H.; Jeong, C.J.; Rhee, B.W.; Choi, J.W.; Boss, C.R.

1997-01-01

The compatibility of DUPIC fuel with the existing CANDU-6 reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will be qualitatively analyzed later. (author)

Assessment of neutron transport codes for application to CANDU fuel lattices analysis

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

1999-08-01

In order to assess the applicability of WIMS-AECL and HELIOS code to the CANDU fuel lattice analysis, the physics calculations has been carried out for the standard CANDU fuel and DUPIC fuel lattices, and the results were compared with those of Monte Carlo code MCNP-4B. In this study, in order to consider the full isotopic composition and the temperature effect, new MCNP libraries have been generated from ENDF/B-VI release 3 and validated for typical benchmark problems. The TRX-1,2,BAPL-1,2,3 pin -cell lattices and KENO criticality safety benchmark calculations have been performed for the new MCNP libraries, and the results have shown that the new MCNP library has sufficient accuracy to be used for physics calculation. Then, the lattice codes have been benchmarked by the MCNP code for the major physics parameters such as the burnup reactivity, void reactivity, relative pin power and Doppler coefficient, etc. for the standard CANDU fuel and DUPIC fuel lattices. For the standard CANDU fuel lattice, it was found that the results of WIMS-AECL calculations are consistent with those of MCNP. For the DUPIC fuel lattice, however, the results of WIMS-AECL calculations with ENDF/B-V library have shown that the discrepancy from the results of MCNP calculations increases when the fuel burnup is relatively high. The burnup reactivities of WIMS-ACEL calculations with ENDF/B-VI library have shown excellent agreements with those of MCNP calculation for both the standard CANDU and DUPIC fuel lattices. However, the Doppler coefficient have relatively large discrepancies compared with MCNP calculations, and the difference increases as the fuel burns. On the other hand, the results of HELIOS calculation are consistent with those of MCNP even though the discrepancy is slightly larger compared with the case of the standard CANDU fuel lattice. this study has shown that the WIMS-AECL products reliable results for the natural uranium fuel. However, it is recommended that the WIMS

Research on nondestructive examination methods for CANDU fuel channel inspection

International Nuclear Information System (INIS)

Soare, M.; Petriu, F.; Toma, V.; Revenco, V.; Calinescu, A.; Ciocan, R.; Iordache, C.; Popescu, L.; Mihalache, M.; Murgescu, C.

1995-01-01

The requirements of the 1994 edition of CAN/CSA-N285.4 Periodic Inspection Standard, which address all known and postulated degradation mechanisms and introduce material surveillance demands, involve a growing need for improved nondestructive examination (NDE) methods and technologies. In order to have a proper technical support in its decisions concerning fuel channel inspections at Cernavoda NPP, the Romanian Power Authority (RENEL) initiated a Research Program regarding the nondestructive characterization of the fuel channels structural integrity. The paper presents the most significant results obtained on this Research Program: the ENDUS experimental system for Laboratory simulation of the fuel channel inspection, ultrasonic Rayleigh-Lamb waves technique for pressure tubes examination, phase analysis technique for near-surface flaws, influence of the metallurgical state of the pressure tube material on the eddy current defectoscopic signals, characterization of plastic deformation and fracture of zirconium alloys by acoustic emission. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Suk, Hoh Chun; Shim, Kee Sub; Byun, Taek Sang; Park, Kwang Suk; Kang, Heui Yung; Kim, Bong Kee; Jung, Chang Joon; Lee, Yung Wook; Bae, Chang Joon; Kwon, Oh Sun; Oh, Duk Joo; Im, Hong Sik; Ohn, Myung Ryong; Lee, Kang Moon; Park, Joo Hwan; Lee, Eui Joon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

This is the `94 annual report of the CANDU advanced fuel design and analysis project, and describes CANFLEX fuel design and mechanical integrity analysis, reactor physics analysis and safety analysis of the CANDU-6 with the CANFLEX-NU. The following is the R and D scope of this fiscal year : (1) Detail design of CANFLEX-NU and detail analysis on the fuel integrity, reactor physics and safety. (a) Detail design and mechanical integrity analysis of the bundle (b) CANDU-6 refueling simulation, and analysis on the Xe transients and adjuster system capability (c) Licensing strategy establishment and safety analysis for the CANFLEX-NU demonstration demonstration irradiation in a commercial CANDU-6. (2) Production and revision of CANFLEX-NU fuel design documents (a) Production and approval of CANFLEX-NU reference drawing, and revisions of fuel design manual and technical specifications (b) Production of draft physics design manual. (3) Basic research on CANFLEX-SEU fuel. 55 figs, 21 tabs, 45 refs. (Author).

Progress of the DUPIC fuel compatibility analysis (II) - thermal-hydraulics

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Choi, Hang Bok

2005-03-01

Thermal-hydraulic compatibility of the DUPIC fuel bundle with a 713 MWe Canada deuterium uranium (CANDU-6) reactor was studied by using both the single channel and sub-channel analysis methods. The single channel analysis provides the fuel channel flow rate, pressure drop, critical channel power, and the channel exit quality, which are assessed against the thermal-hydraulic design requirements of the CANDU-6 reactor. The single channel analysis by the NUCIRC code showed that the thermal-hydraulic performance of the DUPIC fuel is not different from that of the standard CANDU fuel. Regarding the local flow characteristics, the sub-channel analysis also showed that the uncertainty of the critical channel power calculation for the DUPIC fuel channel is very small. As a result, both the single and sub-channel analyses showed that the key thermal-hydraulic parameters of the DUPIC fuel channel do not deteriorate compared to the standard CANDU fuel channel.

R and D activities on CANDU-type fuel in Indonesia

International Nuclear Information System (INIS)

Suripto, A.; Badruzzaman, M.; Latief, A.

1997-01-01

The status of R and D activities in Indonesia with respect of CANDU-type fuel development is presented. The activities have been started since the first feasibility study to introduce nuclear power plants was carried out in 1970s. The early research comprised the in-situ pilot production of yellow-cake in Kalimantan (Borneo) experimental mining site, uranium purification and pellet preparation. This program continued to gain a full support from the Government which culminated in the realisation of the construction by BATAN of a large fuel development laboratory in Serpong, starting from 1984 in co-operation with NIRA Ansaldo of Italy. The laboratory, which is called the Power Reactor Experimental Fuel Element Installation (EFEI) was originally designed as an experimental facility to integrate the acquired domestic R and D results gained so far on the CANDU-type fuel technology and the additional know-how received from NIRA Ansaldo which at that time was engaged, in developing a CANDU-type fuel, called the CIRENE fuel design. In the present days the facility houses the power reactor fuel development activities carried out to build up the national capability on power reactor fuel fabrication technology in anticipation to embark upon the nuclear energy era in the near future. (author)

The burnable poisons utilization for fissile enriched CANDU fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Serghiuta, D; Nainer, O [Team 3 Solutions, Don Mills, ON (Canada)

1996-12-31

Utilization of burnable poison for the fissile enriched fueled CANDU 6 Mk1 core is investigated. The main incentives for this analysis are the reduction of void reactivity effects, the maximization of the fissile content of fresh fuel bundles, and the achievement of better power shape control, in order to preserve the power envelope of the standard 37 rod fuel bundle. The latter allows also the preservation of construction parameters of the standard core (for example: number and location of reactivity devices). It also permits the use of regular shift fueling schemes. The paper makes analyses of MOX weapons-grade plutonium and 1.2% SEU fueled CANDU 6 Mk 1 cores. (author). 6 refs., 4 tabs., 10 figs.

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

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

AECL has developed a suite of technologies for Candu R reactors that enable the next step in the evolution of the Candu family of heavy-water-moderated fuel-channel reactors. These technologies have been combined in the design for the Advanced Candu Reactor TM1 (ACRTM), AECL's next generation Candu power plant. The ACR design builds extensively on the existing Candu experience base, but includes innovations, in design and in delivery technology, that provide very substantial reductions in capital cost and in project schedules. In this paper, main features of next generation design and delivery are summarized, to provide the background basis for the cost and schedule reductions that have been achieved. In particular the paper outlines the impact of the innovative design steps for ACR: - Selection of slightly enriched fuel bundle design; - Use of light water coolant in place of traditional Candu heavy water coolant; - Compact core design with unique reactor physics benefits; - Optimized coolant and turbine system conditions. In addition to the direct cost benefits arising from efficiency improvement, and from the reduction in heavy water, the next generation Candu configuration results in numerous additional indirect cost benefits, including: - Reduction in number and complexity of reactivity mechanisms; - Reduction in number of heavy water auxiliary systems; - Simplification in heat transport and its support systems; - Simplified human-machine interface. The paper also describes the ACR approach to design for constructability. The application of module assembly and open-top construction techniques, based on Candu and other worldwide experience, has been proven to generate savings in both schedule durations and overall project cost, by reducing premium on-site activities, and by improving efficiency of system and subsystem assembly. AECL's up-to-date experience in the use of 3-D CADDS and related engineering tools has also been proven to reduce both engineering and

Safety assessment to support NUE fuel full core implementation in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Fan, H.Z.; Laurie, T.; Siddiqi, A.; Li, Z.P.; Rouben, D.; Zhu, W.; Lau, V.; Cottrell, C.M. [CANDU Energy Inc., Mississauga, Ontario (Canada)

2013-07-01

The Natural Uranium Equivalent (NUE) fuel contains a combination of recycled uranium and depleted uranium, in such a manner that the resulting mixture is similar to the natural uranium currently used in CANDU® reactors. Based on successful preliminary results of 24 bundles of NUE fuel demonstration irradiation in Qinshan CANDU 6 Unit 1, the NUE full core implementation program has been developed in cooperation with the Third Qinshan Nuclear Power Company and Candu Energy Inc, which has recently received Chinese government policy and funding support from their National-Level Energy Innovation program. This paper presents the safety assessment results to technically support NUE fuel full core implementation in CANDU reactors. (author)

Assessment of core characteristics during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

International Nuclear Information System (INIS)

Jeong, Chang Joon; Suk, Ho Chun

2002-01-01

A transition from 37-element natural uranium fuel to CANFLEX-NU fuel has been modeled in a 1200-day time-dependent fuel management simulation for a CANDU 6 reactor. The simulation was divided into three parts. The pre-transition period extended from 0 to 300 FPD, in which the reactor was fuelled only with standard 37-element fuel bundles. In the transition period, refueling took place only with the CANFLEX-NU fuel bundle. The transition stage lasted from 300 to 920 FPD, at which point all of the 37-element fuel in the core had been replaced by CANFLEX-NU fuel bundle. In the post-transition phase, refueling continued with CANFLEX-NU fuel until 1200 FPD, to arrive at estimate of the equilibrium core characteristics with CANFLEX-NU fuel. Simulation results show that the CANFLEX-NU fuel bundle has a operational compatibility with the CANDU 6 reactor during the transition core, and also show that the transition core from 37-element natural uranium fuel to CANFLEX-NU can be operated without violating any license limit of the CANDU 6 reactor

Alternative Concept to Enhance the Disposal Efficiency for CANDU Spent Fuel Disposal System

International Nuclear Information System (INIS)

Lee, Jong Youl; Cho, Dong Geun; Kook, Dong Hak; Lee, Min Soo; Choi, Heui Joo

2011-01-01

There are two types of nuclear reactors in Korea and they are PWR type and CANDU type. The safe management of the spent fuels from these reactors is very important factor to maintain the sustainable energy supply with nuclear power plant. In Korea, a reference disposal system for the spent fuels has been developed through a study on the direct disposal of the PWR and CANDU spent fuel. Recently, the research on the demonstration and the efficiency analyses of the disposal system has been performed to make the disposal system safer and more economic. PWR spent fuels which include a lot of reusable material can be considered being recycled and a study on the disposal of HLW from this recycling process is being performed. CANDU spent fuels are considered being disposed of directly in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System (KRS) which was to dispose of both PWR type and CANDU type, the more effective CANDU spent fuel disposal systems were developed. To do this, the disposal canister for CANDU spent fuels was modified to hold the storage basket for 60 bundles which is used in nuclear power plant. With these modified disposal canister concepts, the disposal concepts to meet the thermal requirement that the temperature of the buffer materials should not be over 100 .deg. C were developed. These disposal concepts were reviewed and analyzed in terms of disposal effective factors which were thermal effectiveness, U-density, disposal area, excavation volume, material volume etc. and the most effective concept was proposed. The results of this study will be used in the development of various wastes disposal system together with the HLW wastes from the PWR spent fuel recycling process.

AECL experience in fuel channel inspection

Energy Technology Data Exchange (ETDEWEB)

Van Drunen, G. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Gunn, R. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Mayo, W.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Scott, D.A. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

1999-06-01

Inspection of CANDU fuel channels (FC) is performed to ensure safe and economic reactor operation. CANDU reactor FCs have features that make them a unique non-destructive testing (NDT) challenge. The thin, 4 mm pressure-tube wall means flaws down to about 0.1 mm deep must be reliably detected and characterized. This is one to two orders of magnitude smaller than is usually considered of significant concern for steel piping and pressure vessels. A second unique feature is that inspection sensors must operate in the reactor core--often within 20 cm of highly radioactive fuel. Work on inspection of CANDU reactor FCs at AECL dates back over three decades. In that time, AECL staff have provided equipment and conducted or supervised in-service inspections in about 250 FCs, in addition to over 8000 pre-service FCs. These inspections took place at every existing CANDU reactor except those in India and Romania. Early FC inspections focussed on measurement of changes in dimensions (gauging) resulting from exposure to a combination of neutrons, stress and elevated temperature. Expansion of inspection activities to include volumetric inspection (for flaws) started in the mid-1970s with the discovery of delayed hydride cracking in Pickering 3 and 4 rolled joints. Recognition of other types of flaw mechanisms in the 1980s led to further expansion in both pre-service and in-service inspections. These growing requirements, to meet regulatory as well as economic needs, led to the development of a wide spectrum of inspection technology that now includes tests for hydrogen concentration, structural integrity of core components, flaws, and dimensional change. This paper reviews current CANDU reactor FC inspection requirements. The equipment and techniques developed to satisfy these requirements are also described. The paper concludes with a discussion of work in progress in AECL aimed at providing state-of-the-art FC inspection services. (author)

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

Behaviour of defective CANDU fuel: fuel oxidation kinetic and thermodynamic modelling

International Nuclear Information System (INIS)

Higgs, J.

2005-01-01

The thermal performance of operating CANDU fuel under defect conditions is affected by the ingress of heavy water into the fuel element. A mechanistic model has been developed to predict the extent of fuel oxidation in defective fuel and its affect on fuel thermal performance. A thermodynamic treatment of such oxidized fuel has been performed as a basis for the boundary conditions in the kinetic model. Both the kinetic and thermodynamic models have been benchmarked against recent experimental work. (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

Removal of a seized fuel channel from the KANUPP reactor

International Nuclear Information System (INIS)

Butt, W.M.; Gunn, R.J.

1995-01-01

In support of the Safe operation of KANUPP program, AECL was commissioned in early 1992 to assist the Karachi Nuclear Power Plant in the design and supply of equipment and procedures for removal of a seized fuel channel from the KANUPP CANDU reactor. In addition AECL was also asked to supply technical site support to assist the KANUPP station staff during the removal of the G-12 channel. The design of a fuel channel removal system presented an interesting challenge. The fuel channel design was unique to KANUPP with no history of previous channel removal, consequently nearly all tools and equipment had to be specially designed. In addition, the seized end fitting posed a special problem requiring the development several contingency tools and techniques. This paper is an account of the design and development of the removal system and the site experiences during the actual fuel channel removal. After the channel had been removed, it was confirmed that a corrosion seizure between the end fitting sleeve bearings was inhibiting normal channel elongation. (author)

CANDU fuel - fifteen years of power reactor experience

International Nuclear Information System (INIS)

Fanjoy, G.R.; Bain, A.S.

1977-05-01

Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) a thorough understanding of the fundamental behaviour of CANDU fuel; (b) data showing that the predicted high utilization of uranium has been achieved; (c) criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to 'CANLUB' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases; (d) proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). As of mid-1976 over 3 x 10 6 individual elements have been built and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification

CANDU fuel cycle economic efficiency assessments using the IAEA-MESSAGE-V code

International Nuclear Information System (INIS)

Prodea, Iosif; Margeanu, Cristina Alice; Aioanei, Corina; Prisecaru, Ilie; Danila, Nicolae

2007-01-01

The main goal of the paper is to evaluate different electricity generation costs in a CANDU Nuclear Power Plant (NPP) using different nuclear fuel cycles. The IAEA-MESSAGE code (Model for Energy Supply Strategy Alternatives and their General Environmental Impacts) will be used to accomplish these assessments. This complex tool was supplied by International Atomic Energy Agency (IAEA) in 2002 at 'IAEA-Regional Training Course on Development and Evaluation of Alternative Energy Strategies in Support of Sustainable Development' held in Institute for Nuclear Research Pitesti. It is worthy to remind that the sustainable development requires satisfying the energy demand of present generations without compromising the possibility of future generations to meet their own needs. Based on the latest public information in the next 10-15 years four CANDU-6 based NPP could be in operation in Romania. Two of them will have some enhancements not clearly specified, yet. Therefore we consider being necessary to investigate possibility to enhance the economic efficiency of existing in-service CANDU-6 power reactors. The MESSAGE program can satisfy these requirements if appropriate input models will be built. As it is mentioned in the dedicated issues, a major inherent feature of CANDU is its fuel cycle flexibility. Keeping this in mind, some proposed CANDU fuel cycles will be analyzed in the paper: Natural Uranium (NU), Slightly Enriched Uranium (SEU), Recovered Uranium (RU) with and without reprocessing. Finally, based on optimization of the MESSAGE objective function an economic hierarchy of CANDU fuel cycles will be proposed. The authors used mainly public information on different costs required by analysis. (authors)

Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

International Nuclear Information System (INIS)

Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M.

2012-01-01

The Enhanced CANDU 6 R (ECo R ) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

Hyperfine 3D neutronic calculations in CANDU supercells

International Nuclear Information System (INIS)

Balaceanu, V.; Aioanei, L.; Pavelescu, M.

2010-01-01

For an accurate evaluation of the fuel performances, it is very important to have capability to calculate the three dimensional spatial flux distributions in the fuel bundle. According this issue, in our Institute, a multigroup calculation methodology named WIMS-PIJXYZ was especially developed for estimating the local neutronic parameters in CANDU cell/supercells. The objective of this paper is to present this calculation methodology and to use it in performing some hyperfine neutronic calculations in CANDU type supercells. More exactly, after a short description for the WIMS-PIJXYZ methodology, the end effect for some CANDU fuel bundles is estimated. The WIMS-PIJXYZ methodology is based on WIMS and PIJXYZ transport codes. WIMS is a standard lattice-cell code and it is used for generating the multigroup macroscopic cross sections for the materials in the fuel cells. For obtaining the flux and power distributions in CANDU fuel bundles the PIJXYZ code is used. This code is consistent with WIMS lattice-cell calculations and allows a good geometrical representation of the CANDU bundle in three dimensions. The end effect consists in the increasing of the thermal neutron flux in the end region and the increasing of power in the end of the fuel rod. The region separating the CANDU fuel in two adjoining bundles in a channel is called the 'end region' and the end of the last pellet in the fuel stack adjacent to the end region is called the 'fuel end'. The end effect appears because the end region of the bundle is made up of coolant and Zircaloy-4, a very low neutron absorption material. To estimate the end effect, the flux peaking factors and the power peaking factors are calculated. It was taken in consideration CANDU Standard (Natural Uranium, with 37 elements) fuel bundles. In the end of the paper, the results obtained by WIMS-PIJXYZ methodology with the similar LEGENTR results are compared. The comparative analysis shows a good agreement. (authors)

Regional overpower protection system analysis for a DUPIC fuel CANDU core

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok; Park, Jee Won

2003-06-01

The regional overpower protection (ROP) system was assessed a CANDU 6 reactor with the DUPIC fuel, including the validation of the WIMS/RFSP/ROVER-F code system used for the estimation of ROP trip setpoint. The validation calculation has shown that it is valid to use the WIMS/RFSP/ROVER-F code system for ROP system analysis of the CANDU 6 core. For the DUPIC core, the ROP trip setpoint was estimated to be 125.7%, which is almost the same as that of the standard natural uranium core. This study has shown that the DUPIC fuel does not hurt the current ROP trip setpoint designed for the natural uranium CANDU 6 reactor

CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

International Nuclear Information System (INIS)

Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J.

1997-01-01

KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

Energy Technology Data Exchange (ETDEWEB)

Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J. [Korea Atomic Energy Research Inst., Yusong, Taejon (Korea, Republic of)

1997-07-01

KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

Once-through uranium thorium fuel cycle in CANDU reactors

International Nuclear Information System (INIS)

Ozdemir, S.; Cubukcu, E.

2000-01-01

In this study, the performance of the once-through uranium-thorium fuel cycle in CANDU reactors is investigated. (Th-U)O 2 is used as fuel in all fuel rod clusters where Th and U are mixed homogeneously. CANDU reactors have the advantage of being capable of employing various fuel cycle options because of its good neutron economy, continuous on line refueling ability and axial fuel replacement possibility. For lattice cell calculations transport code WIMS is used. WIMS cross-section library is modified to achieve precise lattice cell calculations. For various enrichments and Th-U mixtures, criticality, heavy element composition changes, diffusion coefficients and cross-sections are calculate. Reactor core is modeled by using the diffusion code CITATION. We conclude that an overall saving of 22% in natural uranium demand can be achieved with the use of Th cycle. However, slightly enriched U cycle still consumes less natural Uranium and is a lot less complicated. (author)

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

Directory of Open Access Journals (Sweden)

Gabi ROSCA FARTAT

2015-05-01

Full Text Available The objective of this paper is to present a possible solution for the designing of a device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The decommissioning activities are dismantling, demolition, controlled removal of equipment, components, conventional or hazardous waste (radioactive, toxic in compliance with the international basic safety standards on radiation protection. One as the most important operation in the final phase of the nuclear reactor dismantling is the decommissioning of fuel channels. For the fuel channels decommissioning should be taken into account the detailed description of the fuel channel and its components, the installation documents history, adequate radiological criteria for decommissioning guidance, safety and environmental impact assessment, including radiological and non-radiological analysis of the risks that can occur for workers, public and environment, the description of the proposed program for decommissioning the fuel channel and its components, the description of the quality assurance program and of the monitoring program, the equipments and methods used to verify the compliance with the decommissioning criteria, the planning of performing the final radiological assessment at the end of the fuel channel decommissioning. These will include also, a description of the proposed radiation protection procedures to be used during decommissioning. 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. The device shall be designed according to the radiation protection procedures. The decommissioning device assembly of the fuel channel components is composed of the device itself and moving platform support for coupling of the selected channel to be dismantled. The fuel channel decommissioning device is an autonomous device designed for

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

Thermal-hydraulic interfacing code modules for CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

Thermal-hydraulic interfacing code modules for CANDU reactors

International Nuclear Information System (INIS)

Liu, W.S.; Gold, M.; Sills, H.

1997-01-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis

Fuel Management in Candu Reactors Using Tabu Search

International Nuclear Information System (INIS)

Chambon, R.; Varin, E.

2008-01-01

Meta-heuristic methods are perfectly suited to solve fuel management optimization problem in LWR. Indeed, they are originally designed for combinatorial or integer parameter problems which can represent the reloading pattern of the assemblies. For the Candu reactors the problem is however completely different. Indeed, this type of reactor is refueled online. Thus, for their design at fuel reloading equilibrium, the parameter to optimize is the average exit burnup of each fuel channel (which is related to the frequency at which each channel has to be reloaded). It is then a continuous variable that we have to deal with. Originally, this problem was solved using gradient methods. However, their major drawback is the potential local optimum into which they can be trapped. This makes the meta-heuristic methods interesting. In this paper, we have successfully implemented the Tabu Search (TS) method in the reactor diffusion code DONJON. The case of an ACR-700 using 7 burnup zones has been tested. The results have been compared to those we obtained previously with gradient methods. Both methods give equivalent results. This validates them both. The TS has however a major drawback concerning the computation time. A problem with the enrichment as an additional parameter has been tested. In this case, the feasible domain is very narrow, and the optimization process has encountered limitations. Actually, the TS method may not be suitable to find the exact solution of the fuel management problem, but it may be used in a hybrid method such as a TS to find the global optimum region coupled with a gradient method to converge faster on the exact solution. (authors)

A Preliminary Study on the Reuse of the Recovered Uranium from the Spent CANDU Fuel Using Pyroprocessing

International Nuclear Information System (INIS)

Park, C. J.; Na, S. H.; Yang, J. H.; Kang, K. H.; Lee, J. W.

2009-01-01

During the pyroprocessing, most of the uranium is gathered in metallic form around a solid cathode during an electro-refining process, which is composed of about 94 weight percent of the spent fuel. In the previous study, a feasibility study has been done to reuse the recovered uranium for the CANDU reactor fuel following the traditional DUPIC (direct use of spent pressurized water reactor fuel into CANDU reactor) fuel fabrication process. However, the weight percent of U-235 in the recovered uranium is about 1 wt% and it is sufficiently re-utilized in a heavy water reactor which uses a natural uranium fuel. The reuse of recovered uranium will bring not only a huge economic profit and saving of uranium resources but also an alleviation of the burden on the management and the disposal of the spent fuel. The research on recycling of recovered uranium was carried out 10 years ago and most of the recovered uranium was assumed to be imported from abroad at that time. The preliminary results showed there is the sufficient possibility to recycle recovered uranium in terms of a reactor's characteristics as well as the fuel performance. However, the spent CANDU fuel is another issue in the storage and disposal problem. At present, most countries are considering that the spent CANDU fuel is disposed directly due to the low enrichment (∼0.5 wt%) of the discharge fissile content and lots of fission products. If mixing the spent CANDU fuel and the spent PWR fuel, the estimated uranium fissile enrichment will be about 0.6 wt% ∼ 1.0 wt% depending on the mixing ratio, which is sufficiently reusable in a CANDU reactor. Therefore, this paper deals with a feasibility study on the recovered uranium of the mixed spent fuel from the pyroprocessing. With the various mixing ratios between the PWR spent fuel and the CANDU spent fuel, a reactor characteristics including the safety parameters of the CANDU reactor was evaluated

Computer simulation of the behaviour and performance of a CANDU fuel rod

International Nuclear Information System (INIS)

Marino, A.C.

1997-01-01

At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

The travesty of discarding used CANDU fuel

Energy Technology Data Exchange (ETDEWEB)

Ottensmeyer, P. [Univ. of Toronto, Toronto, Ontario (Canada)

2016-09-15

The current plan worldwide for virtually all used nuclear fuels is costly deep burial to attempt to isolate their long-term radiotoxicity permanently. Alternatively Canada's 50,000 tons spent CANDU fuel, of which only 0.74% of the heavy atoms have been fissioned to extract their energy, could supply 130 times more non-carbon energy using proven economical recycling and fast-neutron technologies. The result in this country alone would currently be the creation of $74 trillion of reliable electricity on demand without greenhouse gas emissions. It would avoid adding 475 billion tons CO{sub 2} to the atmosphere compared to the use of coal, to mitigate climate change. Worldwide recycling of stored spent nuclear fuel and replenishing with depleted uranium in fast-neutron reactors could avoid emitting over 20 trillion tons CO{sub 2}, or over six times the current total atmospheric CO{sub 2} content. As added bonus the long-term radiotoxicity of the used CANDU fuel is effectively eliminated, making a long-term deep geological repository unnecessary. Even the shorter-lived radioisotope fission products become valuable stable atoms and minerals that would fetch $3 million per ton. Such an alternative is certainly worth pursuing. (author)

Finite element modelling of different CANDU fuel bundle types in various refuelling conditions

International Nuclear Information System (INIS)

Roman, M. R.; Ionescu, D. V.; Olteanu, G.; Florea, S.; Radut, A. C.

2016-01-01

The objective of this paper is to develop a finite element model for static strength analysis of the CANDU standard with 37 elements fuel bundle and the SEU43 with 43 elements fuel bundle design for various refuelling conditions. The computer code, ANSYS7.1, is used to simulate the axial compression in CANDU type fuel bundles subject to hydraulic drag loads, deflection of fuel elements, stresses and displacements in the end plates. Two possible situations for the fuelling machine side stops are considered in our analyses, as follows: the last fuel bundle is supported by the two side stops and a side stop can be blocked therefore, the last fuel bundle is supported by only one side stop. The results of the analyses performed are briefly presented and also illustrated in a graphical form. The finite element model developed in present study is verified against test results for endplate displacement and element bowing obtained from strength tests with fuel bundle string and fuelling machine side-stop simulators. Comparison of ANSYS model predictions with these experimental results led to a very good agreement. Despite the difference in hydraulic load between SEU43 and CANDU standard fuel bundles strings, the maximum stress in the SEU43 endplate is about the same with the maximum stress in the CANDU standard endplate. The comparative assessment reveals that SEU43 fuel bundle is able to withstand high flow rate without showing a significant geometric instability. (authors)

Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

6. CNS international conference on CANDU maintenance. Proceedings

International Nuclear Information System (INIS)

2003-01-01

The 6th CNS International Conference on CANDU Maintenance took place in Toronto, Ontario on November 16-18, 2003. The theme for the conference was 'Maintenance for Life'. About 270 delegates attended the conference held by the Canadian Nuclear Society. The conference consisted of four parallel sessions, a pattern that continued throughout the conference. Papers were grouped under the following headings: Fuel Channels and End Fittings - Assessments; Fuel Channels and End Fittings - Inspections; Fuel Channels and End Fittings - Maintenance; Fuel Channels and End Fittings - Universal Delivery Machine; Water Upgrading; Performance and Plant Life Improvement; Steam Generator Life Management; Steam Generator Modifications; Steam Generators - Inspections; Steam Generators - Assessments; Maintenance Programs; Feeder Inspections; Feeder Assessment and Mitigation; Valve Maintenance; Instrumentation and Control; Inspection Technology; and Fuel Handling

Neutron-photon energy deposition in CANDU reactor fuel channels: a comparison of modelling techniques using ANISN and MCNP computer codes

International Nuclear Information System (INIS)

Bilanovic, Z.; McCracken, D.R.

1994-12-01

In order to assess irradiation-induced corrosion effects, coolant radiolysis and the degradation of the physical properties of reactor materials and components, it is necessary to determine the neutron, photon, and electron energy deposition profiles in the fuel channels of the reactor core. At present, several different computer codes must be used to do this. The most recent, advanced and versatile of these is the latest version of MCNP, which may be capable of replacing all the others. Different codes have different assumptions and different restrictions on the way they can model the core physics and geometry. This report presents the results of ANISN and MCNP models of neutron and photon energy deposition. The results validate the use of MCNP for simplified geometrical modelling of energy deposition by neutrons and photons in the complex geometry of the CANDU reactor fuel channel. Discrete ordinates codes such as ANISN were the benchmark codes used in previous work. The results of calculations using various models are presented, and they show very good agreement for fast-neutron energy deposition. In the case of photon energy deposition, however, some modifications to the modelling procedures had to be incorporated. Problems with the use of reflective boundaries were solved by either including the eight surrounding fuel channels in the model, or using a boundary source at the bounding surface of the problem. Once these modifications were incorporated, consistent results between the computer codes were achieved. Historically, simple annular representations of the core were used, because of the difficulty of doing detailed modelling with older codes. It is demonstrated that modelling by MCNP, using more accurate and more detailed geometry, gives significantly different and improved results. (author). 9 refs., 12 tabs., 20 figs

Analysis of neutron flux depression across the pellet radius in CANDU fuel elements

International Nuclear Information System (INIS)

Sim, K.S.; Suk, H.C.

1998-08-01

The TUBRNP model, originally developed to perform the analysis of the flux depression across the pellet radius in LWR fuel elements, was improved for the application to CANDU fuel elements. The improved model was verified through comparison with existing CANDU model named FLUXDEP in prediction for various fuel conditions. A sensitivity study was also performed to investigate the effects on the flux depression of fuel initial enrichment and burnup, the contents of isotopes U-234 and U-236 and pellet diameter. (author). 9 refs., 8 figs

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)

CANDU with supercritical water coolant: conceptual design features

International Nuclear Information System (INIS)

Spinks, N.

1997-01-01

An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

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)

The future role of thorium in assuring CANDU fuel supplies

International Nuclear Information System (INIS)

Slater, J.B.

1985-01-01

Atomic Energy of Canada Limited (AECL), in partnership with Canadian industry and power utilities, has developed the CANDU reactor as a safe, reliable and economic means of transforming nuclear fuel into useable power. The use of thorium/uranium-233 recycle gives the possibility of a many-fold increase in energy yield over that which can be obtained from the use of uranium in once-through cycles. The neutronic properties of uranium-233 combine with the inherent neutron economy of the CANDU reactor to offer the possibility of near-breeder cycles in which there is no net consumption of fissile material under equilibrium fuelling conditions. Use of thorium cycles in CANDU will limit the impact of higher uranium prices. When combined with the potential for significant reductions in CANDU capital costs, then the long-term prospect is for generating costs near to current levels. Development of thorium cycles in CANDU will safeguard against possible uranium shortages in the next century, and will maintain and continue the commercial viability of CANDU as a long-term energy technology. (author)

In-situ optical profilometry of CANDU fuel channels

Energy Technology Data Exchange (ETDEWEB)

Jarvis, G.N.; Cornblum, E.O.; Grabish, M.G. [Ontario Hydro Nuclear, Nuclear Technology Services, Pickering, ON (Canada); Mader, D.L.; Kuurstra, J.C.; McNabb, S.C. [Ontario Hydro Technologies, Toronto, ON (Canada)

1996-12-31

This paper describes 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 use of optical profilometry in one form or another is not new. Systems typically make use of scanned laser beams to perform tasks such as weld bead profiling and contour mapping of various solid objects. The specific application of such techniques within the harsh, confined environmental conditions described herein is thought to be somewhat unique and has provided some technical challenges. A brief outline of the fuel channel geometry will be given together with the basics of the profilometry technique employed. This will be followed by a detailed description of the in-channel tooling, key features of the data collection and analysis software and operational experience. (UK).

In-situ optical prolifometry of CANDU fuel channels

Energy Technology Data Exchange (ETDEWEB)

Jarvis, G.N.; Cornblum, E.O.; Grabish, M.G. [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station; Mader, D.L.; Kuurstra, J.C.; McNabb, S.C. [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

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 use of optical profilometry in one form or another is not new. Systems typically make use of scanned laser beams to perform tasks such as weld bead profiling and contour mapping of various solid objects. The specific application of such techniques within the harsh, confined environmental conditions described herein is thought to be somewhat unique and has provided some technical challenges. A brief outline of the fuel channel geometry will be given together with the basics of the profilometry technique employed. This will be followed by a detailed description of the in-channel tooling, key features of the data collection and analysis software and operational experience. (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

ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

International Nuclear Information System (INIS)

Rao, Y.F.; Cheng, Z.; Waddington, G.M.; Nava-Dominguez, A.

2014-01-01

Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles

ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca; Nava-Dominguez, A., E-mail: navadoma@aecl.ca

2014-08-15

Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

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, verification and validation of the fuel channel behaviour computer code FACTAR

Energy Technology Data Exchange (ETDEWEB)

Westbye, C J; Brito, A C; MacKinnon, J C; Sills, H E; Langman, V J [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

FACTAR (Fuel And Channel Temperature And Response) is a computer code developed to simulate the transient thermal and mechanical behaviour of 37-element or 28-element fuel bundles within a single CANDU fuel channel for moderate loss of coolant accident conditions including transition and large break LOCA`s (loss of coolant accidents) with emergency coolant injection assumed available. FACTAR`s predictions of fuel temperature and sheath failure times are used to subsequent assessment of fission product releases and fuel string expansion. This paper discusses the origin and development history of FACTAR, presents the mathematical models and solution technique, the detailed quality assurance procedures that are followed during development, and reports the future development of the code. (author). 27 refs., 3 figs.

The evolution of Candu fuel cycles and their potential contribution to world peace

International Nuclear Information System (INIS)

Whitlock, J.

2001-01-01

The Candu(r) reactor is the most versatile commercial power reactor in the world. It has the potential to extend resource utilization significantly, to allow countries with developing industrial infrastructures access to clean and abundant energy, and to destroy long-lived nuclear waste or surplus weapons plutonium. These benefits are available by choosing from an array of possible fuel cycles. Several factors, including Canada's early focus on heavy-water technology, limited heavy-industry infrastructure at the time, and a desire for both technological autonomy and energy self-sufficiency, contributed to the creation of the first Candu reactor in 1962. With the maturation of the CANDU industry, the unique design features of the now-familiar product - on-power refuelling, high neutron economy, and simple fuel design - make possible the realization of its potential fuel-cycle versatility. Several fuel-cycle options currently under development are described. (authors)

New approach to derive linear power/burnup history input for CANDU fuel codes

International Nuclear Information System (INIS)

Lac Tang, T.; Richards, M.; Parent, G.

2003-01-01

The fuel element linear power / burnup history is a required input for the ELESTRES code in order to simulate CANDU fuel behavior during normal operating conditions and also to provide input for the accident analysis codes ELOCA and SOURCE. The purpose of this paper is to present a new approach to derive 'true', or at least more realistic linear power / burnup histories. Such an approach can be used to recreate any typical bundle power history if only a single pair of instantaneous values of bundle power and burnup, together with the position in the channel, are known. The histories obtained could be useful to perform more realistic simulations for safety analyses for cases where the reference (overpower) history is not appropriate. (author)

MMOSS-I: a CANDU multiple-channel thermosyphoning flow stability model

Energy Technology Data Exchange (ETDEWEB)

Gulshani, P [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Huynh, H [Hydro-Quebec, Montreal, PQ (Canada)

1996-12-31

This paper presents a multiple-channel flow stability model, dubbed MMOSS, developed to predict the conditions for the onset of flow oscillations in a CANDU-type multiple-channel heat transport system under thermosyphoning conditions. The model generalizes that developed previously to account for the effects of any channel flow reversal. Two-phase thermosyphoning conditions are predicted by thermalhydraulic codes for some postulated accident scenarios in CANDU. Two-phase thermosyphoning experiments in the multiple-channel RD-14M facility have indicated that pass-to-pass out-of-phase oscillations in the loop conditions caused the flow in some of the heated channels to undergo sustained reversal in direction. This channel flow reversal had significant effects on the channel and loop conditions. It is, therefore, important to understand the nature of the oscillations and be able to predict the conditions for the onset of the oscillations or for stable flow in RD-14M and the reactor. For stable flow conditions, oscillation-induced channel flow reversal is not expected. MMOSS was developed for a figure-of-eight system with any number of channels. The system characteristic equation was derived from a linearization of the conservation equations. In this paper, the MMOSS characteristic equation is solved for a system of N identical channel assemblies. The resulting model is called MMOSS-I. This simplification provides valuable physical insight and reasonably accurate results. MMOSS-I and a previously-developed steady-state model THERMOSYPHON are used to predict thermosyphoning flow stability maps for RD-14M and the Gentilly 2 reactor. (author). 11 refs., 7 figs.

Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

International Nuclear Information System (INIS)

Lightston, M.F.; Rock, R.

1996-01-01

This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

A feasible approach to implement a commercial scale CANDU fuel manufacturing plant in Egypt

International Nuclear Information System (INIS)

El-Shehawy, I.; El-Sharaky, M.; Yasso, K.; Selim, I.; Graham, N.; Newington, D.

1995-01-01

Many planning scenarios have been examined to assess and evaluate the economic estimates for implementing a commercial scale CANDU fuel manufacturing plant in Egypt. The cost estimates indicated strong influence of the annual capital costs on total fuel manufacturing cost; this is particularly evident in a small initial plant where the proposed design output is only sufficient to supply reload fuel for a single CANDU-6 reactor. A modular approach is investigated as a possible way, to reduce the capital costs for a small initial fuel plant. In this approach the plant would do fuel assembly operations only and the remainder of a plant would be constructed and equipped in the stages when high production volumes can justify the capital expenses. Such approach seems economically feasible for implementing a small scale CANDU fuel manufacturing plant in developing countries such as Egypt and further improvement could be achieved over the years of operation. (author)

Prediction of power-ramp defects in CANDU fuel

International Nuclear Information System (INIS)

Gillespie, P.; Wadsworth, S.; Daniels, T.

2010-01-01

Power ramps result in fuel pellet expansion and can lead to fuel sheath failures by fission product induced stress corrosion cracking (SCC). Historically, empirical models fit to experimental test data were used to predict the onset of power-ramp failures in CANDU fuel. In 1988, a power-ramped fuel defect event at PNGS-1 led to the refinement of these empirical models. This defect event has recently been re-analyzed and the empirical model updated. The empirical model is supported by a physically based model which can be used to extrapolate to fuel conditions (density, burnup) outside of the 1988 data set. (author)

Effect of bundle junction face and misalignment on the pressure drops across a randomly loaded and aligned 12 bundles in CANDU fuel channel

Energy Technology Data Exchange (ETDEWEB)

Suk, H. C.; Sim, K. S.; Chang, C. H.; Lee, Y. O. [Korea Atomic Energy Reaearch Institute, Taejon (Korea, Republic of)

1996-06-01

The pressure drop of twelve fuel bundle string in the CANDU-6 fuel channel is equal to the sum of the eleven junction pressure losses, the bundle string entrance and exit pressure losses, the skin friction pressure loss, and other appendage pressure losses, where the junction loss is dependent on the bundle and faces and angular alignments of the junctions. The results of the single junction pressure drop tests in a short rig show that the most probable pressure drop of the eleven junction was analytically equal to the eleven times of average pressure drop of all the possible single junction pressure drops, and also that the largest and smallest junction pressure drops across the eleven junctions probably occurred only with BA and BB type junctions, respectively, where A and B denote the bundle end sides with an end-plates on which a company monogram is stamped and unstamped, respectively. 5 refs., 7 figs., 1 tab. (author).

Some notes on the Timing of Geological Disposal of CANDU Spent Fuels

International Nuclear Information System (INIS)

Choi, Heui Joo; Kook, Dong Hak; Choi, Jong Won

2010-01-01

CANDU spent fuel is to be disposed of at repository finally rather than recycled because of its low fissile nuclide concentration. But the difficult situation of finding a repository site can not help introducing a interim storage in the short term. It is required to find an optimum timing of geological disposal of CANDU spent fuels related to the interim storage operation period. The major factors for determining the disposal starting time are considered as safety, economics, and public acceptance. Safety factor is compared in terms of the decay heat and non-proliferation. Economics factor is compared from the point of the operation cost, and public acceptance factor is reviewed from the point of retrievability and inter-generation ethics. This paper recommended the best solution for the disposal starting time by analyzing the above factors. It is concluded that the optimum timing for the CANDU spent fuel disposal is around 2041 and that the sooner disposal time, the better from the point of technical and safety aspects.

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

Dynamic Analysis of the Thorium Fuel Cycle in CANDU Reactors

International Nuclear Information System (INIS)

Jeong, Chang Joon; Park, Chang Je

2006-02-01

The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO 2 UO 2 and ThO 2 UO 2 -DUPIC fuels. The recycling is performed through the dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, the thorium fuel CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products of the multiple recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the SF accumulation and save the natural uranium resource compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible. It is recommended, however, the economic analysis should be performed in future

Dynamic Analysis of the Thorium Fuel Cycle in CANDU Reactors

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Park, Chang Je

2006-02-15

The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO{sub 2}UO{sub 2} and ThO{sub 2}UO{sub 2}-DUPIC fuels. The recycling is performed through the dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, the thorium fuel CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products of the multiple recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the SF accumulation and save the natural uranium resource compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible. It is recommended, however, the economic analysis should be performed in future.

Development of CANDU fuel in-bay inspection and dimensional measurement system

Energy Technology Data Exchange (ETDEWEB)

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

2004-09-01

In this report, we describe a spent fuel inspection and dimensional measuring system for CANDU fuel bundles in atmosphere or water. The submissible camera with radiation resistance was used to inspect a spent fuel in water. The design criteria of the dimensional measuring system was set up {+-}0.01mm(10{mu}m) for measuring accuracy. The LVDT (Linear Variable Differential Transformer) was used as measuring sensor in this dimensional measuring system. An LVDT calibration equipment was developed in order to satisfy the required accuracy of the system. Also, aluminum master and CANDU Master with same dimension of fuel bundle was made to calibrate the mechanical error of the dimensional measuring system. The accuracy of the fuel inspection system was examined. The results show that the accuracy in dimensional measurement of fuel rod profile and bearing pad profile, diameter of fuel bundle, fuel rod length, and end plate profile using standard test equipment satisfies the design criteria, i.e., maximum measurement error of {+-}0.01mm(10{mu}m)

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)

Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

EL-Khawlani, Afrah [Physics Department, Sana' a (Yemen); Aziz, Moustafa [Nuclear and radiological regulatory authority, Cairo (Egypt); Ismail, Mahmud Yehia; Ellithi, Ali Yehia [Cairo Univ. (Egypt). Faculty of Science

2015-03-15

The MCNPX (Monte Carlo N-Particle Transport Code System) code has been used for modeling and simulation of a half core of CANDU (CANada Deuterium-Uranium) reactor, both homogenous and heterogeneous model for the reactor core are designed. The fuel is burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor for homogeneous and heterogeneous reactor core is calculated and compared during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the model. The flux and power distributions through channels are calculated.

Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

International Nuclear Information System (INIS)

Clegg, L.J.; Coady, J.R.

1977-01-01

The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)

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

Integrity of spent CANDU fuel during and following dry storage

International Nuclear Information System (INIS)

Villagran, J.E.

2004-01-01

This report examines the issue of CANDU fuel integrity at the back end of the fuel cycle and outlines a program designed to provide assurance that used CANDU fuel will retain its integrity over an extended period. In specific terms, the program is intended to provide assurance that during and following extended dry storage the fuel will remain fit to undergo, without loss of integrity, the handling, packaging and transportation operations that might be necessary until it is placed in disposal containers. The first step in the development of the program was a review of the available technical information on phenomena relevant to fuel integrity. The major conclusions from that review were the following: Under normal storage conditions it is unlikely that the spent fuel will suffer significant degradation during a one-hundred year period and it should be possible to retrieve, repackage and transport the fuel as required, using methods and systems similar to those used today. However, to provide increased confidence regarding the above conclusion, investigations should be conducted in areas where there is higher uncertainty in the prediction of fuel condition and on some degradation processes to which the fuel appears to present higher vulnerability. The proposed program includes, among other tasks, irradiated fuel tests, analytical studies on the most relevant fuel degradation processes and the development of a long-term fuel verification program. (Author)

Coolability of severely degraded CANDU cores. Revised

International Nuclear Information System (INIS)

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

1996-01-01

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

Coolability of severely degraded CANDU cores

International Nuclear Information System (INIS)

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

1995-07-01

Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually re solidify. Thus, the calandria vessel would act inherently as a core-catcher as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author). 48 refs., 3 tabs., 18 figs

Dimensional response of CANDU fuel to power changes

Energy Technology Data Exchange (ETDEWEB)

Fehrenbach, P J [Fuel Engineering Branch, Chalk River Nuclear Laboratories, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Hastings, I J; Morel, P A; Sage, R D; Smith, A D [Fuel Materials Branch, Chalk River Nuclear Laboratories, Atomic Energy of Canada Limited, Chalk River, ON (Canada)

1983-06-01

The introduction of CANLUB-coated fuel cladding, modified fuel management schemes, and revisions to the sequence of control rod movements, have eliminated power ramping fuel failures in CANDU power reactors. However, an irradiation program continues at Chalk River Nuclear Laboratories to determine the effect of various design and operating parameters on the dimensional response of UO{sub 2} fuel elements to power changes, over a range of conditions outside those normally experienced by CANDU power reactor fuel. We have investigated the effect of power changes on element diameter for UO{sub 2} fuel with starting densities of 10.6 and 10.8 Mg/m{sup 3} clad in 0.4 mm thick Zircaloy, at burnups from 0 to 100 MW.h/kg U. Element diameter measurements were obtained at power using an In-Reactor Diameter Measuring Rig (IRDMR). Rates of power change over the range 0.0005 to 0.03 kW.m{sup -1}.s{sup -1} were achieved by a combination of reactor power control and use of a Helium-3 power cycling facility. Total diameter increases in unirradiated elements were about 1% at pellet interface locations for both fuel densities during the initial power increase to 60 kW/m. Diameter changes during subsequent power cycles of these elements from 55 to 100% maximum power were significantly larger for the higher density fuel, ranging from 0.3 to 0.5% compared to less than 0.1% for the standard density (10.6 Mg/m{sup 3}) fuel. In elements pre-irradiated at 27 kW/m to burnups of about 100 MW.h/kg U prior to power ramping, the diameter increases measured after ramping to 55 kW/m also varied with starting fuel density. Diameter changes at pellet interface locations were about 0.9% and 0.6% for higher density and standard density fuel respectively. (author)

Core design and fuel management studies

International Nuclear Information System (INIS)

Min, Byung Joo; Chan, P.

1997-06-01

The design target for the CANDU 9 requires a 20% increase in electrical power output from an existing 480-channel CANDU core. Assuming a net electrical output of 861 MW(e) for a natural uranium fuelled Bruce-B/Darlington reactor in a warm water site, the net electrical output of the reference CANDU 9 reactor would be 1033 MW(e). This report documents the result of the physics studies for the design of the CANDU 9 480/SEU core. The results of the core design and fuel management studies of the CANDU 9 480/SEU reactor indicated that up to 1033 MW(e) output can be achieved in a 480-channel CANDU core by using SEU core can easily be maintained indefinitely using an automated refuelling program. Fuel performance evaluation based on the data of the 500 FPDs refuelling simulation concluded that SEU fuel failure is not expected. (author). 2 tabs., 38 figs., 5 refs

Development of a flow restrictor for CANDU fuel channels

International Nuclear Information System (INIS)

Schroeter, F.; Antonaccio, C.; Masciotra, H.; Klink, A.

2013-01-01

Due to the creep and neutron growth phenomena experienced by the components inside the reactor during operation of CNE it is expected that both the fuel channels and the Liquid Injection lances increase their permanent deformation. One of the deformation types that these two components experiment is the SAG, which is what happens with any beam supported on their extremes to which a load is applied, except for this case that due to creep and neutronic effect growth, part of the deformation is not elastic and increases with time To solve or avoid this condition, two solutions exist, one is to replace the pressure tubes, forcing the calandria tube to recover to a near to original position or to design a device that permits defueling of the channel without modifying the pressure drop and in this way not to affect the distribution of coolant in the core. In some channels it was decided to replace the pressure tube and in others it was decided to defuel them proposing a design for a flow restrictor. (author

Passive heat removal in CANDU

International Nuclear Information System (INIS)

Hart, R.S.

1997-01-01

CANDU has a tradition of incorporating passive systems and passive components whenever they are shown to offer performance that is equal to or better than that of active systems, and to be economic. Examples include the two independent shutdown systems that employ gravity and stored energy respectively, the dousing subsystem of the CANDU 6 containment system, and the ability of the moderator to cool the fuel in the event that all coolant is lost from the fuel channels. CANDU 9 continues this tradition, incorporating a reserve water system (RWS) that increases the inventory of water in the reactor building and profiles a passive source of makeup water and/or heat sinks to various key process systems. The key component of the CANDU 9 reserve water system is a large (2500 cubic metres) water tank located at a high elevation in the reactor building. The reserve water system, while incorporating the recovery system functions, and the non-dousing functions of the dousing tank in CANDU 6, embraces other key systems to significantly extend the passive makeup/heat sink capability. The capabilities of the reserve water system include makeup to the steam generators secondary side if all other sources of water are lost; makeup to the heat transport system in the event of a leak in excess of the D 2 O makeup system capability; makeup to the moderator in the event of a moderator leak when the moderator heat sink is required; makeup to the emergency core cooling (ECC) system to assure NPSH to the ECC pumps during a loss of coolant accident (LOCA), and provision of a passive heat sink for the shield cooling system. Other passive designs are now being developed by AECL. These will be incorporated in future CANDU plants when their performance has been fully proven. This paper reviews the passive heat removal systems and features of current CANDU plants and the CANDU 9, and briefly reviews some of the passive heat removal concepts now being developed. (author)

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.

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

Fuel channel in-service inspection programs program design for maximum cost effectiveness

International Nuclear Information System (INIS)

Van den Brekel, N.C.

1995-01-01

Inspection is an integral part of fuel channel life management strategy. Inspection data is used to assess the state of reactor core integrity and provide the information necessary to optimize long term maintenance programs. This paper will provide an overview of the structured approach to developing fuel channel inspection programs within OHN. The inspection programs are designed to balance the resources utilized (cost, outage time, and dose expenditure) with the benefits provided by the inspection data obtained (improved knowledge of component status, degradation mechanisms and rates, etc..). The CANDU community has yet to have a fuel channel operate for a full 30 year design life. Since research programs can not fully simulate reactor operating conditions, inspections become an essential feature of the life management strategy as the components age. Inspection programs often include activities designed to develop predictive capability for long term fuel channel behaviour and provide early warning of changes in behaviour. It should be noted that although this paper addresses the design of fuel channel inspection programs, the basic principles presented can be applied to the design of inspection programs for any major power plant component or system. (author)

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

A CANDU designed for more tolerance to failures in large components

International Nuclear Information System (INIS)

Spinks, N.J.; Barclay, F.W.; Allen, P.J.; Yee, F.

1988-06-01

Current designs of CANDU reactors have several groups of fuel channels each served by an upstream coolant supply-train consisting of an outlet header, a steam generator, one or more pumps in parallel and an inlet header. Postulated failures in these large components put the heaviest demands on the safety systems. For example, the rupture of a header sets the requirements for the speed of shutdown and for the speed and capacity of emergency coolant injection, and it has a large impact on containment design. A CANDU design is being investigated to reduce the impact of failures in large components. Each group of fuel channels is supplied by more than one train so that if one train fails the rest continue to work. Reverse flow limiters reduce the loss-of-coolant from the unbroken trains to a broken supply train. The paper describes several design options for making the piping connections from multi supply-trains to fuel channels. It discusses progress in design and testing of flow limiters. A preliminary analysis is given of affected accidents

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)

Overview of the CANDU fuel handling system for advanced fuel cycles

International Nuclear Information System (INIS)

Koivisto, D.J.; Brown, D.R.

1997-01-01

Because of its neutron economies and on-power re-fuelling capabilities the CANDU system is ideally suited for implementing advanced fuel cycles because it can be adapted to burn these alternative fuels without major changes to the reactor. The fuel handling system is adaptable to implement advanced fuel cycles with some minor changes. Each individual advanced fuel cycle imposes some new set of special requirements on the fuel handling system that is different from the requirements usually encountered in handling the traditional natural uranium fuel. These changes are minor from an overall plant point of view but will require some interesting design and operating changes to the fuel handling system. Some preliminary conceptual design has been done on the fuel handling system in support of these fuel cycles. Some fuel handling details were studies in depth for some of the advanced fuel cycles. This paper provides an overview of the concepts and design challenges. (author)

Assessment of reactivity devices for CANDU-6 with DUPIC fuel

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok

1998-01-01

Reactivity device characteristics for a CANDU-6 reactor loaded with DUPIC fuel have been assessed. A transport code WIMS-AECL and a three-dimensional diffusion code RFSP were used for the lattice parameter generation and the core calculation, respectively. Three major reactivity devices have been assessed for their inherent functions. For the zone controller system, damping capability for spatial oscillation was investigated. The restart capability of the adjuster system was investigated. The shim operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster rod system. The mechanical control absorber was assessed for the capability to compensate the temperature reactivity feedback following a power reduction. This study has shown that the current reactivity device systems retain their functions when used in a DUPIC fuel CANDU reactor

CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

International Nuclear Information System (INIS)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

1999-04-01

A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

1999-04-01

A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

Reliability assessment of the fueling machine of the CANDU reactor

International Nuclear Information System (INIS)

Al-Kusayer, T.A.

1985-01-01

Fueling of CANDU-reactors is carried out by two fueling machines, each serving one end of the reactor. The fueling machine becomes a part of the primary heat transport system during the refueling operations, and hence, some refueling machine malfunctions could result in a small scale-loss-of-coolant accident. Fueling machine failures and the failure sequences are discussed. The unavailability of the fueling machine is estimated by using fault tree analysis. The probability of mechanical failure of the fueling machine interface is estimated as 1.08 x 10 -5 . (orig.) [de

The Key-Role of shielding analysis in advanced Candu Fuel bundles nuclear safety improvement for some accidental criticality scenarios

International Nuclear Information System (INIS)

Margeanu, C.A.; Rizoiu, A.; Olteanu, G.

2008-01-01

The paper aims to present the source term and photon dose rates estimation for advanced Candu fuel bundles in some accidental criticality scenarios. As reference, the Candu standard fuel bundle has been used. The scenarios take into account for a very short-time irradiated or spent fuel bundles for some configurations closed to criticality. In order to estimate irradiated fuel characteristic parameters and radiation doses, the ORNL's SCALE 5 codes Origin-S and Monte Carlo MORSE-SGC have been used. The paper includes the irradiated fuel characteristic parameters comparison for the considered Candu fuel bundles, providing also a comparison between the corresponding radiation doses

Reactor physics innovations of the advanced CANDU reactor core: adaptable and efficient

International Nuclear Information System (INIS)

Chan, P.S.W.; Hopwood, J.M.; Bonechi, M.

2003-01-01

The Advanced CANDU Reactor (ACR) is designed to have a benign, operator-friendly core physics characteristic, including a slightly negative coolant-void reactivity and a moderately negative power coefficient. The discharge fuel burnup is about three times that of natural uranium fuel in current CANDU reactors. Key features of the reactor physics innovations in the ACR core include the use of H 2 O coolant, slightly enriched uranium (SEU) fuel, and D 2 O moderator in a reduced lattice pitch. These innovations result in substantial improvements in economics, as well as significant enhancements in reactor performance and waste reduction over the current reactor design. The ACR can be readily adapted to different power outputs by increasing or decreasing the number of fuel channels, while maintaining identical fuel and fuel-channel characteristics. The flexibility provided by on-power refuelling and simple fuel bundle design enables the ACR to easily adapt to the use of plutonium and thorium fuel cycles. No major modifications to the basic ACR design are required because the benign neutronic characteristics of the SEU fuel cycle are also inherent in these advanced fuel cycles. (author)

Development of the CATHENA fuel channel model for an integrated blowdown and post-blowdown analysis for a 37-element CANDU fuel channel

International Nuclear Information System (INIS)

Rhee, B.W.; Shin, T.Y.; Yoo, K.M.; Kim, H.T.; Min, B.-J.; Park, J.H.

2006-01-01

The objective of this study is to develop a new fuel channel safety analysis system for covering both the blowdown analysis including the power pulse and the post-blowdown analysis with the same safety analysis code, CATHENA in a consistent manner. This new safety analysis methodology for a fuel channel analysis is expected to be better than the previous one used for the Wolsong 2,3,4 licensing which used CATHENA for the blowdown analysis and CHAN-II for the post-blowdown analysis, in several areas; consistency in the computer codes used and the modeling methods, the degree of uncertainty in the modeling and calculation. For this aim the existing CATHENA subchannel fuel channel model for a post blowdown analysis has been modified, and thus improved, and a processing program that conveys all the final state of the fuel channel at the end of blowdown analysis to the post-blowdown analysis as the initial conditions has been developed, and tested for its proper implementation for the intended purposes. A comparison of the results of this new analysis method with those of the Wolsong 2/3/4 Safety Analysis confirmed that the total heat transfer rate matches well up to 1000 sec, and then that of the new method begins to under-predict it consistently. On the other hand, the fuel temperatures of the center pin, inner ring fuel and the middle ring fuel are predicted by this new method to be lower than the old method by about 200 - 250 o C at the peak time. Considering the differences in these two analyses methodologies, especially the modeling of the fuel ring, a subchannel flow passage with an intermixing, and the radiation among the solid structures by considering every fuel individually, this trend of the results seems to be physically reasonable. However considerable future validation works are necessary to justify this new methodology for a licensing. (author)

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

An improved system to verify CANDU spent fuel elements in dry storage silos

International Nuclear Information System (INIS)

Almeida, Gevaldo L. de; Soares, Milton G.; Filho, Anizio M.; Martorelli, Daniel S.; Fonseca, Manoel

2000-01-01

An improved system to verify CANDU spent fuel elements stored in dry storage silos was developed. It is constituted by a mechanical device which moves a semi-conductor detector along a vertical verification pipe incorporated to the silo, and a modified portable multi-channel analyzer. The mechanical device contains a winding drum accommodating a cable hanging the detector, in such a way that the drum rotates as the detector goes down due to its own weight. The detector is coupled to the multi-channel analyzer operating in the multi-scaler mode, generating therefore a spectrum of total counts against time. To assure a linear transformation of time into detector position, the mechanical device dictating the detector speed is controlled by the multi-channel analyzer. This control is performed via a clock type escapement device activated by a solenoid. Whenever the multi-channel analyzer shifts to the next channel, the associated pulse is amplified, powering the solenoid causing the drum to rotate a fixed angle. Spectra taken in laboratory, using radioactive sources, have shown a good reproducibility. This qualify the system to be used as an equipment to get a fingerprint of the overall distribution of the fuel elements along the silo axis, and hence, to verify possible diversion of the nuclear material by comparing spectra taken at consecutive safeguards inspections. All the system is battery operated, being thus capable to operate in the field where no power supply is available. (author)

An improved system to verify CANDU spent fuel elements in dry storage silos

Energy Technology Data Exchange (ETDEWEB)

Almeida, Gevaldo L. de; Soares, Milton G.; Filho, Anizio M.; Martorelli, Daniel S.; Fonseca, Manoel [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)

2000-07-01

An improved system to verify CANDU spent fuel elements stored in dry storage silos was developed. It is constituted by a mechanical device which moves a semi-conductor detector along a vertical verification pipe incorporated to the silo, and a modified portable multi-channel analyzer. The mechanical device contains a winding drum accommodating a cable hanging the detector, in such a way that the drum rotates as the detector goes down due to its own weight. The detector is coupled to the multi-channel analyzer operating in the multi-scaler mode, generating therefore a spectrum of total counts against time. To assure a linear transformation of time into detector position, the mechanical device dictating the detector speed is controlled by the multi-channel analyzer. This control is performed via a clock type escapement device activated by a solenoid. Whenever the multi-channel analyzer shifts to the next channel, the associated pulse is amplified, powering the solenoid causing the drum to rotate a fixed angle. Spectra taken in laboratory, using radioactive sources, have shown a good reproducibility. This qualify the system to be used as an equipment to get a fingerprint of the overall distribution of the fuel elements along the silo axis, and hence, to verify possible diversion of the nuclear material by comparing spectra taken at consecutive safeguards inspections. All the system is battery operated, being thus capable to operate in the field where no power supply is available. (author)

Fuel management in CANDU reactors: Daniel Rozon's contribution

International Nuclear Information System (INIS)

Rozon, D.; Varin, E.; Chambon, R.

2010-01-01

The CANDU fuel management optimization problem is in many ways different from LWRs fuel management, because of the on-line refueling and the complete 3-D geometry problem. Daniel Rozon was an outstanding leader in the understanding and resolution of this optimization problem and remained during his entire career. Daniel Rozon and his students have used the generalized adjoint formalism implemented in standard mathematical programming methods to solve the optimization of the exit burnup in the reactor as well as the optimization of control rod worth or fuel enrichment. We have summarized here the theoretical basis of fuel management and resolution methods, the latest approaches of optimization and results as obtained using the OPTEX code. (author)

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

Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

Energy Technology Data Exchange (ETDEWEB)

Choi, Chang Bum; Park, Choon Hoh; Park, Chul Joo; Kwon, Woo Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO{sub 2}, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author).

Thermally-induced bowing of CANDU fuel elements

International Nuclear Information System (INIS)

Suk, H.C.; Sim, K.S.; Park, J.H.; Park, G.S.

1995-01-01

Considering only the thermally-induced bending moments which are generated both within the sheath and between the fuel and sheath by an asymmetric temperature distribution with respect to the axis of an element, a generalized and explicit analytical formula for the thermally-induced bending is developed in this paper, based on the cases of 1) the bending of an empty tube treated by neglecting of the fuel/sheath mechanical interaction and 2) the fuel/sheath interaction due to the pellet and sheath temperature variations. In each of the cases, the temperature asymmetries in sheath are modelled to be caused by the combined effects of (i) non-uniform coolant temperature due to imperfect coolant mixing, (ii) variable sheath/coolant heat transfer coefficient, (iii) asymmetric heat generation due to neutron flux gradients across an element and so as to inclusively cover the uniform temperature distributions within the fuel and sheath with respect to the axial centerline. Investigating the relative importance of the various parameters affecting fuel element bowing, the element bowing is found to be greatly affected with the variations of element length, sheath diameter, pellet/sheath mechanical interaction and neutron flux depression factors, pellet thermal expansion coefficient, pellet/sheath heat transfer coefficient in comparison with those of other parameters such as sheath thickness, film heat transfer coefficient, sheath thermal expansion coefficient, and sheath and pellet thermal conductivities. Also, the element bowing of the standard 37-element bundle and CANFLEX 43-element bundle for the use in CANDU-6 reactors was analyzed with the formula, which could help to demonstrate the integrity of the fuel. All the required input data for the analyses were generated in terms of the reactor operation conditions on the reactor physics, thermal hydraulics and fuel performance by using various CANDU computer codes. The analysis results indicate that the CANFLEX 43-element

CANDU fuel performance and development

International Nuclear Information System (INIS)

Hardy, D.G.; Wood, J.C.; Bain, A.S.

1978-12-01

The fuel defect rate in CANDU (Canada Deuterium Uranium) reactors continues to be very low, 0.06% since 1972. The power ramp defects, which constituted the majority of the early defects, have been virtually eliminated by changed fuelling schemes and through the introduction of graphite CANLUB coatings on the inside of the sheath. Laboratory and loop irradiations have demonstrated that the graphite CANLUB layers increase the tolerance to power ramps, but to obtain the maximum benefit, coating parameters such as thickness, adhesion and wear resistance must be optimized. Siloxane CANLUB coated fuel offers greater tolerance to power ramps than most graphite coatings; quality control appears simpler and no instance of localized sheath hydriding has been seen with cured and irradiated coatings. Limited testing has shown that fuel with graphite discs between fuel pellets also has high tolerance to power ramps, but it is more costly and has lower burnup. The number of defects due to faulty components has been extremely small (0.00014%), but improved quality control and welding procedures can lower this number even further. Defects from causes external to the bundle have also been very few. (author)

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

Indian Academy of Sciences (India)

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the ...

Localization of a Robotic Crawler for CANDU Fuel Channel Inspection

Science.gov (United States)

Manning, Mark

This thesis discusses the design and development of a pipe crawling robot for the purpose of CANDU fuel channel inspection. The pipe crawling robot shall be capable of deploying the existing CIGAR (Channel Inspection and Gauging Apparatus for Reactors) sensor head. The main focus of this thesis is the design of the localization system for this robot and the many tests that were completed to demonstrate its accuracy. The proposed localization system consists of three redundant resolver wheels mounted to the robot's frame and two resolvers that are mounted inside a custom made cable drum. This cable drum shall be referred to in this thesis as the emergency retrieval device. This device serves the dual-purpose of providing absolute position measurements (via the cable that is tethered to the robot) as well as retrieving the robot if it is inoperable. The estimated accuracy of the proposed design is demonstrated with the use of a proof-of-concept prototype and a custom made test bench that uses a vision system to provide a more accurate estimate of the robot's position. The only major difference between the proof-of-concept prototype and the proposed solution is that the more expensive radiation hardened components were not used in the proof-of-concept prototype design. For example, the proposed solution shall use radiation hardened resolver wheels, whereas the proof-of-concept prototype used encoder wheels. These encoder wheels provide the same specified accuracy as the radiation hardened resolvers for the most realistic results possible. The rationale behind the design of the proof-of-concept prototype, the proposed final design, the design of the localization system test bench, and the test plan for developing all of the components of the design related to the robot's localization system are discussed in the thesis. The test plan provides a step by step guide to the configuration and optimization of an Unscented Kalman Filter (UKF). The UKF was selected as the ideal

Fuel cycle model and the cost of a recycling thorium in the CANDU reactor

International Nuclear Information System (INIS)

Choi, Hangbok; Park, Chang Je

2005-01-01

The dry process fuel technology has a high proliferation-resistance, which allows applications not only to the existing but also to the future nuclear fuel cycle systems. In this study, the homogeneous ThO 2 -UO 2 recycling fuel cycle in a Canada deuterium uranium (CANDU) reactor was assessed for a fuel cycle cost evaluation. A series of parametric calculations were performed for the uranium fraction, enrichment of the initial uranium fuel, and the fission product removal rated of the recycled fuel. The fuel cycle cost was estimated by the levelized lifetime cost model provided by the Organization for Economic Cooperation and Development/Nuclear Energy Agency. Though it is feasible to recycle the homogeneous ThO 2 -UO 2 fuel in the CANDU reactor from the viewpoint of a mass balance, the recycling fuel cycle cost is much higher than the conventional natural uranium fuel cycle cost for most cases due to the high fuel fabrication cost. (author)

Analyses of fluid flow and heat transfer inside calandria vessel of CANDU-6 reactor using CFD

International Nuclear Information System (INIS)

Yu, Seon Oh; Kim, Man Woong; Kim, Hho Jung

2005-01-01

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a Loss Of Coolant Accident (LOCA) with coincident Loss Of Emergency Core Cooling (LOECC). as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines

Proceedings of the 1. international conference on CANDU fuel handling systems

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

Besides information on fuel loading and handling systems for CANDU and PHWR reactors, the 25 papers in these proceedings also include some on dry storage, modification to fuel strings at Bruce A, and on the SLAR (spacer location and repositioning) system for finding and moving garter springs. The individual papers have been abstracted separately.

Proceedings of the 1. international conference on CANDU fuel handling systems

International Nuclear Information System (INIS)

1996-01-01

Besides information on fuel loading and handling systems for CANDU and PHWR reactors, the 25 papers in these proceedings also include some on dry storage, modification to fuel strings at Bruce A, and on the SLAR (spacer location and repositioning) system for finding and moving garter springs. The individual papers have been abstracted separately

Assessment of Fuel Analysis Methodology and Fission Product Release for 37-Element Fuel by Using the Latest IST Codes during Stagnation Feeder Break in CANDU

International Nuclear Information System (INIS)

Park, Joo Hwan; Jung, Jong Yeob

2009-09-01

Feeder break accident is regarded as one of the design basis accident in CANDU reactor which results in a fuel failure. For a particular range of inlet feeder break sizes, the flow in the channel is reduced sufficiently that the fuel and fuel channel integrity can be significantly affected to have damage in the affected channel, while the remainder of the core remains adequately cooled. The flow in the downstream channel can be more or less stagnated due to a balance between pressure at the break on the upstream side and the reverse driving pressure between the break and the downstream end. In the extreme, this can lead to rapid fuel heatup and fuel damage and failure of the fuel channel similar to that associated with a severe channel flow blockage. Such an inlet feeder break scenario is called a stagnation break. In this report, the fuel analysis methodology and the assessment results of fission product inventory and release during the stagnation feeder break are described for conservatively assumed limiting channel. The accident was assumed to be occurred in the refurbished Wolsong unit 1 and the latest safety codes were used in the analysis. Fission product inventories during the steady state were calculated by using ELESTRES-IST 1.2 code. The whole analysis process was carried out by a script file which was programmed by Perl language. The perl script file was programmed to make all ELESTRES input files for each bundle and each ring based on the given power-burnup history and thermal-hydraulic conditions of the limiting channel and to perform the fuel analysis automatically. The fission product release during the transient period of stagnation feeder break was evaluated by applying Gehl model. The amounts of each isotope's release are conservatively evaluated for additional 2 seconds after channel failure. The calculated fission product releases are provided to the following dose assessment as a source term

Proceedings of the third international conference on CANDU maintenance

International Nuclear Information System (INIS)

1995-01-01

The third international conference on Candu maintenance included sessions on the following topics: predictive maintenance, reliability improvements, steam generator monitoring, tools and instrumentation, valve performance, fuel channel inspection and maintenance, steam generator maintenance, environmental qualification, predictive maintenance, instrumentation and control, steam generator cleaning, decontamination and radiation protection, inspection techniques, maintenance program strategies and valve packing experience, remote tooling/ robotics and fuel handling. The individual papers have been abstracted separately

Proceedings of the third international conference on CANDU maintenance

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-31

The third international conference on Candu maintenance included sessions on the following topics: predictive maintenance, reliability improvements, steam generator monitoring, tools and instrumentation, valve performance, fuel channel inspection and maintenance, steam generator maintenance, environmental qualification, predictive maintenance, instrumentation and control, steam generator cleaning, decontamination and radiation protection, inspection techniques, maintenance program strategies and valve packing experience, remote tooling/ robotics and fuel handling. The individual papers have been abstracted separately.

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

Addressing severe accidents in the CANDU 9 design

International Nuclear Information System (INIS)

Nijhawan, S.M.; Wight, A.L.; Snell, V.G.

1998-01-01

CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

A review of the potential for actinide redistribution in CANDU thorium cycle fuels

International Nuclear Information System (INIS)

Cameron, D.J.

1978-02-01

Actinide redistribution resulting from large radial temperature gradients is an accepted feature of the technology of fast reactor (U,Pu)O 2 fuels. A thorium cycle in CANDU reactors would require the use of oxide fuels with two or more components, raising the question of actinide redistribution in these fuels. The mechanisms proposed to explain redistribution in (U,Pu)O 2 fuels are reviewed, and their relevance to fuels based on ThO 2 is discussed. The fuel primarily considered is (Th,U)O 2 but some reference is made to (Th,Pu)O 2 . At this early stage of thorium fuel cycle technology, it is not possible to consider quantitatively the question of redistribution in specific fuels. However, some guidelines can be presented to indicate to fuel engineers conditions which might result in significant redistribution. It is concluded that redistribution is probably of little concern in high density, CANDU, thorium cycle fuel whose centre temperature is limited to 2350 K. If this centre temperature is exceeded, or if the fuel contains substantial inter-connected porosity, the potential for redistribution is significant and warrants more serious study. (author)

Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

International Nuclear Information System (INIS)

Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

2005-10-01

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor

Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

2005-10-15

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

Enthalpy and void distributions in subchannels of PHWR fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Park, J W; Choi, H; Rhee, B W [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

Enthalpy and void distributions in subchannels of PHWR fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Park, J. W.; Choi, H.; Rhee, B. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

Assessment of CANDU-6 reactivity devices for DUPIC fuel

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok

1998-11-01

Reactivity device characteristics for a CANDU 6 reactor loaded with DUPIC fuel have been assessed. The lattice parameters were generated by WIMS-AECL code and the core calculations were performed by RFSP code with a 3-dimensional full core model. The reactivity devices studied are the zone controller, adjusters, mechanical control absorber and shutoff rods. For the zone controller system, damping capability for spatial oscillation was investigated. For the adjusters, the restart capability was investigated. For the adjusters, the restart capability was investigated. The shin operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster system. The mechanical control absorber was assessed for the function of compensating temperature reactivity feedback following a power reduction. And shutoff rods were also assessed to investigate the following a power reduction. And shutoff rods were also assessed to investigate the static reactivity worth. This study has shown that the current reactivity device system of CANDU-6 core with the DUPIC fuel. (author). 9 refs., 17 tabs., 7 figs

Extending the world's uranium resources through advanced CANDU fuel cycles

Energy Technology Data Exchange (ETDEWEB)

De Vuono, Tony; Yee, Frank; Aleyaseen, Val; Kuran, Sermet; Cottrell, Catherine

2010-09-15

The growing demand for nuclear power will encourage many countries to undertake initiatives to ensure a self-reliant fuel source supply. Uranium is currently the only fuel utilized in nuclear reactors. There are increasing concerns that primary uranium sources will not be enough to meet future needs. AECL has developed a fuel cycle vision that incorporates other sources of advanced fuels to be adaptable to its CANDU technology.

Thermal analysis model for the temperature distribution of the CANDU spent fuel assembly

International Nuclear Information System (INIS)

Choi, Hae Yun; Kwon, Jong Soo; Park, Seong Hoon; Kim, Seong Rea; Lee, Gi Won

1996-01-01

The purpose of this technical is to introduce the methodology and experimental process for the experimental research work with the mock-up test performed to verify and validate the MAXROT code which is a thermal analysis method for Wolsong (CANDU) spent fuel dry storage canister. The experiment was conducted simulating the heat transfer characteristics of combinations of equilateral triangular and square pitch arrays of heater rods, similar to a CANDU spent fuel bundle. After assembly of the heater rod bundle into the containment vessel, the experimental apparatus was operated under the same operating and boundary conditions as an interim dry storage condition at the nuclear power plant site. The reduced data from this experiment has been utilized to verity a model developed to predict the maximum fuel rod surface temperature in a fuel bundle. These test procedures and the experiment can be utilized to establish the fine thermal analysis method applicable to dry storage system for the spent fuel. 12 figs., 5 tabs., 36 refs. (Author) .new

Thermal analysis model for the temperature distribution of the CANDU spent fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Choi, Hae Yun; Kwon, Jong Soo; Park, Seong Hoon; Kim, Seong Rea; Lee, Gi Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-01-01

The purpose of this technical is to introduce the methodology and experimental process for the experimental research work with the mock-up test performed to verify and validate the MAXROT code which is a thermal analysis method for Wolsong (CANDU) spent fuel dry storage canister. The experiment was conducted simulating the heat transfer characteristics of combinations of equilateral triangular and square pitch arrays of heater rods, similar to a CANDU spent fuel bundle. After assembly of the heater rod bundle into the containment vessel, the experimental apparatus was operated under the same operating and boundary conditions as an interim dry storage condition at the nuclear power plant site. The reduced data from this experiment has been utilized to verity a model developed to predict the maximum fuel rod surface temperature in a fuel bundle. These test procedures and the experiment can be utilized to establish the fine thermal analysis method applicable to dry storage system for the spent fuel. 12 figs., 5 tabs., 36 refs. (Author) .new.

CANDU 9 fuelling machine carriage

Energy Technology Data Exchange (ETDEWEB)

Ullrich, D J; Slavik, J F [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

1997-12-31

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

CANDU 9 fuelling machine carriage

International Nuclear Information System (INIS)

Ullrich, D.J.; Slavik, J.F.

1996-01-01

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

Evaluation of fuel performance for fresh and aged CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

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

2012-10-15

Like all other industrial plants, nuclear power plants also undergo degradations, so called ageing, with their operation time. Accordingly, in the recent safety analysis for a refurbished Wolsong 1 NPP, various ageing effects were incorporated into the hydraulic models of a number of the components in the primary heat transport system for conservatism. The ageing data of thermal-hydraulic components for 11 EFPY of Wolsong 1 were derived by using NUCIRC code based on the site operation data and they were modified to the appropriate input data for CATHENA code which is a thermal hydraulic code for a postulated accident analysis. This paper deals with the ageing effect of the PHTS (primary heat transport system) of CANDU reactor on the fuel performance during the normal operation. Initial conditions for fuel performance analysis were derived from the thermal-hydraulic analysis for both fresh and aged core models. Here, fresh core means a core state just right after the refurbishment and the aged core is 11 EFPY state after the refurbishment of Wolsong 1. The fuel performance was analyzed by using ELESTRES code for both fresh and aged core state and the results were compared in order to verify the ageing effect of CANDU HTS on the fuel performance.

Measurement of gap and grain-boundary inventories of 129I in used CANDU fuels

International Nuclear Information System (INIS)

Stroes-Gascoyne, S.; Moir, D.L.; Kolar, M.; Porth, R.J.; McConnell, J.L.; Kerr, A.H.

1995-01-01

Combined gap and grain-boundary inventories of 129 I in 14 used CANDU fuel elements were measured by crushing and simultaneously leaching fuel segments for 4 h in a solution containing KI carrier. From analogy with previous work a near one-to-one correlation was anticipated between the amount of stable Xe and the amount of 128 I in the combined gap and grain-boundary regions of the fuel. However, the results showed that such a correlation was only apparent for low linear power rating (LLPR) fuels with an average linear power rating of 44 kW/m), the 129 I values were considerably smaller than expected. The combined gap and grain-boundary inventories of 129 I in the 14 fuels tested varied from 1.8 to 11.0%, with an average value of 3.6 ± 2.4% which suggests that the average value of 8.1 ± 1% used in safety assessment calculations overestimates the instant release fraction for 129 I. Segments of used CANDU fuels were leached for 92 d (samples taken at 5, 28 and 92 d) to determine the kinetics of 129 I release. Results could be fitted tentatively to half-order reaction kinetics, implying that 129 I release is a diffusion-controlled process for LLPR fuels, and also for HLPR fuels, once the gap inventory has been leached. However, more data are needed over longer leaching periods to gain more understanding of the processes that control grain-boundary release of 129 I from used CANDU fuel

The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

International Nuclear Information System (INIS)

Jeon, Yu Mi; Park, Joo Hwan

2010-09-01

A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect numerical computations on the detailed shape of rod bundle due to challenges with computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers and bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve complex geometry such as a fuel rod bundle. Before applying a solution to the problem of the 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to simple geometry. The split channel method has been proposed with the aim of computing the fully shaped CANDU fuel channel with detailed components. The validity was tested by applying the method to the single channel problem. The average temperature have similar values for the considered two methods, while the local temperature shows a slight difference by the effect of conduction heat transfer in the solid region of a rod. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for future work

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)

Use of dwell time concept in fission product inventory assessment for CANDU reactors

International Nuclear Information System (INIS)

Bae, C.J.; Choi, J.H.; Hwang, H.R.; Seo, J.T.

2003-01-01

A realistic approach in calculating the initial fission product inventory within the CANFLEX-NU fuel has been assessed for its applicability to the single channel event safety analysis for CANDU reactors. This approach is based on the dwell time concept in which the accident is assumed to occur at the dwell time when the summation of fission product inventory for all isotopes becomes largest. However, in the current conservative analysis, the maximum total inventory and the corresponding gap inventory for each isotope are used as the initial fission product inventories regardless of the accident initiation time. The fission product inventory analysis has been performed using ELESTRES code considering power histories and burnup of the fuel bundles in the limiting channel. The analysis results showed that the total fission product inventory is found to be largest at 20% dwell time. Therefore, the fission product inventory at 20% dwell time can be used as the initial condition for the single channel event for the CANDU 6 reactors. (author)

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

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

Validation of the ORIGEN-S code for predicting radionuclide inventories in used CANDU fuel

International Nuclear Information System (INIS)

Tait, J.C.; Gauld, I.; Kerr, A.H.

1995-01-01

The safety assessment being conducted by AECL Research for the concept of deep geological disposal of used CANDU UO 2 fuel requires the calculation of radionuclide inventories in the fuel to provide source terms for radionuclide release. This report discusses the validation of selected actinide and fission-product inventories calculated using the ORIGEN-S code coupled with the WIMS-AECL lattice code, using data from analytical measurements of radioisotope inventories in Pickering CANDU reactor fuel. The recent processing of new ENDF/B-VI cross-section data has allowed the ORIGEN-S calculations to be performed using the most up-to-date nuclear data available. The results indicate that the code is reliably predicting actinide and the majority of fission-product inventories to within the analytical uncertainty. ((orig.))

Tests on CANDU fuel elements sheath samples

International Nuclear Information System (INIS)

Ionescu, S.; Uta, O.; Mincu, M.; Prisecaru, I.

2016-01-01

This work is a study of the behavior of CANDU fuel elements after irradiation. The tests are made on ring samples taken from fuel cladding in INR Pitesti. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory. By metallographic and ceramographic examination we determinate that the hydride precipitates are orientated parallel to the cladding surface. A content of hydrogen of about 120 ppm was estimated. After the preliminary tests, ring samples were cut from the fuel rod, and were subject of tensile test on an INSTRON 5569 model machine in order to evaluate the changes of their mechanical properties as consequence of irradiation. Scanning electron microscopy was performed on a microscope model TESCAN MIRA II LMU CS with Schottky FE emitter and variable pressure. The analysis shows that the central zone has deeper dimples, whereas on the outer zone, the dimples are tilted and smaller. (authors)

The evolution of the CANDU energy system - ready for Europe's energy future

International Nuclear Information System (INIS)

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

2001-01-01

As air quality and climate change issues receive increasing attention, the opportunity for nuclear to play a larger role in the coming decades also increases. The good performance of the current fleet of nuclear plants is crucial evidence of nuclear's potential. The excellent record of Cernavoda-1 is an important part of this, and demonstrates the maturity of the Romanian program and of the CANDU design approach. However, the emerging energy market also presents a stringent economic challenge. Current NPP designs, while established as reliable electricity producers, are seen as limited by high capital costs. In some cases, the response to the economic challenge is to consider radical changes to new design concepts, with attendant development risks from lack of provenness. Because of the flexibility of the CANDU system, it is possible to significantly extend the mid-size CANDU design, creating a Next Generation product, without sacrificing the extensive design, delivery and operations information base for CANDU. This enables a design with superior safety characteristics while at the same time meeting the economic challenge of emerging markets. The Romanian nuclear program has progressed successfully forward, leading to the successful operation of Cernavoda-1, and the project to bring Cernavoda-2 to commercial operation. The Romanian nuclear industry has become a full-fledged member of the CANDU community, with all areas of nuclear technology well established and benefiting from international cooperation with other CANDU organizations. AECL is an active partner with Romanian nuclear organizations, both through cooperative development programs, commercial contracts, and also through the activities of the CANDU owners' Group (COG). The Cernavoda project is part of the CANDU 6 family of nuclear power plants developed by AECL. The modular fuel channel reactor concept can be modified extensively, through a series of incremental changes, to improve economics, safety

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

3D heterogeneous transport calculations of CANDU fuel with EVENT/HELIOS

International Nuclear Information System (INIS)

Rahnema, F.; Mosher, S.; Ilas, D.; De Oliveira, C.; Eaton, M.; Stamm'ler, R.

2002-01-01

The applicability of the EVENT/HELIOS package to CANDU lattice cell analysis is studied in this paper. A 45-group cross section library is generated using the lattice depletion transport code HELIOS. This library is then used with the 3-D transport code EVENT to compute the pin fission densities and the multiplication constants for six configurations typical of a CANDU cell. The results are compared to those from MCNP with the same multigroup library. Differences of 70-150 pcm in multiplication constant and 0.08-0.95% in pin fission density are found for these cases. It is expected that refining the EVENT calculations can reduce these differences. This gives confidence in applying EVENT to transient analyses at the fuel pin level in a selected part of a CANDU core such as the limiting bundle during a loss of coolant accident (LOCA). (author)

Seismic analysis during development stage of CANDU Model 2 fueling machine design

International Nuclear Information System (INIS)

Lee, L.S.S.; Mansfield, R.A.

1989-01-01

The CANDU Model 3 is a new small reactor presently being designed. This reactor is 450 MWe, and as with current operating CANDU's, is based on a heavy water moderated and cooled system using on-power fuelling for the once-through natural uranium fuel cycle. The CANDU 3 Standard plant is designed to be adaptable to a range of world-wide site conditions, i.e. for a peak ground acceleration of 0.3 g and a wide range of soft, medium and hard foundation medium properties. Consequently, a conservatism in the design of structure and equipment is accounted by using enveloped floor response spectra generated by the soil-structure interaction analysis. Seismic qualification of the fuelling machine (F/M) and its support structure are an essential design requirement for maintaining the integrity of the reactor coolant heat transport system (HTS) pressure boundary and the service ports penetrating the containment structure during on-power fueling. This paper deals with the initial conceptual phase of design where the details of the design are in fundamental outline form only and basic mass distribution plus layout geometry is defined

Advanced fuel cycles in CANDU reactors

International Nuclear Information System (INIS)

Green, R.E.; Boczar, P.G.

1990-04-01

This paper re-examines the rationale for advanced nuclear fuel cycles in general, and for CANDU advanced fuel cycles in particular. The traditional resource-related arguments for more uranium nuclear fuel cycles are currently clouded by record-low prices for uranium. However, the total known conventional uranium resources can support projected uranium requirements for only another 50 years or so, less if a major revival of the nuclear option occurs as part of the solution to the world's environmental problems. While the extent of the uranium resource in the earth's crust and oceans is very large, uncertainty in the availability and price of uranium is the prime resource-related motivation for advanced fuel cycles. There are other important reasons for pursuing advanced fuel cycles. The three R's of the environmental movement, reduce, recycle, reuse, can be achieved in nuclear energy production through the employment of advanced fuel cycles. The adoption of more uranium-conserving fuel cycles would reduce the amount of uranium which needs to be mined, and the environmental impact of that mining. Environmental concerns over the back end of the fuel cycle can be mitigated as well. Higher fuel burnup reduces the volume of spent fuels which needs to be disposed of. The transmutation of actinides and long-lived fission products into short-lived fission products would reduce the radiological hazard of the waste from thousands to hundreds of years. Recycling of uranium and/or plutonium in spent fuel reuses valuable fissile material, leaving only true waste to be disposed of. Advanced fuel cycles have an economical benefit as well, enabling a ceiling to be put on fuel cycle costs, which are

Possibility of plutonium burning out and minor actinides transmutation in CANDU type reactor

International Nuclear Information System (INIS)

Gerasimov, A.S.; Kiselev, G.V.; Myrtsymova, L.A.

2000-01-01

The possibility of power or weapon-grade plutonium use as nuclear fuel in CANDU type reactor with simultaneous minor actinides burn-out is studied. Total thermal power is 1900 MW. The fuel lifetime makes 0.24 years, neutron flux density 10 14 neutr/cm 2 s. About 40-45 % of plutonium is incinerated during fuel lifetime. If weapon-grade plutonium is used in fuel channels instead of power one, its consumption is 40% lower. (author)

Validation of the ORIGEN-S code for predicting radionuclide inventories in used CANDU Fuel

International Nuclear Information System (INIS)

Tait, J.C.; Gauld, I.; Kerr, A.H.

1994-10-01

The safety assessment being conducted by AECL Research for the concept of deep geological disposal of used CANDU UO 2 fuel requires the calculation of radionuclide inventories in the fuel to provide source terms for radionuclide release. This report discusses the validation of selected actinide and fission-product inventories calculated using the ORIGEN-S code coupled with the WIMS-AECL lattice code, using data from analytical measurements of radioisotope inventories in Pickering CANDU reactor fuel. The recent processing of new ENDF/B-VI cross-section data has allowed the ORIGEN-S calculations to be performed using the most up-to-date nuclear data available. The results indicate that the code is reliably predicting actinide and the majority of fission-product inventories to within the analytical uncertainty. 38 refs., 4 figs., 5 tabs

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

Quality control of CANDU6 fuel element in fabrication process

International Nuclear Information System (INIS)

Li Yinxie; Zhang Jie

2012-01-01

To enhance the fine control over all aspects of the production process, improve product quality, fuel element fabrication process for CANDU6 quality process control activities carried out by professional technical and management technology combined mode, the quality of the fuel elements formed around CANDU6 weak links - - end plug , and brazing processes and procedures associated with this aspect of strict control, in improving staff quality consciousness, strengthening equipment maintenance, improved tooling, fixtures, optimization process test, strengthen supervision, fine inspection operations, timely delivery carry out aspects of the quality of information and concerns the production environment, etc., to find the problem from the improvement of product quality and factors affecting the source, and resolved to form the active control, comprehensive and systematic analysis of the problem of the quality management concepts, effectively reducing the end plug weld microstructure after the failure times and number of defects zirconium alloys brazed, improved product quality, and created economic benefits expressly provided, while staff quality consciousness and attention to detail, collaboration department, communication has been greatly improved and achieved very good management effectiveness. (authors)

Data base for a CANDU-PHW operating on a once-through, slightly enriched uranium cycle (AECL-6594)

International Nuclear Information System (INIS)

1979-07-01

This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on a once-through fuel cycle with a feed fuel of slightly enriched uranium - 1.2 weight % U-235 in uranium. The effects of varying fuel enrichment, maximum channel power, and economic parameters are also discussed

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)

Beryllium brazing considerations in CANDU fuel bundle manufacture

International Nuclear Information System (INIS)

Harmsen, J.; Pant, A.; Lewis, B.J.; Thompson, W.T.

2010-01-01

'Full text:' Appendages of CANDU fuel bundle elements are currently joined to zircaloy sheaths by vacuum beryllium brazing. Ongoing environmental and workplace concerns about beryllium combined with the continuous efforts by Cameco Fuel Manufacturing in its improvement process, initiated this study to find a substitute for pure beryllium. The presentation will review the necessary functionality of brazing alloy components and short list a series of alloys with the potential to duplicate the performance of pure beryllium. Modifications to current manufacturing processes based on in-plant testing will be discussed in relation to the use of these alloys. The presentation will conclude with a summary of the progress to date and further testing expected to be necessary.

Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

International Nuclear Information System (INIS)

Park, Donghwan; Kim, Youngae; Kim, Sungmin

2016-01-01

CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future

Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

Energy Technology Data Exchange (ETDEWEB)

Park, Donghwan; Kim, Youngae; Kim, Sungmin [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2016-10-15

CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future.

Accelerated reduction of used CANDU fuel waste with fast-neutron reactors: fuel cycle strategy cuts TRU waste lifespan from 400,000 years to less than 80 years

International Nuclear Information System (INIS)

Ottensmeyer, P.

2013-01-01

Canada's 45,000 tonnes of nuclear fuel waste contain over 99% heavy atoms whose nuclear energy can provide $50 trillion of non-carbon electricity in fast-neutron reactors (FNRs), equivalent to 4000 years of nuclear power at present levels. FNRs can utilize the 98.9% uranium in CANDU fuel waste and also the 0.38% transuranic actinides which impose its 400,000-year radiotoxicity. Separation of uranium from CANDU nuclear fuel waste would permit refueling of FNRs primarily with transuranics, hugely accelerating the elimination of the long-term radiotoxicity of the CANDU fuel waste. Practicable separations of uranium would result in the complete elimination of the transuranics in about 80 years using FNRs at current Canadian nuclear energy output, while ideal separations could lower this to 16 years. (author)

Fuelling study of CANDU reactors using neutron absorber poisoned fuel

Energy Technology Data Exchange (ETDEWEB)

Song, J.J.; Chan, P.K.; Bonin, H.W., E-mail: s25815@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

2014-07-01

A comparative fuelling study is conducted to determine the potential gain in operating margin for CANDU reactors incurred by implementing a change to the design of the conventional 37-element natural uranium (NU) fuel. The change involves insertion of minute quantities of neutron absorbers, Gd{sub 2}O{sub 3} and Eu{sub 2}O{sub 3}, into the fuel pellets. The Reactor Fuelling Simulation Program (RFSP) is used to conduct core-following simulations, for the regular 37-element NU fuel, which is to be used as control for comparison. Preliminary results are presented for fuelling with the regular 37-element NU fuel, which indicate constraints on fuelling that may be relaxed with addition of neutron absorbers. (author)

Proceedings of the fourth international conference on CANDU maintenance

International Nuclear Information System (INIS)

1997-01-01

These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance

Proceedings of the fourth international conference on CANDU maintenance

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-07-01

These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance.

Testing a CANDU-fueling machine at the Institute for Nuclear Research Pitesti

International Nuclear Information System (INIS)

Cojocaru, Virgil

2006-01-01

In 2003, as a national and European premiere, the Fueling Machine Head no. 4 (F/M) for the Nuclear Power Plant Cernavoda Unit 2 (NPP) was successfully tested at the Institute for Nuclear Research Pitesti (INR). In 2005, the second Fueling Machine (no. 5) has tested for the Nuclear Power Plant Cernavoda Unit 2. The Institute's main objective is to develop scientific and technological support for the Romanian Nuclear Power Program. Testing the Fueling Machines at INR Pitesti is part of the overall program to assimilate the CANDU technology in Romania. To perform the tests of these machines at INR Pitesti, a special testing rig has built being available for this goal. Both the testing rig and staff had successfully assessed by the AECL representatives during two missions. There was a delivery contract between GEC Canada and Nuclear Power Plant Cernavoda - Unit 2 to provide the Fueling Machines no. 4 and no. 5 in Romania before testing activity. As a first conclusion, the Institute for Nuclear Research Pitesti has the facilities, the staff and the experience to perform possible co-operations with any CANDU Reactor owner

Economic potential of advanced fuel cycles in CANDU

International Nuclear Information System (INIS)

Slater, J.B.

1982-07-01

Advanced fuel cycles in CANDU offer the potential of a many-fold increase in energy yield over that which can be obtained from uranium resources using the current once-through natural uranium cycle. This paper examines the associated economics of alternative once-through and recycle fuelling. Results indicate that these cycles will limit the impact of higher uranium prices and offer the potential of a period of stable constant-dollar generating costs that are only approximately 20% higher than current levels

Remote handling equipment for CANDU retubing

International Nuclear Information System (INIS)

Crawford, G.S.; Lowe, H.

1993-01-01

Numet Engineering Ltd. has designed and supplied remote handling equipment for Ontario Hydro's retubing operation of its CANDU reactors at the Bruce Nuclear Generating Station. This equipment consists of ''Retubing Tool Carriers'' an'' Worktables'' which operate remotely or manually at the reactor face. Together they function to transport tooling to and from the reactor face, to position and support tooling during retubing operations, and to deliver and retrieve fuel channels and channel components. This paper presents the fundamentals of the process and discusses the equipment supplied in terms of its design, manufacturing, components and controls, to meet the functional and quality requirements of Ontario Hydro's retubing process. (author)

Gothic simulation of single-channel fuel heatup following a loss of forced flow

Energy Technology Data Exchange (ETDEWEB)

Chen, X-Q; Tahir, A. [NSS, Dept. of Thermal Hydraulics Analysis, Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, NSATD, Pickering, Ontario (Canada); Kwee, M. [Bruce Power, NSASD, Toronto, Ontario (Canada)

2011-07-01

GOTHIC v7.2 was used to develop a computer model for the simulation of 28- and 37-element fuel heat-up at a loss of forced flow. The model has accounted for the non-uniformity of both axial and radial power distributions along the fuel channel for a typical CANDU reactor. In addition, the model has also accounted for the fuel rods, end-fittings, feeders and headers. Experimental test conditions for both 28- and 37-element bundles at either low or high powers were used for model validation. GOTHIC predictions of the rod and/or pressure-tube temperatures at a variety of test locations were compared with the corresponding experimental measurements. It is found that the numerical results agree well with the experimental measurements for most of the test locations. Results have also shown that the channel venting time is sensitive to the initial temperature distribution in the feeders and headers. An imposed temperature asymmetry at the beginning will cause the channel flow to vent earlier. (author)

CANDU fuel sheath integrity and oxide layer thickness determination by Eddy current technique

International Nuclear Information System (INIS)

Gheorghe, Gabriela; Man, Ion; Parvan, Marcel; Valeca, Serban

2010-01-01

This paper presents results concerning the integrity assessment of the fuel elements cladding and measurements of the oxide layer on sheaths, using the eddy current technique. Flaw detection using eddy current provides information about the integrity of fuel element sheath or presence of defects in the sheath produced by irradiation. The control equipment consists of a flaw detector with eddy currents, operable in the frequency range 10 Hz to 10 MHz, and a differential probe. The calibration of the flaw detector is done using artificial defects (longitudinal, transversal, external and internal notches, bored and unbored holes) obtained on Zircaloy-4 tubes identical to those out of which the sheath of the CANDU fuel element is manufactured (having a diameter of 13.08 mm and a wall thickness of 0.4 mm). When analyzing the behavior of the fuel elements' cladding facing the corrosion is important to know the thickness of the zirconium oxide layer. The calibration of the device measuring the thickness of the oxide layer is done using a Zircaloy-4 tube identical to that which the cladding of the CANDU fuel element is manufactured of, and calibration foils, as well. (authors)

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

The Advanced CANDU Reactor TM (ACR TM ) retains the proven strengths and features of CANDU reactors, and incorporates innovative new features and state-of-the-art technology. In addition to the enhanced emergency core cooling system, the reserve water system is designed to be available to inject reserve water by gravity into the reactor inlet headers after a postulated loss-of-coolant accident (LOCA). To assist in the ACR design and analysis of beyond the design basis events, simulations are needed to demonstrate the effectiveness of these two independent systems on core cooling, and to assess the consequences of the postulated accident coincident with the impairment of either of the two systems. The current paper is subject to an assessment of a postulated large LOCA coincident with loss of the emergency core cooling (LOECC) system. A postulated LOCA/LOECC has very low probability, in the range usually associated with severe core damage events. However, in the CANDU design, including ACR, the presence of moderator water surrounding the fuel channels acts as an effective heat sink, together with other safety features, to prevents severe core damage following a postulated LOCA/LOECC. Therefore, it is possible to analyse LOCA/LOECC using the same deterministic tools that are used for analysis of events with much higher frequencies, in the design basis event range. The assessment is conducted based on the current ACR-700 design. However, the analysis methodology, scope, computer tools, and the results in principle, are applicable to larger ACR designs. This assessment includes system (circuit), fuel channel, and fuel analyses. Some assessment results are needed in subsequent moderator analysis and containment analysis. In the assessment, several simulations were performed to analyse the full circuit and individual fuel channel transient behaviours, as well as the fission product release behaviour. The assessment has captured the key responses of the reactor heat

GENOVA: a generalized perturbation theory program for various applications to CANDU core physics analysis (I)-theory and application

International Nuclear Information System (INIS)

Kim, Do Heon; Choi, Hang Bok

2001-01-01

A generalized perturbation theory (GPT) program, GENOVA, has been developed for the purpose of various applications to Canadian deuterium uranium (CANDU) reactor physics analyses. GENOVA was written under the framework of CANDU physics design and analysis code, RFSP. A sensitivity method based on the GPT was implemented in GENOVA to estimate various sensitivity coefficients related to the movement of zone controller units (ZCUs) existing in the CANDU reactor. The numerical algorithm for the sensitivity method was verified by a simple 2 x 2 node problem. The capability of predicting ZCU levels upon a refueling perturbation was validated for a CANDU-6 reactor problem. The applicability of GENOVA to the CANDU-6 core physics analysis has been demonstrated with the optimum refueling simulation and the uncertainty analysis problems. For the optimum refueling simulation, an optimum channel selection strategy has been proposed, using the ZCU level predicted by GENOVA. The refueling simulation of a CANDU-6 natural uranium core has shown that the ZCU levels are successfully controlled within the operating range while the channel and bundle powers are satisfying the license limits. An uncertainty analysis has been performed for the fuel composition heterogeneity of a CANDU DUPIC core, using the sensitivity coefficients generated by GENOVA. The results have shown that the uncertainty of the core performance parameter can be reduced appreciably when the contents of the major fissile isotopes are tightly controlled. GENOVA code has been successfully explored to supplement the weak points of the current design and analysis code, such as the incapacity of performing an optimum refueling simulation and uncertainty analysis. The sample calculations have shown that GENOVA has strong potential to be used for CANDU core analysis combined with the current design and analysis code, RFSP, especially for the development of advanced CANDU fuels

Optimizing in-bay fuel inspection capability to meet the needs of today's CANDU fleet

International Nuclear Information System (INIS)

St-Pierre, J.; Simons, B.

2013-01-01

With the recent return to service of many CANDU units, aging of all others, increasingly competitive energy market and aging hot cell infrastructure - there exists now a greater need for timely, cost-effective and reliable collection of irradiated fuel performance information from fuel bay inspections. The recent development of simple in-bay tools, used in combination with standardized technical specifications, inspection databases and assessment techniques, allows utilities to characterize the condition of irradiated fuel and any debris lodged in the bundle in a more timely fashion and more economically than ever. Use of these tools and 'advanced' techniques permits timely engineering review and disposition of emerging issues to support reliable operation of the CANDU fleet. (author)

Optimizing in-bay fuel inspection capability to meet the needs of today's CANDU fleet

Energy Technology Data Exchange (ETDEWEB)

St-Pierre, J., E-mail: joe.st-pierre@amec.com [AMEC NSS, Toronto, Ontario (Canada); Simons, B. [Stern Laboratories Incorporated, Hamilton, Ontario (Canada)

2013-07-01

With the recent return to service of many CANDU units, aging of all others, increasingly competitive energy market and aging hot cell infrastructure - there exists now a greater need for timely, cost-effective and reliable collection of irradiated fuel performance information from fuel bay inspections. The recent development of simple in-bay tools, used in combination with standardized technical specifications, inspection databases and assessment techniques, allows utilities to characterize the condition of irradiated fuel and any debris lodged in the bundle in a more timely fashion and more economically than ever. Use of these tools and 'advanced' techniques permits timely engineering review and disposition of emerging issues to support reliable operation of the CANDU fleet. (author)

THERMOSS: a thermohydraulic model of flow stagnation in a horizontal fuel channel

International Nuclear Information System (INIS)

Gulshani, P.; Caplan, M.Z.; Spinks, N.J.

1984-01-01

Following a postulated inlet-side small break in the CANDU reactor, emergency coolant is injected to refull the horizontal fuel channels and remove the decay heat. As part of the accident analysis, the effects of loss of forced circulation during the accident are predicted. A break size exists for which, at the end of pump rundown, the break force balances the natural circulation force and the channel flow is reduced to near zero. The subcooled, stagnant channel condition is referred to as the standing-start condition. Subsequently, the channel coolant boils and stratifies. Eventually the steam flow from the channel heats up the endfitting to the saturation temperature and reaches the vertical feeder. The resulting buoyancy-induced flow then refills the channel. One dimensional, two-fluid conservation equations are solved in closed form to predict the duration of stagnation. In this calculation the channel water level is an important intermediate variable because it determines the amount of steam production

Fouling issues within CANDU fuel bay purification systems

International Nuclear Information System (INIS)

Angell, P.; Murray, J.

2003-01-01

Similar to all nuclear power stations, CANDU stations store their spent fuel in large open fuel bays to allow the fuel bundles to cool down. Over the years there have been a number of minor issues relating to the operation of the fuel bays, and the annual operational and maintenance costs to the stations are in the order of $1-2M. Ontario Power Generation (OPG) stations have been treating the fuel bays of their CANDU stations with hydrogen peroxide. While this treatment has been effective for the control of microbial corrosion and algal problems, an ongoing problem with the rapid plugging of the purification system filters and the presence of a thin translucent film on the surface of the water has been reported. Analysis suggested that the translucent film was of biological origin and initially it was suspected as being responsible for the plugging of the filters. Samples of the bay water, translucent film and filters were collected for microbial analysis. From all samples a number of bacterial species was isolated, all of which had the enzyme catalase that is responsible for the microbial breakdown of hydrogen peroxide, and many of the isolates were found to be capable of producing extracellular polymeric substance (slime). Observation of the filter revealed that no slime layer was present on the outer surface of the filter. This suggested that the plugging was not the result of removal of slime from the bay. Rather it suggested that the plugging was the result of microbial colonization and growth within the matrix of the filter. The large surface area of the filter matrix provides ideal conditions for biofilm growth and the flow of water provides a ready supply of nutrients. Based on a careful review of the chemistry specifications and requirements for the fuel bay water it was suggested that the filters, as currently used, were unnecessary and could be either removed or valved out, thereby eliminating the high costs associated with their operation and

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.

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

Equilibrium fuel-management simulations for 1.2% SEU in a CANDU 6

International Nuclear Information System (INIS)

Younis, M.H.; Boczar, P.G.

1989-06-01

Fuel-management simulations have been performed for 1.2% SEU in a CANDU 6 reactor at equilibrium, for three fuel-management options: axial shuffling; a regular 2-bundling shift with the adjuster rods removed from the core; and a regular 2-bundle shift with the adjuster rods present. Both time-average and time-dependent simulations were performed, from which the physics characteristics of the cores at equilibrium were estimated. Power and power-boost envelopes were derived for both 37-element fuel, and the advanced CANFLEX bundle

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)

THERMOSS: A thermohydraulic model of flow stagnation in a horizontal fuel channel

International Nuclear Information System (INIS)

Gulshani, P.; Caplan, M.Z.; Spinks, N.J.

1984-01-01

A model, called THERMOSS, is developed to compute the duration of stagnation in a CANDU reactor fuel channel with subcooled, stagnant initial conditions. The model solves, in closed form, the one dimensional, two-fluid conservation equations. In the computation of the duration of stagnation, the channel water level is an important intermediate variable because it determines the amount of steam production. A feature of the model is that water level is determined by a momentum balance between frictional pressure drop in the steam phase and hydrostatic head in the liquid phase. This is in contrast to an ealier model in which the level was determined from mass balance considerations. A satisfactory agreement between the predicted and experimentally observed channel water level and duration of stagnation is obtained. (orig.)

Use of enriched uranium as a fuel in CANDU reactors

International Nuclear Information System (INIS)

Zech, H.J.

1976-08-01

The use of slightly enriched uranium as a fuel in CANDU-reactors is studied in a simple parametric way. The results show the possibility of 1) about 30% savings in natural uranium consumption 2) about 35% increase in the utilization of the natural uranium 3) a decrease in fuelling costs to about 70 - 80% of the normal case of natural uranium fuelling. (orig.) [de

Mitigation of end flux peaking in CANDU fuel bundles using neutron absorbers

Energy Technology Data Exchange (ETDEWEB)

Pierce, D.; Chan, P.K., E-mail: dylan.pierce@rmc.ca [Royal Military College of Canada, Kingston ON, (Canada); Shen, W. [Canadian Nuclear Safety Commission, Ottawa ON, (Canada)

2015-07-01

End flux peaking (EFP) is a phenomenon where a region of elevated neutron flux occurs between two adjoining fuel bundles. These peaks lead to an increase in fission rate and therefore greater heat generation. It is known that addition of neutron absorbers into fuel bundles can help mitigate EFP, yet implementation in Canada Deuterium Uranium (CANDU) type reactors using natural uranium fuel has not been pursued. Monte Carlo N-Particle code (MCNP) 6.1 was used to simulate the addition of a small amount of neutron absorbers strategically within the fuel pellets. This paper will present some preliminary results collected thus far. (author)

Root-cause analysis of the better performance of the coarse-mesh finite-difference method for CANDU-type reactors

International Nuclear Information System (INIS)

Shen, W.

2012-01-01

Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, three benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)

Root-cause analysis of the better performance of the coarse-mesh finite-difference method for CANDU-type reactors

Energy Technology Data Exchange (ETDEWEB)

Shen, W. [Candu Energy Inc., 2285 Speakman Dr., Mississauga, ON L5B 1K (Canada)

2012-07-01

Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, three benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)

Monte Carlo assessment of the dose rates produced by spent fuel from CANDU reactors

International Nuclear Information System (INIS)

Pantazi, Doina; Mateescu, Silvia; Stanciu, Marcela

2003-01-01

One of the technical measures considered for biological protection is radiation shielding. The implementation process of a spent fuel intermediate storage system at Cernavoda NPP includes an evolution in computation methods related to shielding evaluation: from using simpler computer codes, like MicroShield and QAD, to systems of codes, like SCALE (which contains few independent modules) and the multipurpose and multi-particles transport code MCNP, based on Monte Carlo method. The Monte Carlo assessment of the dose rates produced by CANDU type spent fuel, during its handling for the intermediate storage, is the main objective of this paper. The work had two main features: -establishing of geometrical models according to description mode used in code MCNP, capable to account for the specific characteristics of CANDU nuclear fuel; - confirming the correctness of proposed models, by comparing MCNP results and the related results obtained with other computer codes for shielding evaluation and dose rates calculations. (authors)

Data base for a CANDU-PHW operating on a once-through, natural uranium fuel cycle

International Nuclear Information System (INIS)

1979-07-01

This report, prepared for INFCE, describes a standard 600 MW(e) CANDU-PHW reactor operating on a once-through natural uranium fuel cycle. Subsequently, data are given for an extrapolated 1000 MW(e) design (the nominal capacity adopted for the INFCE study) operating on the same fuel cycle. (author)

Plutonium dispositioning in CANDU

International Nuclear Information System (INIS)

Boczar, P.G.; Feinroth, H.; Luxat, J.C.

1995-07-01

Recently, the U.S. Department of Energy (DOE) sponsored Atomic Energy of Canada Limited (AECL) to evaluate salient technical, strategic, schedule, and cost-related parameters of using CANDU reactors for dispositioning of weapons-grade plutonium in the form of Mixed OXide (MOX) fuel. A study team, consisting of key staff from the CANDU reactor designers and researchers (AECL), operators (Ontario Hydro) and fuel suppliers, analyzed all significant factors involved in such application, with the objective of identifying an arrangement that would permit the burning of MOX in CANDU at the earliest date. One of Ontario Hydro's multi-unit stations, Bruce A nuclear generating station (4x769 MW(e)), was chosen as the reference for the study. The assessment showed that no significant modifications of reactor or process systems are necessary to operate with a full MOX core. Plant modifications would be limited to fuel handling and modifications necessary to accommodate enhanced security and safeguards requirements. No safety limitations were identified

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

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

International Nuclear Information System (INIS)

1998-08-01

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

Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

International Nuclear Information System (INIS)

Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

2010-01-01

In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

An experimental investigation of heat transfer from a reactor fuel channel to surrounding water

International Nuclear Information System (INIS)

Gillespie, G.E.

An important feature of the CANDU-PHW reactor is that each fuel channel is surrounded by cool heavy-water moderator that can act as a sink for heat generated in the fuel if other means of heat removal were to fail. During postulated loss-of-coolant accidents there are two scenarios in which the primary cooling system may not prevent fuel-channel overheating. These situations arise when: (1) for a particular break size and location, called the critical break, the coolant flow through a portion of the reactor core stagnates before the emergency coolant injection system restores circulation, or, (2) the emergency coolant injection system fails to operate. In either case, the heat generated in the fuel is transferred mainly by radiation to the pressure tube and calandria tube, and then by boiling heat transfer to the moderator. This paper describes a simple one-dimensional model developed to analyse the thermal behaviour of a fuel channel when the internal pressure is high. Also described is a series of experiments in which the pressure-tube segment is pressurized and heated at a constant rate until it contacts a surrounding calandria-tube segment. Predictions of the one-dimensional model are compared with the experimental results

Incentives for improvement of CANDU

International Nuclear Information System (INIS)

Hart, R.S.; Dunn, J.T.; Finlay, R.B.

1988-12-01

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

Experience in the manufacture and performance of CANDU fuel for KANUPP

International Nuclear Information System (INIS)

Salim, M.; Ahmed, I.; Butt, P.

1995-01-01

Karachi Nuclear Power Plant (KANUPP) a 137 MWe CANDU unit is In operation since 1971. Initially, it was fueled with Canadian fuel bundles. In July 1980 Pakistani manufactured fuel was introduced in the reactor core, irradiated to a burnup of about 7500 MWd-teU -1 and successfully discharged in May 1984. The core was progressively fuelled with Pakistani fuel and in August 1990 the reactor core contained all Pakistani made fuel. As of the present, 3 core equivalent Pakistani fuel bundles have been successfully discharged at an average bumup of 6500 MWd-teU -1 . with a maximum burnup of ∼ 10,200 MWd-teU -1 . No fuel failure of Pakistani bundles has been observed so far. This paper presents the indigenous efforts towards manufacture and operational aspects of KANUPP fuel and compares its behaviour with that of Canadian supplied fuel. The Pakistani fuel has performed well and is as good as the Canadian fuel. (author)

A catalogue of advanced fuel cycles in CANDU-PHW reactors

International Nuclear Information System (INIS)

Veeder, J.; Didsbury, R.

1985-06-01

A catalogue raisonne is presented of various advanced fuel cycle options which have the potential of substantially improving the uranium utilization for CANDU-PHW reactors. Three categories of cycles are: once-through cycles without recovery of fissile materials, cycles that depend on the recovery and recycle of fissile materials in thorium or uranium, cycles that depend primarily on the production of fissile material in a fertile blanket by means of an intense neutron source other than fission, such as an accelerator breeder. Detailed tables are given of the isotopic compositions of the feed and discharge fuels, the logistics of materials and processes required to sustain each of the cycles, and tables of fuel cycle costs based on a method of continuous discounting of cash flow

Conceptual designs for very high-temperature CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Bushby, S.J.; Dimmick, G.R.; Duffey, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

2000-07-01

Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310{sup o}C, and exits at {approx}410{sup o}C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed

Conceptual designs for very high-temperature CANDU reactors

International Nuclear Information System (INIS)

Bushby, S.J.; Dimmick, G.R.; Duffey, R.B.

2000-01-01

Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310 o C, and exits at ∼410 o C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed power plants

Key thrusts in next generation CANDU. Annex 10

International Nuclear Information System (INIS)

Shalaby, B.A.; Torgerson, D.F.; Duffey, R.B.

2002-01-01

Current electricity markets and the competitiveness of other generation options such as CCGT have influenced the directions of future nuclear generation. The next generation CANDU has used its key characteristics as the basis to leap frog into a new design featuring improved economics, enhanced passive safety, enhanced operability and demonstrated fuel cycle flexibility. Many enabling technologies spinning of current CANDU design features are used in the next generation design. Some of these technologies have been developed in support of existing plants and near term designs while others will need to be developed and tested. This paper will discuss the key principles driving the next generation CANDU design and the fuel cycle flexibility of the CANDU system which provide synergism with the PWR fuel cycle. (author)

Modeling CANDU type fuel behaviour during extended burnup irradiations using a revised version of the ELESIM code

International Nuclear Information System (INIS)

Arimescu, V.I.; Richmond, W.R.

1992-05-01

The high-burnup database for CANDU fuel, with a variety of cases, offers a good opportunity to check models of fuel behaviour, and to identify areas for improvement. Good agreement of calculated values of fission-gas release, and sheath hoop strain, with experimental data indicates that the global behaviour of the fuel element is adequately simulated by a computer code. Using, the ELESIM computer code, the fission-gas release, swelling, and fuel pellet expansion models were analysed, and changes made for gaseous swelling, and diffusional release of fission-gas atoms to the grain boundaries. Using this revised version of ELESIM, satisfactory agreement between measured values of fission-gas release was found for most of the high-burnup database cases. It is concluded that the revised version of the ELESIM code is able to simulate with reasonable accuracy high-burnup as well as low-burnup CANDU fuel

ITER SAFETY TASK NID-10A:CANDU occupational exposure experience: ORE for ITER fuel cycle and cooling systems

International Nuclear Information System (INIS)

Lee, D.

1995-02-01

This report contains information on TRITIUM Occupational Exposure (Internal Dose) from typical CANDU Nuclear Generating Stations. In addition to dose, airborne tritium levels are provided, as these strongly influence operational exposure. The exposure dose data presented in this report cover a period of five years of operation and maintenance experience from four CANDU Reactors and are considered representative of other CANDU reactors. The data are broken down according to occupational function ( Operators, Maintenance and Support Service etc.). The referenced systems are mainly centered on CANDU Hear Transport System, Moderator System, Tritium Removal Facility and Heavy Water (D20) Upgrading System. These systems contain the bulk part of tritium contamination in the CANDU Reactor. Because of certain similarities between ITER and CANDU systems, this data can be used as the most relevant TRITIUM OCCUPATIONAL DOSE information for ITER COOLING and FUEL CYCLE systems dose assessment purpose, if similar design and operation principles as described in the report are adopted. (author). 16 refs., 8 tabs., 13 figs

Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

International Nuclear Information System (INIS)

Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

1982-01-01

The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

Public health risks associated with the CANDU nuclear fuel cycle

International Nuclear Information System (INIS)

Paskievici, W.; Zikovsky, L.

1983-06-01

This report analyzes in a preliminary way the risks to the public posed by the CANDU nuclear fuel cycle. Part 1 considers radiological risks, while part 2 (published as INFO-0141-2) evaluates non-radiological risks. The report concludes that, for radiological risks, maximum individual risks to members of the public are less than 10 -5 per year for postulated accidents, are less than 1 percent of regulatory limits for normal operation and that collective doses are small, less than 3 person-sieverts. It is also concluded that radiological risks are much smaller than the non-radiological risks posed by activities of the nuclear fuel cycle

Reactor physics data for safety analysis of CANFLEX-NU CANDU-6 core

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Suk, Ho Chun

2001-08-01

This report contains the reactor physics data for safety analysis of CANFLEX-NU fuel CANDU-6 core. First, the physics parameters for time-average core have been described, which include the channel power and maximum bundle power map, channel axial power shape and bundle burnup. And, next the data for fuel performance such as relative ring power distribution and bundle burnup conversion ratio are represented. The transition core data from 0 to 900 full power day are represented by 100 full power day interval. Also, the data for reactivity devices of time-average core and 300 full power day of transition core are given.

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.

Moving towards sustainable thorium fuel cycles

International Nuclear Information System (INIS)

Hyland, B.; Hamilton, H.

2011-01-01

The CANDU reactor has an unsurpassed degree of fuel-cycle flexibility as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle design. These features facilitate the introduction and full exploitation of thorium fuel cycles in CANDU reactors in an evolutionary fashion. Thoria (ThO 2 ) based fuel offers both fuel performance and safety advantages over urania (UO 2 ) based fuel, due its higher thermal conductivity which results in lower fuel-operating temperatures at similar linear element powers. Thoria fuel has demonstrated lower fission gas release than UO 2 under similar operating powers during test irradiations. In addition, thoria has a higher melting point than urania and is far less reactive in hypothetical accident scenarios owing to the fact that it has only one oxidation state. This paper examines one possible strategy for the introduction of thorium fuel cycles into CANDU reactors. In the short term, the initial fissile material would be provided in a heterogeneous bundle of low-enriched uranium and thorium. The medium term scenario uses homogeneous Pu/Th bundles in the CANDU reactor, further increasing the energy derived from the thorium. In the long term, the full energy potential from thorium would be realized through the recycle of the U-233 in the used fuel. With U-233 recycle in CANDU reactors, plutonium would then only be required to top up the fissile content to achieve the desired burnup. (author)

Transport-diffusion coupling for Candu reactor core follow-Up

International Nuclear Information System (INIS)

Varin, E.; Marleau, G.; Chambon, R.

2003-01-01

We couple the finite reactor diffusion code DONJON and the lattice code DRAGON, called for simplicity DD, to perform reactor follow-up calculations using a history-based approach. In order to do this, a new DD module is developed. This module manages the transfer of information between standard DONJON and DRAGON data structures. Moreover, it stores in a history data structure the global and local parameters required for cell calculations as well as the isotopic composition of the various materials present in each cell of the reactor. We then implement in DD a parallel algorithm to perform history-based Candu reactor calculations. Here, we assign to each processor a specific number of fuel channels to be analyzed. The DRAGON cell calculations for each of the fuel bundles associated with the specified channels are performed on the same processor in order to minimize communication time. Only the macroscopic cross section libraries are exchanged between the processor. Since the amount of data exchanged is relatively small, we expect to obtain an ideal speed-up. The coupling is tested for the analysis of a simplified Candu reactor model with 4 x 4 channels each containing 4 bundles. A 100 full-power days core tracking sequence with 16 refueling steps is studied. Results are coherent with those obtained using more approximate approaches. Parallel speed-up is near optimal indicating that the use of this approach for more realistic reactor calculations should be pursued. (authors)

Comparative Analysis of Thermohydraulic Margins in Embalse Power Station, CARA Vs. CANDU with Cobra IV-HW

International Nuclear Information System (INIS)

Daverio, H; Juanico, L

2000-01-01

Comparative analysis of thermohydraulic margins were studied of the CANDU 37 and CARA fuel bundles (FB) in Embalse power station with COBRA IV-HW code ., the geometry of the bundle laying on the channel was particularly modeled and discussing the results in comparison with former calculations with 1/6 simetry .The CARA design with enriched uranium (0.9 %) and extended burn up lets maintain the current thermohydraulic nominal margins , while compared with CANDU 37 rods FB enriched , the CARA design permits widely improve the current margins

Validation of WIMS-CANDU using Pin-Cell Lattices

International Nuclear Information System (INIS)

Kim, Won Young; Min, Byung Joo; Park, Joo Hwan

2006-01-01

The WIMS-CANDU is a lattice code which has a depletion capability for the analysis of reactor physics problems related to a design and safety. The WIMS-CANDU code has been developed from the WIMSD5B, a version of the WIMS code released from the OECD/NEA data bank in 1998. The lattice code POWDERPUFS-V (PPV) has been used for the physics design and analysis of a natural uranium fuel for the CANDU reactor. However since the application of PPV is limited to a fresh fuel due to its empirical correlations, the WIMS-AECL code has been developed by AECL to substitute the PPV. Also, the WIMS-CANDU code is being developed to perform the physics analysis of the present operating CANDU reactors as a replacement of PPV. As one of the developing work of WIMS-CANDU, the U 238 absorption cross-section in the nuclear data library of WIMS-CANDU was updated and WIMS-CANDU was validated using the benchmark problems for pin-cell lattices such as TRX-1, TRX-2, Bapl-1, Bapl-2 and Bapl-3. The results by the WIMS-CANDU and the WIMS-AECL were compared with the experimental data

A model for fission product distribution in CANDU fuel

International Nuclear Information System (INIS)

Muzumdar, A.P.

1983-01-01

This paper describes a model to estimate the distribution of active fission products among the UO 2 grains, grain-boundaries, and the free void spaces in CANDU fuel elements during normal operation. This distribution is required for the calculation of the potential release of activity from failed fuel sheaths during a loss-of-coolant accident. The activity residing in the free spaces (''free'' inventory) is available for release upon sheath rupture, whereas relatively high fuel temperatures and/or thermal shock are required to release the activity in the grain boundaries or grains. A preliminary comparison of the model with the data from in-reactor sweep-gas experiments performed in Canada yields generally good agreement, with overprediction rather than under prediction of radiologically important isotopes, such as I 131 . The model also appears to generally agree with the ''free'' inventory release calculated using ANS-5.4. (author)

CANDU reactors and greenhouse gas emissions

International Nuclear Information System (INIS)

Andseta, S.; Thompson, M.J.; Jarrell, J.P.; Pendergast, D.R.

1999-01-01

This paper was originally presented at the 11th Pacific Basin Nuclear Conference, Banff, Alberta, Canada, May 3-7, 1998. It has been updated to include additional lifecycle data on chemical releases from ore treatment and CANDU fuel fabrication. It is sometimes stated that nuclear power plants can supply electricity with zero emissions of greenhouse gases. In fact, consideration of the entire fuel cycle indicates that some greenhouse gases are generated during their construction and decommissioning and by the preparation of fuel and other materials required for their operation. This follows from the use of fossil fuels in the preparation of materials and during the construction and decommissioning of the plants. This paper reviews life cycle studies of several different kinds of power plants. Greenhouse gases generated by fossil fuels during the preparation of fuel and heavy water used by operating CANDU power plants are estimated. The total greenhouse gas emissions from CANDU nuclear plants, per unit of electricity ultimately produced, are very small in comparison with emissions from most other types of power plants. (author)

CANDU reactors and greenhouse gas emissions

International Nuclear Information System (INIS)

Andseta, S.; Thompson, M.J.; Jarrell, J.P.; Pendergast, D.R.

1998-01-01

This paper was originally presented at the 11th Pacific Basin Nuclear Conference, Banff, Alberta, Canada, May 3-7, 1998. It has been updated to include additional lifecycle data on chemical releases from ore treatment and CANDU fuel fabrication. It is sometimes stated that nuclear power plants can supply electricity with zero emissions of greenhouse gases. In fact, consideration of the entire fuel cycle indicates that some greenhouse gases are generated during their construction and decommissioning and by the preparation of fuel and other materials required for their operation. This follows from the use of fossil fuels in the preparation of materials and during the construction and decommissioning of the plants. This paper reviews life cycle studies of several different kinds of power plants. Greenhouse gases generated by fossil fuels during the preparation of fuel and heavy water used by operating CANDU power plants are estimated. The total greenhouse gas emissions from CANDU nuclear plants, per unit of electricity ultimately produced, are very small in comparison with emissions from most other types of power plants. (author)

Five years of successful CANDU-6 fuel manufacturing in Romania

International Nuclear Information System (INIS)

Galeriu, A.C.; Pascu, A.; Andrei, G.; Bailescu, A.

1999-01-01

This paper describes the evolution of CANDU-6 nuclear fuel manufacturing in Romania at FCN Pitesti, after the completion of the qualification in 1994. Commercial production was resumed early 1995 and fuel bundles produced were entirely delivered to Cernavoda Plant and charged in the reactor. More than 12,000 fuel bundles have been produced in the last five years and the fuel behaved very well. Defective bundles represents less than 0.06% from the total irradiated fuel, and the most defects are associated to the highest power positions. After qualification, FCN focused the effort to improve braze quality and also to maintain a low residual hydrogen content in graphite coated sheaths. The production capacity was increased especially for component manufacturing, appendages tack welding and brazing. A new graphite baking furnace with increased capacity, is under design. In the pelleting area, a rotating press will replace the older hydraulic presses used for pelleting. Plant development taken inter consideration the future demands for Cernavoda Unit 2. (author)

Demonstrating the compatibility of Canflex fuel bundles with a CANDU 6 fuelling machine

Energy Technology Data Exchange (ETDEWEB)

Alavi, P; Oldaker, I E [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Suk, H C; Choi, C B [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

1997-12-31

CANFLEX is a new 43-element fuel bundle, designed for high operating margins. It has many small-diameter elements in its two outer rings, and large-diameter elements in its centre rings. By this means, the linear heat ratings are lower than those of standard 37-element bundles for similar power outputs. A necessary part of the out-reactor qualification program for the CANFLEX fuel bundle design, is a demonstration of the bundle`s compatibility with the mechanical components in a CANDU 6 Fuelling Machine (FM) under typical conditions of pressure, flow and temperature. The diameter of the CANFLEX bundle is the same as that of a 37-element bundle, but the smaller-diameter elements in the outer ring result in a slightly larger end-plate diameter. Therefore, to minimize any risk of unanticipated damage to the CANDU 6 FM sidestops, a series of measurements and static laboratory tests were undertaken prior to the fuelling machine tests. The tests and measurements showed that; a) the CANFLEX bundle end plate is compatible with the FM sidestops, b) all the dimensions of the CANFLEX fuel bundle are within the specified limits. (author). 3 tabs., 3 figs.

Safety research for CANDU reactors

International Nuclear Information System (INIS)

Hancox, W.T.

1982-10-01

Continuing research to develop and verify computer models of CANDU-PHW reactor process and safety systems is described. It is focussed on loss-of-coolant accidents (LOCAs) because they are the precursors of more serious accidents. Research topics include: (i) fluid-dynamic and heat-transfer processes in the heat transport system during the blowdown and refilling phases of LOCAs; (ii) thermal and mechanical behaviour of fuel elements; (iii) thermal and mechanical behaviour of the fuel and the fuel-channel assembly in situations where the heavy-water moderator is the sink for decay heat produced in the fuel; (iv) chemical behaviour of fission gases that might be released into the reactor coolant and transported to the containment system; and (v) combustion of hydrogen-air-steam mixtures that would be produced if fuel temperatures were sufficiently high to initiate the zirconium-water reaction. The current status of the research on each of these topics is highlighted with particular emphasis on the conclusions reached to date and their impact on the continuing program

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.

Flow visualization study of two-phase flow in the horizontal annulus of the fuel-channel outlet end-fitting of a CANDU reactor

International Nuclear Information System (INIS)

Supa-Amornkul, S.; Steward, F.R.; Lister, D.H.

2005-01-01

In CANDU-6 reactors, the pressurized hightemperature coolant flows through 380 fuel channels passing horizontally through the core. In 1996, higher than expected rates of wall thinning of the outlet feeders were ascribed to flow-accelerated corrosion (FAC). Such corrosion is strongly influenced by the hydrodynamics of the coolant. Results of preliminary flow visualization and modelling studies have suggested that flow conditions in the end-fitting annulus upstream of the outlet feeder may influence the pattern of FAC. For a full-scale flow visualization, an acrylic test section was built to simulate the cylindrical end-fitting with its annulus flow path. The tests were performed with water and air at atmospheric pressure and room temperature. The phase distribution along the length of the annulus was recorded with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Significant effects on the flow patterns of spacer buttons in the annulus were observed. A commercial computational fluid dynamics (CFD) code-Fluent 6.1-was used to model the results. (authors)

Element bow profiles from new and irradiated CANDU fuel bundles

International Nuclear Information System (INIS)

Dennier, D.; Manzer, A.M.; Ryz, M.A.

1996-01-01

Improved methods of measuring element profiles on new CANDU fuel bundles were developed at the Sheridan Park Engineering Laboratory, and have now been applied in the hot cells at Whiteshell Laboratories. For the first time, the outer element profiles have been compared between new, out-reactor tested, and irradiated fuel elements. The comparison shows that irradiated element deformation is similar to that observed on elements in out-reactor tested bundles. In addition to the restraints applied to the element via appendages, the element profile appears to be strongly influenced by gravity and the end loads applied by local deformation of the endplate. Irradiation creep in the direction of gravity also tends to be a dominant factor. (author)

An integrated CANDU system

International Nuclear Information System (INIS)

Donnelly, J.

1982-09-01

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

Plutonium Consumption Program, CANDU Reactor Project final report

Energy Technology Data Exchange (ETDEWEB)

1994-07-31

DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

Plutonium Consumption Program, CANDU Reactor Project final report

International Nuclear Information System (INIS)

1994-01-01

DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

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.

International collaboration to study the feasibility of implementing the use of slightly enriched uranium fuel in the Embalse CANDU reactor

International Nuclear Information System (INIS)

Rouben, B.; Chow, H.C.; Leung, L.K.H.; Inch, W.; Fink, J.; Moreno, C.

2004-01-01

In the last few years, Nucleoelectrica Argentina S.A. and Atomic Energy of Canada Limited have collaborated on a study of the technical feasibility of implementing Slightly Enriched Uranium (SEU) fuel in the Embalse CANDU reactor in Argentina. The successful conversion to SEU fuel of the other Argentine heavy-water reactor, Atucha 1, served as a good example. SEU presents an attractive incentive from the point of view of fuel utilization: if fuel enriched to 0.9% 235 U were used in Embalse instead of natural uranium, the average fuel discharge burnup would increase significantly (by a factor of about 2), with consequent reduction in fuel requirements, leading to lower fuel-cycle costs and a large reduction in spent-fuel volume per unit energy produced. Another advantage is the change in the axial power shape: with SEU fuel, the maximum bundle power in a channel decreases and shifts towards the coolant inlet end, consequently increasing the thermalhydraulics safety margin. Two SEU fuel carriers, the traditional 37-element bundle and the 43-element CANFLEX bundle, which has enhanced thermalhydraulic characteristics as well as lower peak linear element ratings, have been examined. The feasibility study gave the organizations an excellent opportunity to perform cooperatively a large number of analyses, e.g., in reactor physics, thermalhydraulics, fuel performance, and safety. A Draft Plan for a Demonstration Irradiation of SEU fuel in Embalse was prepared. Safety analyses have been performed for a number of hypothetical accidents, such as Large Loss of Coolant, Loss of Reactivity Control, and an off-normal condition corresponding to introducing 8 SEU bundles in a channel (instead of 2 or 4 bundles). There are concrete safety improvements which result from the reduced maximum bundle powers and their shift towards the inlet end of the fuel channel. Further improvements in safety margins would accrue with CANFLEX. In conclusion, the analyses identified no issues that

The post-irradiated examination of CANDU type fuel irradiated in the Institute for Nuclear Research TRIGA reactor

International Nuclear Information System (INIS)

Tuturici, I.L.; Parvan, M.; Dobrin, R.; Popov, M.; Radulescu, R.; Toma, V.

1995-01-01

This post-irradiation examination work has been done under the Research Contract No. 7756/RB, concluded between the International Atomic Energy Agency and the Institute for Nuclear Research. The paper contains a general description of the INR post-irradiation facility and methods and the relevant post-irradiation examination results obtained from an irradiated experimental CANDU type fuel element designed, manufactured and tested by INR in a power ramp test in the 100 kW Pressurised Water Irradiation Loop of the TRIGA 14 MW(th) Reactor. The irradiation experiment consisted in testing an assembly of six fuel elements, designed to reach a bumup of ∼ 200 MWh/kgU, with typical CANDU linear power and ramp rate. (author)

The Canadian R and D program targeted at CANDU reactors

International Nuclear Information System (INIS)

Moeck, E.O.

1988-01-01

CANDU reactors produce electricity cheaply and reliably, with miniscule risk to the population and minimal impact on the environment. About half of Ontario's electricity and a third of New Brunswick's are generated by CANDU power plants. Hydro Quebec and utilities in Argentina, India, Pakistan, and the Republic of Korea also successfully operate CANDU reactors. Romania will soon join their ranks. The proven record of excellent performance of CANDUs is due in part to the first objective of the vigorous R and D program: namely, to sustain and improve existing CANDU power-plant technology. The second objective is to develop improved nuclear power plants that will remain competitive compared with alternative energy supplies. The third objective is to continue to improve our understanding of the processes underlying reactor safety and develop improved technology to mitigate the consequences of upset conditions. These three objectives are addressed by individual R and D programs in the areas of CANDU fuel channels, reduced operating costs, reduced capital costs, reactor safety research, and IAEA safeguards. The work is carried out mainly at three centres of Atomic Energy of Canada Limited--the Chalk River Nuclear Laboratories, the Whiteshell Nuclear Research Establishment, and the Sheridan Park Engineering Laboratories--and at Ontario Hydro's Research Laboratories. Canadian universities, consultants, manufacturers, and suppliers also provide expertise in their areas of specialization

Economic case for CANDU life extension projects

International Nuclear Information System (INIS)

Qureshi, S.; Tenev, T.; Lewi, M.

2014-01-01

As CANDU reactors approach their original end of design life utilities are faced with two options: to extend the operating life of the reactor by undergoing a life extension project (LEP), or to commence decommissioning activities. Recent project experience has shown that there is economic merit in extending the life of the operating reactor. There are many benefits to such a decision, the most obvious being the revenue that will be generated from the additional years of electricity production by the utility. Delays in decommissioning are also advantageous since the large costs associated with such a long-term activity are pushed into the future, thereby decreasing the net present value (NPV) of the investment. In addition, relatively few power reactors have been fully decommissioned to date and deferring this activity transfers the associated risks to others that are currently obligated to undertake decommissioning activities sooner. Candu Energy has been involved with the life extension projects of the following CANDU reactors: Point Lepreau (New Brunswick, Canada), Bruce Unit 1 and Unit 2 (Ontario, Canada), and Wolsong Unit 1 (South Korea). These reactors underwent fuel channel replacement programs in addition to replacement of major reactor components. Most recently, both Ontario Power Generation (OPG) and Nucleoelectrica Argentina Sociedad Anonima (NA-SA) have commenced work on life extension projects at the Darlington (Canada) and Embalse (Argentina) sites respectively. The experience gained from previous LEP projects allows Candu Energy to deliver future projects in a timely, efficient, and cost effective manner. (author)